ML20244A634

Regulatory and Technical Reports (Abstract Index Journal). Annual Compilation for 1988

ML20244A634
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Issue date:	05/31/1989
From:	NRC OFFICE OF ADMINISTRATION (ADM)
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NUREG-0304, NUREG-0304-V13-N04, NUREG-304, NUREG-304-V13-N4, NUDOCS 8906120099
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4 Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication. Single copies of this publication are available from National Technical information Service, Springfield, VA 22161 l

NUREG-0304 Vol.13, No. 4 Regulatory and Technical Reports (Abstract Index Journal) Annual Compilation for 1988 Date Published: May 1989 Regulatory Publications Branch Division of Freedom of Information and Publications Services Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 p ~a hfY

1 1 1 l i l 1 1 CONTENTS i i Preface . . . . . .. .. .... . .. .... ........ .. . .. , ,,., ,, , ,,, , y l Index - Tab Main Citations and Abstracts . ... . .. .... . . . ... . ... . . . . ... .. ... 1

• Staff Reports
• Conference Proceedings ,

a Contractor Reports j

                                = International Agreement Reports Secondary Report Number Index . .                  ..... ...           ...... .. . . . ..                                   ... . ..        2     ,

Personal Author index . . . .. . . . . ... . . . . . . . . .3 i Subject index ... . ... ............... . ... .. . ... .. .. ... 4 NRC Originating Organization index (Staff Reports) . . .. ... .. .. ... . . .... .. ... 5 NRC Originating Organization Index (international Agreements) . . . . ..... .6 NRC Contract Sponsor Index (Contractor Reports) . . ... . ... . . .. .. . . 7 Contractor index . . . . . . . . . . . . .. . . .. .. . . ... . . . .. ... 8 International Organization index . . ... . .... . . . . ... ... ....... .... 9 Licensed Facility index . . . . . .. .... ... .. .. .... .10 1 i i 1 1 i; 5lI

PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors, it is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to: 1 Division of Publications Services Policy and Publications Management Branch Publishing and Translations Section Woodmont 537 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indbxes: Secondary Report Number index Personal Author Index Subject index NRC Originating Organization Index (Staff Reports) NRC Originating Organization Index (International Agreements) NRC Contract Sponsor index (Contractor Reports) Contractor index International Organization index Licensed Facility index A detailed explanation of the entries precedes each index. The bibliographic elements of the main citations are the following: Staff Report NUREG-0808: MARK !! CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048. 09570:200. Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for internal NRC use). Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND . RELIABIL.lTY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National I Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070. J Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled  ; the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu- l ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use). Contractor Report  ; NUREG/CR-1556: ETUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R. Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242. Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if j given), and (9) the microfiche address (for NRC internal use). v

                                                                                                              .___----____--a

1 intemational Agreement Report NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA. NEUMANN, U. Kraftwerk Union. August 1986. 223 pp. 8608270424. 37659:138. Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use). The following abbreviations are used to identify the document status of a report: ADD - addendum APP - appendix DRFT - draft ERR errata N - number R - revision S - supplement V - volume Availability of NRC Publications Copies of NRC stcff and contractor reports may be purchased either from the Government Printing Office l (GPO) or from the National Technical information Service, Springfield, Virginia 22161. To purchase i documents from the GPO, send a check or rponey order, payable to the Superintendent of Documents, to the following address: Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275-2171. Non-U.S. customers must make payment in advance either by International Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents. NRC Report Codes  ; The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported. In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/l A is used for intemational agreement reports. All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Publications Services. i vi

Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is an NRC staff-originated report, NUREG/CP-XXXX is an NRC-sponsored conference report, NUREG/CR-XXXX is an NRC contractor-prepared report, and NUREG/lA XXXX is an inter- , national agreement report. The bibliographic information (see Preface for details) is followed by a brief abstract of this report. NUREG-0020 V11 N12: LICENSED OPERATING REACTORS NUREG-0020 V12 N06: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of November STATUS

SUMMARY

REPORT. Data As Of May 31,1988.(Gray 30,1987.(Gray Book 1) SCHWARTZ,l. Division of Computer & Book 1) SCHWARTZ,l. Division of Computer & Telecommunica-

    -Telecommunications Services (Post 870413). January 1988.                tions Services (Post 870413). July 1988. 502pp. 8808080123.

479pp. 88020905D2. 44293:001. 46410:040. The OPERATING UNITS STATUS REPORT - LICENSED OP. See NUREG-0020,V11,N12 abstract. ERATING REACTORS provides data on the operation of nucle-ar units as timely and accurately as possible. This information as NUREG-0020 V12 N07: LICENSED OPERATING REACTORS STATL/S

SUMMARY

REPORT. Data As Of June 30,1988.(Gray collected by the Office of Administration and Resources Man- Book I) SCHWARTZ,l. Division of Computer & Telecommunica-agement from the Headquarters staff of NRC's Office of En- tions Services (Post 870413). August 1988. 533pp. forcement (OE), from NRC's Regional Offices, and from utilities. 8808230411.46573:238. The three sections of the report are: monthly highlights and sta- See NUREG 0020,V11,N12 abstract. tistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each NUREG-0020 V12 N08: LICENSED OPERATING REACTORS unit, provided by NRC's Regional Offices, OE Headquarters and STATUS

SUMMARY

REPORT. Data As Of July 31.,1988.(Gray the utilities; and an appendix for miscellaneous information such Book 1) 1CHWARTZ,t. Office of Administration & Resources as spent fuel storage capability, reactor-years of experience ano Management Director (Post 870413). September 1988. 515pp. non- power reactors in the U.S. It is hoped the report is helpful 88 to all agencies and individuals interested in maintaining an R 0 iW Me-awareness of the U.S. energy situation as a whole-NUREG-0020 V12 N09: LICENSED OPERATING REACTORS NUREG-0020 V12 N01: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of August 31,1988.(Gray Book 1) SCHWART2,1. Office of Administration & STATUS

SUMMARY

REPORT. Data As Of December Resources Management, Director (Post 870413). October 1988. 31,1987.(Gray Book l) SCHWARTZ,l. Division of Computer & 501pp. 8810280077. 47249 348. Telecommunications Services (Post B70413). February 1988. See NUREG-0020,V11,N12 abstract. 510pp. 8803090332. 44653:001. See NUREG 0020,V11,N12 abstract. NUREG-0020 V12 N10: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of September NUREG-0020 V12 NO2: LICENSED OPERATING REACTORS 30,1988.(Gray Book 1) SCHWARTZ,l. Office of Administration & STATUS

SUMMARY

REPORT. Data As Of January Resources Management Director (Post 870413). October 1988. 31,1988.(Gray Book 1) SCHWARTZ,l. Division of Computer & 562pp. 8812020180. 47685:283. Telecomrnunications Services (Post 870413). March 1988. See NUREG-0020,V11,N12 abstract. - 499pp. 8804080125. 45038:189. See NUREG-0020,V11,N12 abstract- NUREG-0020 V12 N11: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of October NUREG-0020 V12 NO3: LICENSED OPERATING REACTORS

                                                                                ,         ay Bod 0 SMWZ,L Mce of Mnistraen &

STATUS

SUMMARY

REPOPT. Data As Of February f988 13pp 881 19009 . 4 803 0 20,1988.(Gray Book 1) SCHWARTZ,l. Division of Computer & See NUREG-0020,V11,N12 &bstract. Telecommunications Services (Post 870413). April 1988. 505pp. 8805020159.45340:145. NUREG-0040 V11 N04: LICENSEE CONTRACTOR AND See NUREG-0020,V11,N12 abstract. VENDOR INSPECTION STATUS REPORT, Quarterly Report, October-December 1987.(White Book)

• Division of Re-NUREG-0020 V12 N04: LICENSED OPERATING REACTORS actor Inspection & Safeguards (Post 870411). February 1988.

STATUS

SUMMARY

REPORT. Data As Of March 31,1988.(Gray 185pp. 8802240187. 44468:084. Book 1) SCHWARTZ,1. Division of Computer & Telecommunica. This penodical covers the results of inspections performed by tions Services (Post 870413). May 1988. 513pp. 8805200063. the NRC's Vendor inspection Branch that have been distributed 45550:091. to the inspected organizations during the period from October See NUREG 0020,V11,N12 abstract. 1987 thru December 1987. Also, included in this issue are the results of certain inspections performed pnor to October 1987 NUREG-0020 V12 N05: LICENSED OPERATING REACTORS that were not included in previous issues of NUREG-0040. STATUS

SUMMARY

REPORT. Data As Of April 30,1968.(Gray Book () SCHWARTZ,l. Division of Computer & Telecommunica- NUREG-0040 V12 N01: LICENSEE CONTRACTOR AND tion: Services (Post 870413). June 1988. 510pp. 8806230155. VENDOR INSPECTION STATUS REPORT, Quarterly Report. January. March 1988.(White Book)

• Division of Reactor 45899:010.

See NUREG-0020,V11,N12 abstract- Inspection & Safeguards (Post 870411). May 1988.131pp. 8806230188. 45897:313. 1

2 Main Citations and Abstracts This penodical covers the results of inspections performed by report also contains information updating some previously re-the NRC's Vendor Inspection Branch that have been distributed ported abnormal occurrences. to the inspected organizations dunng the period from January 1988 through March 1988. Also included in this issue are the NUREG-0090 V11 N01: REPORT TO CONGRESS ON ABNOR-results of certain inspections performed prior to January 1988 MAL OCCURRENCES. January March 1988.

• Office for Analy-  !

that were not included in previous issues of NUREG 0040. . sis & Evaluation of Operational Data, Director. July 1988. 56pp. 8809 M 2 % M m NUREG 0040 V12 NO2: LICENSEE CONTRACTOR AND Section 208 of the Energy Reorganization Act of 1974 identi-VENDOR INSPECTION STATUS REPORT. Quarterly Report, April June 1988.(White Book)

• Division of Reactor in- I'es an abnormal occurrence as an unscheduled incident or spection & Fafeguards (Post 870411). August 1988. 67pp. event which the Nuclear Regulatory Commission determines to 8809150268.46803:001. be significant from the standpoint of public health and safety This periodical covers the resu!!s of inspections performed by and reauires a Quarterly report of such events to be made to the NRC's Vendor inspection Branch that have been distributed Congress. This report covers the period January 1 to March 31, to the inspected organizations dunng the period from April 1988 1988. Dunng the report period, there were three abnormal oc-through June 1988. Also includeo in this issue are the results of currences at nuclear power plants licensed to operate: a poten-certain inspections performed pnor to April 1988 that were not taal for common mode failure of safety-selated components due inclyded in previous issues of NUREG-0040. to a degraded instrument air system at Fort Calhoun; common NUREG-0040 V12 NO3: LICENSEE CONTRACTOR AND a a ma sa a nya s at Nny UnR t and a cracked pipe weld in a safety injection system at Farley VENDOR INSPECTION STATUS REPORT, Quarterly Unit 2. There were six abnormal occurrences at other NRC h-l Report. July-September 1908.(White Book)
• Division of Reactor censees; a diagnostic medical misadministration; a breakdown l Inspection & Safeguards (Post 870411). December 1988.

in management controls at the Georgia institute of Technology 128pp. 8901030130. 47962:058. l This penodical covers the results of inspections performed by reactor facihty; release of Polonium 210 from static elimination  ; I the NRC's Vendor inspection Branch that have been distnbuted devices manufactured by the 3M Company; two therapeutic to the inspected organization during the period from July 1988 medical misadministration; and a significant widespread break-through September 1988. Also included in this issue are the re. down in the radiation safety program at Case Western Reserve suits of certain inspections performed prior to April 1988, May University research laboratories. There was one abnormal oc-1988 and June 1988 that were not included in previous issues currence reported by an Agreement State (Texas) involving radi-of NUREG-0040. ation injury to two radiographer. NUREG-0090 V10 N03: REPORT TO CONGRESS ON ABNOR- NUREG-0090 V11 N02: REPORT TO CONGRESS ON ABNOR-MAL OCCURRENCES. July-September 1987.

• Office for Analy- MAL OCCURRENCES. April-June 1988.
• Office for Analysis &

sis & Evaluation of Operational Data, Director. March 1988. Evaluation of Operational Data, Director. December 1988.42pp. 45pp. 8804280308. 45270:244. 8901040378.47973.:353. Section 208 of the Energy Reorganization Act of 1974 identi-Section 208 of the Energy Reorganization Act of 1974 identi-fies an abnormal occurrence as an unscheduled incident or fies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to event which the Nuclear Regulatory Commission determines to be significant from the candpoint of public health and safety and requires a quarterly report of such events to be made to be significant from the standpoint of public health and safety Congress. This report covers the penod July 1 to September and requires a Quarterly report of such events to be made to 30,1987. Dunng the report period, there were two abnormal oc. Congress. This report covers the period April 1 to June 30, currences at the nuclear power plants licensed to operate. The 1988. For this report period, there were no abnormal occur-first involved a significant degradation of plant safety at Oyster rences at nuclear power plants licensed to operate. There were Creek; and the second involved a steam generator tube rupture two abnormal occurrences at other NRC licensecs: a significant at North Anna Unit 1. There were four abnormal occurrences at , breakdown in management and procedural controls at a medi-the other NRC hcensees. The first involved a therapeutic medi- cat faciRy and a diagnostic medical misadministration. There cal misadministration; the second involved a failure to repo-t di- was one abnormal occurrence reported by an Agreement State agnostic medical misadministration; the third involved the sus- (Texas) involving radioactive material released during a trans-pension of a well logging company's license; and the fourth in- portation accident. The report also contains information updat-volved the suspension of an industnal radiography company's li- ing some previously reported abnormal occurrences. cense. There were two abnormal occurrences reported by an Agreement State (New York). The first involved a hospital con- NUREG-0304 V12 N04: REGULATORY AND TECHNICAL RE-tamination incident and the second involved therapeutic medical PORTS (ABSTRACT INDEX JOURNAL). Annual Compilation misadministration. The report also contains information updat- For 1987.

• Division of Pubhcations Services (870413-880514).

ing some previously reported abnormal occurrences. March 1988.171pp. 8804080179. 45040:137. This journal includes all formal reports in the NUREG series NUREG-0090 V10 N04: REPORT TO CONGRESS ON ABNOR-MAL OCCURRENCES. October-December 1987.

• Office for prepared by the NRC staff and contractors; proceedings of con-Analysis & Evaluation of Operational Data, Director. March ferences and workshops; as well as international agreement re-1988. 34pp. 8804280329. 45270.213. ports. The entries in this compilation are indexed for access by Section 208 of the Energy Reorganization Act of 1974 identi- title and abstract, secondary report number, personal author, fies an abnormal occurrence as an unscheduled incident or subject, NRC organization for staff and international agree-event which the Nuclear Regulatory Commission determines to ments, contractor, intemational organization, and licensed facili-be significant from the standpoint of public health and safety ty.

and requires a quarterly report of such events to be made to Congress This report covers the period October 1 to December NUREG-0304 V13 N01: REGULATORY AND TECHNICAL RE-31,1987. Dunng the report period, there was one abnormal oc. PORTS (ABSTRACT INDEX JOURNAL). Compilation For First currence at the NRC heensees; the item involved the cuspen- Quarter 1988, January-March.

• Division of Freedom of Informa-sion of heense of an oil and gas well tracer company for non. tion & Publications Services (Post 880515). June 1988. 51pp.

compliance with NRC regulatory requirements. There were no 8807110503.46085:073, abnormal occurrences reported by the Agreement States The See NUREG-0304,V12,N04 abstract.

i l l Main Citations and Abstracts, 3 -j NUREG-0304 V13 N02: REGULATORY AND TECHNICAL RE. This Revision 9 of the fourth edition of the NRC Staff Practice PORTS (ABSTRACT INDEX JOURNAL). Compilation For. and Procedure Digest contains a digest of a number of Com .. Second Quarter 1988, April-June.

• Division of Freedom of infor- mission, Atomic Safety and Licensing Appeal Board, and Atomic mation & Publications Services (Post 880515). August 1988J Safety and Licensin9 Board decisions issues dunng the period 51pp. 8809060137. 46690:001, ';

from July 1,1972 to September 30, 1987, interpreting the See NUREG 0304,V12,N04 abstract. NRC's Rules of Practice in 10 CFR Part 2. This Revision 9 re-NUREG-0304 V13 NO3: REGULATORY AND TECHNICAL RE- placqs in pad earher editions and supplements and includes ap-PORTS (ABSTRACT INDEX JOURNAL). Compilation For Third propnate changes reflecting the amendrnents to the Rules of Quarter 1988 July-September,

• Division of Freedom of informa. Practice through September 30,1987. l tion & Publications Services (Post 880515). December 1988.

NUREG-0386 D04 R10: UNITED STATES ' NUCLEAR REGULA. f E 354 f 04 bstract. TORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. Commission. Appeal Board And. Licensing < Board NUREG-0313 R02: TECHNICAL REPORT ON MATERIAL SELEC. Decisions, July 1972. December 1987.

• Office of the General-TVN AND PROCESSING GUIDELINES FOR BWR COOLANT Counsel November 1988. 498pp. 8812190168. 47843:122. . .

  -FAESSURE-                            BOUNDARY         PIPING.Fina!       Report.                                                                               )

HAZELTON.W.S.i KOO,W.H. Division of Engineenng & Systems This Revision 10 of the fourth edition of the NRC Staff Prac- 4 tice and Procedure Digest contains a digest of a number of 4 Technology (Post 870411). January 1988. 47pp. 8802030155: Commission Atomic Safety and Licensing Appeal Board, and  ! 44241:127, l This report updates and supersedes the technical recommen- Atomic Safety and Licensing Board decisions issued during the - dations of NUREG-0313, " Technical Report on Material Selec- period from July 1,1972, to December 31, 1987, interpreting j

  - tion and Processing Guidelines for BWR Coolant Pressure                              the NRC's Rules of Practice in 10 CFR Part 2 This Revision 10.

Boundary Piping," published in July 1977. and its subsequent replaces, in part, earher editions and supplements and includes - revision published in July 1980. This report provides the techru- appropriate changes reflecting the amendments to the Rules of Practice effoctive through December 31,1987. -l cal bases for the NRC staff's revised recommended :aethods to control the intergranular stress corrosion cracking susceptibility of BWR piping. For piping that does not fully comply with the NUREG-0430 V08 N01: LICENSED FUEL FACILITY -STATUS material selection testing, and processing guideline combina. REPORT. inventory Difference Data. January-June 1987.(Gray tions of this documet, varying degrees of augmented inservice Book II)

• Office of Nuclear Material . Safety &

snspection are recommended. This revision also includes guid- Safeguards, Director. Maren 1988. 17pp. 8804060112. J 45029:197 i dn ac ev u tion a d weld riay repa r thod o NRC is committed to the periodic publication of licensed fuel long-term operation or for continuing interim operation of plants facilities' inventory difference data, following agency review of until a more permanent solution is implemented. the information and completion of any related NRC investiga-tions. Information in this report includes inventory difference 4 NUREG-0325 R11: U.S. NUCLEAR REGULATORY COMMISSION data for active fuel fabrication facilities possessing more than [ FUNCTIONAL ORGANIZATION CHARTS.

• Office of Adminis- one effective kilogram of high enriched uranium, low enriched tration & Resources Management, Director (Post 870413). Janu- uranium, plutonium, or uranium-233.

ary 1988. 56pp. 8802030220. 44240:218. Functional organization charts for the U.S. Nuclear Regulatory Commission offices, divisions, and branches are presented. NUREG-0430 V08 NO2: LICENSED FUEL FACILITY STATUS REPORTJnventory Difference Data. July. December 1987.(Gray NUREG-0386 D04 R07: UNITED STATES NUCLEAR REGULA. Book ~ II) Office : of Nuclear Material Safety & TORY COMMISSION STAFF PRACTICE AND PROCEDURE Safeguards, Director. September 1988. 16pp. 8809290026. DIGEST. July 1972 March 1987,

• Office of the General Coun. 46953:071.--

sel December 1987. 412pp. 8801260122. 44129:001, See NUREG-0430,V08,N01 abstract. This Revision 7 of the fourth edition of the NRC Staff Practice and Procedure Digest contains a digest of a number of Com. . NUREG-0525 R14: SAFEGUARDS

SUMMARY

EVENT LIST mission, Atomic Safety and Licensing Appeat Board, and Atomic - (SSEL). Division of Safeguards & Transportation (Post Safety and Licensing Board decisions issued dunng the period 870413). July 1988. 390pp. 8808120181. 46464:207, July 1,1972, to March 31,1987, interpreting the NRC Rules of The Safeguards Summary Event List provides brief summa- l Practice in 10 CFR Part 2. This Revision 7 replaces in part earii. ries of hundreds of safeguards related events involving nuclear 'j er editions and supplements and includes appropriate changes material of facilities regulated by the U.S. Nuclear Regulatory  ! reflecting the amendments to the Rules of Practice effective Commission. Events are described under the categories: bomb-l March 31,1987, related, intrusion, missing / allegedly stolen, transportation-relat-l NUREG-0386 D04 R08: UNITED STATES NUCLEAR REGULA- ed, tampering /vaadalism, arson, firearms-related, radiological ' TORY COMMISSION STAFF PRACTICE AND PROCEDURE sabotage, nonradiological sabotage, alcohol and drug related, OlGEST. July 1972 June 1987.

• Office of the General Coun- and miscellaneous. Information in the event descnptions was ,

sel June 1988. 496pp. 8808180247. 46522:001. obta,ned i from official NRC reports.  ; This Revision 8 of the fourth edition of the NRC Staff Practice and Procedure Digest contains a digest of a number of Com- NUREG-0540 V09 Nit: TITLE LIST OF DOCUMENTS MADE m!ssion, Atomic Safety and Licensing Appeal Board, and Atomic PUBLICLY AVAILABLE. November 1-30,1987,

• Division of Pub-Safety and Licensing Board decisions issued during the period heations Services (870413-880514). January 1988. 317pp. 1' from July 1,1972 to June 30, 1987, interpreting the NRC's 8802170363.44361:327 Rules of Practice in to CFR Part 2. This Revision 8 teplaces in This document is a monthly publication containing desenpJ )

part earher editions and supplements and includes appropnate tions of information received and generated by the U.S. Nuclear ndments to the Rules of Practice ef- Regulatory Commission (NRC). This information includes (1)

   ,      y  *[* * '"3u                   0 98                                          docketed material associated with civihan nuclear power plants .            j and other uses of radioactive matenals, and (2) nondocketed NUREG 0388 D04 R09: UNITED STATES NUCLEAR REGULA.                                      matenal received and generated by NRC portinent to its role as TORY COMMISSION STAFF PRACTICE AND PROCEDURE                                         a regulatory agency. The following indexes are included: Per.

DIGEST. July 1972 September 1987.

• Office of the General sonal AutMr. Corporate Source, Report Number, and Cross Counsel Jufy 1988. 465pp. 6808180243. 46520'212. Reference .o Pnncipaf Documents.

l a I I

                                                                                                                                                                                                               )

4 Main Citabuns and Abstracts NUREG-0540 V09 N12: TITLE LIST OF DOCUMENTS MADE of low-level waste burial facilities, high level waste repositones, PUBLICLY AVAILABLE. December 1 31,1987,* Division of Pub- and uranium mill and mill tailings piles, which are covered in lications Services (870413-880514). February 1988. 379pp. separate rulemaking activities, and decommissioning of uranium 8803090041. 44647:204. mines which are not under NRC jurisdiction. Recommendations See NUREG 0540,V09,N11 abstract. are made as to regulatory decommissioning particulars including such aspects as decommissioning alternatives, appropriate pre-NUREG-0540 V10 N01: TITLE LIST OF DOCUMENTS MADE Iminary planning requirements at the time of commissioning, PUBLICLY AVAILABLE. January 1 31,1988.

• Division of Publi- final planning requirements prior to termination of facility oper-cations Services (870413-880514). March 1988. 314pp. * ' ""*""#* "O * '*

8803250303.44862:164 **""' ** " See NUREG-0540,V09,N11 abstract, NUREG-0540 V10 NO2: TITLE LIST OF DOCUMENTS MADE NUREG-0654 S01 R01: CRITERIA FOR PREPARATION AND PUBLICLY AVAILABLE. February 149,1988.* Division of Publi. EVALUATION OF RADIOLOGICAL EMERGENCY RESPONSE cations Services (870413 880514). April 1988. 346pp. PLANS AND PREPAREDNESS IN SUPPORT OF NUCLEAR POWER PLANTS. Criteria For Utility Offsite Planning And Pre-8804280623.45273:038. See NUREG-0540,V09,N11 abstract, paredness. PODOLAK.E.M. NRC No Detailed Affiliation Given. SANDERS.M.E.; WINGERT,V.L; et al. Federal Emergency Man-NUREG-0540 V10 NO3: TITLE LIST OF DOCUMENTS MADE agement Agency. September 1988. 33pp. 8810030156. FEMA-PUBLICLY AVAILABLE. March 1 31,1988.

• Division of Publica- REP-1. 46972:313.

tions Services (870413 880514). May 1988. 387pp. This document has been developed for use in Nviewing and 8806020034. 45706:303, evaluating utility-prepared offsite emergency plans and pre-See NUREG-0540,V09,N11 abstract. paredness. The document is intended to be used with Section 1 and Appendices 1-5 of the existing NUREG-0654/ FEMA-REP-1, NUREG-0540 V10 N04: TITLE LIST OF DOCUMENTS MADE Rev.1. The document was previously issued in draft for interim PUBLICLY AVAILABLE. April 1 30,1988.

• Division of Freedom use and to invite public review and comment. Changes have of information & Publications Services (Post 880515). June been made to this final document where appropriate. A notice 1988. 353pp. 8807070426. 46012:018.

See NUREG-0540,V09,N11 abstract. has been provided in the Federal Register. Except where spe-cifically modified, the existing licensee-only evaluation criteria of NUREG-0540 V10 N05: TITLE LIST OF DOCUMENTS MADE the current Section il are not affected by this document. For PUBLICLY AVAILABLE. May 1 31,1928.

• Division of Freedom those situations in which State and/or local governments are of information & Publications Services (Post 880515). July 1988. participating in the emergency planning process, the existing 60pp. 8808180280. 46519:152. NUREG 0654/ FEMA-REP-1 Rev.1 evaluation criteria will apply.

See NUREG-0540,V09,N11 abstract. NUREG-0683 S03 DRFT: PROGRAMMATIC ENVIRONMENTAL NUREG-0540 V10 N06: TITLE LIST OF DOCUMENTS MADE IMPACT STATEMENT RELATED TO DECONTAMINATION PUBLICLY AVAILABLE. June 1-30,1988.

• Division of Freedom AND DISPOSAL OF RADIOACTIVE WASTES RESULTING of Information & Publications Services (Post 880515). Septem- FROM MARCH 28.1979 ACCIDENT,THREE MILE ISLAND NU-ber 1988. 451pp. 8810280164. 47248.257. CLEAR STATION, UNIT 2. Docket No. 50-320.(GPU See NUREG-0540,V09,N11 abstract. Nuclear, Incorporated)
• Office of Nuclear Reactor Regulation, NUREG-0540 V10 N07: TITLE LIST OF DOCUMENTS MADE Director (Post 870411). April 1988. 204pp. 8804280348.

PUBLICLY AVAILABLE. July 1-31,1988.

• Division of Freedom 45270:009.

of information & Publications Services (Post 880515). October in accordance with the National Environmental Policy Act, the 1988. 396pp. 8811010235. 47277:001. Programmatic Environmental impact Statement Related to De-See NUREG-0540,V09,N11 abstract. contamination and Disposal of Radioactive Waste from March l 28, 1979 Accident Three Mile Island Nuclear Station, Unit 2 NUREG-0540 V10 N08: TITLE LIST OF DOCUMENTS MADE (PElS) has been supplemented. This draft supplement address-PUBLICLY AVAILABLE. August 1-31,1988.

• Division of Free- es potential environmental impacts associated with the licens-dom of Information & Publications Services (Post 880515). No- ,,s (GPU Nuclear's) proposal to place the TMl-2 facility in a )

vember 1988. 435pp. 8812010368. 47691:073. post-defueling monitored storage mode followed by the comple- 1 See NUREG-0540,V09,N11 abstract. tion of cleanup. The NRC staff has concluded, based on this i NUREG-0540 V10 N09: TITLE LIST OF DOCUMENTS MADE evaluation, that the licensee's proposed plan and the NRC staff. PUBLICLY AVAILABLE. September 1-30,1988.

• Division of identified alternatives for completion of cleanup are within appli-Freedom of Information & Publications Services (Post 880515). cable regulatory limits and could be implemented without signifi-November 1988. 377pp. 8812190333. 47842:105 cant environmental impact. No alternative was found to be See NUREG-0540,V09,N11 abstract. clearly preferable from an environmental impact perspective.

The staff concluded that the benefits of cleanup action out-NUREG-0586: FINAL GENERIC ENVIRONMENTAL IMPACT weigh the small associated impacts. STATEMENT ON DECOMMISSIONING OF NUCLEAR FACILi-TIES.

• Office of Nuclear Regulatory Research. Director (Post NUREG-0713 V07: OCCUPATIONAL RADIATION EXPOSURE AT 860720). August 1988. 509pp. 8810280144. 47251:129. COMMERCIAL NUCLEAR POWER REACTORS AND OTHER This final generic environmental impact statement was pre- FACILITIES 1985. Eighteenth Annual Report. BROOKS,B. Divi-pared as part of the requirement for considenng changes in reg- sion of Regulatory Applications (Post 870413). April 1988.

ulations on decommissioning of commercial nuclear facilities. 168pp. 8804280390. 45271:136. Consideration is given to the decommissioning of pressunzed This report summs-izes the occupational radiation exposure water reactors, boiling water reactors, research and test reac- information that has been reported to the NRC's Radiation Ex. tors, fuel reprocessing plants (FRPs) (currently, use of FRPs in posure Information Reporting System (REIRS) by nuclear power the commercial section is not being considered), small mixed facilities and certain other categories of NRC licensees dunng oxide fuel fabncation plants, uranium hexafluonde conversion the years 1969 through 1985. The bulk of the data presented in plants, uranium fuel fabncation plants, independent spent fuel the report was obtained from annual radiation exposure reports storage installations and non-fuel-cycle facilities for handling submitted in accordance v'ith the requirements of 10 CFR byproduct, source and special nuclear matenals. Excluded here 20.407. Data on workers terminating their employment at certain from consideration for regulation change, are decommissioning NRC licensed facilities were obtained from reports submitted

Main Citations and Abstracts 5 l pursuant to 10 CFR 20.408. The 1985 annual reports submitted See NUREG-0750.V25,IO2 abstract. l by about 500 hcensees indicated that approximately 216,300 in-I dividuals were monitored 94,000 of whom were monitored by NUREG-0750 V26 N01: NUCLEAR REGULATORY COMMISSION nuclear power facilities. They incurred an average individual ISSUANCES FOR JULY 1987.Pages 1-70.

• Division of Publica-dose of 0.22 rem (cSv) and an average measureable dose of tions Services (870413-880514). January 1988. 78pp.

0.43 rem (cSv). Termination radiation exposure reports were 8802100037. 44297:163. analyzed to reveal thct about 77,300 individuals completed their See NUREG 0750,V25,N05 abstract. employment with one or more of the 500 covered licensees dunng 1984. Some 73,200 of these individuals terminated from NUREG-0750 V26 N02: NUCLEAR REGULATORY COMMISSION power reactor facilities, and about 7,400 of them were consid- ISSUANCES FOR AUGUST 1987, Pages 71107.

• Division of ered to be transient workers who received an average dose of Publications Services (870413-880514). January 1988. 44pp.

1.05 rem (cSv). 8802100027. 44297:120. NUREG-0725 R06: PUBLIC INFORMATION CIRCULAR FOR * " " SHIPMENTS OF IRRADIATED REACTOR FUEL

• Division of Safeguards & Transportation (Post 870413). April 1988. 95pp. NUREG-0750 V26 NO3: NUCLEAR REGULATORY COMMISSION ,

8805200083. 45560:346. ISSUANCES FOR SEPTEMBER 1987, Pages 109-248.

• Divi- '

This circular has been prepared in response to numerous re. sion of Publications Services (870413-880514). February 1988. 153pp. 8803090267. 44655:119. Quests for information regarding routes used for the shipment of See NUREG-0750,V25,N05 abstract. irradiated reactor (spent) fuel subject to regulation by the Nucle-ar Regulatory Commission (NRC), and to meet the requirements of Public Law 96-295. The NRC staff must approve such routes NUREG-0750 V26 N04: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR OCTOBER 1987. Pages 249-381.

• Division pnor to their use in accordance with the regulatory provisions of Section 73.37 of 10 CFR Part 73. The information included re-of Publications Services (870413-880514). February 1988.

143pp. 8803150354. 44704:091. flects NRC staff knowledge as of September 30,1987, Spent See NUREG-0750,V25,N05 abstract. fuel shipment routes, primarily for road transportation, but also including three rail routes, are indicated on reproductions of De-partment of Transportation road maps. Also included are the NUREG-0750 V26 N05: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR NOVEMBER 1987. Pages 383-447.

• Division j amounts of material shipped during the appropriate eight-year of Publications Services (870413-880514). March 1988. 74pp.

period that safeguards regulations for spent fuel shipments 8803310244.44920:107. i have been efective. In addition, the Commission has chosen to See NUREG-0750,V25,N05 abstract. provide information in this document regarding the NRC's safety and safeguards regulations for spent fuel shipments as well as l NUREG-0750 V26 N06: NUCLEAR REGULATORY COMMISSION i safeguards incidents regarding spent fuel shipments (of which ISSUANCES FOR DECEMBER 1987. Pages 449 530.

• Division none have been reported to date). This additional information is of Publications Services (870413-880514). March 1988. 89pp.

furnished by the Commission in order to convey to the public a 8804110197. 45071:156. more complete picture of NRC regulatory practices concerning See NUREG-0750,V25,N05 abstract. the shipment of spent fuel than could be obtained by the publi-cation of !he shipment routes and quantities alone. NUREG-0750 V27101: INDEXES TO NUCLEAR REGULATORY { NUREG-0750 V25102: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. January-March 1988.

• Division of

{ COMMISSION ISSUANCES. January-June 1987,

• Division of e m of InWmaton & Mcahom hes Wst 6805%

June 1988. 46pp. 8807080350. 46077:107. Publications Services (870413-880514). March 1988. 86pp. See NUREG-0750,V25.102 abstract.

                                                                                                                                              ]

8803310246. 44920:181. i Digest and indexes for issuances of the Commission, the Atomic Safety and Licensing Appeal NUREG-0750 V27102: INDEXES TO NUCLEAR REGULATORY and Licensing Board Panel, the inistrativeAdm, Panel, the Atomic Law Judge, the Safety COMMISSION ISSUANCES. January June 1988.

• Division of I Directors' Decisions, and the Denials of Petitions for Rulemak. Freedom of Information & Publications Services (Post 880515).

ing are presented- September 1988. 72pp. 8810280096. 47247:300. See NUREG-0750,V25,!02 abstract. I NUREG-0750 V25 N05: NUCLEAR REGULATORY COMMISSION l ISSUANCES FOR MAY 1987.Pages 417-873.

• Division of Pub. NUREG-0750 V27 N01: NUCLEAR REGULATORY COMMISSION lications Services (870413-880514). December 1987, 463pp. ISSUANCES FOR JANUARY 1988. Pages 1-39.
• Division of 8801250244. 44114:153. Publications Services (870413-880514). March 1988. 46pp.

Lega! issuances of the Commission, the Atomic Safety and Li- 8804110195. 45071:246. censing Appeal Panel, the Atomic Safety and L6 censing Board See NUREG-0750,V25,N05 abstract. Panel the Administrative Law Judge, and NRC program offices NU'tEG-0750 V27 NO2: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR FEBRUARY 1988. Pages 41255.

• Division NUREG-0750 V25 N06: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JUNE 1987.Pages 875 997.
• Division of of Publications Services (870413-880514). April 1988. 223pp.

B805090108. 45413:073. Publications Services (870413 880514). January 1988.134pp. See NUREG-0750,V25,N05 abstract. 8802100136.44314:193. See NUREG-0750,V25,N05 abstract. NUREG-0750 V27 NO3: NUCLEAR REGULK10RY COMMISSION NUREG 0750 V26101: INDEXES TO NUCLEAR REGULATORY ISSUANCES FOR , MARCH 1988. Pages 257-334.

• Division of COMMISSION ISSUANCES. July-September 1987.
• Division of Publications Services (870413-880514). May 1988. 84pp.

                  "         ' (870413 880514). March 1988. 50pp.             88 N 880 200204.453 282                                                                e N REG 075 V25,N05 abstract.

See NUREG-0750,V25.102 abstract. NUREG-0750 V27 N04: NUCLEAR REGULATORY COMMISSION NUREG-0750 V26102: INDEXES TO NUCLEAR REGULATORY ISSUANCES FOR APRIL 1988. Pages 335 483.

• Division of COMMISSION ISSUANCES. July-December 1987.
• Division of Publications Services (870413 880514). April 1988. 65pp. Freedom of information & Publications Services (Post 880515).

June 1988.159pp. 8808050268. 46394:224. 8805200092. 45550:026. See NUREG 0750,V25,N05 abstract. I l l

6 Main Citations and Abstracts j j censing Board (ASLB), the intervenor Cituens Association for  ! NUREG 0750 V27 N05: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR MAY 1988. Pages 485 626.

• Division of Sound Energy (CASE), the Comanche Peak Respons>e Team Freedom of information & Pubhcations Services (Post 880515). (CPRT), Cygna Energy Services (Cygna), and the NHC Staff.

July 1988.151pp. 8808080055. 46407:312. The NRC staff concludes that the CAP establishes a compre. See NUREG-0750,V25,N05 abstract. hensive program for resolving the technical concerns identified by the ASLB, CASE, CPRT, Cygna, and the NRC staff and its NUREG-0750 V27 N06: NUCLEAR REGULATORY COMMISSION tSSUANCES FOR JUNE 1988. Pages 627-665.

• Division of implementation ensures that the design of large and small bore Freedom of information & Pubhcations Services (Post 880515). piping and pipe supports at CPSES satisfies the applicable re- i August 1988. 84pp. 8808300264. 46632:303. quirements of 10 CFR Part 50.

J See NUREG-0750,V25,N05 abstract. NUREG-0797 S15: SAFETY EVALUATION REPORT RELATED NUREG-0750 V28 N01: NUCLEAR REGULATORY COMMISSION TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-ISSUANCES FOR JULY 1988.Pages 1-71,

• Division of Free- TRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-dom of Information & Publications Services (Post 880515). Sep- 446(Texas Utilities Generating Company)
• Ofc of Special tember 1988. 78pp. 8810050202. 46990:162. Projects (Pre 881231). July 1988. 183pp. 6808240043.

See NUREG-0750,V25,N05 abstract. 46578.001. NUREG-0750 V28 N02: NUCLEAR REGULATORY COMMISSION Supplement 15 to the Safety Evaluation Report related to the ISSUANCES FOR AUGUST 1988. Pages 73 269.

• Division of operation of the Comanche Peak Steam Electric Station Freedom of information & Publications Services (Post 880515). (CPSES), Units 1 end 2 (NUREG-0797), has been prepared by October 1988. 206pp. 8811220517. 47601:133- the Office of Special Projects of the U.S. Nuclear Regulatory See NUREG-0750,V25,N05 abstract. Commission (NRC). The facility is located in Somervell County, NUREG-0750 V28 NO3: NUCLEAR REGULATORY COMMISSION Texas, approximately 40 miles southwest of Fort Worth, Texas.

ISSUANCES FOR SEPTEMBER 1988. Pages 271-417.

• Divi- This supplement presents the staff's evaluation of the appli-sion of Freedom of Information & Publications Services (Post cants' Corrective Action Program (CAP) related to the design of 880515). November 1988.156pp. 8812020157,47687:337. cable trays and cable tray tungers, The scope and methodolo.

See NUREG-0750,V25.N05 abstract. gies for the CAP workscope as summadzed in Revision 0 to the j cable tray and cable tray hanger projev status report and as NUREG-0750 V28 N04: NUCLEAR REGULATORY COMMISSION detailed in related documents referenced in this evaluation were ISSUANCES FOR OCTOBER 1988. Pages 419-497.

• Division developed to resolve various design issues raisuo by the Atomic of Freedom of information & Publications Services (Post Safety and Licensing Board (ASLB), the intervenor, Citizens M.

880515). November 1988. 90pp. 8812280060. 47911:155. See NUREG-0750,V25,N05 abstract. sociation for Sound Energy (CASE), the Comanche Peak Re-sponse Team (CPRT), CYGNA Energy Services (CYGNA), and NUREG-0781 S05: SAFETY EVALUATION REPORT RELATED the NRC staff. The NRC staff concludes that the CAP works-TO THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 cope for cable trays and cable tray hangers provides a compre-AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And hensive program for resolving the associated technical cork Power Company) Division of Reactor Projects - lit.IV,V & Spe- cerns identified by the ASLB, CASE, CPRT, CYGNA, and the cia P ects (Post 870411). March 1988. 53pp. 8804080171, NRC staff and its implementation ensures that the design of cable trays and cable tray hangers at CPSES satisfies the appli-in April 1986 the staff of the U.S. Nuclear Regulatory Com- cable requirements of 10 CFR 50. mission issued its Safety Evaluation Report (NUREG 0781) re-garding the application of Houston Lighting and Power Compa- NUREG-0797 S16: SAFETY EVALUATION REPORT RELATED ny (applicant and agent for the owners) for a license to operate TO THE GPERATION OF COMANCHE PEAK STEAM ELEC-South Texas Project, Units 1 and 2 (Docket Nos. 50-498 and TRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-50-499). The facility is located in Matagorda County, Texas, west of the Colorado River, 8 miles north- northwest of the 446.(Texas Utilities Generating Company)

• Ofc of Special town of Matagorda and about 89 miles southwest of Houston.

Projects (Pre 881231). July 1988. 164pp. 8808240054. The first supplement to NUREG-0781 was issued in September 46578:184. 1986, the second supplement in January 1987, the third supple. Supplement 16 to the Safety Evaluation Report related to the ment in May 1987, and the fourth supplement in July 1987. This operation of the Comanche Peak Steam Electric Station fifth supplement provides updated information on the issues that (CPSES), Units 1 and 2 (NUREG-0797), has been prepared by had been considered previously as well as the evaluation of the Office of Special Projects of the U.S. Nuclear Regulatory issues that have enson since the fourth supplement was issued. Commission (NRC). The facility is located in Somervell County, The evaluation resolves all the issues necessary to support the Texas, approximately 40 miles southwest of Fort Worth, Texas. issuance of a full- power license for Unit 1. This supplement presents the staff's evaluation of the appli-cants' Corrective Action Program (CAP) related to the design of NUREG-0797 S14: SAFETY EVALUATION REPORT RELATED conduit supports. The scope and methodologies for the CAP TO THE OPERATION OF THE COMANCHE PEAK STEAM workscopes as summarized in Revision 0 to the conduit support ELECTRIC STATION. UNITS 1 AND 2. Docket Nos. 50-445 And 50-446.(lexas Utilities Generat:ng Company)

• Ofc of Special project status reports and as detailed in related documents ref-erenced in this evaluation were developed to resolve various Projects (Pre 881231). March 1988.1,308pp. 8803280039.

44885.199. design issues raised by the Comanche Peak Response Team Supplement 14 to the Safety Evaluation Report related to the (CPRT). CYGNA Energy Services (CYGNA), and the NRC staff. operation of the Comanche Peak Steam Electnc Station The NRC staff concludes that the CAP workscopes for conduit (CPSES), Units 1 & 2 (NUREG-0797), has been prepared by the supports provides a comprehensive program for resolving the Office of Special Projects of the U.S. Nuclear Regulatory Com- associated technical concerns identified by the CPRT, CYGNA, mission (NRC). The facility is located in Somerville County, and the NRC staff and their implementation ensures that the Texas, approximately 40 miles southwest of Fort Worth, Texas. design of conduit supports at CPSES satisfies the applicable re-This supplement presents the staff's evaluation of the appli- quirements of 10 CFR 50. cants' Corrective Action Program (CAP) related to large and small bore piping and pipe supports which was developed to re- NUREG-0797 S17: SAFETY EVALUATION REPORT RELATED solve vanous design issues raised by the Atomic Safety and Li-

i Main Citations and Abstracts 7 j TO THE OPERATION OF COMANCHE PEAK STEAM ELEC- Status Report and as detailed in related documents, were de-TRIC STATION, UNITS 1 AND 2. Docket Nos 50-445 And 50- veloped to resolve various issues raised by the Comanche Peak  ! 446.(Texas Utilities Generating Company) NORKIN.D.P. Ofc of Response Team (CPRT) and the NRv staff to ensure that plant Special Projects (Pre 881231). November 1988. 396pp. equipment is appropriately environmentally and/or seismically 8812160177.47836:345. and dynamically qualified, and documented in accordance with Supplement 17 to the Safety Evaluation Report related to the the validated plant design resulting from other CAP scopes of operation of Comanche Peak Steam Electric Station (CPSES). work for Unit 1 and areas common to Units 1 and 2. The staff Units 1 and 2 (NUREG-0797), has been prepared by the Office concludes that the CAP workscope for equipment qualification of Special Projects of the U.S. Nuclear Regulatory Commission provides a comprehensive program for resolving the concerns (NRC). The facilt/is located in Somervell County, Texas, ap- identified by the CPRT and the NRC staff, and its implementa-proximately 40 miles southwest of Fort Worth, Texas. This sup- tion will assure that equipment at CPSES is appropriately quali- j plement presents the staff's evaluation of the applicant's Cor- fied. The staff will venfy the adequacy of implementation of the rective Action Program (CAP) related to the mechanical, civil / equipment qualification program at CPSES during inspection (s) i structural, electrical, . instrumentation and controis (l&C), and prior to fuel loading. heating, ventilation and air conditioning (HVAC) disciplines. This l l supplement only addresses the systems portion of HVAC. Other NUREG-0797 S20: SAFETY EVALUATION REPORT RELATED aspects of HVAC (e.g., duct and supports) are covered in an. TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-other supplement. The scope and methodologies for these CAP TRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-workscopes, as summarized in project status reports and as de- 446.(Texas Utilities Generating . Company) MCKEE.P.; tailed in related documents referenced in this evaluation, were BROWN.C. Ofc of Special Projects (Pre 881231). November developed to resolve various design issues raised by the Co- 1988. 223pp. 8812190088. 47840:196, manche Peak Response Team (CPRT), CYGNA Energy Serv- Supplement 20 to the Safety Evaluation Report related to the ices (CYGNA), ered tho NRC staff. The NRC staff concludes operation of the Con.anche Peak Steam Electric Station that the CAP workscopes for mechanical, civil / structural, electri- (CPSES), Units 1 and 2 (NUREG-0797), has been prepared by cal, l&C, and HVAC systems provide a comprehensive program the Office of Special Projects of the U.S. Nucloar Regulatory for resoMng the associated technical concerns identified by the Commission (NRC). The facility is located in Somervell County, CPRT, CYGNA, and the NRC staff, and their implementation en- Texas, approximately 40 miles southwest of Fort Worth. Texas. sures that the design of related systems, structures, and com- This supplement presents the staff's evaluation of Comanche ponents at CPSES satisfies the applicable requirements of 10 Peak Response Team (CPRT) implementation of the CPRT Pro-CFR Part 50. gram Plan and the issue-specific action plans (ISAPs), as well NUREG-0797 S18: SAFETY EVALUATION REPORT RELATED as the CPRT's investigations to determine the adequacy of vari-TO THE OPERATION OF COMANCHE PEAK STEAM ELEC. ous types of programs and hardware at CPSES. The results and TRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50 conclusions of the CPRT activities are documented in a results 4464 Texas Utilities Generating Company) TERAO.D. Ofc of report for each ISAP, a Collective Evaluation Report (CER), and Special Projects (Pre 881231). November 1988. 100pp. a Collective Significance Report (CSR). This supplement also 8812160178. 47838:021, presents the staff's safety evaluation of TU Electric's root cause Supplement 18 to the Safety Evaluation Report related to the assessment of past CPSES design deficiencies and weaknese-operation of the Comanche Peak Steam Electric Station es. The NRC staff concludes that the CPRT has adequately im-(CPSES). Units 1 and 2 (NUREG-0797). has been prepared by piemented its investigative and overview activities related to the the Office of Special Projects of the U.S. Nuclear Regulatory design, construction, construction OA/OC, and testing at Commission (NRC). The facility is located in Somervell County, CPSES. The NRC staff further concludes that the CPRT evalua-Texas, approximately 40 miles southwest of Fort Worth, Texas. tion of the results of its investigations is thorough and complete, This supplement presents the staff's evaluation of the appli- and the recommendations for corrective actions are sufficient to cants' Corrective Action Program (CAP) related to the structural resolve identified deficiencies, design cf the heating, ventilation, and air conditioning (HVAC) systems. The scope and methodologies for the CAP workscope NUREG-0800 02,4.2 R3: STANDARD REVIEW PLAN FOR THE as summarized in Revision 0 to the HVAC project status report REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR -I and as detailed in related documents referenced in this evalua. POWER PLANTS. LWR Edition. Proposed Revision 3 To SRP tion were developed to resolve the technical concerns identified Section 2.4.2, " Floods." For Comment. KORNASIEWICZ,R. Di-in the HVAC area. The NRC staff concludes that the CAP vision of Engineenng (Post 870413). October 1988. 9pp. , workscope for HVAC structural design provides a comprehen- 8811110041. 47519:167. l sive program for resolving the associated technical concems The section of the safety analysis report addressed by this j and its implementation ensures that the structural design of the section of the standard review plan (SRP) identifies historical ' HVAC systems at CPSES satisfies the applicable requirements flooding at the proposed site or in the region of the site. It sum- l of 10 CFR 50. marizes arid identifies the ,ndividual I types of flood. producing ] NUREG-0797 Sig: SAFETY EVALUATION REPORT RELATED phenomena, and comb 6 nations of flood-producing phenomena, considered in establishing the flood design bases for safety-re. TO THE OPERATION OF COMANCHE PEAK STEAM ELEC- lated plant features. It also covers the potential effects of local TRIC STATION,UNilS 1 AND 2. Docket Nos. 50-445 And 50-intense precipitation. The flood history and potential for flooding 446.(Texas Utilities Generatng Company) HUBBARD G.; are reviewed. Factors affecting potential runoff (such as urban-MALLOY,M. Ofc of Special Prclects (Pro 881231). November ization, forest fire, or change in agricultural use), erosion, and 1988. 64pp. 8812160181. 47838:121. sediment deposition are considered in the review. Supplement 19 to the Safety Evaluation Report related to the operetion of the Comanche Peak Steam Electric Station NUREG-0800 02,4.3 R3: STANDARD REVIEW PLAN FOR THE (CPSES), Units 1 and 2 (NUREG-0797), has been pmpared by REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR the Office of Special Projects of the U.S. Nuclear Regulatory POWER PLANTS. LWR Edition. Proposed Revision 3 To SRP Commission. The facility is located in Somervell County, Texas. Section 2.4.3, "Probab:e Maximum Flood (PMF) On Streams & approximately 40 miles southwest of Fort Worth, Texas. This Rivers." For Comment. KORNASIEWICZ,R. Davision of Engi-supplement presents the staff's evaluation of the applicant's neenng (Post 870413). October 1988. 18PP. 8811110081. Corrective Action Program (CAP) related to equipment qualifica. 47520.139. tion. The scope and methodology for the CAP workscope, as The section of the safety analysis report addressed by this summarized in Revision 0 to the Equipment Qualification Project section of the standard review plan addresses the hydrometeor-l  ! l

                                                                                                                                                           . _ _ _ ~

l l

 .-                                                                                                                                                                                                              i u

( 8 Main Citations and Abstracts l j ological design basis developed to determine the extent of any . NUREG 0844: NRC INTEGRATED PROGRAM FOR THE RESOJ Dood protection required for those structures, systems, and LUTION OF UNRESOLVED SAFETY ISSUES A 3,A-4 AND A-5 > l components necessary to ensure the capability to shut down REGARDING STEAM GENERATOR TUBE INTEGRITY. Final the reactor and maintain it in a safe shutdown condition. The Report. MURPHY,E. Division of Engineering & Systems Tech-areas of review include the probable maximum precipitation nology (Post 870411). September 1988.152pp. 8810280082. l 47248:105. l (PMP) potential and precipitation losses over the applicable This report presents the results of the NRC integrated pro- '1 drainage area, the runoff response characteristics of the water-shed, the accumulation of flood runoff through river channels gram for the resolution of Unresolved Safety issues (USIs) A 3, j and reservoirs, the estimate of the discharge rate trace (hydro- A-4, and A-5 regarding steam generator tube integrity. A generic graph) of the PMF at the plant site, the deterrnination of PMF risk assessment is provided and indicates that risk from steam water level conditions at the site, and the evaluation of coinci- generator tube rupture (SGTR) events is not a significant con-dent wind- generated wave conditions that could occur with the tributor to total risk at a given site, nor to the total nsk to which 'a PMF. The analyses involve modeling of physical rainfall and. the general public is routinely exposed. This report also identi- l runoff processes to estimate the upper level of possible flood . fies a number of staff recommended actions that the staff finds conditions adjacent to and on the site. Can further enhance the effectiveness of licensee programs in j ensuring steam generator tube integrity and in mitigating the NUREG-0822 S01:lNTEGRATED PLANT SAFETY ASSESSMENT couque,nces of an SGTR. As part of the integrated program, SYSTEMATIC EVALUATION PROGRAM-OYSTER CREEK NU, the staff issued Generic Letter 85-02 encouraging licensees of; CLEAR GENERATING STATION. Docket No. 50-219.(General pressurized water reactors (PWRs) to upgrade their programs, Public Utilities Corporation And Jersey Central Power And Light . as necessary, to meet the intent of the staff- recommended ac. Company)

• Division of Reactor Projects l'Il (Post 870411). tions; however, such actions do not constitute NRC require-Jufy 1988. 70pp. 8808230426. 46594:134 ments. In addition, this report describes a number of ongo.ing The U.S. Nuclear Regulatory Commission (NRC) has pre, staff actions and studies involving steam generator issues which

      . pared Supplement 1 to the final integrated Plant Safety Assess.                                                               are being pursued to provide added assurance that risk from ment Report (IPSAR) (NUREG-0822), under the scope of the SGTR events will continue to be small. The staff concludes that Systematic Evaluation Program (SEP), for the Oyster Creek No.                                                                 with final publication of this report, USls A 3, A 4, and A-5 are clear Generating Station, located in Ocean County, New Jersey technically resolved.

and operated by GPU Nuclear Corporation and Jersey Central Power and Light Company (colicensees). The SEP was initiated NUREG-0933 S07: A PRIORITIZATION OF GENERIC SAFETY by the NRC to review the design of older operating nuclear ISSUES. EMRIT,R.; RIGGS,R.; MILSTEAD,W,; et al. Division of power plants to reconfirm and document their safety. This Regulatory Applications (Post 870413). April 1988. 291pp. report documents the review completed under SEP for those issues that required refined engineering evaluations or the con- 8805060326. 45408:020. tinuation of ongoing evaluations subsequent to issuing the Final The report presents the priority rankings for generic safety issues related to nuclear power plants. The purpose of these IPSAR for the Oyster Creek plant. The review has provided for (1) an assessment of the significance of differences between rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a current technical positions on selected safety issues and those that existed when the Oyster Creek plant was licensed, (2) a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP and have been assigned basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolutions of the plant safety. The final IPSAR and its supplement will form part of the bases for considenng the conversion of tr.e existing pro. safety issues were implemented, and the consideration of un-certainties and other quantitative or qualitative factors. To the visional operating license to a full. term operating license. NUREG-0837 V07 N04: NRC TLD DIRECT RADIATION MONI-TORING NEMORK. Progress Report, October-December 1987. NUREG-0933 S08: A PRIORITIZATION OF GENERIC SAFETY I STRUCKMEYER,R.; MCNAMARA,N.; COHEN.L. Region 1 Ofc ISSUES. EMRIT,R.t RIGGS,R.; MILSTEAD,W.; et al. Division of of the Director. April 1988. 323pp. 8804290185. 45283:099. Regulatory Applications (Post 870413). November 1988.212pp. This report provides the status and results of the NRC Ther- 8812020170. 47687:125. moluminescent Dosimeter (TLD) Direct Radiation Monitoring See NUREG-0933,S07 abstract. Network. It presents the radiation levels measured in the vicimty of NRC licensed facility sites throughout the country for the NUREG-0936 V06 N04: NRC REGULATORY AGENDA.Ouarterly fourth quarter of 1987. Report,0ctober-December 1987. Divisico of Rules & Records (870413-880514). February 1988. 115pp. 8803090295. NUREG-0837 V08 N01: NRC TLD DIRECT RADIATION MONI- 44652:2 0. TORING NETWORK. Progress Report, January-March 1988. The NRC Regulatory Agenda i ,s a compilation of all rules on STRUCKMEYER,R.; MCNAMARA,N. Region 1, Ofc of the Direc. which the NRC has proposed or is considenng action and all tor. June 1988. 227pp. 8807110535. 46078:214. petitions for rulemaking which have been received by the Com-This report provides the status and results of the NRC Ther. mission and are pending disposition by the Commission. The moluminescent Dosimeter (TLD) Direct Radiation Monitoring Regulatory Agenda is updated and issued each quarter. Network. It presents the radiation levels measured in the vicinity of NRC heensed facility sites throughout the country for the first NUREG-0936 V07 N01: NRC REGULATORY AGENDA.Ouarterly quam of M88' Report, January-March 1988.

• Division of Freedom of informa-tion & Publications Services (Post 880515). July 1988.112pp.

NUREG-0837 V08 NO2: NRC TLD DIRECT RADIATION MONI. TORING NETWORK. Progress Report. April-June 1988. 8808080119.46409:288. STRUCKMEYER,R.; MCNAMARA.N. Region 1. Ofc of the Direc- See NUREG-0936,V06,N04 abstract. tor. September 198S. 234pp. 8810050185. 46988:037. NUREG-0936 V07 NO2: NRC REGULATORY AGENDA.Ouarterly This report provides the status and results of the NRC Ther. moluminescent Dosimeter (TLD) Direct Radiation Monitoring Report, April-June 1988.

• Division of Freedom of Information &

Network. It presents the radiation levels measured in the vicinity Pubhcations Services (Post 880515). August 1988. 125pp. of NRC bcensed facility sites throughout the country for the 8809280278.46953:289. second quarter of 1988 See NUREG 0936,V06,N04 abstract. i l l

Main Citations and Abstracts 9 NUREG-0936 V07 NO3: NRC REGULATORY AGENDA.Ouarterly tractor has written a more complete and detailed annuai report Report, July September 1988.

• Division of Freedom of informa- of their work which can be obtained by wnting to NRC; howev.

tion & Publications Services (Post 880515). October 1988. er, we beheve it is useful to have a summary of each contrac-121pp. 8811010244. 47326:232. tor's efforts for the year combined into one volume. See NUREG-0936,V06,N04 abstract. NUREG-1002 S06: SAFETY EVALUATION REPORT RELATED NUREG-0940 V06 N04: ENFORCEMENT ACTIONS:SIGNIFICANT ACTIONS RESOLVED.Ouarterly Progress Report, October-De- TO THE OPERATION OF BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos. 50-456 And 50-457.(Commonwealth Edison cember 1987,

• Ofc of Enforcement (Post 870413). February Company)
• Division of Reactor Projects - Ill,lV,V & Special 1988. 224pp. 8804110204. 45070:180.

Projects (Post 870411). June 1988. 20pp. 8807060087. This compilation summarizes significant enforcement actions 46002:338. that have been resolved during one quanerly period (October - In November 1983, the staff of the Nuclear Regulatory Com-December 1987) and includes copies of letters Notices, and mission issued its Safety Evaluation Report (NUREG-1002) re-Orders sent by the Nuclear Regulatory Commission to licensees garding the application filed by the Commonwealth Edison Com-with respect to these enforcement actions. It is anticipated that pany, as applicant and rwner, for a license to operate Braid-the information in this publication will be widely disseminated to wood Station, Units 1 and 2 (Docket Nos,50-456 and 50-457). managers and emdoyees engaged in activities licensed by the The first supplement to NUREG-1002 was issued in September NRC, so that actions can be taken to improve safety by avoid- 1986; the second supplement was issued in October 1986; the ing future violations similar to those described in this publica- third supplement was issued in May 1987; the fourth supple-tion. ment was issued in July 1987 in support of the full-power li. NUREG-0940 V07 N01: ENFORCEMENT ACTIONS:SIGNIFICANT cense for Unit 1; the fifth supplement was issued in December ACTIONS RESOLVED.Ouarterly Progrecs Report, January-March 1987 in support of the low-power license for Unit 2. This sixth 1988.

• Ofc of Enforcement (Post 870413). June 1988. 315pp. supplement to NUREG-1002 is in support of the full-power it-8807070438.46014:046. cense for Unit 2 and provides the status of items that remained This compilation summarizes significant enforcement actions ynresolved at the time Supplement 5 was published. The facility that have been resolved during one quarterly period (January - is located in Reed Township, Will County, Illinois.

March 1988) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulstory Commission to licensees with NUREG-1032: EVALUATION OF STATION BLACKOUT ACCl-respect to these enforcement actions. It is anticipated that the DENTS AT NUCLEAR POWER PLANTS. Technical Findings Re. information in this publication will be widely disseminated to lated To Unresolved Safety issue A-44. Final Report. managers and employees engaged in activities licensed by the BARANOWSKY,P.W. Office of Nuclear Regulatory Research, Director (Post 860720). BARANOWSKY,P.W. Office of Nuclear NRC, so that actions can be taken to improve safety by avoid-ing future violations similar to those described in this publica- Reactor Regulation, Director (Post 870411). Jane 1988.161pp. tion. 8806100152.45785:188.

                                                                                                     " Station Blackout" which is the complete loss of alternating NUREG-0940 V07 NO2: ENFORCEMENT ACTIONS:SIGNIFICANT                        current (AC) electrical power in a nuclear power plant, has been ACTtONS RESOLVED. Quarterly Progress Report, April-June                designated as Unresolved Safety issue A-44. Because many 1988.

• Ofc of Enforcement (Post 870413). August 1988. safety systems required for reactor c%e decay heat removal 331pp. 8809290018. 46952:100. and co6tainment heat removal depend on AC power, the conse-This compilation summarizes significant enforcement actions quences of a station blackout could be severe. This repon doc-that have been resolved during one quarterly period (April . uments the findings of technical studies performed as part of June 1988) and includes copies of letters, Notices, and Orders the program to resolve this issue. The important factors ana-sent by the Nuclear Regulatory Commission to licensees with lyzed include: the frequency of loss of offsite power; the proba. j respect to these enforcement actions. It is anticipated that the bility that emergency or onsite AC power supplies would be un- 1 information in this publication will be widely disseminated to available; the capability and reliability of decay heat removal l managers and employees engaged in activities licensed by the systems independent of AC power; and the likelihood that off-NRC, so that actions can be taken to improve safety by avoid. site power would be restored before systems that cannot oper-ing future violations similar to those described in this publica. ate for extended periods without AC power fall, thus resulting in i tion. core damage. This report also addresses effects of different de-  !

signs, locations, and operational features on the estimated fre. NUREG-0940 V07 NO3: ENFORCEMENT ACTIONS:SIGNIFICANT ACTIONS RESOLVED.Ouarterly Progress Report. July.Septem- quency of core damage resulting from station blackout events. ber 1988.

• Ofc of Enforcement (Post 870413). December NUREG-1100 V04: BUDGET ESTIMATES. Fiscal Year 1989.
• Di-1988. 300pp. 8901090253. 48112:001.

vision of Budget & Analysis (Post 870413). February 1988. This compilation summarizes significant enforcement actions 149pp. 8803090033. 44648:223. that have been resolved during one quarterly period (July - Sep. This report contains the fiscal year budget justifications to tember 1988) and includes copies of letters, Notices and Orders Congress. The budget provides estimates for salaries and ex. sent by the Nuclear Regulatory Commission to licensees with penses for Fiscal Year 1989. respect to these enforcement actions it is anticipated that the information in this publication will be widely disseminated to NUREG-1109: REGULATORY /BACKFIT ANALYSIS FOR THE managers and employees engaged in activities licensed by the RESOLUTION OF UNRESOLVED SAFETY ISSUE A-44 STA. NRC, so that actions can be taken to improve safety by avoid- TION BLACKOUT, RUBIN,A.M. Office of Nuclear Regulatory ing future violations similar to those described in this publica- Research, Director (Post 860720). RUBIN,A.M. Office of Nucle-bon. ar Reactor Regulation, Director (Post 870411). June 1988. 70pp. 8806100170. 45785:118. NUREG-0975 V06: COMPILATION OF CONTRACT RESEARCH Station blackout is the complete loss of alternating current FOR THE MATERIALS ENGINEERING BRANCH. DIVISION OF (ac) electric power to the essential and nonessential buses in a ENGINEERING. Annual Report For FY 1987.

• Materials Engi- nuclear power plant; it results when both offsite power and the neering Branch. June 1988. 425pp. 8807110528. 46083.319.

onsite emergency ac power systems are unavailable. Because This report presents summanes of the research work per- many safety systems required for reactor core decay heat re-formed dunng fiscal year 1987 by laboratories and organizations moval and containment heat removal depend on ac power, the under contracts administered by the NRC's Materials Engineer- consequences of a station blackout could be severe. Because ing Branch, Office of Nuclear Regula'ory Research. Each con- of the concern about the frequency of loss of offsite power, the

l 1 10 Main Citations and Abstracts i number of failures of emergency diesel generators, and the po- 131). However, acute fatalities or injuries to people offsite due to accidental releases of these materials do not seem plausible. tentially severe consequences of a loss of all ac power, "Sta-tron Blackout" was designated as Unresolved Safety issue (USI) The only other significant accident was identified as a lorvj-term A-44. This report presents the regulatory /backfit analysis for pulsating cnticality at fuel cycle facilities handling high-ennched USl A-44. It includes (1) a summary of the issue, (2) the recom- uranium or plutonium. An important feature of the most serious mended technical resolution, (3) alternative resolutions consid- accidents is that releases are likely to start without prior warn-ered by the Nuclear Regulatory Commission (NRC) staff, (4) an ing. The releases would usually end within about half an hour.

Thus protective actions would have to be taken quickly to be assessment of the benefits and costs of the recommended res- effective. There is not likely to be enough time for dose projec-olution, (5) the decision rationale, (6) the relationship between USl A 44 and other NRC programs and requhements, and (7) a tions, complicated decisionmaking dunng the accident, or the backfit analysis demonstrator <g that the resolution of USl A-44 participation of personnel not in tfie immediate vicinity of the site. The appropriate response by the facility is to immediately complies wcument presents a plan for thermal-hydraulic research for Babcock and Wilcox designed reactor systems. It describes NUREG-1228: SOURCE TERM ESTIMATION DURING INCIDENT the technical issues, regulatory needs, and the research neces-RESPONSE TO SEVERE NUCLEAR POWER PLANT ACCl- sary to address these needs. The plan also discuses the rela-DENTS. MCKENNAT.J.; GitTTER,J. Division of Operational As- tionship between current and proposed research, and provides sessment (Post 870413). October 1988.174pp. 8811010272. a tentative schedule to complete the required work. .,

47275:118. J l The various methods used by the NRC dunng their response NUREG-1249 V01: NRC MODEL SIMULATIONS IN SUPPORT l to a reactor a cident for estimating radionuclides release to the OF THE HYDROLOGIC CODE INTERCOMPARISON (HYDRO-environment (source terms) as a result of an accident at a nu- COIN) STUDY. Level 1 - Code Verification.

• Division of Engi-clear power reactor are discussed. The major factors affecting neering (Post 870413). March 1988. 184pp. 8804080227. $

potential radionuclides releases off site (source terms) as a result 45042:346. J of nuclear power plant accidents are described. The quantifica- HYDROCOIN, an international study for examining ground-tion of these factors based on plant instrumentation also is dis- water flow modeling strategies and their influence on safety as- {$ cussed A range of accident conditions from those within the sessments of geologic repositories for nuclear waste, has been j design basis to the nost severe accidents possible are included divided into three tevels of effort, Level 1 evaluates the accura-  ; in the text. A method of gross estimation of accident source cy of the computer codes by comparing code results to analyti- [ terms and their consequences off site is presented. The goal is cal solutions and intercomparison of code results. Level 2 tests i to present a method of source term estimation that reflects the the codes' capabilities for simulating specific laboratory or field current understanding of source term behavior and that can be expenments. Level 3 addresses the uncertainty of model re- f used during an event. suits, and the sensitivity of those results to changes in either y the modeling approach or input parameters. A greater variety of E NUREG-1231 S01: SAFETY EVALUATION REPORT RELATED analytic approaches and computer codes were compared in a $ i TO BABCOCK AND WILCOX OWNERS GROUP PLANT REAS- systematic fashion than would have been possible due to the SESSMENT PROGRAM. SIEGEL B.L. Office of Nuclear Reactor large number of participants involved and their corporate efforts. ' Regulation. Director (Post 870411). March 1988. 126pp. This report summarizes only the NRC project teams' simulation 8804080214.45037:044. efforts tw We Level 1 benchmarking problems. The codes used Supplement No. I to the Safety Evaluation Report (SER) re- to simul. 4 these seven probisms were SWIFT 11. FEMWATER, lated to the Babcock & Wilcox Owners Group (BWOG) Plant UNSAT2, USGS-3D, and TOUGH. In generaf, linear problems Reassessment Program has been prepared by the Office of Nu- involving scatars such as hydraulic head were accurately simu-clear Reactor Regulation of the U.S. Nuclear Regulatory Com- lated by both finite difference and finite-element solution algo-mission. This supplement contains the NRC staff's evaluation of nthms. Both types of codes produced accurate results even for the BWOG reassessment of the integrated control system /non- complex geometries such as intersecting fractures. Difficulties nuclear instrumentation system, the emergency feedwater initi- were encountered in solving probiems that involved nonlinear ation and control system, reactor trio initiating events, several effects such as density driven flow and unsaturated flow. In additional open items identified in the SER, and the BWOG order to fully evaluate the accuracy of these codes, post-proc-comments on the SER. essing of results using particle tracking algorithms and calculat-ing fluxes were examined. This proved very valuable by uncov-NUREG-1232 V02: SAFETY EVALUATION REPORT ON TEN- enng disagreements among code results even though the hy-NESSEE VALLEY AUTHORITY. Sequoyah Nuclear Perform- draulic-head solutions had been in agreement. ance Plan.

• Ofc of Special Projects (Pre 881231). May 1988.

362pp. 8805260367. 45659:001. NUREG-1252: NUCLEAR POWER PLANT THERMAL-HYDRAU-This Safew Evaluation (SERi on the information submitted by LIC PERFORMANCE RESEARCH PROGRAM PLAN.

• Division the Tennessee Valley Authonty (TVA) in its Sequoyah Nuclear of Reactor & Plant Systems (870413-880716). July 1988. BMp.

Performance Plan, through Revision 2, and supporting dow- 8808180209.46519:036. monts has been prepared by the U S. Nuclear Regulatory Com- This report provides a description of the thermal-hydraulic re-mission staff. The Plan addresses the plant-specific concerns search to be carried out by the Office of Nuclear Regulatory Re. requinng resolution before startup of either of the Sequoyah search. The subject of this plan is plant performance, and the units. In particular, the SER addresses required actions for Unit emphasis is on prevention of severe accidents. The plan de-2 restart. In many cases, the programmatic aspects for Unit 1 fines the major issues, relates the proposed research to these are identical to those for Unit 2; the staff will conduct inspec- issues, defines needed products, provides an historical perspec-tions of implement & tion for those programs. Where the Unit 1 tive, and defines the major interoffice and interdisciplinary inter. program is different, the staff evaluation will be provided in a faces. The plan provides a technical basis for planning and con-supplement to this SER. On the basis of its review, the staff ducting current and future thermal-hydraulic research. The plan concludes that Sequoyah spec:fic issues have been resolved to is consistent with assumptions and guidance in the NRC Five-the extent that would support restart of Sequoyah Unit 2. Year Plan published in March 1988.

\ d 4 Main Citations and Abstracts 13 NUREG-1260 V02: A REPORT TO CONGRESS ON NUCLEAR and misadministration reports that were reported in 1987 and a f REGULATORY RESEARCH. Project Desenptions For FY88.

• brief synopsis of AEOD studies published in 1987. Each volume Office of Nuclear Regulatory Research, Director (Post 860720). contains a list of the AEOD Reports issued for 1980-1987.

f August 1988. 733pp. 8808300255. 40632:349. The Report to Congress on Nuclear Regulatory Research NUREG-1272 V02 NO2: REPORT TO THE U.S. NUCLEAR REGU-r contains information on research projects. It covers objectives, LATORY COMMISSION ON ANALYSIS AND EVALUATION OF ! major considerations, status, significant findings, regulatory ap- OPERATIONAL DATA - 1987.Nonreactors.

• Office for Analysis

( plications, research completed, planned research for the current & Evaluation of Operational Data, Director. October 1988. 80pp. j fiscal year and planced future research. 8812020172.47690:353. NUREG-1263: HYDROLOGIC DESIGN FOR RIPRAP ON EM. See NUREG 1272,V02 N01 abstract. BANKMENT SLOPES. CODELL.R.D. Division of High Level Wast Manage Post 870413). September 1988. 102pp. NUREG-1273: TECHNICAL FINDINGS AND REGULATORY ANALYSIS FOR GENERIC SAFETY ISSUE II.E.4.3, "CONTAIN-Waste impoundments'for uranium tailings and other hazard- MENT INTEGRITY CHECK." SERKlZ,A.W. Division of Reactor ous substances are often protected 'by compacted earth and & Plant Systems (870413-880716). April 1988. 170pp. clay, covered with a layer of loose rock (np-rap). The report out. 8805030120.45343:024. lines procedures that could be followed to design nprap to with- This report contains the technical findings and regulatory l stand forces causad by runoff resulting from extreme rainfall di- analysis for Generic Safety issue ILE.4.3, " Containment integrity rectly on the embankment. The Probable Maximum Precipitation Check." An evaluation of the containment isolation history from for very small areas is developed from considerations of severe 1965 to 1983 reveals that (except for a small number of events) storrns of s+iort duration at mid- latitudes. A two-dimensional containment integrity has been maintained and that the majority finite difference modelis then used to calculate the runoff from of reported events have been events related to exceeding severe rainfall events. The procedure takes into account flow Technical Specification limits (or 0.6 times the allowable leak-both beneath and above the rock layer and approximates the age level). In addition, more recent risk analyses have shown concentration in flow which could be caused by a non- level or that allowable leakage rates even if increased by a factor of 10 slumped embankment. The sensitivity to various assumptions, would not significantly increase nsk Potential method of contin-such as the shape and size of the rock, the thickness of the vous monitonng are identified and evaluated. Therefore, these ( layer, and the shape of the embankment, suggests that peak technical findings and risk evaluations support closure of Gener. runoff from an armored slope could be attenuateu with proper ic issue ILE.4.3. design. Frictional relationships for complex flow regimes are de-veloped on the basis of flow through rock-filled dams and in NUREG-1275 V03: OPERATING EXPERIENCE FEEDBACK mountain streams. These relationships are tested against exper-imental data collected in laboratory flumes; the tests provide ex- REPORT - SERVICE WATER SYSTEM FAILLKRES AND j eeDent results. The resulting runoff is then used in either the DEGRADATIONS. Commercial Power Reactors. LAM.P.; Stephenson or safety factor method to find the stable rock di- LEEDS.E. Office for Analysis & Evaluation of Operational Data, l Director. November 1988.156pp. 8812190085. 47804:308. l ameter. The rock sizes determined by this procedure for a given j flow have been compared with data on the failure of rock layers A comprehensive review and evaluation of service water in expenmental fiumes, again with excellent results. Computer system failures and degradations observed in operating events programs are included for implementing the method- in hght water reactors from 1980 to 1987 has been conducted. The review and evaluation focused on the identification of NUREG-1266 V02: NRC SAFETY RESEARCH IN SUPPORT OF causes of system failures and degradations, the adequacy of REGULATION - 1987.

• Othee of Nuclear Regulatory Research, corrective actions implemented and planned, and the safety sig-Director (Post 860720). May 1988. 59pp. 8805200095. nificance of the operating events. The results of this review end 45549:327.

evaluation indicate that the service water system failures and This report, the third in a series of annual reports, was pre. degradations have significant safety implications. These system pared in response to congressional inquiries conceming how failures and degradations are attributable to a great variety of nudear regulatory research is used. It summarizes the accom-causes, and have adverse impact on a large number of safety-ph;.hments of the Office of Nuclear Regulatory Research dunng related systems and components which are required to mitigate 1967. The goal of this office is to ensure that research provides reactor accidents. The high safety sigmftcance associated with the technical bases for rulemaking and for related decisions in service water system failures and degradations warrants correc-support of NRC heensing and inspection activities. This report live actions to reduce both the frequency and potential conse-describes both the dwoct contnbutions to scientific and technical quences of operating events involving such failures and degra-knowledge with regard to nuclear safety and their regulatory ap- dations. To this end, the Office for Analysis and Evaluation of phcations-Operational Data has developed several recommendations NUREG-1272 V02 N01: REPORT TO THE U.S. NUCLEAR REGU- which are summarized in the report. LATORY COMMISSION ON ANALYSIS AND EVALUATION OF OPERATIONAL DATA - 1987. Power Reactors.

• Office for Anal- NUREG-1283: SAFETY EVALUATION REPORT RELATED TO ysis & Evaluatic, of Operational Data. Director. October 1988. THE RENEWAL OF THE OPERATING LICENSE FOR THE RE-216pp. 8811110038, 47528.287. SEARCH REACTOR AT PURDUE UNIVERSITY,
• Standardize-This annual report of the U.S. Nuclear Regulatory Commis, tion & Non-Power Reactor Project Directorate. April 1988.69pp.

sion's Office for Analysis and Evaluatron of Operational Data 8805090116. 45432:303. (AEOD) is devotsd to the activities performed dunng 1987. The This Safety Evaluation Report for the application filed by l report is published in two volumes. NUREG-1272, Vol. 2, No.1 Purdue University for a renewal of Operating License R 87 to covers Power Reactors end presents an overview of the operat- continue to operate a research reactor has been prepared by ing expenence of the nuclear pcwer industry, with comments re- the Office of Nuclear Reactor Regulation of the U.S. Nuclear garding the trends of some key performance rrmasures The i Regulatory Commission. The facihty is owned by Purdue Univer-report also includes the principal findings and Gsues identified sity and is located on the campus in West Lafayette, Indiana. in AEOD studies over the past year and summanzes information On the basis of its technical Teview, the staff concludes that the from Licensee Event Reports, the NRC's Operations Center, reactor facility can continue to be operated by the University and Diagnostic Evaluations. NUREG-1272, Vol. 2, No. 2, covers l without endangenng the health and safety of the public or the Nonreactors and presents a review of the nonreactor events environment. l

14 Main Citations and Abstracts NUREG-1286 S01: SAFETY EVALUATION REPORT RELATED about the allowed policy of bypassing thermal overload devices TO THE RESTART OF RANCHO SECO NUCLEAR GENERAT- dunng normal conditions. Regulatory Guide 1.106 favors com-ING STATION. UNIT 1,FOLLOWING THE EVENT OF DECEM- promising the function of thermal overload devices in order to BER 26,1985. Docket No. 50-312.(Sacramento Municipal Utility avoid interfering with the safety-related operation of motor-oper. District)

• Division of Reactor Projects - lil,lV.V & Special ated valves. This report describes thermal overload devices and Projects (Post 870411). March 1988. 233pp. 8804080230. their use. It is concluded that the policies stated in Regulatory 4504 t:231. Guide 1.106 are technically correct and allow sufficient flexibility On December 26,1985, the Rancho Seco Nuclear Generat- to allow the use of thermal overload protection without interfer-ing Station expenenced a reactor tnp from 76% ' power, followed ing with safety related functions of motor-operated valves. How-by a rapid overcooling transient and automatic initiation of the ever, it appears that licensees are needlessly bypassing or oth-safety features actuation system. The unit has remained shut erwise compromising the use of thermal overload protection.

down since that time. In response to confirmatory letters from Some licensees are using inadequate design practices to size the NRC Region V Administrator, the licensee, Sacramento Mu- thermal overload devices. The problem of valve motor burnout nicipal Utihty Distnct (SMUD), submitted the "Ranaho Seco is related to a lack of standards and uniform guidance for the Action Plan for Performance improvement" in July 1986. Since design, installation, maintenance, and testing of motor overload then, the licensee has submitted revisions to that action plan protective devices. The NRC's Office of Nuclear Regulatory Re-and numerous other documents and information to support a search will contact several nuclear standards organizations to return of Rancho Seco to power operation. The NRC staff re- suggest that detailed guidance for thermal overload protection viewed the licensee's submittals and other information made of motor-operated valves be developed. available to the staff in support of a restart of Rancho Seco. In October 1987, the NRC staff issued a Safety Evaluation Report NUREG-1297: PEER REVIEW FOR HIGH-LEVEL NUCLEAR (NUREG-1286) relating to the restart of Ranch Seco. Since WASTE REPOSITORIES. Generic Technica! Position. then, the staff has completed its review of all other issues relat- ALTMAN,W.D.; DONNELLY,J.P.; KENNEDY,J.E. Division of ing to the restart effort. The results of this more recently com- High Level Waste Management (Post 870413). February J908. I l pleted review work are contained in this Supplement No.1 to 30pp. 8802240192. 44468:269. NUREG-12B6. This document provides guidance on the use of the peer REGULATORY AND BACKFIT review process in the high-level nuclear waste repository pro-NUREG-1289: ANALYSIS: UNRESOLVED SAFETY ISSUE A-45, SHUTDOWN gram. The applicant must demonstrate in the license application that the applicable health, safety, and environmental regulations DECAY HEAT REMOVAL REQUIREMENTS.

• Division of Safety issue Resolution (Post 880717). November 1988.179pp. in 10 CFR Part 60 have been met. Confidence in the data used 8811220510.47602:142. to support the license application is obtained through a quality All light water reactors require decay heat to be removed sub- assurance (OA) program as described in 10 CFR 60, Subpart G.

sequent to reactor shutdown. Interruption of the decay heat re- Peer reviews may be used as part of the OA actions necessary I moval function could lead to bevore consequences. Concerns to provide confidence in the work being reviewed. Because of about the reliabihty of the systems and components that essist several unique conditions inherent to the geologic repository l in the decay heat removal resulted in establishing the require- program, expert judgment will need to be utilized in assessing ments for decay heat removal as an unresolved safety issue the adequacy of work. Peer reviews are a mechanism by which (USI) designated USl A-45, " Shutdown Decay Heat Removal these judgments are made. This document provides guidance Requirements." This report presents the regulatory analysis for on areas where a peer review is appropriate, the acceptability of USl A-45. It includes (1) a summary of the issue, (2) the pro- peers, and the conduct and documentation of a peer review. posed technical resciution, (3) alternative resolutions consid-ered by the Nuclear Regulatory Comrnission (4) an assessment NUREG-1298: QUALIFICATION OF EXISTING DATA FOR HIGH-of the benefits and costs of all alternatives considered, and (5) LEVEL NUCLEAR WASTE REPOSITORIES. Generic Technical the decision rationale. Position. ALTMAN,W.D.; DONNELLY,J.P.; KENNEDY,J.E. Divi-NUREG-1290 ADD: DIFFERING PROFESSIONAL OPIN- f98827p8802 0325 700 6 IONS.1987 Special Review Panel.

• NRC No Detailed Affili- This document provides guioance on methods of qualifying ation Given. January 1988. 9pp. 8802020172. 44188:347. data not initially collected under a 10 CFR Part 60, Subpart G in November 1987, the five-member Differing Professional quality assurance (OA) program. The license apphcant for a Opinions Special Review Panet established by the Executive Di- geologic repository must demonstrate that the apphcable health, rector for Operations of the U.S Nuclear Regulatory Commis- safety, and environmental regulations in 10 CFR Part 60 have sion to review agency policies and procedures for handling dif- been met. Confidence in the data used to support the license tenng professional opinions (DPOs) presented its findings and appkcation is obtained through a OA program. Some data which recommendations in NUHEG-1290. The issuance of that report have not been initially generated under a 10 CFR Part 60, Sub-completed the first task of the Panels charter. In accordance part G OA program may be needed to support a license appli-with Manual Chapter 4125, Section L, and the charter of the cation to construct and operate a geologic repository. This doc-Special Review Panel, the Panel,s second task was t ument provides guidance on the use and qualification of data

       ... review...the DPOs submitted subsequent to the previous        not iriitially collected under a Subpart G QA program.

Panel's review, in order to identify any employee whose DPO made a significant contribution to the Agency or to the public NUREG-1303: INCIDENT INVESTIGATION MANUAL.

• Office for safety but who has not yet been recognized for such contnbu- Analysis & Evaluation of Operational Data, Director. February tion." This Addendum provides the findings of that review. 1988. 235pp. 8804080191. 45037:220.

NUREG-1296: THERMAL OVERLOAD PROTECTION FOR ELEC- The incicient investigation Manual presenbes guidehnes for TAIC MOTORS ON SAFETY-RELATED MOTOR-OPERATED the conduct of investigative activities of the U.S. Nuclear Regu-VALVES - GENERIC ISSUE 11 E.6.1. ROTHBERG,0. Division of latory Commission (NRC), Incident Investigation Teams (IITs). Engineenng (Post 870413) June 1988. 51pp 8807110511. The purpose of this manual is to provide llTs guidance to 46083:122. ensure that NRC investigations of significant events are timely, The NRC regulatory positions, as stated in Regulatory Guide structured, coordinated, and formally administered. The guide-1.106. Revision 1, have been identified by the Office for Analy- lines are intended to assist the investigation rather than limit the sis and Evaluation of Operational Data (AEOD) as potential con- initiative and good judgment of the team leader or members, inbutors to valve motor burnout. AEOD is particularly concerned they should use their expenence and those techniques that pro-

k ii t d Main Citations and Abstracts 13 0 NUREG-1260 V02: A REPORT TO CONGRESS ON NUCLEAR and misadministration reports that were reported in 1987 and a )y REGULATORY RESEARCH Project Desenptions for FY88.

• brief synopsis of AEOD studies published in 1987. Each volume Office of Nuclear Regulatory Research. Director (Post 860720). contains a list of the AEOD Reports issued for 1980-1987.

August 1988. 733pp. 8808300255. 46632.349. The Report to Congress on Nuclear Regulatory Research NUREG-1272 V02 NO2: REPORT TO THE U.S. NUCLEAR REGU- [ contains information on research projects. It covers objectives, LATORY COMMISSION ON ANALYSIS AND EVALUATICN OF

&        major considerations, status, sigriificant findings, regulatory ap-      OPERATIONAL DATA - 1987.Nonreactors.

• Office for Analysis phcations, research completed, planned research for the current & Evaluation of Operational Data Director. October 1988.80pp.

l fiscal year and planned future research. 8812020172. 47690:353. NUREG-1263: HYDROLOGIC DESIGN FOR RIPRAP ON EM. . See NUREG-1272,V02,N01 abstract. BANKMENT SLOPES. CODELL,R.B. Division of High Level

                                                          "                   NUREG-1273: TECHNICAL FINDINGS AND REGULATORY 810 10307 47 43 3 9                                                    ANALYSIS FOR GENERIC SAFETY ISSUE il.E.4.3, "CONTAIN-Waste impoundments for uranium tailings and other hazard.             MENT INTEGRITY CHECK." SERKlZ,A.W. Division of Reactor ous substances are often protected by compacted earth and                 & Plant Systems (870413-880716). April 1988. 170pp.

clay, covered with a layer of loose rock (np rap). The report out. 8805030120.45343:024. lines procedures that could be followed to design nprap to with- This report contains the technical findings and regulatory stand forces caused by runoff resulting from extreme rainfall di- analysis for Generic Safety issue it.E.4.3, " Containment integrity rectly on the embankment. The Probable Maximum Precipitation Check." An evaluation of the containment isolation history from for very small areas is developed from considerations of severe 1965 to 1983 reveals that (except for a small number of events) storms of short duration at mid- latitudes. A two-dimensional containment integrity has been maintained and that the majority finite difference modelis then used to calculate the runoff from of reported events have been events related to exceeding severe rainfall events. The procedure takes into account flow Technical Specification limits (or 0.6 times the allowable leak-both beneath and above the rock layer and approximates the age level). In addition, more recent risk analyses have shown concentration in flow which could be caused by a non- level or that allowable leakage rates even if increased by a factor of 10 slumped embankment. The sensitivity to vanous assumptions. would not significantly increase nsk. Potentia: method of contin-such as the shape and size of the rock, the thickness of the uous monitonng are identified and evaluated. Therefore, these layer, and the shape of the embankment, suggests that peak technical findings and nsk evaluations support closure of Gener-runoff from an armored slope could be attenuated with proper ic issue (i.E.4.3. design. Frictional relationships for complex flow regimes are de-veloped on the basis of flow through rock-filled dams and in NUREG-1275 V03: OPERATING EXPERIENCE FEEDBACK mountain streams. These relationships are tested against exper. imental data collected in laboratory flumes; the tests provide ex- FEPORT - SERVICE WATER SYSTEM FAILURES AND cellent results. The resulting runoff is then used in either the DEGRADATIONS. Commercial Power Reactors. LAM.P.; Stephenson or safety factor method to find the stable rock di- LEEDS.E. Office for Analysis & Evaluation of Operational Data, Director. November 1988.156pp. 8812190085. 47804:308. ameter. The rock sizes determined by this procedure for a given flow have been compared with data on the failure of rock layers A comprehensive review and evaluation of service water in experimental fiumes, again with excellent results. Computer system failures and degradations observed in operating events programs are included for implementing the method. in light water reactors from 1980 to 1987 has been conducted. The review and evaluation focused on the identification of NUREG-1266 V02: NRC SAFETY RESEARCH IN SUPPORT OF causes of system failures and degradations, the adequacy of REGULATION - 1987.

• Office of Nuclear Regulatory Research, corrective actions implemented and planned, and the safety sig.

Director (Post 860720). May 1988. 59pp. 8805200095. nificance of the operating events. The results of this review and 45549:327. evaluation indicate that the service water system failures and This report, the third in a senes of annual reports, was gere- degradations have significant safety implications. These system pared in response to congressional inquines concerning how failures and degradations are attnbutable to a great variety of nucioar regulatory research is used. It summarizes the accom-causes, and have adverse impact on a large number of safety-plishments of the Office of Nuclear Regulatory Research dunng related systems and components which are required to mitigate 1987. The goal of this office is to ensure that research provides reactor accidents. The high safety significance associated with the technical bases for rulemaking and for related decisions in j service water system iailures and degradations warrants correc- j support of NRC licenang and inspection activities This report tive actions to reduce both the frequency and potential conse- 1 desenbes both the direct contributions to scientific and technical Anowledge with regard to nuclear safety and their regulatory ap-quences of operating events involving such failures and degra-dationn. To this end, the Office for Analysis and Evaluation of l plications. l Operational Data has developed several recommendations I NUREG 1272 V02 N01: REPORT TO THE U.S. NUCLEAR REGU- which are summarized in the report. ' LATORY COMMISSION ON ANALYSIS AND EVALUATION OF OPERATIONAL DATA - 1987. Power Reactors.

• Office for Anal, NUREG 1283: SAFETY EVALUATION REPORT RELATED TO

    " sis & Evaluation of Operational Data, Director. October 1988.            THE RENEWAL OF THE OPERATING LICENSE FOR THE RE-216pp. 8811110038. 47528.287.                                              SEARCH REACTOR AT PURDUE UNIVERSITY.

• Standardize-This annual report of the U.S. Nuclear Regulatory Commis. tion & Non-Power Reactor Project Directorate. April 1988.69pp.

aion's Office for Analysis and Evaluation of Operational Data 8805090116. 45432:303. (AEOD) is devoted to the activities performed dunt g 1987. The This Safety Evaluation Report for the application filed by report is published in two volumes. NUREG-1272 Vol. 2 No.1, Purdue University for a renewal of Operating License R-87 to l covers Power Reactors and presents an overview of the operat. continue to operate a research reactor has been prepared by ing exponence of the nuclear power industry, with comments re. the Office of Nuclear Reactor Regulation of the U.S. Nuclear garding the trends of some key performance measures. The Regulatory Commission. The facility is owned by Purdue Univer. report also includes the pnncipal findings and issues identified sity and is located on the campus in West Lafayette, Indiana. in AEOD studies over the past year eind summanzes information On the basis af its technical review, the staff concludes that the from Licensee Event Reports, the NRC's Operations Center, reactor facility can continue to be operated by the University and Diagnostic Evaluations NUREG-1272, Vol. 2, No. 2, covers j without endangenng the health and safety of the public of the Nonreactors and presents a review of the nonreactor events environment. I

14 Main Citations and Abstracts NUREG-1286 S01: SAFETY EVALUATION REPORT RELATED about the allowed policy of bypassing thermal overload devices TO THE RESTART OF RANCHO SECO NUCLEAR GENERAT- during normal conditions. Regulatory Guide 1.106 favors com-ING STATION,0 NIT 1,FOLLOWING THE EVENT OF DECEM- promising the function of thermal overload devices in order to BER 26,1985. Docket No. 50-312.(Sacramento Municipal Utility avoid interfering with the safety-related operation of motor-oper-District)

• Division of Reactor Projects - fillV,V & Special ated valves. This report describes thermal overload devices and Projects (Post 870411). March 1988. 233pp. 8804080230. their use. It is concluded that the policies stated in Regulatory 4504t231. Guide 1.106 are technically correct and allow sufficient flexibiWty On December 26,1985, the Rancho Seco Nuclear Generat- to allow the use of thermal overload protection without interfer-ing Station experienced a reactor inp from 76*4 power, followed ing with safety-related functions of motor operated valves. How-by a rapid overcooling transient and automatic initiation of the ever, it appears that licensees are needlessly bypassing or oth-safety features auuation system. The unit has remained shut erwise compromising the use of thermal overload protection.

down since that time. In response to confirmatory letters from . Some licensees are using inadequate design practices to size the NRC Region V Administrator, the licensee, Sacramento Mu- thermal overload devices. The problem of valve motor burnout nicipal Utility District (SMUD), submitted the " Rancho Seco is related to a lack of standards and uniform guidance for the Action Plan for Performance Improvement" in July 1986. Since design, installation, maintenance, and testing of motor overload then, the licensee has submitted revisions to that action plan protective devices. The NRC's Office of Nuclear Regulatory Re-and numerous other documents and information to support a search will contact several nuclear standards organizations to return of Rancho Seco to power operation. The NRC staff re- suggest that detailed guidance for thermal overload protection viewed the licensee's submittals and other information made of motor operated valves be developed. available to the staff in support of a restart of Rancho Seco in October 1987, the NRC staff issued a Safety Evaluation Report NUREG-1297: PEER REVIEW FOR HIGH-LEVEL NUCLEAR (NUREG-1286) relating to the restart of Ranch Seco. Since WASTE REPOSITORIES. Generic Technical Position. then, the staff has completed its review of all other issues relat- ALTMAN,W.D.: DONNELLY,J.P.; KENNEDY,J.E. Division of ing to the restart effort. The results of this more recently com- High Level Waste Management (Post 870413). February 1988. pleted review work are contained in this Supplement No.1 to 30pp. 8802240192. 44468:269. < NUREG-1286. This document provides guidance on the use of the peer l l AND BACKFIT review process in the high-level nuclear wasto repository pro- 1 NUREG-1289: REGULATORY ANALYSIS: UNRESOLVED SAFETY ISSUE A-45. SHUTDOWN gram. The applicant must demonstrate in the license application that the applicable health, safety, and environmental regulations ] DECAY HEAT REMOVAL REQUIREMENTS.

• Division of Safety issue Resolution (Post 880717). November 1988.179pp. in 10 CFR Part 60 have been met. Confidence in the data used i to support the license application is obtained through a quality 8811220510.47602:142.

All light water reactors require decay heat to be removed sub- assurance (OA) program as described in 10 CFR 60, Subpart G. sequent to reactor shutdown. Interruption of the decay heat re. Peer reviews may be used as part of the QA actions necessary I moval function could lead to severe consequences. Concerns to provide confidence in the work being reviewed. Because of ) about the reliability of the systems and components that assist several unique conditions inherent to the geologic repository in the decay heat removal resulted in establishing the require- program, expert judgment will need to be utilized in assessing ments for decay heat removal as an unresolved safety issue the adequacy of work. Peer reviews are a mechanism by which (USI) designated USl A-45, Shutdown Decay Heat Removal these judgments are made. This document provides guidance ] Requirements." This report presents the regulatory analysis for on areas where a peer review is appropriate, the acceptability of USl A 45. It includes (1) a summary of the issue, (2) the pro- peers, and the conduct and documentation of a peer review. posed technical resolution, (3) alternative resolutions consid-ered by the Nuclear Regulatory Commission, (4) an assessment NUREG-1298: OUAllFICATION OF EXISTING DATA FOR HIGH-of the benefits and costs of all attematives considered, and (5) LEVEL NUCLEAR WASTE REPOSITORIES. Generic Technical Position. ALTMAN,W.D.; DONNELLY,J.P.: KENNEDY,J.E. Divi-the decision rationale. NUREG-1290 ADD: DIFFERING PROFESSIONAL OPIN- f988 7 p 8802 40325 4470 6 IONS.1987 Special Review Panel.

• NRC - No Detailed Affili- This document provides guidance on methods of qualifying ation Given. January 1988. 9pp. 8802020172. 44188:347. data not initially collected under a 10 CFR Part 60, Subpart G in November 1987, the five. member Differing Professional quality assurance (OA) program. The license applicant for a Opinions Special Review Panel established by the Executive Di- geologic repository must demonstrate that the applicable health, rector for Operations of the U.S. Nuclear Regulatory Commis- safety, and environmental regulations in 10 CFR Part 60 have sion to review agency policies and procedures for handhng dif- been met. Confidence in the data used to support the license tenng professional opinions (DPOs) presented its findings and application is obtained through a OA program. Some data which recommendations in NUREG-1290. The issuance of that report have not been initially generated under a 10 CFR Part 60, Sub-completed the first task of the Panels charter. In accordance art G OA program may be needed to support a license appli-with Manual Chapter 4125, Section L, and the charter of the cation to construct and operate a geologic repository. This doc-Special Review Panet, the Panel,s second task was to ument provides guidance on the use and qualification of data

       ... review...the DPOs submitted subsequent to the previous          not initially collected under a Subpart G OA program.

Paners review, in order to identify any employee whose DPO made a significant contribution to the Agency or to the public NUREG-1303: INCIDENT INVESTIGATION MANUAL

• Office for safety but who has not yet been recognized for such contribu- Analysis & Evaluation of Operational Data, Director. February tion. ' This Addendum provides the findings of that review. 1988. 235pp. 8804080191. 45037:220.

NUREG 1296: THERMAL OVERLOAD PROTECTION FOR ELEC- The incident investigation Manual prescribes guidelines for TAIC MOTORS ON SAFETY RELATED MOTOR OPERATED the conduct of investigative activities of the U.S. Nuclear Regu-VALVES GENERIC ISSUE II.E.6.1. ROTHBERG.O. Division of latory Commission (N9C), incident investigation Teams (IITs). Engineenng (Post 870413). June 1988. 51pp. 8807110511. The purpose of this manual is to provide llTs guidance to 46083:122. ensure that NRC investigations of significant events are timely, The NRC regulatory positions, as stated in Regulatory Guide structured, coordinated, and formally administered. The guide-1.106, Revision 1, have been identified by the Office for Analy- lines are intended to assist the investigation rather than limit the sis and Evaluation of Ope itional Data (AEOD) as potential con- initiat?ve and good judgment of the team leader or members; tributors to valve motor burnout. AEOD is particularly concerned they should use their experience and those techniques that pro-

Main Citations and Abstracts 15 vide the most confidence in assunng the llT objectives are missioning cost estimates that is acceptable to the NRC, and achieved. contains values for the escalation of radioactive waste burial costs, by site and by year. The licensees may use the formula, NUREG-1304: REPORTING OF SAFEGUARDS EVENTS. DWYER,P.A.; ERVIN,N E. Division of Reactor inspection & the coefficients, and the burial escalation factors from this report Safeguards (Post 870411). February 1988. 25pp. 8802240225. in their escalation analysis, or may use an escalation rate at 44469:004. least equal to the esca!ation approach presented herein. On June 9,1987, the Commission published in the " Federal Register" a final rule revising the reporting requirements for NUREG 1308: RADIOACTIVE MATERIAL IN THE WEST LAKE LANDFILL. Summary Report.

• Division of industrial & Medical safeguards events. Safegue.rds events include actual or at' Nuclear Safety (Post 870729). April 1988. 22pp. 8805090165.

tempted theft of special nuclear matenal (SNM); actual or at-tem,cted acts or events which interrupt normal operations at 45412:328. power reactors due to unauthonzed use of or tampenng with The West Lake Landfill is located near the city of St. Louis in machinery, components, or controls; certet threats made Bridgeton, St. Louis County, Missouri. The site has been used against facilities possessing SNM; and safer 'ards system fail- s nce 1962 for disposing of municipel refuse, industrial solid and ures impacting the effectiveness of the system. The revised rule liquid wastes, and construction demolition debris. This report was effective October 8,1987. On September 14,1987, the summarizes the circumstances of the radioactive material found NRC held a workshop in Bethesda, MD to answer affected 11- n the West Lake Landfill. Primary emphasis is on the radiologi-censees' qucstions on the final rule. This report documents cal environmental aspects as they relate to potential disposition questions discussed at the September 14 meeting, reflects a d N Merid' completed staff review of the answers, and supersedes previ-ous oral comment on the topics covered. NUREG-1308 R01: RADIOACTIVE MATERIAL IN THE WEST LAKE LANDFILL. Summary Report.

• Division of Industrial &

NUREG-1305: TECHNICAL SPECIFICATIONS FOR SOUTH Medical Nuclear Safety (Post 870729). June 1988. 23pp. TEXAS PROJECT, UNIT 1. Docket No. 50-498.(Houston Lighting 8808050231,46389 004. And Power Company)

• Division of Reactor Projects - lit.IV,V & The West Lake Landfill is located near the city of St. Louis in Special Projects (Post 870411). March 1988. 446pp. Bridgeton, St. Louis County, Missourt. The site has been used 8804080208.45035:318. since 1962 for disposing of municipal refuse, industrial solid and The South Texas Project, Unit No.1, Technical Specifications liquid wastes, and construction demolition debns. This report were prepared by the U.S. Nuclear Regulatory Commission to summenzes the circumstances of the radioactive material in the set forth the limits, operating conditions, and other requirements West Lake Landfill. The radioactive material resulted f:om the applicable to a nuclear reactor facility as set forth in Section processing of uranium ores and the subsequent sale by the ,

30.36 of 10 CFR 50 for the protection of the health and safety Atomic Energy Commission of the processing residues. Primary I of the public. emphasis is on the radiological environmental aspects as they NUREG-13% NRC SAFET) SIGNIFICANCE ASSESSMENT relate to potential disposition of the material. It is concluded TEAM REPORT ON ALLEGATIONS RELATED TO THE SOUTH that remedial action is called for. TEXAS PROJECT, UNITS 1 & 2.

• Division of Reactor Projects -

Ill.IV,V & Special Projects (Post 870411). March 1988.151pp' NUREG-1309: THE U.S. NUCLEAR REGULATORY COMMISSION 8804200527,45274:151 PROGRAM WITH STATE AND LOCAL GOVERNMENTS AND This report provides the results of a review by the Safety Sig- INDIAN TRIBES. DROGGITIS,S.C. State, Local & Indian Tribe nificance Assessment Team (SSAT) of the Nuclear Regulatory Programs. March 1988. 50pp. 8804080200. 45037:170. Commission (NRC) of alleged construction irregulanties at The April 12,1987 reorganization of the Nuclear RegJiatory Houston Lighting and Power Company's South Texas Project Commission created State Local and Indian Tribe Programs (STP), Units 1 and 2 (Docket Nos. 50-498 and 50-499), located (SLITP) within the Offico of Govemmental and Put;lic Affairs. in Matagorda County, Texas. These allegations were provided The creation of SLITP and the goals and objectives stated in to the NRC by the Govemment Accountabihty Protect (GAP) the NRC Strategic Plan concerning State initiatives provided an which received them from approximately 35 current and former opportunity to examine NRC's relations with State and local employees of STP, and covered a wide range of concems with governments and Indian Tribes, and to refocus them, as appro-hardware and quality assurance and control, and issues of man- pnate. The result of this review is attached. agement, harassment and intimidation and wrongdoing. Only those concems considered by the SSAT to be technically-ori- NUREG-1310: NATURALLY OCCURRING AND ACCELERATOR-ented were selected for review based on their possible safety PRODUCED RADIOACTIVE MATERIALS.1987 Review. significance, genene implications, specificity to a particular plant AUSTIN.J.H. Office of Nuclear Material Safety & component, system or structure, and to provide a multidiscipline Safeguards, Director. March 1988. 74pp. 8804080174. overview of the implementation and effectiveness of the STP 45201:180 Quality Assurance Program. The SSAT review of GAP's allega-From time to time, the issue as to whether the U.S. Nuclear tions has ,dentified i no substantive safety issue that would war- Regulatory Commission (NRC) shou!( seek legislative authority ran de y in the NRC s consideration of a full-power licunse for to regulate naturally occurring and accelerator-produced radio-active matenals (NARM) is raised. Because NARM erists in the environment, in homes, in workplaces, in medical institutions, NUREG-1307: REPORT ON WASTE BURIAL and in consumer products, the issue of Federal controls over CHARGES. Escalation Of Decorrmissioning Waste Disposal NARM is very old and very complex. This report presents a Costs At Low-Level Waste Durial Facilities.

• Division of Engi- review of NARM sources and uses as well as incidents and l

neering (Post 870413). Juh 1988. 33pp. 8809230425. prolems associated with those rnaterials. A review of previous 3 46576.023. congressional ar.d Federal agency actions on radiation protec-One of the requirements placed upon nuclear power reactor ti- tion matters, in general, and on NARM, in particular, is provided censees by the U.S. Nuclear Regulatory Commission (NRC) is to develop an understanding of existing Federal regulatory activ-for the licensees to penodically adjust the estimate of the cost ity in ionizing radiation and in control of NARM. In addition, of decommissioning their plant, in dollars of the current year, as State controls over NARM are reviewed. Eight Questions are ex. part of the process to provide reasonable assurance that ade- amined in terms of whether the NRC should seek legislative au-quate funds for decommissioning will be available when thonty to regulate NARM. The assessment of these questions needed This report, which is scheduled to be revised annually, serves as the basis for developing and evaluating five options. contains the development of a formuta for escalating decom- The evaluation of those oDtions leads to two recommendations.

I i l l 16 Main Citations and Abstracts  ! l l NUREG-1311: FUNDING THE NPC TRAINING PROGRAM FOR containment and isolation requirements are defined as "impor-STAT ES. LUBENAU,J.O.; CORLEY,J.H.; KAMMERER,C. State, tant to waste isolation." These structures, systems, compo-Local & Indian Tnbe Programs. Juna 1988. 7'pp. 8807070455. nents, and barriers, and the activities related to their character. 46011:276. ization, desip, construction and operation are required to meet j On February 3,1988, the Commission received a briefing on quality assurance (QA) criteria to provide confidence in the per-the State, Local and Indian Tribes Programs of the NRC Office formance of the geologic repository. The list of structures, sys-of Governmental and Public Affairs. The bnefing included dis- tems, and components important to safety and engineered bar. cussion of the Agreernent State program and, more particularly, riers important to waste isolation is referred to as the "O-List" the training provided to State personnel to help them maintain and lies within the scope of the OA program. programs which are adequate to protect public health and safety and compatible with the Commission's program. The NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT Commission endorsed the NRC State training program but ANALYSIS HANDBOOK. AYER,J.E.; CLARK.A.T.; LOYSEN,P.; questioned the long-standing practice of paying the travel and et al. Division of Industrial & Medical Nuclear Safety (Post per diem of State personnel approved to attend the NRC spon- 870729). May 1988. 527pp. 8806100179. 45788:258. sored training. The staff was requested by the Commission t The Nuclear Reguletory Commission (NRC) has sponsored a provide a report on this aspect. This report is the staff s re- research program to develop improved methods for realistically sponse. It includes an evaluation of the practice of funding g State travel and per diem costs for personnel attending NRC courses and other options to make this program more cost ef- cycle facilities. These methods, along with supporting informa-fective including 'utihzing whenever possible minimal cost Feder. P - al and commercial training facilities. Four types of facilities: fuel manufacturing, fuel reprocessing, waste storage and solidification, and spent fuel storage: and six NUREG-1313: SAFETY EVALUATION REPORT RELATED TO accident types: fires, explosions, spills, tornadoes, nuclear criti-THE EVALUATION OF LOW-ENRICHED URANIUM SILICIDE' cahties, and equipment failures, are cover 6d. Both computer ALUMINUM DISPERSION FUEL FOR USE IN NON-POWER and hand calculations are presented to estimate the source REACTORS.

• Standardization & Non-Power Reactor Project term from potential accidents. This source term information is Directorate. July 1988. 202pp. 8808300273. 46631:308. used as input to computer codes which calculate the transport Low-enriched uranium siheide-aluminum dispersion plate-type and release of mass, ener9y, and material throughout the facili-fuels have been extensively researched and developed under

           'he international program, Reduced Enrichment in Research            ty and to the environment, and Test Reactors. The international effort was led by Argonne NUREG-1325: DISPOSITION OF RECOMMENDATIONS OF THE National Laboratory (ANL) in the United States. This evaluation NATIONAL RESEARCH COUNCIL IN THE REPORT "REVITAL-is based primarily on reports issued by ANL that discuss and summanze the developmental tests and expenments, including           IZING NUCLEAR SAFETY RESEARCH"

• Office of Nuclear postirradiation examinations, of both miniature and full-sized Regulatory Research, Director (Post 860720). June 1988. 43pp.

8807060093.46002:293, plates of prototypical fuel compositions. This evaluation con. cludes that plate type fuels suitable and a::ceptable for use in On December 8,1986, the Committee on Nuclear Safety Re-research and test reactors can be fabricated with U(3)Si(2)-Al search of the National Research Council submitted its report, dispension compacts with uranium densities up to 4.8 g/cm(3). " Revitalizing Nuclear Safety Research," to the U.S. Nuclear Regulatory Commission (NRC). The Commission and its staff NUREG-1317 DRFT FC: REGULATORY OPTIONS FOR NUCLE- have carefully reviewed the Committee s report and have exten-AR PLANT LICENSE RENEWAL. Draft For Comment.

• Division sively examined the planning, implementation, and management of Safety issue Resolution (Post 880717). August 1988. 65pp.

of NRC research programs in order to respond most effectively 8809140371.46802:053. to the Committee's recommendations. This report presents the The NRC is developing regulations for relicerising nuclear power plan;s. In recognition of the need to resolve the issues Commission's view of the Committee's report and describes the affecting public health and safety in a timely manner, the NRC actions that are under way in response ot its recommendations. issued a Federal Regist;:r notice in November 1986 requesting public comments on thc license renewal policy development NUREG-1328: USE OF PERIMETER ALARMS AT FUEL FABRI-effort. Fifteen topics of concern have been identified from the CATION FACILITIES USING OR POSSESSING FORMULA public's request. The topics have been categonzed as: tehnolo. QUANTITIES OF STRATEGIC SPECIAL NUCLEAR MATERIAL. gical, environmental, and procedural. The review and analysis of DWYliR,P.A.; GASKIN,C.E. Division of Safeguards & Transpor-these topics have resulted in the characterization of regulatory tation (Post 870413). December 1988. 22pp. 8812280105. issues and the identification of ways for dealing with certain 47911:273. issues. This report presents the rtatus of this effort and is being This guidance document presents information on installation, issued for pubhc commerd The comments would help focus on system objectives, maintenance, and tesung of penmeter intru-the issues that should be addressed in the proposed rule on li- sion detection systems at fuel fabrication facilities using or pos-cense renewal. setsing formula quantities of strategic nuclear material. NUREG-1318: TECHNICAL POSITION ON ITEMS AND ACTIVI- NUREG-1329: ENTRYtEXIT CONTROL AT FUEL FABRICATION TIES IN THE HIGH-LEVEL WASTE GEOLOGIC REPOSITORY FACILITIES USING OR POSSESSING FORMULA QUANTITIES PROGRAM SUBJECT TO QUALITY ASSURANCE REOUIRE- OF STRATEGIC SPECIAL NUCLEAR MATERIAL, DWYER,P.A. MENTS. DUNCAN,A.B.; BILHORN,S G.; KENNEDY,J E. Division Division of Safeguards & Transportation (Post 870413). Decem-of High Leml Waste Management (Post 870413). April 1988. 28pp. 8805200014. 45549:155. ber 1988.17pp. 8812280087. 47911:295. This document provides guidance on how to identify items This NUREG document presents information on entry / exit and activities subject to Quahty Assurance in the hig41evel nu- control at fuel fabncation facilities using or possessing formula clea* waste repository program for pre closure and post ciosure quantities of strateg!c special nuclear matenal. It describes NRC phases of the repository. In the pre closure phase, structures, requirements and methods for conducting personnel, package systems, and components essential to the prevention or mitiga- and vehicle searches at these facilities. An example of a testing tion of an accident that could result in an off-site radiation dose method for determining the detection capability for firearms, ex-of 0.5 rom or greater are termed "important to safety." in the plosives, and metal detectors is provided. post-closure phase, the bamers which are relied on to meet the

l Main Citations and Abstracts 17 NUREG-1330: PERSONNEL AND VEHICLE BARRIERS AT FUEL stainless steel pipe in combined axial tension and bending - was FABRICATION FACILITIES USING OR POSSESSING STRATE- addressed. The applied bending moments at crack growth initi-GlC OUANTITIES OF SPECIAL NUCLcAR MATERIAL. l ation and at fracture instability were sought. Seven estimation ' RENTSCHLER.RR Division of Safeguards & Transportation type solutions were performed along with a benchmark elastic-(Post 870413). December 1988. 28pp. 8812280076. 47911:245. plastic finite element solution. It was learned that precise speci-This document provides information on use and availability of fication of the material stress-strain curve must be made to barriers to deny unauthorized personnel and vehicle entry into obtain meaningful results. But, when applied under controlled fuel fabrication facilities using or possessing formula quantities conditions, the different estimation method solutions do provide of strategic special nuclear material. reasonably consistent results. These results appear to be con-NUREG 1332: REGULATORY ANALYSIS FOR THE RESOLU- servative in companson with an elastic-plastic finite element so-TION OF GENERIC ISSUE 125.11.7, " REEVALUATE PROVI. lution that was performed to provide a comparison with these SiON TO AUTOMATICALLY ISOLATE FEEDWATER FROM results. y STEAM GENERATOR DURING A LINE BREAK," BASDEKAS.D.L Division of Safety issue Resolution (Post NUREG/CP-0089: PROCEEDINGS OF THE CSNI SPECIAll0T 880717). September 1988. 27pp. 8810110235. 47044:259. MEET!NG ON TRAINING OF NUCLEAR REACTOR PERSONNEL. Held At Orlando, Florida, April 2124,1987.

• Divi-Generic issue 125.II.7 addresses the concern related to tho automatic isolation of Auxiliary Feedwater (AFW) to a steam sion of Licensee Performance & Ouality Evaluation (Post generator with a broken steam or feedwater line. This regulatory 870411). January 1988. 498pp. 8802030342. 44242:154.

analysis provides quantitative assessment of the costs and ben- This report provides the papers which were presented at the efits associated with the removal of the AFW autometic isola- CSNI Specialist Meeting on Training of Nuclear Reactor Person-tion, and concludes that no new regulatory requirements are nel in Orlando, Flonda from April 21-24, 1987. Topics covered ) warranted. include approaches to Training and regulatory practices in the f various CSNI countries, instructional methods, evaluation, diag- l NUREG-1333 DRFT FC: MAINTENANCE APPROACHES AND nostics and team training. l PRACTICES IN SELECTED FOREIGN NUCLEAR POWER ' PROGRAMS AND OTHER U.S. INDUSTRIES. Review And Les- NUREG/CP-0091 Vot: PROCEEDINGS OF THE FIFTEENTH sons Leamed Draft Report For Comment. DEY,M.K. Division of WATER REACTOR SAFETY INFORMATION MEETING. Regulatory Applications (Post 870413). November 1988.156pp. WEISS.A.J. Brookhaven National Laboratory. February 1988. 8812020311.47692:148. 478pp. 6803100008. 44658:210. The Commission recently published a Notice of Proposed This six-volume report contains 140 papers out of the 164 Rutemaking on Maintenance of Nuclear Power Plants spelling that were presented at the Fifteenth Water Reactor Safety infor-out NRC's expectations in maintenance. In prepanng the pro- mation Meeting held at the National Bureau of Standards. - I posed rule, the NRC reviewed maintenance practices in other Gaithersburg, Maryland, during the week of October 26-29, countries and considered maintenance approaches in other in- 1987. The papers are printed in the order of their presentation dustries in this country. As a (esult of the examination of the in each session and describe progress and results of programs benefits of vanous regulatory approaches, it is concluded that a in nuclear safety research conducted in this country and reguletory approach similar to that adopted by the Federal Avia- abroad. Foreign participation in the meeting included twenty-two tion Administration is most appropriate br NRC's proposed rule- different papers presented by researchers from Belgium, making. As a result of the review of maintenance practices, it is ] Czechoslovakia, Germany, Italy, Japan, Russia, Spain Sweden, j concluded that certain practices in the following areas have The Netherlands and the United Kingdom. The titles of the been found to contribute significantly to effective maintenance: papers and the names of the authors have been updated and (i) systems approach; (ii) maintenance program effectiveness' may differ from those that appeared in the final program of the monitoring; (iii) technician quahfications and motivation; and (iv) meeting, maintenance organization. NUREG/CP-0064: SECOND CNSI WORKSHOP ON DUCTliti NUREG/CP-0091 V02: PROCEEDINGS OF THE FIFTEENTH FRACTURE TEST METHODS. LOSS,F.J. Matenals Engineering WATER REACTOR SAFETY INFORMATION MEETING. Associates, Inc. August 1988. 327pp. 8809220043. MEA-2313. WEISS,A.J. Brookhaven National Laboratory. February 1988. 46879:010. 463pp. 8803090179. 44646:101. See NUREG/CP-0091,V01 abstract. This report is a compilation of papers presented at the Second CSNI Workshop on Ductile Fracture Test Methods, held at OECD Headquarters Pans, France, on April 17-19, 1985. NUREG/CP-0091 V03: PROCEEDINGS OF THE FIFTEENTH WATER REACTOR SAFETY INFORMATION MEETING. The contributors addressed advances in test methods to char- WEISS.A.J. Brookhaven National Laboratory. February 1988. acterize the fracture toughness of structural steels. Sessions 414pp. 8803100004. 44659:328. were held on new and improved test techniques, standardized See NUREG/CP-0091,V01 abstract. J-R curve test procedures, expenence and problems with exist-ing techniques, and use of fracture mechanics by the nuclear NUREG/CP-0091 V04: PROCEEDINGS OF THE FIFTEENTH industry. Summaries of the individual sessions have been pre- WATER REACTOR SAFETY INFORMATION MEETING. pared by the session chairmen. The meeting identified progress WEISS.A.J. Brookhaven National Laboratory. February 1988. in test methods since the first workshop was held in 1982. A 607pp. 8803280097. 44860:269. clear movement to standardize J-R curve tests is now apparent. See NUREG/CP-0091/M1 abstract. However, there exists a continuing need to improve elastic-plastic fracture test methods. NUREG/CP-0091 V04 ADD: PROCEEDINGS OF THE FIFTEENTH NUREG/CP-0075: PROCEEClNGS OF CSNI/NRC WORKSHOP WATER REACTOR SAFETY INFORMATION MEETING. ON DUCTILE PfPING WEISS.A.J. Brookhaven National Laboratory. September 1988. FRACTURE MECHANICS. 32pp. 8811010192. 47271:319. KANNINEN.M.F. Southwest Research Institute. May 1988. See NUREG/CP-0091,V01 abstract. 164pp. 8807110512. CSNI 97. 46082:318. This report contains the papers presented at a workshop NUREG/CP-0091 V05: PROCEEDINGS OF THE FIFTEENTH meeting that was conducted to compare the vanous different elastic-plastic fracture mechanics analysis methods that can be WATER REACTOR SAFTEY INFORMATION MEETING. WEISS A.J. Brookhaven National Laboratory. February 1988. applied to assess the margin of safety in cracked nuclear plant 470pp. 8803150379. 44690:108. pipes. A specific problem - a circumferentially cracked Type 304 See NUREG/CP-0091,V01 abstract.

18 Main Citations ancf Abstracts NUREG/CP-0091 V06: PROCEEDINGS OF THE FIFTEENTH correspond to each of the presentations make up the body of WATER REACTOR SAFETY INFORMATION MEETING. this report. The workshop was hosted by Sandia National Lab-WEISS.A.J. Brookhaven National Laboratory. February 1988. oratones under the sponsorship of the U.S. Nuclear Regulatory 259pp. 8803090298. 44654:151. Commission. Principal organizers for the workshop were Walter See NUREG/CP 0091,V01 abstract. von Riesemann and Toni D. Molna from Sandia National Lab-a s u a d James F. Costello of the U.S. Nu-NUREC/CP-0092: PROCEEDINGS OF THE SEMINAR ON LEAK. BEFORE BREAK. Progress in Regutatory Policies And Support-

                                                                                              ,             9     mb ing Research. KASHIMA K. Central Research institute of Elec-        NUREG/CP-0096: TRANSACTIONS OF THE SIXTEENTH inc Power Industry. WILKOWSKI G.M. Battelle Memorial Insti-          WATER REACTOR SAFETY INFORMATION MEETING.                            ,

tute, Columbus Laboratories March 1988. 539pp. 8804080282. WEISS,A.J. Office of Nuclear Regulatory Research, Director l 45065:050. (Post 860720). October 1988. 286pp. 8810280158. 47246:034. . The third in a senes of international Leak-Before Break (LBB) This report contains summaries of papers on reactor safety l Seminars supported in part by the U.S. Nuclear Regulatory research to be presented at the 16th Water Reactor Safety in- l Commission was held at TEPCO Hall in the Tokyo Electne formation Meeting at the National Bureau of Standards in Gaith-Power Company's (TEPCO) Electric Power Museum on May 14 ersburg, Maryland, October 24-27, 1988. The summaries bnefly and 15,1987. The semirar updated the intemational policies describe the programs and results of nuclear safety research and supporting research on LBB. Attendees included represent- sponsored by the Office of Nuclear Regulatory Research, stives from regulatory agencies, electric utility representatives. USNRC. Summaries of invited papers concerning nuclear safety fabricators of nuclear power plants, research organizations, and issues are included from the Offices of Nuclear Reactor Regula-university professors. Regulatory policy was the subject of pres- tion and Nuclear Matenal Safety and Safeguards, USNRC, in entations by Mr. G. Arlotto (U.S. NRC, U.S.A). Dr. H. Schultz addition to summaries of invited papers that cover the highlights ) (GRS, W. Germany), Dr. P. Milella (ENEA-DISP, Itafy), Dr. C. of reactor safety research conducted by the Department of Faidy, P. Jamet, and S. Bhandari, (EDF/Septen, CEA/CEN, and Energy (DOE), the electric utilities through the Electric Power Framatome, France), and Mr. T. Fukuzawa (MITI, Japan). Dr. F. Research Institute (EPRI), the nuclear industry, and the re-Nilsson presented revised nondestructive inspection require- search of government and industry in Europe and Japan. The ments relative to LBB in Sweden. In addition, several papers on summaries have been compiled in one report to provide a basis the supporting researcn programs discussed regulatory policy. for meaningful discussion and information exchange dur,ng the Questions following the presentations of the papers focused on course of the meeting. and are given in the order of their pres-the impact of vanous LBB policies or the impact of research entation in each session. findings. Supporting research programs were reviewed on the first and second day by several participants from the U.S., NUREG/CP-0099: PROCEEDINGS OF THE PUBLIC WORKSHOP Japan, Germany, Canada, Italy, Sweden, England, and France. FOR NRC RULEMAKING ON MAINTENANCE OF NUCLEAR l POWER PLANTS. TABATABAl,A.; SCOTT W. Battelle Memorial NUREG/CP-0093: PROCEEDINGS OF THE MEETING ON UL. Institute, Pacific Northwest Laboratory. DEY,M.K. Division of TRASENSITIVE TECHNIQUES FOR MEASURE,JENT OF URA-Regulatory Applications (Post 870413). November 1988.372pp. NIUM IN BIOLOGICAL SAMPLES AND THE NEPHROTOXI. CITY OF URANIUM. KATHREN,R.L.; WEBER,J.R. Battelle Me. 8812020308. PNL-SA 16308. 47692:304. monal institute, Pacific Northwest Laboratory. April 1988.172pp. This is the proceedings of the Public Workshop for NRC 8805260245. PNL 6511. 45632:120. Rulemaking on the Mruitenance of Nuclear Power Plants. It Edited transenpts are provided of two public meetings spon. was held at the Mayflower Hotel in Washington, D.C. on July sored by the Occupational Radiation Protection Branch of the 11 13, 1988. The purpose of the workshop was to solicit early Division of Radiation Programs and Earth Sciences, Nuclear public and industry participation in and comments on a rufemak-Regulatory Commission. The first meeting, held on December 3, ing process recently initiated by the United States Nuclear Reg-1985, included nine presentations covenng ultrasensitive tech. ulatory Commission (NRC) to improve the conduct and effec-niques for measurement of uranium in biological specimens. tiveness of maintenance practices in U.S. commercial nuclear Topics include laser-spectrometne techniques for uranium bio- power plants. The first day of the workshop was devoted to assay, correlation of urinary uranium samples with air sampling general discussions by NRC s'e'f describing the proposed rule-results in industnal settings, delayed neutron counting, laser ki- making alternatives and presentations by industry representa-notic phosphometry, isotope diluten mass spectrometry, reso. tives regarding industry's position on the maintenance rufemak-nance ionization spectroscopy, fission track analysis, laser in. ing initiative. On the second day, the workshop was divided into duced fluorescence, and costs of sampling and processing. The four working groups to focus discussion on specific aspects of nine presentations of the second meeting, on December 4, maintenance and the proposed rule. Results of the working 1985, deelt with the nephrotoxicity of uranium. Among the group discussions were presented to all workshop participants topics were the physiology of the kidney, the effects of heavy on the third day. The third day also provided an open forum for metals on the kidney, animal studies in uranium nephrotoxicity, questions, answers, discussion and comments on the rulemak. compansons of kidney histology in nine humans, renal effects in ing initiative. uranium mill workers, renal damage from different uranium iso- NUREG/CR-0130 ADD 04: TECHNOLOGY, SAFETY AND COSTS topes, and Canadian studies on uranium toxicity. Discussions OF DECOMMISSIONING A REFERENCE PRESSURIZED following the presentations are included in the edited tran. WATER REACTOR POWER STATION. Technical Support For senpts. Decommissioning Matters Related To Preparation Of The Final NUREG/CP-0095: PROCEEDINGS OF THE FOURTH WORK- Decommissioning Rule. KONZEK,G.J.; SMITH R.I. Battelle Me-SHOP ON CONTAINMENT INTEGRITY. COCHRELL.R. Sandia monal Institute, Pacific Northwest Laboratory. July 1988.101pp. National Laboratones. November 1988. 639pp. 8812280114. 8808120069.46455:155. SANDB81836. 47912:001. Preparation of the final Decommissioning Rule by the Nuclear The Fourth Workshop on Containment intognty was held in Regulatory Commission (NRC) staff has been assisted by Pacif-Arlington, Virginia, on June 14-17,1988. The workshop pro?ded ic Northwest Laboratory (PNL) staff familiar with decommission-a forum for exchanging information on the integnty of contain- ing matters. These efforts have included updating previous cost monts at nuclear power plants. The behavior of containments estimates developed during the senes of studies of conceptually dunng severe accidents was of pnmary interest to the partici- decommissioning reference licensed nuclear facilities for inclu-pants. There were 145 participants at the workshop including 56 sion in the Final Generic Environmental Impact Statement foreign attendees from 13 countnes. Written contnbutions that (FGEIS) on Decommissioning; documenting the cost updates;

Main Citations and Abstracts 19 evaluating the cost and dose impacts of post TMI-2 backfits on using correlation tables from the Sequence Coding and Search decommissioning; performing revised scaling factor analyses System. concoming reactor plants different in size from the reference PWR desenbed in the earlier studies; determining the formula NUREG/CR-2000 V07 N1: LICENSEE EVENT REPORT (LER) for adjusting current cost estimates to reflect escalation in labor, COMPILATION:For Month Of January 1988.

• Oak Ridge Na-matenals, and waste disposal costs; and completing a study of tional Laboratory. February 1988.167pp. 8803090261. ORNL/

recent PWR steam generator replacements to determine realis- NSIC-200. 44649:139 See NUREG/CR-2500,V06,N12 abstract. tic estimates for time, cost, and radiation doses associated with steam generator removat during decommissioning. This report NUREG/CR-2000 V07 N2: LICENSEE EVENT REPORT (LER) presents supporting information in four of the aforementioned COMPILATION;For Month Of February 1988.

• Oak Ridge Na-areas concerning decommissioning the reference PWR: 1) up* tional Laboratory. March 1988.152pp. 8804080232. ORNL/

dating the previous cost estimates to January 1986 dollars,2) NSIC-200. 45041:079. assessing the cost and dose impacts of post-TMI-2 backfits,3) See NUREG/CR-2000,V06,N12 abstract. developing scaling and escalation formulae, and 4) assessing the impact of recent steam generator replacements. NUREG/CR-2000 V07 N3: LICENSEE EVENT REPORT (LER) COMPILATION.For Month Of March 1988.

• Oak Ridge Nation-  ;

NUREG/CR 0672 ADD 03: TECHNOLOGY, SAFETY AND COSTS at Laboratory. April 1988.163pp. 8805090093. ORNL/NSIC- ' OF DECOMMISSIONING A REFERENCE BOILING WATER 200.45412:001. REACTOR POWER STATION.Technictd Support For Decom- See NUREG/CR-2000,V06,N12 abstract. missioning Matters Related To Preparation Of The Final Decom- . missioning Rule. KONZEK,G.J.; SMITH,R.I. Battelle Memorial In- NUREG/CR-2000 V07 N4: LICENSEE EVENT REPORT (LER) l stitute, Pacific Northwest Laboratory. July 1988. 67pp. COMPILATION.For Month Of April 1988.

• Oak Rdge National I 8808120112.46453:262. Laboratory. May 1988.133pp. 8806020043. ORNL/NSIC-200.

Preparation of the final Decommissioning Rule by the Nuclear 45706:089. 1 Regulatory Commission (NRC) staff has been assisted by Pacif. See NUREG/CR-2000,V06,N12 abstract. ic Northwest Laboratory (PNL) staff familiar with decommission-ing matters. These efforts have included updating previous cost NUREG/CR-2000 V07 NS: LICENSEE EVENT REPORT (LER) J COMPILATION:For Month Of May 1988.

• Oak Ridge National i estimates developed dunng the series of studies of conceptually Laboratory. June 1988.150pp. 8807110514. ORNL/NSIC-200.

decommissioning reference licensed nuclear facilities for inclu- 46083:169. sion in the Final Generic Environmental impact Statement See NUREG/CR-2000,V06,N12 abstract. (FGEIS) on Decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on NUREG/CR-2000 V07 N6: LICENSEE EVENT REPORT (LER) decommissioning; performing revised scaling factor analyses COMPILATION:For Month Of June 1988.

• Oak Ridge National conceming reactor plants different in size from the reference Laboratory. July 1988.128pp. 8808080062. ORNL/NSIC-200.

BWR described in the earlier studies; and determining the for- 46408:103. mula for adjusting current cost estimates to reflect escalation in See NUREG/CR-2000,V06,N12 abstract. supp rtn int mat o i t ee of he afore n io d are s NUREG/CR-2000 V07 N7: LICENSEE EVENT REPORT (LER) conceming decommissioning the reference BWR: 1) updating COMPILATION:For Month Of July 1988.

• Oak Ridge National the previous cost estimates to January 1986 dollars,2) assess- Laboratory. August 1988. 87pp. 8809150272. ORNL/NSIC-200.

eng the cost and dose impacts of post-TMI-2 backfits, and 3) 46802:273. developing scaling and escalation formulae. See NUREG/CR-2000,V06,N12 abstract. NUREG/CR-2000 V06N12: LICENSEE EVENT REPORT (LER) NUREG/CR-2000 V07 N8: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of December 1987

• Oak Ridge Na- COMPILATION:For Month Of August 1988. Oak Ridge Nation- j tional Laboratory. January 1988.138pp. 8802030142. ORNL/ al Laboratory. September 1988.151pp. 8811010186. ORNL/

NSIC-200. 47275:292. - Th s mont y repo t contains Licensee Event Report (LER) See NUREG/CR-2000,V06,N12 abstract. operational information that was processed into the LER data file of the Nuclear Safety Information Center (NSIC) during the NUREG/CR 2000 V07 N9: LICENSEE EVENT REPORT (LER) ) COMPI..ATION-For Month Of September 1988.

• Oak Ridge '

one month period identified on the cover of the document. The Nationa Laboratory. October 1988. 113pp. 8811070099. LERs, from which this information is denved, are submitted to ORNL/NSIC-200. 47390:032. the Nuclear Regulatory Commission (NRC) by nuclear power See NUREG/CR-2000,V06,N12 abstract. plant heensees in accordance with federal regulations. Proce-dures for LER reporting for revisions to those events occurring NUREG/CR-2000 V07N10: LICENSEE EVENT REPORT (LER) prior to 1984 are desenbed an NRC Regulatory Guide 1.16 and COMPILATION:For Month Of October 1988.

• Oak Ridge Na-NUREG 1061, " Instructions for Preparation of Data Entry tional Laboratory. November 1988. 92pp. 8812190133. ORNL/

Sheets for Licensee Event Reports." For those events occurring NSIC-200. 47806:016. on and after January 1,19fi4, LERs are being submitted in ac. See NUREG/CR-2000,V06.N12 abstract. I cordance with the revised rule contained in Titio 10 Part 50.73 l of the Code of Federal Regulations (10 CFR 50.73 - Licensee NUREG/CR-2000 V07N11: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of November 1988.

• Oak Ridge Na-Event Report System) which was published in the Federal Reg-ister (Vol. 48, No.144) on July 26,1983. NUREG 1022, "Li- tional Laboratory. December 1988. 97pp. 8901090323. ORNL/

censee Event Report System - Description of Systems and NSIC-200. 48110:167. See NUREG/CR-2000,V06,N12 abstract. Guidelines for Reporting." provides supporting guidance and in-formation on the revised LER rule. The LER summanes in this NUREG/CR-2331 V7N2-3: SAFETY RESEARCH PROGRAMS report are arranged alphabetically by facility name and then SPONSORED BY OFFICE OF NUCLEAR REGULATORY chronologically by event date for each facility. Component, RESEARCH. Progress Report, April-September 1987. WEISS.A.J. system, keyword, and component vendor indexes follow the Brookhaven National Laboratory. June 1988. 270pp. j summanes. Vendors are those identified by the utihty when the 8810040125. BNL-NUREG 51454. 46975:260. LER form is initiated; the keywords for the component, system, The Advance and Water Reactor Safety Research Programs and general keyword indexes are assigned by the computer Quarterly Progress Reports have been combined and are in-

20 Main Citations and Abstracts cluded in this report entitled, " Safety Research Programs Spon- erage dose commitments from the airborne pathways. The total sored by the Office of Nuclear Regulatory Research - Progress dose commitments (from both liquid and airbome pathways) for Report." This progress report will describe current activities and each wie ranged from a high of 110 person-rem to a low of technical progress in the programs at Brookhaven National Lab- 0.002 person-rem for the sites with plants operating throughout oratory sponsored by the Division of Regulatory Applications, the year with an arithmetic mean of 5 person-rem. The total Division of Engineering, Division of Reactor Accident Analysis, population dose for all sites was estimated at 280 person-rem and Division of Reactor and Plant Systems of the U.S. Nuclear for the 100 million people considered at risk. The site average Regulatory Commission, Office of Nuclear Regulatory Research individual dose commitment from all pathways ranged from a following the reorganization in February 1987. low of 6 x 10(6) mrem to a high of 0.04 mrom. No attempt was NUREG/CR-2331 V8N1-2: SAFETY RESEARCH PROGRAMS made in this study to determine the maximum dose commitment SPONSORED BY OFFICE OF NUCLEAR REGULATORY received by any one individual from the radionuclides released RESEARCH. Progress Report, January-June 1988. WEISS,A.J. at any of the sites. Brookhaven National Laboratory. December 1988. 279pp. 8901090383. BNL-NUREG-51454. 48109:216. NUREG/CR-2850 V07: POPULATION DOSE COMMITMENTS This progress report describes current activities and technical DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER progress in the programs at Brookhaven National Laboratory PLANT SITES IN 1985. BAKER,D.A. Battelle Memorial Institute, j sponsored by the Division of Regulatory Applications, Division Pacific Northwest Laboratory. August 1988. 154pp. of Engineering, Division of Safety issue Resolution, and Division 8809220023. PNL-4221. 46883:330. of Systems Research of the U.S. Nuclear Regulatory Commis- Population radiation dose commitments have been estimated sion, Office cf Nuclea- Regulatory Research following the reor. from reporteo radionuclides releases from commercial power re-ganization in July 1988. The previous reports have covered the actors operating dunng 1985. Fifty-year dose commitments from penod October 1,1976 through December 31,1987. a one-year exposure were calculated from both liquid and at-NUREG/CR-2336: STEAM GENERATOR TUBE INTEGRITY mospheric releases for four population groups (infant, child, PROGRAM. Phase ll Final Report. KURTZ,R.J.; BICKFORD.R.L.; teen-ager and adult) residing between 2 and 80 km from each CLARK,R.A.; et al. Battelle Memorial Institute, Pacific Northwest of 61 sites. This report tabulates the results of these calcula-Laboratory. August 1988. 94pp. 8808230421. PNL-4008. tions, showing the dose commitments for both liquid and air-46575:051, borne pathways for each age group and organ. Also included j The SGTIP was a three phase program conducted for the for each of the sites is a histogram showing the fraction of the NRC by PNL. The first phase involved burst and collapse test- total population within 2 to 80 km around each site receiving ing of typical steam generator tubing with machined def9 cts. vanous average dose commitments from the airbome pathways. The second phase of the SGTIP continued the integnty Ostin9 The total dose commitments (from both liquid and airborne work of Phase 1, but tube specimens were degraded by chemi- pathways) for each site ranged from a high of 73 person-rem to cal means rather than machining methods. The third phase of a low of 0.011 person-rem for the sites with plants operating the program used a removed-from-service steam generato" as a throughout the year with an arithmetic mean of 3 person-rem. test bed for investigating the reliability and effectiveness of in- The tota! population dose for all sites was estimated at 200 service nondestructive eddy-current inspection methods anu as erson-rem for the 110 million people considered at risk. The a source of service degraded tubes for validating the Phase I and Phase 11 data on tube integrity. This report desenbes the site average individual dose commitment from all pathways results of Phase ll of the SGTIP. The object of this effort includ- ranged from a low of 5 x 10(-6) mrom to a high of 0.02 mrem. ed burst and collapse testing of chemically defected PWR No attempt was made in this study to determine the maximum steam generator tubing to validate empirical equations of re- dose commitment received by any one individual from the ra-maining tube integrity developed during Phase 1. Three types of dionuclides released at any of the sites. defect geometries were investigated: stress corrosion cracking l (SCC), uniforrn thinning and ethptical wastage. In addition, a NUREG/CR 2907 V06: RADIOACTIVE MATERIALS RELEASED l review of the publicly available leak rate data for steam genera- FROM NUCLEAR POWER PLANTS. Annual Report For 1985. tor tubes with axial and circumferential SCC and a comparison TICHLER,J.; NORDEN.K.; CONGEMI,J. Brookhaven National with an analytical leak rate modet is presented. Lastly, nonde. Laboratory. January 1988. 286pp. 8802180012. BNL-NUREG-structive eddy current (EC) measurements of defect severity are 51581. 44364:260. reported. Laboratory EC measurements to determine accuracy Releases of radioactive materials in airbome and liquid ef-of defect depth sizing using conventional and afternate stand- fluents from commercial light water reactors during 1985 have ards is described. To supplement the laboratory EC data and been compiled and repor:ed. Data on solid waste shipments as obtain an estimate of EC capability to detect and size SCC, a well as selected operating information have been included. This mini-round robin utilizing several fi rms that routinely perform in- report supplements earlier annual reports issued by the former service inspections was conducted. Atomic Energy Commission and the Nuclear Regulatory Com-NUREG/CR-2850 V06: POPULATION DOSE COMMITMENTS mission. The 1985 release data are summarized in tabular forn.. DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER Data covenng specific radionuclides are summarized. ic Northwest borato au 198 1 NUREG/CR-2907 V07: RADIOACTIVE MATERIALS RELEASED 8802030210. PNL-4221. 44241:212. FROM NUCLEAR POWER PLANTS. Annual Report For 1986. Population radiation dose commitments have been estimated TICHLER J.; NORDEN,K.; CONGEMIJ. Brookhaven National from reported radionuclides releases from commercial power ro. Laboratory. November 1988.305pp.8812010047. BNL-NUREG-actors operating dunng 1984. Fifty-year dose commitments from 51581. 47684:034. e one-year exposure were calculated from both liquid and at. Releases of radioactive materials in airborne and liquid ef-mospheric releases for four population groups (infant, child, fluents from commercial light water reactors during 1986 have teen-ager and adult) residing between 2 to 80 km from each of been compiled and reported. Data on solid waste shipments as 56 sites. This report tabulates the results of these Calculations, well as selected operating information have been included. This showing the dose commitments for both liquid and airbome repon supplements earlier annual reports issued by the former pathways for each age group and organ. Also included for each Atomic Energy Commission and the Nuclear Regulatory Com-of the sites is a histogram showing the fraction of the total pop- mission. The 1986 release data are summanzed in tabular form. ulation witnin 2 to 80 km around each site receiving various av- Data covering specific radionuclides are summanzed.

l I Main Citations and Abstracts 21 NUREG/CR-3145 V06: GEOPHYSICAL INVESTIGATIONS OF of funds for decommissioning. The report also concludes that THE WESTERN OHlO-INDIANA REGION. Annual the NRC should recommend changes in bankruptcy laws, in-Report, October 1986 - September 1987. SCHWARTZ,S.Y ; cluding dec3mmissioning obligations in utility prospectuses, and CHRISTENSEN.D.; LAY,T ; et af. Michigan, Univ. of, Ann Arbor, conduct periodic financial reviews of nuclear utilities due to MI. February 1988. 63pp. 8808080111. 46409:040. changing economic conditions. Earthquake activity in the Western Ohio - Indiana region has been monitored with a precision seismograph network consist- NUREG/CR 3908: SURVEY OF THE STATE OF THE ART IN ing of nine stations located in west-central Ohio and four sta- MITIGATION SYSTEMS CASTLE,J.N - CATTON,1 - tions located in Indiana. One local and eleven near regional DOOLEY,J.L.; et al. R&D Associates. Janudry 1988. 129pp. earthquakes have been recorded during this report penod. The 88 2 8 4 local event had a duration magnitude of 0.7 and was not felt. Its location is close to the larger (m(b) = 4.5) July 12,1986 St. tract in the area of LWR severe accident mitigation. This report Marys Ohio earthquake. Many of the regional events were felt summarizes the current state of the art of mitigation devices with magnitudes ranging from m(bLg) = 2.7-4.9. The two larg- and systems that are potentially applicable to LWRs. The de-est of these events (March 27,1987 in northeaster Tennes-see, m(b)Lg = 4.2, ar,d June 10,1987 in southeastern Illinois, provide such as core retention, hydrogen control, vent systems m(b)Lg = 4.9) produced minor damres. P(n) travel time residu- p als, computed for all well-recorded egional events since de- eM ployment of the Anna Seismic Network, display a strong azi-muthat dependence with positive residuals (slow arrival times) NUREG/CR-3950 V04: FUEL PERFORMANCE ANNUAL REPORT FOR 1986. BAILEY,W.J. Battelle Memorial institute, obtained from ovents with northeasterly through southerly back Pacific Northwest Laboratory. WU.S. NRC - No Detailed Affili-azimuths and negative residuals (fast errival times) from events ation Given. March 1988. 183pp. 8804080131. PNL 5210. with westerly back azimuths. This pattern has larger residuals. 45034:199. but is similar to that displayed in the teleseismic P wave residu- This annual report, the ninth in a series, provides a brief de-als, supporting an interpretation that shallow structural hetero- scription of fuel performance during 1986 in commercial nuclear geneity is responsible for the P(n) residual pattern. power plants. Onef summaries of fuel design changes, fuel sur-NUREG/CR-3444 V05: THE IMPACT OF LWR DECONTAMINA- veillance programs, fuel operating experience and trends, fuel TIONS ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCI. problems, high-bumup fuel experien,;e, and items of general sig-ATED OCCUPATIONAL EXPOSURE. Annual Report, FY 1987. nificance are provided. References to additional, more detailed . ADAMS J.W.; SOO,P. Brookhaven National Laboratory, June information and related NRC evaluations are included. 1988.70pp.8901030121. BNL-NUREG-51699. 47961:349. A study was carned out to determine the extent of corros. ion NUREG/CR-4000 V02: THE MESORAD DOSE ASSESSMENT MODEL. Computer Code. RAMSDELL,J.V.; ATHEY,G.F.; for a range of low-level waste container materials in contact BANDER.T.J.; et al. Battelle Memorial Institute, Pacific North-with simulated resin bead decontamination wastes. The maten-als included metals and high density polyethylene. Tests were west Laboratory. October 1988. 387pp. 8811070057. PNL-5219. 47389:005. also conducted on mechanical stability of the decontamination resin wastes, solidified in cement and vinyl ester-styrene. dunng MESORAD is a dose assessment model for emergency re-thermal cycling between 60 degrees C and -40 degrees u sponse applications that is designed to be run on minicom-puters. It has been developed by the Pacific Northwest Labora-NUREG/CR 3509: POWER SPECTRAL DENSITY FUNCTIONS tory for use as part of the intermediate Dose Assessment CC"MPATIBLE WITH NRC REGULATORY GUIDE 1.60 RE. System in the U.S. Nuclear Regulatory Commission Operations SPONSE SPECTRA. SHINOZUKA M.; MOCHIO,T.; Center in Washington, D.C., and the Emergency Management SAMARAS.E.F. Columbia Univ., New York, NY. June 1988. System in the U.S. Department of Energy Unified Dose Assess. 71pp. 8807080360. 46077:033. ment Center in Richland, Washington, This volume describes Among other possibilities, the Kanai-Tajimi power spectral the MESORAD computer code and conte. ins a listing of the density enhanced at the higher frequency range is found to be code. The technical basis for MESORAD is described in the first useful for generating ground acceleration time histories that sat. Volume of this report (Scherpelz et al 1986). A third volume of  ; isfy NRC RG 1.60 requirements. The values of the parameters the documentation is planned. That volume will contain utility { involved in the spectral density function are recommended for programs and input and output files that can be used to check this purpose. Also, a suggestion is made as to in what way the the implementation of MESORAD. power spectral density requirements can be placed in combina-tion with those of the current NRC RG 1.60 to ensure both the NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM - PHASE f response and power spectrum requirements. ll.S th Program Report. October 1986 - September 1987. i WILKOWSKI,G.M.; AHMAD J.; BARNES,C.R.; et at Battelle Me- I NUREG/CR-3899 S01: UTILITY FINANCIAL STABILITY AND morial Institute, Columbus Laboratories. April 1988. 333pp. THE AVAILABILITY OF FUNDS FOR DECONIMISSIONING.An 8805060305. BMI-2120. 45431:012. Analysis Of Intemal And External Funding. SIEGEL J.J. Engi- Presented herein is an Annual Report of the U.S. NRC's De-neenng & Economics Research, Inc. June 1988. 3Bpp. graded Piping Program - Phase 11. This is the sixth program 8807110543.46078:173. report on this program. Pnor reports were semiannual reports. The NRC is currently developing final rules in the area of de- The intent of this program is to experimentally validate and en-commissioning nuclear facilities. A part of that rulemaking effort hance available analytical methods for evaluating the mechani-is assuring that funds will be available at the time of decommis- Cal behavior of nuclear power plant piping containing circumfer. sioning of power reactors. Previous NRC reports, including entially oriented defects. Fifty-seven pipe experiments have NUREG/CR-3899, published September,1984, have examined been conducted to date. These and approximately fifty addition-this issue by studying various funding methods. This report pro- al pipe expenments from other programs have been analyzed, vides an update by considering public comments received on in the analytical effort, a screening critorion has been devel-the NRC's proposed rule on decommissioning (published Febru- oped to show when the net-section-collapse analysis is valid. ary,1985) and by analyzing the relative level of assurance of Th,s shows that even tough materials such as stainless steel internal and external reserves. In its analysis, the report makes can fail at less than net-section-collapse loads if the pipe diam-use of specific case utility financial situations. The report con- eter is sufficiently large. Numerous predictive J-estimation cludes that from a financial standpoint, with the exception of schemes have been evaluated and modified. A finite length sur-PSNH, intemal reserves currently provide sufficient assurance face-cracked pipe estenation scheme has also been developed

                                                                                                                                                ]

t i I

22 Main Citations and Abstracts and incorporated into a computer code called NRCPIPE. This to resist severe core-melt accidents to limit the release of radio-code provides a convenient way of analyzing cracked pipe with active materials, incentives, goals, costs and sources of funding a number of currently accepted analytical methods. Supporting are discussed, and a series of possible implementation steps research efforts involve investigating geometry effects on J-R are presented. curves, as well as characterizing the matenal properties for NUREG/CR-4312 V01: RELAPS/ MOD 2 CODE MANUAL. Volume each pipe tested. The significance of all of the efforts to date 1: Code Structure, Systems Models And Solution Methods. relat ve ppe fracture analyses and flaw assessment enteria RANSOM,V.H.; WAGNER,R.J.; TRAPP,J.A.; et al. EG&G Idaho,

                   ~

Irs. (subs. of EG&G, Inc.). August 1985. 377pp. 8808120075. NUREG/CR-4219 V04 N2: HEAVY-SECTION STEEL TECHNOL- EGG-2396. 46457;135. OGY PROGRAM. Semiannual Progress Report For April-Sep- The principal objective of the RELAPS project is to provide tember 1987, CORWIN W.R. Oak Ridge Nakonal Laboratory. the United States Nuclear Regulatory Commission (USNRC) April 1988. 336pp. 8806020048. ORNL/TM-9593. 45705:113. with a fast running and user convenient light water reactor The Heavy-Section Steel Technology (HSST) Program is an system transient analysis code for use in rule making, licensing engineenng research activity conducted by the Oak Ridge Na- audit calculations, evaluation of operator guidelines, and as a tional Laboratory for the Nuclear Regulatory Commission. The basis for a nuclear plant analyzer. The RELAPS/ MOD 2 code Program comprises studies related to all areas of the technolo- has been developed for best estimate transient simulation of gy of matenals fabncated into thick section primary-coolant con- pressurized water reactors and associated systems. The code tainment systems of light-water-cooled nuclear power reactors. modeling capabilities are simulation of large and small break The investigation focuses on the behavior and structural integri- loss-of-coolant accidents, as well as operational transients such ty of steel pressure vessels continuing cracklike fiaws. Current as anticipated transient without SCRAM, loss-of- offsite power, j work is organized into twelve tasks: (1) program management, loss of feedwater, and loss of flow. A generic modeling ap- j (2) fracture-methodology and analysis, (3) matenal characterize- proach utilizes as much of a particular system to be modeled as tion and properties, (4) environmentally assisted crack-growth necessary. RELAPS/ MOD 2 extends the modehng base and ca-studies, (5) crack-arrest technology, (d) irradiation effects stud- pabilities offered by previous versions of the code. In particular, ses, (7) cladding evaluations, (8) intermediate vessel tests and MOD 2 contains two energy equations, reflood heat transfer, a analysis, (9) thermal-shock technology, (10) pressurized ther. two-step numerics option, a gap conductance model, reWsed mal-shock technology, (11) Pressure Vessel Research Users' constitutive models, and additional component and control Facility, and (12) shipping-cask material evaluations. system models. NUREG/CR-4219 VOS N1: HEAVY-SECTION STEEL TECHNOL- NUREG/CR-4312 V02 R1: RELAPS/ MOD 2 CODE OGY PROGRAM. Semiannual Progress Report For October MANUAL. Volume 2: Users Guide And input Requirements. 1987 - March 1988. CORWIN,W.R. Oak Ridge National Labora- RANSOM,V.H.; WAGNER,R.J.; TRAPP.J.A.; et al. EG&G Idaho, tory. August 1988. 283pp. 8810110288. ORNL/TM-9593. Inc. (subs. of EG&G, Inc.). March 1987, 444pp. 8808120082. 47042:078. EGG-2396. 46456:051, See NUREG/CR-4219,V04,N02 ebstract- The RELAP5/ MOD 2 code has been developed for best esti-mate transient simulation of pressurized water reactors and as-NUREG/CR-4242: SURVEY OF LIGHT WATER REACTOR CON. , TAINMENT SYSTEMS, DOMINANT FAILURE MODES AND sociated systems. The code modeling capability includes simu- ) MITIGATION OPPORTUNITIES. Final Report. CASTLE J.N.; lation of large and small break loss-of-coolant accidents as well CATTON,l.; DOOLEY,J.L.; et al. R&D Associates. January 1988. as operational transients such as anticipated transient without 290pp. 8802040315. RDATR127301-002. 44257:001. SCRAM, loss-of- offsite power, loss of feedwater, and loss of This is one of five major reports prepared under this contract flow. A genenc modeling approach is utilized, which permits as in the area of LWR severe accident mitigation. This report dis. much of a particular system to be modeled as necessary. Con- l cusses the effect of severe accidents on the five major reactor trol system pnd secondary system components are included to I containment structures found in the United States, including the permit modeling of plant controls, turbines, condensers, and dominant modes in which each containment type is expected to secondary feedwater conditioning systems. fait, the degree of public risk due to such failures and the feasi- NUREG/CR-4315 V09 R1: EVALUATION OF NUCLEAR FACILl-bility of modifying the containment systems to prevent failures TY DECOMMISSIONING PROJECTS. Summary Status or mitigate the consequences of failures. Report Three Mile Island Unit 2, Radioactive Waste And Laundry NUREG/CR-4243: VALUE/ IMPACT ANALYSIS FOR EVALUAT. Shipments. DOERGE,D.H.; HAFFNER,D.R. Westinghouse Han-ING ALTERNATIVE MITIGATION SYSTEMS. ford Co. June 1988. 48pp. 8807110504. 46085:120. KASTENBERG W.E.; CATTON,l.; CASTLE,J.N.; et al R&D As. This document summarizes information concerning radioac-sociates. January 1988. 44rp. 3802050017. 44267:120. tive waste and laundry shipments from the Three Mile Island This is one of five major reports prepared under this contract Nuclear Station Unit 2 to radioactive waste disposal sites and to in the area of LWRs severe accident mitigation. This report de- protective clothing decontamination facilities Oaundries) since scribes methods for use in evaluating the cost effectiveness of the loss of Coolant accident experienced on March 28, 1979. systems and strategies proposed for mitigating the effects of Data were collected from radioactive shipment records, sunima-severe accidents. By estimating the reduction in overall popula- nzed, and placed in a computerized information retrieval / manip-tion dose provided by the implementation of a specific mitiga. ulation system which permits extraction of specific information tion scheme and using this as the value or benefit with the cost as required. Information contained in this report includes: waste being the estimated dollar cosd of the improvement, cost-benefit disposal site locations, dose rates, cune content, waste descrip-values are calculated. Other cc,nsiderations such as post-acci- tion, container type and number, volumes and weights. dent recovery costs are addressed also. NUREG/CR-4508: BEHAVIOR OF A CORIUM JET IN HIGH NUREG/CR.4244: STRATEGIES FOR IMPLEMENTING A MtTI- PRESSURE MELT EJECTION FROM A REACTOR PRESSURE GATION POLICY FOR UGHT WATER REACTORS. VESSEL. FRID W. Sandia National Laboratones. April 1988. KASTENBERG,W.E.; HAMMOND,R.P.; CATTON.I.; et al. R&D 180pp. 8808080113. SAND 851726. 46409.108. Associates. January 1988. 57pp. 8802050068. 44267:064. This report provides results from analytical and expenmental This is one of five reports prepared under this contract in the investigations on the behavior of gas supersaturated molten jet area of LWR severe accident mitigation. In this report, possible expelled from a pressunzed vessel. Models are developed for strategies are discussed for implementing a regulatory policy re- jet expansion, pnmary breakup of the jet, and secondary frag-quiring that power reactor containment enclosures be modified mentation of melt droplets. The jet expansion model is based

Main Citations and Abstracts 23 on a general relation for bubble grewth which includes both in- The Zion Risk Rebaselining Study (ZRRS) is conducted with ertial-controlled and diffusion-controlled growth phase. The two broad objectives: (1) to provide a severe accident risk reba-model is able to predict the jet void traction, jet radius as a selining perspective for the Zion nuclear power plant based function of axial distance from the pressure vessel bubble size, upon the methodology established as a part of the Severe Acci-and bubble pressure. Predicted fragment sizes are in reason- dent Risk Reduction Program (SARRP) at the Sandia National ably good agreement with the data. Parametric calculations for Laboratory (SNL), and (2) to examine the potential for reduction a TMLB' accident show that the corium jet is disrupted within a in risk within the framework of cost / benefit analysis. These ob-few initial jet diameters from the vessel. The radius of the jectives have been achieved and several important results and corium jet at the level of the reactor cavity floor is predicted to insights gained are summarized in this report. be in the range 0.8m - 2.6m. l NUREG/CR-4555 Rot: GENERIC COST ESTIMATES FOR THE NUREG/CR-4573: CLOSEOUT OF IE BULLETIN 80-13: CRACK

• DISPOSAL OF RADIOACTIVE WASTES. CLARK,R.;

ING IN CORE SPRAY SPARGERS. FOLEY,W.J.; DEAN,R.S.; KNUDSON R.; SCIACCA.F.; et al. Science & Engineering Asso-HENNICK.A. Parameter, Inc. January 1988.37pp.8802030177. ciates, Inc. September 1988.171pp. 8810030200. SEA 87-288-PARAMETER IE153. 44241:174. 04A1. 46974:332. Between late 1978 and early 1980, the licensees of Oyster NRC regulatory impact analyses address the costs and bene-Creek and Pilgrim nuclear power stations notified the NRC that fits associated with proposed regulatory requirements. Many of , cracks had been found in core spray spargers. In early 1979, these requirements will result in physical modifications to exist- l General Electne (GE) requested licensees of boiling water reac- ng structures and systems at nuclear power plants. This report tors (BWRs) to inspect spargers for visual indications of crack- provides a methodology and data needed to estimate the ge-ing, in March 1980, representatives of GE and the NRC met to neric costs of disposing of radioactive wastes that may be gen-discuss sparger cracking. IE Bulletin 80-13 was issued May 12- erated as a result of NRC regulations requiring modifications or 1980, to require more intensive inspection of these safety relat- repairs to nuclear facilities. Also presented are descriptions of ed systems. Core spray spargers are provided as engineered typical low-level radioactive wastes generated at nuclear power , safety features, for emergency core cooling. Licensees of oper. plants and the various processes used to treat the wastes in 1 ating BWRs were required to take four specific actions. Evalua- preparation for shipment and burial. The wcste disposal cost es- ' tion of licensees' responses and inservice inspection reports, timates included in this report cover pil the major elements that NRC/IE inspection reports and NRC correspondence shows contribute to the overall costs. The key factors that influence that the bulletin can be closcd out for all of the 23 BWR operat- the costs are discussed. Pertinent ranges of values for the key ing facilities which were issued the bulletin for action. Examina- variables are explored and important sensitivities identified. The tion of spargers at 22 operating BWRs is required every refuel- cost implications of the burial surcharges authorized by the ing outage. The licensees have incorporated this exaraination Low-Level Radioactive Waste Policy Amendments Act of 1985 l into their Inservice inspection programs. Techniques for inspec- are covered. Occupational radiation exposure associated with tion of spargers have been improved during the period of bulle- in-plant handling of the wastes is also discussed. This report tin activity. Generic Letter 84-11 establishes the requirement for updates and revises information presented in NUREGvCR. an ongoing program for inspection of BWR stainless steel 4555. piping. NUREG/CR-4527 V02: AN EXPERIMENTAL INVESTIGATION OF NUREG/CR-4597 V02: AGING AND SERVICE WEAR OF AUXIL-INTERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT lARY FEEDWATER PUMPS FOR PWR NUCLEAR CONTROL CABINETS.Part II: Room Effects Tests. PLANTS. Volume 2. Aging A,ssessments And Monitonng Method CHAVEZ,J.M.; NOWLEN S.P. Sandia National Laboratones. No. Evaluations. KITCH.D.M.; SCHLONSKI,J.S.; SOWATSKEY,P.J.; vomber 1988. 66pp. 8812190075. SAND 86-0336. 47804:249. et al. Oak Ridge National Laboratory. June 1988. 136pp. This report presents the findings of the secOnd part of a two. 8808120131. ORNL-6262. 46458:152. part serier, of full-scale electrical cabinet fire tests conducted by Failure causes attributable to aging and service wear of auxil-Sandia National Laboratones for the U.S. Nuclear Regulatory iary feedwater pumps (AUXFPs) are given and ranked in terms Commission. The first part of this test series investigated the ef- of importance. Cause identifications are made on the bases of facts of various cabinet parametews on a cabinet fne. The experience, post service examinations, and in situ assessments. second part of the test series, described here, investigated the Measurable parameters related to failure causes are identified. effects of such a fire on a large (18.3x12.2x6.1-m or 60x40x20 ISCM methods are also identified, evaluations are made based I ft) enclosure. Five tests involving a fire in a control cabinet were on Westinghouse experience, and recommendations are given. ] conducted under Part 2 of the 1est series. These tests investi- The methods are intended to yield required capabilities for es-gated the effects of fuel type, cabinet configuration, and encio- tablishing operational readiness, as well as for detecting and sure ventilation rate on the development of the enclosure envi. tracking deg*3dation. "he role of maintenance in mitigating ronment. Although fires as large as 1300 kW resulted, encio aging and service wear effects is discussed, and the relation-sure peak temperatures (outside the fire plume itself) were typi- ahip of maintenance to ISCM methods is identified. Predictive, preventive, and corrective maintenance practices are discussed cally less than 150 degrees C, with significant vertical thermal stratification observed. The most significant impact on the test and evaluated. Appendixes are included that contain failure data enclosure environment was that dense smoke, in all cases, re- base Information, AUXFP installation lists, discussion of low-flow suited in total obscuration of the enclosure within 6-15 min of testing, auxiliary feedwater system desenption, AUXFP mini-fire ignition. Enclosure ventilation rates as high as 8 room air mum flow-rate criteria, and proposed guidelines for full-flow test-changes per hour were found to be ineffective in purging the eng. TNs report was produced under the U.S. Nuclear Regula-smoke from this large enclosure. Similar obscuration problems tory Commission's Nuclear Plant Aging Research Program. had also been observed in the Part 1 tests, which utilized a NUREG/CR-4605: TRAINING MANUAL ON STATISTICAL METH-smmr enclosure with ventilation rates as high as 15 room air ODS FOR NUCLEAR MATERIAL MANAGEMENT. JAECH,J L. cheyes por hour. Battelle Memorial Institute, Pacific Northwest Laboratory. April NUREG/CR 4551 V5 DRF: EVALUATION OF SEVERE ACCI- 1988. 465pp. 8806230207. PNL 5855. 45896:128. DENT RISKS AND POTENTIAL FOR RISK REDUCTION. ZION This training manual on statistical methods for nuclear materi-POWER PLANT. Draft Report For Comment. KHATIB-RAHBAR, al management is a companion publication to a reference book PARK,Ca CHUN M.; et al. Brookhaven National Laboratory. entitled " Statistical Methods for Nuclear Material Management" February 1987. 1,938pp. 8802090229. BNL-NUREG.52029. edited by W. Michael Bowen and Carl A. Bennett. The training j 44301:001. manual follows the reference book in its terminology, notatron , l l

24 Main Citations and Abstracts and methodology. It is intended to be used with the reference micro. computer. Users can perform data base searches to fur-book either as a guide for self-paced instruction or as a problem nish HEP estimates and HCFD rates. In this manner, the NU-manual for formal courses in statistics on this subject. It con- CLARR system can be used to support a vanety of nsk assess-sists of a number of example problems for each chapter, follow- ment activities. This volume, Volume lit of a S volume senes, Ing the chapters in the reference book, and worked solutions to presents the procedures used to process HEP and HCFD for all problems. The matenal is presented in loose-leaf form with entry in NUCLARR and describes how to modify the existing each problem and solution starting on a new page so that in NUCLARR taxonomy in order to add equipmerit types of action connection with its usage with short courses the problems may verbs. Volume ill also specifies the various roles of the adminis-be handed the class participants first and the solutions later. trative staff on assignment to the NUCLARR Cleannghouse who NUREG/CR-4625: THE POSTIRRADIATION EXAMINATION OF are tasked with maintaining the data base, dealing with user re-THE DC MELT DYNAMICS EXPERIMENTS. FRYER,C.P.; quests, and processing NUCLARR data. 1 HITCHCOCK,J.T. Sandia National Laboratories. July 1988.  ! 53pp. 8808120165. SAND 86-1102. 46481:070. NUREG/CR-4639 V03 P2: NUCLEAR COMPUTERIZED LIBRARY The results of the postirradiation examination of two dry FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Guide debns beds composed of (1) pure UO(2) and (2) mixed UO(2) To Data Processing And Revision.Part 2: Hurnan Error Probabili-and stainless steel are presented. In the UO(2) oed, approxi- ty Data Entry And Revision Procedures. GILMORE,W.E.; mately 50% of the bed formed a molten pool which was sur- GERTMAN,D.L; GILBERT.B.G.; et al. EG&G Idaho, Inc. (subs. rounded by a high density crust. Lenticular pore formation and of EG&G, Inc.). November 1988.171pp. 8812190037, EGG-migration due to UO(2) vapor transport in the strong thermal 2458.47806:156. gradient were seen m the crust. In the mixed UO(2) and steel See NUREG/CR 4639,V03,P01 abstract. I bed, all of the steel had been molten. Steel and UO(2) migration were observed, both were thought to be caused by vapor trans- NUREG/CR-4639 V03 P3: NUCLEAR COMPUTERIZED LIBRARY port mechanisms. FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Guide NUREG/CR439 V01: NUCLEAR COMPUTERIZED LIBRARY To Data Processing And Revision. Part 3: Hardware Component FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Volume Failure Data Entry And Revision Procedures. GILMORE,W.E.;

1. Summary Desenption. GERTMAN,D.l.; GILMORE,W.E.; GERTMAN D.l.; GILBERT,B.G.; et al. EG&G Idaho, Inc. (subs.  !

GALYEAN.W.J.; et al. EG&G Idaho, Inc. (subs. of EG&G, Inc.)- of EG&G,. Inc.). November 1988. 95pp. 881219006d. EGG-February 1988. 30pp. 8805090146. EGG-2458. 45413:043. 2458.47804:154. A data management system has been implemented which See NUREG/CR-4639.V03 P01 abstract. supports a vanety of nsk-related analyses and provides a repos-itory of hardware component failure and human error probability NUREG/CR-4639 V04 P1: NUCLEAR COMPUTERIZED LIBRARY data to the nsk analyst. The Nuclear Computerized Library for FOR ASSESSING REACTOR RELIABILITY (NUCLARR). User's Assessing Reactor Rehability, NUCLARR, is an interactive, Guide,Part 1: Overview Of NUCLARR Data Retrieval graphically onented system which resides on a personal com- GILMORE,W.E.; GENTILLON,C.D.; GERTMAN.D.I.; et al. EG&G puter (PC) or PC-compatible environment. An overview of the Idaho, Inc. (subs. of EG&G, Inc.). June 1988. 38pp. data management system, including a desenption of data col-lection, specification, data structure, and taxonomies, is present- 8808050257. EGG-2458. 46395:338-ed in Volume I of this report. Programming activities, procedures The Nuclear Computerized Library for Assessing Reactor Re-for processing data, user's guide, and hard copy data manual liability (NUCLARR) is an automated data base management are presented in Volumes 11 through V. system for processing and storing human error probability and hardware component failure data. The NUCLARR system soft-NUREG/CR 4639 V02: NUCLEAR COMPUTERIZED LIBRARY ware resides on an IBWI (or compatible) personal micro-comput-FOR ASSESSING REACTOR RELIABILITY er. NUCLARR can be used by the end user to furnish data (NUCLARR). Programmer's Guide. CALL,0.J.; JACOBSON.J.A. mputs for both human and hardware reliability analysis in sup-EG&G Idaho, Inc. (subs. of EG&G, Inc.). September 1988. 298pp. 8810050266. EGG.2458. 46989:086. port of a variety of risk assessment activities. The NUCLARR The Nuclear Computenzed Library for Assessing Reactor Re- system is documented in a five-volume series of reports. liabmty (NUCLARR) is an automated data base agement " Volume IV: User's Guide" is presented in three parts. "Part 1: system for processing and stonng human error probabihty and Ovmiew of NUCLARR Data Retrieval," provides an introducto-hardware component fai;ure data. The NUCLARR system soft. ry overview to the system's capabilities and procedures for data ware resides on an IBM (or compatible) personal micro.comput. retneval "Part 2: Guide to Operations" contains the instructions er and can be used to furnish data inputs for both human and and basic procedures for using the NUCLARR software. "Part hardware rehabihty analysis in support of a vanety of nsk as. 3: NUCLARR System Description" provides an in depth discus-sessment activities. The NUCLARR system is documented in a sion of the design characteristics and special features of the five-volume senes of reports. Volume 11 of this senes is the Pro- NUCLARR software. Drammer's Guide for maintaining the NUCLARR system soft-ware. This Programmer's Guide provides, for the software engi. NUREG/CR-4639 V04 P2: NUCLEAR COMPUTERIZED LIBRARY neer, an onentation to the software elements involved, dis- FOR ASSESSING REACTOR RELIABILITY (NUCLARR). User's cusses maintenance methods, and presents useful aids and ex- Guide,Part 2: Guide To Operations. GILMORE,W.E.; amples. GENTILLON C.D.; GERTMAN.D.I.; et al. EG&G idaho, Inc. NUREG/CR-4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY (subs. of EG&G, Inc). June 1988.117tg. 8808050281. EGG-2458.46393:238. FOR ASSESSING REACTOR RELIABILITY (NUCLARR) Guide To Data Processing And Revision. Part 1: Technical Overview. See NUREG/CR-4639,V04,P01 abstract. GILMORE.W.E.; GERTMAN D.I.; GlLBERT,B.G.; et a!. EG&G Idaho, Inc. (subs. of EG&G, Inc.). November 1988. 48pp. NUREG/CR-4639 VD4 P3: NUCLEAR COMPUTERIZED LIBRARY 8812190018. EGG-2458. 47806:108. FOR ASSESSING REACTOR RELIABILITY (NUCLARR). User's The Nuclear Computenzed Library for Assessing Reactor Re. Guide.Part 3: NUCLARR System Desenption. GILMORE,W.E.; fiabihty (NUCLARR) is an automated data base management GENTILLON.CP.; GERTMAND.L; et al. EG&G Idaho, Inc. system for processing and stonng human error probability (HEP) (subs. of EG&G, Inc.). June 1988. 229pp 8808050242. EGG-and hardware component failure data (HCFD). The NUCLARR N58. 46393:355. system software resides on an IBM (or compatible) personal See NUREG/CR-4639,V04.P01 abstract.

Main Citations and Abstracts 25 NUREG/CR-4639 V05 P1: NUCLEAR COMPUTERIZED LIBRARY lish the seismic frr.gility of nuclear pswer plant equipment by j FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data use of existing test data and demonstrated its ' application by Manual,Part 1: Summary Desenption. GERTMAN,D.I.; considering two equipment pieces. In Phase !! of the program, GILBERT,B.G.; GILMORE.W.E.; et al. EG&G Idaho, Inc. (subs. BNL has collected additional test data, and has further ad-of EG&G, inc4 June 1988.123pp 8808050298. EGG-2458. vanced and is applying the methodology to determine the fragili-46391:298. ty levels of selected essential equipment categories. The data j This volume of a five-volume series summarizes those data evaluation of four equipment families, namely, motor control ' currently resident in the first release of the Nuclear Computer. center, switchboard, panelboard and power supply has been l tzed Library for Assessing Reactor Reliability (NUCLARR) data completed. Fragility levels have been determined for various base. The raw human error probability (HEP) and hardware failure modes of each equipment class and the deterministic re-component failure data (HCFD) contained herein are accompa- sults are presented in terms of test response spectra. In addi-nied by a glossary of terms and the HEP and hardware taxono- tion, the test data ' nave been analyzed for determination of the 1 mies used to structure the data. Instructions are presented on respective probabilistic fragility levels. The zero period accelera- j how the user may navigate through the NUCLARR data man- tion and the average spectral acceleration over a frequency I agement system to find anchor values to assist in solving risk- range of ir.terest are used as inputs in the statistical analysis. related problems. " Volume V: Data Manual" wil! be updated on The resulting fragility parameters are presented in terms of a a periodic basis so that nsk analysts without 8" cess to a com- median valse, an uncertainty coefficient and a randomness co-puter may have access to the latest NUCLARR data. Those efficient. Ultimately, each fragility level is expressed in terms of users wishing to learn rnore regarding the computer-based inter- a single desenptor called an HCLPF value corresponding to a activo search and report- generation capabilities of the NU- high (95%) confidence of a low (5%) probability of failure. The CLARR system are referred to the other volumes in the important observations and recommendations for future re-NUREG/CR-4639 series, e g., " Volume I: Summary Desenp- search work in the fragility area are included in this report. One tion" or " Volume IV: User's Guide." of the important needs is to study the applicability of the fragility NUREG/CR-4639 V05 P2: NUCLEAR COMPUTERIZED LIBRARY results to the earlier vintage equipment for which little or no test data exist. FOR ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual,Part 2: Human Error Probability (HEP) Estimates-GERTMAN D.I.; GILBERT,B.G.; GILMORE,W.E.; et al. EG&G NUREG/CR-4662: CLOSEOUT OF IE BULLETIN 80-18.MAINTE-NANCF OF ADEOUATE M!NIMUM FLOW THRU CENTRIFU-Idaho, Inc. (subs. of EG&G, Inc.). June 1988. 817pp. GAL CHARGING PUMPS FOLLOWING SECONDARY SIDE 8808050309. EGG 2458. 46389:201. See NUREG/CR 4639,V05 P01 abstract. HIGH ENERGY LINE RUPTURE. FOLEY,W.J.; DEAN,R.S.; HENNICK,A. Parameter, Inc. January 1988.36pp.BBC2030234. NUREG/CR-4639 V05 P3: NUCLEAR COMPUTERIZED LIBRARY PARAMETER IE159. 44240:324. FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data On May 6,1980, Westinghouse reported to the NRC/IE per Manual,Part 3: Hardware Component Failure Data (HCFD). Title 10 CFR Part 21 that one or more cent" fugal charging GERTMAN.D.I.; GILBERT,B.G.; GILMORE W.E.; et al. EGaG pumps could be damaged by low flow at certain plants before Idaho, Inc. (subs. of EG&G, Inc.). June 1988. 537pp. satisfactory termination of safety injection after a secondary 8808050291. EGG-2458. 46392:061. side high energy line rupture. The plants of concern had been See NUREG/CR-4639,V05,P01 abstract. notified, a calculational method of evaluation and plant-specific reviews had been recommended, and interim modifications had NUREG/CR-4639 V05 P4: NUCLEAR COMPUTERIZED LIBRARY been proposed by Westinghouse. The NRC/IE issued IE Bulle-FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data tin 80-18 on July 24,1980 because of this potential safety-relat-Manual,Part 4: Summary Aggregations. GERTMAN,D.I.; GILBERT,B.G.; GILMORE,W.E.; et al. EG&G Idaho, Inc. (subs. ed problem. Licensees and near-term licensees of pressurized of EG&G, Inc.). June 1988. 315pp. 8808050249. EGG-2458. water reactors (PWRs) were required to take specific actions and submit written responses. Utilities with PWRs under con- 'i NUREG/CR-4639,V05,P01 abstract- structon were issued the bulletin for information, in preparation for the licensing process. Evaluation of utihty responses and NUREG/CR-4651 V02: DEVELOPMENT OF RIPRAP DESIGN NRC/ Region inspection reports shows that the bulletin can be CRITERIA BY RIPRAP TESTING IN FLUMES. Phase ll. Followup closed out per specific criteria for 44 (96%) of the 46 facilities Investigations. ABT,S.R.; WITTLER.R.J.; RUFF,J.F.; et al. Oak to which it was issued for action. A followup item for the remain-Ridge National Laboratory. September 1988. 117pp. ing dual plant is proposed for use by the NRC to ensure satis-l 8810040075. ORNL/TM-10100. 46975:143. factory completion of required and corrective actions. Flume studies were conducted in which r!prap embankments were subjected to overtopping flows. Embankment slopes of 1, NUREG/CR-4665: CLOSEOUT OF IE BULLETIN 83-08:ELECTRI-2,8,10, and 20% were protected with riprap containing median CAL CIRC,UIT BREAKERS WITH AN UNDERVOLTAGE TRIP stone sizes of 1, 2, 4, 5, and/or 6 in. Riprap layer thickness FEATURE IN USE IN SAFETY-RELATED APPLICATIONS ranged from 1.5 D(50) to 4 D(50). Riprap design enteria for OTHER THAN THE REACTOR TRIP SYSTEM. FOLEY,W.J.; overtopping flows were developed in terms of unit discharge at DEAN,R.S.; HENNICK A. Parameter, Inc. April 1988. 41pp. failure, interstitial velocities and discharges through the nprap 8805090133. PARAMETER IE162. 45413:001. layer, resistance to flow over the nprap surface, effects of nprap The NRC/IE issued Dulletin 83-08 December 28,1983 be-layer thickness and gradation on nprap stability, and potential cause of concern about circuit breaker deficiencies reported per impacts of integrating soil into the nprap layer for nprap stabili- previ us bulletins 83-01 and 83-04. The object of IEB 83-08 ration A riprap design procedure is presented for overtopping was to assure proper operation of circuit breakers with under-flow conditions. voitage inp attachments (UVTAs) in all safety-related applica-tions other than use as reactor trip breakers (RTBs). The bulle-NUREG/CR-4659 V02: SEISMIC rRAGILITY OF NUCLEAR tin was issued for action to all ficensees and holders of con-POWER PLANT COMPONENTS (PHASE il) Motor Control struction permits of power reactors. Evaluation of utility re-Center. Switchboard.Panelboard And Power Supply. BANDYO- sponses and NRC/ Region inspection reports shows that the PADHYAY; HOFMAYER.C.H.; KASSIR.M.K.; et al. Brookhaven bulletin can be closed out per specific criteria for 123 (99%) of National Laboratc,ry. January 1988. 65pp. 8803080476. BNL- the 124 facilities to which it was issued. A followup item is pro- < NUREG-52007. 44631:2B7. In Phase I of the Component Fragility Program, Brookhaven posed for use by the NRC to assure completion of required ac- ) tions at the remaining facility. Circuit breakers with UVTAs were i National Laboratory (BNL) has developed a proce.1ure to estab-used in safety-related applicat;ons other than the reactor trip l i l l

26 Main Citations and Atstracts system in six (6) facilities. Malfunctions of the UTVAs were re- served with either method. The elution peaks obtained with nep-ported only for the facility for which folicwup is proposed. tunium and uranium were asymmetrical and the shapes were - often complex, observations which suggest irreversibilities in the j NUREG/CR-4667 V04: ENVIRONMENTALLY ASSISTED CRACK- sorption reaction. An expenment was performed to provide in. ING IN LIGHT WATER REACTORS. Semiannual formabon on the compositions of the first groundwater that will Report, October 1986 - March 1987. SHACK,W.J.; contact waste canisters in a tuff-hosted repository after very KASSNER.T.F.; MAlYA,P.S.; et al. Argonne National Laboratory. near field temperature, have cooled to below 100 degrees C. February 1988. 77pp. 8804080134. ANL-87-41. 45039:328. This progress report summanzes work performed by Argonne NUREG/CR-4728: EQUIPMENT QUALIFICATION RESEARCH National Laboratory on environmentally assisted cracking in TEST OF A HIGH-RANGE RADIATION MONITOR. hght water reactors during the six months from October 1986 - RICHARDS.E.H.: JACOBUS.M.J.; DROZDA P.M.; et al. Sandia March 1987. National Laboratories. February 1988. 64pp. 8806230158. SAND 86-1938. 45898:305. NUREG/CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1986,A STATUS REPORT. A high-range radiation detector was tested in a simultaneous MINARICK.J.W.; HARRIS,J.D.; AUSTIN,P.N.; et al. Oak Ridge steam and radiation environment simulating a postulated loss-National Laboratory. May 1988.183pp. 8805260278. ORNL/ of- coolant accident (LOCA) to assess possible synergistic ef-iocts that may be important to its performance in an accident. NOAC-232. 45650:260. The detector, manufactured by General Atomic, was simulta-Thirty-five operational events, reported in licensee event re-neously subjected to a simulated accident environment includ-ports and occurring at commercint LWRs dunrm 1986, are con-sidered to be precursors to potential severo core damage. ing 171 degree C (340 degrees F) steam at 410 kPa gage (60 These are described along with associated significance esti- Psig) and 4 Mrad /hr gamma radiation while its performance was monitored. Test results showed that the detector successfully mates, categorization, and subsequent analyses. This study is a continuation of earlier work, which evaluated the 1969-1981 and operated at the high dose rate and temperature, without evi. 19%1985 events. The report discusses (1) the general ration- dence of synergisms. However, at reduced radiation levels in a ale for this study, (2) the selection and documentation of events saturated steam environment, the detector signal at the readout as precursors, (3) the estimation and use of conditional prob- module detenorated in accuracy or ceased altogether, The abilities of subsequent severe core damage to rank precursor cause of these anomalies is attributed to leakage currents and/ events, and (4) the initial conclusions from the assessment of or possible galvanic action in the coaxial connections and/or 1986 events and from the collective assessment of 1984 1986 cabfes. WUREG/CR-4735 V03: EVALUATION AND COMPILATION OF NUREG/CR-4674 V06: PRECURSORS TO POTENTIAL SEVERE DOE WASTE PACKAGE TEST DATA. Biannual CORE DAMAGE ACCIDENTS:1986,A STATUS REPORT, Report. February-July 1087. !NTERRANTE,C.; ESCALANTE,E.; MINARICK,J.W.; HARRIS.J.D.; AUSTIN,P.N.; et al. Oak Ridge FRAKER,A.; et al. National Institute of Standards & Technology National Laboratory. May 1988. 315pp. 8805260286. ORNL/ (formerly National Bureau of Standa. May 1988. 16tpp. NOAC 232. 45631:093. 8806070059.45727:023. See NUREG/CR-4674,V05 abstract. This report summarizes evaluations by the National Bureau of Standards (NBS) of Department of Energy activities on waste NUREG/CR-4688 V02: OUANTIFICATION AND UNCERTAINTY packages designed for containment of high-level nuclear waste ANALYSIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN (HLW). The &aste package is a proposed engineered barrier LIGHT WATER REACTORS (OUASAR) Part it Sensitivity Analy. that is part of a permanent repository for HLW. Metal alloys are l sis Techniques. ISHIGAMI,T, Japan Atomic Energy Research In. stitute. CAZZOLI E.; KHATIB-RAHBAR; et al. Brookhaven Na. the principal barners within the engineered barrier system. The corrosion of candidate container materials, particularly carbon tional Laboratory. October 1987 54pp. 8802240354. BNL- steels, stainless steels and cooper, is discussed. The level of NUREG-52008. 44470:081. understanding of several proposed container materials is ques-Existing methods for sensitivity analysis are described and troned for the candidate repository in tuff. Three issues are ad-new techniques are proposed. These techniques are evaluated dressed: (1) The possibility of stress induced failure of Zircaloy, through consideration relative to the OUASAR program. Ments and limitations of the various approaches are examined by a de- (2) possible corrosion of copper and copper alloys and (3) the lack of site-specific charactenzation data. The Basalt Waste iso-tailed application to the Suppression Pool Aerosol Removal lation Project (BWIP) section discusses localized corrosion and Code (SPARC), environmentally- assisted cracking of AISI 1020 steel at elevat-NUREG/CR 4708 VD2: PROGRESS IN EVALUATION OF RADIO- ed temperatures (150 degrees C). For the proposed salt site, NUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY the importance of the duration of corrosion tests and conditions DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE that may preclude prompt initiation of long-term testing are dis-PROJECTS Reoort For April 1966 - September 1987. cussed. NBS work related to vitrification of HLW horosilicata MEYER,R E.; ARNOLD,W.D.; HLENCOE,J.G.; et al. Oak Ridge glass at the West Valiey Demonstration Project (WVDP) and at National Laboratory. July 1988. 74pp. 8808300246. ORNL/TM- the Defense Waste Processing Facility (DWPF) is discussed. 10147. 46631:241. Activities of the Matenals Characterization Center (MCC) are information that is being developed by projects within the De- presented. To stress the necessity of independent data interpre. partment of Energy (DOE) pertinent to the potential geochemi- tations, an illustration is given of a conclusion drawn from pub-cal behavior of radionuclides at the Yucca Mountain, Nevada. lished data that is different from the conclusion reached by the candidate site for high-leni radioactive waste repository is investigator, being evaluated by Oak Ridge National Laboratory (ORNL) for the Nuclear Regulatory Commission (NRC). Batch sorption ratio NUREG/CR-4735 V04: EVALUATION AND COMPILATION OF determ: nations were conducted for strontium, cesium, uranium, DOE WASTE PACKAGE TEST DATA bannual Report; August and technetium onto samples of tuff using real and synthetic 1987 - January 1988. INTERRANTE,C.; ESCALANTE,E.; groundwater J-13. Columns 1 cm in diameter and about 5 cm FRAKER A.; et al. National institute of Standards & Technology long were constructed, and expenments were conducted with (formerly National Bureau of Standa. August 1988. 183pp. the objective of correlating the results of batch and the column 8810110232.47044:286. expenments. For strentium and cesium, fairly good correlation This report summanzes evaluations made in the penod between values of the sorption ratio obtained by the two meth- August 1987 to January 1988 by the National Bureau of Stand-ods was observed. Little or no technetium sorption was ob- ards (NBS) of Department of Energy (DOE) activities on waste

Main Citations and Abstracts '27 packages for containment of high-level nuclear waste (HLW). NUREG/CR-4763i SAFETY-RELATED EQUIPMENT SURVIVAL , The waste package will be part of a propot6ed engineered bar- IN HYDROGEN BURNS IN LARGE DRY PWR CONTAINMENT rier system (EBS) in the permanent repository for HLW. Metal - BUILDINGS. KING,D.B.; NICOLETTE,V.F.; DANDINI,V.J.; et al. alloys are principal barriers within the EBS. The Budget Recon- Sandia National Laboratories. March 1988. 299pp. 88042ft0588, ciliation Act for Fiscal Year 1988 provided that only the Yucca SAND 86-2280. 45272:099. Mountain, NV, site (in which tuff is the geologic medium) shall Analytical and experimental investigations of equipment sur-be characterized for use as a HLW repository. Five reviews vival in hydrogen burns in large dry PWR containment buildir.us - weie completed for tuff covering ferrous alloys, copper, ground- have been conducted. Both atmospheric and subatmospheric water chemistry, and glass. Two issues were identified for the containments were considered. Two sets of analytical studies Yucca Mountain site: (1) the approach used to calculate corro- were carried out for atmospheric large dry containments. One-sion rates for ferrous alloys, and (2) the observation of crevice' sei analyzed the hydrogen burn that occurred as a result of the. corrosion in a copper-nickel alloy. It is noted that pluts.nium can . _ March 1979, accident at Three Mile island. The other set con-form pseudo- colloids that may facilitate transport. NBS work re- sidered a bybrid power plant consisting of the Zion reactor lated to the vitrification of HLW borosilicate glass at the West housed in the TMi-2 containment building. An analytical study of

                                      . Valley Demonstration Project (WVDP) and the Defense Waste                             subatmospheric containments was also carried out using a Processing Facility (DWPF) and activities of the DOE Material                         model of the Surry nuclear power plant. To complement the Characterization Center (MCC) is also included. Appended are                          analyses, a series of experiments simulating hydrogen bums in NBS reviews of twenty other DOE technical reports and select-                         large dry containments was also conducted using the Sandia -

ed other reviews which, although conducted under the basalt severe combined environment test chamber (SCETCh). The ex-and salt programs, are considered relevant to the present work periments investigated the survivability of thermally and radi-on the tuff program. For these former candidate sites, technicci ation aged nuclear qualified Brand Rex power and control cable discussions are given for the corrosion of metals proposed for and a Barton '763 pressure transmitter in a simulated LOCA/hy-the canister, particularly carbon steels, stainless steels, and drogen burn environment. NUREG/CR-4775: GUIDE FOR PREPARING' OPERATING PRO-NUREG/CR-4740: NUCLEAR PLANT-AGING RESEARCH ON C,EDURES FOR SHIPPING PACKAGES. WITTE,M.C. Lawrence REACTOR PROTECTION SYSTEMS. MEYER,L.C. EG8G Livermore National Laboratory. December 1968. 32pp. Idaho, Inc. (subs. of EG&G,'Inc.). MEYER,L.C. Idaho National 8 O Eng n g Labora ory. January 1988.131pp. 8804080236. ep es g ting manuals of operating rocedures for shipping packages used for radioactive materi- ~ This report presents the results of a review of the Reactor als. Guidelines are provided which are based on the broad ex-Trip System (RTS) and the Engineered Safety Feature Actuating shipping industry and are consistent Ejt G 10 7 System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data NUREG/CR-4777: STEAM ' OXIDATION OF ZlRCALOY CLAD-base, Nuclear Plant Reliabihty Data System, and plant mainte-DING IN THE ORNL FISSION PRODUCT RELEASE TESTS. nance records. Our purpose is to evaluate the potential signifi- YAMASHITA.T. Oak Ridge National Laboratory / March 1988. cance of aging, including cycling, trips and testing as contrib* 58pp. 8805200101, ORNL/TM-10272. 45560:051. tors to degradation of the RTS and ESFAS. Tables are present-- A simple model (ZlRCOX) has been developed to calculate ed that show the percentage of events for RTS and ESFAS the extent of oxidation of Zircaloy cladding in the vertical fur-classitied by cause, components, and subcomponents for each nace tests conducted at ORNL. The model is baseci on the fact ' of the Nuclear Steam Supply System vendors. A representative that the oxidation of Zircaloy is governed by the parabolic rate ' Babcock and Wilcox plant was selected for detailed study. The law. By introducing the equivalent time t*, the model is able to U.S. Nuclear Regulatory Commission's Plant Aging Research handle nonisothermal oxidation as well as' isothermal. The guidehnes were followed in performing the detailed atudy that degree of cladding oxidation and H(2) production were calculat-identified materials susceptible to aging, stressors, environmen- ed using three sets of rate constants for the Zircaloy oxidation, tal factors, and failure modes for the RTS and ESFAS as gener- and these results were compared with those measured experi-ic instrumentation and control systems. Funchonal indicators of mentally. The calculated .results shownd excellent agreement degradation are listed, testing requirements evaluated, and reg- with the experiments The rate constant for Zircaloy oxidation ulatory issues discussed. that gave the best agreement was one based upon combined Pawel et al. and Prater and Courtright data. The ternperature NUREG/CR-4747 V02: AN AGING FAILURE SURVEY OF LIGHT difference between the cladding and the zirconia fumace tube WATER REACTOR SAFETY SYSTEMS AND COMPONENTS- liner was estimated by the HEATING 6 heat transfer program. MEALE,D.M.; SATTERWHITE.D. EG&G idaho, Inc. (subs. of EG&G, Inc.). July 1988. 557pp. 8809220097. EGG-2473. and was found to be about 50 K for the maximem oxidation j rate. The heating atmosphere during test VI-1 was also evaluat. 46875:333. ed in terms of the oxygen potential. This report describes the methods, analyses, results, and conclusions of two different aging studies. The first study con- NUREG/CR-4778: PRELIMINARY STUDIES OF THE MORPHOL-'

                                    - sists of a survey of light water reactor component failures asso-                      OGY OF THERMAL GRADIENT TUBE DEPOSITS FROM FIS .

ciated with 15 selected safety and support systems. Analysts SiON PRODUCT RELEASE EXPERIMENTS. WISBEY,S.J. Oak used computerized sorting techniques to classify component Ridge National Laboratory. March 1988. 57pp. 8805090156. failures into generic failure categones. The second ciudy con- ORNL/TM-10273. 45413:296. ' sists of careful examination of component failures records to Sections of thermal gradient tubes and deposits from filters idemify and categorize the reported cause of component fail- used as collectors in several fission product release tests at ures. The systems cvaluated in the failure- cause analysis were ORNL.have been examined by scanning electron microscopy the auxihary feedwater, Class 1E electrical power distribution, with elemental identification by energy dispersive X ray analysis, i high. pressure injection, and service water. Tables and figures Shape, size, and composition of the deposits are reported; cor- I are presented, indicating the systems and the components relations with experimental conditions such as gas composition within those systems most affected by aging. Also provided are

                                                                                                                                                                                                   )

and temperature, as well as with independent analyses, have I engineering insights drawn from the data. This report is the been made where possible. A wide variety of shapes and struc- i second of two volumes and presents all of the Volume 1 data tures, apparerstly dependent on the deposition temperature, i from FY 86 combined with the data gathered ir, FY-87, were photographed and elemental analyses were recorded. Al- ) j i

28 Main Citations and Abstracts though some fission products (Cs, Ba, Ag) were detected, struc- flexibility factors associated with piping system branch connec-tural and impurity elements (Sn, Si, S. W, Pt) were predominant tions and nozzles. Recommendations are given for developing in most cases. Recommendations for analytical procedures and needed improvements. handling of similar samples are made for the future. NUREG/CR-4780 V01: PROCEDURES FOR TREATING ANNUAL REPORT. July December 1980. Volume 36. . Sandia ," COMMON CAUSE FAILURES IN SAFETY AND RELIABILITY 0" bora on STUDIES. Procedural Framework And Examples. MOS EH,A.; Sp g65[00 FLEMING,K.; PARRY,G.; et al Pickard, Lowe & Garrick, Inc. S D8 Sandia National Laboratories is conducting, under USNRC i January 1988. 205pp. 8802030161. EPRI NP-5613. 44240:013. s onsorship, phenomenological research related to the safety This report presents a framework for the inclusion of the impact of common cause failures in risk and reliability evalua- eddumNMNNMMh experiments to simulate the phenomenology of the accident tions. Common cause failures are defined as that subset of de- conditions and the development of analytical models, verified by pendent failures for which causes are not explicitly included in the logic model as basic events The emphasis here is on pro-tems performance and behavior under abnormal conditions. The viding procedures for a practical, systematic approach that can be used to perform and clearly document the analysis. The framework comprises four major stages: (1) system logic model and analytical methods to (1) identify and define safety issues, development, (2) identification of common cause component (2) understand the progression of nsk- significant accident se-groups, (3) common cause modeling and data analysin, and (4) quences, and (3) conduct safety assessments. The collective system quantification and interpretation of results. The frame- NRC-sponsored effort at Sandia National Laboratories is direep work and the methods discussed for performing the different ed at enhancing the technology base supporting licensing deci-stages of the analysis integrate insights obtained from engineer. sions. ing assessments of the system and the historical evidence from multi le tai e vents into a systematic, reproducible, and de- NUREG/CR-4807: SURFACE-COMPLEXATION MODELING OF RADIONUCLIDES ADSORPTION IN SUBSURFACE ENVIRON-MENTS. SIEGEL,M.D. Sandia National Laboratories. KENT,D.B.; NUREG/CR-4784: INFLUENCE OF GROUNDWATER ON SOIL- TRIPATHI,V.S.; et at Stanford Univ., Stanford, CA. March 1988. STRUCTURE INTERACTION. COSTANTINO,C.J.; PHILIPPACO- 141pp. 8803220121. SAND 86-7175. 44774:149. POULO Brookhaven National Laboratory. December 1987. Requirements for applying the surface-complexation modeling 68pp.8803090227. BNL-NUREG.52039. 44646:035- approach to simulating radionuclides adsorptian onto geologic This report presents a i,ummary of the second year's effort materials are discussed. Accurate description of adsorption be-on the subject of the influence of foundation ground water on havior requires that chemical properties of both adsorbent and the SSI phenomenon. A finite element computer program, de- adsorbate be characterized in conjunction with determinations veloped during the first year's effort, was used to study the of extent of adsorption. Critical chemical properties of adsorb-impact of depth to the ground water surface on the SSI prob. ents include dissolution and oxidation / reduction behavior, types lem. The formulation used therein is based on the Biot dynami? and densities of adsorption sites, and interaction of sites with equations of motion for both the solid and fluid phases of a typi- solution components. Important adsorbate properties include cal soi Frequency dependent interaction coefficients were then hydrolysis, complexation, oxidation / reduction, and oligomeriza-generated for the two dimensional plane problem of a ngid sur- tion. Adsorption behavior is described by a set of chemical re-face footing moving against a linear soi The soil is considered actions and binding constants between: adsorption sites and dry above the GWT and fully saturated below. The results indi- solution components, adsorbate and solution components, and cate that interaction coefficients are significantly modified as adsorbate and adsorption sites Methods for implementing such compared to the comparable values for a dry soil, particularly an approach are discussed; examples based on selute adsorp-for the rocking mode of response, if the GWT is close to the t on onto oxides are presented. Implementation of the surface-  ! foundation. As the GWT moves away from the foundation' coniplexstion modeling approach would greatly improve the pre-these effects decrease in a relaLvely orderly fashion for both dictability of the role of adsorption in regulating radionuclides the horizontal and rocking modes of response. For the vertical transport in subsurface environments. Interaction coefficients, the rate of convergence to the dry solu-tion is frequently dependent. Calculations were made to study NUREG/CR-4811: THE ECONOMIC COSTS OF RADIATION-IN-the impact of the modified interaction coefficients on the ts- DUCED HEPA EFFECTS. Estimation And Simulation. sponse of a typical nuclear reaction building. The amplification NIEVES.L.A.; %VIL,J.J. Battelle Memorial institute, Pacific factors for a slick model placed atop a dry and saturated soil Northwest Laboratory. August 1988. 201pp. 8809280245. PNL-were computed. It was found that pore water caused the rock- 6097. 46953 088. ing response to decrease and translationel response to in- The purpose of this report is to improve the quantitative infor-crease over the frequency range of interest, as compared to the mation available for use in evaluating actions that alter health response on dry soil. risks due to population exposure to ionizing radiation. To project NUREG/CR-4785: REVIEW AN. 'ALUATION OF DEulGN the potential future costs of changes in health effects risks, Pa-ANALYSIS METHODS FOR CALCULATING FLEXIBILITY OF cific Northwest Laboratory (PNL) constructed a probabilstic NOZZLES AND BRANCH CONNECTIONS. MOORE,S E.; computer model, Health Effects Costs Model (HECOM), which RODABAUGH,E.C.; MOKHTARIAN,K.; et al. Oak Ridge National utiiizes the health effect incidence estimates from accident con-Laboratory. December 1987.108pp. 9804010215. ORNL-6339. sequeacss modo!s. It calculates the discounted sum of the eco-44974:004. nomic costs associated with population exposure to ionizing ra-Modern piping system design generally includes an analytical diation. Costs of three major types of health effects are estimat-determination of displacements, rotations, moments, and reac- ed: acute radiation injuries (and fatalities), latent cancers, and tion forces at various positions nlong the piping system by impairments due to genetic effects. The economic costs of means of a so- called flexibility andysis. The analytscal model is health effects estimated by HEOOM represent both the value of normally based on a strength-of matenals description of the resources consumed in diagnosing, treating, and canng for the l piping system as an interconnected set of straight and curved patient and the value of goods not produced because of illness l beams along with " flexibility factors" that are used to compen- or premature death due to the health effect. Additional costs to l sate for inaccuracies in the model behavior. This report gives an society, such as pain and suffenng, are not included in the PNL j in-depth evaluation of the vanous analytical desenptions of the economic cost measures. I 1 l l _____________ _ _ _ _ _J

1 1 Main Citations and Abstracts - 29 5 I NUREG/CR-4813 R01: ASSESSMENT OF LEAK DETECTION full- scope PRA of the LaSalle nuciear power station. Thus, in SYSTEMS FOR LWRS. KUPPERMAN,D.S. Argonne National . addition to the review, this report contains recommendations for Laboratory. October 1988. 57pp. 8811010260. ANL-86-52. a suitable uncertainty analysis methodology for the LaSailo i 47271:227. PRA. The topical report summartzes work performed by-the Ar. - gonne National Lavoratory as subcontractor on on-line leak NUREG/OR-4857: CADET:A DECISION SUPPORT SYSTEM FOR . monitonng of LWRs during the 12 montns from October 1987 to LIGHT WATER . REACTOR SAFETY. NICOLOSI,S.L; 1 September 1988. HESSE.D.J. Battelle Memorial Institute, Columbus Laboratories. September 1988.155pp. 8809220058. BMI-2146. 46880:095. NUREG/CR-4628: FATIGUE CRACK GROWTH OF PART- CADET (Computer Aided Decision Tool)is a decision support THROUGH CRACKS IN PRESSURE VESSEL AND PIPING system for light water reactor safety which is designed for use i STEELS. Air Environment Results CULLEN W.H.; JOLLES,M.R. on personal computers. As a decision support system, it pro. l Materials Engineering Associates, Inc. October 1988. 64pp. vdes a user-friendly data base program complemented with 8812190123. MEA-2198. 47805:320. several computational capabilities. The data base component of This report describes the first phase of a program to evaluate the program provides users with pertinent data from a variety of the effect of PWR water environment on the fatigue crack sources. The computational portion of the program provides growth in part-through surface-crack specimens en pressure measures of consequence and risk, and a means for performing vessel and piping steels, unciad and clad. The objective of the "what if" analyses with selected elements of the data base, program is to compare the crack growth rate data from a two- Predictive capabihties incorporated into the present version of . dimensional crack with that generated from the more popular CADET include effects of time of containment failure, effects of compact specimens. The results of room air environment tests, l containment leak rate, and influences of filtered venting on the J at ambient 'emperature and 288 degrees C are presen'ed. The outcome of accident sequences. materials used, the description of the cladding process, the test specimen preparation for the direct current potential drop NUREG/CR-4860 R01: FLAW DENSITY EXAMINATIONS OF A (DCPD) technique, and the computerized data acquisition sys- CLAD BOILING WATER REACTOR PRESSURE VESSEL SEG-tems are described. The computational techniques for the proc. l

                                                                                . MENT. COOK,K.V.: MCCLUNG,R.W. Oak Ridge National Labo-                 '

essing of the DCPD data are detailed, and the crack extension ratory. February 1988. 34pp. 8801080170. ORNL/TM-10364. plots are presented. The crack growth rate data from these air 45038:156. environrrent tests are compared with data sets from tests of As part of the Oak Ridge National Laboratory's Heavy-Sec-compact tension specimens in the same orientations as the tion Steel Technology Program, studies have been ccnducted to major and minor axes of the semiefliptical part-through crack. determine flaw density in a section of reactor pressure vessel NUREG/CR-4834 V02: RECOVERY ACTIONS'IN PRA FOR THE cut from the Hope Creek Unit 2 vessel. This boiling water reac-RISK METHODS INTEGRATION AND EVALUATION PRO . tor vessel was never in service. One objective was to evaluate GRAM (RMIEP).Voiume 2: Application Of The Data-Based the approximate 0.7 by 3-m (2 by 10-ft) segment of the vessel Method. WHITEHEAD,0.W. Sandia National Laboratories. De- provided using ultrasonic flaw detection methods performed cember 1987. 45pp. 8803080461. SANDB7 017p. 44634:057. with both ASME Code techniques and supplemental ultrasonic in a Probabilistic Risk Assessment (PRA) for a nuclear power inethods. A second objective was to evaluate the inner surface plani, the analyst identifies a set of potential core damage stainless steel cladding for cracks with a high sensitivity pene-events and their estimated probabilities of occurrence. These  : trant examination. Both objectives were successfully completed. events include both equipment failures and human errors. If op- Five Code-recordable indications were detected ultrasonically; erator recovery from an event within some specified time is con. however, all were found to be anomalies associated with the sidered, the probability of this recovery can be included in the cladding. One flaw was detected by the supplemental ultrasonic PRA. This report provides PRA analysts with a step by-step tests, and it was analyzed destructively. This flaw was a pipelike methodology for including recovery actions in a PRA. The re. indication, about 20 mm (0.8 in.) long extending along the covery action is divided into two distinct phases: a Diagnosis length of the longitudinal weld in which it was located and was Phase (realizing that there is a problem with a entical parameter about 20 mm below the cladding surface. The flaw had a and deciding upon the correct course of action) and an Action through-wall dimension (or length) of about 6 mm (0.24 ~in.) for Phase (physically accomplishing the required action). In this an approximate 3-mrn (0.1-in.) distance along the 20-mm major methodology, time-reliability curves, which were developed from length. No flaws were detected by the penetrant examination of simulator data on potentially dominant accident scenarios, are the cladding surface, used to provide estimates for the Diagnosis Phase, and other existing methodologies are used to provic'e estimates for the NUREG/CR-4864 V01: THERMODYNAMIC TABLES FOR NU-Action Phase. CLEAR ' WASTE ISOLATION. Aqueous Solutions Database. PHILLIPS,S.L.; HALE,F.V.; SILVESTER,L.F.; et al. Lawrence NUREG/CR-4836: APPROACHES TO UNCERTAINTY ANALYSIS Berkeley Laboratory. June 1988.192pp. 8807110498. LBL-IN PROBABILISTIC RISK ASSESSMENT. BOHN,M.P.; 22860.46082:118. WHEELER T.A.; PARRY,G.W. Sandia National Laboratones- Tables of consistent thermodynamic property values for nu-January 1988.108pp. 8805090083. SAND 87-0871. 45449:001. clear waste isolation are given. The tables include critically as-An integral part of any probabilistic risk assessment (PRA) is sessed values for Gibbs energy of formation, enthalpy of forma-l the performance of an uncertainty analysis to quantify the un- tion, entropy and heat cupacity for generic minerals; solids; certainty in the point estimates of the nsk measures considered- aqueous ions; ion pairs and complex ions of selected actinide While a variety of classical methods of uncertainty analysis and fission decay products at 25 degrees C and zero ionic exist, application of these methods and developing new tech- strength. These intnnsic data are used to calculate equilibrium niques consistent with existing PRA data bases and the need constants and standard potentials which are compared with typ. for expert (subjective) input has been an area of considerable ical experimental measurements and other work. Recommenda-interest since the pioneenng Reactor Safety Study (WASH- tions for additional research are given. 1400) m 1975. This report presents the results of a critical I review of existing methods for performing uncertainty analyses NUREG/CR-4873: BENCHMARK STUDY OF THE l-DYNEV for PRAs. with special emphasis on identifying data base limita- EVACUATION TIME ESTIMATE COMPUTER CODE. tions on the vanous methods. Both classical and Baysian ap- URBANIK,T.: MOELLER M.P.; BARNES,K. Battelle Memorial in-proaches have been examined. This work was funded by the stitute, Pacific Northwest Laboratory. June 1988. 60pp. U S. Nuclear Regulatory Commission in support of its ongoing 9808080066. PNL-6171. 46408:349.

30 Main Citations and Abstracts This report compares observed vehicle movement on a high- mated from hydrologic / radionuclides transport models. This com-way network during periods of peak commuter traffic with a sim. parison has provided valuable insights into the applicability of ulation of the traffic flow made with the I-DYNEV computer transport modeling, and in determining what level of effort is model. The purpose of the comparison is to determine if the needed in site characterization at locations similar to the Nitrate model can accurately simulate the patterns of vehicular move- Disposal Pit to provide the desired degree of predictive capabili-ment and delay during congested enmmuter traffic. The results ties. indicate that the 1-DYNEV model adequately simulates the pat-

                                                                     " NUREG/CR-4880 V01: CHARACTER 12ATION OF IRRADIATED evacuaton, rovided tha he modefs apa i y d c io facto                 CURRENT-PRACTICE WELDS AND A533 GRADE B CLASS 1 is an input parameter. The current 1-DYNEV model automatically        PLATE FOR NUCLEAR PRESSURE VESSEL SERVICE.

reduces capacity by 15% of input capacity to account for con. MCGOWAN,J.J.; NANSTAD,R.K.; THOMS,K.R. Oak Ridge Na-gestion-induced losses in capacity. Because the study roadway tional Laboratory. August 1988.153pp. 8808300235. ORNL-did not have any capacity reduction due to congestion, the 6484.46632:150. model underestimated capacity during congestion. Therefore Studies of the effects of neutron irradiation on fracture tough-the use of a capacity reduction factor should be a decision ness properties of steels have generally included R minimum made by the analyst, not the model. When l-DYNEV was used number of tests for each material condition. The present . study with a capacity reduction factor appropriate to the data set used attempts to apply statistical analyses, with multiple testing at se- l (i.e., no reduction in capacity),1-DYNEV produced reasonable lected temperatures, to assess the accuracy and reliability of . results. the results. Fracture toughness test specimens were irradiated { in the Bulk Shielding Reactor at Oak Ridge National Laboratory i NUREG/CR-4874: THE SENSITIVITY OF EVACUATION TIME at 288 degrees C to target neutron fluences of 2 x 10(23) neu- i ESTIMATES TO CHANGES IN INPUT PARAMETERS FOR trons/m(2) (greater than 1 MeV). The materials were ASTM THE l-DYNEV COMPUTER CODE. URBANIK,T.; A533 grade B class 1 plate (HSST Plate 02) and four sub-MOELLER,M.P.; BARNES,K. Battelle Memorial Institute, Pacific merged-arc welds representing the current fabrication practice Northwest Laboratory. June 1988. 70pp. 8808080115. PNL- for nuclear pressure vessels. Both unirradiated and irradiated , 6172.46408:285- specimens were tested by two separate laboratories, and multi- i l ,A study performed by the Pacific Northwest Laboratory (PNL) pie tests were conduc'ed at selected temperatures. Statistical l with the assistance of the Texas Transportation Institute for the analyses permitted the determination of material and test varia-  ! U.S. Nuclear Regulatory Commission (NRC) identifies the key bility and an interlaboratory comparison. Behavior in both the input parameters to I-DYNEV affecting evacuation time esti- transition and ductile-shelf regions was studied. The results mates (ETEs). In addition, this study attempts to determine the demonstrated the relatively low radiation sensitivity of low sensitivity of ETEs to changes in those parameters when ap-plied to two different evacuation networks. This information copper / nickel welds and the qualitative agreement between could then be used to determine parameters requinng additional Charpy impact and fracture toughness observations of tough-research and to assist in the evaluation of ETEs submitted by ness degradation. licensees and applicants. The parameters analyzed for the NUREG/CR-4880 V02: CHARACTERIZATION OF IRRADIATED study included vehicle population, neturk capacity, loading CURRENT-PRACTICE WELLS AND A533 GRADE B CLASS 1  ; time, the capacity reduction factor, the time interval of process- PLATE FOR NUCLEAR PRESSURE VESSEL SERVICE. I ing, and free-flow velocity. These parameters were applied to MCGOWAN.J.J.; NANSTAD,R.K.; THOMS,K R. Oak Ridge Na-two evacuation networks simulating a rural and an urban area. tional Labordory. July 1988. 479pp. 8811010198. ORNL-6484. Changes in each of the six input parameters evaluated for this 47278:037, study affected to some degree the estimates of evacuation See NUREG/CR-4880,V01 abstract. times. In addition, an algorithm within the I DYNEV model was determined to route traffic in a potentially unreasonable manner NUREG/CR 4881: FISSION PRODUCT RELEASE CHARACTER- 1 in order to balance the system demand. In general, however, ISTICS INTO CONTAINMENT UNDER DESIGN BASIS AND l the results obtained revealed that the sensitivity o' the evacu- SEVERE ACCIDENT CONDITIONS. NOURBAKHSH,H.P.; ation time estimates to changes in the input parameters is con- KHATIB-RAHBAR; DAVIS,R.E. Brookhaven National Laboratory. sistent with traffic modeling theory and documented algonthms March 1988. 122pp. 8807070446. BNL-NUREG-52059, included in the model, 46013.284. I A new raMogeal ease eshmaks fm J NUREG/CR-4879 V01: DEMONSTRATION OF PERFORMANCE light water reactor accident sequences is presented as a basis { MODELING OF A LOW-LEVEL WASTE SHALLOW LAND f r development of a simplified approach for prediction of char-BURIAL SITE.A Companson Of Predictive Radionuclides Trans-port Modeling Versus Field Observations At The Nitrate Dispos- amnes d ra% cal rebans into mntahnts under al Pit Site, Chalk River Nuclear Labs. ROBERTSON.D.E ,? gn asis an s e accht mnMons fu M Ressun i ' BERGERON,M.P.; MYERS,D.A.; et al. Battelle Memorial Insti-ized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Resulting source term estimates are also compared tute, Pacific Northwest Laboratory. November 1987. 137pp. with parallel results using the Source Term Code Package 6805090112. PNL-6175. 45412:164. (STCP) methodology. Before a license can be obtained to construct a facility for the shallow-land bunal of low-level waste, the U.S. Nuclear Regula- NUREG/CR-4888: PRESSURIZED-THERMAL SHOCK TEST OF l tory Commission must be assured that the facilities will meet 6-INCH THICK PRESSURE VESSELS.PTSE-2: Investigation Of I both performance objectives and prescriptive requirements set Low Tearing Resistance And Warm Prestressing. BRYAN,R.H.; I forth in 10CFR61, " Licensing Requirements for Land Disposal BASS,B.R.; BOLT,S.E.; et al Oak Ridge National Laboratory, of Radioactive Waste " Subpart D, Section 61.50(a)(2) of December 1987. 306pp. 8804010275. ORNL-6377. 44949:179. 10CFR61 states that a " disposal site shall be capable of being The second pressurized-thermal-shock test of a 148-mm-thick characterized, modeled, analyzed and monitored." In order to steel pressure vessel with a 1-m-long flaw was performed to in-test the concept of " site modelability" a 30-year old low level vestigate fracture behavior of a vessel under conditions relevant radioa::tive waste disposat site at Chalk River Nuclear Laborato- to a flawed nuclear reactor pressure vessel dur ng an overcool- i ries (CRNL), Canada, was used as a field location for evaluating ing accident. The objectives were to observe . transitional crack I the process of site characterization and the subsequent model- behavior in a steel with low tearing resistance and the effects of ing predictions of radionuclides transport from the site by ground- warm prestressing on crack initiation. Tw. combinations of water. This evaluation was performed by companng the actual pressure and thermal transient conditions were imposed on the measured radionuclides migration with predicted migration esti- vessel with initial vessel temperatures of about 300 and 275 ce-

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l 1 i j

1 i Main Citations and Abstracts 31 goes C. The first . transient, designed for studying warm pres- tions for the hydrogen combustion model, if necessary, are up-tressing, required a depressurization from about 63 MPa and graded. In this report, we present HECTR's analyses of the repressurization to about 52 MPa. The crack propagated after large-scale premixed hydrogen combustion experiments at the being warm prestressed while K(1) > K(ic). Warm prestressing Nevada Test Site (NTS) and its comparison with the test re- j elevated the load at fracture significantly above K(Ic). The suits. The existing correlations in HECTR version 1.0, under cer- ' second transient produced a cleavage crack propagation and tain conditions, have difficulty in predicting accurately the com-arrest followed immediately by unstable tearing Stable ductile bustion completeness and burn time for the NTS experiments. teanng preceded both brittle fractures and also followed the first By combining the combustion data obtained from the NTS ex-crack arrest. The final cleavage propagation arrested at high- periments with other experimental data (FITS, VGES, ACUREX, K(l) levels (about 420 MPaWTn) in spite of the low ductile and Whiteshell), a set of new and better combustion correla-tearing resistance. tions were generated. HECTR prediction of the containment re-NUREG/CR-4898: RESULTS OF SEMISCALE MOD 2C FEED- sponses, using single-compartment model and EPRI-provided combustion completeness and burn time, compares reasonably WATER AND STEAM LINE BREAK (S-FS) EXPERIMENT well against the test results. However HECTR prediction of the SERIES. Bottom Main Feodwater Line Break Accident Experi-ments. BOUCHER,T.J. EG&G Idaho, Inc. (subs. of EG&G inc.). containment responses using a multicompartment model does February 1988.190pp. 8803150377, EGG-2503. 44691:218. not compare well with the test results. This discrepancy shows This report presents the results of four experiments simulating the deficiency of the homogeneous burning model used in 100,50 'nd 14.3% bottom mein feedwater line break accidents HECTR. To overcome this deficiency, a flame propagation perfnr aed at high pressure and temperature in the Semiscale model is highly recommended. Mod 4C facility. The primary and secondary thermal. hydraulic responses are characterized (including local secondary convec- NUREG/CR-4917: DCH-2:RESULTS FROM THE SECOND EX-tive heat transfer) and the influence of the break size on the re- PERIMENT PERFORMED IN THE SURTSEY DIRECT HEAT-sponses is discussed. A definite deficiency is identified in exist- ING TEST FACILITY. TARBELL,W.W.; NICHOLS,R.T/ ing forced convection boiling heat transfer correlations and the BROCKMANN.J.E.; et al. Sandia National Laboratories. January conservatism of FSAR heat transfer degradation assumptions is 1988.100pp. 8803030010. SAND 87-0976. 44558:165. shown to be questionable. The effectiveness of the recovery This test involved 80 k9 of molten core debns simulant e1'ect-operations in enaintaining control of the system is addressed, ed under pressure into a 1:10 linear scale model of a reactor and the system response (including local secondary convective cavity. The apparatus was placed in the Surtsey Direct Heating heat transfer) to voided secondary refill operations are dis- Test Facility to allow direct measurement of the temperature cussed. Feedwater line break issues are discussed and conclu. and pressure rise of the contained atmosphere. The molten ma-sions are drawn based on the results of the analysis. Finally, terial was ejected from the cavity as a dense cloud of particles recommendations are made for further utilization of the data and gas. The dispersed debns caused a rapid pressunzation of ano considerations for future code calculations. the 103 m(3) atmosphere. Peak pressures ranged from 0.22 to 0.31 MPa above the ambient level. Peak temperatures were NUREG/CR-4914: THE INFLUENCE OF SELECTED CONTAIN-from 759 degrees C to 1335 degrees C, with the highest values MENT STRUCTURES ON DEBRIS DISPERSAL AND TRANS-recorded near the top of the chamber. Much of the debris PORT FOLLOWING HIGH PRESSURE MELT EJECTION (about 70%) was found adhered to the top and sides of the FROM THE REACTOR VESSEL. PILCH M.: TARBELL,W.W.; BROCKMANN.J.E. Sandia National Laboratories. September steel chamber. The pattern of the retained matenal suggested 1988. 99pp. 8810280090. SAND 87 0940. 47248:006 that the debris field propagated around the chamber following . High pressure expulsion of molten core debris from the reac- the contour of the vessel Aerosol measurement indicated that for pressure vestel may result in dispersal of the debris from about 1% to about 6.6% of the ejected mass was in the size the reactor cavity. In most plants, the cavity exits into the con- range less than 10 micrometers aerodynamic diameter. tainment such that the debns impinges on structures. Retention of the debns on the structures may affect the further tran%: NUREG/CR-4918 V02: CONTROL OF WATER INFILTRATION of the debns throughout the containment. Two tests were done INTO NEAR SURFACE LLW DISPOSAL UNITS. Task Report - A with scaled structural shapes placed at the exit of 1:10 linear Discussion. SCHULZ,R.K. California, Univ, of, Berkeley, CA. scale models of the Zion cavity. The results show that the RIDKY,R.W. Maryland, Univ. of, College Park, MD. debns does not adhere significantly to structures. The lack of O'DONNELL E. NRC - No Detailed Affiliation Given. March i 1988. 33pp. 8804080272. 45066:311. ' retention is attributed to splashing from the surface and reen-trainment in the gas flowing nver the surface. These processes The principal pathway for water entry into LLW disposal units  ! are shown to be applicable to reactor scale. A third experiment n the humid eastern United States is through their covers. Two was done to simulate the annular gap between the reactor types of sub surface features that may be constructed to en-vessel and cavity wall. Debns collection showed that the frac- hance run-off (surface or sub-surface run-off) and thus reduce tion of debris exiting through the gap was greater than the gap- percolation are: (1) the " resistive layor" barrier, and (2) the i to. total flow area ratio. Film records indicate that dispersal was " conductive layer barrier" The " resistive layer" barrier is the l primarily by entrainment of the molten debris in the cavity. welkknown compacted rail or compacted clay layer and de-pends on compaction of permeable porous matenal to obtain NUREG/CR 4916: HECTR ANALYSES OF THE NEVADA TEST low flow rates. The " conductive layer" barrier is a special case SITE (NTS) PREMIXED COMBUSTION EXPERIMENTS. of the capillary barrier. Use is made of the capillary barrier phe-WONG.C.C. Sandia National Laboratories. November 1988. nomenon not only to increase the moisture content above an 176pp. 8812190322. SANDB7-0956. 47841:289. interface but to divert water away from the waste. During such ! The HECTR (Hydrogen Event: Containment Transient Re- diversion the water is at all times at negative capillary potential I sponse) computer code has been developed at Sandia National or under tension in the " flow layer". A very effective barrier l Laboratones to predct the transient pressure and temperature system might be constructed by placing a " resistive barrier" responses within reactor containments for hypothetical acci- over a " conductive barrier", A note of caution: such a system dents involving the transport and combustion of hydrogen. Ah must fail if appreciable subsidence takes place. An alternate though HECTR was designed primanly to investigate these phe- procedure called "bioengineenng management" utilizes features

                       .1omena in LWRs, it may also be used to analyze hydrogen               at the surface (as opposed to the subsurface) to ensure ade-transport and combustion experiments as well. It is in this            quate run-off. The engineered features are combined with manner that HECTR is assessed and empincal correlations,               stressed vegetation, that is, vegetation in an overdraft condition, such as the Combustion completeness and flame speed correla-           to control deep percolation.

32 Main Citations and Abstracts NUREG/CR-4920 V01: ASSESSMENT OF SEVERE ACCIDENT tainment loads were also identified. In addition, those features PREVENTION AND MITIGATION FEATURES:BWR, MARK i of a BWR Mark lil, which are important for preventing core CONTAINMENT DESIGN. PRATT,W.T.: ELTAWILA,F.; damage and are available for mitigating fission-product release PERKINS.K.R.; et al. Brookhaven National Laboratory. July to the environment were also identified. This report is issued to 1988.BBpp.8809060193. BNL-NUREG.52070. 46688:007. provide focris to the analyst examining an individual plant. This Plant features and operator actions, which have been found report calls attention to plant features and operator actions and to be important in either preventing or mitigating severe acci- provides a hst of deterministic attributes for assessing those dents in BWRs with Mark I containments (BWR Mark rs) have features and actions found to be helpful in reducing the overall been identified. These features and actions were oeveloped risk for Grand Gulf and other Mark 111 plants. Thus, the guidance from insights derived from reviews of in-depth risk assessments is offered as a resource in examining the subject plant to deter. performed specifically for the Peach Bottom plant and from as- mine if the same, or similar, plant features and operator actions sessment of other relevant studies. Accident sequences that will be of value in reducing overall plant risk. This report is in-dominate the core-damage frequency and those accident se- tended to serve solely as guidance. quences that are of potentially high consequence were identi-fied. Vulnerabilities of the BWR Mark i to severe accident con- NUREG/CR-4920 V04: ASSESSMENT OF SEVERE ACCIDENT tainment loads were also identified. In addition, those features PREVENTION AND MITIGATION FEATURES:PWR,LARGE of a BWR Mark I, which are important for preventing core DRY CONTAINMENT DESIGN. PERKINS,K.R.; HSU.C.J.; damage and are available for mitigating fission-product release LEHNER,J.R.; et al Brookhaven National Laboratory. July 1908. to the environment were also identified. This report is issued to 129pp.8809060157. BNL-NUREG-52070. 46688:093. provide focus to an analyst examining an individual plant. This Plant features and operator actions which have been found to report calls attention to plant features and operator actions and be important in either preventing or mitigating severe accidents provides a list of deterministic attributes for assessing those in PWRs with large dry containments have been identified. features and actions found to be helpful in reducing the overall These features and actions were developed from insights de-nsk for Peach Bottom and other Mark I plants. Thus, the guid- rived frorn reviews of risk assessments performed specifically ance is offered as a resource in examining the subject plant to for the Zion plant and from assessments of other relevant stud-determine if the same, or similar, plant features and operator ies. Accident sequences that dominate the core-damage fre-actions will be of value in reducing overall plant risk. This report quency and those accident sequences that are of potentially is intended to serve solely as guidance. high consequence ' vere identified. Vulnerabilities of the large NUREG/CR-4920 V02: ASSESSMENT OF SEVERE ACCIDENT dry containment w severe accident containment loads were PREVENTION AND MITIGATION FEATURES.BWR, MARK ll also identified. In addition, those features of a PWR with a large CONTAINMENT DESIGN. LEHNER,J.R.; HSU,C.J.; dry containment, which are important for preventing core ELTAWILA,F.; et al. Brookhaven National Laboratory. July 1988. damage and are available for mitigating fission-product release 108pp.8809060181. BNL-NUREG 52070. 46687:001, to the environment were identified This report is issued to pro-Plant features and operator actions, which have been found vide focus to the analyst examining an individual plant. This to be important in either preventing or mitigating severe acci. repc't calls attention to plant features and operator actions and dents in BWRs with Mark 11 containments (BWR Mark ll's) have provides a list of deterministic attributes for assessing those been identified. These features and actions were developed features and actions found to be helpful in reducing the overall from insights denved from reviews of in-depth risk assessments risk for Zion and other PWRs with large dry containments. Thus, performed specifically for the Limenck and Shoreham plants the guidance is offered as a resource in examining the subject and from other relevant studies. Accident sequences that domi- Plant to determine if the same, or similar, plant features and op-nate the core damage frequency and those accident sequences erator actions will be of value in reducing overall plant risk. This that are of potentially high consequence were identified. Vulner, report is intended to serve solely as guidance. abilities of the BWR Mark 11 to severe-accident containment loads were also noted. In addition, those features of a BWR NUREG/CR-4920 VOS: ASSESSMENT OF SEVERE ACCIDENT PREVENTION AND MITIGATION TEATURES:PWR,1CE-CON-Mark,11 which are important for preventing core damage and DENSER CONTAINMENT DESIGN. HSU.C.J.; PERKINS,K.Rc l are available for mitigating fission-product release to the envi" i ronment were also identified. This report is issued to provide LUCKAS,W.J.; et al. Brookhaven National Laboratory. July 1988.138pp.8809060147. BNL-NUREG 52070. 46688:222. focus to an analyst examining sq individual plant. This report Plant features and operator actions which have been found to calls attention to plant features and operator actions and pro-vides a list of deterrr.inistic attnbutes for assessing those fea-be important in either preventing and mitigating severe acci-tures and actions found to be helpful in reducing the overall risk dents in PWRs with ice condenser containments have been identified. Thus features and actions were developed from in-for Mark 11 plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, s ghts derived from reviews of risk assessments performed spe-cifically for the Sequoyah plant and from assessments of other plant features and operator actions will be of value in reducing relevant studies. Accident sequences that dominate the core-overall plant risk. This report is intended to serve solely as guid-damage frequency ar.d those accident sequences that are of potentially high consequence were identified. Vulnerabilities of NUREG/CR-4920 VD3: ASSESSMENT OF SEVERE ACCIDENT the ice-condenser containment to severe accident containment PREVENTION AND MITIGATION FEATURES:BWR, MARK lli loads were also identified. In addition, those features of a PWR CONTAINMENT DESIGN. FITZPATRICK,R.; LEHNER,J.R.; with an ice <ondenser containment, which are important for pre. ELTAWILA.F.; et al. Brookhaven National Laboratory. July 1388. venting core damage and are available for mitigating fission-77pp.8809060174. BNL-NUREG-52070. 46687:109. product release to the environment were identified. This report . Plant features and operator actions, which have been found is issued to provide focus to an analyst examining an individual to be important in either preventing or mitigating severe acci- plant. The report calls attention to plant features and operator dents, in BWRs with Mark ll1 containments (BWR Mark lil's), actions and provides a list of deterministic attributes for assess-have been identified. These features and actions were devel- ing those features and actions found to be helpful in reducing oped from insights derived from reviews of in-depth risk assess- the overall risk of Sequoyah and other PWRs with ice-condens-ments performed specifically for the Grand Gull plant and from er containments. Thus, the guidance is offered as a resource in assessments of other relevant studies. Accident sequences that examining the subject plant to determine if the same, o similar, i dominate the core damage frequency and those accident so- plant features and operator actions will be of value in weing quences that are of potentially high consequence were identi- overall plant nsk. This report is intended to serve solely as guid-feed. Vulnerabilhies of the BWR Mark 111 to severe accident con- ance.

Main Citations and Abstracts 33 NUREG/CR-4924: SEISMIC CATEGORY l STRUCTURES Lucie 1, the licensee elected to continue to use the subject PROGRAM. Final Report. Fiscal Year 1983 - 1984. DOVE R.C.; relays and presented acceptable justification. BENNETT,J G.; FARRAR.C.; et al. Los Alamos National Labora-tory. September 1987, 73pp. 8802240287. LA-11013-MS. NUREG/CR-4935: CLOSEOUT OF IE BULLETIN 85 02:UNDER-  ! 44469:176. VOLTAGE TRIP ATTACHMENTS OF WESTINGHOUSE DB-50 This report summanzes the results obtained from a series of TYPE REACTOR TRIP BREAKERS. FOLEY,W.J.; DEAN R.S.; simulated seismic tests on scale models of a prototypical Cate- HENNICK A. Parameter, Inc. September 1988. 36pp. gory i nuclear power plant auxiliary building, representing a rein- 8810050250. P/,RAMETER IE170. 46988:271, forced concrete, diesel generator building. Two sizes of model Documentation is provided in this report to close out IE Bulle-structures were used: 1/10 scale and 1/30 scale. Model con- tin 85 02 on the subject of undervoltage trip attachments struction, test methods, instrumentation, data reduction tech-niques, experimental results, companson of experimental and (UVTAs) of Westinghouse DB 50 Type reactor trip breakers computed results, and conclusions are presented in this report. (RTBs). The bulletin was issued to require the owners of Wes. Values of structural stiffness obtained from both static and dy- tinghouse operating power reactors to provide assurance that namic tests are found to be significantly lower than values of their DB-50 Type RTBs with UVTAs were operating properly

• stiffness computed using the usual design methods. Values of modal frequency obtained from dynamic tests are compared tion. Review of utility responses and NRC documents shows with computed values. Decreasing modal frequencies with in. that the concern about unmodified RTBs applied to eight facili-creasing seismic input are reported. The effective damping of ties. Evaluation of utility responses and NRC/ Region inspection these test structures is determined from the test results. The re. reports shows that reliability of UVTAs at these eight facilities is suits obtained from the two different size (1/10- and 1/30- ensured by means of procedures, tests and schedules. Further-scale) models are compared. more, the NRC regions verify that the recommended automatic NUREG/CR-4932: CLOSEOUT OF IE BULLETIN 80-03: LOSS OF shunt trip attachment has been or will be implemented at these CHARCOAL FROM STANDARD TYPE ll,TWO-INCH, TRAY AD- remaining eight facilities. Background information is supplied in SORBER CELLS. DEAN,R.S.; FOLEY,W.J.; HENNICK A. Param, the introduction and Appendix A of this report.

eter nc. April 1988. 28pp. 8805200093. PARAMETER IE167. NUREG/CR-4939 V01: IMPROVING MOTOR RELIABILITY IN Because of concem about defective charcoal tray adsorber NUCLEAR POWER PLANTS. Volume 1. Performance Evaluation cells found in certain ventilation systems at the Sequoyah Nu. And Maintenance Practices. SUBUDHi,M.; GUNTHER,W.E.; clear Plant, the NRC/IE issued IE Bulletin 80-03 on February 6, TAYLOR,J.H.; et al. Brookhaven National Laboratory. November 1980. Some charcoal cells are used in ventilation systems asso. 1987.153pp.8802030212. BNL-NUREG-52031. 44242:001. ciated with engineered safety features, which are provided for This report constitutes the first of the three volumes under protection from abnormal events. Others are installed to control this NUREG. The report presents recommendations for devel-radioactive materials during expected operations. Licensees of oping a cost- effective program for performance evaluation and operating power reactors and holders of permits for those under maintenance of electric motors in nuclear power plants. These construction were required to take specific actions. Evaluation recommendations are based on current industry practices, avail-  ! of utility responses and NRC/ Region inspection reports shows that the bulletin can be closed out by means of specific criteria able techniques for monitoring degradation in motor compo-for 123 (99%) of the 124 facilities with operating licenses or nents, manufacturer's recommendations, operating experience, construction permits. A followup item is proposed for the only and results from two laboratory tests on aged motors. Two lab-facility with open bulletin status, for use by the NRC in ensuring oratory test repcrts on a small and a large motor are presented satisfactory completion of corrective action. The cells with rivet- in separate volumes of this NUREG. These provide the basis ed screens which were identified at Sequoyah were not found for the various functional indicators recommended for mainte-at any other facility. Although cells with miscellaneous defects nance programs in this report. Implementation of the informa-were found at nine facihties other than Sequoyah, there were no tion provided in this report, will improve motor reliability in nucle-l Charcoal problems. ar power plants. The study indicates the kinds of tests to con. duct how and when to conduct them, and to which motors the NUREG/CR-4933: CLOSEOUT OF IE BULLETIN 80-19. FAIL- tests should be applied. It should be noted that the recommen. URES OF MERCURY-WETTED MATRIX RELAYS IN REAC-dations and conclusions provided in this report are based on re. TOR PROTECTIVE SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED BY COMBUSTION ENGINEER- search findings, and as such should not be co.nstrued as re9ula. ING. FOLEY,W.J.: DEAN,R.S.; HENNICK,A. Parameter, Inc. Wy or staMory mqWemems for motors in nuclear power April 1988. 26pp. 8805090125. PARAMETER IE168. 45412:301. #ams. The NRC/IE issued IE Bulletin 80-19 initially on July 31,1980, and issued Revision 1 of the bulletin on August 13,1980.The NUREG/CR-4939 V02: IMPROVING MOTOR RELIABILITY IN bulletin was issued to all licensees and holders of construction NUCLEAR POWER PLANTS. Volume 2. Functional indicator permits of power reactors, because of numerous reports about Tests On A Small Electric Motor Subjected To Accelerated single and multiple failures of C.P. Clare Model HG2X 1011 Aging. SUBUDHl M.; TAYLOR,J.H.; LOFARO,R.: et al Brookha-mercury wetted matrix relays in reactor protective systems. The ven National Laboratory. November 1987. 78pp. 8802030222. concern based on those reports was the build up of coincident BNL-NUREG-52031. 44241:050. {

  " failed closed" tailures of certain sets of relays could result in                  A ten horsepower electric motor was artificially aged by plug             i inp failures for off-normal events. Evaluation of utility responses             reverse cycling for test purposes. The motor was manufactured and NRC/ Region inspection reports shows that the bulletin can                  in 1967 and was in service at a commercial nuclear power plant                t be closed out per specific cnteria for 100% of the 123 power                    for twelve years. Various tests were performed on the motor                   !

facilities with operating licenses or construction permits. In throughout the aging process. The motor failed after 3.79 million effect, all of the responses and inspection reports apply to Revi- reversals ( 3 seconds per reversal) over seven months of test. l sion 1 of tne bulletin. All except the three following facilities ing. Each test parameter was trended to assess its suitability in  ; l either do not have the subject relays in the reactor protective monitoring aging and service wear degradation in motors. Re-i system, or have changed the relays to an acceptable type as a suits and conclusions are discussed relative to the applicability I result of the bulletin. An inspection report closed the bulletin for of the tests performed to nuclear power plant motor mainte. Calvert Cliffs 1 & 2 based on licensee commitments to replace nance programs. the mercury wetted relays with the dry-contact type. For St. c i

34 Main Citations and Abstracts NUREG/CR-4939 V03: IMPROVING MOTOR RELIABILITY IN improvements to control room habitability systems are provided NUCLEAR POWER PLANTS. Volume 3. Failure Analysis And Di- in the report. agnostic Tests On A Naturally Aged Large Electric Motor. SUBUDHl.M.; TAYLOR J.H. Brookhaven National Laboratory. NUREG/CR-4971: RESULW OF SEMISCALE MOD-2C FEED-SHEETS M.W. NUTECH Engineers, Inc. November 1987. 52pp. WATER AND STEAM LINE BREAK (S-FS) EXPERIMENT 8802030137. BNL-NUREG-52031. 44241:001, SERIES. Main Steam Line Break Accident Expenments. Stator coils of a naturally failed 400 hp motor from the Brook. BOUCHER,T.J. EG8G Idaho, Inc. (subs. of EG&G, Inc.). March haven National Laboratory test reactor facility were tested for 1988.112pp. 8804280505. EGG-2516. 45274:302. their dielectric integnties. The motor was used to drive the pri- This report presents the results of two experiments conduct-mary reactor coolant pump for the last 20 years. Maintenance ed in the Semiscale Mod-2C facility which simulated main steam activities en this motor during its entire service life were mini- line break accidents at high pressure and temperature. Tests S-mal, with the exceptinn of meggenng it periodically. The stator FS-1 and S-FS-2 simulated double-ended offset shears of the consisted of ninety individual coils which were separated for main steam line downstreatn and upstream, respectively, of the testing. Seven different dielectric tests were performed on the flow restrictor. Initial and boundary conditions were scaled from, coils indicated a spectrum of variation depending on their aging and compounding failures and assumptions simulated, those conditions and characteristic 6. By companng the test data to conditions utilized for Final Safety Analysis Report (FSAR) cal-baseline data, the test methods were assessed for application culations. Primary and secondary thermal- bydraulic responses i are characterized, and the influence of the break size or loca-to motor maintenance programs in nuclear power plants. Also included in this study are results of an investigation to deter. tion on the responses is discussed. The limiting of primary-to-  ; mine the cause of this motor failure. Recommendations are pro- secondary heat transfer by conduction heat transfer is shown to I vided on the aged condition of a second identical primary pump Produce a trend of increased pnmary cooling with decreased , motor which is the same age and presentiy in operation. Rec- break size, pointing out the need for further analysis for smaller l commendations are also presented relating to each of the dielec- break sizes. The degree of conservatism inherent in FSAR sep-arator performance and break size and location assumptions is

                                                                                                                                        ~

tric test methods applicable to motor maintenance shown to be questionable, and the FSAR assumption of a loss NUREG/CR-4947: ANALYSIS OF THE A3028 AND A533B of offsite power is shown to be nonconservative. The effective-STANDARD REFERENCE MATERIALS IN SURVEILLANCE ness of the recovery operations in regaining and maintaining i CAPSULES OF COMMERCIAL POWER REACTORS. control of the system is addressed; and main steam line break l STALLMANN,F.W. Oak Ridge National Laboratory. January issues are discussed. Finally, conclusions are d. awn and recom- I 1988.128pp.8803140060. ORNL/TM-10459. 44673:013. mendations are made for further utilization of the data. l The results of Charpy tests for A302B and A533B Standard a Reference Materials in surveillance capsules of commercial NUREG/CR-4984: DEVELOPMENT OF A THREE-DIMENSIONAL j power reactors are analy:ad. The NDT shifts at 30 ft-lb (41 J) FLUX SYNTHESIS PROGRAM AND COMPARISON WITH 3-D q generally follow the predictions provided by the draft Revision 2 TRANSPORT THEORY RESULTS. KAM.F.B. Oak Ridge Nation- 1 of Reg. Guide 1.99. Deviations from the predicted values are al Laboratory. CHOWDHURY,P.; WILLIAMS M.L Louisiana { correlated to similar deviations for the accompanying materials State Univ., Baton Rouge, LA. January 1988. 102pp. j in the same capsules, but large random fluctuations prevent 8002240224. ORNb'TM-10503. 44471:057. i precise quantitative determination. Difference in capsule tem. A threa-d mensional (3-D) flux synthesis program called l peratures is a likely explanation for deviations from Reg. Guide DOTSYN has been developed to estimate neutron fluence in l 1.99 predictions, firadiations of the same materials ir. the ORR. pressure vessels of light water reactors. The synthesis uses PSF experiment show similar results, although somewhat larger lower dimensional flux values obtained from 2-D and 1-D trans-than predicterf embrittlement in the high-fluence, high-fluence port theory calculations to approximate the desired 3-D fluxes. rate SSC2 capsule, which appears to be a fluence-rate effect. In order to verify the accuracy and efficiency of the method, a Shifts in opper-shelf energy and other measures of embrittle- simulated pressurized water reactor is modeled in 3-D cylindri-ment are not considered in this report. cal coordinates, and calculations are performed with the three-dimensional transport code TORT as well as with DOTSYN. Re-NUREG/CR-4960: CONTROL ROOM HABITABILITY SURVEY OF suits by TORT are used as a 3-D " benchmark" solution, and LICENSED COMMERCIAL NUCLEAR POWER GENERATING synthesis results at varioub locations are compared to this refer-STATIONS. DRISCOLL.J.W. Argonne National Laboratory. Oc- ence solution. The enor between TORT and DOTSYN results at tober 1988. 262pp. 8812010073. ANL-87-22. 47683:132. the peak flux locations (theta = 9 degrees) at the inner curface This document presents the results of a survey of control of the prassure vessel varies between 0.35% at the midplane room habitability systems at twelve couercial nuclear generat- and 4.73% at the bottom of the nozzle. The worst agreement ing stations. The survey, conducted by Argonne National Lab > was found to be about 19.8% at the bottom of the nozzle at an ratory (ANL) is part of an NRC program initiated in response to azimuth of about 39.5 degrees. It is concluded that the 3-D syn-concerns and recommendations of the Advisory Committee on thesis method is an efficient method for estimating reactor pres-Reactor Safeguards. The major conclusion of the report is that sure vessel iluence, which gives acceptable accuracy within the the numerous types of potentially significant discrepancies vessel even at axial locations far above the active core height. found among the surveyed plants may be indicative of similar discrepancies throughout the industry. The report provides NUREG/CR-4991: EVALUATION AND PROPOSED IMPROVE-plant-specific and generalized findings regarding safety func- MENTS TO EFFECTIVENESS OF U.S. NUCLEAR REGULA-tions with respect to the consistency of the design, construc- TORY COMMISSION GENERIC COMMUNICATIONS. tion, operation and testing of control room habitability systems THURBER,J.A.: MELBER.B.D.; GEISENDORFER,C.; et af. Bat-and corresponding Technical Specifications compared with de- telle Human Affairs Research Centers. January 1988, 66pp. senptions provided in the keense basis documentation (licensee 8802030152. PNL-6289. 44243:292. NUREG 0737 Item Ill D.3 4 submittals and updated Safety Anal- This report describes an evaluation of NRC generic communi-ysis Reports, and NRC Safety Evaluation Reports) including as- cations with iridustry about nuclear power plant operating expe-sumptions in the operator toxic gas concentrations & radiation rience. The analysis builds on the findings presented in the dose calculations. Calculations of operator toxic gas concentra- 1986 AEOD Special Study Report, "An Overview of Nuclear tions and radiation doses required by 10 CFR 50 Appendix A, Power Plant Operating Experience Feedback Programs" General Design Cntenon 19, Control Room, used in the survey (AEOD/S602). The pnmary objective of the report is to present to determine the adequacy of occupant protection were provid- practical recommendations for improving NRC's documents and ed in the license basis documentation. Recommendations for generic communications system. The report is based upon a

l Main Citations and Abstracts 35 systematic review and evaluation of NRC and industry operating K(a) values obtained using the proposed test method and those experience documents and an analysis of interviews with licens- obtained from large-scale tests. ee personnel at five utilities and their nuclear power plants and NRC regional and headquarters managers and staff. NRC and NUREG/CR-4997: METHODS FOR DESCRIBING AIRBORNE ticonsee personnel interviewed are generally satisfied with the FRACTIONS OF FREE FALL SPILLS OF POWDERS AND LlO-current NRC- industry communications system; however, sever- UlDS. BALLINGER.M.Y.: OWCZARSKI.P.C.; BUCK,J.W.; et al. al problems and solutions to those problems wore discovered. Battelle Memonal Institute, Pacific Neithwest Laboratory. Janu-The report makes seven recommendations for improvement in ary 1988. 50pp. 8803310288. PNL-6300. 44918:201. the effectiveness of NRC-industry generic communications This report desenbes methods for estimating the fractions of about nuclear power plant operating expen oce. liquid and powder that produce aerosols in hypothetical spills. The charactenstics of these aerosols are also described. The NUREG/CR-4992 VD1: AGING AND SERVICE WEAR OF MUL- methods developed were based on previous data on the free TISTAGE SWITCHES USED IN SAFETY SYSTEMS OF NU- fall spills in static air of aqueous solutions of uranyl nitrate, CLEAR POWER PLANTS. Operating Experience And Failure uranine, sucrose solutions and slurries, and titanium dioxide Identification. ROBERTS,G.C.; BACANSKAS,V.P.; TOMAN,G.J.; powders (Sutter et al.1981, Ballinger and Hodgson 1986). After et al. Oak Ridge National Laboratory. September 1987. 63pp. the liquid spill data were corrected for evaporation and settling,  ; 8802240217. ORNLSUB83289155. 44472:291. regression anafysis of the data was produced correlating the  ; An assessment of the types and uses of multistage switches fraction of a spill make airborne with dimensionless numbers 1 in nuclear power plant safety related service is provided. containing the following parameters: spill height, liquid viscosity, Through a desenption of the operation of each type of switch, liquid density, air density, and volume of liquid spilled. Statistical j combined with knowledge of nuclear power plant applications descriptions of the log normal distributions were provided for l and operational occurrences, the significant stressors responsi- parameters of liquid aerosols. Whereas liquid aerosols formed ' ble for multistage switch deterioration are identified. A review of on impact, powder aerosols formed as the powder fell. A dimen-operating exporience (failure data) leads to identification of po- sional analysis of the powder that dispersed while falling corre-tential monitonng techniques for early detection of incipient fail- lated well with the Galileo number, Ga, a dimensionless number ures. Although the operating experience does not justify exten. containing the parameters mass of powder spilled, powder bulk sive deterioration monitoring of multistage switches, nondestruc- density, air density, and air viscosity. A computer code was de-tive testing methods that could be used to evaluate the condi- veloped to pred;ct the amount airborne at specified time steps tion of switches are identified. after a powder spill. NUREG/CR-4993: A STANDARD PROBLEM FOR HECTR-MAAP NUREG/CR-4998: THE SEISMIC CATEGORY I STRUCTURES COMPARISON. Incomplete Burning. WONG,C.C. Sandia Nation- PROGRAM.Results For Fiscal Year 1985. BENNETT,J.G.; al Laboratories. August 1988.118pp. 8809220049. SAND 87- DOVE,R.C.; DUNWOODY,W.E. et al. Los Alamos National Lab-1858.46879:337. oratory. December 1987, 46pp. 8802240234. LA 11117 MS. To assist in the resolution of differences between the NRC 44469:031. and IDCOR on the hydrogen combustion issue, a standard in FY 1985 a new effort was begun to resolve an issue that problem has been defined to compare the results of HECTR became the " stiffness difference issue." This issue came about and MAAP analyses of hydrogen transport and combustion in a from reporting the results from testing both isolated shear walls nuclear reactor containment. The first part of this standard prob- and box-like shear deformation-dominated scale models that lem, which addresses incomplete burning of hydrogen in the showed a consistent reduction in structural stiffness measured lower and upper compartments, has been completed, and the experimentally. This structural stiffness was different trorn that results will be presented in this report. A critical review and which would be calculated analytically at loads associated with comparison of the combustion models in HECTR and in MAAP operating basis earthquake levels. Several possible explana-show that HECTR's predictions are better than MAAP's when tions were proposed for the experimental / analytical difference compared against test results of the VGES, FITS and NTS ex. (most likely attributable to cracking of the concrete models). penments. The model in MAAP overpredicts the burn time and Nssibilities are microcracking at very low loads, microconcrete underpredicts the steam inerting effect on ignition. For the effects, such as shrinkage cracks, unaccounted for dynamic ef-standard problem, HECTR predicts that pressure generated due fects in the analyses, and low stress level, low-cycle, fatigue to incomplete burning in the lo ver and upper compartments will degradation of microconcrete properties. A new configuration have a sharper nse, shorter duration and higher peak value was proposed by the Technical Review Group (TRG) for this than that predicted by MAAP. MAAP prediction resembles a program and, in FY 1985, a prototyoe structure was designed. A standing diffusion flame, rather than a flame propagating 1/4-scale microconcrete model of the prototype structure was , through a homogeneous mixture. constructed and tested. This report details that investigation, but I it does not report the resolution of the " stiffness difference" NUREG/CR-4996: A REPORT ON THE ROUND ROBIN PRO- issue, an investigation that was ongoing at the end of FY 1985. GRAM CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD TEST METHOD FOR DET6RMINING THE PLANE NUREG/CR-4999: ESTIMATION OF RISK REDUCTION FROM l STRAIN CRACK ARREST FRACTUPE TOUGHNESS,K(IA).OF IMPROVED PORV RELIABILITY IN PWRS. Final Report. FERRITIC MATERIALS. BARKER.D B.; CHONA,R.; HSU.C.; PERKINS.K.; YOUNGBLOOD,R. Brookhaven National FOURNEY,W.L.; et al. Oak Ridge National Laboratory. January Laboratory. March 1988. 82pp. 8804080267. BNL-NUREG-1998.193pp.8803140054. ORNLSUB797778/4. 44667:159. 52101. 45066:344. < The experimental results from a Round Robin Program that An analysis was performed to explore the risk potential of im- I was conducted dunng 1983-1985 to evaluate a proposed ASTM proving the pressunzer PORV and block valve reliability for two Test Method for Crack Arrest Fracture Toughness of Ferritic representative PWR plants, Indian Point 3, and Oconee 3. At-Materials are summarized. Of a total of twenty-seven partici- tention was focused upon particular transient scenarios includ-pants from the United States, Canada, Europe and Japan., ang steam generator tube rupture (SGTR), a stuck-open PORV, twenty-one completed their testing and forwarded results for in- cooldown to cold shutdown, and the use of PORV in low-tem-clusion and discussion in this report. Each organization tested perature overpressunzation (LTOP) protection. The feasibility of three specimens of A514 bndge steel, three specimens of A588 using the PORV as a high point vent to supplement the function bridge steel, and six specimens of A533 grade B class 1 steel. of Reactor Vessel Head Vent System (RVHVS) was also stud-ACalyses of the data show that the procedure can provide re- ied. The pertinent event trees, fault trees, and the basic data i spctable results and that good agreement occurs between presented in the indian Point Prow.cilistic Safety Study (IPPSS) ' l I ___

36 Main Citations and Abstracts and the Oconce PRA were utilized to quantify the benefit of an are developed for combining exponence data into composite improved PORV and block valve reliability in terms of potential specifications for qualification of equipment that can be shown reduct6on in cora-melt frequencies. For indian Point 3, independ- to be physically similar to the reference equipment. Other pro-ent core-melt frequency calculations were made based on the cedures are developed for extending qualifications beyond the Boolean expressions derived for various plant damage states. original specifications under certain conditions. Some examples l For Oconee 3, the components of the dominant cut sets and for application of the procedures and venfication of them are the detailed fault trees shown in the Oconee PRA were given given for certain cases that can be approximated by a two thorough scrutiny to determine their relevance to the hardware degree of freedom simple primary / secondary system. Other ex-or operational failures of the PORV and its block valve. With the amples are based on use of actual test data available from pre-exception of LTOP, the core melt frequencies attributable to vious qualifications. Relationships of the developments with PORV or block valve failures were found to be relatively insignif- other previously-published methods are discussed. The develop-icant, only a very small fraction of the total core-melt frequency ments are intended to elaborate on the rather broad revised due to intemal events. For the case of LTOP, the core me!t fre-guidelines developed by the IEEE 344 Standards Committee for quency and associated risk appear to be srnali for Indian Point equipment qualificotton in new nuclear plants. Howeveri the re- i 3 and Oconee 3 since the vessels have not hr.d a substantial sults also contribute to filling a gap that exists between the I fraction of their estimated lifetime irradiation. The results of a conservative estimation of health effect, however, indicate that IEEE 344 methodology and that previously developed by the the public nsk from LTOP events may become more significant late in plant life when the aging effects of the vessel contribute suits to safety margin technology is also discussed. to increased vulnerability. These effects are being studied in more depth in the NRC's studies for Generic Issue Number 94. NUREG/CR-5013: FATIGUE LIFE CHARACTERIZATION OF SMOOTH AND NOTCHE D PIPING STEEL SPECIMENS IN 288 NUREG/CR-5000: METHODOLOGY FOR UNCERTAINTY ESTi- DEGREES C AIR ENMONMENTS. TERRELL,J.B. Materials MATION IN NUREG 1150 (DRAFT). Conclusions Of A Review Engineenng Associates, Inc. May 1988. 77pp. 8805200115. Panel. KOUTS H.; CORNELL,A.; FARMER,R.; et al. Brookhaven MEA-2232. 4556t256. National Laboratory. December 1987. 23pp. 6806230139. BNL' Fatigue strain-life tests were conducted on ASME SA 106-B piping steel at 24 degrees C (76 degrees F) and at PWR oper-A re i been und rtaken by a panel of experts, of the methodology for estimation of uncertainty in severe accident ating temperature,288 degrees C (550 degrees F), under com-nsk resutting from accidents to nuclear power plants as present- pletely reversed loading. Smooth specimens were tested at ed in the Draft NUREG 1150 report. This report provides de- both temperatures whereas spe:imens containing notches of tailed discussions and conclusions resulting from this review vanous aculties were tested af 288 degrees C. Fatigue limits at i process. 10(7) cW.ies were estimated to be 185 MPa (26.8 ksi) et 24 de-l grees C and 232 MPa (33.7 ksi) at 268 degrees C. The differ-l NUREG/CR-5009: ASSESSMENT OF THE USE OF EXTENDED ence in fatigue strength observed at the PWR temperature is BURNUP FUEL IN LIGHT WATER POWER REACTORS. postulated to be due to dynamic strain aging processes. Howev-BAKER.D.A.; BAILEY,W.J.; BEYER C.E.; et al. Battelle Memonal er, there is a reduction in low cycle fatigue strength at this tem-Institute, Pacific Northwest Laboratory. February 1988. 06pp. 8802240276. PNL-6258. 44469:080. perature which results in a decrease in the intended safety This study has been conducted by Pacific Northwest Labora-factor of the ASME Section lit design curve for carbon steels. tory for the U.S. Nuclear Regulatory Commission to review the Notch strain histories were estimated for the notched specimen environmental and economic impacts associated with the use of 9 'E extended burnup nuclear fuel in light water power reactors. It ciuded that the use of the fatigue notch concentration factor has been proposed that current batch everage burnup levels of (Kf) in the Neuber relation in conjunction with the untaxial cyclic 33 GWd/t uranium be increased to above 50 GWd/t. The envi- stress-strain curve provided the best correlation of notched ronmental effects of extending fuel bumup dunng normal oper- specimen fatigue data with results obtained from smooth speci-ations and dunng accident events and the economic effects of men tests. The notched specimen strain-life results derived cost changes on the fuel cycle are discussed in this report The from the application of Neuber's rule alone proved to be con-physical effects of extended burnup on the fuel and the fuel as- servative when compared to smooth specimen test results to sembly are also presented as a basis for the environmental and such an extent Nat Neuber generated notch stress and strain economic assessments. Environmentally, this burnup increase amplitudes cannot be compared to the ASME Section lit fatigue would have no significant impact over that of normal burnup. curves for carbon steels. Economically, the increased burnup would have favorable ef- 1 fects, consisting pnmarily of a reduction in 1) total fuel require- NUREG/CR 5015: IMPROVED RELIABILITY OF RESIDUAL ments,2) reactor downtime for fuel replacement,3) the number HEAT REMOVAL CAPABILITY IN PWRS AS RELATED TO of fuel shipments to and from reactor sites, and 4) repository RESOLUTION OF GENERIC ISSUE 99. CHU,T.L; , storage requirements. FITZPATRICK,R.; YOON,W.H.; et al Brookhaven National Lab-  ! NUREG/CR 5012: SIMILARITY PRINCIPLES FOR EQUIPMENT oratory. May 1988.161pp. 8806070054. BNL-NUREG-52121. QUALIFICATION BY EXPERIENCE. KAN A,D.D.; 45727:184. POMERENING,0.J. EG&G Idaho, Inc. (subs. of EG&G Inc4

• This report summarizes a study performed by Brookhaven 1 Southwest Research institute. July 1988.147pp. 8808120173. National Laboratory for the Reactor and Plant Safety issues l EGG 2521. 46465;237. Branch, RES cf the U S. Nuclear Regulatory Commission in pur-A methodology is developed for seismic qualification of nucle- suit of the resolution of NRC Generic issue 99. Generic Issue

                                                                                                                                                                                 ]

1 ar plant equipment by applying similanty pnnciples to exis'ang 99 focuses on the risk associated with loss of residual heat re- j j exponence data. Experience data are available from previous moval events at PWRs while shut down. Numerous loss of re- < i qualifications by analysis or testing, or from actual earthquake sidual heat removal events have occurred at pressurized water j events. Similarity pnnciples are defined in terms of excitation, reactors (PWRs) in the USA, which were terminated prior to equipment physical characteristics, and equipment response. damaging the reactor core. This study estimates the nsk from j Physical similanty is further defined in terms of a entical transfer loss of residual heat removal events and investigates ways of function for response at a location on a primary structure. lowenng this risk. l whose response can be assumed directly related to ultimate fra-gility of the item under elevated levels of excitation. Procedures I i i _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __)

Main Citations and Abstracts 37 y i NUREG/CR-5016: COMPENDIUM AND COMPARISON OF NUREGICR-5020; A

SUMMARY

OF ENVIRONMENTALLY AS-INTERNATIONAL PRACTICE FOR PLUGGING, REPAIR AND SISTED CRACK-GROWTH STUDIES PERFORMED AT WES-INSPECTION OF STEAM GENERATOR TUBING. CLARK,R.A.; TINGHOUSE ELECTRIC CORPORATION.Under Funding From XURTZ.R.J. Battelle Memonal Institute, Pacific Northwest Labo- The Heavy-Section Steel Technology Program. BAMFORD,W.H. ratory. April 1988. 71pp. 8805090114. PNL-6341. 45432.233. Oak Ridge National Laboratory. BAMFORD W.H. Westinghouse l The Committee on the Safety of Nuclear installations (CSNI) Electnc Corp. May 1988. 271pp. 8808050237. ORNL- .! of the Organization for Economic Cooperation and Development SUB82215981. 46396:012. - Nuclear Energy Agency (OECD-NEA) is an international body This report presents a collection of the information developed , of scientists and engineers with responsibilities for nuclear in the crack-growth studies at Westinghouse Electric Corpora-  ! safety research and nuclear licensing. The CSNI fosters interna- l'on from 1969 to 1986 under a subcontract with the Heavy- ' tional cooperation in nuclear safety amongst OECD member Section Steel Technology Program. Although progress reports countnes. In July 1986, the Committee's Pnncipal Working were issued regularly over the entire period, the program results Group on Primary Circuit integnty agreed it would be useful to have not previously been integrated in a single document. A I those setting cntena and making decisions about plugging of Isrge number of tests were performed to characterize the ef-degraded steam generator tubes to have a better appreciation fects of pressurized-water environment on the enhancement of of the criteria presently employed in other members countries fatigue crack growth. Specimens were taken from plateb, forg-and their technical bases. The United States Nuclear Regulatory ings, welds and heat-affected zones. This work provided a data base for development of ASME Code reference crack-growth { Commission (U.S. NRC) offered to arrange the preparation of a j comparative summary for CSNI based on responses to a ques- curves, as well as insights into some of the mechanisms of en- l tionnaire circulated by the OECD-NEA among member coun- vironmental enhancement. Static-load crack-growth-rate tests j tries. The following report is based upon the information ob- were also conducted, with some specimens loaded in cycess of q tained from nine countnes currently operating pressurized water 11 years. Extensive fractography was performed and the obser-reactors (PWRs)' vations were correlated with the microstructure and the level of  ; environmental enhancement. All the data obtained in this pro-gram were compared with results obtained et other laboratories. l NUREG/CR-5018: URANIUM OXIDE-IRON OXIDE MIXED AERO- l SOL EXPERIMENTS IN STEAM-AIR ATMOSPHERES.NSPP Tests 611,612,613 And 631. Data Record Report. TOBIAS,M.L.; NUREG/CR-5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS ADAMS,R.E. Oak Ridge Nat:onal Laboratory. January 1988. NUCLEAR ONE - UNIT 1. Volume 1, Program For inspectors. 63pp.6804010223. ORNL/TM-10588. 44974:192. CAMPBELL D.J.; GUTHRIE,V.H.; K!RCHNER,J.R.; et al. Oak This data record report summanzes the results from three Ridge National Laboratory. March 1988.192pp. 8805090144. ORNL/TM-10604. 45434:154. I tests involving mixed aerosbis of uranium oxide and iron oxide in a steam air ensaronment and one test in a dry air environ- This user's guide is a two-volume document designed to ment. This research sponsored by the U.S. Nuclear Regulatory teach NRC inspectors and NRC regulators how to access prob. i Commiss;on was ronducted in the Nuclear Safety Pilot Plant at abihstic risk assessment information from the two Plant Risk l Status information Management System (PRISIM) programs de- l the Oak Ridge National Laboratory. The purpose of this project is to provide a data base on the behavior of aerosols in contain- veloped for Arkansas Nuclear One Unit One (ANO 1). Volume I 1 describes how the PRA information available in Version 1.0 of I ment under conditions assumed to occur in postulated LWR ac-PRISIM is useful for planning inspections. Using PRISIM, in- l cident sequencos; this data base will provide expenmental vali-spectors can quickly access PRA information and use that infor-dation of aerosol behavioral codes under development. In the mation to update nsk analysis results, reflecting a plant's status < report a bnei desenption is given of each test together with the at any particular time. Volume 2 desenbes how the PRA infor- ' results in the form of tables and graphs. Included are data on mation available in Version 2.0 of PRISIM is useful as an eval-aerosol mass concentration, aerosol fallout and plateout rates, uation tool for regulatory activities. Using PRISIM, regulators total mass fallout and plateout, aerosol particle size, vessel at- can both access PRA information and modify the information to mosphere pressure, vessel atmosphere temperatures, tempera- assess the impact these changes may have on plant safety, ture gradients near the vessel wall, and steam condensation Both volumes are stand-alone documents; each volume pre-rates on the vessel wall. sents several sample computer sebsions designed to lead the user through a variety of PRISIM applications used to obtain NUREG/CR-5019: NEUTRON EXPOSURE PARAMETERS FOR PRA-related information for monitonng and controlling plant risk. THE METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATION NUREG/CR-5021 V02: USER'S GUIDE FOR PRISIM ARKANSAS SERIES CAPSULES. MILLER,LF.; BALDWIN,C.A.; NUCLEAR ONE - UNIT 1. Volume 2, Program For Regulators. STALLMANN.F.W.: et al. Oak Ridge National Laboratory. March CAMPBELL.D.J.; GUTHRIE,V H.; KIRCHNER J.R.: et al. Oak 1908.127pp.8804280633. ORNL/TM-10582. 45274'024. Asdge National Laboratory. March 1988. 224pp. 8805090170. The Heavy Section Steel Technology (HSST) Program, sup- ORNL/TM-10604. 45433.290. ported by the U.S. Nuclear Regulatory Commission, has com- This user's guide is a two-volume document designed to pleted the Series 5 (HSSTS) irradiation expenments. Twelve teach NRC inspectors and NRC regulators how to access prob-capsules which contain rnetallurgical test specimens have been abilistic risk assessment information from the two Plant Risk irradiated at the Oak Ridge Research Reactor located at the Status information Management System (PRISIM) programs de. Oak Ridge National Laboratory. These capsules have been dis- veloped for Arkansas Nuclear One - Unit One (ANO-1). Volume assembled, internal dosimeters have been analyzed, and expo. 1 coscribes how the PRA information available in Version 1.0 of sure parameters are presented for each irradiation test speci- PRISIM is useful for planning inspections. Using PRISIM, in-men. This report cosenbes the computational methodology for spectors can quickly access PRA information and use that infor-the least-squares adjustment of the dosimetry data with neu, mation to update nbk analysis results, reflecting a plant's status I tronics calculations, and it presents exposure parameters at at any padecular time. Volume 2 desenbes how the PRA infor-each test specimen location for the fluence rate greater than mation available in Version 2.0 of PRISIM is useful as an eval-l 1.0 MeV, fluence rate greator than 0.1 MeV, and displacements uation tool for regulatory activities. Using PRISIM, regulators I por atom. The specific activity of each dosimeter at the end of can both access PRA information and modify the information to irradiation is listed in the Appendix. essess the impact these changes may have on plant safety. Each volume is a stand-alone document. l l

38 Main Citations and Abstracts NUREG/CR-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAIL. clear grado cabinets was taken to identify components which URE PREDICTION METHODS FOR PIPING. SEVERUD,L.K.; experienced a rattling environment. The INEL has used several ANDERSON.M.J.: LINDOUIST,M.R.; et al. Westinghouse Han- different methods to reduce that data and has determined the ford Co. January 1988. 222pp. 8802030326. WHC-EP-0081. existence of a number of device anomalies in the presence of 44244:001. high frequency cabinet response to earthquake-type excitation Seismic response and failure analyses were performed for motion. However, causahty between the high frequency content four piping systems that were shake-tested to high-level nonlin- and the malfunctions could not be conclusively confirmed. ear and inelastic response levels. Both pre- and post test anely- Phase 11 of the study consisted of shake table testing for the ses were accomplished, A number of simplified elastic, elasto- most prevalent malfunction discovered in the survey, relay chat-plactic, and inelastic transient dynamic analysis methods were . ter, with excitation frequency content in the seismic range and utilized. Desenptions of these methods, with their special struc^ higher. This report will document the results of Phase I and 11 of tural parameters and compansons of predictions using each the study, insight into the susceptibility of electrical devices to method to test data, are provided. Reasonably useful, but con-rattling and characterization of relay chatter mechanisms offers servative, methods were found for predicting the hi0h-level in-guidance in addressing rattling effects during qualification. elastic response and the failure modes. NUREG/CR-5024: TENSILE AND J-R CURVE CHARACTERIZE- NUREG/CR-5032: MODELING TIME TO RECOVERY AND INITi-TlON OF THERMALLY AGED CAST STAINLESS STEELS. ATING EVENT FREQUENCY FOR LOES OF OFFSITE POWER HISER.A.L. Materials Engineenng Associates, Inc. September INCIDENTS AT NUCLEAR POWER PLANTS. IMAN,R.L.; 1988. 208pp. 0810110239. MEA-2229. 47044:051, HORA,S.C. Sandia National Laboratories. January 1988. 61pp. Although cast stainless steels have excellent properties in the 8804080121. SAND 87-2428. 45038:095. as-received or preservice condition, significant degradation of industry data representing the time to recovery of loss of off-the properties, pnncipally fracture toughness (J-R curve), can site power at nuclear power plants for 63 incidents caused by occur after extended exposure to elevated temperatures typical plant- centered losses, grid losses, or severe weather losses of service conditions. The NRC is sponsonng work to study the are fit with exponential, lognormal, gamma and Weibull probabil-significance of in-service embnttlement of thermally aged, cast ity models. A Bayesian analysis is used to compare the adequa-stainless steel. This report summtJizes the results of tensile and J-R curve tests of commercial and expenmental heats of thes cy of each of these models and to provide uncertainty bounds steels. The materials were supplied by Argonne National Labo- dNM ekAem MdM Me ratory (ANL), who have conducted microstructural studies to the probability models fitted to each of the three sources of identify the degradation mechanisms. The loss in Charpy V data is presented as a method for predicting the time to recov-upper shelf energy can be quite large, up to 57% for aging at ery of loss of off-site power. The composite model is very gen-450 degrees C for 9980 h. The decrease in fracture toughness, eral and can be made site specific by making adjustments on specifically J levels on the J-R curve, can be even more severe, the models used, such as might occur due to the type of switch-with a reduction of 75% in some cases. Data from this study yard configuration or type of grid, and by adjusting the weights j are accessible through the NRC's Piping Fracture Mechanics on the individual models, such as might occur with weather con- l Data Base (PIFRAC), an on-line system available at MEA. ditions exbting at a particular plant. Adjustments in the compos- l ite model are shown or different models used for switchyard NUREG/CR-5029: MELT PROGRESSION IN SEVERELY DAM-configuration and for different weights due to weather. Balesian AGED REACTOR CORES. DOSANJH.S.S. Sandia National Lab-oratones. December 1987. 82pp. 8804080274. SAND 87 2384. approaches are also presented for modehng the frequency of 45066:229. initiating events leading to loss of off-site power. One Bayesian A model of melt formation and relocation in a two-dimension, model assumes that all plants share a common incidence rate al core rubble bed is developed in this report. The analysis in- for loss of off-site power, while the other Bayesian approach cludes mass conservation equations for the species of interest models the incidence rate for each plant relative to the inci-(UO(2) and ZrO(2)); a hquid momentum equation (z,r,) that in, dence rates of all other plants. Combining the Bayesian models corporates the effects of drag, gravity and capillary forces; and for the frequency of the initiating events with the composite an energy equation that includos internal heat generation by Bayesian model for recovery provides the necessary vehicle for decay heating, convection by the liquid and the solid (as it col- a complete model that incorporates uncertainty into a probabi-lapses) as well as conduction and raciation through the bed. An listic risk assessment. equilibrium UO(2FZrO(2) phase diagram is prescribed and radi-ateve heat transfer through the bed is incorporated utilizing a NUREG/CR 5033:

SUMMARY

DESCRIPTION OF THE SCALE temperature-dependent thermal conductivity. Models developed MODULAR CODE SYSTEM. PARKS,C.V. Oak Ridge National l in this work will be implemented in the MELPROG computer Laboratory. December 1987.134pp.8803020241. ORNL/CSD/ l code that is boing developed by Sandia and Los Alamos Na- TM-252, 44544:202. l tional Laboratories. The modified version of MELPROG will then SCALE - a modular code system for Standardized Computer be used to calculate melt progression, crust growth, pool forma- Analyses for licensing [ valuation - has been developed at Oak tion, crust failure and the relocation of debns material into the Ridge National Laboratory at the request of the U.S. Nuclear lower plenum dunno the Three Mile Island accident and other Regulatory Commission staff. The SCALE system utihres well-nuclear reactor accidents. estabbshed computer codes and methods within standard ana-NUREG/CR-5031: SIGNIFICANCE OF IN-STRUCTURE GENER- lytic sequences that (1) allow simplified free-form input, (2) auto-ATED MOTION IN SEISMIC OUAllFICATION TESTS OF CABl. mate the data processing and coupling between codes, and (3) NET MOUNTED ELECTRICAL DEVICES. THINNES.G.; provide accurate and reliable results. System development has GLOZMAN V. EG&G Idaho, Inc. (subs. of EG&G Inc.). July been directed at criticality safety, shielding, and heat transfer 1988. 453pp. 8809220092. EGG 2523. 46877:170. analysis of spent fuel transport and/or storage casks. However, The Idaho National Engineenng Laboratory (IN8EL) has been only a few of the sequences (and none of the individual func-conducting a research study to assist the United States fduclear tional modules) are restricted to cask applications. This report Regulatory Commission (USNRC) in determining susceptibihty of will provide a background on the history of the SCALE develop-electncal devices to in-structure generated motion sometimes ment and review the components and their function within the present in electiical cabinets. In Phase 1 of this study, a aurvey system. The available data libranes are also discussed, together of past seismic qualification tests conducted at Wyle Laborato- with the automated features that standardize the data process-nes on vanous electncal and control equipment housed in nu- ing and system analysis.

i j i Main Citations and Abstracts 39 NUREG/CR-5038: OPTIMIZATION OF THE CONTROL OF CON. NUREG/CR-5042 S01: EVALUATION OF EXTERNAL HAZARDS -! TAMINATION AT NUCLEAR POWER PLANTS. KHAN,T.A.: TO NUCLEAR POWER PLANTS IN THE UNITED ~ ~! BAUM,J.W. Brockhaven National Laboratory. May 1988.148pp. STATES. Seismic Hazard. PRASSINOS,P.G. : Lawrence Liver- ] 8806230171. BNL-NUREG-52073. 45898:084. . more National Laboratory. April 1988,64pp. 8805030107. UCID-  ; A methodology is described for the optimization of the actions 21223. 45343:194. taken to control contamination. It deals with many aspects of As part of the research program supporting the implementa-  ! contamination, such as the monetary value assigned to a unit of . tion of the NRC Policy Statement on Severe Accidents, the l radiation dose, the treatment of skin and extremity dose, and Lawrence Livermore National Laboratory (LLNL) has performed- 4

          ' the inefficiencies introduced from working in a contaminated on-        a study of the risk of core damage to nuclear power plants in veronmont. The optimization method is illustrated with two case .       the United States due to seismic initiated events. The broad ob -

studies based on cleanup projects at nuclear power plants. jective has been to gain an understanding of whether or not Guidelines for the use of protective apparet, and for monitoring seismic events are among the major potential accident initiators radiation and contamination at various levels of contamination that may pose a threat of severe reactor core damage or of are presented. The report concludes that additional research is ' large radioactive release to the environment from the reactor, required to quantify the effect of a contaminated environment The analysis was based on two figures-of. merit, one based on on work efficiencies. core damage fiequency and the other based on the frequency of large radioactive releases. Using these two figures-of-merit NUREG/CR-5039 V01: REACTOR ~ SAFETY RESEARCH SEMI- as evaluation criteria, it has been possible to ascertain that the ANNUAL REPORT. January June 1987. Volume 37

• Sandia risk from seismic initiated accidents is an important contributor National Laboratories. January 1988f 342pp. 8803150370. to overall risk for the U.S. nuclear power plants studies.

SANDB7-2411. 44692:048.  ; Sandia National Laboratories is conducting, under USNRC NUREG/CR-5043: CONTAINhlENT PENETRATION SYSTEM sponsorship, phenomenological research related to the safety (CPS) TESTS UNDER ACCIDENT LOADS. CRAPO,H.S.; of commercial nuclear power reactors. The research includes . STEELE.R. EG&G Idaho, Inc. (subs. of EG&G, Inc.). August experiments to simulate the phenomenology of the accident 1988. 78pp. 8809140306. EGG-2524. 46802:121. conditions and the development of analytical models, verified by This report provides the.results of accident simulation tests of experiment, which can be use to predict reactor and safety sys-tems performance and three typical light water reactor containment penetration sy - N under abnormal conditions. The tems to provide a technical basis for the support and develop-objective of this work is to povide NRC requisite data bases ment of equipment qualification procedures at design basis load and analytical methods to (1) identify and define safety issues, levels and to determine safety margins at severe accident load (2) understand the progression of risk- significant accident so-quences, and (3) conduct safety assessments. The collective levels. The three systems tested were (a) an 0-in. gate valve NRC-sponsored effort at Sandia NationC Laboratories is direct- sys?em modeling a containment sprsy system; (b) an 8-in. but. . terfly valve system modeling a purge and vent system; and (c) a e,d at enhancing the technology base supporting licensing deci-sions. 2-in. glove valve system modeling the numerous smalLbore ppng systems. The valve types, valve sizes, piping configura-NUREG/CR-5039 V02: REACTOR SAFETY RESEARCH SEML ti ns, penetrations, and supports used for the tests are typical ANNUAL REPORT. July-December 1987. Reactor Safety Re- of those found in commercial U.S. nuclear power plants for con-search Program. WALKER,J.V. Sandia National Laboratories. tainment isolation applications. The three systems tested were November 1988. 337pp. 8901040369. SANDS 7-2411. m unted in a fixture and exposed to simulated severe accident 47974:158 mechanicalloads by displacing the penetrations relativ6 to the See NUREG/CR-5039,V01 abstract. piping. Thermal and pressure loads were also applied. The test ' l results indicate that valve, penetration, or piping system failure , NUREG/CR-5041 VD2: RECOMMENDATIONS TO THE NRC during hypothesized accident events is unlikely due to accident- j FOR REVIEW CRITERIA FOR ALTERNATIVE METHODS OF inoucedloads.  ! l LOW-LEVEL RADIOACTIVE WASTE DISPOSAL. Task 2b: Earth- - Mounded Concrete Bunkers. DENSON,R.H.; BENNETT.R.D.; NUREG/CR-5044: ESTIMATION TECHNIQUES FOR COMMON WAMSLEY,R.M.; et al. Army, Dept. Of, Army Engineer Water- CAUSE FAILURE EVENTS. KELLY.E.J.; HEMPHILL,G.M.- Los I we s Ex riment Station. January 1986.118pp. 8804010092. Alamos National Laboratory. March 1988. 53pp. 8806100174. i LA.11179-MS. 45788:205. I

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Recommendations are provided for general design enteria Cnmmon cause failure probability estimation techniques, in- l and specific design review criteria covenng the design, con- cluding B-factor, basic parameter, binomial failure rate, multiple l struction and operation of the earth-mounded concrete bunker Greek, and C-factor estimators, are evaluated and compared  ! (EMCB) alternative method of low-level radioactive waste (LLW) using simulation data that captures the real world problem of disposal. An EMCB consists of a below grade reinforced con. sparse data from different plants. The effects on the estimators' crete bunker and an above grade tumulus. The reinforced corp performances from underlying factors such as common cause crete bunker is constructed in an excavation that is located shock rates, lethal shock rates probability of failing given a below the freeze line on a previous foundation blanket. Pervious shock, independent failure rates, and system operational time '  ! fill materialis compacted adjacent to the bunker in the excavat. are discussod. Worst case results are reported, and it is seen  ; ed area and the top of the bunker is covered with a low perme. that for extremely small common cause failure probabilities the ' abihty matenal The above grade tumulus built over the bunker binomial failure rate estimators are best. However, these esti-consists of stacked waste containers that are covered initially mators can underestimate the true probabilities when the fail with pervious fill and then a layered waste cover system. The ures deviate from the binomial failure rate model. The B-factor l EMCB includes filter and drainage systems that are connected technique is shown to be conservative, and in some cases to ' to monitorings sumps. Eight major review criteria categories are overestimate the true probabihty by several orders of magni-identified in the report. The categories include (1) the loads and fude. When there are observed failures for each failure event, I load combinations to be used in design, (2) structural design *he basic parameter technique is best and is easily calculated.

      - and analytical methods. (3) construction matenal quality > J du-         'his estimator is investigated in detail and is used to develop an rability, (4) construction and operations, (5) quahty assurance,         estimator for the probability of K or more units failing due to a (6) structural performance monitonng (7) filter and drainage             common cause. Uncertainty limits for this probability are also systems, and (8) waste cover system.                                     developed.                                                                 i L_ _ _ -                  ^

40 Main Citations and Abstracts NUREG/CR-5045: KANSAS-NEBRASKA SEISMICITY STUDIES tions performing RPV fluence determinations. Another purpose USING THE KANSAS-NEBRASKA M:CROEARTHOUAKE of this report is to provide supporting documentation for any NETWORK. Final Report. STEEPLES.D.W.; HILDEBRAND,G.M.; proposed regulatory guide on this subject. BENNETT,B.C.; et al. Kansas, Univ. of Lawrence, KS. March 1988. 78pp. 8806020039. 45706:222. NUREG/CR-5050: ANNOTATED BIBLIOGRAPHY OF RELIABIL-ITY AND RISK DATA SOURCES. HESTER,0.V.; BROWN S.R.; The Kansas Geological Survey (KGS) operates a 15 station GENTILLON.C.D. EG&G Idaho, Inc. (subs. of EG&G, Inc.). seismograph network with stations located in northeast Kansas March 1988.195pp. 8804050264. EGG-REO 7827. 45009:050. and southeast Nebraska. The network is supported in part by This document is an annotated bibliography of nuclear, non-funding from the United States Nuclear Regulatory Commission (NRC). This report discusses operation of the network and sum. nuclear, and foreign data sources that are useful in nuclear Power plant reliability and risk analysis applications. A brief de-marizes the results of research that allows a better understand- senption of the contents, areas of usefulness, access informa-ing of the seismicity of the region and the link between the seis- tion, and the name and address of a contact is provided for macity and the tectonic setting of the region. Results from a data sources of all types. In addition, for nuclear data sources, crustal seismic refraction survey along a 500 km long line are tabular compansons are made. These comparisons include the discussed in Appendix B to the report. scope of the data sources; their operational, special-purpose NUREG/CR-5047: RADIONUCLIDES ACCUMULATION BY operational, pedigree, aggregated, and derived data; the oper. AOUATIC BIOTA EXPOSED TO CONTAMINATED WATER IN ational and design data each data source originates from; and ARTIFICIAL ECOSYSTEMS BEFORE AND AFTER ITS PAS- access information. Probabilistic risk assessments (PRAs) are SAGE THROUGH THE GROUND. CUSHING.C.E.; profiled separately. For each PRA, background information that RICKARD,W.H.; WATSON.D.G. Battelle Memorial Institute, Pa- describes the PRA and the plant itself is provided. Also the cific Northwest Laboratory. February 1988. 25pp. 8802180003. input daia sources used to support each PRA are identified. Of PNL 5590. 44364.234. special interest is the identification of unique plant-specific data Aquatic biota (fish, clams, algae) and an aquatic emergent sou9es that have evolved from the PRA analyses. To the vascular plant were expenmentally exposed to mixtures of ra- extent possible, how the data sources were used in the analysis dionuclides in three artificial streams for time periods extending is also discussed. up to 86 days. The experimental streams consisted of industrial water discharged directly into a teaching trench, and the same NUREG/CR-5051: DETECTING AND MITIGATING BATTERY water after it had migrated through the ground for a distance of CHARGER AND INVERTER AGING. GUNTHER,W.E.; 260 meters. The third stream was river water. After migrating LEWIS,R.; SUBUDHI,M. Brookhaven National Laboratory. through the ground, the teach water still contained measurable August 1988. 120pp. 8901040389. BNL-NUREG-52108. amounts of (60)Co and (90)Sr. These were accumulated by all 47974:036. classes of aquatic biota and by tomato p; ants grown in a quasi- This report constitutes the second of the two-phase approach hydroponic system. This expenrnent indicated that (60)CC and for assessing the safety and operational aspects of battery (90)Sr can be transported along with ground water flow and are charger and inverter aging in nuclear power plants. This work available for biological uptake by aquatic biota- was conducted under the auspices of the U.S. NRC Nuclear Plant Aging Research (NPAR) Program. The analyses present-NUREG/CR-5048: REVIEW OF THE NATURAL CIRCULATION ed include an assessment of the recent operating experiences EFFECT IN THE VERMONT YANKEE SPENT-FUEL with battery chargers and inverters and a presentation of battery POOLDocket No. 50-271.(Vermont Yankee Nuclear Power charger and inverter reliability improvements that may be Corp) WHEELER,C.L Battelle Memonal Institute Pacific North-achieved through modification of the equipment's configuration west Laboratory. January 1988.144pp. 8801250223. PNL-6388. and an increased inspection frequency. The resutts are evalust. 44113:158. ed from a survey of the current maintenance and test practices A 7,429-node, three dimensional computer model of the Ver, used for battery chargers and inverters in nuclear power plants, mont Yankee spent-fuel pool was set up and run using the along with the manufacturer's recommendations for maintaining porous media model of the TEMPEST computer code. The re. equipment operability. Advanced designs for uninterruptible sults of this analysis show that natural circulation is sufficient to power systems, subcomponent improvements, and current mon-ensure adequate cooling, regardless of the loading pattern used itoring and protective equipment are described and related to or the orientation of the cooling system discharge nozzle. i their potential applicability in nuclear power plants. A naturally NUREG/CR-5049: PRESSURE VESSEL FLUENCE ANALYSIS aged inverter and battery charger were tested at BNL to evalu-l ' AND NEUTRON DOSIMETRY. KAM,F.B.; MAERKER,R.E.; ate the naturally aged condition, the effectiveness of condition l WILLIAMS,M.L.; et al. Oak Ridge National Laboratory. Decem- monitonng techniques, and the practicality of implementing se- ! ber 1981. 66pp. 8802240213. ORNL/TM-10651, 44468.300. lected maintenance and monitonng recommendations. A review of the vanous methodologies used by industnes and research institutes for reactor pressure vessel (RPV) fluence de- NUREG/CR 5052: OPERATING EXPERIENCE AND AGING AS-termination shows that most organizations employ are analysis SESSMENT OF COMPONENT COOLING WATER SYSTEMS sequence consisting of three steps. These include transport cal- IN PRESSURIZED WATER REACTORS. HIGGINS,J.; culations, dosimetry measurements, and a statistical procedure LOFARO,R.; SUBUDHl M.; et al. Brookhaven National Laborato-to combine the calculations and measurements to arrive at a ry. July 1988. 201pp. 8811070038. BNL-NUREG-52117. fluence value which has a smatter uncertainty than the original 47387:106. calculations. An accurate determination of damage fluence ac- An aging assessment of Component Cooling Water (CCW) cumulated by the RPV as a function of space and time is es- systerns in Pressurized Water Reactors (PWRs) was performed sential in order to ensure the vessel integrity for beth pressur- as part of the Nuclear Plant Aging Research (NPAR) Program. 12ed thermal shock transients and end-of- life considerations. The objectives of the NPAR program are to provide a technical The desired accuracy for neutron exposure parameters such as basis for the identification and evaluation of degradation caused displacements per atom or neutron fluence (E greater than 1.0 by age in nuclear power plant applications. The information gen-MeV) is on the order of plus or minus 10% to plus or minus erated will be used to assess the impact of aging en plant 15% (1 standard deviation). These types of accuracies can orily safety and to develop effective mitigating actioris. Aging in the be obtained realisocally by validation of the entire ana!ysis se- CCW system was characterized using the Aging and Life Exten-quence in benchmark expenments. This report identifies a sion Assessment Program (ALEAP) Systems Level Plan devel-standardized procedure based on benchmark calculations data, oped by Brookhaven Nationa! Laboratory. Failure data from vari-and dosimetry measurements, which could be used by organiza- ous national data bases were reviewed and analyzed to identify

l 4 l Main Citations and Abstracts 41 predominant failure modes, causes and mechanisms in CCW NUREG/CR-5055: ATMOSPHERIC DIFFUSION FOR CONTROL i systems. Time-dependent failure rates for major components ROOM HABITABILITY ASSESSMENTS. RAMSDELLJ.V. Bat-were calculated to identify aging trends. Plent specific data telle Memorial institute, Pacific Northwest Laboratory. May j were obtained and evaluated to supplement data base results. 1988.120pp. 8805260384. PNL-6391. 45632:292. ' A computer program (PRAAGE) was developed and implement. This report presents the results of an evaluation of the proce-ed to model a typical CCW system design and perform Probabi- dure used by the NRC staff to assess nuclear reactor control listic Risk Assessment (PRA) calculations. Time-dependent fail. room habitability. The evaluation is based on experimental data ure rates were input to the program to evaluate the effects of collected in seven sets of field experiments at nuclear power aging on component importance and system unavailability. plant sites. The procedure is generally conservative, but the Changes in component importance and system unavailability models in the procedure show little skill in predicting the effects j with age were observed anc are discussed. of different atmospheric conditions on maximum effluent con-centrations in building wakes. Two alternative building-wake NUREG/CR-5053: OPERATING EXPERIENCE AND AGING AS. models have been developed using the experimental data. The j SESSMENT OF MOTOR CONTROL CENTERS. SHIER.W.; first model differs significantly from current nodels in the SUBUDHl.M. Brookhaven National Laboratory. June 1988. manner in which wind speed enters the model. The second 85pp.8812010086. BNL-NUREG-52118. 47685:154.  ; modei is an extension of the first model that has more desirable As part of the NRC Nuclear Aging Research Program asymptotic behavior and includes consideration of the mitigating (NPAR), an assessment was made of the characteristics of effect of plume rise on concentrations in building wakes. A cet aging and service wear of motor control centers (MCCs). MCCs of non-mathematical guidelines is offered for use in evaluating perform an important function in the operation and control of a Potential control room air intake locations. large number of motors that are included in safety-related sys-tems; thus, the operability and reliabihty of MCCs can impact NUREG/CR-5058: PRA APPLICATIONS PROGRAM FOR IN-the overall safety of operating nuclear plants. This report fol- SPECTION AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. lows the established NPAR strategy and investigates the oper. 50-313.(Arkansas Power And Light Company) VO,T.V.; k ational performance, the design and manufactunng methods, HARRIS,M.S.; GORE.B.F. Battelle Memorial Institute, Pacific and the current maintenance, surveillance and monitonng tech-Northwest Laboratory. March 1988. 74pp. 8804050160. PNL- ] i 6394. 44972:032. $ naques of MCCs. A significant result described in this report con-cerns the identification of important MCC failure modes, causes- The ANO 1 PRA has been anatyred to identify plant systems j and mechanisms from plant operational experience. Frequer" and components important to minimizing pubhc nsk, and to identify and primary failure modes of these components. This in- ] cies of failure determined for the various subcomponents of l MCCs also are described. In addition, recommendations are formation has been tabulated, and correlated with inspection 1 modules fro.n the NRC inspection and Enforcement Manual. provided for functional indicators to monitor the performance of J MCCs prior to failure. These functional indicators will be evalu- The report precents a series of tables, organized by system and l priontized by public risk, which identify components associated ated during Phase 2 of the program. with 98% of the inspectable nsk due to plant operation. The systems addressed, in descending order of risk importance, are: NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR DC Power, High Pressure injection, Low Pressure injection, REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW- Service Water, Reactor Protection, Emergency Feedwater, LEVEL RADIOACTIVE WASTE DISPOSAL Environmental Moni-tonng And Surveillance Programs. Emergency AC Power, Safety Relief Valves, Power Conversion, DENHAM.D.H.; STENNER,R.D.; EDDY,P.A.; et al. Battelle Memorial Institute. and Emergency Feedwater initiation and Control. This ranking is j Pacific Northwest Laboratory. July 1988.197pp. 8808120123. based on the Fussel-Vesely measure of ranking importance ap- l plied to core melt probability, i.e., the fraction of the core melt PNL-6553. 46458:288. probability which involves failures of the system of interest. Licensing of a facihty for low-level radioactive waste disposal requires the review of the environmental monitonng and surveil. NUREG/CR-5061: THREE-FREQUENCY EDDY-CURRENT IN-

                                                                                                                                                            )

i lance programs. A set of review criteria is recommended for the STRUMENT. DODD.C.V,; CHITWOOD,L.D. Oak Ridge National U.S. Nuclear Regulatory Commission (NRC) staff to use in each Laboratory. May 1988.155pp. 8808120118. ORNL/TM 10663. monitoring phase - preoperational, operational, and postopera. 46455:256. tional - for evaluating radiological and selected nonradiological A three-frequency eddy-current instrument has been con-parameters in proposed environmental monitoring and surveil- structed for general multiple property applications, with particu-lance programs at low- level waste disposal facihties. Applicable tar emphasis on 1.ght water reactor steam generator tubing ex-regulations, industry standards, and technical guidance on low, amination. A desenption is given of the overall operating princi-level radioactive waste are noted throughout the document. In p es of the complete instrument and of the operation of the dif-the preoperational phase, the applicant must demonstrate that forent modules in the instrument. Also included are wiring and  ; the environmental monitoring program identifies radiation levels pnnted circuit diagrams and the necessary computer programs and radionuclides concentrations at the site and also provides to run the instrument. adequate basic data on the disposal site, in the operational phase, the applicant must demonstrate that considerable care NUREG/CR-5063: DEVELOPMENT OF A MECHANISTIC UN-has been taken in designing and implementing the environmen- DERSTANDING OF RADIATION EMBRITTLEMENT IN REAC-tal monitoring programs and that data obtained during the first TOR PRESSURE VESSEL STEELS. Final Report. EBRAHIMI,F.; phase are reflected in the design of those programs. The oper- HOELZER.D.T.; VENABLES,D.; et al. Materials Engineering As-sociates, Inc. January 1988. 91pp. 8801270208. MEA-2268. ational phase must also be technically sound and broad enough 44149:040. to address potential issues that may be raised by the public. The microstructure of a senes of reactor pressure vessel The postoperational phase requires continued sampling and (RH/) steels and model iron alloys with various Cu, Ni, and P measurements of those media that may provide a future expo-sure pathway to the public, perhaps at a reduced frequency contents were examined in unirradiated and neutron-irradiated dunng the long-term care penod, based on the data obtained conditions using high resolution analytical microscopy. Fractog-raphy techniques were also applied. Objectives were to isolate during the operatiohat phase. Review checklists are provided for and sdentify the mechanisms by which these elements effect NRC use in evaluating the adequacy of environmental monitor-steel radiation embrittlement sensitivity as evidenced by notch ing and surveillance programs for compliance with applicable regulations. ductility and tensile strength changes. Radiation hardening of a reference iron alloy was found associated with the formation of

42 Main Citations and Abstracts pnsmatic dislocation loops of an interstitial nature. A very low A numerical simulation of the multidimensional thermal-hy-density of loops was observed for a low Cu-tow P steel having a draulic charreteristics in a 3-loop PWR during a hypothetical low radiation sensitivity. Copper decreased the size and in- TMLB' accident scenario has been performed by means of the creased number density of observable defects; however, the COMMIX computer code. The operating conditions, as well as enhancement of radiation sensitivity by Cu is due to a radiation- the modeling approaches used in the simulation, are discussed. induced formation of Cu-rich clusters / precipitates. No synergism Selected results, which show the natural- circulation phenom-between Cu and Ni was four.d in the model iron alloys except ena and compare the thermal-hydraulic behavior in the different for an enrichment of Cu clusters with Ni. Alloys containing ap- loops, are presented. preciable P did not show intergranular fracture. The detrimental effect of this impunty on radiation sensitivity for the case of a NUREG/CR-5071: TRAC SUPPORT SOFTWARE. JENKS,R.P.; low Cu content is due to a radiation-induced clustering of P. MARTINEZ.V. Los Alamos National Laboratory. June 1988.106. Phosphorus clustering was found absent in a high Cu-high P 8808120057. LA-11214-MS. 46478:252. iron alloy, consistent with the apparent inactive role of P in radi- This manual provides users of the Transient Reactor Analysis 1 ation sensitivity of high Cu steels The effect cf alloying / impurity Code (TRAC) with information about computer codes that can elements on the evolution of defect structures during irradiation be used to support their analysis efforts. These codes are col-is ascribed to vacancy trapping by solute atoms, leading to an lectively referred to as TRAC support software. Specifically, increased number density of defects. documentation is provided to allow users to implement, apply, ' aqd interpret the output of the support software that h avail-NUREG/CR-5065: TIME AND VOLUME AVERAGED CONSER- able. The TRAC code is an advanced best-estimate code for j VATION EOUATIONS FOR MULTIPHASE FLOW ICNG analyzing transients m thermal-hydraulic systems. Considerable MASS-WEfGHTED VELOCITY AND INTERNAL ENERGY. input data are required for modeling a trrge thermal-hydraulic SHA W.T.; CHAO.B.T.: SOO.S.L Argonne National Laboratory. system and considerable output data are produced from the January 1988. 84pp. 8804010116. ANL-87-51. 44950:125. computer calculation for analysis; the support software assists Conservation equations of mass, momentum, and energy for users in their management of both the input and the output multiphase flow, formulated on the basis of local volume aver, data. Initially, this document contains sections that describe aging followod by time-averaging for turbulent flows, are pre. EXCON, TRAP, EXTRACT, and CVRT, major support software sented. They are differential equations of transport with area in. used routinely at Los Alamos National Laboratory. j tegrals associated with interfacial transport. Because the spatial averaging theorems used in the analysis are subject to certain length scale restrictions, the resulting equations are best suited NUREG/CR-5072: DECAY HEAT REMOVAL USING FEED-AND-for dispersed systems. The tocal instantaneous variable is de. BLEED FOR U.S. PRESSURIZED WATER REACTORS. LOOMIS.G.; COZZUOL,J.M. EG&G ldaho, Inc. (subs. of EG8G, ! composed as a linear combination of its local intnnsic volume average and a spatial deviation. Use of the mass-weighted, Inc.).

• Idaho National Engineering Laboratory. June 1988.

l 48pp. 8807110521. EGG-2526. 46085:024. ! volume averaged velocity and internal energy simplified certain relationships between the volume average of products and the As part of a United States Nuclear Regulatory Commission product of volume averages. Recognition of the fact that the (USNRC) evaluation of current decay heat removal methods, spatial deviation component takes on positive and negative the adequacy of feed-and-bleed decay heat removal has been values within the averaging volume makes further simplifications assessed for United States pressurized water reactors feasible. Inasmuch as information is always lost as a result of (USPWRs). Use of feed-and-bleed for decay heat removal be-averaging, be it volume-averaging or time-averaging or both, the comes necessary in a pressurized water reactor (PWR) system lost informaison must somehow be replaced before the equa- if there is a loss of steam generator heat sink capability. The tions can be solved. This is commonly done by the develop- feed and-bleed technique involves passing hot fluid out of the merit of appropriate constitutive relations, which, however, is primary system through a pressurizer power operated relief i t not treated in this report. ' valve (PORV), while simultaneously feeding the primary system NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF charging flow, or both. This document summarizes the results of COMBINED ALPHA AND BETA IRRADIATION OF THE available experimental and analytical work that has been per. LUNG. Phase ll Report. SCOTT,B.R.; HAHN,F.F.; SNIPES,V.B.; formed to investigate the thermal- hydraulic phenomena associ-et al. Lovelace Biomed & Environmental Research Institute. ated with the feed-and-bleed technique. These results are then March 1988. 41pp. 8804280444. LMF 119,45269:001. synthesized into a statement of adequacy of the feed and-bleed This report summanzes an inhalation exposure experiment technique as a means of decay heat removal it is concluded that concems early and continuing effects of combined alpha that feed-and-bleed can be a viable alternative form of decay and beta irradiation of the lung of rats. Both morbidity at 18 heat remnval for USPWRs, but that successful use of feed-and-months and mortality within 18 months after exposure were ex. bleed is contingent upon the existence and use of procedures amined for rats exposed to the beta. emitter -(147)Pm. the alpha- for its implementation, as well as upon the specifics of plant emitter (238)Pu, or both combined. The results were used to design. validate hazard-function models that were developed (1) for pul-monary functional morbidity at 18 months and (2) for lethality NUREG/CR 5073: OUANTIFICATION OF MARGINS IN PIPING from radiation pneumonitis and pulmonary fibrosis within 18 SYSTEM SEISMIC RESPONSE. Methodologies And Damping.

• months. Both models were found to adequately predict the ex- Lawrence Livermore National Laboratory. JOHNSON,J.J -

perimental observations after combined chronic alpha and beta SENDA,B.J. EOE, Inc. February 1988.146pp. 8802240271' irradiation of the lung. A relative biological effectiveness of ap UCRL 21000. 44469:270-proximately 7 was obta ned for (238)Pu &%e radiation com- The conservation of seismic analysis and design of pip.ing pared to (147)Pm beta radiation for both pulmonary functional l systems due to analysis methodologies and damping values i morbidity and lethality from radiation pneumonitis and pulmo-was quantified. Envelope sesponse spectrum analyses, inde-nary fibrosis. J pendent support motion response spectrum analyses, and multh j NUREG/CR-5070: ANALYSIS OF NATURAL CONVECTION PHE- support time history analysis methodologies were evaluated. 1 NOMENA IN A 3 LOOP PWR DURING A TMLB' TRANSIENT Constant damping, ranging from 1% to 10%, and PVRC famp. USING THE COMMIX CODE. DOMANUS,H.M.; SHA,W.T. Ar- ing were considered. Conservatism were evaluated with re-gonne National Laboratory. January 1988. 53pp. 8803280243. spect to best estimate responses of the entire seismic analysis ANL 87 54. 44890.105. chain and of the piping system alone.

Main Citations and Abstracts 43 NUREG/CR 5075: THE SAFT-UT REAL TIME INSPECTION NUREG/CR-5078 V02: A RELIABILITY PROGRAM FOR EMER. SYSTEM - OPERATIONAL PRINCIPLES AND IMPLEMENTA- GENCY DIESEL GENERATORS AT NUCLEAR POWER TION. HALL,T.E.; REID L.D.; DOCTOR.S.R. Battelle Memonal PLANTS. Maintenance. Surveillance And Condition Monitoring. Institute, Pacific Northwest Laboratory. June 'iO88. 174pp. LOFGREN,E.V. Science Applications international Corp. (for-8808050278. PNL-6413. 46389:027. merly Science Applications, Inc.). HENDERSON.W.;

• This document provides a technical description of the real. BURGHARDT,D.: et al. Trident Engineers, Inc. December 1983.

time imaging system developed for rapid fiaw detection and 09pp. 8901090375. SAND 87-7176. 48109:146. - characterization utilizing the synthetic aperture focusing tech. This report is a companjort report on NUREG/CR-5078, nique for ultrasonic testing (SAFT-UT). The complete fieldabie Volume 1, "A Reliability Program for Emergency Diesef Genera- , system has been designed to perform inservice inspection of tors at Nuclear Fower Plants: Frogram Structure. The purpose light-water reactor components. Software was wntten on a DEC of this report is to provide technical findings and insights related LSI 11/23 computer system to control data collection. The un- a: failure evaluaaon, troubleshooting, maintenance, surveil-processed data is transferred to a VAX 11/730 host computer lance, and condition monitoring. ramples and recommenda- i to perform data processing and image display tasks. A parallel tions are provided for each of these areas based on actual I architecture peripheral to tne host computer, referred to as the emergency diesel generator (EDG) operating experience and l Real-Time SAFT Processor, rapidy performs the SAFT process-the opinions of desel generator experts. This report expands the more general guidance provided in Volume 1. In addition, a ing function. From the host's point of view, this device operates discusainn of EDG interactions with other plant systems (e.g., ) on the SAFT data in such a way that one may censider it to be instrumem air, service water, de power) is provided since experi- I a specialized or SFT array processor. A guide to SAFT-UT ence has shown that these support systems and their operation theory and conventions is included, along with a detailed de- can adversely affect EDG reliability Portions of this report have  ; scription cf the operation of the software, how to install the soft-been designed for use by onsite personnel for evaluating oper-  ! ware, and a detailed hardware description. ational characteristics of EDGs. NUREG/CR-5076: AN APPROACH TO THE QUANTIFICATION NUREG/CR-5080: A STUDY OF NEW ENGLAND SEISMICITY OF SEISulC MARGINS N NUCLEAR POWER PLANTS.The WITH EMPHASIS ON MASSACHUSETTS AND NEW  ; importance Of BWR Plant Systems And Functions To Seismic HAMPSHIREJinal Report Covering The Period 1976-1985. I Margins. AMICO,P.J. Lawrence Uvermore National Laboratory. . TOKSOZ,M.N.; V.ADINSKY-CADE Massachusetts institute of 1 Applied Risk Technology Carp. May 1988. 40pp. 8806230148. Technology, Canbndge, MA. January 1988. 135pp. UCRL-15985. 45896:060. 8803290408.44895:190. In NUREG/CR-4334 ("An Approach to the Quanbfication of This is the final repor', for U.S. Nuclear Regulatory Commis-Seismic Margins in Nuclear Power Plants"), the Export Panel on s on Contract No. NRC-04-76-209 with the Massachusetts insti-Quantification of Seismic Margins precented a technique for tute of Technology (M.I.T.) entitled "A Study of New England studying the issue of quantifying seismic margins. As part of Seismicity with Emphasis on Massachusetts and New Hamp-that technique, the panel included methods for simplifying the Wro". The contract period was from 1976 to 1985. During that margins assessment by screening out Components and systems time network daily activities progressed from determining phase using both systems and fragilities screening out components arrival times on analog records from a handful of field stations and systems using both systems and fragilities screening guide- to operation of an efficier:t realAme data acquisition system lines. At the time of that report, the panel was able to develop with advanced seismic data trislysis capabilities. Quarterly Progress Reports have provided contireous reporting of seismic fragilities screening guidelines for all plants, however the sys-tems screening guidelines applied only to PWRs (due to a activity in our area for monitonng purposes. Phase data from the M.I.T. network have regularly been included h the North-shortage of BWR seismic PRAs upon which to base BWR sys-tems screening guidelines). TI,is report develops the BWR sys- eastern U.S. Seismic Network (N.E.U.S.S.N.) bulletins. This report summarizes daily operations for the time period of the tems screening guidelines by utilizing the results of a number of above contract, and then describes some d the scientific re-BWR PRAs which have become available since the publication of NUREG/CR-4334. suits obtained from data provided 4 this and surrounding re-gioral seismic networks. The cientsfic results can be divided into several categories or to?ics: velocity structure models, NUREG/CR-5078 V01: A RELIABILITY PROGRAM FOR EMER- earthquake hazard studies, an0 the determination of earthquake GENCY DIESEL GENERATORS AT NUCLEAR POWER .nechanisms and focal depths, crustal stresses and seismic PLANTS. Program Structure. LOFGREN,EN.; DEMOS 8.G.M.; wave attenuation. Key components in our ability to complete FRAGOLA,J R.; et al. Sandia Natier.at Laboratones. April 1988. these studies have been the reasonably close station spacing 155pp. 8804280386. SAND 87-7176. 45271:304. (about 50 krn) in the northeastern U.S. and the continuous oper. The purpose of the report is to develop guidance for the NRC ation of the networks over a period of several years. staff to evaluate emergency diesel generator (EDG) reliability programs. Such reviews will likely result foiiowing the resolution NUREG/CR-5082: SIMULATION EXPERIMENTS ON TWO-of USI A-44 and GSI B-56. The diesel generator reliab;hty pro- PHASE NATURAL CIRCULATION IN A FREON 113 FLOW VIS-gram is a management systerr, for achieving and maintaining a UAUZATION LOOP. LEE,S.Y.; ISHil,M. Argonne National Labo-selected (or target) level of reliability. This can be achieved by: rator3. January 1988. 58pp. 8805200070. ANL-88-1. 45549:096. (1) understanding the factors that control the EDG reliability and In order to study the two-phase natural circulation and flow (2) then applying reliability and maintenance techniques in termination danng a small break loss of coolant accident in proper proportion to achieve selected performance goals. The LWR, simulation experiments have been performed using a concepts ano guidelines discussed in this report are concepts Freon-113 flow visuakzation bop. The main focus of the present and approaches that have been successful in applications expenment was placed on the two-paase flow behavior in the where high levels of reliability must be maintained. Both the hot-leg U-bend typical of B&W LWR systems. The loop was EDG reliability program process and a set of review items for built based on the two phase flow scaling criteria developed NRC use are provided. The review items represent a checklist under this program to find out the effect of fluid properties, for reviewing EDG reliability programs. They do not, in them- phase changes and couplir'g between hydrodynamic arid heat transfer phenomena. Significantly different flow behaviors have selves, cons'.itute a reliability program. Rather, the review items been observeo due to the non-equilibrium phase change phe-are those distinctive features of a reliability program that must nomena such as the flashing .and condensation in the Freon be present for the program to be acceptuble. loop in comparison with the previous adiabatic expenment. The

44 Main Citations and Abstracts phenomena created much more unstable hydrodynamic conds- NUREG/CR-5090: EFFECTS OF TEMPERATURE AND HUMIDI-tions which lead to cyclic or oscillatory flow behaviors. Also, the TY ON RESPIRATOR FIT UNDER SIMULATED WORK CONDI-void ,1:stnbution and pnmary loop flow rate were measured in TIONS. SKAGGS,B.J.; LOlBL,J.M.; CARTER,K.D.; et al. Los detail in addition to the important key parameters, such as the Alamos National Laboratory. July 1988.44pp.8809150286.LA-power input, loop friction and the liquid level inside the simulat- t 1236. 46803:095. ed steam generator. A study conducted at the Los Alamos National Laboratory compared quantitative fit factors and simulated work factors and NUREG/CR-5083: DESIGN CONSTRUCTION AND INSTRUMEN- determined the effects of temperature and humidity on respira-TATION OF A 1/6-SCALE REINFORCED CONCRETE CON . tor fit. This study used a commercially available fit chamber and - TAINMENT BUILDING. HORSCHELO.S. Sandia National Lab- an environmental chamber set at six conditions to simulate U.S. oratories. August 1988. 308pp. 8809220018. SAND 88-0030. work environments. Seven respirators were tested on a limited 46881:004. test panel of 10 subjects. The test results indicate that the per. This report describes the design, construction, and instrumen- formance of one powered air- purifying respirator (PAPR) helmet tation of a 1/6-scale reinforced concrete containment Duilding is significantly degraded during the simulated work exercises, that has been built at Sandia National Laboratories in Albuquer- whereas the halbmask PAPR performance was not affected. que, New Mexico. The model Light Water Reactor (LWR) con- Tight-fitting facepiece PAPRs provide higher protection than tainment building was designed and built to the American Socie- loose fitting PAPRs. The performance of the negative- pressure ty of Mechanical Engineers (ASME) Code by United Engineers half-mask and full-facepiece respirators is degraded during fit , I and Constructors, Inc. The containment model will be tested to tests at high humidity and high temperature. The degradation of failure to determine its response to static intemal overpressure- the fit factors for the negative-pressure half-mask during high zation. The results from testing the heavily instrumeisted con- humidity at ambient and high temperatures is probably due to tainment will be used to assess the capability of analytical facepiece slippage caused by sweating. More dynamic exer-methods for predicting the performance of containments subject cises, including motions in which the individual bends over and to severe accident loads as part of the U.S. Nuclear Regulatory stands up repeatedly, are recommended to develop quantitative Commission's program on containment integrity. The scaled di- fit factors that adequately simulate work factors. Tight-fitting mensions of the cylindrical wall and hemispherical dome are facepiece PAPRs should not be classified with loose-fitting typical of a fulLsize containment. Features representative of a PAPRs. 1 prototypical containment and included in the heavily reinforced - model are equipment hetches, personnel airlocks, several small NURF.G/CR-5094: APPLICATION OF STOCHASTIC METHODS piping penetrations, and a thin steel liner attached to the con- TO THE SIMULATION OF LARGE SCALE UNSATURATED FLOW AND TRANSPORT. POLMANN.D.J.; VOMVORIS,E.G.; crete by headed studt Over 1200 channels of instrumentation will be used to assess the model's behavior during testing. Sev- MCLAUGHLIN,D.; et al. Massachusetts institute of Technology, eral video and still camera stations are also used dunng testing Cambridge, MA. September 1988, 138pp. 8810050198. of the containment for both data gathering purposes and for 46990:024. support in conoucting the test. Numerical simulations are used to demonstrate features of large- scale flow and transport in heterogeneous unsaturated NUREG/CR-5084: lFCI: AN J.NTEGRATED CODE FOR CALCULA- soils using effective expressions derived from a stochastic TION OF ALL PHASES OF FUEL-GOOLANTJNTERACTIONS. theory developed at MIT, The case of stratified soil is examined YOUNG.M.F. Sandia National Laboratories. September 1987. for one- and two- dimensional (2D) flow and 2D transport. 56pp. 8802170359. SAND 87-1048. 44362:340. Finite-difference methods are used in the flow analysis and a IFCI, a code developed to calculate all phases of .FCis random-walk algorithm is used for transport. Input parameters rnechanistically, includes dynamic fragmentation and surface are derived from statistics of soil samples collected at the site area transport. This report desenbes the IFCI code and the re- of a large-scale tracer expeninent conducted by New Mexico sults of the first IFCI test problem, a simulation of a FITS experi- State University (NMSU). This study examines spatial variability ment. The calculation compared reasonably to the expenment. of saturated and unsaturated hydraulic conductivity regarding ef-facts on the bulk character of moisture flow and solute transport NUREG/CR-5086: PLATINUM CAT ALYTIC !GNITERS FOR LEAN (directional rates of movement and spreading of tne plume). HYDROGEN-AIR MIXTURE % THORNE,L.R.; VOLPONI,J.V.; The numerical simulations demonstrate tension-dependent ani-MCLEAN,W.J. Sandia National Laboratories. September 1988- sotropy, hysteresis and macrodispersion derived in the stochas-40pp. 8810110243, SAND 88-8201. 47043:268. tic theory based on this spatial variability. These features We have developed a prototype catalytic igniter for lean hy- cannot be explained using conventional deterministic mooets. drogen-air mixtures that could have important applications in nu- Preliminary results suggest that the stochastic theory is better clear reactor safety. The igniter has two useful charactenstics able to simulate the bulk character of the NMSU flow experi-related to these applications: it requires no electncal power and ments compared to a deterministic model. Sensitivity analyses it can ignite mixtures as loan as 5.5% hydrogen. The ignition in- identify major factors of soit variability which control moisture duction time ranges from 20 s to 400 s depending on the hydro" flow and spreading. Results of this study indicate the need for gen concentration, gas flow velocity, gas temperature and rela- careful experimental design and soils data collection. tive humidity of the gas mixture. Induction times are shorter for mixtures with higher hydrogen concentrations, higher flow ve- NUREG/CR-5095 V01: THERMODYNAMIC NONEOUILIBRIUM IN locities. higher gas temperatures and lower relative humidity. POST-CRITICAL-HEAT-FLUX BOILING IN -A ROD The igniter operates successfully under conditions that may be BUNDLE. Description Of Experiments And Sample Results. present dunng a loss of- coolant accident (LOCA) at a light TUZLA,K.: UNAL.C.; BADR,0.; et al. Lehigh Univ., Bethlenem, water nuclear reactor. In the event of a .LOCA, large quantities P A. June 1988.153pp. 8807110494. 46085:169. of hydrogen may be produced very rapdly and the catalytic ig. This report describes the post-CHF heat transfer experiments niter could provide a means of igniting it before dangerously in a 3 x 3 rod bundle. The objective was to obtain measure-high concentrations are attained, even in the event that electri- ments of thermodynamic nonequilibrium in the post CHF regime cal power required for conventional igniters is not available. The and to charactenze its effects on two-phase heat transfer for igniter has not be tested under all possible LOCA conditions. venfication of models used in thermal-hydraulic codes. A two-High gas velocities, water spray, steam and iodine- containing phase loop was constructed. The nine rod test bundle incorpo-compounds may be present during a LOCA and will defeat the rated a heated shroud to simulate the operating characteristics prototype igniter. However, shielding and semipermeable coat- of a large rod bundle. An original" hot patch" technique was de-ings on the igniter could overcome lhese difficulties. veloped to achieve steady-state post-CHF conditions in a rod

I Main Citations and Abstracts 45 l bundle. Steam temperature probes, developed earlier at Lehigh rosity, three-dimensional characteristics of fluid flow at the for tests in single tubes, were modified for use in the rod Apache Leap Tuff Site. Also, the variable saturation which bundle. Each test provided measurements of system pressure, j exists within fractures and the matrix at the site will also be re-coolant flow rate, wall heat flux, wall temperatures, two- pNise produced. Simulation scenarios including constant head and flux i equilibrium qualities, and vapor superheat temperatures. These surface boundary conditions, as well as slug and cyclic surface primary data permitted determination of wall heat transfer coeffe- boundary conditions are recommended, cients, nonequilibnum vapor qualities, and quench front propa-gation velocities. Expenments were conducted in three different NUREG/CR-5099: EVALUATION OF MATERIALS OF CON-modes: a) steady-state with fixed CHF location, b) reflood with STRUCTION FOR THE REINFORCED CONCRETE REACTOR advancing CHF locations (propagating quench front), c) boil-off CONTAINMENT MODEL KNOROVSKY,G.A.; HATCH,P.W.; with retreating CHF locations. Tests were expected over the fol. GUTIERREZ,M.R. Sandia National Laboratories. September lowing range of conditions: Coolant mass flux, 0.1 to 26 kg/ 1988. 59pp. 8810030190. SAND 88-0052,46972.251. m(2)s, (lnlet) quality,40 degrees C subcooled to 0.40, Pressure, This report summarizes the chemical analysis, metallography, 105 to 120 KPa. Heat flux,5 to 43 KW/m(2). and tensile test results obtained from steel matenals and welds NUREG/CR-5095 V02: THERMODYNAMIC NONEOUILIBRIUM IN used in the construction of a 1/6-scale model reinforced con-POST-CRITICAL-HEAT FLUX BOILING IN A ROD crete nuclear reactor containment. The purpose of building such BUNDLE. Data For Stabilized Quench Front Tests. TUZLA,K.; a model is to expenmentally venfy the ability of numerical 1 UNAL,C.; BADR,0.; et al. Lehigh Univ., Bethlehem PA. June models to predict deformation and failure in full-size contain- I 1988. 207pp. 8807110524. 46080:293. ments. Naturally, the predictions of such models are strongly in-See NUREG/CR-5095,V01 abstract- fluenced by the constitutive models used for the containment materials. The program reported on here was iretended to pro-NUREG/CR-5095 V03: THERMODYNAMIC NONEOUILIBRIUM IN vide such data. Besides providing tensih test data on the sheet, POST-CRITICAL-HEAT-FLUX BOLLING IN A ROD plate, and rebar materials used for dome, cylinder, cylinder in- ] BUNDLE. Data For Advancing Quench Front Tests. TUZLA K.; serts, reinforcements and penetrations, pull tests on sev3ral UNALC.; BADR,0.; et al. Lehigh Univ., Bethlehem, PA June weld geometries have been obtained. Standard tensils test data 19BB. 571pp. 8807110530. 46079:081. See NUREG/CR-5095,V01 abstract. derived from load vs. extensometer output records (i.e. engi-neering stress vs. engineering strain) are supplemented with I

                                                                                           *            *" *"                      "     "O * "#'

NUREG/CR-5095 V04: THERMODYNAMIC NONEOUILIBRIUM IN POST-CRITICAL-HEAT-FLUX BOILING IN A ROD cross sociional areas during interrupted tensile tests. Additional-BUNDLE. Data For Retreating Quench Front Tests. TUZLA K.; ly, 7" value measurements (ratio of width to thickness strain) UNAL,C.; BADR,0.; et al. Lehigh Univ., Bethlehem, PA. June which can be used to provide information about anisotropic mul-tiaxial flow conditions are derived from the true strain data and NURdG/ R 095, 01 ac~ rep rted. A small number of tests which included holds at con-stant load levels were also performed and analyzed to yield am- 1 NUREG/CR-5096: EVALUATION OF SEALS FOR MECHANICAL bient temperature creep equatinns. Finally, conditions resulting PENETRATIONS OF CONTAINMENT BUILDINGS. from construction practices (such as residual stresses and the BRINSON,0.A.; GRAVES.G.H. Sandia National Laboratories.

• Bauschinger effect) which may affect the prediction of yield and ERC International, Inc. August 1988. 137pp. 8810110264. fracture stresses from this data are bnefly discussed. ]

SAND 88-7016. 47043:038. This report describes tests of elastomeric seals that are used NUREG/CR-5005: RESPONSE MARGINS INVESTIGATION OF in the mechanical penetrations of nuclear power plant contain- PIPING DYNAMIC ANALYSES USING THE INDEPENDENT ments. These tests assessed the effects of thermal aging, radi- SUPPORT MOTION METHOD AND PVRC DAMPING. ation and thermal eging, sealing surface separation, and BEZLER,P.; WANG,Y.K.: REICH.M. Brookhaven National Labo-squeeze on the performance of several gasket designs: 0-ring, ratory. March 1988. 393pp. 8805260282. BNL-NUREG-52137. gum drop, double dog ear, and tongue and groove. Both ethyl- 45633:052. ene propylene rubtier and silicone rubber gaskets were tested. An evaluation of Independent Support Motion (ISM) response The environment for testing enveloped a hypothetical severe spectrum methods of analysis coupled with the Pressure Vessel accident: 143 psig and up to 700 degrees F. The seal's per- Research Committee (PVRC) recommendation for damping, to formance is quantifieo in terms of the leakage onset point on compute the dynamic component of the seismic response of the time-temperature curve. piping systems, was completed. Response estimates for five NUREG/CR-5097: SIMULATION OF LIQUID AND VAPOR MOVE- piping / structural systems were developed using fourteen var. MENT IN UNSATURATED FRACTURED ROCK AT THE iants of the ISM response spectrum method, the Uniform Sup-APACHE LEAP TUFF SITE.Models And Strategies. YEH,T.- port Motion response spectrum method and the ISM time histo. C. JIM; RASMUSSEN.T.C.: EVANS,0.D. Arizona, Univ. of. ry analysis method, all based on the PVRC recommendations Tucson, AZ. March 1988. 84pp. 8803220116. 44774:066. for damping. The tSM/PVRC calculational procedures were The physical, hydraulic and pneumatic properties of vanably iound to exhibit orderly charactenstics with levels of conserv-siturated, fractured tuff are currently being evaluated at the 8tism comparable to those obtained with the ISM / uniform Apache Leap Tuff Site, located near Superior, Arizona. Nine in- damping procedures. Using the ISM /PVRC response spectrum clined boreholes, the deepest of which penetrate to a depth of a,othod with absolute combination between group contributions thirty meters, have yieldeo' over 270 m of oriented core. Field provided consistently conservative results while using the ISM / , and laboratory data with regard to matrix and fracture properties PVRC response spectrum method vith square root sum of l are being collected which will be used to characterize the site. squares combination betwesn group contributions provided esti-1 A description of the characterization parameters as well as field mates of response which were deemed to be acceptable. l and laboratory techniques used to collect the parameters is pre-sented. To extend the charactenzation to larger scales, as well NUREG/CR-5106: USER'S GUIDE FOR THE TACTS COMPUTER i CODE. WEST.D.B.; GtLPIN,H.E. Science Applications interna-as to interpret collected data, computer simulation modeling will be performed. A review and desenption of available computer tional Corp. (formerly Science Applications, Inc.). June 1988. 206pp. 8807070447. SAIC-88/3023. 46013:011. models is presented. Recommendations for site charactenzation The TACT 5 computer code, a successor to TACT 111 and ear-includes the use of analytic stochastic models, equivalent porous media modes, and discrete fracture network models. her versions of TACT (an acronym for Transport of Activity), Such models may accurately reproduce the expected dual po- simulates the movement of radioactivity hypothetically released from a reactor core as it migrates through user-defined regions

46 Main Citations and Abstracts (nodes) of the containment, is immobihzed by filters and sprays, scribes a two dimensional (r z) debris meltdown model that has beer developed for use in the MELPROG code. Of interest in and is released to the outside environment. The code has been modsfied to run on a personal computer (PC). A series of inter- this study is melt progression in LWR accidents. The analys:s active BASIC pre-processor codes assists the user in compiling includes coupled mass, momentum and energy equations. a nuclide input data file and a plant model data file. The plant Phase diagrams are used to model Fe-Zr and U-Zr-O interac-model data file specifies a dynamic compartment model, which tions. Solutions are qualitatively similar to the post-accident con-is represented by systems of ordinary differential equations with figuration of the Three-Mile Island (TMI-2) core. Key results are constant coefficients. Tne equations are solved explicitly by (1) a dense metallic crust is created near the bottom of the bed as molten matenals flow downward and freeze; (2) liquid accu-matrix transformation methods. A code run carries out the inte-gration of these systems of equations over a succession of time mulates above the blockage and if zirconium is present, the intervals following reactor shutdown, with the interval bound- pool grows rapidly as molten Zr dissolves both UO(2) and aries corresponding to transitions of system parameter values ZrO(2) particles; (3a) if the melt wets the solid, a fraction of the which must be conetant withiti each time interval. Output gener- melt flows radially outward under the action of capillary forces i I ated includes the level of radioactivity in each node of the con- and freezes near the radial boundary; (36) in a nonwetting fainment and in the environment, and radiation doses to refer- system, all of the malt flows into the bottom of the bed; and (4) ence individuals at up to three different receptor points. when Zr and Fe are in intimate contact and the Zr atomic frac-tion is greater than 0.33, these metals can liquefy and flow out NUREG/CR-5107: HYDROGEOLOGIC CHARACTERl2ATION OF of the bed very early in the meltdown sequence. Major uncer. BASALTS.The Northern Rim Of The Columbia Plateau Physio- tainties in the analysis are identified and validation experiments graphic Province And Of The Creston Study Aree. Eastern are discussed. Washington. STEELE,T.D.; PASCHIS.J.A.; KOENIG,R.A. In-Situ, Inc. March 1988. 242pp. 8804080218. 45042:104. NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON This report provides a general bu; comprehensive character

• THE DRAFT REACTOR RISK REFERENCE ization of hydrogeologic and hydrogeochemical baseline conds- DOCUMENT,NUREG-1150. KASTENBERG.W.E.;

tions for the Creston area located along the northem fim of the APOSTOLAKIS.G.; BICKEL,J.H.; et at Lawrence Livermore Na-Columbia Plateau physiographic province. Historical as well as tional Laboratory. May 1988. 212pp. 8806230152. UCID-21346. recent data and other available information from previous stud- 45900:160. ies and alternative sources have been considered in this base- This report contains the fiading of the "NUREG-1150 Peer line hydrogeological characterization. These include data and in- Review Panet" on the methodology and results presented in the formation on water levels, aquifer characteristics, and water Draft Reactor Risk Reference Study (NUREG-1150) and sup-quality for shallow basalt units comprising the Wanapum Forma- porting documentation. The report consists of two principal tion and the Grande Ronde Formation in the Creston study area parts: the chairman's report, which summarizes the principal and for the general region surrounding this stuoy area. The issues of the review, and the individual member's reports, which overall goal of this hydrologic characterization was to provide supply the individual findings and technical details of the four-useful information leading to the selection of the Roza Member teen reviewers. of the Wanapum Formation as the study's basalt horizon and for other related, subsequent study components of in Situ's re- NUREG/CR-5119: METALLOGRAPHIC EXAMINATION OF THE search project. SEVERE FUEL DAMAGE SCOPING TEST (SFD-ST) FUEL ROD BUNDLE CROSS SECTIONS. HOFMANN,P. EG&G Idaho, NUREG/CR-5108: Th2RMODYNAMIC PROPERTIES OF TC(IV) Inc. (subs. of EG&G Inc.). July 1988.122pp. 8808180220. OXIDES.Solubilit.es And The Electrode Potential Of The Tc(Vil)/ EGG-2537. 46535:345. Tc(IV)-Oxide Couple. MEYER,R.E.; ARNOLD,W.D.; CASE,F.I.; et A series of severe fuel damage (SFD) tests was performed in al. Oak Ridge National Laboratory. May 1988. 33pp. the Power Burst Facility (PBF) at the Idaho National Engineering 8808050226. ORNL-6480. 46388:330. Laboratory (INEL) to obtain data necessary to understand: (a) l Solubihties of precipitated Tc(fV) oxides have been deter. fission product release, deposition, and transport; (b) hydrogen mined in solutions of Nacl in the pH range 7 to 9. Technetium release; and (c) fuel behawor and coolability during severe fuel oxide was prepared by precipitation upon purified sand with a damage accidents. This report presents the results from the 30*4 excess of hydrazine. The oxide-covered sand was then metallographic posttest examination of the fuel rod bundle cross transferred to a small column and solutions of Nacl varying in s sections of the first experiment in the SFD series, the Scoping concentration from 0.1 mol/L to 2.6 mol/L were continuously Test (SFD-ST). The metallographic appearance of the zircaloy cycled through the column. The concentrations of Tc(IV) spe- cladding, solidified melt, UO(2) fuel, inconel spacer grids, and cies in equilibrium with the oxide were determined and found to other instrumentation and structural components is discussed. be approximately constant in the Nacl concentration range 0.1 The observed structures are compared to those documented in mol/L to 2.6 mol/L and averaged (2.61 plus or minus 1.00) x well defined, out of-pile, separate- effects reference experi-10(-9) mol/L. A solubility product of (1.63 plus or minus 0.62) x ments. The chemical interactions that take place at high tem-10(-33) was calculated from the measured solubihties of Tc(fV) peratures between the fuel bundle components are described; oxide. Because there is some disagreement among values of E and estimates of peak test temperatures are made, based only degree for the Tc(Vit) - Tc(IV) redox couple in the titerature, a on metallography since scanning electron microscopy results redetermination of the standard potential was completed. The were not available. slopes of the graphs of electrode potential vs pH and concen. tration were confirmed, and the value of E degree was deter. NUREG/CR-5120: A MODEL FOR THE TRANSPORT AND mined to be 0.747 plus or minus 0.004 V. These data can be CHEMICAL REACTION OF MOLTEN DEBRIS IN DIRECT CON-used to estimate Tc(IV) solubilities for cases where solubihty TAINMENT HEATING EXPERIMENTS. MARX,K.D. Sandia Na-limits transport of technetium in reducing environments of high- tional Laboratories. May 1988. 54pp. 8808230424. SANDB8-level waste repositories. 8213. 46581:083. NUREG/CR-5109: RELOCATION OF METALLIC CONSTITUENTS A computer model is described which simulates the effects of IN CORE DEBRIS BEDS. DOSANJH,S S. Sandia National Lab- releasing molten debris into a gas filled containe . This work is oratories. September 1988. 86pp. 8810050256. SAND 88-0535. motivated by studies of direct containment heat ng due to the 46987:081. dispersal of debris produced in certain nuclear ri tector accident The MELPROG computer code is being developed to provide scenarios. The model consists of a finite-differer ce scheme for mechanistic treatment of Light Water Reactor (LWR) accidents the gas flow coupled with a Lagrangian particle transport algo-from accident initiation through vessel failure. This paper de- nthm. It computes the transpod of the debris through the gas

I 1 1 Main Citations and Abstracts 47 and evaluates radiative and convective heat transfer effects. It provided. This includes a study of plugs that have dried, as well also accounts for the chemical reaction of the debris with the as of plugs that have remained wet throughout the testing oxygen in the atmosphere, including the concurrent heat re- period. An introductory literature review indicates that deep un-lease. The resulting computer code is used to simulate experi- derground structures in competent rock are safer than surface ments in the Surtsey Direct Heating Test Facihty. It is found that structures, openings at shallow depth, and openings in fractured the computational results agree well wtth expenment for modest rocks, when subjected to earthquakes and subsurface blasts. debns fluxes. It is further shown that the simulation of configura- Cement plugs are installed in 2.5 cm diameter coaxial holes in tions with large fluxes can be improved with better submodels 15 cm diameter granite cylinders. Water is injected under pres-to describe the debns behavior. The description f the interac- sure on top of the plugs and is collected below the plugs. Hy-tion of the debris with the container walls is of pubcular impor- draulic conductivities are calculated. Once a long-term steady-tance. state flow trend has been established, the sample are subjected NUREG/CR-5123: STUDIES 0F THE PATTERN AND AGES OF to dynamic loading on a shaking table. Shaking is performed at POST-METAMORPHIC FAULTS IN THE PIEDMONT OF VIR. accelerations up to 2 g and for up to 300 seconds. Wet cement GINIA AND NORTH CAROLINA. GLOVER,L.; COSTAIN,J.K.; seals are less permeable than intact Charcoal granite. Sealing CORUH.C. Virginia Polytechnic Institute & State Univ., Blacks. performance can deg'ade severely when cement seals are al-burg, VA. April 1988. 67pp. 8805060296. 45430:226. Iowed to, dry. Dye injection shows very limited and uniform pen-A geologic corridor from the Blue Ridge to the eastern Pied. etration into wet plugs, but strongly preferential flow along the mont of Virginia is integrated into a tectonic model and extrapo. plug / rock interface of dried plugs. The permeability of wet and lated downward 10 to 15 km by means of seismic reflection and of rewetted previous;y dned cement seals does not change sig-gravity studies. The Blue Ridge appears to be a hinge zone that nificantly after the application of dynamic loads. Sealing in an faced a rift- generated lapetus Ocean. An Eastem continent unsaturated environment may affect the drying (curing, aging) with an Eocambrian and Cambrian magmatic arc and sediments conditions of cementitious seals, as well as the structure of of the same age, collided with the North American continental earthen seals. An unsaturated environment will need to be inte-margin in the Middle and Late Ordovician. Subsequent Devono. grated realistically into sealing performance tests and analyses. , Mississippian and Mississippian-Permian orogenesis continued i to dnve thin thrust nappes onto North America. Early Mesozoic NUREG/CR-5130: BENTONITE BOREHOLE PLUG FLOW TEST-rift basins record the beginning of the Atlantic basin and, from ING WITH FIVE WATER TYPES. GAUDETTE,M.V.; Middle Jurassic to Present, the margin of North America was DAEMEN,J.J. Anzona, Univ. of, Tucson, AZ. April 1988.223pp. covered by Coastal Plain sediments. Several constrained hypo. 8805090139.45432:010. centers of the central Virginia seismic zone, adjacent to a re- The hydraulic conductivity has been determined of plugs con-flection profile, show an apparent relation to structure. We ten- structed with commercial precompressed bentonite pellets. Ben-tatively conclude that flat and ramp faults formed dunng Paleo- tonite has been hydrated and tested with waters of five different zoic nappe emplacement are currently being reactivated. The chemical compositions, includiag one groundwater (Ogallala sq-reactivation may be largely aseismic on the old thrust faults, but uifer, Texas). The groundwater contained a significant amount seismicity appears to be related to high angle transcurrent faults of solids; waters prepared in the laboratory did not. Prepared where new rock breakage may be occumng. waters used for testing included distilled water, a high (1000 NUREG/CR-5126: TAC 2D STUDIES OF MARK i CONTAINMENT ppm) and a low (45 ppm) calcium solution, and a 39 ppm DRYWELL SHELL MELT-THROUGH. WEINGARDT,J.J.; s dium water. Uncompacted plugs were constructed by drop-BERGERON,K.D. Sandia National Laboratories. August 1988. ping bentonite tablets into waterfilled cyhnders, or by mixing powdered bentonite with preselected water volumes in order to 67pp. 8809140357. SAND 881407. 46802:203. i A series of parametnc calculations of the thermal attack of obtain controlled initial water contents. The hydraulic conductive molten corium on a steel shell has been performed with the ty of all plugs tested with all waters would result in a classifica-TAC 2D computer code in order to elucidate uncertainties about tion of practically impervious, by conventional soil mechanics j the survivability of the BWR Mark I containment boundary in the standards. Variations of several orders of magnitude of the hy- J event of a core-melt accident. Since TAC 2D is a two dimension- draulic conduct'vity are observed. Clay plugs constructed from al heat conduction code, it is not possible to capture some of bentonite tablets hydrated with unfiltered Ogallala groundwater the complexities of the corium spreading process or the debris- exhibited reduced swelling and lower hydraulic conductivity than l concrete interactions which would occur in this accident scenar- similar plugs constructed from tablets and distilled water. The  ; io. However, the two-dimensional transient nature of the thermal differences in observed conductivities may be governed by fac-attack is modeled better with TAC 2D than is possible with exist- tors such as swelling characteristics and permeant colloidal, ing debris-concrete interaction codes. This study was therefore matter clogging plug pore spaces. Constant pressure injection undertaken as a supplement to earlier work with debris concrete and transient pulse testing methods have been tried to deter-interaction codes (like CORCON-MOD 2), with the intentsors of mine hydraulic conductivity. Especially in constant pressure in-assisting members of expert panels assessing uncertainty in jection tests, outflow volumes may require adjustment to ac-severe accident phenomena for the US Nuclear Regulatory count for consolidation drainage. Attempts made here at such Commission's NUREG-1150 project. A total of 23 calculations corrections have not been successful. Consolidation testing re-are reported, consisting of two base cases (one with overlying quires extremely long time periods to approach a constant limit. water and one without) and numerous sensitivity variations it appears probable that the flow tests have not been pysued about each case. Sensitivities investigate 1 ' Jde mixed versus for a sufficient time to assure complete determination of water layered conum, heat transfer paramete' ward and down- chernistry effects on hydraulic conductivity. ward, initial conum temperature, chemie aesting rate, heat transfer conditions in the gap outside the shell, and corium NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANS-depth. FER AND FLOW VISUALIZATION STUDIES AND TECH-NIQUES RELEVANT TO THE STUDY OF EROSlON-CORRO-NUREG/CR-5129: EXPERIMENTAL ASSESSMENT OF THE IN- SiON OF REACTOR PIPING SYSTEMS. KUZAY,T.M. FLUENCE OF DYNAMIC LOADING ON THE PERMEABILITY HALLE H J.; KASZA.K.E. Argonne National Laboratory. July OF WET AND OF DRIED CEMENT BOREHOLE SEALS. 1988. 28pp. 8809060198. ANL-88-20. 46687:339. ADISOMA.G.; DAEMEN J.J. Anzona, Univ. of, Tucson, AZ. April This report provides some background information on the 1988. 279pp. 8805090129. 45433:011. failed piping at the Surry-2 reactor; a summary of pertinent liter-An expenmental sealing performance assessment of cement sture on mass transfer in related geometries; and a description borehole plugs that have been subjected to dynamic loading is of methodologies for visualization and erosion rate measure- _ ___o

48 Main Citations and Abstracts ments in laboratory model studies that can provide greater in- tion and hence overly conservative (rapid) vessel cooldown sight into the role of flow geometry in erosion corrosion. rates. Using a corrected version of this code, our new calcula-tions now exhibit flow circulation. However, parametric analysis NUREG/CR-5132: SEVERE ACCIDENT INSIGHTS REPORT. of less likely (more equ:pment failure PORVs,HPl pumps) sce-PRATT,W.T. Brookhaven National Laboratory. April 1988.42pp. . narios revealed that ' flow stagnation was indeed possible but 8809150278. BNL-NUREG-52143. 46803:061. could only occur at lower pressures. This simple mapping pro-This report describes the conditions and events that nuclear cedure has been favorsbly benchmarked against the (TRAC) power plant personnel maw encounter during the latter stages of

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system calculations. This tool is therefore useful for screening a severe core damage accident and what the consequences possible risk dominant SBLOCA scenarios in various PWR de-might be of actions they may take during these latter stages. signs. The report also describes what can be expected of the perform. ance of the key barriers to fission product release (primarily NUREG/CR-5136: FATIGUE STRENGTH OF SMOOTH AND containment systems), what decisions the operating staff may NOTCHED SPECIMENS OF ASME SA 106-B STEEL IN PWR face during the course of a severe accident, and what could ENVIRONMENTS. TERRELL.J B. Materials Engineering Associ- 1 result from these decisions based on our current state of knowl- ates, Inc. September 1988. 93pp. 8810110250. MEA-2289. j edge of severe accident phenomena. 47043:175. j NUREG/CR-5133: A COMPUTATIONAL MODEL FOR C3lTICAL Fatigue strain-hfe tests were conducted on ASME SA 106-B j FLOW THROUGH INTERGRANULAR STRESS CORROSION piping steel base metal and weld metal specimens in 288 de-CRACKS. SCHROCK V.E.; REVANKAR,S.T.; LEE.S.Y.; et al. grees C (550 degrees F) pressurized water reactor (PWR) envi-Califomia. Univ. of, Berkeley, CA. November 1988. 103pp. ronments as a function of strain amplitude, strain ratio, notch ~ 8812010060. LBL-21967. 47684:339. acuity, and cyclic frequency. Notched base metal specimens The presence of intergranular stress corrosion cracks tested at 0.017 Hz in 1.0 part per billion (ppb) dissolved oxygen (IGSCC) in thermal stressed zones in stainless steel piping and environments nearly completely used up the margins of safety associated components is of much concern in reactor safety. of 2 on stress and 20 on cycles incorporated into the ASME The prediction of leak rates through the cracks is important in Section ill design rurve for carbon steels. Tests conducted with assessing the plant reliability. An analytical model has been de- specimens having theoretical notch concentration (K(t)) values veloped to predict flow rates of initially subcooled or saturated ranging from 2 to 6 showed virtually no effect of notch acuity on water through these cracks. The model assumes the flow in the fatigue life. Tests conducted with smooth base and weld metal cracir to be homogeneous and in thermal equi;ibrium. The crack specimens at 1.0 Hz showed virtually no degradation in cycles geometry was idealized as a convergent straight slit of constant to failure when compared to 288 degrees C air test results. In gap thickness. The fluid is assumed to enter the crack without all cases, however, the effect of temperature alone reduces the separation. The one rtimensional model accounts for the chang- margin of safety offered by the design curve in the low cycle ing cross sectional area in the flow direction. The effects of wall regime for the test specimens. Comparison between the fatigue inction, expansions / contractions and tortuosity of the actual life results of smooth and notched specimens suggests that fa-flow path are lumped into an equivalent friction. The numerical tigue crack initiation is not significantly affected by 1.0 ppb dis-scheme developed for the model solutior' has been pro- solved oxygen, and that most of the observed degradation may grammed into a Fortran computer code called SOURCE. A be attributed to crack growth acceleration. These results sug-companion subroutine STEAM provides the saturated fluid prop- gest that the ASME Section 111 methodology should be revised erties. Inputs to SOURCE are the upstream stagnation pressure to account for PWR environment variables which degrade fa-and temperature, the crack geometry specification, and the tigue life of pressure retaining components. j equivalent friction factor. NUREG/CR-5137: BIODEGRADATION TESTING OF TMI-2 idREG/CR-5134: APPLICATION OF ACOUSTIC LEAK DETEC- EPICOR-Il WASTE FORMS. ROGERS.R.D.; MCCONNELL,J.W. TION TECHNOLOGY FOR THE DETECTION AND LOCATION EG&G idaho, Inc. (subs. of EG&G, Inc.). June 1988. 40pp. OF LEAKS IN LIGHT WATER REACTORS. KUPPERMAN,D.S.; PRINE.D.; MATHIESON,T. Argonne National Laboratory. Octo- 8807070421. EGG-2540. 46011:344. ber 1988.146pp. 8811110044. ANL 88-21. 47523:062. Waste-form testing is being conducted by the Three Mile Island Unit 2 (TML2) EPICOR-il Resin / Liner investigation; Low-This report presents the results of a study to evaluate the Level Waste Data Base Development Program of the U.S. Nu-adequacy of leak detection systems in hght water reactors. The sources of numerous reported leaks and methods of detection clear Regulatory Commission (NRC) in accordance with the NRC Branch Technical Position on Waste Form. Waste forms have been documented. Research to advance the state of the which were tested contain ion exchange resins solidified with art of acoustic leak detection is presented, and procedures for vinyl ester-styrene and Portland Type I-11 cement. This report implementation are discussed. describes the biodegradation testing of those waste forms, pre-NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS sents the test results, and provides recommendations for alter. RELATIVE TO PRESSURIZED THERMAL SHOCK. nate tes; methodology. THEOFANOUS.T.G. Cahfornia, Univ. of, Santa Barbara, CA. LA CHANCE.JL.; WILLIAMS,K.A.; et al. Science Applications intor- NUREGICR-5138: VALIDATION OF GENERIC COST ESTi-national Corp. (formerly Science Applications, Inc.). October MATES FOR CONSTRUCTION RELATED ACTIVITIES AT NU-1988. 55pp. 8812020179. LA-UR-88 2871. 47685:239. CLEAR POWER PLANTS. Final Report. SIMION,G.; The U.S. Nuclear Regulatory Commission (NRC) Pressurized SCIACCA,F.; CLAIBORNE,E,; et al Science & Engineering As-Thermal Shock (PTS) study had previously identified Small sociates, Inc. May 1988. 53pp. 8805200123. SEA 87-253-04A1. l 45561:334. Break Loss of Coolant Accidents (SBLOCAs) as a risk dominate accident scenario due to (numerically calculated) primary loop This report represents a validation study of the cost method- 1 flow stagnation at high pressure. The objectives of the present ologies and quantitative factors derived in " Labor Productivity ) effort were two-fold. First to develop a physically-based under- Adjustment Factors" (NUREG/CR 4546) and " Generic Method-  ; standing of controlling thermal- hydraulic phenomena producing ology for Estimating the Labor Cost Associated with the Remov. such PTS SBLOCA stagnation scenanos. Secondly, to use al of Hardware, Materials, and Structures From Nuclear Power i these insights in developing a simple (computationally efficient) Plants" (SEA Report 84- 116 05 A:1). This cost methodology

• mapping" tool to quantify the occurrence and thermal behavior was developed to support NRC analysts in determining genene of such high pressure flow stagnation regimes. Review of the estimates of removal, installation, and total labor costs for con-previous (TRAC) calculations revealed that inaccurale modeling struction related activities at nuclear generating stations. In ad-of vapor condensation erroneously produced the flow stagna- diion to the validation discussion, this report reviews the gener-t

_ _ - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . _ _____ --_ _ _ a

i I i Main Citations and Abstracts .49 j ic cost analysis methodology employed. It afsn discusses each Central to 1h'e understanding of nuclear reactor containment j of the individual cost factors used in estimating the costs of vessel fracture safe design is the development of techniques to

     - physical modifications at nuclear power plants. The generic esti-         charactenze and predict the fracture of ductile metals in the mating approach presented uses the "greenfield" or new plant              upper region of the ductile to brittle toughness transition. This construction installation costs compiled in the Energy Economic           project has utilized the J integral parameter to develop tough-Data Base (EEDB) as a baseline. These baseline costs are then                                                                                l ness transition information for static and dynamic loading rates   q adjusted to account for labor productivity, radiation fields, leam-    ' throughout the toughness transition for an A533 grade B class -     a
    , ing curve ettects, and impacts on ancillary systems or compo.              1 containment vessel steel. The J at cleavage (J(civ)) is evalu-nents. For comparisons of estimated vs actual labor costs, ap .           ated when cleavage occurs, and the J(Ic) is evaluated when proximately four dozen actual cost data points (as reported by           ductile crack extension is present, in many specimens, for both 14 nuclear utilities) were ot;tained. Detailed background informa-       static and dynamic loading rates, both J(Ic) and J(civ) are evalu-tion was collected on each individual data point to give the best         ated and plotted as a function of test temperature and test rate._

understanding possible so that the labor productivity factors, re- A principal conclusion is that if dynamic loading rates are used, moval factors, etc. could judiciously be chosen. This study con- very little size dependence seems to be present, though larger cludes that cost estimates that are typically within 40% of the specimens produce lower bound results. Close agreement is actual values can be generated by prudently using the method- also shown between dynamic initiation cleavage toughness ologies and cost factors investigated hwein. values and the arrest tought.ess K(la) values obtained by NBS (Gaithersburg) dynamic crack arrest tests conducted on large NUREG/CR-5140: VALUE-lMPACT ANALYSIS FOR EXTENSION  : edge cracked plates of this same steel alloy. OF NRC BULLETIN 85-03 TO COVER ALL SAFETY-RELATED . MOVS. HIGGINS,J.C.; RUGER,C.J.; MACDOUGALL,E.A.: et al. NUREG/CR-5144: ACOUSTIC EMISSION SYSTEM CAltBRA-Brookhaven National Laboratory. July 1988.90pp.8901030137. TlON AT WATTS BAR . UNIT 1. NUCLEAR E REACTOR. BNL-NUREG-52145. 47961:252. HUTTON,P.H.; FRIESEL,M.A.; DAWSON J.F.; et al. Battelle Me. BuIletin 85-03 is entitled Motor Operated Valve (MOV) morial Institute, Pacific Northwest Laboratory. August 1988. Common Mode Failures During Plant Transients Due to improp- 38pp. 8811010179. PNL-6549. 47271:285, er Switch Settings. The Bulletin requires that NRC licensees im- An acoustic emission system has been installed on TVA's plement a detailed program to ensure that MOV operator switch Watts Bar Unit 1 reactor to monitor selected areas of the pres-settings are set properly in two particular safety-related systems sure boundary during cold hydrostatic testing, hot functional per operating nuclear power plant. The proposal analyzed testing, and ultimately, during reactor startup and operation. The herein requires licersees to extend this detailed program to woik is part of a cooperative effort between TVA and NRC Re-ensure proper MOV switch settings for all safety- related MOVs. search to test and demonstrate AE technology. This technology A Value-impact Ana'ysis was performed, according to the guide- has been develop (d under an NRC Research program to vali-lines in NUREG/CR-3568, "A Handbook for Value-Impact As- date the application of AE techniques for continuous, on-line sessment," to determine the cost effectiveness of extending monitoring of reactor pressure boundaries to detect cracking. NRC Bulletin 85-03, consistent with providing protection of the The performance of and results from a special calibration test public health and safety, to cover all safety-related MOVs. The of the AE system using simulated AE signals to evaluate the ac-results show that both a very favorable dose aversion and dollar curacy of signal source location by the system is discussed in savings can be realized if the proposed actions are implement- this report. ed Also, the proposed actions can be justified based on exist-ing requirements (for example,10CFR50 Apper. dix B and Ge- NUREG/CR-5145: FAILURE INVESTIGATION OF 3M SERIES neric Letter 83-28). 900 STATIC ELIMINATORS. CZAJKOWSKl,C. Brookhaven Na-tional Laboratory. July 1988. 60pp. 8808300251. BNL-NUREG-NUREG/CR 5141: AGINC AND OUAllFICATION RESEARCH ON 52146.46631:196. SOLENOID OPERATED VALVES. BACANSKAS,V.P ; Numerous instances of facility contamination by polonium 210 TOMAN.G.J.; CARFAGNO,S.P. Franklin Institute. August 1988. microspheres have been reported. These contamination events 177pp, 8808230415. 46575:145. appear to be caused by leakage or ejection of the radioactive . A research program was conducted on the aging and qualifi-cation of solenoid operated valves (SOVs). Some SOVs had material from static elimination devices manufactured by the q Minnesota Mining and Manufacturing Company (3M). The Ash- 1 been aged naturally through service in nuclear power plants and land Chemical Company (Earton, Pennsylvania) was the first fa-  ! others were subjected to accelerated aging. Thermal aging was cility determined to be contaminated and was subsequently sub-conducted both with air and nitrogen as the process gas. Oper-jected to a comprehension review by the U.S. Nuclear Regula-ational aging was simulated by putting the specimens through tory Commission (USNRC). Therefore, devices from this facility operational cycles. The program included simulation of a design were chosen for examination. A failure investigation has been basis event (DBE), that consisted of DBE gamma irradiation and performed on six static eliminators The investigation consisted a 30-d main steam line break / loss of coolant accident (MSLB/ of visual inspection, sectioning and scanning electron microsco-LOCA) simulation. A naturally aged. ASCO SOV with Buna N i py of six devices. It is concluded that non-uniform and imperfect i seals and a new ASCO SOV, with EPDM seals, subjected to ac- microspheres appear to have been manufactured and installed in l celerated aging with nitrogen as the process gas, were the only - devices. Rough handling may induce loosening of tt e micros = ones to perform satisfactorily throughout the test program. Faib i pheres and the epoxy binder used in device manufacture may j ures to transfer of other ASCO SOVs appeared to be caused by not be suitable for the varied service conditions. coil deterioration, not by seal deterioration. Valcor SOVr suf- t fered from sticking of the shaft seal O-rings, making it impossi. NUREG/CR-5146: DEBRIS DISPERSAL FROM REACTOR CAV- I ble to complete accelerated thermal aging. Seal deterioration in ITIES DURING HiGH-PRESSURE MELT EJECTION ACCIDENT i the Valcor SOVs caused leakage and delays in transferring fol- SCENARIOS. TUTU,N.K.; GINSBERG,T.: FINFROCK,C.; et al. towing DBE irradiation. Valcor SOVs performed satisfactorily Brookhaven National Laboratory. July 1986.72pp.8811070088. j during several hours of the MSLB/LOCA simulation, but mal- BNL-NUREG-52147. 47387:307, 4 functioned during most of the rest of the test. This report presents the results of a scoping experimental study of the " extent of molten debris dispersal" from PWR re. NUREG/CR-5142: DUCTILE TO BRITTLE TOUGHNESS TRANSh actor cavities under direct containment heahng conditions. Sim-TION CHARACTERIZATION OF A533B STEEL. JOYCE,J.A. utated high-pressure melt ejection experiments were conducted David W. Taylor Naval Research & Development Center. June

' 1988. 37pp. 8807110516. 46099:318.                                        using 1/42nd-scale models of reactor cavities and were de-              1 signed to employ low-temperature melt simulants. Three "repre.

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50 Main Citations and Abstracts sentative" reactor cavities were selected: Zion, Surry, and Watts This report describes the concept of performance-based in. Bar. A scaling analysis of the debns dispersal phenomena was spections that is being taught in NRC's training course, "in-carried out and was employed to extrapolate the expenmental specting for Performance." This concept has been endorsed results for " debris dispersal fraction" to accident conditions. and is being implemented by the Nuclear Regulatory Commis-The experimental results, along with the interpretation based sion (NRC'). NRC performance-based inspections concentrate upon the scaling analysis, suggest that the three reactor cav- on activities that impact plant reliability and safety. The inspec-itses studied here, i.e., Zion, Surry, and Watts Bar, would retain tions begin with performance- based observations, and then the little, if any, of the melt ejected into them from the reactor inspectors let discrepancies or uncertainties lead to inspection vessel following the high-pressure steam blowdown which of other areas, such as quality verification organization effec-would follow vessel failure and men release. tiveness, training adequacy, and procedural controls. This in-spection approach departs from past NRC practices that em-NUREG/CR-5147: FUNDAMENTAL ATTRIBUTES OF A PRACTI- phasized documentation and prograti, review as a means to CAL CONFIGURATION MANAGEMENT PROGRAM FOR NU- measure operational mfety. The first goal of performance-based CLEAR PLANT DESIGN CONTROL, KLEIN,S.M. ERC Intema- inspections is to improve the NRC inspector's ability to accu-tional, Inc. June 1988. 80pp. 8807080362. 46077:265. rately evaluate plant safety and reliability. This goal will be This report summarizes the results of an evaluation of find- achieved by increasing the inspector's emphasis on actual ob-ings identified during a number of Safety-System Functional in- servation of ongoing facility work activities and reducing the em-spections and Safety System Outage Modification Inspections phasis on document and program reviews. The secondary goal which are related to configuration management. A computerized of performance-based level inspections is to encourage licens-database of these findings was generated from a review of the ces to manage their facilities in a performance-based manner. design inspection reports. Based on the results of the evalua-tion, attributes of a configuration management program were NUREG/CR-5153: THE TEACHABILITY AND MECHANICAL IN-developed which are responsive to minimizing these types of in- TEGRITY OF SIMULATED DECONTAMINATION RESIN spection findings. Incorporation of these key attributes is con- WASTES SOLIDIFIED IN CEMENT AND VINYL ESTER STY-sidered good practice in the development of a configuration RENE. SOO,P.; MILIAN,L.W.; PICIULO.P.L. Brookhaven Nation-management program at operating nuclear plants. al Laboratory. May 1988. 88pp. 8811070049. BNL-NUREG-52149.47390:145. NUREG/CR-5149: EROSION-CORROSION OF PWR FEED- The rates of release of organic doct .nination reagents WATER PIPING SURVEY OF EXPERIENCE, DESIGN, WATER CHEMISTRY AND MATERIALS. JONAS,0. Argonne National were measured for mixed bed resins solified in Pertland cement and vinyl ester-styrene. The i,ntent is to evaluate the effect of Laboratory. JONAS.O. Jonas, Inc. June 1988. 101pp. composite size on the release rate to determine if experiments 8809060200. ANL-88-23. 46687.238. on small laboratory samples can be used to predict the behav-This report presents the results of a survey of 26 PWRs in ior of full size waste forms. The effects of chelating agents on the U.S. The survey determined the frequency of erosion-corro-sion in these reactors and the corresponding design and operat- the release of Co-60 from cementitious waste forms is ad-dressed. Cracking during leaching was also investigated and ing parameters of interest, including (1) feedwater velocities, reae,ons for its occurrence outlined. temperatures, and pressures, (2) water chemistry charactens-tics, and (3) the materials of construction. These data are re- NUREG/CR-5154: EXPERIMENTAL ASSESSMENT OF DAMP. viewed in terms of the known effects of these parameters on ING IN LOW ASPECT RATIO, REINFORCED CONCRETE er ion orrosion and used to develop recommendations for SHEAR WALL STRUCTURES. FARRAR,C.R.; BENNETT.J.G. Los Alamos National Laboratory. August 1988. 40pp. NUREG/CR-5150: STEAM GENERATOR OPERATING 8808240028. LA-11325-MS. 46578:348. EXPERIENCE. Update For 1984 1986. FRANK,L.; STOKLEY,J. This report summarizes the experimental data obtained from Science Applications international Corp. (formerly Science Ap- the Seismic Category i Structures Program concerning damping plications, Inc.). June 1988. 85pp. 8807110505. SAIC 87/3014, in low aspect ratio, reinforced concrete shear wall structures. 46099.233. This program, that is sponsored by the United States Nuclear This report summarizes operatcnal events and degradation Regulatory Commission, has tested 37 shear wall structures mechanisms affecting pressunzed water reactor steam genera- and structures and structural elements both statically (mono-tor integnty, provides updated inspection results reported in tonic and cyclic) and dynamically (sine sweep, random, simulat-1984,1985, and 1986, and highlights both prevalent problem ed seismic, and impulse). Data from these tests have been ana-areas and advances in improved equipment test practices, pre. lyzed by four different methods to determine equivalent viscous l ventive measures, repair techniques, and replacement proce. damping ratios that can be used in the analysis of shear wall j structures. These methods are: (1) Frequency response function dqres. It describes equipment design features of the three major suppliers and discusses 68 plants in detail. Steam generator analysis, (2) The log decrement method, (3) The hysteretic degradation mechanisms include intergranular stress corrosion energy loss method, and (4) The flow response spectra match-cracking, primary water stress corrosion cracking, pitting, inter. ing method. The flow response matching method is, to the au-granular attack, and vibration wear that affects tube integnty thor's knowledge, new and provides the most general method and causes leakage Plugging, sleeving, heat treatment, peen. for assessing a variety of damping mechanisms. Results from ing, chemical cleaning, and steam generator replacements are the various methods were generally consistent and the damping described and regulatory instruments and inspection guidelines values w3re found to be in the range specified by current regu-for non-destructive evaluations and girth weld cracking are dis. latory guides. A discussion of the various damping mechanisms, cussed. The report concludes that although degradation mecha- how damping mechanisms affect the equations of motion, the nisms are generally understood, the ehmination of unscheduled effects of the type of loading on the vanous methods used to plant s' Otdowns and costly repairs resulting from leaking tubes determine the damping, and other investigators' results are also has not 5een achieved. Highlights of Fleam generator research presented. and unre solved safety issues are discussed. NUREG/CR-5156: REVIEW OF EROSION-CORROSION IN NUREG/CR-5151: PERFORMANCE-BASED INSPECTIONS. SINGLE PHASE FLOWS. CRAGNOLINO,G.; CZAJKOWSKl,C.; HAWKINS,F. NRC - No Detailed Affibation Given. JOHNSON.J. SHACK,W.J.; et al. Brookhaven National Laboratory. June 1988. Weirich & Associates. LINER.R.T ; et al. Science Appl 6ons 107pp. 8809220039. ANL-88 25. 46878:263. Intemational Corp. (formerly Science Applications, Inc.). June This report contains two literature reviews (prepared by 1988. 27pp. 8806230162. SAIC-88/3014. 45907:234. Brookhaven National Laboratory and Argonne National Labora-

l Main Citations and Abstracts 51 tory, respectively) on the available data and current mechanistic used to predict wear and fatigue damage. Reducers were found understanding of erosion-corrosion, and a failure analysis (pre- to cause little or no performance degradation. Elbow effects pared by Brookhaven National Laboratory) of a tee-elbow joint must be considered when located within 5 diameters of the from the Surry Unit 2 reactor that failed by erosion-corrosion in check valve, while severe turbulence sources have significant December 1986. It also includes suggestions for additional re- effect at distances to 10 diameters. Clearway swing check de-search that should be performed by the USNRC to increase the signs were found to be particularly sensitive to manufacturing j capability to rank plants and/or location within plants in terms of tolerances and installation variables making them likely candi- i susceptibility to erosion- corrosion and to ensure that proposed dates for premature failure. Reducing the disk full opening angle inspection and mitigation programs are soundly based. on these clearway designs results in significant performance im-provement. NUhEG/CR-5157: THE DEVELOPMENT OF APRIL. MOD 2 - A COMPUTER CODE FOR CORE MELTDOWN ACCIDENT NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB I ANALYSIS OF BOILING WATER NUCLEAR REACTORS. AT THE SUB-COMPONENT AND SUBSYSTEM LEVEL. Final KIM,S.H.; KIM.D.H.; KOH,B.R.; et el Oak Ridge National Labo- Report. ROBINSON,T.; SIMlON,G.; SCIACCA,F.: et al. Science ratory. June 1988. 285pp. 8808180257. ORNLSUB819089/3. & Engineering Associates, Inc. May 1988 72pp. 8806230166. SEA 87-253-06A1. 45898:232. Th s r port summarizes the thermal-hydraulic models used in The Energy Economic Data Base (EEDB) provides complete APRILMOD2. This digital computer code was developed at construction cost estimates for boiling water reactors (BWRs) Rensselaer Polytechnic institute (RPl) under joint sponsorship by the Empire State Electric Energy Research Corporation and pressurized water reactom (PWRs). The genenc cost esti-(ESEERCO) and the Boiling Water Reactor Severe Accident mating methods developed for the NRC utilize the EEDB cost Technology Program at Oak Ridge National Laboratory (ORNL). data as a basis for et,timating the costs of physical modifica-The APRIL. MOD 2 code can be used for the mechanistic analy, tions to nuclear plants. Such modifications may be mandated by sis of severe accident phenornena in Boiling Water Nuclear Re_ new or revised NRC regulatory requirements. The costs pre-actors (BWRs) having either Mark-l or Mark-Il containment sys. sented in the EEDB are often given at a rettatively high level of tems. aggregation. The definition of material, labor, and equipment costs is seldom given at the same level as that encountered in NUREG/CR-5158 V01: WORLDWIDE ACTIVITIES ON THE RE- plant modification projects. Additional definition is needed to DUCTION OF OCCUPATIONAL EXPOSURE AT NUCLEAR adequately estimate the costs of many proposed changes. This POWER PLANTS. KHAN,T.A.; BAUM.J.W. Brookhaven National report presents material and equipment costs and labor man-Laboratory. June 1988.46pp.8809220014. BNL-NUREG-52086. hours / costs at the component, subcomponent, and subsystem 46884:193. level that support and correspord to the more highly aggregat-This report is based on analysis of aninformational data base ed data presented in the EEDB. It is intended to be a supple-set up at the Brookhaven National Laboratory ALARA Center. It ment to the EEDB. Specifically, unit labor and material / equip-is pad of a project sponsored by the U.S. Nuclear Regulatory ment costs are defined for the following types of material and Comrnission to monitor and evaluate research on dose reduc- equipment: Rotating machinery, piping related commodities tion at nuclear power plants in the U.S. and abroad. The main (piping, valves, hangers, and insulation), instrumentation and benefits to be expected from reducing occupational exposures control, lighting and service power, skid mounted equipment. are highlighted in the report, the chief causes of elevated doses are identified, and effective approaches to minimize radiatio,n NUREG/CR-5162: CHARM:A MODEL FOR AEROSOL BEHAV-exposures are proposed. A wide range of research activity is IOR IN TIME VARYING THERMAL-HYDRAULIC CONDITIONS. covered, including plant chemistry, cobalt reduction WHEATLEY,C.J. Sandia National Laboratories. August 1988. techniques, stress corrosion cracking, decontamination, remote 165pp. 8810030122. SAND 88-0745. 46973:012. tools and devices, and, robotics. Advanced reactors, which are CHARM is a computer model for the behaviot of 8 one com- i designed for low radiation exposures, are examined, and health ponent aerosol in a single region with time-varying external cort  ! physics techw ogy programs which have been effective in re-ducing occupahonal exposure at various utilities are discussed. ditions. It treats particle gravity and turbulence, andagglomeration duedue particle deposition to to Brownian Browniarmotion $ The highlights of the programs on dose reduction conducted by motion, gravity, turbulence, thermophoresis and diffussophoresis. ' a number of countries are described, and compansons are Turbulence properties are estimated for flow through a region of made of the collective occupational radiation dose equivalents arbitrary cross- sectional shape, with aerodynamically rough or l for selected countries. The short and long term trends such studies are pointing to are evaluated. It is concluded that the smooth walls at any Reynolds number. The gas can be o' f pay  ! i efforts to improve dose reduction, both in the U.S. and abroad, remain in a heal h state but require continuing encouragement are: the temperature, pressure and velocity of the gas, wall P 'n-standard deviation of the source. The model is simply modified NUREG/CR-5159: PREDICTION OF CHECK VALVE PERFORM- to enable this list to be extended if needed. A new method of ANCE AND DEGRADATION IN NUCLEAR POWER PLANT solving the governing equations, based on the finite element SYSTEMS. KALSI M.S.; HORST,C.L.; WANG,J.K. Kalsi Engi- collocation method, enables the time-varying conditions to be neenng, Inc. May 1988. 78pp. 8806230210. KEl 1559. treated accurately and economically. We describe in detail: the 45897:233. models, the numerical methods, the execution of the computer Degradation and failure of swing check valves and resulting code (including how to write the input data file and interpret re-damage to plant equipment has led to a need to develop a suits), and how to make simple modifications to the model. We method to predict performance and degradation of these valves discuss how the model could be implemented as a sub-model in nuclear power plant systems. This Phase I investigation de- of a larger one and what further work needs to be done to veloped methods which can be used to predict the stability of enable It to efficiently treat multicomponent aerosols, and con-the check valve disk wnen piping disturbances such as elbows, densation onto and evaporation from parLcles. reducers, and generalized turbulence sources are present up-stream of the valve within 10 pipe diameters. Major findings in- NUREG/CR-5164: A SIMPLIFIED MODEL FOR CALCULATING clude the flow velocity required to achieve a full open, stable EARLY OFFSITE CONSEQUENCES FROM NUCLEAR REAC-disk position, the magnitude of disk motion developed with TOR ACCIDENTS. MADNI,l.K.; CAZZOLI,E.G.: KHATIB-these upstream disturbances with flow velocities below full open RAHBAR Brookhaven National Laboratory. July 1988.173pp. condrSons, as well as disk natural frequency data which can be 8809280214. BNL-NUREG-52153. 46954:054.

52 Main Citations and Abstracts A personal computes based model, SMART, has been devel- NUREG/CR-5170: A REVIEW OF RESEARCH CONDUCTED BY oped that uses an integral approach for calculating early offsite LOS ALAMOS NATIONAL LABORATORY FOR THE NRC consequences from nuclear reactor accidents. The solution pro- WITH EMPHASIS ON THE MAXEY FLATS,KY, SHALLOW - cedure uses simplified meteorology and involves direct analytic LAND WASTE BURIAL SITE. FOWLER E.B.; POLZER.W.L. Los integration of air concentration equations over time and posi- Alamos National Laboratory. August 1988. 69pp. 8809220040. tion. This is different from the discretization approach currently LA-11304-MS. 46884:124. used in the CRAC2 and MACCS codes. The SMART code is Studies to determine the impact of the Maxey Flats low-level fast-running, thereby providing a valuable tool for sensitivity and waste burial site on the environment have been conducted uncertainty studies. The code was benchmarking against both since 1963. Neither the migration or lack of migration of waste MACCS version 1.4 and CRAC2. Results of benchmarking and radionuclides to the off site by subsurface flow has been un-detailed sensitivity / uncertainty analyses using SMART are pre- equivocally proved; the migration of tritium may be an excep-sented. ton. Some radionuclides, e.g. plutonium, did migrate short dis-tances from a burial trench. The movement of "only a short dis-NUREG/CR-5165: SEISMOLOGICAL INVESTIGATION OF tance" was attributed to the importance of a biopopulation in EARTHQUAKES IN THE NEW MADRID SEISMIC ZONE AND the sod and to the importance of soluble iron as a competitor THE NORTHEASTERN EXTENT OF THE NEW MADRID SEIS. . for the chelate of the potentially mobile plutonium / chelate MIC ZONE. Final Report September 1981 - December 1986. system. In both cases a soluble plutonium complex is degraded HERRMANN.R.B.; TAYLOR,K.; NGUYEN.B. St. Louis Univ., St. and the plutonium is released in a form that is sorbed by the Louis, MO. July 1988. Sipp. 8808080121. 46408:231. soil. Soil moisture data in conjunction with tritium data indcate Earthquake activity in the Central Mississippi Valley has been that infiltration into the trench was predominately through the monitored by a seismograph network of eight stations in the trench cap. Tritium data also indicate subsurface migration of Wabash River Va!'ey and six stations in the New Madrid seismic trmun zone. This network is a component of a larger network jointly NUREG/CR-5171: FLOW VISUALIZATION STUDY OF POST sponsored by the NRC, USGS, universities and states. From CRITICAL HEAT FLUX REGION FOR INVERTED October 1981 to December 19861206 earthquakes were locat- BUBBLY, SLUG AND ANNULAR FLOW REGIMES. ed, of which 808 were in the New Madrid, Missouri area. Focal DENTEN,J.G.; ISHil,M. Argonne National Laboratory. November mechansms have been calculated for the June 10,1987 south- 1988. 85pp. 8812190120. ANL-88-27. 47805:148, I em Illinois earthquake using both P-wave first motions and long- A visual study of film boiling using still photographic and high-period surface-wave spectral amplitude data. The solution which speed motion picture methods was carried out in order to ana-best fit the surface wave data together with P-wave first motion lyze the post-CHF hydrodynamics for steady-state inlet pre-CHF data is one with a focal depth of 10 plus or minus 1 km, a seis- two-phase flow regimes. Pre-CHF two-phase flow regimes were i mic moment of 3.1 x 10(23) dyne-cm, and a focal mecharism established by introducing Freon 113 liquid and nitrogen gas characterized by a pressure axis that trends 89 degrees and into a jot core injection nozzle. An idealized, post-CHF two-plunges 4 degrees and a terision axis that trends 357 degrees phase core initial flow geometry (cylindrical multiphase jet core and plunges 24 degrees. The long-penod surface-wave and surrounded by a coaxial annulus of gas) was established at the strong ground motion accelerogram recordings of the January nozzle exit by introducing nitrogen gas into the annular gap be-31,1986 northeastern Ohio earthquake were used to estimate tween the jet nozzle two-phase effluent and the heated test the focal mechanism and source time function of the source. section inlet. For the present study three basic post-CHF flow j The surface-wave solution requires a source with a depth of 7 regimes have been observed: the rough wavy regime (inverted  ; km, a seismic moment of 1.1 x 10(23) dyne-cm, and a focal annular flow preliminary break down), the agitated regime (tran-mechanism characterized by a pressure axis that trends 336 de- sition between inverted annular and dispersed droplet flow), and grees and plunges 21 degrees and a tension axis that trends 70 the dispersed ligament / droplet regime. For pre-CHF bubbly flow degrees and punges 7 degrees. in the jet nozzle, the post-CHF flow (beginning from jet nozzle exit / heated test section inlet) consists of the rough wavy NUREG/CR-5166: ELECTROCHEMICAL EVALUATION OF regime, followed by the agitated and then the dispersed liga-SOLID STATE PH SENSORS FOR NUCLEAR WASTE CON. ment / droplet regime. In the same way, for pre-CHF slug flow in TAINMENT, HUANG,P.H.; KREIDER,K.G. National institute of the jet core, the post-CHF flow is compnsed of the agitated Standards & Technology (formerly National Bureau of Standa. regime at the nozzle exit, followed by the dispersed regime. Pre-July 1988. 29pp. 8808180204. NBSIR 88-3790. 46519:122. CHF annular jet core flow results in a small, depleted post CHF This report contains a literature review for electrochemical agitated flow regime at the nozzle exit, immediately followed by evaluation of solid state pH sensors. The requirements of pH the dispersed ligament / droplet regime, electrode for geochemical fluids in a nuclear waste repository site are rather difficult to fulfil, that ,s, i the electrode must have NUREG/CR-5178: EVALUATION OF GENERIC ISSUE 125.II.7, REEVALUATE PROVISION TO AUTOMATICALLY ISO-stability at temperatures up to 250 degrees C, low ionic and LATE FEEDWATER FROM STEAM GENERATOR DURING A redox interferences, corrosion resistance, and robustness- LINE BREAK. BRUSKE.SJ.; WELLAND,H.J.; CATHEY,N.G.; et Amo..g the potential electrode materials, the IrO(2) emerges at al. EG&G idaho, Inc. (subs. of EG&G, Inc.). July 1988. 55pp. the most promising because of its consistent near Nernstian

• 8808120105. EGG-2544. 46453:329.

sponse and its property of low impedance in thin films. Hovvv- This report presents the evaluation of the potential safety er, improved understanding of the electrochemical behavior at concerns identified in Generic issue 125.11.7, related to the auto-l the IrO(2)-solution interface is essential in order to optimize the matic auxiliary feedwater (AFW) isolation from a steam genera-performance of IrO(2) electrodes for pH sensing at high tem- tor during a main steam or feedwater line break. For this review, peratures and pressures. This report reviews theoretical models existing probabilistic nsk assessments (PRAs) were ovaluated to of the oxide-solution interfaces based on the theory of the elec- identify specific event tree sequences where the AFW system tric double layer. Electrochemistry of IrO(2) films with emphasis had failed. These sequences were utilized to calcula" De con-on the properties of anodic films is summanzed. A plan for pH tribution of the AFW isolation system to the acc%nt sequence testing of sputtered iridium oxide films TSIROF) for geochemical frequency. By utilizing this methodology, the change in risk measurements at a Tuff Repository is described. could be calculated for a plant with an automatic AFW isolation system compared to the same plant with the automatic AFW isolation system removed. The review evaluated one Westing-

Main Citations and Abstracts 53 house plant, one Combustion Engineenng plant, and two Bab- and the methods and procedures used for f pecimen removal cock and Wilcox design versions of a plant. from the generator are reported. Results from examinations of NUREG/CR-5180: CHEMICAL DECONTAMINATION AND CHEM- these specimens are presented and discussed. These examina-ICAL CLEANING OF LWR COMPONENTS AND POSSIBLE tions include visual inspection of all specim9ns and metallogra-INTERACTIONS WITH METALLURGICAL AGING EFFECTS. phic examination and burst testing of selected specimens. Sta-DIERCKS.D.R. Argonne National Laboratory. November 1988, tistical analysis of the combined metallographic and EC data to 52pp. 8812190153. ANL-88-30. 47805:104. dsterrnine the probability of detection (POD) and sizing accura-The basis for a recently initiated program on the chemical de- cy are reported along with a discassion of the factors which in-contamination of nuclear reactor components and its possible fluenced the EC results. Finally, listings of the metallographic i impact on extended. life service are desenbed. The economic in- and corresponding EC data bases are given. I centives to the utilities of extending the life of nuclear power plants beyond the present 40-year hmit are noted, and the aging NUREG/CR-5189: CLOSEOUT OF IE BULLETIN 79-26: BORON degradation processes that may be accentuated in extended-life LOSS FROM BWR CONTROL BLADES. DEAN,R.S.; service are described. The CAN-DECON(TM) and LOMi chemi- FOLEY.W.J.; HENNICK,A. Parameter, Inc. August 1988. 22Wp. caf decontamination processes for nuclear plant primary sys- 8810030117. PARAMETER IE171,46972:349. tems are summarized, and the results of prior studies on their Documentation is provided in this report for the closeout of IE corrosive effects on reactor structural alloys are reviewed in Bulletin 79-26 on the safety-related subject of boron loss from detail. Available experience with the Electric Powe tiescarch boiling water reactor (BWR) control blades. Closeout is based Institute Steam Generator Owner's Group chemical cleaning on the implementation of the required actions by licensees for process for the secondary side of pressurized water reactor all 25 General Electric (GE) facilities to which the bulletin was steam generators is also reviewed. issued for action. The bulletin was issued initially on November NUREG/CR-5182: THE SEISMIC CATEGORY t STRUCTURES 20,1979 and in slightly revised form on August 29.1980.The PROGRAM.Results For FY 1986. BENNETT,J.G.; DOVE,R.C.; NRC's concern was the safety impact of a loss of boron poison DUNWOODY,W.E.; et al. Los Alamos National Laboratory. Sep. material from control blades on shutdown capability and scram tamber 1988. 62pp. 8810050192. LA-11377-MS. 46987:160. reactivity. The failure mode had been identified and investigated The accomplishments of the Seismic Category I Structures by GE. After examination of GE's hot cell test results, calcula-Program for FY 1986 are reported. The background leading to tions, assumptions and conclusions, the NRC required all BWR the FY 1986 Program Plan is summarized and the design of a licensees to take five detailed actions, including submittal of new geometric configuration of a reinforced concrete shear wall written reports. On October 22,1985, the NRC granted GE's re-test structure is described. The report discusses static and seis- quest that " advanced longer-life control rods" (ALLCRs) be mic testings of two of these structures, a 1/4-scale,1 in.-thick exempt from the requirements of the bulletin, except for track-sheai wall model of microconcrete and a 4-in. thick shear wall ing control rod life. prototype. Results and conclusions regarding degrading stiff-ness characteristics, natural frequencies, and scalability of mi- NUREG/CR 5190: CLOSEOUT OF lE BULLETIN 80-14:DEGRA. croconcrete with actual concrete are compared with past fiscal DATION OF BWR SCRAM DISCHARGE VOLUME CAPABILITY. year results. Possible base rotation effects for the large struc- FOLEY,W.J.; DEAN,R.S.; HENNICK,A. Parameter, Inc. Novem-ture are examined analytically. Finally, tentative conclusions are ber 1908. 30pp. 8812010349. PARAMETER IE172. 47690:233. stated regarding the degrading stiffness and scaling of these Documentation is provided in this report for the closecut of IE structures and recommendations are made about future seismic Bulletin 80-14 regarding degradation of safety-related scram dis-testing of large structures. charge volume (SDV) systems in General Electric (GE) boiling NUREG/CR 5183: A USER'S MANUAL FOR THE CONTAMI- water reactors (BWRs). Closeout is based on the implementa-NANT TRANSPORT MODULE OF THE MiGRAT CODE. tion and verification of six required licensee actions. Evaluation NESTOR,CW.; PIN,F.G.; REISTER D.B.; et al. Oak Ridge Na- of utility responses and NRC/ Region inspection reports indi-tional Laboratory. October 1988. 243pp. 8812020t66. ORNL/ cates that the bulletin is closed for all of the 23 facilities to TM-10830. 47688:290. which it was issued for action. The bulletin was issued June 12, The TRUST computer code, which simulates the movement 1980 as a result of SDV events at two plants. Furthermore, of moisture in the unsaturated zone, has been modified to the studies of related anticipated transients without scram (ATWS) MiGRAT code that provides the capability of simulating the si- raised concem that the SDV function may bs degraded by the multaneous transport of up to two retarded and decaying con- undetected presence of fluid in the SDV. The bulletin was for-taminants. The numerical method implemented for contaminant warded for information only to all GE power reactor facilities transport in this code has been venfied by comparison with an with a construction permit. analytic solution for a simple transport problem. This verification demonstrates excellent agreement over the range of model NUREG/CR-519h CLOSEOUT OF IE BULLETIN 80-17: FAILURE input parameters used in the verification. OF 76 OF 185 CONTROL RODS TO FULLY INSERT DURING NUREG/CR 5185: STEAM GENERATOR GROUP PROJECT. Task A SCRAM AT A BWR. DEAN,R.S.: FOLEY,W.J.; HENNICK,A. 13 Final Report: Nondestructive Examination (NDE) Validation. Parameter, Inc. December 1988. 55pp. 8901090317. PARAME-BRADLEY,E.R.; DOCTOR,P.G.; FERRIS.R.H.; et al. Battelle Me- TER IE173. 48110.113. monal Institute, Pacific Northwest Laboratory. August 1988. Documentation is provided in this report for the closeout of IE 340pp. 8810280153. PNL-6399. 47246:320. Bulletin 8017 on the subject of safety related failures of control

      - The Steam Generator Group Project (SGGP) was a multi-task         rods to insert fully at boiling water reactors (BWRs). Closeout is effort using a retired-from-service steam generator as a test bed      based on the implementation and venfication of numerous ac-to investigate the reliability of in-service inspection. This report   tions required for compliance with the initial bulletin and five describes the results and analysis from Task 13 NDE Valida-            supplements. Evaluation of utility responses and NRC/ Region tion. The primary r)jective of Task 13 was to validate the EC          inspection reports indicates that the bulletin is closed for all of inspection results ob*ained under Tasks 7 and 9 and thereby            the 23 operating BWRs to which it was issued for action. The establish the re'%:ity of EC inspection to detect and size tube         rr min reason for incomplete insertion of control rods was found defects. Another objeClive was to measure the remaining integ.          to be retention of a significant amount of undetected water in rity of degraded specimens by burst testing More than 550               the scram discharge volume (SDV). The bulietin and its supple-specimens were removed from the generator and included in               ments were forwarded for information only to all BWRs with a the validation studies. The bases for selecting the specimens           construction parmit.

54 Main Citations and Abstracts NUREG/CR-5192: TESTING OF A NATURALLY AGED NUCLE- NUREG/CR-5198: INHALED (239)PUO(2) AND/OR TOTAL-AR POWER PLANT INVERTER AND BATTERY CHARGER. BODY CAMMA RADIATION.Early Mortality And Morbidity in GUNTHER,W.E. Brookhaven National Laboratory. September Rats And Dogs. FILIPY,R.E.; DECKER J.R.; LAl,Y.L.; et al. Bat-1988.55pp.8901090342. BNL-NUREG-52158. 48110:323. telle Memorial Institute, Pacific Northwest Laboratory. August A naturally aged inverter and battery charger were obtained 1988. 305pp. 8809060245. PNL-6586. 46689:001. from the Shippingport facility. This equipment was manufactured Rats and beagle dogs were given doses of (60)Co gamma ra-in 1974, and was installed at Shippingport in 1975 as part of a diation and/or body burdens of (239)PuO(2) within lethal ranges major plant modification. Testing was performed on this equip- in an experiment to determine and compare morbidity and mor-mrmt under th9 auspices of the NRC's Nuclear Plant Aging Re- tality responses of both species within 1 year after exposure. search (NPAR) program to evaluate the type end extent of deg- Radiationi 7 nduced morbidity was assessed by measuring changes in body weights, hematologic parameters, and pulmo-radation due to a9in9, and to determine the effectiveness of nary-function parameters. Gamma radiation caused transient condition monitoring techniques which could be used to detect morbidity, reflected by immed;ately depressed blood cell con-aging effects. Steady state testing was conducted over the centrations and by long-term loss of body weights and dimin-equipment's entire operating range. Step load changes were ished puimonary function in animals of both species that sur-also initiated in order to monitor the eiectrical response. During vived the acute gamma radiation syndrome. Inhaled plutonium this testing, component temperatures were monitored and cir- caused a loss of body weight and diminished pulmonary func-cuit waveforms analyzed. Results indicated that aging had not tion in both species, but its only effect on blood cell concentra-substantially affected equipment operation. On the other hand, tions was lymphocytopenia in dogs. Combined gamma irradia-when compared with original acceptance test data, the monitor- tson and plutonium lung burdens were synergistic, in that ani-ing techniques employed were sensitive to changes in measura- mais receiving both radiation insults had higher morbidity and ble component and equipment parameters indicating the viabili- mortality rates than would be predicted based on the effect of , ty of detecting degradation prior to catastrophic failure. either kind of radiation alone. Rats were less sensitive than  ! dogs to both kinds of radiatio whether administered alone or NUREG/CR-5194: RELAPS/ MOD 2 MODELS AND CORRELA- in combination. It was concl@ J that the beagle dog is a better TIONS. DIMENNA,R.A.; LARSON.J.R.; JOHNSON,R.W,; et al. model than the rat for predicting the effect of similar radiation EG&G Idaho, Inc. (subs. of EG&G, Inc.). August 1988. 400pp. insults in humans. 8809220096. EGG-2531. 46881:312. A review of the RELAP5/ MOD 2 computer code has been NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH performed to assess the basis for the modele and correlations AOT AND STI REQUIREMENTS AT THE ANO-1 NUCLEAR l POWER PLANT, SAMANTA.PK; WONG S.M.; CARBONARO J. comprising the code. The review has included verifbation of the onginal data base, including thermodynamic, thermal-hydraulic, Brookhaven National Laboratory. August 1988. 78pp. 8809220046. BNL-NUREG-52024. 46883:092. and geometric cunditions; simplifying assumptions in implemen- This report presents an evalation of the core melt frequency tation or application; and accuracy of implementation compared contributions associated with Allowed Outage Times (AOTs) to documented descriptions of each of the models. An effort and Surveillance Test intervals (STis) at Arkansas Nuclear One has been made to provide the reader with an understanding of - Unit 1 (ANO-1). The results show that the core-melt frequency what is in the code and why it is there and to provide enough contributions from present AOTs and STis vary by more than information that an analyst can assess the impact of the corre- four orders of magnitude (a factor of 10,000). This wide range lation or model on the ability of the code to represent the phys- of variation indicates the wide range of the risk importance of ics of a reactor transient. Where assessment of the implement- present AOTs and STis. The core-melt contributions from spe-ed versions of the models or correlations has been accom- cific AOTs and STis can be used to prioritize those components piished and published, the assessment results have been in- which should be focused on for inspection activities, personnel cluded. training, ead reliability program activities that are involved with NUREG/CR-5196: SUBMISSION FOR THE CSNI/GREST BENCHMARK EXERCISE ON CHEMICAL THERMODYNAMIC NUREG/CR-5201: EXPERIMENTAL ASSESSMENTS OF GUN- .' MODELING IN CORE CONCRETE INTERACTION RELEASES DREMMINGEN RPV ARCHIVE MATERIAL FOR FLUENCE OF RADIONUCLIDES. POWERS,D.A. Sandia National Laborato. RATE EFFECTS STUDIES. HAWTHORNE,J.R.; HISER,A.L. Ma-ries. October 1988. 53pp. 8811010250. SANDBfb1920. terials Engineering Associates, Inc. October '988. 314pp. 47275:064. 8811110092. MEA 2286. 47527:333. A submission for the CSNI/PWG-4/GREST standard problem The 250-MW boiling water Gundremmingen reactor, KRB-A, on chemical thermodynamic modeling in core-concrete interac- In the Federal Republic of Germany (FRG) has been decommis-sioned. A joint USA /FRG/UK study is underway to evaluate ma-tion releases of radionuclides is described. Part A of the exer-  ; cise is a highly defined benchmark calculation in which data tand tenal removed from the vessel for a critical assessment of - p wer reactor vs. test reactor environment effects. The vessel speciation are specified. The problem is, however, ambiguous operated at 288 degrees C; the inner wall fluence estimate at concerning the definition of an ideal solution. Consequently, two omrn sMng was aw 1 x 6 nkmm, E greaw man solutions are provided, in one solution, specified species are 1 MeV. This report describes test reactor irradiation assess-treated as molecular entitles to define the ideal solution, in the ments of a forging segment believed to be archive material from second, mixing is assumed to occur ideally on cationic and ani- the KRB-A vessel fabrication. Charpy-V (C(v)), compact tension onic lattices. Tne different results obtained in these calculations (CT) and tension test specimens were evaluated in five as-irra-illustrate the importance of condensed phase modeling in the diated and two postirradiation annealed conditions. With 288 analyses of high temperature melt interactions with concrete. deg ses C irradiations, the eMvation in 100 MPa m temperature Part B of the exercise consists of six problems in which the way 'ound to match the elevation in 41-J temperature tests temperatures, pressures and bulk compositions of the melts are within 12 degrees C. The latter elevation was predicted well by specified Data and speciation are to be supplied as parts of the Regulatory Guide 1.99. The L-C onentation data for the archive solutions to the problems. Results of calculations for these six matenal vs. the vessel trepans suggest a fuence-rate effect. problems are presented. Additional solutions are provided to il- The C-L orientation data, however, do not. A test orientation de-lustrate the effects of speciation in the condonsed oxide phase, pendence of radiation embnttlement sensitivity, described by the non-ideality in the condensed metal phase and uncertainty in trepan material but not the archive material, is responsible and the thermodynamic properties of gas phase species. is anomalous.

Main Citations and Abstracts 55 NUREG/CR-5203: DYNAMIC AMPLIFICATION OF ELECTRICAL gentsat shear strength in the containment walls. While testing CABINETS. BANDYOPADHYAY; HOFWIAYER,C.H.; programs indicate that significant shear strength is available in KASSIR.M.K.; et al. Brookhaven National Laboratory. June cracked reinforced concrete, the testing also demonstrates that 1988. 56pp. 8809060195. BNL-NUREG-52159. 46689:306. shear stiffness reduces significantly after cracking. The need to Dynamsc amplification of electrical cabinets is discussed in consider the reduced shear stiffness is discussed. Recommen-this report. The amplification factor is defined as the ratio of the dations for revised design provisions are summarized and output and the input spectral accelerations of the time histories design examples are provided. recorded at a device location and at the base of the cabinet, respectively. High level test data of nine motor control centers NUREG/CR-5210: TECHNICAL FINDINGS DOCUMENT FOR GE. and ten switchgear assemblies have been studied. The amplifi- NERIC ISSUE 51: IMPROVING THE RELIABILITY OF OPEN. cation values at various frequencies are graphically presented CYCLE SERVICE-WATER SYSTEMS. NEITZEL,0.A.; as amplification spectra. A probabilistic estimate of the amplifi- JOHNSON,K.l. Battelle Memorial Institute, Pacific Northwest cation factor for each equipment class is included. The influ- Laboratory. August 1968. 114pp. 8809220089. PNL-6623. ence of various parameters on the arnplification values is also 46880:250. discussed. This report summarizes information needed to prepare a foul-ing surveillance and control program for a nuclear power plant. NOREG/CR-5204: LOW-LEVEL RADIOACTIVE WASTE SOURCE The safety significance of bivalve and other fouling is reviewod. TERM MODEL DEVELOPMENT AND TESTING. Many safety-related systems are cooled either directly by the SULLfVAN,T.M.; KEMPF.C.R.; SUEN.C.J.; et al. Brookhaven Na- open-cycle water system or indirectly through intermediate cool-tional Laboratory. August 1988. 208pp. 8811070043. BNL- ing loops. Residual heat-removal heat exchangers, containment NUREG 52160. 47388:019. cooling units, diesel-generator coolers, fire-protection systems, The Low Level Waste Source Term Evaluation Project has and safety-related equipment coolers have been fouled by bi-the objective to develop a system model capable of predicting valves, sediment, or corrosion. The biological characteristics of radionuclides wiease rates from a shallow land burial facility. The bivalves enhance their ability to foul service-water systems. The previous topical report for this project discussed the framework design of the service- water system provides areas where sedi-and methodology for developing a system model and divided ments can accumulate and where bivalves can settle and grow. the problem into four compartments: water flow, container dog- Surveillance and control systems are available to reduce the oc radation, waste form leaching, and radionuclides transport. Each currence of bivalve sediment, and corrosion fouling. No one of these compartments is described by submodels which will be technique seems to provide the best answer, A workable sur-coupled into the system model. From February,1987 to March, veillance and control program requires using several surveil-1988, computer mode:s have been selected to predict water lance and control attematives. Utility expenence has shown that flow (FEMWATER) and radionuclides transport (FEMWASTE) continues low-level chlorination of the service-water system is and separate models have been developed to predict pitting ne of the most effective means of minimizing the safety signifi-corrosion of steel containers and leaching from porous waste cance of macrofouling. forms contained in corrodible containers. This report discusses each of the models in detail and presents results obtained from NUREG/CR-5212: EMERGENCY ENVIRONMENTAL SAMPLING applying the models to shallow land burial trenches over a AND ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES. range of expected cortlitions. STOETZEL,G.A.; LYNCH,T.P. Battelle Memorial institute, Pacific NUREG/CR-5207: FRACTURE EVALUATION OF SURFACE Northwest Laboratory. August 1988. 85pp. 8809220053. PNL-CRACKS EMBEDDED IN REACTOR VESSEL 6625.46883:170. CLADDING. Material Property Evaluations. MCCABE,0.E. Materi. This report provides information that could be used by radio-als Engineenng Associates, Inc. September 1988. 36pp. active material facilities for developing or improving environmen-8810110299. MEA-2285. 47043:001. tal sampling and analysis programs for emergency conditions. The materials that are present in the local region of the clad Areas that need to be addressed during the planning phase of layer of RPV steel were evaluated for tensile properties and such a program include 1) emergency organization, 2) sample fracture toughness before and after irradiation damage. Residu. measurement and collection locations, 3) required equipment al stresses in the clad region were determined. The information and supplies,4) cample collection procedures,3) field measure-described herein was used to understand the behavior of sur- ment methods,6) recordkeeping methods, and 7) quality assur-face cracks embedded in the clad layer in beam tests conduct. ance prograrn. Emphasis is placed on the need for these fecili-ed in another phase of this investigation. ties to coordinate monitoring activities with any supporting agen-cies, such as state and federal monitoring teams who might re-NUREG/CR-5209: DESIGN PROVISIONS FOR TANGENTIAL spond. The report also reviews the responsibilities and current SHEAR IN CONTAINMENT WALLS. OESTERLE,R.G. Construc- capabilities of radioactive material facilities, state and local gov-tion Technology Laboratories August 1988.61pp.8808230423. emment agencies, and federal agencies with regard to environ-46575:322. mental sampling and analysis in an emergency situation. The purpose of the work accomplished in preparing this report was to synthesize results of available research concem- NUREG/CR-5214: ANALYSES OF NATURAL CIRCULATION eng the capacity of cracked reinforced containment walls to DURING A SURRY STATION BLACKOUT USING SCDAP/ transfer tangential shear stresses while in a state of biaxial ten- RELAP5. BAYLESS,P.D. EG&G Idaho, Inc. (subs. of EG&G, sion from internal pressurization. A review of experimental work inc.). October 1988. 161pp. 8811010254. EGG-2547. is presented. Results of experimental work indicate that the cur- 47276:083. rent A',ME-ACI Code Provisions for Tangential Shear Stresses The effects of reactor coolant system naturai circulation on are very conservative. Recommendations for redefinition and re- the response of the Surry nuclear power plant during a station vised use of the terms Vc " concrete contribution," and Vs blackout transient were investigated. A TMLB' sequence (loss

     " steel contribution" are provided. Results of testing programs        of all ac power, immediate loss of auxiliary feedwater) was sim-are used to formulate revised design provisions for diagonal           ulated from transient initiation until after fuel rod relocation had tensile strength. Significant y higher shear stresses can be al-        begun. Integral analyses of the system thermal-hydraulics and lowed without inclined reinforcement. Also, an a%Iytical study          the core damage behavior were performed using the SCDAP/

based on recent testing programs is used to define a conserva- RELAPS computcr code and several different models of the tive maximum limit for tangential shcar stress. The maximum plant. Three scoping calculations were performed in which the limit is dependent on the relative amounts of crthogonal rein- complexity of the plant model was progressively increased to forcement and inclined reinforcement used to provide the tan- determine the overall effects of inmessel and hot leg natural cir-

56 Main Citations and Abstracts culation flows on the plant response. The natura; circulation the most damaging form of waterhammer and its diagnosis is flows extended the transient, slowing the core heatup and de- complicated by the complex nature of the underlying phenom-laying core damage by transferring energy from the core to ena. In Volume 1, the guidebook groups condensation-induced structures in the upper plenum and coolant loops, increased waterhammers into five event classes which have similar phe-temperatures in the ex-core structures indicated that they may nomena and levels of damage. Diagnostic guidelines focus on fail, however. Nine sensitivity calculations were then performed locating the event center where condensation and slug accel-to investigate the effects of modeling uncertainties on the multi- eration take pig.ce. Diagnosis is described in three stages: an dimensional natural circulation flows and the system response. initial assessment, detailed evaluation and final confirmation. Creep rupture failure of the pressurizer surge line was predicted Graphical scoping analyses are provided to evaluate whether an to occur in eight of the calculations, with the hot leg failing in event from one of the event classes could have occurred at the the ninth. The failure time was fairly insensitive to the param- event center. Examples are provided for each type of water-eters varied. The failures occurred near the time that fuel rod hammer. Special instructions are provided for walking down relocation began, well before failure of the reactor vessel would damaged piping and evaluating damage due to waterhammer, be expected. A calculation was also performed in which creep To illustrate the diagnostic methods and document past experi- i ruptuta failure of the surge line was modeled. The subsequent ence, six case studies have been compiled in Volume 2. These blowdown led to rapid accumulator injection and quenching of ) case studies, based on actual condensation-induced waterham- ] the entire core. mer events at nuclear plants, present detailed data and work ' NUREG/CR-5218: FINANCIAL QUALIFICATIONS REVIEW OF through the event diagnosis using the tools introduced in the APPLICANTS FOR NUCLEAR POWER PLANT CONSTRUC- first volume. TION PERMITS. HENDRICKSON,P.; MULLEN,M.F.; CARR.D.B. Battelle Memorial Institute, Pacific Northwest Laboratory. Sep- NUREG/CR-5220 V02: DIAGNOSIS OF CONDENSATION-IN-tember 1988. 80pp. 8810050261. PNL-6632. 46989:007. DUCED WATERHAMMER. Case Studies. IZENSON,M.G.; The NRC and its predecessor the AEC have had a regulatory ROTHE,P.H.; WALLIS G.B. Creare, Inc. October 1988.139pp. requirement since 1956 that utilities seeking a construction 8812020159. CREARE TM 1189. 47690:019. permit for a nuclear power plant be financially qualifted to con. See NUREG/CR.5220,V01 abstract. qui me s w re ade o e the years nc udi an te pt NUREG/CR-5223: SCINTILLATION FIBER DETECTOR FOR IN. 1982 to drop financial qualification review for electric utilities. VIVO ENDOSCOPIC INTERNAL DOSIMETRY, This attempt was subsequently found invalid by a federal court. PFTTlJOHN,R.R. Center for Planning & Research, Inc. October Nevertheless, financial qualification reviews consume significant 1988. 32pp. 8811110069. 47523:027. amounts of NRC staff time and time at Atomic Safety and Li. This document reports on a feasibility study to design and censing Board hearings. The analysis reported in this study was construct a scintillation fiber radiation detector for in-vivo endos-conducted to determine whether there is any empirical evidence copic internal dosimetry. The instrument design utilizes an of a relationship between a utility's financial health at the time alpha / beta particle-sensitive plastic scintillation fiber to detect of its construction permit application and the subsequent safety and measure internally deposited radionuclides that cannot be performance of the opera $ng plant. The principal financial suitably assayed using other techniques. Scintillations are opti-measures used to test for this relationship were bond rating, in. cally guided to a photomuttiplier for detection. The computerized terest coverage ratio, debt / asset ratio, debt / equity ratio, and electronics provides preamplification, amplification, analog-to-rate of return on equity. The principal safety measure was the digital conversion and storage in a multi-channel analyzer. The long-term average of the scores assigned the utility in four key rrncrocomputer provides data storage, retrieval, and graphical areas by the NRC under the Systematic Assessment of Licens- display Alpha detection and attenuation were conducted with a ee Performance program. The results of the analysis showed no calibrated P 210 source. Ten scintillation fibers of varying diam-evidence of a relationship between financial health at tne time eters (0.5 to 3 mm) degrees and properties were tested to de-of the constructiori permit and subsequent safety performance. termine their relative sensitivity for alpha particle detection as a NUREG/CR-5219: THE MIXING OF IMMISCIBLE LIOUlD function of distance of the source to the photomultiplier tube. LAYERS BY GAS BUBBLING. SUO-ANTTILA,A. Sandia Nation- The sensitivity in counts per minute was approximately propor-at Laboratories. November 1988. 55pp. 8812010353. SAND 88- tional to the diameter of the fiber for samples of the same com-7118.47690:252. position and manufacturer. A mechanistic model of the mixing of two layers of immiscible fluids by gas : mhng is presented. The model is based upon NUREG/CR-5225: AN OVERVIEW OF BWR L1 ARK-1 CONTAfN-the conceptual framework of a bundle lifting a droplet of heavy MENT VENTING RISK IMPLICATIONS. WAGNER,K.C.; fluid into the light fluid and releasing it when the bubble breaks DALLMAN,R.J.; GALYEAN,W.J.; et al. EG&G Idaho, Inc. (subs. at the upper surface. The model predicts the degree of entrain- of EG&G, Inc.). November 1988. 83pp. 8812f 90094. EGG. ment for any rate of gas flow. The rate of heat transfer between 2548.47805:236. the two layers is also predicted. Comparison of the model pre- Venting of boiling water reactors with Mark-l containments dictions and experimental data is found to be good. A compari- has been suggested as a way to prevent catastrophic failure son with expenment of an existing correlation by Calderbank for and/or mitigate the consequences resulting from a severe acci-emulsification by aercation is also presented. Applying the dent. Based on phenomenological, human factors, and nsk con-model to molten core-concrete interactions indicates that a siderations, the potential benefits and downsides of venting three-layered concept, heavy oxide / metallic / light oxide, is un- Mark-l containments were analyzed. Several generic venting j stable. A more likely situation would be a completely entrained systems and two proposed utility systems were reviewed. Based { system or a partially entrained two layer, oxide metallic, system on generic considerations, the offsite consequences during nsk 1 depending upon the gas velocity. dominant accidents were qualitatively assessed for four different vont systems. A quantitative risk study of an early venting strat-NUREG/CR-5220 V01: DIAGNOSIS OF CONDENSATION-IN- egy was periormed, based on the existing Peach Bottom hard-DUCED WATERHAMMER. Methods And Background, ] ware and the draft NUREG-1150 results for Peach Bottom Ap-tZENSON,M.G.; ROTHE,P.H.; WALLIS,G B. Creare, Inc. October pendices are also included which corttain reviews of the Pilgnm l 1988.157pp. 8812020164. CREARE TM-1189. 47688:133. and Vermont Yankee venti.1; submittals, a response to the This guidebook provides reference matenal and diagnostic seven questions from the NRC sout the Pilgrim venting strate-procedures concerning condensation-induced waterhammer in gy, and a review of the venting strategy directed by Revision 4 nuclear power plants. Condensation-induced waterhammer is of the Boiling Water Reactor Emergency Procedures Guidelines.

i Main Citations and Abstracts ti7 NUREG/CH-5227: FITNESS FOR DUTY IN THE NUCLEAR A fire modeling computer code has been developed with suffi-POWER INDUSTRY.A Review Of Technical issues. BARNES V.; cient flexibility for accurate representations of geometry, ventila-FLEMING,l.; GRANT,T.; et al. Battelle Human Affairs Research tion and other conditions as may be present in cable rocms, Centers. September 1988.174pp. 8810050265. BHARC700/ control rooms, and other enclosures in nuclear power plants. , 88/018.46987:223. The computer code is capable of three-dimensional, transient,  ! This report presents information gathered and analyzed in turbulent flow and heat transfer calculations with chemical reac-support of the United States Nuclear Regulatory Commission tion and radiation. The code has a modular structure, specifical-(NRC's) efforts to develop a rule that will ensure that workers ly designed for fire problems. The code employs the latest rele-with unescorted access to protected areas in nuclear power vant finite-volume solution techniques. The code has been ap-plants are fit for duty. The primary potential fitness-for. duty con- plied to a series of benchmark problems and a recent fire test corn addressed in the report is impairment caused by substance problem. This has confirmed the feasibility of the fire code. Con-abuse, although other sources of impairment on the job are dis- siderable further work is needed to enhance the physical cussed. The report eumines the prevalence of fitness-for-duty models (to improve the realism of predicted solutions) and to problems and discusses the use and effects of ilhcit drugs, pre- validate and document the final code. Specific recommenda. senption drugs, over the-counter preparations and alcohol. The tions are made for Phases 11 and ll1 of the Project. j ways in which fitness-for-duty concerns are being addressed in both public- and private-sector industries are reviewed and a NUREG/CR-5240: COMPARATIVE EVALUATION OF SELECTED desenption is provided of fitness-for-duty practices in six organi- CONTINUUM AND DISCRETE-FHACTURE MODELS. Emphasis zations that, like the nuclear industry,. are regulated and whose On Dispersivity Calculations For Application To Fractured Geo-operations can affect public health and safety. Methods of en- logic Media, Creston Study Area. Eastern Washington. sunng fitness for duty in the nuclear industry are examined in KUNKEL.J.R.; WAY,S.C.; MCKEE,C.R. In-Situ, Inc. November detail. The report also addressos methods of evaluating the ef- 1988. 69pp. 8812010069. 47685:082. festiveness of fitness-for-duty programs in the nuclear power in. Transverse dispersivity calculated using an ar.alytical solution dustry. to the dispersion equation is compared to transverse dispersi-e ussg a dscehadum mW W four cases W NUREG/CR-5229: ANNUAL REPORT OF THE TMI-2 EPICOR-il volv ng tw sets of parallel, planar, cor*tinuous fractures inter. RESIN / LINER INVESTIGATION. Low-Level Waste Data Base se ting at various angles. Both models preserved the functional Development Program For Fiscal Year 1988. ROGERS.R.D' relationships of transverse dispersivity when the angle of the MCCONNELL,J.W. EGAG Idaho, Inc. (subs. of EG&G, Incj' auk gradent changM M mspect to me two hactum sets, DAVIS E.C.; et at Oak Ridge National Laboratory. Uecember the spacing between fractures changed, the angle between l 1988. 47pp. 8901090263. EGG-2553. 48111:317 l fracture sets changed, or the hydraulic proporties of one of the  ! The TMI-2 EPICOR il Resin / Liner investigation: Low-Levei , sets changed. The results demonstrate that dispersivity calculat. Waste Data Base Development Program, funded by the U.S. ) ed by a discrete-fracture model agrees well with theory. Scale ) Nuclear Regulatory Commission (NRC), is studying the degrada- effects of transverse dispersivity indicated that a relatively con. tion effects in EPICOR-il organic ion exchange resins caused by stant value was reached after particles had traveled about 10 radiation; examining the adequacy of test procedures recom- times the fracture spacing of the most widely spaced fracture rnanded in the Branch Technical Position on Waste Forms to family. A field example compares dispersivities estimated from I meet the requirements of 10 CFR 61 using solidified EPICOR-li two natural-gradient tracer tests to dispersivities calculated from I resins; obtaining performance information on solidified EPICOR-rock core and other data. The example demonstrates that there il ion exchange resins in a disposal environment; and determin-

                                                                                    ' ood agreement between the theory and actual field tracer ing the condition of EPICOR-il liners. This report summarizes           tes '

accomplishments of Fiscal Year 1988. NUREG/CR-5232: UNCERTAINTIES IN MODELING AND SCAL. NUREG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PER. , ING IN THE PREDICTION OF FUEL STORED ENERGY AND FORMANCE INDICATORS. OLSON,J.; CHOCKIE,A.D.; THERMAL RESPONSE. WULFF,W. Brookhaven National Labo- GEISENDORFER,C.; et al. Battelle Human Affairs Research ratory. September 1988. 53pp. 8812010333. BNL-NUREG. Centers. October 1988. 163pp. 8811220515 PNL-668% 52164.47690:303. 47601:339. The steady state temperature distribution and the stored Th.s rtport summarizes a series of analyses of available plant energy in nuclear fuel elements is computed by analytical meth, perfonaanco data to determine if the data can be used to con-ods and used to rank, in the order of importance, the effects on struct indicators of performance for several important plant stored energy from statistical uncertainties in modeling param- functions. Data conceming the backlog of generic safety issues, eters, in boundary and in operating conditions An integral tech. operator examination scores, causes of events, and repeat nique is used to calculate the transient fuel temperature and to equipment failures are reviewed and analyzed. The anafysis in-estimate the uncertainties in predicting the fuel thermal re, dicates that generic safety issue backlog data can be used to sponse and the peak clad temperature dunng a large-break loss assess some aspects of management performance. Operator of coolant accident. The uncertainty analysis presented here is exam scores do not appear to constitute good summary meas- , an important part of evaluating the applicability, the uncertain- ures of the quality of training. Licensee Event Reports were  ! ties and the scaling capabilities of computer codes for nuclear found to be codable through the Sequence Coding and Search j reactor safety analyses. Specifically, there are 16 modeling pa. System into meaningful and useful cause codes for assessing

rameters identified to govem initially the fuel stored energy, and performance in several programmatic areas. Available data on i three parameters describing transient boundary conditions for repeat equipment failures were found to be inadequate for es-I the fuel el6 ment. It is shown that the blowdown peak clad tern- tablishing a clear measure of maintenance performance. The perature is dominated by fuel stored energy or, equivalently, by report -concludes that additional data are necessary to assure linear heating rate. Gap conductance, peaking tactors and fuel the valid and reliable assessment of programmatic performance.

thermal conductivity are the three most important fuel modeling parameters affecting peak clad temperature uncertainty. S1eam NUREG/CR-5242: A FAST BOTTOM-UP ALGORITHM FOR cooling limits the blowdown peak clad temperature. COMPUTING THE CUT SETS OF NONCOHERENT FAULT TREES. CORYNEN.G C. Lawrence Livermore National Labora-NUREG/CR-5233: A COMPUTER CODE FOR FIRE PROTEC- tory. October 1988. 64pp. 8811110022. UCRL 53828. TION AND RISK ANALYSIS OF NUCLEAR PLANTS 47520:166. SINGHAL,A.K.; HABCHl,S.D.; PRZEKWAS A.J. CFD Research An efficient procedure for finding the cut sets of large fault Corp. September 1988. 80pp. 8809220060. 46883:248. trees has been developed. Designed to address coherent or

58 Main Citations and Abstracts noncoherent systems, dependent events, shared or common- eftemately with the epicenter in order to use different weights cause events, the method - called SHORTCUT - is based on a fr.r the depth computation. Weights and partial derivatives of the fast algorithm for transforming a noncoherent tree into a quasi- travel time were based on the velocity gradient crustal model, coherent tree (COHERE), and on a new algogrithm for reducing With this technique, all events were relocated and examined for cut sets (SUBSET). To assure sufficient clanty and precision, consistent travel-time residuals. Significant residuals of about the procedure is discussed in the language of simple sets, 0.2 s were observed at stations near a sedimentary basin. Pre-which is also developed in this report. Although the new method liminary analysis of focal mechanisms suggests that earth-has not yet been fully implemented on the computer, we report quakes in the most active zone in southeaster Tennessee are theoretical worst case estimates of its computational complex- consistently sinke slip and at depths of 15 to 20 km. Earth-ity. quakes in the surrounding area are shallower and often show significant components of normal or reverse movement. This NUREG/CR-5248: PRIORITIZATION OF TIRGALE%RECOM. MENDED COMPONENTS FOR FURTHER AGING FESEARCH. suggests that the central area reacts directly to regional plate LEVY,1.S.; JARRELL.D.B. Battelle Memorial institute, Pacific stress, whereas the surrounding earthquakes occur on existing Northwest Laboratory. COLLINS,E.P.: et al. Science Applica- planes of weakness. Seismic activity near Lake Sinclair and tions international Corp. (formerly Science Applications, Inc.). Richard B. Russell Lake was also studnid A few of near Richard B. Russell Lake appear to haveinduced been by

                                                                                                                                                                              ,the events November 1988. 206pp. 8812010343. PNL-6701. 47689:173.

The " Plan for Integration of Aging and Life Extension," devel. filling of this lake, ens on R X y 9 dent f ed he s ety e a NUREG/CR-5264: GUIDE FOR LICENSING EVALUATIONS nuclear power plant structures and components (S/C) that USING CRAC2.A Computer Program For Calculating Reactor Accident Consequences. WHITE.J.E.; ROUSSIN,R.W.; should be priontried for further evaluation by the NRC's Nuclear Plant Aging Research Program (NPAR). This report documents GILPIN,H. Oak Ridge National Laboratory. December 1988. the results of an expert panel workshop established to perform 150pp.6901090271. ORNL/TDMC-3. 48111:052. the S/C priontization activity. Prioritization was primarily based A version of the CRAC2 computer code applicable for use in l upon cnteria denved from a specially-developed nsk-based analyses of consequences and risks of reactor accidents in methodology. This methodology incorporates the effect upon case work for environmental statements has been implemented plant risk of both componont aging and the effectiveness of cur. for use on the Nuclear Regulatory Commission Data General rent industry aging management practices in mitigating that MV/8000 computer systems. Input preparation is facilitated aging. An additional set of criteria used to categorize the S/C is through the use of an interactive computer program which oper-the importance of aging research on S/Cs to the resolution of stes on an IBM personal con.puter. The resulting CRAC2 input generic safety issues (GSI) and/or to identify NRC/NRR user deck is transmitted to the MV/8000 by using an error-free fue needs. The resultant S/C categorization was to provide addi. transfer mechanism. To facilitate the use of CRAC2 at NRC, rel-tional information to decision makers, but was not used to cal- evant background material on input requirements and model de-culate final S/C ranks. scriptions has been extracted from four reports " Calculations of Reactor Accident Consequences," Version 2, NUREG/CR-NUREG/CR-5255: STABLE ISOTOPES OF AUTHIGENIC MINER" 2326(SAf*0811994); "CRAC2 Model Descriptions," NUREG/ ALS IN VARIABLY-SATURATED FRACTURED TUFF. CR-2552(SAND 82-0342);"CF.AC Calculations for Accident Sec-WEBER,0.S.; EVANS,D.D. Arizona, Univ. of, Tucson, AZ. No-tions of Environmental Statements " NUREG/CR-vember 1988. 79pp. 8811220506. 47602:321. 2001(SAND 82-1693); and " Sensitivity and Uncertainty Studies identifying stable isotope variation and mineralogical changes of the CRAC2 Computer Code," NUREG/CR-4038(ORNL-in fractured rock may help establish the history of climatic and 6114). When this background information is combined with in-geomorphological processes that might affect the isolation properties of a waste repository site. This study examines the structions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific use of the stable isotope ratios of oxygen ((18)O/(16)O) and onentation toward applications on the MV/8000' carbon ((13)C/(12)C) in authigenic minerals as hydrogeochemi-cal tools tracing low-temperature rock-water interaction in vari-ably-saturated fractured tuff. Isotopic compositions of fracture- NUREG/CR-5277: THE TENSORIAL NATURE OF EFFECTIVE POROSITY AND LARGE-SCALE DISPERSION l filling and rock matrix minerals in the Apache Leap tuff, near COEFFICIENTS. Application To The Creston Study Area, Eastern Supenor, Anzona were concordant with geothermal tempera- Washington. MCKEE.C.R.; WAY,S.C. In-Situ, Inc. December tures and in equilibrium with water isotopically similar to 1988. 61pp. 8901090332,4B110:265. present-day meteone water and groundwater. Oxygen and To describe f6w in a complex system of fractures requires an carbon isotope ratios of fracture-fillings, in unsaturated fractured understanding of the effects of direction or orientation on sever-tuff, displayed an isotopic gradient believed to result from near-al hydrologic characteristics, such as hydraulic conductivity, po-surface isotopic ennchment due to evaporation rather than the rosity, and dispersion coeftscient. The theory for hydraulic con-effects of rock-water interaction. Oxygen isotope rateos of rock ductivity is well understood; this paper deals with the effects of matnx opal samples exhibited an esotopic gradient believed to fracture onentation on porosity and dispersion coefficient. The result from teaching and reprecipitation of silica at depth. Meth-tensorial nature of effective porosity was examined and was ods and results can be used to further define primary flowpaths and? movement of water in variably saturated fractured rock. found to be a second rank tensor in fractured rock units. Porosi-ty vanes at a fixed point, deper. ding on its onentation. A method NUREG/CR-5258 V01: GEORGIA / ALABAMA REGIONAL SEIS- to calculate a dispersion coefficient from field tracer tests is de-MOGRAPHIC NETWORK. Annual Report. August 1985 - June scribed. The components of the dispersion coefficient can be 1986. LONG L.T. Georgia instrtute of Technology, Atlanta, GA. calculated from the concentration profiles observed in downgra-December 1988. 40pp. 8901090250. 48111:277. dient observation wells. The method provides a procedure for Earthquakes in southeastern Tennessee were analyzed using study;ng the dispersion effect in large-scale field testing. The a velocity gradient crustal model. S-P times were used to first application of this method was successfully demonstrated in a compute ongin time independent of location. Then the epicenter tracer test performed in the research wellfield at the Creston was found using the standard methods. Depth was computed study area, Lincoln County, Washington.

Secondary Report Number index This index lists, in alphabetical order, the perforrning organization-issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross-referenced to the NUREG number for the report and to the 10-digit NRC Document Control System accession number. SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER ANL 86-52 NUREG/CR-4813 R01 EGG-2467 NUREG/CR-4740 - ANL 87 22 NUREG/CR-4960 EGG-2473 NUREG/CR-4747 V02 ANL-87-41 NUREG/CR4667 V04 EGG-2503 NUREG/CR-4898 ANL 87-51 NUREG/CR-5065 EGG 2516

• NUREG/CR-4971 ANL-87-54 NUREG/CR-5070 EGG 2521 NUREG/CR 5012 ANL-881 NUREG/CR-5082 EGG-2523 NUREG/CR 5031.

ANL-88-20 NUREG/CR 5131 EGG 2524 NUREG/CR 5043 ANL 88-21 NUREG/CR-5134 EGG 2526 NUREG/CR-5072 ' ANL-88-23 NUREG/CR-5149 EGG 2531 NUREG/CR-5194

 .ANL-88 25                 NUREG/CR-5156             EGG-2537                    NUREG/CR-5119 ANL-88-27                  NUREG/CR-5171             EGG-2540                    NUREG/CR-5137 ANL 88-30                  NUREG/CR-5180             EGG 2544                    NUREG/CR-5178 BHARC700/87/016            NUREG /CR-4991            EGG-2547                    NUREG/CR-5214 BHARC700/88/018          - NUREG/CR-5227             EGG 2548                    NUREG/CR-5225 BHARC700/08/022            NUREG/CR-5241             EGG-2553                    NUREG/CR 5229 BMI-2120                   NUREG/CR 4082 V06         EGG-REO-7827 '              NUREG/CRSOSO BMi-2146                   NUREG/CR4857              EPRI NP 5613                NUREG/CR-4780 V01 BNL-NUREG-51454            NUREG/CR 2331 V7N2-3      FEMA-REP-1                  NUREG 0654 S01 ROI BNL NUREG-51454            NUREG/CR-2331 V8N1-2      IEB-70-26                   NUREG/CR-5189 BNL-NUREG 51581            NUREG/CR-2007 V06         IEB 80-03                   NUREG/CR-4932 BNL-NUREG 51581            NUREGICR-2907 V07         IEB-8019                    NUREG/CR4933 BNL NUREG-51699            NUREG/CR-3444 VOS         lEB-83-08                   NUREG/CR 4665 BNL-NUREG-52007            NUREG/CR-4659 V02         IEB-85-002                  NUREG/CR-4935 BNL-NUREG-52008 -          NUREG/CR 4688 V02         IEB-85-003                  NUREG/CR-5140 BNL NUREG 52024            NUREG/CR 5200             KEl1559                     NUREG/CR-5159 BNL-NUREG 52029            NUREG/CR-4551 V5 DRF      LA-11013-MS                 NUREG/CR4924 BNL NUREG 52031            NUREG/CR 4939 V03         LA-11117-MS                 NUREG/CR-4908 BNL-NUREG-52031            NUREG/CR4939 V01          LA 11179-MS                 NUREG/CR-5044 BNL NUREG 52031            NUREG/CR4939 V02          LA.11214 MS                 NUREG/CR-5071 BNL-NUREG 52039            NUREG/CR-4784             LA 11236                    NUREG/CR-5090 BNL NUREG 52059            NUREG/CR-4881             LA 11325-MS                 NUREG/CR-5154 BNL-NUREG 52070            NUREG/CR-4920 V05         LA 11354-MS                 NUREG/CR-5170 BNL-NUREG-52070            NUREG/CR 4920 V04         LA 11377-MS                 NUREG/CR-5182 BNL NUREG 52070            NUREG/CR-4920 V03         LA-UR-88-2871               NUREG/CR-5135 BNL NUREG-52070            NUREG/CR-4920 V02         LBL-21967                   NUREG/CR-5133 BNL-NUREG-52070            NUREG/CR-4920 V01         LBL-22860                   NUREG/CR4864 V01 BNL-NUREG 52073            NUREG/CA-5038             LMF-119                     NUREG/CR 5067 BNL NUREG 52086            NUREG/CR-5158 V01         MEA-2198                    NUREG/CR4828 BNL-NUREG-52101            NUREG/CR-4999             MEA-2229                    NUREG/CR-5024 BNL-NUREG-52108            NUREG/CR-5051             MEA-2232                    NUREG/CR-5013 BNL-NUREG 52117            NUREG/CR-5052             MEA 2268                    NUREG/CR-5063 BNL-NUREG-52118            NUREG/CR-5053             MEA-2285                    NUREG/CR-5207 BNL NUREG-52119            NUREG/CR-5000             MEA-2286                    NUREG/CR-5201 BNL NUREG 52121            NUREG/CR-5015             MEA-2289                    NUREG/CR 5136 BNL NUREG-52137            NUREG/CR-5105             MEA-2313                    NUREG/CP-0064 BNL-NUREG 52143            NUREG/CR 5132             NBSIR 88-3790               NUREG/CR-5166 BNL-NUREG-52145            NUREG/CR-5140             ORNL 6282                   NUREG/CR-4597 V02 BNL NUREG-52146            NUREG/CR-5145             ORNL-6339                   NUREG/CR-4785 BNL-NUREG-f 2147           NUREG/CR-5146             ORNL-6377                   NUREG/CR4888 BNL-NUREG 52149            NUREG/CR 5153             ORNL 6480                   NUREG/CR-5108 BNL-NUREG 52153            NUREG/CR 5164             ORNL 6484                   NUREG/CRABB0 V01 BNL NUREG-52158            NUREG/CR-5192             ORNL-6484                   NUREG/CR4880 V02 BNL-NUREG-52159            NUREG/CR-5203             ORNL/CSD/TM 252             NUREG/CR-5033 BNL NUREG 52160             NUREG/CR-5204             ORNL/NOAC-232               NUREO/CR4674 VOS BNL NUREG-52164            NUREG/CR-5232             OANL/NOAC-232               NUREG/CR4674 V06 CREARE TM-1189              NUREG/CR-5220 V02         ORNL/NSIC-200               NUREG/CR-2000 V06N12 CREARE TM-1189              NUREG/CR 5220 V01         ORNL/NSIC-200               NUREG/CR 2000 V07 N1 CSNI105                     NUREG/CP-0064             ORNL/NSIC-200               NUREG/CR-2000 V07 N2 CSNI97                      NUREG/CP 0075             ORNL/NSIC-200               NUREG/CR 2000 V07 N3 EGG 2396                    NUREG/CR 4312 VD1         ORNL/NSIC 200             ' NUREG/CR 2000 V07 N4 EGG 2396                    NUREG/CR 4312 V02 R1      ORNL/NSIC 200               NUREG/CR-2000 V07 N5 EGG 2458                    NUREG/CR-4639 V01         ORNL/NSIC-200               NUREG/CR-2000 V07 N6 EGG-2458                    NUREG/CR4639 VO4 P3       ORNL/NSIC-200               NUREG/CR 2000 V07 N7 EGG-2458                    NUREG/CR 4639 VOS P4      ORNL/NSIC 200               NUREG/CR 2000 V07 NB EGG-2458                    NUREG/CR-4639 V04 P1      ORNL/NSIC-200               NUREG/CR-2000 V07 N9 EGG-2458                  . NUREG/CR4639 V04 P2       ORNL/NSIC-200               NUREG/CR 2000 V07N10 EGG-2458                    NUREGICR-4639 /05 P3      ORNL/NSIC 200               NUREG/CR 2000 V07N11 EGG-2458                    NUREG/CR4639 VOS P1       ORNL/TDMC-3                 NUREG/CR-5264 EGG 2458                    NUNEG/CR-46?> VOS P2      ORNL/TM-10100               NUREG/CR4651 V02 EGG-2458                    NUREG/CR-4fJ9 V02         ORNL/TM-10147               NUREG/CR-4708 V02 EGG 2458                    NUREG/CR 4639 V03 P1      ORNL/TM 10272               NUREG/CR4777 EGG-2458                    NUREG/CR-4639 V03 P2      ORNL/TM-10273               NUREG/CR 4778 EGG 2458                    NUREG/CR-4639 V03 P3      ORNL/TM 10364               NUREG/CR4860 R01
                                                                                                         )

59 i

60 Secondary Report Number index SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER ORNL/TM-10459 NUREG/CR4947 PNL-6632 NUREG/CR-521B ORNL/TM-10503 NUREG/CR-4984 PNL-6652 NUREG/CR-5227 ORNL/TM 10582 NUREG/CR 5019 PNL-6680 NUREG/CR 5241 ORNL/TM-10588 NUREG/CR-5018 PNL 6701 NUREG/CR-5248 ORNL/TM-10604 NUREG/CR-5021 V01 PNL SA-16308 NLIREG/CP-0099 ORNL/TM 10604 NUREG/CR-5021 V02 RDATR127301-002 NUREG/CR4242 ORNL/TM-10651 NUREG/CR-5049 EAIC-87/3014 NUREG/CR-5150 ORNL/TM-10663 NUREG/CR-5061 SAIC 88/3014 NUREG/CR-5151 ORNL/TM-10830 NUREG/CR-5183 SAIC-88/3023 NUREG/CR-5106 ORNL /TM-9593 NUREG/CR-4219 V04 N2 SAND 85-1726 NUREG/CR 4508 ORNL/TM-9593 NUREG/CR4219 VOS N1 SAND 86-0336 NUREG/CR-4527 V02 ORNLSUB797778/4 NURE G/CR-4996 SAND 861102 NUREG/CR 4625 ORNLSUB819089/3 NUREG/CR-5157 SAND 86-1938 NUREG/CR-4728 ORNLSUB82215981 NUr4EG/CR5020 ORNLSUB83289155 NUREG/CR-4992 V01 U V02 SANDB6 7175 NUREG/CR-4807 PARAMETER IE153 NUREG/CR4523 SAND 87-0179 NUREG/CR 4834 V02 PAR AMETER lE159 NUREG/CR4662 SAND 87-0323 NUREG/CR-4864 V01 PARAMETER IE162 NUREG/CR-4665 SAND 87 0871 NUREG/CR 4836 PARAMETER lE167 NUREG/CR-4932 SAND 87-0940 NUREG/CR-4914 PARAMETER lE168 NUREG/CR-4933 SAND 87-0956 NUREG/CR-4916 PARAMETER IE 170 NUREG/CR4935 SAND 87 0976 NUREG/CR 4917 PARAMETER lE171 NUREG/CR-5189 SAND 87-1048 NUREG/CR-5084 PARAMETER IE172 NUREG/CR-5190 SAND 871858 NUREG/CR-4993 PARAMETER IE173 NUREG/CR-5191 SAND 87 2384 NUREG/CR-5029 PNL4008 NUREG/CR-2336 SAND 87 2411 NUREG/CR-5039 V01 PNL 4221 NUREG/CR-2850 V06 SAND 87-2411 NUREG/CR-5039 V02 PNL-4221 NUREG/CR 2850 V07 SAND 87-2428 NUREG/CR-5032 PNL 5210 NUREG/CR-3950 V04 SAND 87 7176 NUREG/CR 5078 V01 PNL 5219 NUREG/CR 4000 V02 SAND 87-7176 NUREG/CR-5078 V02 i PNL 5500 NUREG/CR-5047 SAND 88-0030 NUREG/CR-5083 I PNL-5855 NUREG/CR 4605 SAND 88-0052 NUREG/CR-5099 PNL-6097 NUREG/CR-4811 SAND 88-0535 NUREG/CR-5109 PNL-6171 NUREG/CR4873 SAND 88 0745 NUREG/CR-5162 PNL-6172 NUREG/CR-4874 SAND 88-1407 NUREG/CR-5126 PNL-6175 NUREG/CR-4879 V01 SAND 881836 NUREG/CP-0095 PNL 6258 NUREG/CR-5009 SAND 88-1920 NUREG/CR-5196 PNL 6289 NUREG/CR-4991 SAND 88-7016 NUREG/CR 5096 PNL-6300 NUREG/CR 4997 SAND 88-7118 NUREG/CR-5219 PNL-6341 NUREG/CR-5016 SAND 88 8201 NUREG/CR-5086 PNL-6388 NUREG/CR-5048 SANDB8-8213 NUREG/CR-5120 PNL 6391 NURE G/CR-5055 SEA 87 253-04A1 NUREG/CR-5138 PNL-6394 NUREG/CR 5058 SEA 87-253-06A1 NUREG/CR-5160 PNL-6399 NUREG/CR-5185 SEA 87 288 04A1 NUREG/CR-d555 RO1 PNL f 413 NUREG/CR-5075 UCID-20820 NUREG/CR 4775 PNL 6511 NUREG/CP-0093 UCID-21223 NUREG/CR-5042 Sol PNL-6549 NUREG/CR 5144 UCID-21346 NUREG/CR-5113 PNL-6553 NUREG/CR-5054 UCRL 15985 NUREG/CR-5076 PNL-6586 NUREG/CR-5190 UCRL 21000 NUREG/CR-5073 PNL-6623 NUREG/CR-5210 UCRL-53828 NUREG/CR-5242 PNL-6625 NUREG/CR-5212 WHC-EP 0081 NUREG/CR-5023

i Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of -{i the report (s) prepared by the author. If further information is needed, refer to the main cita- l tion by the NUREG number. ABEL.K.H. ARNOLD,W.D. NUREG/CR-4879 V01: DEMONSTRATION OF PERFORMANCE MOD- NUREG/CR-4708 V02: PROGRESS IN EVALUATION OF RADIONU- 1 ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL SITE.A CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- j

        . Compenson Of Predictive Radionucide Transport Modeling Versus                   LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For Frid Observations At The Nitrate Disposal Pit Site. Chalk River Nuclear         Apnf 1986 - September 1987.

Labs. NUREG/CR-5108: THERMODYNAMIC PROPERTIES OF TC(IV) ABT.S.R. OXIDES. Solubilities And The Electrode Potential Of The Tc(Vil)/Tc(IV)- Oxide Couple. NUREG/CR 4651 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY RIPRAP TESTING IN FLUMES. Phase ll. Followup investigations. ATHEY,G.F.

                 ~                                                                     NUREG/CR-4000 V02: THE MESORAD DOSE ASSESSMENT NUR /CR 3444 V05: THE IMPACT OF LWR DECONTAMINATION                                 MODEL. Computer Code.

ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU- AUSTIN J.H. PATIONAL EXPOSURE. Annual Report, FY 1987. NUREG 1310: NATURALLY OCCURRING AND ACCELERATOR-PRC.s-ADAMS,R.E. DUCED RADIOACTIVE MATERIALS.1987 Review. NUREG/CR-5018: URANIUM OXIDE-IRON OXIDE MIXED AEROSOL 1 3 And 631. Data Re Repo NU E'G/CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1986.A STATUS REPORT. ADISOMA.G- NUREG/CR 4674 V06: F 4ECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-5129: EXPERdNTAL ASSESSMENT OF THE INFLUENCE DAMAGE ACCIDENTS:.286.A STATUS REPORT. OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF DRIED CEMENT BOREHOLE SEALS. AYER.J.E. NUREG 1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS AHMAD.J. HANDBOOK. I NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM - PHASE ll. Sixth i Program Report. October 1986 - September 1987. BACANSKAS,V.P. AINSWORTH.D.L. NUREG/CR 4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE NUREG/CR 5041 V02: RECOMMENDATIONS TO THE NRC FOR SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER Pt ANTS Operating Experience And Failure identification. REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOACTIVE WASTE DISPOSAL. Task 2b. Earth-Mounded Concrete NUREG/CR-5141: AGING AND OUALfFICATION RESEARCH ON SOLE-Bunkers. NOID OPERATED VALVES. ALLENSPACH,F. BADR.O. NUREG-12 84 R04: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT. NUREG/CR 5095 Vot: THERMODYNAMIC NONEOUILlBRfUM IN 'OST-IC ASSESSMENT OF LICENSEE PERFORMANCE. CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE.Desenpdon Of i Expenments And Sample Results.  ! ALTMAN W.D. NUREG/CR 5095 V02: THERMODYNAMIC NONEQUILIBRIUM IN POST. NUREG.1297: PEER REVIEW FOR HIGH LEVEL NUCLEAR WASTE RE- CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Stabi-POSITORIES. Genenc Technical Posmon. hred Quench Front Tests. NUREG 1298. QUALIFICATION OF EXISTING DATA FOR HIGH LEVEL NUREG/CR 5095 V03: THERMODYNAMIC NONEOUILIBRIUM IN POST. NUCLEAR WASTE REPOSITORIES.Genene Technical Position. CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Advanc-ing Quench Front Tests. AMICO,P.J. NUREG/CR-5095 V04: THERMODYNAMIC NONEQUILfBRIUM IN POST-NUREG/CR-5076.- AN APPROACH TO THE OUANTIFICATION OF SEIS- CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Retreat-MIC MARGINS IN NUCLEAR POWER PLANTS The Importance Of Ing Quench Front Tests. BWR Plant Systems And Functions To Seesmic Margins. BAILEY,W.J. ANDERSON.C.A. NUREG/CR.4924 SEISMIC CATEGORY l STRUCTURES NUREG/CR-3950 V04: FUEL PERFORMANCE ANNUAL REPORT FOR 1986.

     ' PROGRAM. Final Report, Fiscal Year 1983 1984.

NUREG/CR 5009: ASCESSMENT OF THE USE OF EXTENDED - l ANDERSON.M.J. " ^ " ^ ' NUREG/CR-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE BAKER,D.A. PREDICTION METHODS FOR PIP!NG. NUREG/CR-2850 V06: POPULATION DOSE COMMITMENTS DUE TO S NDERSON.N.R. RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES

                                                                           ^ *'        N
        'S S       DES N CRITERI . Ora Re                     m n                           E /C'R-2850 V07: POPULATIOli DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES APOSTOLAKIS.G.                                                                          IN 1985.

NUREG/CR-5113. FINDINGS OF THE PEER REVIEW PANEL ON THE NUREGICR-5009: ASSESSMENT OF THE USE OF EXTENDED DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG 1150. BURNUP FUEL IN LIGHT WATER POWER REACTORS. APPlGNANI.P.L. BALDWIN.C.A. NUREG/CR-5078 VOI: A RELIABILITY PROGRAM FOR EMERGENCY NUREG/CR-5019: NEUTRON EXPOSURE PARAMETERS FOR THE DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Program METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SEC-Structure TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES. i 61 1

62 Personal Author Index BALL,N.B. NUREG/CR-4639 V04 P3: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR 4807: SURFACE. COMPLEXATION MODELING OF RADIO- ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide.Part 3: NUCLIDE ADSORPTION IN SUBSURFACE ENVIR1NMENTS. NUCLARR System Desenption. BALLINGER.M.Y. BENDA,B.J. NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS NUREG/CR.5073: OUANTIFICATION OF MARGINS IN PIPING SYSTEM SEISMIC RESPONSE. Methodology?s And Damping. NUREG/C 4 97: METHODS FOR DESCRIBING AIRBORNE FRAC-TlONS OF FREE FALL SPILLS OF POWDERS AND LIQUIDS. BENNETT B.C. BAMFORD,W.H. NUREG/CR-5045: KANSAS-NEBRASKA SEISMICITY STUDIES USING NUREG/CR-5020: A

SUMMARY

OF ENVIRONMENTALLY ASSISTED THE KANSAS. NEBRASKA MICROEARTHOUAKE NETWORK. Final CRACK-GROWTH STUDIES PERFORMED AT WESTINGHOUSE Report. ELECTRIC CORPORATION.Under Funding From The Heavy Section Steel Technology Program. BENNETT.J.G. NUREG/CR 4924: SEISMIC CATEGORY t STRUCTURES BANDER,T.J. PROGR AM. Final Report, Fiscal Year 1983 - 1984. NUREG/CR 4000 V02: THE MESORAD DOSE ASSESSMENT NUREG/CR 4908: THE SEISMIC CATEGORY I STRUCTURES MODELComputer Code. PROGRAM.Results For Fiscal Year 1985. NUREG/CR-5154: EXPERIMENTAL ASSESSMENT OF DAMPtNG IN LOW ASPECT RATIO, REINFORCED CONCRETE SHEAh WALL NUREG/ 46 V02: SEISMIC FRAGILITY OF NUCLEAR POWER PLANT COMPONENTS (PHASE II) Motor Control Center. Switchboard.Panelboard And Power Sappiy. NURE / R 85 THE SEISMIC CATEGORY I STRUCTURES NUREG/CR-5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABl. PROGRAM.Results For FY 1986. BENNETT,R.D. BARANOWSKY,P.W. NUREG/CP-5041 V02: RECOMMENDATIONS TO THE NRC FOR NUREG 1032: EVALUATION OF STATION BLACKOIJT ACCIDENTS AT REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL NUCLEAR POWER PLANTS. Technical Findings Related To Unre- RADIOACTIVE WASTE DISPOSAL. Task 2b Earth-Mounded Concrete

   /    solved Safety lasue A-44. Final Report.                                                                                                                           Bunkers.

BARKER,0.B. BERGERON,K.D* NUREG/CR 4996: A REPORT ON THE ROUND ROBIN PROGRAM CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD NUREG/CR-5126: TAC 2D STUDIES OF MARK I CONTAINMENT DRYWELL SHELL MELT-THROUGH.  ; TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK ARREST FRACTURE TOUGHNESS,K(IALOF FERRITIC MATERIALS. BERGERON,M.P. BARNES.C.R. NUREG/CR-4879 V01: DEMONSTRATION OF PERFORMANCE MOD-NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM - PHASE II. Sixth ELiNG OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A Program Report. October 1986 September 1987. Comparison Of Predictwe Radionuclides Transport Modeling Versus Field Observations At The Narate Disposal Pit Site Chalk Rwer Nuclear BARNES,K. Labs. NUREG/CR 4873: BENCHMARK STUDY OF THE l-DYNEV EVACU-ATION TIME ESTIMATE COMPUTER CODE. BEYER C.E. NUREG/CR-4874: THE SENSITIVITY OF EVACUATION TIME ESTI- NURt:G/CR-5009: ASSESSMENT OF THE USE OF EXTENDED MATE 3 TO CHANGES IN INPUT PARAMETERS FOR THE l-DYNEV BURNUP FP . UGHT WATER POWER REACTORS. COMPUTER CODE. O " BARNES V NUREG/CR-5227; FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG' /CR-5105: RESPONSE MARGINS INVESTIGATION OF PIPING INDUSTRY.A Review Of Technical Issues. DYNAMIC ANALYSES USING THE INDEPENDENT SUPPORT MOTION METHOD AND PVRC DAMPING. BASDEKAS,0.L NUREG-1332: REGULATORY ANALYSIS FOR THE RESOLUTION OF BICKEL,J.H. GENERIC ISSUE 125.11.7. " REEVALUATE PROVISION TO AUTO- NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE MATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG-1150. DURING A LINE BREAK." BICKFORD,R.L BASS,B.R. NUREG/CR-2336: STEAM GENERATOR TUBE INTEGRITY NUREG/CR-4888: PRESSURIZED THERMAL-SHOCK TEST OF 6 INCH PROGRAM. Phase ll F!nal Report. . THICK PRESSURE VESSELS.PTSE-2: Investigation Of Low Teanng Re-J sistance And Warm Prestressing- BILHORN.S.G. NUREG 13's: TECHNICAL POSITION ON ITEMS AND ACTIVITIES IN BAUM,J.W[CR.5038. NUREG OPTIMIZATION OF THE CONTROL OF CONTAMINA. THE HIGH-LEVEL WASTE GEOLOGIC REPOSITORY PROGRAM g TION AT NUCLEAR POWER PLANTS. SUBJECT TO OUALITY ASSURANCE REQUIREMENTS. NUREG/CR-5158 V01: WORLDWIDE ACTIVITIES ON THE REDUCTION OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS. BLENCOE,J.G. NUREG/CR-4708 V02: PROGRESS IN EVALUATION OF RADIONU-BAYLESS,P.O. CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- i NUREG/CR 5214 ANALYSES OF NATURAL CIRCULATION DURING A LEVEL NUCLEAR WASTE REDOSITORY SITE PROJECTS. Report For j SURRY ST ATION BLACKOUT USING SCDAP/RELAPS. Apnl 1986 - September 1987. BEAN,0.L BLOND,R M NUREG/CR-5041 V02- RECOMMENDATIONS TO THE NRC FOR REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL NUREG/CR 5113: FINDINGS OF THE PEER REVIEW PANEL ON THE RADIOACTIVE WASTE DISPOSAL Task 2b Earth-Moundej Concrete DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG 1tSO-Bde* BOARD.S.J. BEERS,G.H. NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE NUREG/CR-4630 V04 P1: NUCLEAR COMPUTERIZED LIBRARY FOR DRAFT REACTOR RISK REFERENCE DOCUMENT.NUREG-1150. ASSESSING REACTOR RELIABluTY (NUCLARR). user's Guide.Part 1: Overview Of NUCLARR Data Retneval BOCCIO J. NUREG/CR 4639 V04 P2. NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5078 V01: A RELIABILITY PROGRAM FOR EMERGENCY ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide,Part 2: DIESEL GENERATORS AT NUCLEAR POWER PLANTS Program Guide To Operations Structure. I l

                                                                                                                                                     )

Personal Author Index 63 90ECKER,B.B. BURGHARDT,D. NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED NUREG/CR-5078 V02. A RELIABILITY PROGRAM FOR EMERGENCY ALPHA AND BETA IRRADIATION OF THE LUNG. Phase 11 Report. DIESEL GENERATORS AT NUCLEAR POWER COHN.M.P. aintenance,SMance W Mion Wndonng. NUREG/CR-4836: APPROACHES TO UNCERTAINTY ANALYSIS IN BUSCH80M R.L PROBABILISTIC RISK ASSESSMENT. NUREG/Chl-5198: INHALED (239)PUO(2) AND/OR ' TOTAL BODY SOLD.F.C. GAMMA RADIATION.Early Mortality And Morbidity in Rats And Dogs. NUREG/CR 5009: ASSESSMENT OF THE USE OF EXTENDED BURNUP FUEL IN LIGHT WATER POWER REACTORS. CALL,0.J. NUREG/CR-4639 V02: NUCLEAR COMPUTERIZED LIBRARY FOR AS . NURE'G/CR4888: PRESSURIZED THERMAL-SHOCK TEST OF 6 lNCH THICK PRESSURE VESSELS.PTSE 2. Investigation Of Low Teanng Re- CAMPBELL.D.J. sistance And Warm Prestressing. NUREG/CR-5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU-CONNEY,R.F CLEAR ONE - UNIT 1. Volume 1. Program For inspectors. NUREG/CR-5178: EVALUATION OF GENERIC ISSUE NUREG/CR-5021 V02: USER'S GUIDE FOR PRISIM ARKANSAS NU-12517, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE CLEAR ONE - UNIT 1. Volume 2. Program For Regulators. FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. BOUCHER,T.J. . . NUREG/CR-5190: INHALED '(239)PUO(2) ' AND/OR TOTAL-BODY NUREG/CR-4898: RESULTS OF SEMISCALE MOD-2C FEEDWATER GAMMA RADIATION.Early Mortality And Morbedity in Rats And Dogs. AND STEAM LINE BREAK (S-FS) EXPERIMEf4T SERIES. Bottom Main Feedwater une Break Accident Expenments. CARBONARO,J. NUREG/CR 4971: RESULTS OF SEMISCALE MOD-2C FEEDWATER NUREG/CR 5200: EVALUATION OF RISKS ASSOCIATED WrTH AOT. AND STEAM LINE BREAK (S-FS) EXPERIMENT SERIES Main Steam Line Break Accident Expenments. AND STI REQUIREMENTS - AT THE ANO 1 NUCLEAR POWER PLANT. ERADLEY,E.R. NUREG/CR-5185: STEAM GENERATOR GROUP PROJECT. Task 13 CWAN,SP, Final Report Nondestructive Examination (NDE) Validation. NUREG/CR-5141: AGING AND QUALIFICATION RESEARCH ON SOLE-NOID OPERATED VALVES ERINSON.D.A. NUREG/CR-5096: EVALUATION OF SEALS FOR MECHANICAL PENE- CARR,D.B. TRATIONS OF CONTAINMENT BUILDINGS. NUREG/CR-5218: FINANCIAL QUALIFICATIONS REVIEW OF APPLl- J EROCKMANN,J.E. NUREG/CR-4914: THE INFLUENCE OF SELECTED CONTAINMENT CARTER,K.D STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW-ING HIGH PRESSURE MELT EJECTION FROM THE REACTOR NUREG/CR 5090: EFFECTS OF TEMPERATURE AND HUMIDITY ON j VESSEL. RESPIRATOR FIT UNDER SIMULATED WORK CONDITIONS, NUREG/CR4917: DCH 2:RESULTS FROM THE SECOND EXPERIMENT PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. CASE,F.L EROOKS,B. NUREG/CR-5108: THERMODYNAMIC PROPERTIES OF TC(IV) . OXIDES.Solubihties And The Electrode Potential Of The Tc(Pil)/Te(IV)- NUREG 0713 V07: OCCUPATIONAL RADIATION EXPOSURE AT COM- Oxide Couple. MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES 1985. Eighteenth Annual Report. CASTLEJ.N. i NUREG/CR 3908: SURVEY OF THE STATE OF THE ART IN MITIGA. NU 0797 S20: SAFETY EVALUATION REPORT RELATED TO THE TION SYSTEMS. OPERATION OF COMANCHE PEAK STEAM ELECTRIC NUREG/CR-4242: SURVEY OF LIGHT WATER REACTOR CONTAIN-4 STATION. UNITS 1 AND 2. Docket Nos. 50 445 And 50446.(Texas Utili- MENT SYSTEMS. DOMINANT FAILURE MODES AND MITIGATION ties Generating Company) OPPORTUNITIES.Finat Report. NUREG/CR4243: VALUE/ IMPACT ANALYSIS FOR EVALUATING AL-EROWN,S.R. TERNATIVE MITIGATION SYSTEMS. NUREG/CR-5050: ANNOTATED BIBLIOGRAPHY OF REUABILITY AND NUREG/CR4244: STRATEGIES FOR IMPLEMENTING A MITIGATION RISK DATA SOURCES. POLICY FOR LIGHT WATER REACTORS. ERUSKE.S.J. CATHEY,N.G. NUREG/CR-5178: EVALUATION OF GENERIC ISSUE NUREG/CR-5178: EVALUATION OF . GENERIC ISSUE 12517, REEVALUATE PROVtSION TO AUTOMATICALLY ISOLATE 12517, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK-GRUST,F. CATTON,L Pr a e October 19 Se 987 NUREG/CR 3908: SURVEY OF. THE STATE OF THE ART IN MITIGA- l TION SYSTEMS. 4 ERYAN R.H. NUREG/CR-4242: SURVEY OF LIGHT WATER FIEACTOR CONTAIN-l NUREG/CR 4888: PRESSURIZED-THERMAL-SHOCK TEST OF 6 INCH MENT SYSTEMS, DOMINANT FAILURE MODES AND MITIGATION ' . THICK PRESSURE VESSELS.PTSE 2: Investigation Of Low Teanng Re- OPPORTUNITIES. Final Report. sistance And Warm Prestressing. l NUREG/CR-4243: VALUE/ IMPACT ANALYSIS FOR EVALUATING AL-TERNATIVE MITIGATION SYSTEMS. .! E3YSONJ.W. NUREG/CR4BB8: PRESSURIZED THERMAL SHOCK TEST OF 6-INCH NUREG/CR4244: STRATEGIES FOR IMPLEMENTING A MITIGATION l POLICY FOR LIGHT WATER REACTORS. ', THICK PRESSURE VESSELS.PTSE.2: investigation Of Low Teanng Re-sistance And Warm Prestressing. CAZZOLI,E.  ! EUCHANANJ.A. NUREG/CR-4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR 5185: STEAM GENERATOR GROUP PROJECT. Task 13 RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER . Final Report: Nondestructive Examination (NDE) Vahdation_ P " * "F R MB E ATION AND UNCERTAINTY ANALY. EUCKJ.W. SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT I NUREG/CR-4997: METHODS FOR DESCRIBING AIRBORNE FRAC. WATER REACTORS (OUASAR).Part II: Sensitivity Analysis Techw { TIONS OF FREE F ALL SPILLS OF POWDERS AND LIQUIDS. niques. J I

                                                                                                                                                 .q.

j i 64 . Personal Author index l CAZZOLI,E.G. CHUNG,Y. NUREG/C45164: A SIMPUFIED MODEL FOR CALCULATING EARLY NUREG/C45045: KANSAS-NEBRASKA SEISMICITY STUDIES USING ' OFFSITE CONSEQUENCES FROM NUCLEAR REACTOR ACCl- THE KANSAS NEBRASKA MICROEARTHOUAKE NETWORK. Final DENTS. Report. CER&RSKI,W.V. CLAISORNE,E- . . NUREG/C44597 V02: AGING AND SERVICE WEAR OF AUXILIARY NUREG/CR-5138: VALIDATION OF GENERIC COST ESTIMATES FOR FEEDWATER PUMPS FOR PWR NUCLEAR PLANTS. Volume 2.Agin9 CONSTRUCTION RELATED ACTIVITIES AT NUCLEAR POWER Assessments And Monitonng Method Evaluations. PLANTS. Final Report. NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDD AT THE SUB COMPONENT AND SUBSYSTEM LEVELFinal Report. NU E dR 4879 V01: DEMONSTRATION OF PERFORMANCE MOD-ELING OF A LOW-LEVEL WASTE SHALLOW LAND BURIAL SITE.A CLARK.A.T. Companson Of Predictive Radionuchde Transport Modeling Versus NUREG-1320: NUCLEAR FUEL CYCLE FACluTY ACCIDENT ANALYSIS Field Observations At The Nitrate Disposal Pit Site, Chalk River Nuclear HANDBOOK.' Labs. CLARK,R. NUR CR-5065: TIME. AND VOLUME AVERAGED CONSERVATION NUREG/CR-4555 RO1: GENERIC COST ESTlMATES FOR THE DIS-' POSAL OF RADIOACTIVE WASTES. EQUATIONS FOR MULTIPHASE FLOW USING MASS-WElGHTED VE-LOCITY AND INTERNAL ENERGY. CLARK,R.A. CHAVEZ,J.M. NUREG/CR2336: STEAM GENERATOR TUBE INTEGRfTY NUREG/CR-4527 V02: AN EXPERIMENTAL INVESTIGATION OF IN- PROGRAM. Phase il Final Report. TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CONTROL NUREG/CR-5016: COMPENDfUM AND COMPARISON OF INTERNA-CABINETS.Part II. Room Effects Tests. TIONAL PRACTICE FOR PLUGGING. REPAIR, AND INSPECTION OF STEAM GENERATOR TUBING. CHEN.J.C. NUREG/CR-5095 V01: THERMODYNAMIC NONEOUILIBRIUM IN POST. CLETCHER,J.W. - CRITICAL.HEATTLUX BOILING IN A ROD BUNDLE.Desenption Of . NUREG/CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE Expenments And Sample Results. DAMAGE ACCIDENTS:1986,A STATUS REPORT. NUREG/CR-5095 V02: THERMODYNAMIC NONEQUILIBRIUM IN POST. NUREG/CR-4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE. Data For Stabi- DAMAGE ACCIDENTS:1986,A STATUS REPORT. lized Quench Front Tests. NUREG/CR5095 V03: THERMODYNAMIC NONEOUILIBRIUM IN POST- COCHRELL,R. CRITICAL. HEAT FLUX BOILING IN A ROD BUNDLE. Data For Advanc. NUREG/CP-0095: PROCEEDINGS OF THE FOURTH WORKSHOP ON . inDOvench Front Tests. CONTAINMENT INTEGRfTY. NUMEG/CR 5095 V04. THERMODYNAMIC NONEOUILIBRIUM IN POST-CRITICAL-HEATfLUX BOLLING IN A ROD BUNDLE. Data For Retreat- CODELL,R.B. InD Quench Front Tests. NUREG-1263; HYDROLOGIC DESIGN FOR RIPRAP ON EM:.lANKMENT SLOPES. CHITWOOD.LD. NUREG/CR-5061; THREE-FREQUENCY EDDY CURRENT INSTRU- COHEE,B.P. MENT. NUREG/CR-3145 V06: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1986 - N EG/CR-4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES;PWR.LARGE DRY CON- COHEN,L ' TAINMENT DESIGN. NUREG-0837 V07 N04: NRC TLD DIRECT RADIATION MONITORING NUREC/C44920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE- NETWORK. Progress Report. October-December 1987. VENTION AND MITIGATION FEATURES.PWR lCE CONDENSER CONTAINMENT DESIGN. COHEN,S. NUREG/CR-4555 R01: GENERIC COST ESTIMATES FOR THE DIS-N RE d 5241: DEVELOPMENT OF PROGRAMMATIC PERFORM-

                                                                                                             ^

ANCE INDICATORS. COLLINS,E.P. NUREG/CR-5248: PRIORITIZATION OF TIRGALEX-RECOMMENDED CHOKSHIN.C'3 NUREGI 123 DRFT FC: REGULATORY ANALYSIS FOR USl A-40, COMDONENTS FOR FURTHER AGING RESEARCH.

      " SEISMIC DESIGN CRITERIA." Draft Report For Comment-CONGEMI,J.

CHONA R. NUREG/CR-2907 V06: RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR.4996: A REPORT ON THE ROUND ROBIN PROGRAM NUCLEAR POWER PLANTS. Annual Report For 1985. CONDUCTED TO EVALUATE THE PROPOSED ASTM bTANDARD NUREG/CR 2007 V07: RADIOACTIVE MATERIALS RELEASED FROM TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK NUCLEAR POWER PLANTS. Annual Report For 1986. ARREST FRACTURE TOUGHNESS.K(IA).0F FERRITIC MATERIALS. COOK,K.V. CHOWDHURY,P. NUREG/CR 4860 RO1: FLAW DENSITY EXAMINATIONS OF A CLAD NUREG/CR 4984: DEVELOPMENT OF A THREE-DIMENSIONAL FLUX BOILING WATER REACTOR PRESSURE VESSEL SEGMENT, SYNTHESIS PROGRAM AND COMPARISON WITH 3 D TRANSPORT THEORY RESULTS. CORLEY,J.H. NUREG-1311: FUNDING THE NRC TRAINING PROGRAM FOR CHRISTENSEN.D. STATES. NUREG/CR3145 V06. GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION Annual ReportOctober 1986 CORNELL,A. September 1967. NUREG/CR-5000: METHODOLOGY FOR UNCERTAINTY ESTIMATION IN NUREG-1150 (DRAFT). Conclusions Of A Review Panel CHU,T.L NUREG/CR 5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE. CORUh,C. I MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF NUREG/CR 6123: STUDIES OF THE PATTERN AND AGES OF POST. i GENERIC ISSUE 99. METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND NORTH CAROLINA. CHUN.M. NUREG/CR-4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT CORWIN,W.R. RISKS AND POTENTIAL FOR FI;SK REDUCTION.2 ION POWER NUREG/CR-4219 V04 N2: HEAVY-SECTION STEEL TECHNOLOGY PLANT. Draft Report For Comment. PROGRAM. Semiannual Progress Report For April-September 1937. i l' l

Personal Author index 65 NUREG/CR 4219 V05 N1: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR-4662: CLOSEOUT OF IE BULLETIN 8018. MAINTENANCE PROGRAM. Semiannual Progress Report For October 1987 - March OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING 1988. PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP-NUREG/CR 4888: PRESSURIZED-THERMAL-SHOCK TEST OF 6-lNCH TURE. THICK PRESSURE VESSEL S PTSE-2: Investigation Of Low Teanng Re- NUREG/CR.4665: CLOSEOUT OF IE BULLETIN 83-08ILECTRICAL sistance And Warm Prestressing. CIRCUlT BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN CORYNEN.G.C. USE IN SAFETY.RELATED APPLICATIONS OTHER THAN THE RE-ACTOR TRIP GYSTEM NUREG/CR-5242: A FAST BOTTOM-UP ALGORITHM FOR COMPUT. ING THE CUT SETS OF NONCOHERENT FAULT TREES. NUREG/CR 4932: CLOSEOUT OF IE BULLETIN BO-03: LOSS OF CHAR. COAL FROM STANDARD TYPE il,TWO-INCH, TRAY ADSORBER COSTAINJ.K. CELLS. NUREG/CR-5123: STUDIES OF THE PATTERN AND AGES OF POST. NUREG/CR.4933: CLOSEOUT OF IE BULLETIN 8019. FAILURES OF METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND MERCURY-WETTED MATRIX RELAYS IN REACTOR PROTECTIVE NORTH CAROLINA. SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED BY COMBUSTION ENGINEERING. COSTANTINO,C.J. NUREG/CR-4935: CLOSEOUT OF IE BULLETIN 8502:UNDERVOL. NUREG/CR-4784: INFLUENCE OF GROUNDWATER ON SOIL-STRUC- TAGE TRIP ATTACHMENTS OF WESTINGHOUSE DB-50 TYPE RE. TURE INTERACTION. ACTOR TRIP BREAKERS. C0ZZUOLJ.M. NUREG/CR-5189: CLOSEOUT OF IE BULLETIN 79-26: BORON LOSS FROM BWR CONTROL BLADES. NUREG/CR-5072: DECAY HEAT REMOVAL USING FEED AND-BLEED FOR U.S. PRESSURIZED WATER REACTORS NUREG/CR 5190: CLOSEOUT OF fE BULLETIN 8014: DEGRADATION OF BWR SCRAM DISCHARGE VOLUME CAPABILITY. CRAGNOLINO.G NUREG/CR-5191: CLOSEOUT OF lE BULLETIN 80-17: FAILURE OF 76 NUREG/CR-5156: REVIEW OF EROSION CORROSION IN SINGLE- OF 185 CONTROL RODS TO FULLY INSERT DURiNG A SCRAM AT l PHASF FLOWS. A BWR. CRAPO,H.S. DECKERJ.R. NUREG/CR-5043: CONTAINMENT PENETRATION SYSTEM (CPS) NUREG/CR-5198: INHALED (239)PUO(2) AND/OR TOTAL-BODY TESTS UNDER ACCIDENT LOADS. GAMMA RADIATION.Early Mortality And Morbidity in Rats And Dogs. CULLEN.W.H. DELARCHE G. NUREG/CR-4828: FATIGUE CRACK GROWTH OF PART-THROUGH NUREG/CR-5078 Vot: A RELIABILITY PROGRAM FOR EMERGENCY CRACKS IN PRESSURE VESSEL AND PIPING STEELS. Air Environ- DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Program ment Resutts. Structure. CUSHING,C.E. DEMOSS,G. NUREG/CR-5047: RADIONUCLIDES ACCUMULATION BY AQUATIC BIOTA EXPOSED TO CONTAMINATED WATER IN ARTIFICIAL ECO- NUREG/CR-5248- PRIORITLZATION OF TIRGALEX-RECOMMENDED SYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH THE COMPONENTS FOR FURTHER AGING RESEARCH' GROUND-DEMOSS,G.M. CZAJKOWSKI.C. fWUREG/CR-5078 V01: A RELIABILITY PROGRAM FOR EMERGENCY I l NUREG/CR 5145: FAILURE INVESTIGATION OF 3M SERIES 900 DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Program STATIC ELIMINATORS. Structure. NUREG/CR-5156: REVIEW OF EROSION CORROSION IN SINGLE-PHASE FLOWS. DENHAM D.H. DAEMENJ.J. NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW LEVEL RADIOAC. NUREG/CR.5129 EXPERIMENTAL ASSESSMENT OF THE INFLUENCE TiVE WASTE DISPOSAL Environmental Monitonng And Surveillance OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF Programs. DRIED CEMENT BOREHOLE SEALS. NUREG/CR 5130: BENTONITE BOREHOLE PLUG FLOW TESTING DENSON,R.H. WITH FIVE WATER TYPES. NUREG/CR-5041 V02: RECOMMENDATIONS TO THE NRC FOFv DALLMAN R.J REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL l ' NUREG/CR 5225: AN OVERVIEW OF BWR MARK 1 CONTAINMENT RADIOACTIVE WASTE DISPOSAL Task 2b EarttvMounded Concrete VENTING RISK IMPLICATIONS. Od 8 ' DANDINI,V.J. DENTENJ.G. NUREG/CR 4763: SAFETY-RELATED EQUIPMENT SURVIVAL IN HY- NUREG/CR-5171: FLOW VISUAll2ATION STUDY OF POST CRITICAL DROGEN BURNS IN LARGE DRY PWR CONTAINMENT DJILDINGS. HEAT FLUX REGION FOR INVERTED BUDBLY, SLUG AND ANNULAR FLOW REGIMES. DAVIS.E.C. NUREG/CR-5229. ANNUAL REPORT OF THE TMi-2 EPICOR-il RESfN/ DEY,M.K. LINER INVESTIGATION. Low Level Weste Data Base Development NUREG.1333 DAFT FC: MAINTENANCE APPROACHES AND PRAC-ProDram For Fiscal Year 1988. TICES IN SELECTED FOREIGN NUCLEAR POWER PROGRAMS AND DAVIS.P. OTHER U S. INDUSTRIES. Review And Lessons Learned. Draft Report For Comment NUREG/CR.4b20 VO1: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES BWR. MARK I CONTAIN. NUREG/CP 0099: PROCEEDINGS OF THE PUBLIC WORKSHOP FOR MENT DESIGN- NRC RULEMAKING ON MAINTENANCE OF NUCLEAR POWER PLANTS. N E /CR-4881: FlSSION PRODUCT RELEASE CHARACTERISTICS *

• INTO CONTAINMENT UNDER DESIGN BAS!S AND SEVERE ACCl- NUREG/CR-5180: CHEMICAL DECONTAMINATION AND CHEMICAL DENT COhlDrilONS' CLEANING OF LWR COMPONENTS AND POSSIBLE INTERACTIONS WITH METALLURGICAL AGING EFFECTS.

DAWSON J.F. NURE G/CR-5144 ACOUSTIC EMISSION SYSTEM CALIBRATION AT DIMENNA R.A. WATTS BAR UNIT 1 NUCLE AR RE ACTOR. NUREG/CR-5194. RELAP5/ MOD 2 MODELS AND CORRELATIONS. DEAN.R.S. DOCTOR.P.C. NUREG/CR-4523 CLOSEOUT OF IE BULLETIN 8013 CRACKING IN NUREG/CR.5185. STEAM GENERATOR GROUP PROJECT. Task 13 CORE SPRAY SPARGERS Fanal Report: Nondestructwe Examination (NDE) Validatton.

                                                                                                                                                                                                                     ] l l

l "66. Personal Author index L l DOCTOR,5.R. . . NUREG-1329: ENTRY / EXIT CONTROL AT FUEL FABRICATION FACILi-NUREG/CR-5075: THE SAFT UT REAL TIME INSPECTION SYSTEM - TIES USING OR POSSESSING FORMULA QUANTITIES OF STRATE-OPERATIONAL PPINCIPLES AND IMPLEMENTATION. GIC SPECIAL NUCLEAR MATERIAL DODD,C.V- . DYCUS F.M. NUREG/CR 5061: THREE-FREQUENCY EDDY. CURRENT .INSTRU- NUREG/CR-5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU. MENT. CLEAR ONE - UNIT 1 Volume 1, Program For inspectors.

                                                                                                            . NUREG/CR-5021 V02: USER'S GUIDE FOR PRISIM ARKANSAS NU-
                                                                                                                           ~

UR G CR4315 V09 'R1: EVALUATION OF NUCLEAR FACILITY DE-COMMISSIONING PROJECTS. Summary Status Report,Three Mile - EBRAHIMI,F. , . Island Unit 2, Radioactive Waste And Laundry Shipments. NUREG/CR 5063: DEVELOPMENT OF A MECHANISTIC UNDER-STANDING OF RADIATION EMBRITTLEMENT IN FIEACTOR PRES-SURE VESSEL STEELS. Final Report. UR R 5'070: ANALYSIS OF NATURAL-CONVECTION PHENOM-ENA IN A 3 LOOP PWR DURING A TMLB' TRANSIENT USING THE EDD Y,P.A-COMMIX CODE. NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW DONNELLYJ.P. CRITERIA FOR ALTERNATIVE METHODS OF LOW LEVEL RADIOAC-NUREG 1297: PEER REVIEW FOR HIGH LEVEL NUCLEAR WASTE RE. TIVE WASTE DISPOSAL Environmental Monitonng And Surveillance Programs. - FOSITORIES. Genene Technical Position. NUREG-1298 QUALIFICATION OF EXISTING DATA FOR HIGH-LEVEL - NUCLEAFt WASTE REPOSITORIES.Genenc Technical Position. D F DONOVAN.R.W. . . ALPHA AND BETA IRRADIATION OF THE LUNG. Phase ll Report.

      ' NUREG 0654 S01 RO1: CRITERIA FOR PREPARATION AND EVALUA.

TlON OF RADIOLOG! CAL EMERGENCY RESPONSE PLANS AND ELLISON B.C. . PREPAREDNESS' IN SUPPORT OF NUCLEAR - POWER NUREG/CR-5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU-PLANTS. Criteria For Utility Offsite Planrung And Preparedness. CLEAR ONE UNIT 1. Volume 1. Program For inspectors NUREG/CR 5021 V02; USER'S GUIDE FOR PRISIM AHKANSAS NU : DOOLEYJL CLEAR ONE UNIT 1. Volume 2. Program For Regulators. NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA-TION SYSTEMS. ~ELTAWILA,F. . NUREG/CR4242: SURVEY OF LIGHT WATER REACTOR CONTAIN- NUREG/CR4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE-MENT SYSTEMS, DOMINANT FAILURE MODES AND MITIGATION VENTION AND MITIGATION FEATURES:BWR. MARK 1 CONTAIN-OPPORTUNITIES. Final Report. MENT DESIGN. NUREG/CR4243 VALUE/ IMPACT ANALYSIS FOR EVALUATING AL' NUREG/CR 4920 V02: ASSESSMENT OF SEVERE ACCIDENT PRE. NU /C 42 TR T I IMPLEMENTING A MITIGATION ME T ES GN POLICY FOR LIGHT WATER REACTORS. NUREG/CR-4920' V03: ASSESSMENT OF SEVERE ACCIDENT PRE-DOSANJH.S.S. VENTION AND MITIGATION FEATURES:BWR. MARK lli CONTAIN-NUR G/CR 5029 MELT PROGRESSION IN SEVERELY DAMAGED RE-NURE / 4920' V04: ASSESSMENT OF SEVERE ACCIDE:JT PRE-NUREG/CR-5109I RELOCATION OF METALLIC CONSTITUENTS IN VENTION AND MITIGATION FEATURES:PWR,LARGE DRY CON. CORE DEBRIS BEDS. TAINMENT DESIGN. NUREG/CR-4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE-DOVE.R.C. VENTION AND MITIGATION FEATURES:PWR.lCE-CONDENSER NUREG/CR4924: SEISMIC CATEGORY I STRUCTURES CONTAINMENT DESIGN. PROGRAM. Final Report. Fiscal Year 1983 - 1984 NUREG/CR-4998: THE SEISMIC CATEGORY l STRUCTURES EMRIT,R. PROGRAM.Results For Fiscal Year 1985. NUREG-0933 S07: A PRIORITIZATION OF CENERIC SAFETY ISSUES. NUREG/CR-5182: THE SEISMIC CATEGORY I STRUCTURES NUREG 0933 SOB: A PRIORfTIZATION OF GENERIC SAFETY ISSUES. PROGRAM.Results For FY 1086. EPSTEIN,M. DRISCOLL.J.W. NUREG/CR-5113; FINDINGS OF THE PEER FtEVIEW PANEL ON THE NUREG/CR-4960: CONTROL ROOM HABITABILITY SURVEY OF L1- DRAFT FIEACTOR FilSK REFERENCE DOCUMENT,NUREG-1150. , CENSED COMMERCIAL NUCLEAR POWER GENERATING STA- J TIONS. ERVtN N.E. NUREG-1304: REPORTING OF SAFEGUARDS EVENTS. J NUREG-1309. THE U.S. NUCLEAR REGULATORY COMMISSION PRO- ESCALANTE E. GRAM WITH STATE AND LOCAL GOVERNMENTS AND INDIAN NUREG/CR4735 V03: EVALUATION AND COMPILATION OF DOE TRIBES. WASTE PACKAGE TEST DATA. Biannual Report: February-July 1987. NUREG/CR-4735 V04: EVALUATION AND COMPILATION OF DOE annual NponAugust M87 Janw NUREG Fid728: EQUIPMENT QUALIFICATION RESEARCH TEST OF gg' A HIGH RANGE RADIATION MONITOR. DUNCAN,A.B. IVA"8'O O-NUREG-1318: TECHNICAL POSITION ON ITEMS AND ACTIVITIES IN NUREG/CFI-5097: SIMULATION OF LIOUID AND VAPOR MOVEMENT i THE HIGH LEVEL WASTE GEOLOGIC REPOSITORY PROGRAM IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP SUBJECT TO QUALITY ASSURANCE FIEQUIREMENTS. /C U 5 ABLE O PES OF AUTHIGENIC MINERALS IN DUWWOODY.W.E. VARIABLY-SATURATED FRACTURED TUFF. NUREG/CR 4998: THE SEISMIC CATEGORY l STRUCTURES PROGRAM.Results For Fiscal Year 1995. FARMER.R. NUREG/CR-5182. THE SEISMIC CATEGORY l STRUCTURES NUREG/CR.5000: METHL90 LOGY FOR UNCERTAINTY ESTIMATION  ; PROGRAM.Results For FY 1986. IN NUREG-1150 (DRAFT).'.onclusions Of A Review Panel. DWYER,P.A. F AROUAHAR90N J. NUREG-1304: REPORTING OF SAFEGUARDS EVENTS. NUREG/CR.5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU-NUREG-1328: USE OF f'ERIMETER ALARMS AT FUEL FABRICATION CLEAR ONE UNIT 1. Volume 1. Program For inspectors. FACILITIES USING OR POSSESSING FORMULA QUANTITIES OF NUREG/CR-5021 V02: USER'S GUIDE FOR PRISIM ARKANSAS NU. STRATEGIC SPECIAL NUCLEAR MATERIAL CLEAR ONE UNIT 1. Volume 2.Prrrgram For Regulators. .g L

Personal Author index 67-FARRAR,C. NUREG/CR-5189: CLOSEOUT OF IE BULLETIN 79-26: BORON LOSS NUREG/CR4924; SEISMIC CATEGOF:Y l STRUCTURES FROM BWR CONTROL BLADES. PROGRAM. Final Report, Fiscal Year 1983 -1984. NUREG/CR-5190: CLOSEOUT OF IE DULLETIN 8014: DEGRADATION . OF BWR SCRAM DISCHARGE VOLUME CAPABILITY. FARRAR.C.R. NUREG/CR-5191: CLOSEOUT OF lE BULLETIN 80-17; FAILURE OF 76 NUREG/CR 4998: . THE SEISMIC CATEGORY l_ STRUCTURES OF 185 CONTROL RODS TO FULLY INSERT DURING A SCrlAM AT PROGRAM Results For Fiscal Yeer 1985. A BWR' NUREG/CR-5154: EXPERIMENTAL ASSESSMENT OF DAMPlNG IN LOW ASPECT RATIO, REINFORCED CONCFIETE ' SHEAR WALL FOURNEY,W.L. STRUCTURES- NUREG/CR4996: A REPORT ON THE ROUND ROBIN PROGRAM NUREG/CR-5182: THE SEISMIC CATEGORY I STRUCTURES CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARO PROGRAM Results For FY 1986. TEST METHOD FOR DETERMINING THE 3 LANE STRAIN CRACK FEINAUER,LR. ARREST FRACTURE TOUGHNESS.K(IA),OF FERRITIC MATERIALS. NUREG/CR-4312 V01; RELAp5/ MOD 2 CODE MANUAL. Volume .1: Code FOWLER,E.B Structure. Systems Models And Solution MetNxis. NUREG/CFi-5.70 A REVIEW OF RESEARCH CONDUCTED BY LOS - FERRIS,R.H. ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHASIS NUREG/CR-5185: STEAM GENERATOR GROUP PROJECT. Task 13 ON THE MAXEY FLATS KY, SHALLOW LAND WASTE BURIAL SITE. Fsnal Report; Nondestructive Examinaten (NDE) Vahdaten. FRAGOLA,J.R. FILIPY,R.E. NUREG/CR-5078 V01: A RELIABILITY PROGR*M FOR EMERGENCY NUREG/CR-5190: INHALED (239)PUO(2) AND/OR TOTAL-BODY DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Program GAMMA RADIATION.Early Mortahty And MortHdity in Rats And Dogs. _ Structure. FINFROCK,C. FRAKER,A. NUREG/CR-5148: DEBRIS DISPERSAL FROM PEAL: TOR CAVITIES NUREG/CR-4735 V03: EVALUATION AND COMPILATION OF DOE DURING HIGH. PRESSURE MELT EJECTION ACCIDENT SCENAR- WASTE PACKAGE TEST DATA. 3 annual Re .: February July 1987. IOS. NUREG/CR-4735 V04: EVALUATION AND PILATION OF DOE WASTE PACKAGE TEST DATA. Biannual Report: August 1987 - Janu-FITZPATRICK,R. ary 1988. NU% CR4WO V01: ASSESSMENT OF SEVERE ACCIDENT PRE-VhrvtlON AND MITIGATION FEATURES BWR, MARK I CONTAIN- FRANK,L. MENT DESIGN, NUREG/CR 5150; STEAM GENERATOR OPERATING NUREG/CR-4920 V02: ASSESSMENT OF SEVERE ACCIDENT PRE- EXPE91LNCE.Uptlate For 1984-1986. VENTION AND MITIGATION FEATURES:BWR. MARK ll CONTAIN-MNT DFPGN. FRID.W. NUREG/CR 920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG/CR4508: BEHAVIOR OF A CORIUM JET IN HIGH PRESSURE VENTION AND MITtGATION FEATURES:BWR. MARK lli CONTAIN- MELT EJECTION FROM A REACTOR PRESSURE VESSEL. MENT DESIGN. NUREG/CR 4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE- FRIESEL,M.A. VENTION AND MITIGATION FEATURES.PWR.LARGE DRY CON- NUREG/CR 5144: ACOUSTIC EMISSION SYSTEM CALIBRATION AT N 4 20 05: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES:PWR,1CE-CONDENSER FRYER,C.P. CONT AINMENT DESIGN. NUREG/CR4625: THE POSTlRRADIATION EXAMINATION OF THE DC NUREG/CR-5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE- MELT DYNAMICS EXPERIMENTS. MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF GENERIC ISSUE 90. FULLWOOD.R. NUREG/CR-5052: OPERATING EXPERIENCE AND AGING ASSESS-UREG/ R 021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU- WATER EA ORS CLE AR ONE . L' NIT 1.Volurae 1, Program For inspectors. NUREG/CR.5021 V02: USER'S GUIDE FOR PRlSIM ARKANSAS NU. GALYEAN,W.J. CLEAR ONE - UNIT 1. Volume 2, Program For ReDulators-NUREG/CR-4639 t/01: NUCLEAR COMPUTERIZED LIBRARY FOR AS-FLEMING,0. SESSING REACTOR RELIABILITY (NUCLARR). Volume LSummary De-NU R v Techn cal ues' NU -4639 VD? P1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide,Part 1: j FLEMING.K. Overview Of NUCLARR Data Retneval. 1 NUREG/CR-4780 V01: PROCEDURES FOR TREATING COMMON NUREG/CR-4639 V04 P2: NUCLEAR COMPUTERIZED LIBRARY FOR , CAUSE FAILURES IN SAFETY AND RELIABILITY ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide,Part 2:  ; STUDIES. Procedural Framework And Examples. jRud To at ns. FOLEY,W.J. ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide.Part 3: l NUREG/CR-4523: CLOSEOUT OF IE BULLETIN 8013: CRACKING IN NUCLARR System Desenption. -  ! CORE SPRAY SPARGERS. NUREG/CR 4633 V05 P1: NUCLEAR COMPUTERIZED LIBRARY FOR 4 l- NtIREG/CR-4662: CLOSEOU~ OF IE BULLETIN 8018: MAINTENANCE ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part 1: _i 07 ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING Summary Desenption. PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP. NUREG/CR-4639 V05 P2. NUCLEAR COMPUTERIZED LIBRARY FOR  ; TURE. . ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part 2: l' NUREG/CR-4865: CLOSEOUT OF IE BULLETIN 83 08: ELECTRICAL Human Error Probability (HEP) Estimates. CIRCULI BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN NURFG/CR-4639 V05 P3: NUCLEAR COMPUTERIZED LIBRARY FOR USE IN SAFETY RELATED APPLOATiONS OTHER THAN THE RE- ASGESSING REACTOR RELIABILITY (NUCLARR). Data Manual Part 3: j ACTOR TRIP SYSTEM. Hardware Component Failure Datt (HCFDL - NUREG/CR-4932: CLOSEOUT OF IE BULLETIN 8003 LOSS OF CHAR. NUREG/CR4639 V05 P4: NUCLEAR COMPUTERIZED LIBRARY FOR l COAL FROM STANDARD TYPE II.TWO.lNCH, TRAY ADSORBER - ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 4 CELLS, Summary Aggregations. 4 NUREG/CR4933: CLOSEOUT OF IE BULLETIN 80-19-FAILURES OF NUREG/CR 6225: AN OVERVIEW OF BWR MARK 1 CONTAINMENT i MERCURY WETTED MATRIX RELAYS IN REACTOR PROTECTIVE VENTING RISK IMPLICATIONS. 1 SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DE91GNED J SY COMBUSTION ENGINEER!NG. . CASKIN.C.E. . j NUREGICR4935. CLOSEOUT OF IE BULLETIN 85 02:UNDERVOL- NUREG-1S20: USE OF PERIMETER ALARMS AT FUEL FABRICATION l TAGE TRIP ATTACHMENTS OF WESTINGHOUSE DB-50 TYPE RE- FACILITIES USING OR POSSESSING FORMULA QUANTITIES OF j AC7OR TRIP BREAKERS. STRATEGIC SPECIAL NUCLEAR MATERIAL J i l

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68 Personal Author inder GAUDETTE,M.V, Processing And Revision.Part 2: Human Error Probab6hty Data Entry NUREG/CR-5130: BENTONITE BOREHOLE PLUG FLOW TESTING And Revision Procedures. WITH FIVE WATER TYPES. NUREG/CR4639 V03 P3: NUCLEAR COMPUTERIZED UBRARY FOR GEtSENDORFER,C. ASSESSING REACTOR RELIABILITY (NUCLARR) Guide To Data

                                                                           ')rocessing And Revision. Part 3: Hardware Component Failure Data NUREG/CR 4991: EVALUATION AND PROPOSED IMPROVEMENTS TO                 Entry And Revision Procedures.

EFFECTIVENESS OF U.S. NUC" EAR REGULATORY COMMISSION NUREG/CR4639 V04 P1: NUCLEAR COMPUTERIZED UBRARY FOR HUREG/C 24 : EL NT OF PROGRAMMATIC PERFORM-ANCE INDICATORS. U RR a NUREG/CR-4639 V04 P2: NUCLEAR COMPUTERIZED LIBRARY FOR GELHAR LW. ASSESSING REACTOR REllA91LfTY (NUCLARR). User's Guide,Part 2: NUREG/CR.5094: APPLICATION OF STOCHASTIC METHODS TO THE Guide To Operatens. SIMULATION OF LARGE SCALE UNSATURATED FLOW AND NUREG/CR 4639 V04 P3: NUCLEAR COMPUTEF!' ZED UBRARY FOR TRANSPORT. ASSESSING REACTOR REUABILITY (NUCLMr UWs Guide.Part 3: NUCLARR System Desenption. GENTILLON,C.D. NUREG/CR-4639 V05 P1: NUCLEAR COMPUTERIZED UBRARY FOR NUREG/CR 4639 V0t NUCLEAR COMPUTERIZED UBRARY FOR AS- ASSESSING REACTOR REUABILITY (WUCLARA). Data ManuaLPart 1: SESSING REACTOR REUABILIFY (NUCLARR). Volume !. Summary De. Summary Desenplion. senpton. NUREG/CR4639 VOS P2: NUCLEAR COMPUTERIZED UBRARY FOR NUREG/CR4639 V04 P1: NUCLEAR COMPUTERIZED UBRARY FOR ASSESSING REACTOR REUABluTY (NUCLARR). Data Manual.Part 2 ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide,Part 1: Human Error Probability (HEP) Estimates.  ! Overview Of NUCLARR Data Retneval- NUREG/CR 4639 VOS P3: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR4639 V04 P2: NUCLEAR COMPUTERIZED UBRARY FOR ASSESSING REACTOR REUABILITY (NUCLARR). Data Manual,Part 3: ASSESSING REACTOR REUABlWTY (NUCLARR). User's Guide.Part 2: Haidware Component Failure Data (HCFD). Guide To Operatons. NUREG/CR-4639 V04 P3: NUuLEAR COMPUTERIZED UBRARY FOR NUREG/CR4639 VOS P4: NUCLEAR COMPUTER 1 ZED UBRARY FOR ASSESSING REACTOR REUABILITY (NUCLARR). Data Manual,Part 4: l ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide.Part 3: NUCLARR System Desenpton. Sumniary Apgahons. NUREG/CR 5050: ANNOTATED BIBLIOGRAPHY OF REUABILITY AND RISK DATA SOURCES. N REG I 4639 V01: NUCLEAR COMPUTERIZED UBRARY FOR AS-GERTMAN,D,l. SESSING REACTOR REUABluTY (NUCLARR).'/olume l aummary De- i NUREG/CR-4639 V01: NUCLEAR COMPUTERIZED UBRARY FOR AS. senption. j SESSING REACTOR RELIABILITY (NUCLARR). Volume LSummary De- NUREG/CR4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY FOR senption. ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data NUREG/CR4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY FOR Processing And Revision. Part 1: Technical Overview. ASSES $ LNG REA!, TOR RELIABILITY (NUCLARR) Guide To Data NUREG/CR4639 V03 P2: NUCLEAR COMPUTERIZED UBRARY FOR Processmg And Revision. Part 1: Technical Overview ASSESSING REACTOR REUABILITY (NUCLARR). Guide To Data 1 NUREG/CR-4639 V03 P2: NUCLEAR COMPUTERIZED UBRARY FOR Processmg And Revision.Part 2: Human Error Probability Data Entry ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data And Revision Procedures. Processing And Revison.Part 2: Human Error Probabihty Data Entry NUREG/CR 4639 V03 P3: NUCLEAR COMPUTERIZED LfBRARY FOR And Revision Procedures. ASSESSING REACTOR REUABluTY (NUCLARR)Gude To Data 1 NUREG/CR-4639 V03 P3: NUCLEAR COMPUTERIZED UBRARY FOR Processing And Revision. Part 3: Hardware Component Failure Data l AENSSING REACTOR RELIABILITY (NUCLARR).Gude To Data Entry And Revision Procedures. Prassing And Revision. Part 3: Hardware Component Failure Data i NUREG/CR-4639 V04 P1: NUCLEAR COMPUTERIZED LIBRARY FOR I i NF C- 39 4 NN. EAR COMPUTERIZED UBRARY FOR Ove e O NUCLARR Da a ASSESSING REACTOR REUABtWTY (NUCLARR). User's Guide,Part 1: r al l NUREG/CR-4639 V04 P2: NUCLEAR COMPUTERIZED UBRARY FOR NU /CR 9 ASSESSING REACTOR REUABILITY (NUCLARR). User's Guide,Part 2: 2 UL F COMPUTERIZED UBRARY FOR ASSESSING REACTOR REUABluTY (NUCLARR) User's Guide,Part 2: NUR /CR 39 P3: NUCLEAR COMPUTERIZED UBRARY FOR NURE / 4$9 0 P3: NUCLEAR COMPUTERIZED UBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide Part 3: NC RR t m Descr p on NUREG/CR463 P1 N LEAR COMPUTERIZED LIBRARY FOR NUREG/CR-4639 V05 P1: NUCLEAR COMPUTERIZED UBRARY FOR AGSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part 1: ASSESSING REACTOR RELtABluTY (NUCLARR). Data Manual.Part 1: Summary Desenption. Summary Desenption. NUREG/CR-4639 V05 P2: NUCLEAR COMPUTERIZED UBRARY FOR NUREG/CR 4839 VOS P2: NUCLEAR COMPUTERIZED UBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part 2: ASSESSING RZACTOR REUABILITY (NUCLARR). Data Manual,Part 2: Human Error Probability (HEP) Estimates. Human Error Probabihty (HEP) Estimates. NUREG/CR-4639 VOS P3 NUCLEAR COMPUTERIZED UBRARY FOR NUREG/CR4639 V05 P2: NUCLEAR COMPUTERIZED UBRARY FOR ASSESSING REACTOR REUABluTY (NUCLARR). Data Manual.Part 3: ASSESSING REACTOR REUABILITY (NUCL ARR). Data Manual,Part 3: Hardware Component Failure Data (HCFD). Hardware Component Failure Data (HCFD) NUREG/CR4639 V05 P4: NUCLEAR COMPUTERIZED UBRARY FOR NUREGICR4639 VOS P4. NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR REUABluTY (NUCLARR). Data Manual Part 4: ASSESSING REACTOR REUA81LITY (NUCLARR). Data Manval,Pa,14: Summary Aggregations. c Summary Aggregations. GILPIN H. CHADI All N. NUREG/CR 5264: GUIDE FOR UCENSING EVALUATIONS USING NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM - PHASE ILSixth CRAC2.A Computer Program For Calculating Reactor Accident Conse-Program Report October 1986 September 1987. quences. RilTTER,J. GILPIN,H.E. NUREG-1220: SOURCE TERM ESilMATION DURING INCIDENT RE-SPONSE TO SEVERE NUCLEAR POWER PLANT ACCIDENTS. NUREG/CR-5106: USER'S GUIDE FOR THE TACTS COMPUTER CODE. CILBERT B.G. GINSBERG.T. NUREG/CR4639 VD1: N'; CLEAR COMPUTERIZED UBRARY FOR AS. NUREG/CR-5146: DEBRIS DISPERSAL FROM REACTOR CAVITIES SESSING REACTOR REUABILITY (NUCLARR)V0lume LSummary Do. DURING HIGH-PRESSURE MELT EJECTION ACCIDENT SCENAR. senption IOS. NUREG/CR4639 V03 P1: NUCLEAR COMPU'ERIZED UBRARY FOR ASSESSING REACTOR REUABluTY (NUCLARR) Guide To Data GLOVER,L Processing And Revisio.. Par' 1: Technical Overview NUREG/CR-5123. STUDIES OF THE PATTERN AND AGES OF POST-NUREG/CR-4639 V03 P2: NUCLEAR COMPUTERIZED USRARY FOR METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND ASSESSING REACTOR REUABILITY (NUCLARR) Guide To Data NORTH CAROLINA.

l i Personal Author index 69-GLOZMAN,V. HALE,F.V. .

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NUREG/CR.5031: SIGNIFICANCE OF IN-STRUCTURE GENERATED NUREG/CR 4864 V01: THERMODYNAMIC TABLES FOR NUCLEAR i MOTION IN SEISMIC QUALIFICATION TESTS OF CABINET MOUNT. WASTE ISOLATION. Aqueous Solutions Database. j ED ELECTRICAL DEVICES. ' HALL,D. GODFREY,P. NUREG/CR-4735 V03: EVALUATION AND COMPILATION OF DOE NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE WASTE PACKAGE TEST DATA. Biannual Report-February-July 1987. SUB COMPONENT AND SUBSYSTEM LEVELFinal Report. HALL,T.E. GOLDMAN P. NUREG/CR4998: 'THE SEISMIC CATEGORY l STRUCTURES NUREG/CR-5075: THE SAFT UT REAL TIME INSPECTION SYSTEM - PROGRAM.Results For Fiscal Year 1985 OPERATIONAL PRINCIPLES AND IMPLEMENTATION. NUREG/CR-5182: THE SEISMIC CATEGORY I STRUCTURE 3 HALLE,H.J. . i PROGRAM.Results For FY 1986. NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND GORE B.F. FLOW VISUALIZATION STUDIES AND TECHNIQUES RELEVANT TO NUREG/CR 5058. PRA APPLICATIONS PROGRAM FOR 'NSPECTION THE STUDY OF EROSION-CORROSION OF REACTOR PIPING SYS-TEMS. ) AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50-313.(Arkansas . Power And Light Company) HAMMICK,E.M. GRANT,T. NUREG/CR-5094: APPLICATION OF STOCHASTIC METHODS TO THE a NUREG/CR 5227: FITNESS FOR DUTY IN THE NUCLEAR POWER SIMULATION OF LARGE-SCALE UNSATURATED FLOW AND I INDUSTRY.A Review Of Technicallasues. TRANSPORT. GRAVES,G.H. HAMMOND,R.P. . NUREG/CR-5096: EVALUATION OF SEALS FOR MECHANICAL PENE- NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA- ' TRATIONS OF CONTAINMENT BUILDINGS. TION SYSTEMS. GREGORY,W.S. NUREG/CR4242: SURVEY OF LIGHT WATER REACTOR CONTAIN. MENT SYSTEMS, DOMINANT FAILURE MODES AND MITIGATION NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANAL' YSIS OPPORTUNITIES. Final Report. HANDBOOK. NURFG/CR4243; VALUE/ IMPACT ANALYSIS FOR EVA'.UATING AL. TEFINATIVE MITIGATION SYSTEMS. N EG CR-4639 V01: NUCLEAR COMPUTERIZED LIBRARY FOR AS. SESSING REACTOR RELIABILITY (NUCLARR)Nolume l. Summary De. PO CY G AT EA S scnp90n- HANAUER,5. ' NUREG/CR 5000: METHODOLOGY FOR UNCERTAINTY ESTIMATION I NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE IN NUREG-1150 (DRAFT) Conclusions Of A Review Panel. l DRAFT REACTOR RISK REFERENCE DOCUMENT.NUREG-1150. HANSON.R.G. GUERHIERI.D. NUREG/CR-5194: RELAP5/ MOD 2 MODELS AND CORRELATIONS. NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM PHASE 11 Sixth HARRIS,J.C Program Report. October 1986 - September 1987. NUREG/CR-5144: ACOUShC EMISSION SYSTEM CALIBRATION AT SUNTHER,W.E. WATTS BAR UNIT 1 NUJLEAR REACTOR. NUREG/CR 4039 V01: IMPROVING MOTOR RELIABILITY IN lulCLEAR PO PLANTS. Volume 1. Performance Evaluation And Maintenance HARRl ,J.D NUREG/CR.5051: DETECTING AND MITIGATING BATTERY CHARGra DAMAGE ACCIDENTS:1986,A STATUS REPORT. 1 AND INVERTER AGING. NUREG/CR-4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE  ! NUREG/CR-5192: TESTING OF A NATURALLY AGED NUC'NR DAMAGE ACCIDENTS:1986.A STATUS FIEPORT. 1 POWER PLANT INVERTER AND BATTERY CHARGER. GUTHRIE,V.H. NUREG/CR-5058: PRA APPLICATIONS PROGRAM FOR INSPECTION i NUREG/CR 5021 V01: USER'S GUIDE FOR PR!SIM ARKANSAF NU- AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50-313.(Arkansas CLEAR ONE - UNIT 1. Volume 1 Program For Inspectors Power And Light Company) hUREG/CR-5021 VD2: USER'S DUIDE FOR PRislM ARKANSAS NU-CLEAR ONE - UNIT 1Nolume 2, Program For Regulators. HARRISON,S. NUREG/CR4735 V03: EVALUATION AND COMPILATION OF DOE i G 'R 99: EVALUATION Or MATERIALS OF CONSTRUCTION annual NWhaWuy N. FOR THE REINFORCED CONCRETE REACTOR CONTAINMENT HATCH.P.W. MODEL NUREG/CR-5099: EVALUATION OF MATERIALS OF CONSTRUCTION GWALTNE Y,R.C. F0F? THE REINFORCED CONCRETE REACTOR CONTAINMENT NUREG/CR 4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS EL METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND HAUTH,J. BRANCH CONNECTIONS. NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER HABCHi,$.D. INDUSTRY.A Review Of Technicalissues. NUREG/CR-5233: A COMPUTER CODE FOR FIRE PROTECTION AND F FilSK ANALYSIS OF NUCLEAR PLANTS. HAWKINS,dR 5151: PERFORMANCE BASED INSPECTIONS. NUREG/ NURE /C'R 315 VOD R1: EVALUATIOf9 OF NUCLEAR FACILITY DE. HAWTHORNE.JA COMMISSIONING PROJECTS. Summary Status Report,Three Mile NUREG/CR-5201: EXPERIMENTAL ASSESSMENTS OF GUNDREM-Island Unit 2, Radioactive Waste And Laundry Shipments. MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS STUDIES. HAGEN.E.W. NUREG/CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE HAZELTON,W.S. ( DAMAGE ACCIDENTS 1986.A STATUS REPORT NUREG/CR-4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE NUREG 0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE DAMAGE ACCIDENTS:1986,A STATUS REPORT. BOUNDARY PIPING Final Report. HAHN.F.F. HTMPHILL,G.M-NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED NUREG/CR-5044 ESTIMATION TECHNIQUES FOR COMMON CAUSE i ALPHA AND BET A IRRADIATION OF THE LUNG. Phase 11 Report. FAILURE EVENTS. '

70 Personal Author index HENDERSON W. HLASTALA,M.P.  ! NUREG/CR-5078 V02: A RELIABILITY PROGRAM FOR EMERGENCY NUREG/CR 5198: INHALED (239)PUO(2) AND/OR TOTAL-BODY l DIESEL GENE 9.ATORS AT NUCLEAR POWER GAMMA RADIATION.Early Mortahty And Mortadity in Rats And Dogs. PLANTS Maintenance. Surveillance And Condihon Monitonng. HOELZER.D.T. HENDRICKSON,J. NUREG/CR-5063: DEVELOPMENT OF A MECHAN!STIC UNDER. NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES-INDUSTRY.A Review Of Technicalissues. SURE VESSEL STEELS. Final Report. HENDRICKSON.P. HOFMANN'P" NUREG/CR 5218: FINANCIAL QUALIFICATIONS REVIEW OF APPLi-CANTS FOR NUCLEAR POWER PLANT CONSTRUCTION PERMITS. NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE NUREG/CR-5227; FITNESS FOR DUTY IN THE NUCLEAR POWER DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG-1150. INDUSTRY.A Review Of Technicalissues. NUREG/CR 5119. METALLOGRAPHIC EXAMINATION OF THE SEVERE FUEL DAMAGE SCOPING TEST (SFD-ST) FUEL ROD BUNDLE HENNICK,A. CROSE SEC* IONS. NUREG/CR 4523: CLOSEOUT OF IE BULLETIN 80-13. CRACKING IN CORE SPRAY SPARGERS HOFMAYER,C.H. MUREG/CR-4662: CLOSEOUT OF IE BULLETIN 80-18. MAINTENANCE NUREG/CR 4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING PLANT COMPONENTS (PHASE II). Motor Control PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP- Center. Switchboard.Panelboard And Power Supply. TURE. NUREG/CR-5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABl. MUREG/CR-4665. CLOSEOUT OF IE BULLETIN 83 08. ELECTRICAL NETS. CIRCUIT BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN USE IN SAFETY RELATED APPLICATIONS OTHER THAN THE RE- HORA.S.C. ACTOR TRIP SYSTEM. NUREG/CR-5032. MODELING TIME TO RECOVERY AND INITIATING NUREG/CR 4932: CLOSEOUT OF IE BULLETIN 80 0310SS OF CHAR' EVENT FREQUENCY FOR LOSS OF OFFSITE POWER INCIDENTS COAL FROM STANDARD TYPE II.TWO-INCH, TRAY ADSORBER AT NUCLEAR POWER PLANTS. CELLS. NUREG/CR-4933: CLOSEOUT OF IE BULLETIN 80-19. FAILURES OF HORSCHEL,D.S. MERCURY WETTED MATRIX RELAYS IN REACTOR PROTECTIVE NUREG/CR-5083: DESIGN. CONSTRUCTION AND INSTRUMENTATION SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED OF A 1/6-SCALE REINFORCED CONCRETE CONTAINMENT BUILD-BY COMBUSTION ENGINEERING ING' NUREG/CR 4935: CLOSEOUT OF IE BULLETIN 85-02:UNDERVOL-TAGE TRIP ATTACHMENTS OF WESTINGHOUSE DB 50 TYPE RE' HORST,C.L NURE /C 0 L UT OF IE BULLETIN 79-26: BORON LOSS NUREG/CR 5159: PREDICTION OF CHECK VALVE PERFORMANCE l FROM BWR CONTROL BLADES. AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. MUREG/CR-5190: CLOSEOUT OF IE BULLETIN 80-14: DEGRADATION l OF BWR SCRAM OISCHARGE VOLUME CAPABILITY. HSU.C. I F'UREG/CR-5191: CLOSEOUT OF IE BULLETIN BO 17. FAILURE OF 76 NUREG/CR-4999: ESTIMATION OF RISK REDUCTION FROM IM-l OF 1EE CONTROL RODS TO FULLY INSERT DURING A SCRAM AT PROVED PORV RELIABILITY IN PWRS. Final Report. HSU.C.J. HERRMANN.R.B. NUREG/CR-4920 V02: ASSESSMENT OF SEVERE ACCIDENT PRE-MUREG/CR 5165: SEISMOLOGICAL INVESTIGATION OF EARTH- VENTION AND MITIGATION FEATURES:BWR. MARK 11 CONTAIN-QUAKES IN THE NEW MADRID SEISMIC ZONE AND THE NORTH- MENT DESIGN. EASTERN EXTENT OF THE NEW MADRID SEISMIC ZONE Final NUREG/CR 4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE-Report. September 1981 December 1986. VENTION AND MITIGATION FEATURES.PWR.LARGE ORY CON-TAINMENT DESIGN. HESSE,D J. NUREG/CR-4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE-NUREG/CR 4057: CADET.A DECISION SUPPORT SYSTEM FOR LIGHT VENTION AND MITIGATION FEATURES:PWR.lCE-CONDENSER WATER REACTOR SAFETY- CONTAINMENT DESIGN. HESTER.O.V. HUANG,P.H MUREG/CR-5050: ANNU". ATED BIBLIOGRAPHY OF RELIABILITY AND RISK DATA SOURCES. NUREG/CR-5166: ELECTROCHEMICAL EVALUATION OF SOLID STATE PH SENSORS FOR NUCLEAR WASTE CONTAINMENT. HIGGINS J. NUREG/CR-5052: OPERATING EXPERIENCE AND AGING ASSESS. HUBBARD,G. MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR. NUREG-0797 SID: SAFETY EVALUATION REPORT RELATED TO THE IZED WATER REACTORS. OPERATION OF COMANCHE PEAK STEAM ELECTRIC STW N. UNITS 1 AND 2. Docket Nos. 50-445 And 50 446.(Texas Llile HIGGINS.J.C. e ,erahng Company) WUREG/CR-5140: VALUE-lMPACT ANALYSIS FOR EXTENSION OF NRC BULLETIN 8403 TO COVER ALL SAFETY RELATED MOVS. HUSZAGH,D. NUREG/CR-5140. VALUE lMPACT ANALYSIS FOR EXTENSION OF MR CI KANSAS NEBRASKA SEISMICITY STUDIES USING THE KANSAS NEBRASKA MICROEARTHOUAKE NETWORK. Final HUTTON,P.H. Report- NUREG/CR-5144: ACOUSTIC EMISSION SYSTEM CALIBRATION AT HINKLE.N.E. b0 #I NUREG/CR 4651 V02- DEVELOPMENT OF RIPRAP DESIGN CRITERIA HYATT.E.C. BY RIPRAP TESTING IN FLUMES. Phase ll. Followup invest gations. NUREG/CR-5090: EFFECTS OF TEMPERATURE AND HUMOITY ON HISER.A.L. RESPIRATOR FIT UNDER SIMULATED WORK CONDITIONS. NUREG/CR-5024: TENSILE ANC .,-R CURVE CHARACTERIZATION OF THERMALLY AGED CAST STAINLESS STEELS. IMAN.R.L NUREG/CR-5201: EXPERIMENTAL ASSESSMENTS OF GUNDREM. NUREG/CR-5032: MODELifG TIME TO RECOVERY AND INITIATING MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS EVENT FREQUENCY FOR LOSS OF OFFSITE POWER INCIDENTS STUDIES. AT NUCLEAR POWER PLANTS. HITCHCOCKJ.T. INTERRANTE,C. NUREG/CR 4625: THE POSTlRRADIATION EXAMINATION OF THE DC NUREG/CR-4735 V03: EVALUATION AND COMPILATION OF DOE MELT DYNAMICS EXPERIMENTS. WASTE PACKAGE TEST DATA. Biannual Report. February-July 1987.

Personal. Author index 711 NUREG/CR-4735 V04: EVALUATION AND COMPILATION OF DOE JONAS,0. WASTE PACKAGE TEST DATA. Biannual Report. August 1987. Janu. NUREG/CR-5149: EROSION CORROSION OF PWR ~ FEEDWATER ~ ary 1988. PIPING SURVEY OF EXPERIENCE. DESIGN WATER CHEMISTHY AND MATERIALS. IRWIN,G.R. NUREG/CR4996: A REPORT ON THE ROUND ROBIN PROGRAM JOYCE,J.A. CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD NUREG/CFr5142: DUCTILE TO BRITTLE TOUGMNESS TRANSITION TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK CHARAC.TERIZATION OF A533B STEEL. ARREST FRACTURE TOUGHN"!SS.K(IA).OF FERRITIC MATERIALS. . ISHIGAMI T. NUREG/CR 5000: A STUDY OF NEW ENGLAl1D SEISMICITY WITH NUREG/CR 4688 V02: OUANTIFICATION AND UNCERTAINTY ANALY, EMPHASIS ON MASSACHUSETTS AND NEW HAMPSHIRE. Final - SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT Report Covenn0The Penod 19761985. WATER REACTORS (OUASAR).Part II. Sensitivity Analysis Tech-rwques. NUREG/CR-5159: PREDICTION OF CHECK VALVE PERFORMANCE ISHILM. AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. NUREG/CR-5082: SIMULATION EXPERIMENTS ON TWO PHASE NAT. URAL CIRCULATION IN A FREON.113 FLOW VISUALIZATION LOOP. KAM,F.B.

     . NUREG/CR-5171: FLOW VISUALIZATION STUDY OF POST CRITICAL              NUREG/CR-4984: DEVELOPMENT OF A THREE-DIMENSIONAL FLUX HEAT FLUX REGION FOR INVERTED BUBBLY, SLUG AND ANNULAR               SYf#HESIS PROGRAM AND COMPARISON WITd 3-D TRANSPORT FLOW REGIMES.                                                        THEORY RESULTS.

NUREG/CR-5019: NEUTRON EXPOSURE FARAMETERS FOR THE IZENSON,M.G. METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY-SEC-NUREG/CR-5220 V01: DIAGNOSIS OF CONDENSATION-lNDUCED TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES. WATERHAMMER. Methods And Background. NUREG/CR-5049: PRESSURE VESSEL FLUENCE ANALYSIS AND NUREG/CR-5220 V02; DIAGNOSIS OF CONDENSATION-INDUCED NEUTRON DOSIMETRY. WATERHAMMER. Case Studies. KAMMERER C. . JACOBSON,J.A. NUREG 1311: FUNDING THE NRC TRAINING PROGRAM FOR NUREG/CR 4639 V02: NUCLEAR COMPUTERIZED LIBRARY FOR AS- STATES. SESSING REACTOR RELIABILITY (NUCLARR). Programmer's Guide. JACOBUS,M.J. NUREG/CR-5012: SIMILARITY PRINCIPLES FOR EOUlPMENT QUAll-NUREG/CR-4728: EQUIPMENT QUALIFICATION RESEARCH TEST OF - FICATION BY EXPERIENCE. A HIGH-RANGE RADIATION MONITOR. KANNINEN.M.F. JAECH J.L. NUREG/CP-0075: PROCEEDINGS OF CENI/NRC WORKSHOP ON NUREG/CR-4605: TRAINING MANUAL ON STATISTICAL METHODS DilC11LE PIPING FRACTURE MECHANICS. . FOR NUCLEAR MATERIAL MANAGEMENT. KASHIMA,K. JAQUISH,R.E. . NUREG/CP-0092: PROCEEDINGS OF THE SEMINAR ON LEAK- 3

      .NUREG/CP-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW                    BEFORE-BREAK. Progress in Regulatory Policies And Supporting Re- 'j CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOAC.                search.                                                            -

TiVE WASTE DISPOSAL Environmental Monitonng And Surveillance Programs. KASSIR.M.K. - NUREG/CR.4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER JARRELL.D.B. PLANT COMPONENTS (PHASE II). Motor Contros NUREG/CR-5248; PRIORITIZATION OF TIRGALEX-RECOMMENDED Center, Switchboard,Panelboard And Power Supply.' 'J COMPONENTS FOR FURTHER AGING RESEARCH. NUREG/CR-5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABI-NETS. JASTROW,J.D. NUREG/CR-5229 ANNUAL REPORT OF THE TMI-2 EPICOR-il RESIN / KASSNER,T,F, LINER INVESTIGATION. Low Level Waste Data Base Development NUREG/CR-4667 V04: ENVIRONMENTALLY ASSISTED CRACKING IN Program For Fmcal Year 1988. LIGHT WATER REACTORS. Semiannual Report, October 1986 March 1987. JENKS,R.P. NUREG/CR-5071: TRAC SUPPORT SOFTWARE. KASTENBERG,W.E. JOHNSEN G.W. NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA-TION SYSTEMS. NUREG/CR-4312 V01:RELAPS/ MOD 2 CODE MANUALVolpe 1: Code NUREG/CR4242: SURVEY OF LIGHT WATER REACTOR CONTAIN-Structure. Systems Models And Solution Methods. MENT SYSTEMS, DOMINANT FAILURE MODES AND MITIGATION l NUREG/CR4312 V02 R1: RELAPS/ MOD 2 CODE MANUAL. Volume 2: OPPORTUNITIES. Final Report. - i Users Guide And input Requirements. NUREG/CR 4243: VALUE/ IMPACT ANALYSIS FOR EVALUATING AL; JOHNSON,J. TERNATIVE MITIGATION SYSTEMS. NUREG/CR-4244: STRATEGIES FOR IMPLEMENTING A MITIGATION i NUREG/CR-5 tS1: PERFORMANCE BASED INSPECTIONS. POLICY FOR LIGHT WATER REACTORS. JOHNSON J.J- NUREG/CR-5113. FINDINGS OF THE PEER REVIEW PANEL ON THE DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG 1150. . NUREG/CR-5073. QUANTIFICATION OF MARGINS IN PIPING SYSTEM < SElSMIC RESPONSE, Methodologies And Dampin0 KASZA K.E.  ; JOHNSON,K.f. NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND FLOW VISUALIZATION STUDIES AND TECHNIQUES RELEVANT TO - ~l l NUREG/CR-5210: TECHNICAL FINDINGS DOCUMENT FOR GENERIC . THE STUDY OF EROSION CORROSION OF REACTCM PIPING SYS. l ISSUE 41: IMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE- TEMS. WATER SYSTEMS. KATHREN R.L. JOHNSON,R.W.

                                                                          - NUREG/CP-0093: PROCEEDINGS OF THE MEETING ON ULTRASEN-NUREG/CH-5194: RELAPS/ MOD 2 MODELS AND CORRELATIONS.                                                                                        t SITIVE TECHNIOUES FOR MEASUREMENT OF URANIUM IN BIO.

LOGICAL SAMPLES AND THE NEPHROTOXICITY OF URANIUM. I NUREG/CR4828. FATIGUE CRACK GROWTH OF PART-THROUGH KELLY,E.J. CRACKS IN PRESSURE VESSEL AND PIPING STEELS. Air Environ- NUREG/CR-5044: ESTIMATION TECHNIQUES FOR COMMON CAUSE i

      ' ment Results                                                           F ALLURE EVENTS.
                                                                                                                                                    ~

I i i d

 '72'        Personal Author index KEMPF,C.R.                        .

NUREG/CR-4312 V02 R1; RELAPS/ MOD 2 CODE MANUAL. Volume 2: NUREG/CR 5204: . LOW-LEVEL RADIOACTIVE WASTE SOURCE TEFIM Users Guide And input Atauirements. MODEL DEVELOPMENT AND TESTING. NUREG/CR 5194: RELAPS/ MOD 2 MODELS AND CORRELATIONS. KITCH,D.M. I KENNEDY,J.E. NUREG 1297: PEER FIEVIEW FOR HtGH-LEVEL NUCLEAR WASTE RE- NUREG/CR 4597 V02: AGING AND SERVICE WEAR OF AUXILIARY POSITORIES. Genesic Technical Position. FEEDWATER PUMPS FOR PWR NUCLEAR PLANTS Volume 2. Aging NUREG-1298: QUALIFICATION OF EXISTING DATA FOR HIGH-LEVEL Assessments And Monitonng Method Evaluations. l NUCLEAR WASTE REPOSITORIES Generic Technical Position.  ! NUREG 1318: TECHNICAL POSITION ON ITEMS AND ACTIVITIES IN KLAGES,J. THE HIGH-LEVEL WASTE GEOLOGIC REPDSITORY PROC 2 RAM NUREG/CR 5146: DEBRIS DISPERSAL FROM REACTOR CAVITIES SUBJECT TO QUALITY ASSURANCE REQUIREMENTS. DURING HIGH-PRESSURE MELT EJECTION ACCIDENT SCENAR-IOS.  ; KENT,D.B. NUREG/CR-4807: SURFACE-COMPLEXATION MODELING OF RADIO- S.M. ' NUCLIDE ADSORPTION IN SUBSURFACE ENVIRONMENTS. KLEIN'EG/CR-5147: NUR FUNDAMENTAL ATTRIBUTES OF A PRACTICAL' KHAN,T.A. CONFIGURATION MANAGEMENT PROGRAM FOR NUCLEAR PL ANT  ; f5 REG /CR4038: uPiiMtZATION OF THE CONTROL OF CONTAMINA- DESIGN CONTROL. l

                                                                                                                                                                                          ?

TlON AT NUCLEAR POWER PLANTS. NUREG/CR-5158 V01: WORLDWIDE ACTIVITIES ON THE FIEDUCTION ' KNAPP.R. .j OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS. NUREG/CR-5045: KANSAS-NEBRASKA SEISMICITY STUDIES USING  ; THE KANSAS-NEBRASKA MICROEARTHOUAKE NETWORK. Final i KHATIB-RAHSAR Report 1 NUREG/CR-4551 V5 DAF: EVALUATION OF SEVERE ACCIDENT J KNOROVSKY,G.A.' j RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER PLANT. Draft Report For Comment. NUREG/CR-5099: EVALUATION OF MATERIALS OF CONSTRUCTION i NUREG/CR-4688 V02: QUANTIFICATION AND UNCERTAINTY ANALY- FOR THE REINFORCED CONCRETE REACTOR CONTAINMENT

      ' SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT                                          MODEL WATER REACTORS (OUASAR).Part it: Sensitivity Analysis Tech-                                                                                                                       i niques.                                                                               KNUDSON R.                                                                                  1 NUREG/CR 4881: FISSION PRODUCT RELEASE CHARACTERISTICS                                     NUREG/CR 4555 R01: GENERIC COST ESTIMATES FOR THE DIS.

INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCI- POSAL OF RADIOACTIVE WASTES. DENT CONDITIONS. NUREG/CR-5164: A SIMPLIFIED MODEL FOR CALCULATING EARLY KOENIG,R.A. OFFSITE CONSEQUENCES FROM NUCLEAR REACTOR ACCI- NUREG/CR-5107: . HYDROGEOLOGIC CHARACTERIZATION OF DENTS. BASALTS.The Northern Ram Of The Columbia Platriau Physiographic KEATTAK.M.S. - NUREG/CR-4651 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA KOH,8.R. ' BY RIPRAP TESTING IN FLUMES. Phase ILFollowup investigations. NUREG/CR-5157:THE DEVELOPMENT OF APRILMOD2. A COMPUT. ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL-KILLEY,R.W' NUREG/CR- 4879 V01: DEMONSTRATION OF PERFORMANCE MOD- ING WATER NUCLEAR REACTORS. ELING OF A LOW-LEVEL WASTE SHALLOW LAND BURIAL SITE.A 3 Comparison Of Predictive Radionuchde Transport Modeling Versus KONO'EG/CR NUR 5227: FITNESS FOR DUTY IN THE NUCLEAR POWER Field Observations At The Nitrate Disposal Pit Site. Chalk River Nuclear INDUSTRY.A Heview Of TechnicalIssues. Labs. KONZEK,GJ. KIM.D.H. NUREG/CR 0130 ADD 04: TECHNOLOGY. SAFETY AND COSTS OF N'JREG/CR-5157: THE DEVELOPMENT OF APRILMOD2 - A COMPUT-DECOMMISSIONING A REFERENCE PRESSURIZED WATER REAC-ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOtL. TOR PCWER STATION. Technical Support For Decommissioning Mat-ING WATER flVCLEAR REACTORS. ters Related To Preparation Of The Final Decommissioning Rule. KIM,S.H. NUREG/CR-0672 ADD D3: TECHNOLOGY, SAFETY AND COSTS OF NUREG/CR 5157: THE DEVELOPMENT OF APRIL. MOD 2 - A COMPUT. ' DECOMMISSIONING A REFEREWCE BOILING WATER REACTOR POWER STATION. Technical Support For Decommissioning Matters  ! ER CODE FOR CORE MELTDOWN ACCIDEffT ANALYSIS OF BOIL-ING WATER NUCLEAR REACTORS. Related To Preparation Of The Final Decommissioning Rule. KING D.B. KOO.W.H. NUREG/CR-4763: SAFETY-RELATED EOU!PMENT SURVIVAL IN HY- NUREG-0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION .l DROGEN BURNS IN LARGE DRY PWR CONTAINMENT DulLDINGS. ANC, PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE j BOUNDARY PIPING. Final Report.  : KINQf.K. - NUREG/CR4113: FINDINGS OF THE PEER REVIEW PANEL ON THE KORNASIEWICZ,R. j DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG-1150. NUREG-0B00 02.4.2 R3: STANDARD FIEVIEW PLAN FOR THE REVIEW I OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER l oposed Rewsion 3 h W Mon 2M,

           $G-1226: DEVELOPMENT AND UTILIZATION OF THE NRC                                           p       ,

o, POLICY STATEMENT ON THE REGULATION OF ADVANCED NU- NUR5G-0800 02.4.3 R3: STANDARD REVIEW PLAN FOR THE REVIEW j CLEAR POWER PLANTS. OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER l PLANTS (WR Editior..Propored Revision 3 To SRP Sect on 2.4.3, l KIRCHNERJ.R. .j NUREG/CR-5021 V01: LISER'S GUIDE FOR PRISIM ARKANSAS NV. " Probable Maximum Flood 9MF) On Streams & Rivers." For Com-CLEAR C5NE UNIT 1. Volume 1 Program For insp.ectors. ment l NUREG/CR4021 V02: USER'S 6UIDE FOR PRISIM ARKANSAS NU.  ! CLEAR ONE UNIT 1. Volume 2.Progra;n For Regulato<s. 8 Rd"G'/CR-5000: METHODOLOGY FOR UNCERTAINTY EST KIRKMAN,J.Q. IN NUREG-1150 (DRAFT). Conclusions Of A Review Panel. NUREG/CR.5021 V01: USER'S GulCE FOR PRISIM ARKANSAS NU. CLE AR ONE - UNIT 1 Volume LProgram For inspectors KRAMER,G. NUREG/CR4021 V02: USER'S GUIDE FOR PRISIM ARKANSAS NU- NUREG/CR-4082 V06: DEGRADFD PIPING PROGRAM + PHASE II. Sixth CLEAR ONE - UNIT LVolume 2. Program For Regulatc<s. Program Report. October 1986 September 1987. KISER.D.M. KRElDER,K.G. ' . NUREG/CFi 4312 VD1: HELAP5/ MOD 2 CODE MANUALVolume 1: Code NUREG/CR-5166: ELECTHUCHEMICAL . EVALUATION OF SOLID I I Structure. Systems Models And Solution Methods. STATE PH SENSORS FOR NUCLEAR WASTS CONTAINMENT.

                                                                                                                                                                                    ~

Personal Author index 73 KRIPPS.L. . . LECKIE J,0. NUREG/CR-5078 V02: A RELIABILITY PROGRAM FOR EMERGENCY VJREG/CR-4807 SURFACE-COMPLEXATION MODELING OF RADIO-DIESEL GENERATORS AT NUCLEAR POWER NUCLIDE ADSORPTION IN SUBSURFACE ENVIRONMENTS. PLANTS. Maintenance. Surveillance And Condition Monitoring. LEE,0.W. KRISHNAMOORTHY NUREG/CR-5063: DEVELOPMENT OF A MECHANISTIC UNDER- NUREG/CR-4351 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA - BY RIPRAP TESTING IN FLUMES. Phase ILFollowup investigations. STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES-SURE VESSEL STEELS. Final Report LEE R.Y. NUREG 1236: THERMAL-HYDRAULIC RESEARCH PLAN FOR - BAB-NUREG/ -5198: INHALED (239)PUO(2) AND/OR TOTAL. BODY GAMMA RADIATION.Early Mortahty Act Morbedity in Rats And Dogs Lgg,g,y, KUNKEL.J.R NUREG/CR-5082: SIMULATION EXPERIMENTS ON TWO-PHASE NAT. NUREG/CR-5240: COMPARATIVE EVALUATION OF SELECTED CON- URAL CIRCULATION IN A FREON-113 FLOW VISUALIZATION LOOP. TINUUM AND DISCRETE-FRACTURE MODELS Emphasis On Disper- NUREG/CR 5133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW

                                  - sivky Calculations For Apphcation To Fractured Geologic Media, Cres-                     THROUGH INTERGRANULAR STRESS CORROSION CRACKS.

ton Study Area. Eastern Washington. LEEDS,E. KUPPERMAN.D.S. NUREG-1275 V03: OPERATING EXPERIENCE FEEDBACK REPORT . NUREG/C44813 RO1: ASSESSMENT OF LEAK DETECTION SYSTEMS SERVICE WATER SYSTEM FAILURES AND' FOR LWRS. DEGRADATIONS Commercial Power Reactors. NUREG/CR-5134: APPLICATION OF ACOUSTIC LEAK DETECTION TECHNOLOGY FOR THE DETECTION AND LOCATION OF LEAKS IN LEHNER,J.R. LIGHT WATER REACTORS. NUREG/CR4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE. VENTION AND MITIGATION FEATURES:BWR, MARK I CONTAIN-KURTZ,RJ. MENT DEStGN. NUREG/CR-2336: STEAM GENERATOR TUBE INTEGRITY - NUREG/CR-4920 V02: ASSESSMENT OF SEVERE ACCIDENT PRE-PROGRAM. Phase 11 Final Report. VENTION AND MITIGATION FEATURES:BWR, MARK #1 CONTAIN. NUREG/C45016: COMPENDIUM AND COMPARISON OF INTERNA- MENT DESIGN. TIONAL PRACTICE FOR PLUGGING, REPAIR AND INSPECTION OF NUREG/C44920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE. STEAM GENERATOR TUBING. VENTION AND MITIGATION FEATURES:BWR, MARK 111 CONTAIN-KUZAY,T.M. MENT DESluN. NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND NUREG/CR-4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE. FLOW VISUAUZATION STUDIES AND TECHNIQUES RELEVANT TO VENTION AND MITIGATION FEATURES:PWR LARGE DRY CON-TAINMENT DESIGN. THE STUDY OF EROSION CORROSION OF REACTOR PIPING SYS* TEMS. NUREG/CR-4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION . FEATURES:PWR,1CE-CONDENSER LA CHANCE,J.L. CONTAINMENT DESIGN. NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS RELA. LEVY LS TlVE TO PRESSURIZED THERMAL SHOCK' NUREG/C45248: PRIORITIZATION OF TIRGALEX. RECOMMENDED LAGRONE.D.L. COMPONENTS FOR FURTHER AGING RESEARCH. NUREG/CR-4651 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY RIPRAs' TESTING IN FLUMES. Phase ll. Followup investigations. d LAHEY,R.T. NUREG CR.4728: EQUIPMENT QUALIFICATION RESEAPOH TEST OF A HIGH-RANGE RADIATION MONITOR.- NUR8EG/CR 5157; THE DEVELOPMENT OF APRILMOD2. A COMPUT-ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL, LEWISJt. ING WATER NUCLEAR REACTORS. NUREG/CR-5051: DETECTING AND MihGATING BATTERY CHARGER AND INVERTER AGING. LAl,Y.L. NUREG/CR.5198: INHALED (239)PtJO(2) AND/OR TOTAL-BODY LIGGETT.W. GAMMA RADIATION.Early Mortality And Mortndity in Rats And Dogs. NUREG/CR-4735 V03: EVALUATION AND COMPILATION OF DOE LAM,P. WASTE PACKAGE TEST DATA. Biannual ReportFebruary-July 1987 NUREG-1275 V03: OPERAT'NG EXPERIENCE FEEDBACK REPORT - LINDOUIST,M.R. SERVICE WATER SYSTEM FAILURES AND NUREG/CR 5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE DEGRADATIONS Commercial Power Reactors. PREDICTION METHODS FOR PIPING. LAND,J.F. ! LINER.R.T. NUREG/CR-47DB VC2: PROGRESS IN EVALUATION OF RADIONU-NUREG/CR.5151: PERFORMANCE-BASED INSPECTIONS. CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-i.EVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report For LINZER M. . April 1986. September 1987. . NUREG/CR-4735 V03: EVALUATION AND COMPILATION OF DOE WASTE PACKAGE TEST DATA. Biannual Report: February-July 1987, NUREG/CR 4082 V06: DEGRADED PIPING PROGRAM . PHASE ll. Sixth LOBBIN,F. Prograr : Report. October 1986 - September 1987 NUREG/CR.4555 R01: GENERIC COST ESTIMATES FOR THE DIS-POSAL OF RADIOACTIVE WASTES. NUREG/CR-5194: RELAP5/ MOD 2 MODELS AND CORRELATIONS- LOFARO,R. LARSON,T.K. NUREG/CR 4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR f61 REG /C45194. RELAP5/ MOD 2 MODELS AND CORRELATIONS. POWER PLANTS. Volume 2.Functonal Indicator Tests On A Small Electnc Motor Sublected To Accelerated Aging. LAUHALA.K.E. NUREG/CR-5052: OPERATING EXPERIENCE AND AGING ASSESS-NUREG/CR 5198: INHALED (239)PUO(2) AND/OR TOTAL BODY MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR-GAMMA MADIATION.Early Mortahty And Morbedity 11, Rats And Dogs. IZED WATER REACTORS. LAY,T. LOFGREN E.V. NUREG/CR-3145 V06: GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-5078 V01: A RELIABILITY PROGRAM FOR EMERGENCY WESTERN OHIO INDIANA REGION. Annual Report, October 1986 - September 1987. DIESEL GENERATORS AT IKUCLEAR POWER PLANTS. Program Structure.

l 74 Personal Author Index NUREG/CR 5078 V02: A REUABILITY PROGRAM FOR EMERGENCY - MARSCHALL.C.W. DIESEL GENERATORS AT NUCLEAR POWER NUREG/CR-4082 V06: DEGPADED PIPING PROGRAM - PHASE li.Sath . PLANTS Maintenance. Surveillance And Condition Monitoring. Program Report. October 1986 September 1987. LOl5L.J.M. MARTINEZ,V. NUREG/CR 5090: EFFECTS OF TEMPERATURE AND HUMIDITY ON NUREG/CR-5071: TRAC SUPPORT SOFTWARE.- l RESPIRATOR FIT UNDER SIMULATED WORK CONDITIONS. !. MARK,K.D. l LONG.L.T. NUREG/CR-5120: A MODEL FOR THE TRANSPORT AND CHEMICAL l NUREG/CR 5258 V01: GEORGIA / ALABAMA REGIONAL SEISMO- REACTION OF MOLTEN DEBRIS IN DIRECT CONTAINMENT HEAT-GRAPHIC NETWORK. Annual R6 port, August 1985 June 1986. ING EXPERIMENTS. LOOMIS.G. MATHIESON,T. I NUREG/CR 5072: DECAY HEAT REMOVAL USING FEED.AND-BLEED NUREG/CR-5134: APPLICATION OF ACOUSTIC LEAK DETECTION FOR U.S. PRESSURIZED WATER REACTORS. TECHNOLOGY FOR THE DETECTION AND LOCATION OF LEAKS IN l pj LIGHT WATER REACTORS. j NUREG/CP-0064: SECOND CNSI WORKSHOP ON DUCTILE FRAC- MAUDERLY,J.L TURE TEST METHODS. NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED ALPHA AND PETA IRRADIATION OF THE LUNG Phase il Report. LOYSEN P. NUREG 1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS MCCABE,D.E. l HANDBOOK. NUREG/CR-5207: FRACTURE EVALUATION OF SURFACE CRACKS i LUBENAU,J.O. EMBEDDED IN REACTOR VESSEL CLADDING Material Property Eval-NUREG-1311: FUNDING THE NRC TRAINING PROGRAM FOR uations. STATES. MCCLUNG,R.W. LUCERO.D A, NUREG/CR-4860 R01: FLAW DENSITY EXAMINATIONS OF A CLAD NUREG/CR 4917: DCH-2:RESULTS FROM THE SECOND EXPERIMENT BOILING WATER REACTOR PRESSURE VESSEL SEGMENTv PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. LUCKAS,W.J. NUREG/CR-5137: BIODEGRADATION TESTING OF TMI-2 EPICOR-il 1 NUREG/CR 4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE- WASTE FORMS. -l VENTION AND MITIGATION FEATURES:BWR MARK I CONTAIN- NUREG/CP,5220: ANNUAL REPORT OF THE TMi-2 EPICOR-ll RESIN /  ! MENT DESIGN. LINER INVESTIGATION. Low-Level Waste Data Base Development NUREG/CR-4920 V02: ASSESSMENT OF SEVERE ACCILINT PRE- Program For Fiscal Year 1988. VENTION AND MITIGATION FEATURES:BWR, MARK ll CONTAIN-MENT DESIGN. - MCGEE,0.R. NUREG/CR-4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG/CR-5198: INHALED (239)PUO(2) AND/OR TOTAL BODY VENTION AND V"".iATION FEATURES:BWR, MARK 11! CONTAIN- GAMMA RADIATION.Early Mortality And Morbidity in Rats And Dogs. MENT DESIGN. NUREG/CR-4920 Vu ASSESSMENT OF SEVERE ACCIDENT PRE- MCGOWAN,J.J. VENTION AND MniGATION FEATURES'PWR.LARGE DRY CON-NUREG/CR 4880 V01: CHARACTERIZATION OF IRRADIATED CUR-TAINMENT DESIGN. RENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUPEG/CR-4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE- . NUCLEAR PRESSURE VESSEL SERVICE. j VENTION AND MITIGATION FEATURES:PWR,1CE CONDENSER NUREG/CR-4880 V02: CHARACTERIZATION OF 1RRADIATED CUR-CONTAINMENT DESIGN. RENT-PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUCLEAR PRESSURE VESSEL SERVICE. l NUREG/CR 5212: EMERGENCY ENVIRONMENTAL SAMPLING AND MCGUIRE,S.A. ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES- NUREG 1140. A' REGULATORY ANALYSIS ON EMERGENCY PRE- , PAREDNESS FOR FUEL CYCLE AND OTHER RADIOACTIVE MATE- I NU EG CR 4 VALUE-lMPACT ANALYSIS FOR EXTENSION OF NRC BULLETIN 85-03 TO COVER ALL SAFETY-RELATED MOVS. MCKEE.C.R-MADNi,1 K NUREG/CR.5240: COMPARATIVE EVALUATION OF SELECTED CON. TINUUM AND DISCRETE-FRACTURE MODELS. Emphasis On Disper-NUREG/CR-5164: A GIMPLIFIED MODEL FOR CALCULATING EARLY sivity Calculations For Application To Fractured Geologic Media, Cres-OFFSITE CONSEQUENCES FROM NUCLEAR REACTOR ACCl- ton Study Area. Eastern Washington. DENTS ~ NUREG/CR-5277: THE TENSORIAL NATURE OF EFFECTIVE POROSI-MAERKER,R.E. TY AND LARGE SCALE DISPERSION COEFFICIENTS.Apphcation To I, NUREG/CR-5049. PRESSURE VESSEL FLUENCE ANALYSIS AND The Creston Study Area. Eastern Washington. ] NEUTRON DOSIMETRY, MCKEE,P. MALVA,P.S. NUREG-0797 S20: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR 4667 V04: ENVIRONMENTALLY ASSISTED CRACKING IN OPERATION OF COMANCHE PEAK STEAM ELECTRIC LIGHT WATER REACTORS. Semiannual Report, October 1986 March STATION, UNITS 1 AND 2. Docket Nos. 50 445 And 50 446.(Texas Utili-1987. ties Generating Company) MALLOY,M. MCKENNA,T.J. NUREG 0797 S19: SAFETY EVALUATION REPORT RELATED TO THE NUREG-1228: SOURCE TERM ESTIMATION DURING INCIDENT RE-OPERATION OF CCMANCHE PEAK STEAM ELECTRIC SPONSE TO SEVERE NUCLEAR POWER PLANT ACCIDENTS. STATION, UNITS 1 AND 2. Docket Nos. 50 445 And 50 446.(Texas Utili-tes Generating Company) MCLAUGHLIN.D. , NUREG/CR-5094: APPLICATION OF STOCHASTIC METHODS TO THE i MALY,J.A- SIMULATION OF LARGE SCALC UNSATURATED FLOW AND

                                                                                                                                                            ~

NUREGICR-4920 V04: ASSESSMENT OF SEVERE ACCIDONT PRE- TRANSPORT. VENTION AND MITIGATION FEATURES:PWR LARGE DRY CON-TAINMENT DESIGN. MCLAUGHLIN,M. < I NUREG/CR 4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE- NUREGiCR-5138: VAllDATION OF GENERIC COST ESTIMATES FOR VENTION AND MIT!GATION FEATURES:PWA.lCE4CNDENSER CONSTRUCTION-RELATED ACTIVITIES AT NUCLEAR POWER CONTAINMENT DESIGN. PLANTS. Final Report. 3 1

                                                                                                                                                     -a

Personal Author index . 75 MCLEAN,W.J. MOKHTARIAN,K. NUREG/CR-5086: PLATINUM CATALYTIC IGNITERS FOR LEAN HY- NUREG/CR-4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS DROGEN-AIR MIXTURES. METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND BRANCH CONNECTIONS. MCNAMARA.N. NUREG.0837 V07 N04: NRC TLD DIRECT RADIATION MONITORING MOLTYANER,G.L NETWORK. Progress Report, October-December 1987 NUREG/CR 4879 Voi: DEMONSTRATION OF PERFORMANCE MOD-NUREG.0837 V08 N01: NRC TLD DIREC7 RADIATION MONITORING ELING OF A LOW-LEVEL WASTE SHALLOW-LAND BURIAL SITE.A NETWORK. Progress Report. January-March 1988. Companson Of Predective Radionuclides Transport Modeling Versus NUREG-0037 V08 NO2: NRC TLD DIRECT RADtATION MONITORING Field Observations At The Nitrate Disposal Pit Site, Chalk River Nuclear NETWRK. Progress Report. April June 1988. Labs. MEALE.B.M. MOON,D. NUREG/CR4747 V02: AN AGING FAILURE SURVEY OF LIGHT NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER WATER REACTOR SAFETY SYSTEMS AND COMPONENTS- INDUSTRY.A Review Of Technicalissues. MELDER.B.D. MOORE C. NUREG/CR4991: EVALUATION AND PROPOSED llWPROVEMENTS TO NUREG/CR 5227: FITNESS FOR DUTY IN THE NUCLEAR POWER EFFECTIVENESS OF U.S. NUCLEAR REGULATORY COMMISSION INDUSTRY.A Review Of Technical lasues. GENERIC COMMUNICATIONS. MOORE,5.E. MERKLE J.G. NUREG/CR4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS NUREG/CR4888: PRESSURIZEDTHERMAL-SHOCK TEST OF 6-INCH METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND THICK PRESSURE VESSELS.PTSE 2: Investigation Of Low Teanng Re-BRANCH CONNECTIONS. 2:ctance And Warm Prestressmg. MORRIS C.J. NURNG/bR4740: NUCLEAR PLANT-AGING RESEARCH ON REACTOR PR RA h se 11 F epo PROTECTION SYSTEMS. MOSLEH,A. NURNG 5R4708 V02: PROGRESS IN EVALUATION OF RADIONU-CLlDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-USE FA RS SA ETY AND RE I I - LEVEL NUCLEtR WASTE REPOSITORY SITE PROJECTS. Report For S ESAWural Ranmod And Exarnpes. April 1988 September 1987. gggggggggg,g,g, NUREG/CR-S108: THERMODYNAMIC PROPERTIES OF TC(IV) NUREG/CR-5204: LOW-LEVEL RADIOACTIVE WASTE SOURCE TERM OXIDES. Solubilities And The Electrode Potential Of The Tc(Vil)/Tc(IV)' MODEL DEVELOPMENT AND TESTING. Oxide Couple. MULLEN,M.F. NURN dR-5153: THE TEACHABILITY AND MECHANICAL INTEGRITYNUREG/CR-5218: FINANCIAL QUALIFICATIONS REVIEW OF APPLl-OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED CANTS FOR NUCLEAR POWER PLAf.T INSTRUCTION PERMITS. IN CEMENT AND VINYL ESTER-STYRENE- . NUREG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PERFORM-ANCE INDICATORS. MILLER,C.S. MURPHY,E. NUREG/CR-4312 V02 R1: RELAPS/ MOD 2 CODE MANUAL. Volume 2: NUREG-0844: NRC INTEGRATED PROGRAM FOR THE RESOLUTION NUR / 194 E PS/M DELS AND CORRELATIONS. OF UNRESOLVED SAFETY ISSUES A 3.A 4 AND A 5 REGARDING STEAM GENERATOR TUBE INTEGRITY. Final Report. MILLER,LF. E A* NUREG/CR-5019: NEUTRON EXPOSURE PARAMETERS FOR THE METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SEC- NURE'G/CR4879 VD1: DEMONSTRATION OF PERFORMANCE MOD-TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES. ELING OF A LOW-LEVEL WASTE SHALLOW-LAND BURIAL SITE.A Companson Of Predictive Radionuclides Transport Modeling Versus MILLER,R.D. Field Observations At The Nitrate Disposal Pit Site Chalk River Nuclear NUREG/CR-5045: KANSAS NEBRASKA SEISMICITY STUDIES USING Labs. THE KANSAS NEDRASKA MICROEARTHOUAKE NETWORK.Fmal NAKAGAKI,M.

                                   *P NUREG/CR4082 V06: DEGRADED PIPING PROGRAM - PHASE ILSixth MILSTEAD,W.                                                                                   Program Report. October 1986. September 1987.

NUREG 0933 S07: A PRIORITIZATION OF GENERIC SAFETY ISSUES. NUREG-0933 S08: A PRIORITIZATION OF GENERIC SAFETY ISSUES. NANSTAD,R K MINARICK.J.W. RENT-PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUREG/CR 4L/4 V05: PRECURSORS TO POTENTIAL SEVERE CORE NUCLEAR PRESSURE VESSEL SERVICE. DAMAGE ACCIDENTS 1986,A STATUS REPORT. NUREG/CR4880 V02: CHARACTERIZATION OF IRRADIATED CUR-NUREGICR4674 V08: PRECURSORS TO POTENTIAL SEVERE CORE RENT. PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR .. DAMAGE ACCIDENTS:19B6 A STATUS REPORT. NUCLEAR PRESSURE VESSEL SERVICE. NUREG/CR4888: PRESSURIZED-THERMAL-SHOCK TEST OF 6-INCH MISHIMA.J. THICK PRESSURE VESSELS.PTSE 2: Investigation Of Low Tearing Re. NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS sistance And Wann Prestressmg. HANDDOOK. l NEITZEL.D.A. I MOCHIO.T. NUREG/CR-5210: TECHNICAL FINDINGS DOCUMENT FOR GENERIC ' NUREG/CR 3509: POWER SPECTRAL DENS'TY FUNCTIONS COM- ISSUE 51: IMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE. PATIBLE WITH NRC REGULATORY G iDE 1.60 RESPONSE SPEC- WATER SYSTEMS. TRA. NELSON.J.D. MOELLIR M.P. NUREG/CR4651 V02: DEVELOPMENT OF RIPRAP DESIGN CROERIA NUREG/CR.4873: BENCHMARK STUDY OF THE l-DYNEV EVACU- BY RIPRAP TESTING IN FLUMES. Phase ll. Followup investigations. ATION TIME ESTIMATE COMPUTER CODE. NUREG/CR-4874: THE SENSITIVITY OF EVACUATION TIME ESTI. NEMERGUT,J. MATES TO CHANGES IN INPUT PARAMETERS FOR THE l DYNEV NUREG/CR-4555 ROI: GENERIC COST ESTIMATES FOR THE DIS-COMPUTER CODE. POSAL OF RADIOACTIVE WASTES.

76 Personal Author index NESTOR,C.W. OLIVER.M.S. NUREG/CR-5183: A USER'S MANUAL FOR THE CONTAMINANT NUREG/CR-4917: DCH-2 RESULTS FROM THE SECOND EXPERIMENT TRANSPORT MODULE OF THE MIGRAT CODE. PERFORMED IN THE SURTSEY [dRECT HEATING TEST FACILITY, NETI,S. OLSON,J. NUREG/CR-5005 V01: THERMODYNAMIC NONEOUILIBRIUM IN POST- NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER CRfTICAL HEAT FLUX BOILING IN A ROD BUNDLE.Desenption Of INDUSTRY.A Review Or Technical Issues. NUREG/CR 52 DEVELOPMENT OF PROGRAMMATIC PERFORM-N /CN O V .TE I YNAMIC NONEQUILtDRIUM IN POST- ' CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Stabi-hred Quench Front Tests OLSON R NUREG/CR-5095 V03. THERMODYNAMIC NONEOUILIBRIUM IN POST. NUREG/CR 4082 V06: DEGRADED PIPING PROGRAM - PHASE 11 Sixth CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Advanc-Program Report. October 't986 - September 1987. ing Quench Front Tests. NUREG/CR-5095 V04. THERMODYNAMIC NONEOUILIBRIUM IN POST-CRI ICA HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Retreat-NUR G/CR-4735 V04: EVALUATION AND COMPILATION OF DOE WASTE P/.CKAGE TEST DATA Biannual Report. August 1987 - Janu-NEWTON,G.J. ary 1988. NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED ALPHA AND BETA IRRADIATION OF THE LUNG. Phase 11 Report. OSTRACH,S. NUREG/CR-St13. FINDINGS OF THE PEER REVIEW PANEL ON THE NGUYEN.B. DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG 1150. NUREG/CR S165: SEISMOLOGICAL INVESTIGATION OF EARTH-OUAKES IN THE NEW MADRID SEISMIC ZONE AND THE NORTH- OWCZARSKl,P.C. EASTERN EXTENT OF THE NEW MADRID SEISMIC ZONE. Final NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS Report. September 1981 - December 1986. HANDBOOK. NUREG/CR 4997: METHODS FOR DESCRIBING AIRBORNE FRAC. RE - 320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS HANDBOOK- PAPASPYROPOULOS NICHOLS,R.T. NUREG/CR 4082 V06: D'c' GRADED PIPING PROGRAM PHASE ll Sixth Program Report. October 1986 - September 1987 NUREG/CR 4917: DCH-2.RESULTS FROM THE SECOND EXPERIMENT PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. PARK,C. NICOLETTE V.F. NUREG/CR-4551 V5 DRF: EVALUATION OF SEVERE ACCJDENT NUREG/CR.4763: SAFETY RELATED EQUlPMENT SURVIVAL IN HY. RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS PL/.NT. Draft Report For Comment. NICOLOSI,s.L PARK J.F. NUREG/CR-4857: CADET.A DECISION SUPPORT SYSTEM FOR LIGHT NUREG/CR-5198: INHALED (239)PUO(2) AND/OR TOTAL BODY WATER REACTOR SAFETY. GAMMA RADIATION Early Mortality And Morbidity in Rats And Dogs. NIEVES,LA. PARK,J.Y. NUREG/CR 4811: THE ECONOMIC CCdTS OF RADIATION-INDUCED NUREG/CR-4667 V04. ENVIRONMENTALLY ASSISTED CRACKING IN HEALTH EFFECTS. Estimation And Simulation- LIGHT WATER REACTORS. Semiannual Report. October 1986 - March NORDEN,K. 1987. NUREG/CR-2007 V06: RADIOACTIVE MATERIALS RELEASED FROM PARKS C.V NURI /CR 2 7 A O lY AT IALS RELEASED FROM NUREG/bR 6033:

SUMMARY

DESCRIPTION OF THE SCALE MODU- l NUCLEAR POWER PLANTS Annual Report For 1986. LAR CODE SYSTEM. ' NORKIN,0.P. PARRY,G. NUREG.0797 S17: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR 4780 V01: PROCEDURES FOR TREATING COMMON OPERATION OF COMANCHE PEAK STEAM ELECTRIC CAUSE FAILURES IN SAFETY AND RELIABILITY STATION. UNITS 1 AND 2 Docket Nos 50-445 And 50 446.(Texas Utih- STUDIES Procedural Framework And Examples. ties Generating Company) NOURBAKHSH,H.P. NUREG/CR 4836: APPROACHES TO UNCERT AINTY ANALYSIS IN NUREG/CR-4BB1: FIS$10N PRODUCT RELEASE CHARACTERISTICS PROBABILISTIC RISK ASSESSMENT. INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCI-DENT CONDITIONS. PASCHIS,J.A. NUREG/CR 5107: HYDROGEOLOGIC CHARACTERIZATION OF NOWLEN,S.P* BASALTS.The Northern Rim Of The Columbia Plateau Physiographic NUCIEG/CR 4527 V02: AN EXPERIMENTAL INVESTIGATION OF IN- Province And Of The Creston Study Area, Eastern Washington. TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CONTROL CABINETS Part il Room Effects Tests- PAULA,H. O'DONNELL,E S A CMM NUREG/CR 4918 V02. CONTROL OF WATER INFILTRATION INTO CAUSE FAILURES IN SAFETY AND RELIABILITY NEAR SURFACE LLW DISPOSAL UNITS Task Report - A Discussion. STUDIES. Procedural Framework And Examples. O'KELLEY,G.O. PAULA,H.M. NUREG/CR 4708 V02: PROGRESG IN EVALUATION Or RADIONU. NUREG/CR 5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU-CLIDE GEOCHEMICAL INFORMATION DEVELOPED DY DOE HIGH. CLEAR ONE - UNIT 1. Volume 1. Program For inspectors. LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For NUREG/CR-5021 V02: UdIR'S GUIDE FOR PRISIM APKANSAS NU-Antil 1986 - September 1987 CLEAR ONE - UNIT 1. Volume 2f. gram For Regulators. NUI'tEG/CR 5108. THERMODYNAMIC PROPERTIES OF TC(IV) PEPPER,S.E. OXIDES Solubilities And The Electode Potential 9f The Tc(Vil)/Tc(IV). Oxioe Couple. NUREG/CR-4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER PLANT COMPONENTS (PHASE II). Motor Control OESTERLE.R.G. Center, Switchboard.Panelboerd And Power Supply. NUREG/CR-5209 DESIGN PROVISIONS FOR TANGENTIAL SHEAR IN NUREG/CR-5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABI-CONTAINMENT WALLS NETS.

i. Personal Author index 77' 'I PERKINS,K. . POMERENING.D.J. NUREG/CR-4999: ESTIMATION OF RISK REDUCTION FROM IM- NUREG/CR-5012: SIMILARITY PRINCIPLES FOR EQUIPMENT OUAll. PROVED PORV RELIABILITY IN PWRS.Fenal Report. FICATION BY EXPERIENCE. j PERKINS,K.R. POWERS,0.A. ' NUREG/CR-4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE-NUREG/CR-5196: SUBMISSION FOR THE CSNt/GREST BENCHMARK VENTION AND MITIGATION FEATURES.BWR. MARK 1 CONTAIN-EXERCISE ON CHEMICAL THERMODYNAMIC MODELING IN CORE-NUREG/ 4 0 V02: ASSESSMENT OF SEVERE ACCIDENT PRE-

                                                                                                             ^                      '

VENTION AND MITIGATION FEATURES:BWR. MARK ll CONTAIN- PRASSINOS,P.G. MENT DESIGN- J NUREG/CR-4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG/CR-5042 S01: EVALUATION OF EXTERNAL HAZARDS TO NU. 1 VENTION AND MITIGATION FEATURES BWR, MARK lli CONTAIN- CLEAR POWER PLANTS IN THE UNITED STATES. Seismic Hazard. MENT DESIGN- PRATT,W.T. NUREG/CR 4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE" ) VENTION AND MITIGATION FEATURES;PWR.LARGE DRY CON- NUREG/CR 4920 V01: ASSESSMENT'OF SEVERE ACCIDENT PRE- 1 VENTION AND M:TIGATION FEATURES BWR. MARK I CONTAIN. f NI 4 20 5: ASSESSMENT OF SEVERE ACCIDENT PRE- MENT DESIGN. VENTION AND MITIGATION FEATURES:PWR' ICE-CONDENSER NUREG/CR 4P20 V02: ASSESSMENT OF SEVERE ACCIDENT PRE-CONTAINMENT DESIGN. VENTION AND MITIGATION FEATURES:BWR, MARK 11 CONTAIN-MENT DESIGN. - , PERSINKO,D. NUREG/CR 4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE- l NUREG-1214 R03; HISTORICA'. DATA

SUMMARY

OF THE SYSTEMAT, VENTION AND MITGATION FEATURES.BWR. MARK lli CONTAIN-IC ASSESSMENT OF LICENSEE PERFORMANCE. MENT GN R gg PESSANHA.J. VENTION AND MITIGATION FEATURES:PWR,LARGE DRY CON- 1 NUREG/CR 5157: THE DEVELOPMENT OF APRIL. MOD 2 - A COMPUT. TAINMENT DESIGN. 'k ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL- NUREG/CR 4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE-

        ' ING WATER NUCLEAR REACTORS.                                       VENTION AND MITIGATION FEATURES.PWR,lCE CONDENSER                     I CONTAINMENT DESIGN.

PETTIJOHN R.R. NUREG/CR-5132: SEVERE ACCIDENT INSIGHTS REPORT. NUREG/CR-5223: SCINTILLATION FIBER DETECTOR FOR IN VIVO ENDOSCOPIC INTERNAL DOSIMETRY. PRINE,D. NUREG/CR 5134: APPLICATION OF ACOUSTIC LEAK DETECTION - UR CR 784 INFLUENCE OF GROUNDWATER ON SOIL.STRUC-TURE INTERACTION. GHT C R PHILLIPS.S.L PRZlKWAS,A.J. NUREG'/CR-4864 VO1: THERMODYNAMIC TABLES FOR NUCLEAR NUREG/CR 5233: A COMPUTER CODE FOR FIRE PROTECTION AND .. WASTE ISOLATION. Aqueous Solutions Database. RISK ANALYSIS OF NUCLEAR PLANTS. PICluLO.P.L PUTNAM,C.H. NUREG/CR 5153: THE TEACHABILITY AND MECHANICAL INTEGRITY NUREG/CR-5151: PERFORMANCE BASED INSPECTIONS. OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED IN CEMENT AND VINYL ESTER-STYRENE. M i N RE'G/dR-5198; INHALED (239)PUO(2) AND/OR TOTAL-BODY PILCH M. GAMMA RADIATION.Early Mortality And Morbidity in Rats And Dogs. ] NUREG/CR 4914: THE INFLUENCE OF SELECTED CONTAINMENT STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW. RAMSDELL.J.V. ING HIGH PRESSURE MELT EJECTION FROM THE REACTOR NUREG/CR.4000 V02: THE MESORAD DOSE ' ASSESSMENT VESSEL. MODEL. Computer Code. PIN,F.G. NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW CRnERIA FOR ALTERNATIVE METHODS OF LOW LEVEL RADIOAC-NUREG/CR-5183: A USER'S MANUAL FOR THE CONTAMINANT TIVE WASTE DISPOSAL Environmental Monitonng And Surveillance TRANSPORT MODULE OF THE MiGRAT CODE. Programs. NUREG/CR-5055: ATMOSPHERIC DIFFUSION FOR CONTROL ROOM UR G 0933 S07: A PRIORITIZATION OF GENERIC SAFETY ISSUES. NUREG 0933 S08 A PRIORITIZATION OF GENERIC SAFETY ISSUES. RANSOM,V.H.  ! PLANTE.E. NUREG/CR 4312 V01: RELAP5/ MOD 2 CODE MANUALVolume 1: Code NUREG/CR-4735 V04. EVALUATION AND COMPILATION OF DOE Structure. Systems Models And Solution Methods. WASTE PACKAGE TEST DATA. Biannual Report. August 1987 - Janu- NUREG/CR 4312 V02 R1: RELAPS/ MOD 2 CODE MANUALVolume 2: ary 1988. Users Guide And input Requirements. d PODOLAK,E.M. RASMUSON,0. NUREG 0654 S01 RO1: CRITERIA FOR PREPARATION AND EVALUA. NUREG/CR-4780 V01: PROCEDURES FOR TREATING COMMON TION OF RADIOLOGICAL EMERGENCY RESPONSE PLANS AND CAUSE FAILURES iN SArETY AND RELIABILITY ) PREPAREDNESS IN SUPPORT OF NUCLEAR POWER STUDIES. Procedural Framework And Examples. ' PLANTSfntem For Utihty Offsite Planrnng And Preparedness. RASMUSSEN,N. PODOWSKl.M.Z. NUREb/CR-5000. METHODOLOGY FOR UNCERTAINTY ESTIMATION NUREG/CR-5157: THE DEVELOPMENT OF APRIL. MOD 2 A COMPUT. IN NUREG 1150 (ORAFT). Conclusions Of A Review Panel. ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL. ING WATER NUCLEAR RE ACTORS. NA PO!.MANN D.J. NUREG/CR 97: SIMULATION OF LIOUlO AND VAPOR MOVEMENT IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP < IdUREG/CR-5 % APPLICATION OF STOCHASTIC METHODS TO THE TUFF SITE.Models And Strategies. ' SIMULATION OF LARGE-SCALE UNSATURATED FLOW AND TRANSPORT. REECE W.J. POLZER.W.L - NUREG/CR-4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data fdVREG/CR-5170. A REVIEW OF RESEARCH CONDUCTED BY LOS Processing And Revision Part 1: Technical Overview. ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHAS:S ON THE MAXEY FLATS.KY, SHALLOW LAND WASTE BURIAL S'TE. NUREG/CR 4639 V03 P2: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data

78 Personal Author Index Processing And Revision.Part 2: Human Error Probability Data Entry ROBINSON,G.C. And Revision Procedures. NUREG/CR-4888: PRESSUR'?ED-THERMAL-SHOCK TEST OF 6-lNCH NUREG/CR4639 V03 P3: NUCLEAR COMPUTERIZED LIBRARY FOR THICK PRESSURE VESSELS.PTSE.2: Investigation Of Low Teanng Re-ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data sistance And Warm Prestressing. Processing And Revision. Part 3: Hardware Component Failure Data Entry And Rev'sion Procedmes. ROBINSON,T, NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE SUB-COMPONENT AND SUBSYSTEM LEVELFinal Report. NUREG CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE DRAFT REACTOR RISK REFERENCE DOCUMENT.NUREG 1150- RODABAUGH,E.C. NUREG/CR-4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS M REICH'Ed/CR-5105: NUR RESPONSE MARGINS INVESTIGATIONMETHODS OF PIPING BRANCH COfulECTIONS. FOR CALCul ATING FLEXIBILITY OF NOZZLES AND DYNAMIC ANALYSES USING THE INDEPENDENT SUPPORT MOTION METHOD AND PVRC DAMPlNG. R W RS M RDD,L.D. NUREG/CR 5137: BIODEGRADATION TESTING OF TMI-2 EPICOR-il NUREG/CR-5075: THE SAFT UT REAL-TIME INSFECTION SYSTEM - WASTE FORMS. OPERATIONAL PRINCIPLES AND IMPLEMENTATtW NUREG/CR-5229: ANNUAL REPORT OF THE TMI 2 EPICOR il RESIN / LINER INVESTIGATION. Low Level Waste Data Base Development REISTER,0.9. Program For Fiscal Year 1988. NUREG/CR.5183: A USER'S MANUAL FOR THE CONTAMINANT TRANS" PORT MODULE OF THE MiGRAT CODE. ROSENFELD,M. NUREG/CR 4082 V06: DEGRADED PIPING PROGRAM - PHASE II. Sixth Program Report. October 1986 September 1987. N RE b PERSONNEL AND VEHICLE BARRIERS AT FUEL FABRI-CATION FACILITIES US.NG OR POSSESSING STRATEGIC QUANTI- poss,J,w, TIES OF SPECIAL NUCLEAR MATERIAL NUREG/CR-4917: DCH-2:RESULTS FROM THE SECOND EXPERIMENT PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. REVANKAR.S.T. NUREG/CR 5133: A COMPUTATIONAL MODEL COR CRITICAL FLOW ROTHBERG,0 THROUGH INTERGRANULAR STRESS CORROSION CRACKS. NUREG-1296: THERMAL OVERLOAD PROTECTION FOR ELECTRIC RICHARDS.E.H. MOTORS ON SAFETY RELATED MOTOR-OPERATED VALVES GE-NUREG/CR-4728. EQUIPMENT QUALIFICATION RESEARCH TEST OF NERIC ISSUE II.E.6.1. A HIGH. RANGE RADIATION MONITOR. RICKARD,W.H. NUREG/CR-52" V01: DIAGNOSIS OF CONDENSATION lNDUCED NUREG/CR 5047. RADIONUCLIDES ACCUMULATION BY AOUATIC WATERHAMh? Methods And Background. BIOTA EXPOSED TO CONTAMINATED WATER IN ARTIFICIAL ECO. NUREG/CR-5220 h,7 DIAGNOSIS OF CONDENSATION-INDUCED SYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH THE WATERHAMMER.Cas Studies. GROUND. ROTHLEDER,B. RICKER,R. NUREG/CR-5078 V02: A n ? LIABILITY PROGRAM FOR EMERGENCY NUREG/CH-4735 V03: EVALUATION AND COMPILATION OF DOE D!ESEL GENERATO S AT NUCLEAR POWER WASTE PACKAGE TEST DATA. Biannual Re : February-July 1987. PLANTS.Memtenance.Sunn ilance And Condition Monitonng. NUREG/CR.4735 V04: EVALUATION AND PILATION OF DOE WASTE PACKAGE TEST DATA. Biannual Report. August 1987 - Janu- ROUS$1N,R.W. ary 1988- NUREG/CR-5264: GUIDE FOR LICENSING EVALUATIONS USING CRAC2.A Computer Program For Calculatmg Reactor Accident Conse-RIDKY,R.W. quences. NUREG/CR.4918 V02: CONTROL OF WATER INFILTRATION INTO NEAR SURFACE LLW D!SPOSAL UNITSTask Report - A Discussion. RUBIN,A.M. RIEMKE,R.A. NUREG-1109: REGULAmRY/BACKFIT ANALYSIS FOR THE RESOLU-NUREG/CR-4312 y01: RELAP5/ MOD 2 CODE MANUALVotume 1: Code TION OF UNRESO' .,,,I SAFETY ISSUE A 44, STATION BLACKOUT. Structure. Systems Modsis And Solution Methods. RUFF,J.F. NUREG/CR-4312 V02 Rt: RELAP5/ MOD 2 CODE MANUAL. Volume 2: NUREG/CR-4651 Voi y EVELOPMENT OF RIPRAP DESIGN CRITERIA Users Guide And input Requirements. BY RIPRAP TESTING H FLUMES. Phase ll. Followup investigations. RIGGS,R. NUREG 0933 S07: A PRIORITI2ATION OF CENERIC SAFETY ISSUES. RUGER.C.J. NUREG-0933 S08. A PRIORITIZATION OF GENERIC SAFETY ISSUES. NUREG/CR 5140: VALUE lMPACT ANALYSIS FOR EXTENSION OF RIORDAN,B. NUREG/CR-5138: VALIDATION OF GENERIC COST ESTIMATES FOR RUSPi,J. CONSTRUCTION RELATED ACTIVITIES AT NUCLEAR POWER NUREG/CR-4735 V03: EVALUATION ant' COMPILATION OF DOE PL'.NTS Final Report WASTE PACKAGE TEST DATA. Biannual Report: February-July 1987. 4 NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE NUREG/CR 4735 V04: FVALUATION AND COMPILATION OF DOE i SUB-COMPONENT AND SUBSYSTEM LEVELFinal Report. WASTE PACKAGE TEST DATA. Biannual Report. August 1987 - Janu-RITZMAN,R.L. ary 1988. NUREG/CR-5113. FINDINGS OF THE PEER REVIEW PANEL ON THE R DRAFT REACTOR RISK REFERENCE DOCLWENT,NUREG-1150. UREG CR 4667 V04: ENVIRONMENTALLY ASSISTED CRACKING IN ROBERTF G.C. LIGHT W ATER REACTORS. Semiannual Report,0ctober 1986 - March i NUREGICR 4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE 1987. l SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER SAARI,L. PLANTS Operating Expenence And Failure Identification. NUREG/CR 5227: FITNESS FOR DUTY IN THF. NUCLEAR POWER l ROBERTSON,D.E. INDUSTRY.A Review Of Technicat issues.  ! I NUREG/CR-4879 V01: DEMONSTRATION OF PERFORMANCE MOD-ELING OF A LOW-LEVEL WASTE SHALLOW-LAND BURIAL SITE.A SAMANTA,P.K.  ! Compenson Of Predictive Radionuchde Transport Modeling Versus NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT I Field Observations At The Nitrate Disposal Pit Site, Chalk River Nuclear AND STI REQUIREMENTS AT THE ANO 1 NUCLEAR POWER l Labs. PLANT, ' _ a

Personal Author Index 79 SAMARAS.E.F. SCOTT,W. NUREG/CR 3509. POWER SPECTRAL DENSITY FUNCTIONS COM- NUREG/CP 009D: PROCEEDINGS OF THE PUBLIC WORKSHOP FOR PATIBLE WITH NRC REGULATORY GUIDE 1.60 RESPONSE SPEC + NRC RULEMAKING ON MAINTENANCE OF NUCLEAR POWER TRA. PLANTS. NUREG/CR-5227. FITNESS FOR DUTY IN THE NUCLEAR POWER SANDERS.M.E. INDUSTRY.A Rewew Of Techrwcal issues. NUREG-0654 Sci R01: CRITERIA FOR PREPARATION AND EVALUA-TlON OF RADIOLOGICAL EMERGENCY RESPONSE PLANS AND SERKlZ,A.W. PREPAREDNESS IN SUPPORT OF NUCLEAR POWER NUREG-1273: TECHNICAL FINDINr3S AND REGULATORY ANALYSIS PLANTS.Critena For Utility Offsite Plannin0 And Preparedness. FOR GENERIC SAFETY ISSUE li.EA3, " CONTAINMENT INTEGRITY SA1TERWHITE,D. NUREG/CR-4747 V02: AN AGING FAILbHE SURVEY OF LIGHT SE VERUD.LK. WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. NUREG/CR-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE SCHERPELZ,R.I. PREDICTION METHODS FOR PIPING. NUREG/CR-4000 V02: THE MESDRAD DOSE ASSESSMENT SHA,W.T. MODEL. Computer Code. NUREG/" "i65 TIME AND VOLUME AVERAGED CONSERVATION SCHLONSKI,J.S. EQU/ 00A FOR MULTIPHASE FLOW USING MASS-WEIGHTED VE-NUREG/CR-4597 V02: AG'NG AND SERVICE WEAR OF AUXILIARY LOCITY AND INTERNAL ENERGY. i FEEDWATER PUMPS FOR PWR NUCLEAR PLANTS Volume 2. Aging NUREG/CR-5070: ANALYSIS OF NATURAL CONVECTION PHENOM- , ! Assessments And Monitonng Method Evaluations. ENA IN A 3 LOOP PWR DURING A TMLB' TRANSIENT USING THE COMMIX CODE. SCHROCK,V.E. NUREG/CR-5133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW SHACK,W.J. THROUGH INTERGRANULAR STRESS CORROSION CRACKS. NUREG/CR 4667 V04: ENVIRONMENTALLY ASSISTED CRACKING iN LIGHT WATER REACTORS. Semiannual Report October 1986. March . SCHULZ,R.K. 1987. l NUREG/CR-4918 V02: CONTROL OF WATER INFILTRATION lPTO NUREG/CR 5156: REVIEW OF EROSION-CORROSION IN SINGLE. i NEAR SURFACE LLW DISPOSAL UNITS. Task Report - A Discus 4 ton. PHASE FLOWS. SCHW ARTZ l. SHAUKAT,S.K. NUREG 0020 V11 N12: LICENSED OPERATING REACTORS STATUS NUREG-1233 DRFT FC: REGULATORY ANALYSIS FOR USl A 40,

SUMMARY

REPORT. Data As Of November 30,1987.(Gray Book 1) " SEISMIC DESIGN CRITERIA." Draft Report For Comment. NUREG 0020 V12 N01: LICENSED OPERATING REACTORS STATUS GUMMARY REPORT. Data As Of December 31,1987.(Gray Book 1) SHEETS.M.W. NUREG-0020 V12 NO2: LICENSED OPERATING REACTORS STATUS NUREG/CR-4939 V01: IMPROVING MOTOR RELIABILITY IN NUCLEAR N EG 2 V NO C NS E NG R A TATUS POWER PLANTS. Volume 1. Performance Evaluation And Maintenance NU EG 2 V N CENS D 0 ENA I R TATUS NU E Fk-4939 V02: IMPROVING MOTOR REUABILITY IN NUCLEAR

SUMMARY

REPORT. Data As Of March 31.1988.(Gray Book I) POWER PLANTS. Volume 2. Functional Indicator Tests On A Small NUREG-0020 V12 N05: LICENSED OPERATING REACTORS STATUS Electne Motor Sublected To Accelerated Aging.

SUMMARY

REPORT. Data As Of April 30,1988.(Gray Book l) NUREG/CR-4939 V03: IMPROVING MOTOR RELIABILITY IN NUCLEAR NUREG-0020 V12 N06: LICENSED OPERATING REACTORS STATUS POWER PLANTS. Volume 3. Failure Analysis And Diagnostic Tests On

SUMMARY

REPORT Data As Of May 31,1988 (Gray Book f) A Naturally Aged Large Electnc Motor. NUREG 0020 V12 N07: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of June 30.1dBB.(Gray Book 1) SHIER,W. NUREG-0020 V12 NOB: LICENSED OPERATING REACTORS STATUS NUREG/CR 5053: OPERATING EXPERIENCE AND AGING ASSESS-SUMMAHY REPORT. Data As Of July 31.1988.(Gray Book 1) MENT OF MOTOR CONTROL CENTERS. i NUREG-0020 V12 N09: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of AuDust 31,1988.(Grav Book 1) SHINOZUKA,M. NUREG 0020 V12 N10. LICENSED OPERATING REACTORS STATUS NUREG/CR 3509. POWER SPECTRAL DENSITY FUNCTIONS COM-

SUMMARY

V12 REPORT. Data As Of September 30,198BJGray PATIBLE WITH NRC REGULATORY GUIDE 1.60 RESPONSE SPEC-Book 1)TUS NUREG.0020 N11: LICENSED OPERATING REACTORS STA TRA.

SUMMARY

REPORT. Data As Of October 31,19BB.(Gray Book 1) SHULL,R. NUR G C 3545 V06: GEOPHYSICAL INVESTIGATIONS OF THE A NA CONANN N DOE WESTERN OHIO-INDIANA REGION. Annual Report. October 1986 . amal RepoMeWaW HBL September 1987. S!AHMED,E. SCHWARZ,C.E NUREG/CR-5157: THE DEVELOPMENT OF APRIL. MOD 2. A COMPUT-NUREG/CR$146: DEBRIS DISPERSAL FROM REACTOR CAVITIES ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL-DURING HIGH-PRESSURE MELT EJECTION ACCIDENT SCENAR- ING WATER NUCLEAR REACTORS. lOS. SIEGEL.B.L. SCIACCA,F. NUREG-1231 S01: SAFETY EVALUATION REPORT RELATED TO BAB-NUREG/CR-4555 RO1: GENERIC COST ESTIMATES FOR THE DIS, COCK AND WILCOX OWNERS GROUP PLANT REASSESSMENT POSAL OF RADIOACTIVE WASTES PROGRAM. NUREG/CR-5138: VALIDATION OF GENERIC COST ESTIMATES FOR CONSTRUCTION-RELATED ACTIVITIES AT NUCLEAR POWER SIEGEW PLANTS Final Report. NUREG/CR-3899 S01: UTILITY FINANCIAL STABILITY AND THE NUFIEG/CR 5160: GUIDEUNES FOR THE USE OF THE EEDB AT THE AVAILABILITY OF FUNDS FOR DECOMMISSIONING.An Analysis Of SUB COMPONENT AND SUBSYSTEM LEVELFinal Report. Internal And External Funding. SCOTT,BP. SIEGEL,M.D. NUREG/CR 5067: EARLY AND CONTJNUING EFFECTS OF COMBINED NUREG/CR 4807: SURFACE COMPLEXATION MODELING OF RADIO-ALPHA AND BETA IRRADIATION OF THE LUNG Phase il Report. NUCLIDE ADSORPTION IN SUBSURFACE ENVIRONMENTS. NUREG/CR-5198, INHALED (239)PUO(2) AND/OR TOT AL BODY NUREG/CR-486e V01: THERMODYNAMIC TABLES FOR NUCLEAR GAMMA RADIATION.Early Mortality And Morbidity in Rats And Dogs. WASTE ISOLATION. Aqueous Solutions Database. SCOTT,P. SILVESTER LF. i NUREG/CR-4082 V06- DEGRADED PIPING PROGRAM - PHASE 11 Sixth NUREGICR 4864 V01: THERMODYNAMIC TAELES FOR NUCLEAR Program Report. October 1986 - September 1987. WASTE ISOLA RON. Aqueous Solutions Database.

80 Pe:sonal Author index SIMlON,G. STENNER,R.D. NUREG/CR 4555 RO1: GENERIC COST ESTIMATES FOR THE DIS- NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW POSAL OF RADIOACTIVE WASTEa CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOAC. NUREG/CR-5138: VALIDATION OF GFNERIC COST ESTIMATES FOR TiVE WASTE DISPOSAL Environmental Monitonng And Surveillance CONSTRUCTION RELATED ACTIVhlES AT NUCLEAR POWER Programs. PLANTS. Final Report. NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE STETKAR,J.W. SUB COMPONENT AND SUBSYSTEM LEVELFinal Report- NUREG/CR 5113: FINDINGS OF THE PEER REVIEW PANEL ON THE DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG-1150. NUREG/CR 2336: STEAM GENERATOR TUBE INTEGRITY STOETZEL,G.A. PROGRAM. Phase ll Final Report NUREG/CR-52t2: EMERGENCY ENVIRONMENTAL SAMPLING AND SINGHAL,A.K. ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES. NUREG/CR-5233 A COMPUTER CODE FOR FIRE PROTECTION AND STOKLEY,J RISK ANALYSIS OF NUCLEAR PLANTS. NUREG/CR-5150: STEAM GENERATOR OPERATING SKAGGS,B.J. EXPERIENCE. Update For 1984-1986. NUREG/CR-5090: EFFECTS OF TEMPERATURE AND HUMIDITY ON RESPIRATOR FIT UNDER SIMULATED WORK CONDITIONS. ST NUREG/bR 5194: RELAP5/ MOD 2 MODELS AND CORRELATIONS. SKREINER,K.M. NUREG/CR-4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR STRUCKMEYER,R. POWER PLANTS. Volume 2.Funcuonal Indicator Tests On A Small NUREG 0837 V07 N04: NRC TLD DIRECT AADIATION MONITORANG j Electnc Motor Subjected To Accelerated Aging. NETWORK. Progress Report, October-December 1987.  ! NUREG4837 V08 N01: NRC TLD DIRECT RADIATION MONITORING SMITH.R.L NETWORK. Progress Report, January-March 1988. l NUREG/CR-0130 ADD 04: TECHNOLOGY. SAFETY AND COSTS OF NUREG-0837 V06 NO2: NRC TLD DIRECT RADIATION MONITORING DECOMMISSIONING A REFERENCE PRESSURIZED WATER REAC- NETWORK. Progress Report. Apnl June 1988. l TOR POWER STATION. Technical Support For Decommissiocung Mat-l ters Related To Preparation Of The Final Decommissioning Rule. SUBUDHi,M. NUREG/CR-0672 ADD 03: TECHNOLOGY, SAFETY AND COSTS OF NUREG/CR-4939 V01: IMPAOVING MOTOR RELIABILITY IN NUCLEAR DECOMMISSIONING A REFEPENCE BOILING WATER REACTOR POWER PLANTS. Volume 1. Performance Evaluation And Maintenance POWER STATION. Technical Support For Decommissioning Matters Practices. Related To Preparation Of The Final Decommissioning Rule. NUREG/CR-4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR SNIPES,M.B. POWER PLANTS. Volume 2. Functional indicator Tests On A Small Electnc Motor Subjected To Accelerated Aging NUREG/CR-5067: F ARLY AND CONTINUING EFFECTS OF COMBINED NUREG/CR-4939 V03: IMPROVING MOTOR RE'LIABI;1TY IN NUCLEAR ALPHA AND BETA IRRADIATION OF THE LUNG. Phase 11 Report. POWER PLANTS. Volume 3. Failure Analysis And Diagnostic Tests On A Naturally Aged Large Electnc Motor. SOO.P'EG/CR-3444 NUR V05: THE IMPACT OF LWR DECONTAMINATION NUREG/CR-5051: DETECTING AND MITIGATING BATTERY CHARGER

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ATION X UR'E Annu I Re rt FY 987. NUREG/CR- 052 OPERATING EXPERIENCE AND AGING ASSESS. NUREG/CR-5153: THE TEACHABILITY AND MECHANICAL INTEGRITY MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR-OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED IZED WATER REACTORS. IN CEMENT AND VINYL ESTER-STYRENE. NUREG/CR-5053: OPERATING EXPERIENCE AND AGING ASSESS-MENT OF MOTOR CONTROL CENTERS. SOO,SL NUREG/CR-5065: TIME. AND VOLleME AVERAGED CONSERVATION SUEN,C.J. EQUATIONS FOR MULTIPHASE FLOW USING MASS WElGHTED VE. NUREG/CR 5204: LOW-LEVEL RADIOACTIVE WASTE SOURCE TERM LOCITY AND INTERNAL ENERGY. MODEL DEVELOPMENT AND TESTING. SOWATSKEY,P.J. SUGARMAN,A.C. NUREG/CR 4597 V02: AGING AND SERVICE WEAR OF AUXIL!ARY NUREG/CR-4939 VOI: IMPROVING MOTOR RELIABILITY IN NUCLEAR FEEUWATER PUMPS FOR PWR NUCLEAR PLANTS. Volume 2.Agmg POWER PLANTS. Volume 1. Performance Evaluation And Maintenance Assessments And Monttonng Method Evaluations. Practices. SPLETZER,B.L. NUREG/CR-4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR POWER PLANTS. Volume 2. Functional indicator Tests On A Small NUREG/CR-4763. SAFETY RELATED EQUIPMENT SURVlVAL IN HY- Electric Motor Subjected To Accelerated Aging. DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS. SULLfvAN,T.M. STALLMANN,F.W. NUREG/CR-5204: LOW-LEVEL RADIOACTIVE WASTE SOURCE TERM NUREG/CR-4947: ANALYSIS OF THE A302B AND A5338 STANDARD MODEL DEVELOPMENT AND TESTING. REFERENCE MATERIALS IN SURVEILLANCE CAPSULES OF COM-MERCIAL POWER REACTORS- SUO-ANTTILA A. NUREG/CR-5019: NEUTRON EXPOSURE PARAMETERS FOR THE METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY-SEC* NUREG/CR-5219: THE MIXING OF IMMISCtBLE LIQUlO LAYERS BY GAS BUBBLING' TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES. NUREG/CR-5049 PRESSURE VESSEL FLUENCE ANALYSIS AND SWANSON,D. NEUTRON DOSIMETRY. NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA-  ! STEELE R. TION SYSTEMS. NUREG/CR-5043: CONTAINMENT PENETRATION SYSTEM (CPS) NUREG/CR-4242: SURVEY OF LIGHT WATER RE/ (10R CONTAIN-TESTS UNDER ACCIDENT LCADS. MENT SYSTEMS. DOMINANT FAILURE MODES AND MITIGATION OPPORTUNITIES. Final Report. STEELE.T.D- I l NUREG/CR-5107: HYDROGEOLOGIC CHARACTERIZATION OF SZUKIEWICZ,A.J. l BASALTSThe Northern Rim Of The Columbia Plateau Physiographic NUREG 1217 DRFT FC: EVALUATION OF SAFETY IMPLICATIONS OF ' Province And Of The Creston Study Area. Eastern Washington. CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS Technical Findings Related To Unresolved Safety lasue A 47. Draft Report For STEEPLES.D.W. Comment. ) NUREG/CR-5045; KANSAS-NEBRASKA SEISMICITY STUDIES USING NUREG-1218 DRFT FC: REGULATORY ANALYSIS FOR PROPOSED THE KANSAS-NEBRASKA MICROEARTHOUAKE NETWORK. Final RESOLUTION OF USl A-47. Safety imphcations Of Control Report. Systems. Draft Rept For Comment. l

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i t Personal Author index ' 81 l TABATABAI,A. THURBERJ.A. ~ NUREG/CP-0099: PROCEEDINGS OF THE PUBLIC WORKSHOP FOR NUREG/CR-4991: EVALUATION AND PROPOSED IMPROVEMENTS TO l NRC RULEMAKING ON MAINTENANCE OF NUCLEAR POWER EFFECTIVENESS OF U.S. NUCLEAR NEGULATORY COMMISSION PLANTS. GENERIC COMMUNICATIONS. TARBELL,W.W. 13CHLERJ. NUREG/CR-4914: THE INFLUENCE OF SELECTED CONTAINMENT ' NUREG/CR-2907 V06: RADIOACTIVE MATERIALS RELEASED FROM STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW- NUCLEAR POWER PLANTS. Annual Report For 1985. ING HIGH PRESSURE MELT EJECTION FROM THE REACTOR NUREG/CR 2907 V07: RADIOACTIVE MATERIALS RELEASED FROM VESSEL NUCLEAR POWER PLANTS. Annual Report For 1986. NUREG/CR 4917: DCH.2:RESULTS FROM THE SECOND EXPERIMENT PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. TINGLE.A. NUREG/CR 4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT-TAWILJ.J. - RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER NUREG/CR.4811: THE ECONOMIC COSTS OF RADIATION-INDUCED PLANT. Draft Report For Comment. I HEALTH EFFECTS. Estimaton And Simulation. NUREG/CR 5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE. '! NUREG/CR 5009: ASSESSMENT OF THE USE OF EXTENDED MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF BURNUP FUEL IN LIGHT WATER POWER REACTORS. GENERIC LSSUE 99.  ! TAYLOR,J. TOSIAS,M.L - NUREG/CR4052: OPERATING EXPERIENCE AND AGING ASSESS-NUREG/CR-5018: URANIUM OXIDE IRON OXIDE MIXED AEROTOL MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR-IZED WATER REACTORS. EXPERIMENTS IN STEAM-AIR ATMOSPHERES.NSPP - Tests 611,612.613 And 831, Data Record Report TAYLOR.J.H. TOKSOZ,M.N. NUREG/CR-4939 V01: IMPROVING MOTOR RELIABILITY IN NUCLEAR POWER PLANTS. Volume 1. Performance baluston And Maintenance NUREG/CR 5080 A STUDY OF NEW ENGLAND SEISMICITY WITH l EMPHASIS ON MASSACHUSETTS AND NEW HAMPSHIRE. Final 1 NURE Cb4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR ep rt ng ed 19M85. ' POWER PLANTS. Volume 2. Functional indicator Tests On A Small TOMAN,G.J. Electnc Motor Subtected To Accelerated Agmg-3 NUREG/CR-4939 V03: IMPROVING MOTOR RELIABILITY IN NUCLEAR NUREG/CR-4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE I POWER PLANTS Votume 3. Failure Analysis And Diagnostic Tests On SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER A Naturally Aged Lar9e Electne Motor. PLANTS. Operating Experience And Failure identifica;on. NUREG/CR-5141: AGING AND OVALIFICATION RESEARCH ON SOLE-TAYLOR,K. NOID OPERATED VALVES. NUREG/CR-5165: SEISMOLOGICAL INVESTIGATION OF EARTH" TOOUAM,J OUAKES IN THE NEW MADRIO SEISMIC ZONE AND THE NORTH-EASTERN EXTENT OF THE NEW MADRID SEISMIC ZONE. Final NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER Report. September 1981 December 1986. INDUSTRY.A Review Of Technicallanues. TERAO D. ~ ^ A~ NUREG 0797 S18: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-4312 V01: RELAP5/ MOD 2 CODE MANUAL. Volume 1: Code OPERATION OF COMANCHE PEAK STEAM ELECTRtc NUR G C STATION, UNITS 1 AND 2. Docket Nos. 50 445 And 50-446.(Texas utili- 12 02 1 LA 2 DE MANUALVolume 2: ties Generating Company) Users Guide And input Requirements. TERRELL,J.B. TRIPATHt,V.S. . . NUREG/CR 5013. FATlGUE LIFE CHARACTERIZATION OF SMOOTH NUREG/CR 4807: SURFACE COMPLEXATION MODELING OF RADIO- , AND NO'CHED PIPING STEEL SPECIMENS IN 288 DEGREES C AIR NUCLIDE ADSORPTION IN SUBSURFACE ENVIRONMENTS. j ENVIRONMENTS. TUTU,N.K* NUREG/CR-5136: FATIGUE STRENGTH OF SMOOTH AND NOTCHED f1 SPECIMENS OF ASME SA 106 B STEEL IN PWR ENVIRONMENTS. NUREG/CR 5146: DEBRIS DISPERSAL ^ FROM REACTOR CAVITIE'S 1 DURING HIGH-PRESSURE MELT EJECTION ACCIDENT SCENAR- q THEOFANOUS,T.G. IOd i NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE l DRAFT REACTOR RISK REFERENCE DOCUMENT.NUREG 1150. TUZLA,K. .) NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS RELA. NUREG/CR 5095 V01: THERMODYNAMIC NONEQUILIBRIUM IN POST. l TNE TO PRESSURIZED THERMAL SHOCK. CRITICAL HEAT FLUX BOILING IN A ROD BUNDLE.Desenpton Of Expenments And Sample Results. THINNES,0. NUREG/CR-5095 V02: THERMODYNAMIC NONEQUILIBRIUM IN POST-NUREG/CR 5031: SIGNIFICANCE OF IN-STRUCTURE GENERATED CRITICAL HEAT. FLUX BOILING IN A ROD BUNDLE. Data For Stabi-MOTION IN SEISMIC QUALIFICATION TESTS OF CABINET MOUNT. lized Quench Front Tests. ED ELECTRICAL DEVICES. NUREG/CR-5095 V03: THERMODYNAMIC NONEQUILIBRIUM IN POST- l THOMAS,C.W. CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Advanc-Ing Quench Front Tests. NUREG/CR-4879 V01: DEMONSTRATION OF PERFORMANCE MOD- NUREG/CR-5095 V04: THERMODYNAMIC NONEQUILlBRIUM IN Pool-ELING OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE. Data For Retreat-Companson Of Predictive Radionuchde Transport Modehng Versus inD Ouench Front Tests.- i Feld Observations At The Nitrate Disposal Pit Site. Chalk River Nuclear Labs. UNAL,C. i THOMS,KA NUREG/CR-5095 VD1: THERMODYNAMIC NONEQUILibRIUM IN POST-CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE.Desenption Of l NUREG/CR 4880 VO1: CHARACTER 12ATION OF IRRADIATED CUR. Expenments And Sample Results. - RENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUREG/CR-5095 V02: THERMODYNAMIC NONEOUILIBRIUM IN POST. i NUCLEAR PRESSURE VESSEL SERVICE. CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLEdata For Stabi-NUREG/CR 4880 V02: CHARACTERIZATION OF IRRADIATED CUR- lized Quench Front Tests. RENT-PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUREG/CR 5095 V03: THERMODYNAMIC NONEOUILIBRIUM IN POST-NUCLEAR PRESSURE VESSEL SERVICE. CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Advanc-' ing Quench Front Tests. THORNE,L.R. - NUREG/CR-5095 V04: THERMODYNAMIC NONEOUILIBRIUM IN POST-NUREG/CR-5006: PLATINUM CATALYTIC (GNITERS FOR LEAN HY- CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Retreat. DROGEN-AIR MIXTURES. mg Quench Front Tests. l

82 Personal Author index UNWIN,S. WANG,J.K. . NUREG/CR 4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR-5159: PREDICTION OF CHECK VALVE PERFORMANCE RISKS AND POTENTIAL FOR RISK REDUC'llON. ZION POWER AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. - PLANT. Draft Report For Comment. WANG,Y K.' UNWIN S.D. NUREG/CR-5105: RESPONSE MARGINS INVESTIGATION OF PIPING NUREG/CR-4688 V02: QUANTIFICATION AND UNCERTAINTY ANALY- DYNAMIC ANALYSES USING ' THE INDEPENDENT SUPPORT SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT MOTION METHOD AND PVRC DAMPING. WATER REACTORS (OUASAR)Part II: Sensitiwty Analysis Tech-naques. WATLINGTON,B. -)' NUREG/C45138: VAUDATION OF GENERIC COST ESTIMATES FOR

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URE /dR.4873. BENCHMARK STUDY OF THE l DYNEV EVACU- S F nel ATION TIME ESTIMATE COMPUTER CODE. NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE NUREG/CR 4874: THE SENSITIVITY OF EVACUATION TIME ESTi- SUB COMPONENT AND SUBSYETEM LEVEL. Final Report. ' MATES TO CHANGES IN INPUT PARAMETERS FOR THE l-DYNEV COMPUTER CODE. WATSON,D.G. VALENTILN NUREG/CR-5047: RADIONUCLIDES ACCUMULATION BY AQUATIC NUREG/Cb 5178'- EVALUATION OF GENERIC ISSUE BIOTA EXPOSED TO CONTAMINATED WATER IN ARTIFICIAL ECO-125.lL7, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE SYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH THE FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK . GROUND. VALLARIO,R.W. WAY,S C. l NUREG/CR-4991: EVALUATION AND PR14 OSED IMPROVEMENTS TO NUREG/C45240: COMPARATIVE EVALUATION OF SELECTED CON- I EFFECTIVENESS OF U.S. NUCLEAR REGULATORY COMMISSION TINUUM AND DISCRETE. FRACTURE MODELS. Emphasis On Disper-GENERIC COMMUNICATIONS. sivity Calculations For Apphcation To Fract. ed Geologic Media, Cres-NUREG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PERFORM- ton Ftudy Area, Eastern Washington. ANCE IND CATORS. NUREG/CR-5277: THE TENSORIAL NATURE OF EFFECTIVE POROJ TY AND LARGE SCALE DISPERSION COEFFICIENTS.Apphcation To NURE /d 5063: DEVELOPMENT OF A MECHANISTIC UNDE 4 STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES- WEBER,D.S. SURE VESSEL STEELS. Final Report. NUREG/CR-5255: STABLE ISOTOPES OF AUTHIGENIC MINERALS IN

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VISKANTA R. NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE WEDER,J.R. DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG-1150. NUREG/CP-0093: PROCEEDINGS OF THE MEETING ON ULTRASEN-VO T.V. SITIVE TECHNIQUES FOR MEASUREMENT OF URANIUM IN BIO-NUREG/CR-5058: PRA APPLICATIONS PROGRAM FOR INSPECTION LOGICAL SAMPLES AND THE NEPHROTOXICITY OF URANIUM. AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50-313.(Arkansas WEINER,E.O Power And Ught Company) NUREG/Clk-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE VOLPONI,J.V. PREDICTION METHODS FOR PIPING. NUREG/C45086: PLATINUM CATALYTIC IGNITERS FOR LEAN HY-WEINGARDT,J.J. DROGEN. AIR MIXTURES. NUREG/CR-5126: TAC 2D STUDIES OF MARK 1 CONTAINMENT VOMVORIS E.G. DRYWELL SHELL MELT THROUGH. NUREG/CR-5094: APPUCATION OF STOCHASTIC METHODS TO THE SIMULATION OF LARGE-SCALE UNSATURATED FLOW AND WEISS.A.J. TRANSPORT. NUREG/CP-0091 V01: PROCEEDINGS OF THE FIFTEENTH WATER REACTOR SAFETY INFORMATION MEETING. WAGNER,K.C. NUREG/CP.0091 V02: PROCEEDINGS OF THE FIFTEENTH WATER NUREG'C45225: AN OVERVIEW OF BWR MARK-1 CONTAINMENT REACTOR SAFETY INFORMATION MEETING. VENTING RISK IMPUTATIONS. NUREG/CP-0091 V03: PROCEEDINGS OF THE FIFTEENTH WATER REACTOR SAFETY INFORMATION MEETING. WAGNER,R.J. NUREG/CP-0091 V04: PROCEEDINGS OF THE FIFTEENTH WATER NUREG/CR 4312 V01: RELAP5/ MOD 2 CODE MANUALVolume 1: Code REACTOR SAFETY INFORMATION MEETING. Structure. Systems Models And Solution Method 9 NUREG/CP-0091 V04 ADD- PROCEEDINGS OF THE FIFTEENTH NUREG/CR 4312 V02 H1: RELAPS/ MOD 2 CONE MANUALVolume 2: WATER REACTOR SAFETY INFORMATION MEETING. Users Guide And input Requirements. NUREG/CP 0091 V05: PROCEEDINGS OF THE FIFTEENTH WATER REACTOR SAFTEY INFORMATION MEETING. G/dRI5023: HIGH LEVEL SEISMIC RESPONSE AND FAILURE REACTOR SAFETY INFORMATION MEETING. PREDICTION METHODS FOR PtPING. NUREG/CP-0096: TRANSACTIONS OF THE SIXTEENTH WATER RE. WALKER,J.V. ACTOR SAFETY INFORMATION MEETING. NUREG/CR-5039 V02: REACTOR SAFETY RESEARCH SEMIANNUAL NUREG/C42331 V7N2 3: SAFETY RESEARCH PROGRAMS SPON-REPORT. July December 1987. Reactor Safety Research Program. SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Progress Report April. September 1987. WALLIS,G.B. NUREG/CR-2331 V8N1-2: SAFETY RESEARCH PROGRAMS SPON. NUREG/C45220 V01: DIAGNOSIS OF CONDENSATION-INDUCED SORED BY OFFICE OF NUCLEAR REGULATORY WATERHAMMER. Methods And Background. RESEARCH Progress Report, January-June 1988. 9VREG/CR-5220 V02: DIAGNOSIS OF CONDENSATION-INDUCED WATERHAffMEH Case Studies. WELLAhD,H.J. NUREG/CR-5178: EVALUATION OF GENERIC ISSUE WAMBLFY,R.M. 125.II.7. REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE NUREG/CR.5041 V02: RECOMMENDATIONS TO THE NRC FOR FEEDWATER FROM STEAM GENERATOR DURING A UNE BREAK. REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOACTIVE 5 ASTE DISPOSAL Task 2ttEarth-Mounded Concrete WEST,D.B. Bunkers. NUREG/CR-5106: USER'S GUIDE FOR THE TACT 5 COMPUTER CODE. WANG.C-H. WHEATLEY,C.J. NUREG/CR-5133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW NUREG/CR.5162. CHARM.A MODEL FOR AEROSOL BEHAVIOR IN THROUGH INTERGRANULAR STRESS CORROSION CRACKS. TIME VARYlNG THERMAL-HYDRAUUC CONDITIONS.

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I  ? Personil Author index 83 1 A,C.L WOLFORD,A. EG/CR-5048: REVIEW OF THE NATURAL CIRCULATION EFFECT NUREG/CR-5248: PRIORITIZATION OF TIRGALEX. RECOMMENDED A THE VERMONT YANNEE SPENT-FUEL POOL. Docket No. 50- COMPONENTS FOR FURTHER AGING RESEARCH. 271.(Vermont Yankee Nuclear Power Corp) WHEELER,K.R. NUREG/CR-4916: HECTR ANALYSES OF THE NEVADA TEST SITE NUREG/CR-2336: STEAM GENERATOR TUBE INTEGRITY (NTS) PREMIXED COMBUSTION EXPERIMENTS. PROGRAM. Phase li Final Report. NUREG/CR 4993: A STANDARD PROBLEM FOR HECTR-MAAP COMPARISON. incomplete Burning. WHEELER,T.A. NUREG/CR-4836: APPROACHES TO UNCERTAINTY ANALYS!S IN WONG.S.M. PROBABILISTIC RISK ASSESSMENT. NUREG/CR.5200: EVALUATION OF RISKS ASSOCIATED WITH AOT AND STI REQUIREMENTS AT THE ANO-1 NUCLEAR POWER WHtTE.J.E. PLANT. NUREG/CR-5264 GUIDE FOR LICENSING EVALUATIONS USING CRAC2.A Computer Program For Calculating Reactor Accident Conse. WORLEDG'" D. quences. NUREG/CR 4780 V01: PROCEDURES FOR TREATING COMMON CAUSE FAILURES IN SAFETY AND RELIABluTY WHITEHEA3,D.W. STUDIES. Procedural Framework And Examples. NUREG/CR 4834 V02: RECOVERY ACTIONS IN PRA FOR THE RISK METHODS INTEGRATION AND EVALUATION PROGRAM WREATHALL,J. (RMIEP). Volume 2:Apphcation Of The Data-Based Method. NUREG/CR-5248: PRIORITIZATION OF TIRGALEX RECOMMENDED COMPONENTS FOR FURTHER AGING RESEARCH. WlERINGA,0. NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER WU,$. INDUSTRY.A Review Of Technicalissues. NUREG/CR-3950 V04: FUEL PERFORMANCE ANNUAL REPORT FOR 1986. WILKOWSKI,G.M. NUREG/CP 0092: PROCEEDINGS OF THE SEMINAR ON LEAK. WULFF,W. BEFORE BREAK. Progress in Regulatory Pohcies And Supporting Re. NUREG/CR-5232: UNCERTAINTIES IN MODELING AND SCALING IN search. THE PREDICTION OF FUEL STORED ENERGY AND THERMAL RE-NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM - PHASE ll. Sixth SPONSE. Program Report. October 1986 Septer.iber 1987. YAMASHITA,T. WILLIAMS,K.A. NUREG/CR-4777: STEAM OXIDATION OF ZlRCALOY CLADDING IN NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS RELA. THE ORNL FISSION PRODUCT RELEASE TESTS. TIVE TO PRESSURIZED THERMAL SHOCK. V ANIV,S.S. WILLIAMS M.L NUREG/CR-5190: INHALED (2.9)PUO(2) AND/OR TOTAL-BODY NUREG/CR-4984: DEVELOPMENT OF A THREE-DIMENSIONAL FLUX GAMMA RADIATION.Early MortL4y And Morbidity in Rats And Dogs. SYNTHESIS PROGRAM AND COMPARISON WITH 3-D TRANSPORT THEORY RESULTS. YEH,T.-C. JIM NUREG/CR-5049: PRESSURE VESSEL FLUENCE ANALYSIS AND NUREG/CR-5097: SIMULATION OF LIOUlO AND VAPOR MOVEMENT NEUTRON DOSIMETRY. IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP TUFF SITE.Models And Strategies. NUREG-1226. DEVELOPMENT AND UTILIZATION OF THE NRC YOON.W.H. POLICY STATEMENT ON THE REGULATION OF ADVANCED NU. NUREG/CR-5015. IMPROVED RELIABILITY OF RESIDUAL HEAT RE-CLEAR POWER PLANTS. MOVAL CAPABILITY IN PWRS AS nELATED TO RESOLUTION OF GENERIC ISSUE 99. WINGERT,V.L NUREG-0654 S01 RO1. CRITERIA FOR PREPARATION AND EVALUA- YOST,P. TION OF RADIOLOGICAL EMERGENCY RESPONSE PLANS AND NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER PREPAREDNESS IN SUPPORT OF NUCLEAR POWER INDUSTRY.A Review Of Technicalissues. PLANTS Critena For Utihty Offsite Planning And Preparedness. WISBE Y,s.J. NUREG/CR 4879 V01: DEMONSTRATION OF PERFORMANCE MOD-hUREG/CR 477H: PRELIMINARY STUDIES OF THE MORPHOLOGY OF ELING OF A LOW-LEVEL WASTE SHALLOW-LAND BURIAL SITE.A THERMAL GRADIENT TUBE DEPOSITS FROM FISSION PRODUCT Companson Of Predictive Radionuchde Transport Modehng Versus FIELEASE EXPERIMENTS. Field Observations At The Nitrate Disposal Pit Site, Chalk River Nuclear Labs. WITTE M.C. NUREG/CR 4775: GUIDE FOR PREPARING OPERATING PROCE. YOUNG,M.F. DURES FOR SHIPPING PACKAGES. NUREG/CR-5084: IFCI. ' 'NTEGRATED CODE FOR CALCULATION OF ALL PHASES OF F~ 100LANT INTERACTIONS. NUREG/CR-5183. A USER'S MANUAL FOR THE CONTAMINANT YOUNGBL')OD,R. TRANSPORT MODULE OF THE MIGRAT CODE. NUREG/CR-4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER WITTLER.R.J. PLANT. Draft Report For Comment. NUREG/CR 4651 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA NUREG/CR-4999: ESTIMATION OF RISK REDUCTION FROM IM-BY RIPRAP TESTING IN FLUMES. Phase lifollowup Investigations. PROVED PORV RELIABLITY IN PWRS Final Report.

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                           ' .N         THE VERMONT                                    YANKEEG/CR.5048:                                                    REVIEW                  OF T 271.(Vermont WHEELER,K.R.                         EYankee SPENT        Nucl FUHE NATURAL                              CIRCULATION NUREG/CR 2336:

ear Power Corp) EL POOLDocket N0EFFECT 50-WOLFORD,A. tJUREG/CR424B: Personil Author83index PROGRAMPhase lf FSTE/M anat Report. GENERATOR WHEELER,TA TUBE INTEGRITY COMPONENTS FOR OF FURTHEPRIORtTIZA NUREG/CR-4836: NOREG/CR-4916: R AGING . RESEARCHTIRGA PROBABlUSTIC WHfTE,J.E. RISK SESSMENT. UNCERTAINTY ASAPPROACHES TO NUREG/CR-4903:(NTS) PREMtXED COMB A NUREGICR4264: CRAC2 GUIDE FOR ANALYSIS IN COMPARtSON. WONG,5.M. STANDAROUSTION incomplete rrun0 PROBLEM . Bu A TEST SITEEXPER FOR Quences.A Computer LICENSING Program EVALUATIONS For Calculat NUREG/CR-5200: AND HECTR-MAAP WHfTEHEAD.D.W. ing Reactor Accident ConsUSING PLANT, STI REQUIREMENTSEVALUATION OF RI e. NUREGICR-4834 METHODS V02: WORLEDGE.D. AT THE ASSOCIATED ANO-1 WTEGP RECOVERY WITH AOT (RMlEP). Volume 2 App 4 TION ANDc NUCLEAR POWER WlErdNGA,D. ACTIONS EVALUATION IN PRA E RISK FORNUREG/CR-4780 TH CAUSE V01: u ation Of 1he Data Ba FAILURES NUREG/CR4227: - sed Method. PROGRAM STUDIES. Procedural WREA THALL,.i. IN SAFETYFORFramPROCEDURE TREATING INDUSTRY.A Review Of TFITNESS FOR DUTY WILKOWSKI,G.M. NUREG/CR 5248: ework And Examples AND COMMON REUABluTY NUREG/CP-0092: echnicatissues.IN THE NUCLEAR WU,S.COMPONENTS POWER FOR FURTHPRtORITIZATION BEFORE BREAK. Progress search. OF PROCEEDINGS And cTHENUREG/CR.3950 ER AGING RESEARCHOF ED TIR n Regulatory Policies SEMfNAR V04: FUEL P 1986. NUREG/CR-4062 Program Report Octob V00: DEGR ON LEAK. WOLFF,W. WILLIAMS,KA er 1966. September 1967ADED PIPING PROGRAM ERFORMANCE apportingORT Re. ANNUAL FOR REP

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SPONSE. AINTIES IN MODEUNG AN NUREG/CR-4984: ERMAL SHOCK. SBLOCAS RELA.MAL HYDRAUUCS OF L STORED ENERGY D SCAUNG IN AND THERMAL RE-THEORY RESULTSSYNTHESIS NUREG/CR4049: . AND COMP NUREG/CR-4777. PROGRAMDEVELOPMENT VANIV.S S. XIDATION OF STEAM OF A THRE ZlRCALOY OTHE NEUTRON DOSIMETRYPRESSURE . ARISON WITH 3-D ORT VESSEL TRANSPEDIMENSIONAL NUREG/CR4198: FL CT RELEASE TESTS. FLUX CLADDING IN NUREG 1226: UENCE ANALYS6S AND GAMMA RAD YEH,T. C, JIM (ATIONEartINHALED

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POLICY STATEMENTDEVELOPMENT AND rtality And AND/OR Morbidity In B TOTAL EODY CLEAR POWER PLANTSON NUREG/CR-5097: IN OF THE THE REGULATIONUTILtZATIONUNSATURATED ION OFSIMULAT ata And Dogs. WINGERT V.L. ' . OF UQUID NOREG-0654 TtON OF S01 R01: ADVANCEDNRC NU. TUFF SITE.Models YOON,W.H. rategies. And StFRACTUR HE APACHE LEAP PREPAREDNESSRADIOLOGICAL IN EMERGENCRITERIA NUREG/CR4015: FOR PREPARAT IMPROV O PLANTS.Cntena WISBEY,5.J. y SUPPORT For f%ne Planning OF Utilit CY RESPONSE And Pr NUCLEAR O AND PLANSI GENERIC YOST,P. N AND EVALUA-

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AS RELATED TO TION OF RESOLUL HEAT NUREG/CR.4778. RELEASE EXPERIMENTS. PPREUMINTHERMAL INDUSTRY.A Review YOUNG.J1. GFIADIENT OFITNESS FOR TUD DUTY IN WITTE,M.C. NUREG/CR 4775: ARY STUDIES N PRODUCT OFNUREG/CR-4879 THE MORE EUNG OF t Technicallanues. V01: DEPOSITS THEFROM NUCLEAR FtSPO DURES FOR KAGES, SHIPPING ARtNG PACGUIDE FOR CompensonPREP A LOW LEVEL R Of Predictwe WASTE DEMONSTRATION OF Field Observations AtHALLOW-LAND Labs. Th NUREG/CR4A tB3: USER'S OPERATING PHOCE. YOUNG,M.F. edonuchde Transport ITE.A e Nitrate Disposa! PitingSite ModelBURIAL Ch l PERFO Versus TRANSPORT MODULE WITTLER.RJ. MiGRAT THE OF THEMANUAL NUREG/CR FOR 5084: a k Rwer Nuclear CODE. CONTAMINANT OF ALL PHASES OF FUEL CIFCt: AN INTEGRATE NUREG/CR-4651 LOPMENT MES. Phase 11 V02: OFFollowRIPRAP DEVEBY pin IGN DES YOUNGBLOOD,R. NUREG/CR RISNS R1PRAP TESTING IN FLU CRITERIA 4551 - OOLANTINTERACTIONSDE FOR CALCU V5 DRF: u vestgatons. PLANT. Draft Report FoAND POTENTIAL FOR REVALUATION O NUREG/CR-4999 SEVERE PROVED ESTtMATIONr PORV REUABlU OF RISK CommentPOWER ISK REDUCTtONZlONACC TYIN PWRS.FinalRepo r. tREDUCTION IM-FROM

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            .W THE VERMONT YANKEE SPENT-FULL POOL. Docket No. 50-                   COMPONENTS FOR FURTHER AGING RESEARCH.
      - ' 271.(Vermont Yankee Nuclear Power Corp)

WHEELER.K.R. NUREG/CR-2336: NUREG/CR-4916: HECTR ANALYSES OF THE NEVADA TEST SITE STEAM GENERATOR TUBE INTEGRITY (NTS) PREMIXED COMBUSTION EXPERIMENTS. PROGRAM. Phase ll Final Report. NUREG/CR-4993. A STANDARD PROBLEM FOR HECTR MAAP COMPARISON. Incomplete Burning. NUREG/CR 4836: APPROACHES TO UNCERTAINTY ANALYSIS IN WONG,S.M. PROBABILISTIC RISK ASSESSMENT, NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT WHITE JK. AND STI REQUIREMENTS AT THE ANO.I NUCLEAR POWER PLANT. NUREG/CR-5264: GUIDE FOR LICENSING EVALUATIONS USING CRAC2.A Computer Program For Calculating Reactor Acculent Conse. WORLEDGE D. quences. NUREG/CR-4780 V01: PROCEDURES FOR TREATING COMMON CAUSE FAILURES IN SAFETY AND RELIABILITY WHITEHEAD,D.W. STUDIES. Procedural Framework And Examples. NUREG/CR4834 YO2: RECOVERY ACTIONS IN PRA FOR THE RISK METHODS INTEGRATION AND EVALUATION PROGRAM WREATHALL,J. (RMIEP). Volume 2.Apphcation Of The Data-Based Method. NUREG/CR-5248: PRIORITIZATION OF TIRGALEX RECOMMENDED WlERINGA,D. COMPONENTS FOR FURTHER AGING RESEARCH. NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER WU,S. INDUSTRY.A Review Of Technicalissues. NUREG/CR-3990 V04: FUEL PERFORMANCE ANNUAL REPORT FOR 1986. WILKOWSKl,G.M. NUREG/CP4092: PROCEEDINGS OF THE SEMINAR ON LEAK. WULFF.W. BEFORE-BREAK. Progress in Regulatory Policies And Supportmg Re. NUREG/CR-5232: UNCERTAINTIES IN MODELING AND SCALING IN search. THE PREDICTION OF FUEL STORED ENERGY AND THERMAL RE-NUREG/CR-4082 V06: DEGRADED PIPING PROGRAM - PHASE ll. Sixth SPONSE. Program Report. October 1986 - September 1987. WILLIAMS.K.A. NUREG/CR-4777: STEAM OXIDATION OF ZlRCALOY CLADDING IN NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS RELA. THE ORNL FISSION PRODUCT RELEASE TESTS. TIVE TO PRESSURIZED THERMAL SHOCK. YANfV,S.S. WILLIAMS M.L NUREG/CR-5198. INHALED (239)PUO(2) AND/OR TOTAL-BODY NUREG/CR4984: DEVELOPMENT OF A THREE. DIMENSIONAL FLUX GAMMA RADIATION.Early Mortahty And Mortxdity in Rats And Dogs. SYNTHESIS PROGRAM AND COMPARISON WITH 3 D TRANSPORT THEORY RESULTS. YEH,T.-C. JIM NUREG/CR-5049: PRESSURE VESSEL FLUENCE ANALYSIS AND NUREG/CR 5097: SIMULATION OF LIOUlO AND VAPOR MOVEMENT NEUTRON DOSIMETRY. IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP TUFF SITE # odds And Strategies. NURLC 1226: DEVELOPMENT AND UTILIZATION OF THE NRC YOON,W.H. POUCY STATEMENT ON THE FIEGULATION OF ADVANCFD NU. NUREG/CR 5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE-CLEAR POWER PLANTS. MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF GENERIC ISSUE 99. WINGERT.V.L NUREG-0654 S01 F101: CRITERIA FOR PREPARATION AND EVALUA. YOST,P, TION OF RADIOLOGICAL EMERGENCY RESPONSE PLANS AND NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER - PREPAREDNESS IN SUPPORT OF NUCLEAR POWER INDUSTRY.A Review Of Technicalissues. PLANTS.Cntena For Utility Offsite Planning And Preparedness. WISBEY,s.J. NUREG/CR4879 V0t DEMONSTRATION OF PERFORMANCE MOD-NUREG/CR-4778: PRELIMINARY STUDIES OF THE MORPHOLOGY OF ELING OF A LOW LEVEL WASTE SHALLOW. LAND BURIAL SITE.A THERMAL GRADIENT TUBE DEPOSITS FROM FISSION PRODUCT Companson Of Predictive Radionuchde Transport Modehng Versus RELEASE EXPERIMENTS, Field Observations At The Nitrate Disposal Pit Site, Chalk River Nuclear WITTE M.C. NUREG/CR 4775; GUIDE FOR PREPARING OPERATING PROCE. YOUNG.M.F. DURES FOR SHIPPING PACKAGES. NUREG/CR-5084: IFCI: AN INTEGRATED CODE FOR CALCULATION OF ALL PHASES OF FUEL. COOLANT INTERACTIONS. NUREG/CR 5183: A USER'S MANUAL FOR THE CONTAMINANT YOL.NGBLOOD,R. TRANSPORT MODULE OF THE MiGRAT CODE. NUREG/CR-4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT WITTLER,R.J. RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER PLANT Draft Report For Comment. NUREG/CR-4651 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY RIPRAP TESTING IN FLUMES Phase li. Followup investigations. NUREG/CR-4999: ESTlMATION OF RISK REDUCTION FROM IM-PROVED PORV RELIABILITY IN PWRS. Final Report.

S bject Index This index was developed from keywords and word strings in titles and abstracts. During this development period, there will be some redundancy, which will be removed later when a rea-sonable thesaurus has been developed through experience. Suggestions for improvements are welcome. 3-D Transport Theory Accident Analysis NUREG/CR 4984: DEVELOPMENT OF A THREE DIMENSIONAL FLUX NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS SYNTHESIS PROGRAM AND COMPARISON WITH 3-D TRANSPORT HANDBOOK. THEORY RESULTS. Accident Condition REG CR 5070. ANALYSIS OF NATURAL-CONVECTION PHENOM- R[PRT u De e ENA IN A 3 LOOP PWR DURING A TMLB' TRANSIENT USING THE 9 6 Volu 36 COMMIX CODE. Accident Load A5338 Steel NUREG/CR 5043. CONTAINMENT PENETRATION SYSTEM (CPS) NUREG/CR-5142: DUCTILE TO BRITTLE TOUGHNESS TRANSITION TESTS UNDER ACCIDENT LOADS. CHARACTERIZATION OF A5338 STEEL. Accident Sequence ACRS Reports NUREG/CR4674 VOS: PRECURSORS TO POTENTIAL SEVERE CORE NUREG-1125 V09: A COMPILATION OF REPORTS OF THE ADVISORY DAMAGE ACCIDENTS:1986.A ST ATUS REPORT. COMMITTEE ON REACTOR SAFEGUARDS 1987 Annual NUREG/CR4674 V06: PRECURSORS TO POTENTML SEVERE CORE DAMAGE ACCIDENTS 1986,A STATUS REPORT. AEOD NUREG-1272 V02 NO1: REPORT TO THE U S. NUCLEAR REGULA- NUREG/CR-5039 V01: REACTOR SAFETY RESEARCH SEMIANNUAL REPORT. January-June 1987. Volume 37. TORY COMMIS$lON ON ANALYSIS AND EVALUATION OF OPER. ATIONAL DATA .1987. Power Reactors Acoustic EmissLon NUREG 1272 V02 NO2: REPORT TO THE U S. NUCLEAR REGULA-TORY COMMISSION ON ANALYSIS AND EVALUATION OF OPER- NUREG/CR-5144: ACOUSTIC EMISSION SYSTEM CALIBRATION AT ATIONAL DAT A .1987.Nonreactors. WATTS BAR UNIT 1 NUCLEAR REACTOR. APRIL Acoustic Leak Detect 6on NUREG/CR 5157: THE DEVELOPMENT OF APRILMOD2. A COMPUT. NUREG/CR-5134: APPLICATION OF ACOUSTIC LEAK DETECTION ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL- TECHNOLOGY FOR THE DETECTION AND LOCATION OF LEAKS IN ING WATER NUCLEAR REACTORS. LIGHT WATER REACTORS. ASME Code Adsorption NUREG/CR 4785 REVIEW AND EVALUATION OF DESIGN ANALYSIS NUREG/CR4807: SURFACE-COMPLEXATION MODELING OF RADIO-METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND NUCLIDE ADSORPTION IN SUBSURF ACE ENVIRONMENTS. BRANCH CONNECTIONS. NUREG/CR4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS Advanced Reactor METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND NUREG-1226: DEVELOPMENT AND UTILIZATION OF THE NRC BRANCH CONNECTIONS. POUCY STATEMENT ON THE REGULATION OF ADVANCED NU-ASTM NUREG/CR4996: A REPORT ON THE ROUND ROBIN PROGRAM Aerosol CONDUCTED TO EVALUATE THE PROPOSED NSTM STANDARD TEST METHOD FOR DETERMINING THE plt ME STRAIN CRACK NUREG/CR-4508. BEHAVIOR OF A CORIUM JET IN HIGH PRESSURE ARREST FRACTURE TOUGHNESS,K(IA).OF FEl.RITIC MATERIALS- MELT EJECTION FROM A REACTOR PRESSURE VESSEL-NUREG/CR4917: DCH-2:RESULTS FROM THE SECOND EXPERIMENT Abnormal Occurrence PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. NUREG-0000 V10 NO3. REPORT TO CONGRESS ON ABNORMAL NUREG/CR4997: METHODS FOR DESCRIBING AIRBORNE FRAC-TIONS OF FREE F ALL SPILLS OF POWDERS AND LIQUIDS. OCCURRENCES. NUREG-0090 V10 N04.July-Septemtwr REPORT TO 1987' CONGRESS ON ABNORMAL NUREG/CR-5018: URANIUM OXIDE lRON OXIDE MIXED AEROSOL OCCURRENCES October December 1987. EXPERIMENTS IN STEAM-AIR ATMOSPHERESESPP Tests NUREG-0090 V11 N01: REPORT TO CONGRESS ON ABNORMAL 611,612.613 And 631, Data Record Report. OCCURRENCES. January-March 1988. NUREG/CR-5162: CHARM.A MODEL FOR AEROSOL BEHAVIOR IN NUnEG 0000 V11 NO2: REPORT TO CONGRESS ON ABNORMAL TIME VARYING THERMAL-HYDRAUUC CONDITIONS. OCCURRENCES. Apol-June 1988. Ag6ng Abstract NUREG-0304 V12 N04 REGULATORY AND TECHNICAL REPORTS NUREG/CR-4597 V02: AGING AND SERVICE WEAR OF AUXILIARY FEEDWATER PUMPS FOR PWR NUCLEAR PLANTS. Volume 2. Aging NU 30 01 REGU O N H C EPORTS Assessments And Monitonng men haluations. (ABSTRACT INDEX JOURNAL) Compilation For First Quarter NUREC/CR-4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER Nt G3 V 02: REGULATORY AND TECHNICAL REPORTS PLANTS. Operating Exponence And Failure identification. (ABSTRACT INDEX JOURNAL) Compilation For Second Quarter NUREG/CR-5051: DETECTING AND MITIGATING BATTERY CHARGER i D08. April June AND INVERTER AGING. NUREG 0304 V13 NO3 REGULATORY AND TECHNICAL REPORTS P'UREG/CR-5141: AGING AND QUALIFICATION RESEARCH ON SOLE-(A S IND JOURNAL) Compilation For Third Quarter N D P RAT D L ES CLE ANING OF LWR COMPONENTS AND POSSIBLE INTERACTIONS Accelerator WITH MET ALLURGICAL AGING EFFECTS. NUREG 1310: NATURALLY OCCURRING AND ACCELERATOR-PRO. NUREG/CR-5192: TESTING OF A NATURALLY AGED NUCLEAR DUCED RADIOACTIVE MATERIALS 1987 Review. POWER PLANT INVERTER AND BATTERY CHARG5R. 85

1 1 1 86 Subject index Aging Degradation B-Factor Technique NUREG/CR-5053: OPERATING EXPERIENCE AND AGING ASSESS- NUREG/CR-5044: ESTIMATION TECHNIQUES FOR COMMON CAUSE MENT OF MOTOR CONTROL CENTERS- FAILURE EVENTS. Agreement States BWR NUREG 1309: THE U.S. NUCLEAR REGULATORY COMMISSION PRO- NUREG 0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION GRAM WITH STATE AND LOCAL GOVERNMENTS AND INDIAN AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE TRIBES BOUNDARY PIPING. Final Report. NUREG-1311: FUNDING THE NRC TRAINING PROGRAM FOR STATE.S. . NUREG 1217 DRFT FC: EVALUATION OF SAFETY IMPLICATIONS OF CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS.Techrwcal Airborne Findings Related To Unresolved Safety issue A 47. Draft Report For NUREG/CR 4997: METHODS FOR DESCRIBING AIRBORNE FRAC- Comment. TIONS OF FREE FALL SPILLS OF POWDERS AND LIQUIDS. NUREG/CR 0672 ADD 03: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR Algorithm POWER STATION. Technical Support For Decommissiorung Matters NUREG/CR-5242: A FAST BOTTOM UP ALGORITHM FOR COMPUT- Related To Preparation Of The Final Decommissioning Rule. ING THE CUT SETS OF NONCOHERENT FAULT TREES. NUREG/CR-2000 V07 NB: LICENSEE EVENT REPORT (LER) COMPILATION.For Month Of August 1988. Allowed Outage Time NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT NUREG/CR-2000 V07 N9: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of September 1988. AND STI REQUlHEMENTS AT THE ANO 1 NUCLEAR POWER PLANT. NUREG/CR-4060 RO1: FLAW DENSITY EXAMINATIONS OF A CLAD BOLLING WATER REACTOR PRESSURE VESSEL SEGMENT. Alpha Emitter NUREG/CR-4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE. NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED VENTION AND MITIGATION FEATURES.BWR, MARK I CONTAIN-l ALPHA AND BETA IRRADIATION OF THE LUNG, Phase 11 Report- MENT DESIGN. NUREG/CR 4920 V02: ASSESSMENT OF SEVERE ACCIDENT PRE. UE/ 5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABI- ME T DES GN NETt* NUREG/CR-4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES:BWR, MARK 111 CONTAIN-Analysis t id Evaluation Of Operational Data NUREG .C "02 N01: REPORT TO THE U S. NUCLEAR REGULA. MENT DESIGN. TORY COMMISSION ON ANALYSIS AND EVALUATION OF OPER. NUREG/CR-5076: AN APPROACH TO THE QUANTIFICATION OF SEIS-ATIONAL DATA 1987. Power Reactors. MIC MARGINS IN NUCLEAR POWER PLANTS.The importance Of NUREG 1272 V02 NO2: REPORT TO THE U.S. NUCLEAR REGULA- BWR Plant Systems And Functions To Seismic Margins. TORY COMMISSION ON ANALYSIS AND EVALUATION OF OPER- NUREG/CR-5126: TAC 2D STUDIES OF MARK I CONTAINMENT ATIONAL DATA 1987.Nonreactors. DRYWELL SHELL MELT-THROUGH. NUREG/CR 5157: THE DEVELOPMENT OF APRIL. MOD 2 - A COMPUT-Annual Report ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL-NUREG 0975 V06: COMPILATION OF CONTRACT RESEARCH FOR ING WATER NUCLEAR REACTORS. THE MATERIALS ENGINEERING BRANCH. DIVISION OF NUREG/CR 5160: GUIDELINES FOR THE USE OF THE EEDB AT THE N EE IN yRE 1 y U n""5 $ G TORY COMMISSION 1987 SUB-COMPONENT AND SUBSYSTEM LEVEL. Final Report NUREG/CR 5189. CLOSEOUT OF IE BULLETIN 79-26: BORON LOSS ANNUAL REPORT' FROM BWR CONTROL BLADES. Aquatic Blota NUREG/CR-5190: CLOSEOUT OF IE BULLETIN 8014: DEGRADATION NUREG/CR-5047: RADIONUCLIDES ACCUMULATION BY AQUATIC OF BWR SCRAM DISCHARGE VOLUME CAPABILITY. NUREG/CR 5191: CLOSEOUT OF IE BULLETIN BO-17; FAILURE OF 76 BIOTA EXPOSED TO CONTAMINATED WATER IN ARTIFICIAL ECO. SYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH THE OF 185 CONTROL RODS TO FULLY INSERT DURING A SCRAM AT A BWR. l GROUND. NUREG/CR-5225: AN OVERVIEW OF BWR MARK 4 CONTAINMENT Asiatic Clam VENTING RISK IMPLICATIONS. l l NUREG/CR 52'D: TECHNICAL FINDINOS DOCUMENT FOR GENERIC ISSUE 51:lMPROVING THE RELIABILITY OF OPEN-CYCLE SERVICE- Babcock And Wilcox WATER SYSTEMS- NUREG1231 S01: SAFETY EVALUATION REPORT RELATED TO BAB- , Atmospheric Diffusion GR

                                             '" "        "'           NUREG-1216'  : THERMAL-HYDRAULIC RESEARCH PLAN FOR BAB-HA TABIL Y SSESSMENT                                               COCK AND WILCOX PLANTS.

Attenuation Basatt Waste isolation Project NUREG/CR-5080: A STUDY OF NEW ENGLAND SEISMICITY WITH , EMPHASIS ON MASSACHUSETTS AND NEW HAMPSHIRE, Final NUREG/CR 5107: HYDROGEOLOGIC CHARACTER 12ATION OF 1 Report Covenng The Period 1976-1985. BASALTS.The Northern Rim Of The Columbia Plateau Physiographic j Provmee And Of The Creston Study Area. Eastern Washington. Authigenic Minerale NUREG/CR 5255: ST ABLE ISOTOPES OF AUTHIGENIC MINERALS IN Battery Charger VARIABLY-SATURATED FRACTURED TUFF. NUREG/CR-5051: DETECTING AND MITIGATING BATTE3Y CHARGER i ANDINVERTER AGING. l Auxiliary Feedwater NUREG/CR-5192: TESTING OF A NATURALLY AGED NUCLEAR l NUREG-1332: REGULATORY ANALYSIS FOR THE RESOLUTION OF POWER PLANT INVERTER AND BATTERY CHARGER' GENERIC tSSUE 125.lL7. " REEVALUATE PROVISION TO AUTO-MATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR Bentonite Pellet NUREG/CR 5130: BENTONITE BOREHOLE PLUG FLOW TESTING NU E R 4 47 02 AN AGING FAILUR8E SURVEY OF LIGHT WITH FIVE WATER TYPES. j WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. NUREG/CR-5178. EVALUATION OF GENERIC ISSUE l Beta Emitter j 12511.7. REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED ALPHA AND BETA lRRADIATION OF THE LUNG, Phase 11 Report. Auxillary Feedwater Pump Biodegradation I NUREG/CR 4597 V02: AGING AND SERVICE WEAR OF AUXILIARY  ! FEEDWATER PUMPS FOR PWR NUCLEAR PLANTSVolume 2.Agmg NUREG/CR 5137: BIODEGRADATION TESTING OF TMI-2 EPICOR-Il j Assessments And Monitonng Method Evaluations. WASTE FORMS. l

( l Subject index 87 j B6ological Sample CADET NUREG/CP-0093: PROCEEDINGS OF THE MEETING ON ULTRASEN- NUREG/CR4857: CADET:A DECISION SUPPORT SYSTEM FOR LIGHT SITIVE TECHNIOUES FOR MEASUREMENT OF URANIUM IN BIO. WATER REACTOR SAFETY, LOGICAL SAMPLES AND THE NEPHROTOXICITY OF URANIUM. CHARM Blackout NUREG/CR-5162; CHARM.A MODEL FOR AEROSOL BEHAVIOR IN NUREG/CR.5214: ANALYSES OF NATURAL CIRCULATION DURING A TIME VARYING THERMAL HYDRAULIC CONDITIONS. , SURRY STATION BLACKOUT USING SCDAP/RELAP5. ' COMMIX Code Bolling Water Reactor NUREG/CR 5070: ANALYSIS OF NATURAL CONVECTION PHENJM-NUREG-0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION ENA IN A 3-LOOP PWR DURING A TMLB' TRANSIENT USING THE AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE COMMIX CODE. BOUNDARY PIPING Final Report. NUREG-1217 DRFT FC: EVALUATION OF SAFETY IMPLICATIONS OF CRAC2 CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS. Technical NUREG/CR-5264: GUIDE FOR LICENSING EVALUATIONS USING Findings Related To Urresolved Safety issue A47. Draft Report For CRAC2.A Computer Program for Calculating Reactor Accident Conse-NU EG C 0672 ADD 03: TECHNOLOGY SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR CSNI POWER STATION. Technical Support For Decommissioning Matters NUREG/CP-0075: PROCEEDINGS OF CSN1/NRC WORKSHOP ON i Related To Preparation Of The Final Decommissioning Rule. DUCTILE PIPING FRACTURE MECHANICS. NUREG/CR-2000 V07 NB: LICENSEE EVENT REPORT (LER) NUREG/CP-0089: PROCEEDINGS OF THE CSNI SPECIALIST MEET. COMPILATION:For Month Of August 1988. ING ON TRAINING OF NUCLEAR REACTOR PERSONNEL. Held At NUREG/CR-2000 V07 N9: LICENSEE EVENT REPORT (LER) Orlando,Flonda. Aprit 2124.1987. COMPILATION:For Month Of September 1988. NUREG/CR4860 RO1: FLAW DENSITY EXAMINATIONS OF A CLAD NUREG/CR 5016: COMPENDIUM AND COMPARISON OF INTERNA-TIONAL PRACTICE FOR PLUGGING, REPAIR AND INSPECTION OF BOILING WATER REACTOR PRESSURE VESSEL SEGMENT. NUREG/CR-4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE- STEAM G7NERATOR TUBING. VENTION AND MITIGATION FEATURES.BWR. MARK i CONTAIN- Cabinet Fire NUREG/ 92 V02: ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG/CR4527 V02: AN EXPERIMENTAL INVESTIGATION OF IN-VENTION AND MITIGATION FEATURES:BWR MARK ll CONTAIN- TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CONTROL MENT DESIGN. CAB! NETS.Part itRoom Ettacts Tests. NUREG/CR4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE-Ca NTION AND MITIGATION FEATURES:BWR. MARK lli CONTAIN-RE /CR-4527 V02: AN EXPERIMENTAL INVESTIGATION OF IN-NUREG/CR-5076: AN APPROACH TO THE QUANTIFICATION OF SEIS. TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CONTROL MIC MARGINS IN NUCLEAR POWER PLANTS.The importance Of CABINETS.Part II. Room Effects Tests. BWR Plant Systems And Functions To Seismic Margins. NUREG/CR-5126: TAC 2D STUDIES OF MARK 1 CONTAINMENT Capsule DRYWELL SHELL MELT-THROUGH. NUREG/CR-5019: NEUTRON EXPOSURE PARAMETERS FOR THE NUREG/CR-5157: THE DEVELOPMENT OF APRIL. MOD 2 - A COMPUT- METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY-SEC. ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL- TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES. ING WATER NUCLEAR REACTORS. NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE Catalyt6c Igniter SUB-COMPONENT AND SUBSYSTEM LEVEL. Final Report. NUREG/CR 5086: PLATINUM CATALYTIC IGNITERS FOR LEAN HY-NUREG/CR-5189: CLOSEOUT OF IE BULLETIN 79-26: BORON LOSS DROGEN-AIR MIXTURES. FROM BWR CONTROL BLADES. NUREG/CR-5190: CLOSEOUT OF IE BULLETIN 80-14: DEGRADATION Cement OF BWR SCRAM DISCHARGE VOLUME CAPABILITY. NUREG/CR 5153: THE TEACHABILITY AND MECHANICAL INTEGRITY NUREG/CR-5191: CLOSEOUT OF IE BULLETIN 8017: FAILURE OF 76 OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED OF 185 CONTROL RODS TO FULLY INSERT DURING A SCRAM AT IN CEMENT AND VINYL ESTER-STYRENE. A BWR. NUREG/CR-5225: AN OVERVIEW OF BWR MARK 1 CONTAINMENT Cement Borehole Plug VENTING RISK IMPLICATIONS. NUREG/CR 5120: EXPERIMENTAL ASSESSMENT OF THE INFLUENCE OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF NUREG/ R 5130: BENTONITE BOREHOLE PLUG FLOW TESTING WITH FIVE WATER TYPES. Cement Permeability Borehole ' Sealing NUREG/CR-St20: EXPERIMENTAL ASSESSMENT OF THE INFLUENCE NUREG/CR-5129: EXPERIMENTAL ASSESSMENT OF THE INFLUENCE OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF DRIED CEMENT BOREHOLE SEALS. DRIED CEMENT BOREHOLE SEALS- Centrifugal NUR /R 30 BENTON TE BOREHOLE PLUG FLOW TESTING

                                                ~

NUREG/CR4062. CLOSEOUT OF IE BULLETIN 80-18. MAINTENANCE OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING Boron Loss PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP-NUREG/CR-5189 CLOSEOUT OF IE BULLETIN 79-26. BORON LOSS TURE. FROM BWR CONTROL BLADES. Charcoat Tray Adsorber Cell Branch Connection NUREG/CR 4932: CLOSEOUT OF IE BULLETIN BO-03: LOSS OF CHAR. NUREG/CR 4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS COAL FROM STANDARD TYPE ll.TWO-INCH. TRAY ADSORBER METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND CELLS. BRANCH CONNECTIONS. Charging Pump Brittle Toughness WUREG/CR4662: CLOSEOUT OF IE BULLETIN 80-18 MAINTENANCE NUREG/CR-5142: DUCTILE TO BRITTLE TOUGHNESS TRANSITION OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING CHARACTER 12ATION OF A5338 STEEL. PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP-TURE. Budget NUREG/CR4834 V02: RECOVERY ACTIONS IN PRA FOR THE RISK Charpy Test METHODS INTEGRATION AND EVALUATION PROGRAM NUREG/CP 0064: SECOND CNSI WORKSHOP ON DUCTILE FRAC-(RMIEP) Volume 2.Apphcation Of The Data-Based Method. TURE TEST METHODS. l

88 Subject Index NUREG/CR 4947; ANALYSIS OF THE A302B AND A5330 STANDARD Processing And Revision.Part 2: Human Error Probabikty Data Entry j REFERENCE MATERIALS IN SURVEILLANCE CAPSULES OF COM- And Revision Procedures. MERCIAL POWER REACTORS. NUREG/CR 4639 V03 P3. NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data Check Valve Processing And Revision. Part 3: Hardware Component Failure Data NUREG/CR-5159: PREDICTION OF CHECK VALVE PERFORMANCE Entry And Revision Procedures. AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS- NUREG/CR-4639 V04 P1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR FIELIABILM Y (NUCLARR). User's Guide,Part 1: Chemical Analysis Overview Of NUCLARR Data Retneval NUREG/CR-5009: EVALUATION OF MATERIALS OF CONSTRUCTION NUREG/CR 4639 V04 P2: NUCLEAR COMPUTERIZED LIBRARY FOR FOR THE REINFORCED CONCRETE REACTOR CONTAINMENT ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide.Part 2: MODEL Guide To Operations. Chem 6 cal Cleaning NUREG/CR-4639 V04 P3: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR 5180: CHEMICAL DECONTAMINATION AND CHEMICAL ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide.Part 3:

                                   "       ^                         NUREG/CR 463 VD W TH E A LURG CA G G EF                S' LEAR COMP"TERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 1:

Circuit Breaker Summary Desenption. NUREG/CR-4665: CLOSEOUT OF IE BULLETIN 83-08 ELECTRICAL NUREG/CR 4639 V05 P2: NUCLEAR COMPUTERIZED LIBRARY FOR CIRCUlT BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN ASSESSING FIEACTOR RELIAP.t!TY (NUCLARR). Data Manual,Part 2: USE IN SAFETY RELATED APPLICATIONS OTHER THAN THE RE. Human Error Probabihty (HEP) Estmates. ACTOR TRIP SYSTEM. NUREG/CR-4639 V05 P3: NUCLE AR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIAEalTY (NUCLARR) Data Manual,Part 3: Cladd6ng Hardware Component Failure t'ata (HCFD). NUREG/CR 4828: FATIGUF CRACK GROWTH OF PART THROUGH NUREG/CR 4639 V05 P4: NUCLEAR COMPUTERIZED LIBRARY FOR CRACKS IN PRESSURE VESSEL AND PIPING STEELS.Aer Environ. ASSESSING REACTOR RELIABluTY (NUCLARR). Data Manual. Par

• 4:

ment Results Summary Aggregations j NUREG/CR-5207: FRACTURE EVALUATION OF SURFACE CRACKS NUREG/CR-4993. A STANDARD PROBLEM FOR HECTR-MAAP EMBEDDED IN REACTOil VESOEL 08 ADDING.Matenal Property Eval- COMPARISON. incomplete Burning uaNns. NUREG/CR-5071: TRAC SUPPORT SOFTWARE. NUREG/CR-5194. RFLAPS/ MOD 2 MODELS AND CORRELATIONS. Cleanup eseration NUREG/CR-5233: A COMPUTER CODE FOR FIRE PROTECTION AND WUREG M83 S03 DRFT PROGRAMMATIC ENVIRONMENTAL IMPACT RISK ANALYSIS OF NUCLEAR PLANTS. STATEMNT RELATLD TO DECONTAMINATION AND DISPOSAL OF RADIOAC'lVE V,ASTES RESULTING FROM MARCH 28,1979 Computer Based Model ACCIDENT,REE MILE ISLAND NUCLEAR STATION, UNIT 2. Docket NUREG/CR 5164. A SIMPLIFIED MODEL FOR CALCULATING EARLY No. 50-320.(GPU Nuclear, incorporated) OFFSITE CONSEQUENCES FROM NUCLEAR FIEACTOR ACCl-DENTS. Closecut NUREG/CR-4523 CLOSEOUT OF IE BULLETIN 80-13 CRACKING IN Concrete Interaction CORE SPRAY SPARGERG. NUREG/CR 5196: SUBMtSSION FOR THE CSNI/GREST BENCHMARK NUREG/CR 4662: CLOSEOUT OF IE BULLETIN 8018: MAINTENANCE EXERCISE ON CHEMICAL THERMODYNAMIC MODELING IN CORE-OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING CONCRETE INTERACTION RELEASES OF RADIONUCLIDES. PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP- NUREG/CR-5219: THE MIXING OF IMMISCIBLE LIQUID LAYERS BY TURE. GAS BUBBLING. NUREG/CR 4665: CLOSEOUT OF IE BULLETIN 83 08 ELECTRICAL CIRCUIT BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN Condensation-induced USE IN SAFETY RELATED APPLICATIONS OTHER THAN THE RE- NUREG/CR 5220 V01: DIAGNOSIS OF CONDENSATION-INDUCED ACTOR TRIF SYSTEM WATERHAMMER. Methods And Background. NUREG/CR-4932: CLOSEOUT OF IE BULLETIN 80 03 LOSS OF CHAR- NUREG/CR-5220 V02: DIAGNOSIS OF CONDENSATION-INDUCED COAL FROM STANDARD TYPE ll,TWO-INCH TRAY ADSORBER WATERHAMMER Case Studies. CELLS NUREG/CR-4933: CLOSEOUT OF IE BULLETIN 8019: FAILURES OF Condenser MERCURY-WETTED MATRIX RELAYS IN REACTOR PROTECTIVE NUREG/CR 4312 V02 Ft1: RELAP5/ MOD 2 CODE MANUAL. Volume 2: SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED Users Guide And input Requirements. BY COMBUSTION ENGINEERING. NUREG/CR-4935: CLOSEOUT OF IE BULLETIN 85-02:UNDERVOL- Conductive Layer TAGE TRIP ATTACHMENTS OF WESTINGHOUSE DB-50 TYPE RE- NUREG/CR-4918 V02: CONTROL OF WATER INFILTRATION INTO ACTOR TRIP BRE AKERS. NEAR SURFACE LLW DISPOSAL UNITS. Task Report . A Discussion. NUREG/CR-5189: CLOSEOUT OF IE BULLETIN 79-26 BORON LOSS FROM BWR CONTROL BLADES. Configuration Management NUREG/CR-5190 CLOSEOUT OF IE BULLETIN 80-14 DEGRADATION OF BWR SCFIAM DISCHARGE VOLUME CAPABluTY. NUREG/CR-5147: FUNDAMENTAL ATTRIBUTES OF A PRACTICAL NUREG/CR 5191: CLOSEOUT OF IE BULLETIN 6017. FAILURE OF 76 CONFIGURATION MANAGEMENT PROGRAM FOR NUCLEAR PLANT DESIGN CONTROL-OF 185 CONTROL RODS TO FULLY INSERT DURING A SCRAM AT A BWR. Conservation Equation Combustion Experiment NUREG/CR 5065: TIME AND VOLUME-AVERAGED CONSERVATION NUREG/CR-4916: HECTR ANALYSES OF THE NEVADA TEST SITE EQUATIONS FOR MULTIPHASE FLOW USING MASS-WElGHTED VE. (NTS) PREMIXED COMBUSTION EXPERIMENTS. LOCITY AND INTERNAL ENERGY. i Component Aging construction NUREG/CR 524B: PRIORITIZATION OF TIRGALEX-RECOMMENDED NUREG/CR-5083: DESIGN. CONSTRUCTION AND INSTRUMENTATION COMPONENTS FOR FURTHER AGING RESEARCH. OF A 1/6-SCALE REINFORCED CONCR'ITE CONTAINMENT BUILD-ING. Computer Code WUREG/CR 4639 V01: NUCLEAR COMPUTERIZED LIBRARY FOR AS- Construction Permit SESSING REACTOR RELIAB!LITY (NUCLARR) Volume i Summary De- NUREGICR-5218: FINANCIAL QUALIFICATIONS REVIEW OF APPL 1-senption CANTS FOR NUCLEAR POWER PLANT CONSTRUCTION PERMITS. NUREG/CR 4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR) Guide To Data Containment Processing And Rewsion. Part 1 Technscal Overview NUREG/CR-4242 SURVEY OF LIGHT WATER FIEACTOR CONTAIN-NUREG/CR-4639 V03 P2: NUCLEAR COMPUTERIZED LIBRARY FOR MENT SYSTEMS, DOMINANT F AILURE MODES AND MITIGATION ASSESSING REACTOR RELIABILITY (NdCLARR). Guide To Data OPPORTUNITIES Final Report.

Subject index 89 NUREG/CR-4763: SAFETY-RELATED EQUIPMENT SURVIVAL IN HY- Control Rod DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS. NUREG/CR-5191: CLOSEOUT OF IE BULLETIN 80-17: FAILURE OF 76 NUREG/CR-4881: FtSSION PRODUCT RELEASE CHARACTERISTICS OF 185 CONTROL RODS TO FULLY INSERT DURING A SCRAM AT INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCl- A BWR, DENT CONDITIONS NUREG/CR-5043. CONTAINMENT PENETRATION G ofEM (CPQ Control Room TESTS UNDER ACCIDENT LOADS. US GU O. TACT OMPUTE NUREG/CR 4960. CONTROL ROOM HABITABILITY SURVEY OF Lt-Qj/ p AND CHE CENSED COMMERCIAL NUCP EAR POWER GENERATING STA-REACTION OF MOLTEN DEBRIS IN DIRECT CONTAINMENT HEAT-NU E CR-5233: A COMPUTER CODE FOR FIRE PROTECTION AND N /C -5126: TAC 2D STUDIES OF MARK i CONTAINMENT RISK ANALYSIS OF NUCLEAR PLANTS. DRYWELL SHELL MELT THROUGH NUREG/CR-5157: THE DEVELOPMENT OF APRIL. MOD 2 - A COMPUT- Control Room Habitability ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL, NUREG/CR-5055 ATMOSPHERIC DIFFUSION FOR CONTROL ROOM ING WATER NUCLEAR REACTORS. HABITABILITY ASSESSMENTS-NUREG/CR 5166: ELECTROCHEMICAL EVALUATION OF SOLID STATE PH SENSORS FOR NUCLEAR WASTE CONTAINMENT- Co" NR 1 8 DRFT FC: REGULATORY ANALYSIS FOR PROPOSED Containment Building RESOLUTION OF USl A 47. Safety implications Of Control NUREG/CR-5083. DESIGN. CONSTRUCTION AND INSTRUMENTATION Systems. Draft Rept For Comment. OF A 1/6-SCALE REINFORCED CONCRETE CONTAINMENT BUILD-lNG Cooling water System NUREG/CR-5096: EVALUATION OF SEALS FOR MECHANICAL PENE- NUREG/CH-5052: OPERATING EXPERIENCE AND AGING ASSESS-TRATIONS OF CONTAINMENT BUILDINGS. MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR- '5 Conta6nment Design 12ED WATER REACTORS. NUREG/CR-4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE- Core Damage VENTION AND MITIGATION FEATURES.BWR MARK I CONTAIN- NUREG 1032: EVALUATION OF STATION BLACKOUT ACCIDENTS AT NR / 42 V02: ASSESSMENT OF SEVERE ACCIDENT PRE- o ed Sa et sueA4 Fa po VENTION AND MITIGATION FEATURES BWR. MARK 11 CONTAIN-MENT DEStGN. NUREG/CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR.4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE- DAMAGE ACCIDENTS 1986 A STATUS REPORT VENTION AND MITIGATION FEATURES.BWR. MARK lli CONTAIN- NUREG/C&4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1986,A STATUS REPORT. NU E / 92 V03 ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG/CR-5042 S01: EVALUATION OF EXTERNAL HA2ARDS TO NU-VEN 10 P 'N T E E c Hazard. MITIGATION FEATURES BWR. MARK lli CONTAIN. NUR A5 AC D G R bR NUREG/CR 4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE- Core Damage Event VENTION AND MITIGAT40N FEAlVRES.PWR.LARGE DRY CON-NUREG/CR 4834 V02: RECOVERY ACTIONS IN PRA FOR THE RISK N1 G A 20 0 ASSESSMENT OF SEVERE ACCIDENT PRE- METHODS INTEGRATION AND EVALUATION PROGRAM VENTION AND MITIGATION FEATURES PWR,lCE CONDENSER (RMIEP). Volume 2: Application Of The Data Based Method. CONTAINMENT DESIGN Core DeMs Containment Environment NUREG/CR 5029: MELT PROGRESSION IN SEVERELY DAMAGED RE-NUREG/CR-5038 OPTIMl2ATION OF THE CONTROL OF CONTAMINA. ACTOR CORES. TION AT NUCLEAR POWER PLANTS. NUREG/CR-5109: RELOCATION OF METALLIC CONSTITUENTS IN CORE DEBRIS BEDS. Containment integrity NUREG 1273: TECHNICAL FINDINGS AND REGULATORY ANALYSIS Core Reflooding FOR GENERIC SAFETY ISSUE ILE da " CONTAINMENT INTEGRITY NURE G/CR-5171: FLOW VISUAll2ATION STUDY OF POST CRITICAL CHECK " HEAT FLUX REGION FOR INVERTED BUBBLY, SLUG AND ANNULAR NUREG/CP 0095 PROCEEDINGS OF THE FOURTH WORKSHOP ON FLOW REGIMES. CONTAINMENT INTEGRITY. Core Spray Sparger Containment Structure NUREG/CR 4523: CLOSEOUT OF IE BULLETIN 80-13 CRACKING IN NUREG/CR 4914: THE INFLUENCE OF SELECTED CONTAINMENT CORE SPRAY SPARGERS. STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW-ING HIGH PRESSURE MELT EJECTION FROM THE REACTOR Corrosion VESSEL. NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND Containment Vessel FLOW VISUAll2ATION STUDIES AND TECHNIQUES RELEVANT TO NUREG/CR 5142: DUCTILE TO BRITTLE TOUGHNESS TRANSITION THE STUDY OF EROSION CORROSION OF REACTOR PIPING SYS-TEMS. CHARACTER 12ATION OF A5338 STEEL NUREG/CR 5149. EROSION-CORROSION OF PWR FEEDWATER Contalement Wall PIPING SURVEY OF EXPERIENCE. DESIGN, WATER CHEMISTRY A RIALS NUREG/CR 5209: DESIGN PROVISIONS FOR TANGENTIAL SHEAR IN CONTAINMENT WALLS-Cost Contaminant Transport Module NUREG/CR 4555 RO1: GENERIC COST ESTIMATES FOR THE DIS-NUREG/CR 5183: A USER'S MANUAL FOR THE CONTAM!NANT POSAL OF RADIOACTIVE WASTES. TRANSPORT MODULE OF THE MiGRAT CODE. Cost Analysis Contamination NUREG/CR-5160. GUIDELINES FOR THE USE OF THE EEDB AT THE NUREG/CR 5145: FAILURE INVESTIGATION OF 3M SERIES 900 SUB-COMPONENT AND SUBSYSTEM LEVELFanal Report. ST ATIC ELIMINATORS. Crack Contamination Contrcl NUREG/CP-0092: PROCEEDINGS OF THE SEMINAR ON LEAK-NUREG/CR 5038 OPTIM12ATION OF THE CONTROL OF CONTAMINA. BEFOREJ1REAK Progress in Regulatory Pohcres And Supporting Re-TION AT NUCLEAR POWER PLANTS. search. Contractor NUREG/CF04813 A01: ASSESSMENT OF LEAK DETECTION SYSTEMS FORLWRS NUREG 0975 V06. COMPILATION OF CONTRACT RESEARCH FOR NUREG/CR 5134 APPLICATION OF ACOUSTIC LEAK DETECTION THE MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEERING. Annual Report For FY 1987. TECHNOLOGY FOR THE DETECTION AND LOCATION OF LEAKS IN LIGHT WATER REACTORS.

90- Subject.index i I Crack Arrest TOR POWER STATION. Technical Support For Decommissioning Mat - l NUREG/CR-4888: PRESSURIZED THERMAL-SHOCK TEST OF 6 INCH ters Related To Preparation Of The Final Decommissioning Rule. i' THICK PRESSURE VESSELS.PTSE-2: Investigation Of Low Teanng Re- NUREG/CR.0672 ADD 03. TECHNOLOGY, SAFETY AND COSTS OF sistance And Warm Prestressing DECOMMISSIONING A REFERENCE BOILING WATER REACTOR  ; NUREG/CR-4996. A REPORT ON THE ROUND ROBIN PROGRAM . POWER STATION. Technical Support For Decommissioning Matters ,j CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD Related To Preparation Of The Final Decommissiorung Rule. TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK NUREG/CR-3899 S01: UTILITY FINANCIAL STABILtTY AND THE ARREST FRACTURE TOUGHNESS.K(IA).0F FERRITIC MATERIALS. AVAILABILITY OF FUNDS FOR DECOMMISSIONING.An Analysis Of l

                                                                                                                                                                                           )

Internal And Extemal Fundrng. Crack Growth NUREG/CR-4315 V09 R1: EVALUATION OF NUCLEAR FACILITY DE. l NUREG/CR4867 V04: ENVIRONMENTALLY ASS'STED CRACKING IN COMMISSIONING PROJECTS. Summary Status ReportThree Mne . LIGHT WATER REACTORS. Semiannual Report,0ctober 1986 March Island Unit 2, Radioactive Waste And Laundry Shipments.' l1 1987. i NUREG/CR-5020: A

SUMMARY

OF ENVIRONMENTALLY ASSISTED Decontamination  ; CRACK GROWTH STUDIES PERFORMED AT WESTINGHOUSE NUREG 0683 S03 DRFT: PROGRAMMATIC ENVIRONMENTAL IMPACT i ELECTRIC CORPORATION.Under Funding From The Heavy-Section STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL OF Steel Technology Program. RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 ACCIDENT,THREE MILE ISLAND NUCLEAR STATION. UNIT 2. Docket N R G/C 4082 V08: DEGRADED PIPING PROGRAM . PHASE 11 Smth NURE C 55 TH H L AND MECHANICAL INTEGRITY ' Program Report. October 1986. September 1987, OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED Cracking IN CEMENT AND VINYL ESTER STYRENE. j NUREG/CR 4523: CLOSEOUT OF IE BULLETIN 60-13: CRACKING IN Decontaminetton Weste NU / 5 4 A T EMISSION SYSTEM CALIBRATION AT NUREG/CR-3444 V05: THE IMPACT OF LWR DECONTAMINATION ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU-WATTS BAR UNIT 1 NUCLEAR REACTOR. PATIONAL EXPOSURE. Annual Report, FY 1987. DOE NUREG/CR 4735 V03. EVALUATION AND COMPILATION OF DOE Degrade %n WASTE PACKAGE TEST DATA. Biannual Report February July 1987. NUREG/CR-4597 V02: AG!NG AND SERVICE WEAR OF AUXILIARY - NUREG/CR-4735 V04: EVALUATION AND COMPILATION Of: DOE FEEDWATER PUMPS FOR PWR NJCLEAR PLANTS. Volume 2. Aging WASTE PACKAGE TEST DATA Biannual Report: August 1987 - Janu. Assessments And Monitonna Method Evaluations. ary 1988. NUREG/CR 4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER ' DOTSYN PLANTS. Operating Exponence And Failure identification. NUREG/CR-4984: DEVELOPMENT OF A THREE DIMENSIONAL FLUX NUREG/CR 5159: PREDICTION OF CHECK VALVE PERFORMANCE SYNTHESIS PROGRAM AND COMPARISON WITH 3 D TRANSPORT AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. THEORY RESULTS. Degraded Pipin0 Damping NUREG/CR-4082 V06: DEGRADED PlPING PROGRAM PHASE ILSmth NUREG/CR-5105: RESPONSE MARGINS INVESTIGATION OF PIPING ProDram Report. October 1986 - September 1987. DYNAMIC ANALYSES USING THE INDEPENDENT SUPPORT MOTION METHOD AND PVRC DAMPING. Design Control NUREG/CR-5154: EXPERIMENTAL ASSESSMENT OF DAMPING IN NUREG/CR-St47: FUNDAMENTAL ATTRIBUTES OF A PRACTICAL'. LOW ASPECT RATIO, REINFORCED CONCRETE SHEAR WALL CONFIGURATION MANAGEMENT PROGRAM FOR NUCLEAR PLANT STRUCTURES. DESIGN CONTROL Debris Diesel Generator Reliability NUREG/CR-4914: THE INFLUENCE OF SELECTED CONTAINMENT NUREG/CR-5078 V01: A RELIABILITY PROGRAM FOR EMERGENCY STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW- DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Program ING HIGH PRESSURE MELT EJECTION FROM THE REACTOR Structure. VESSEL NUREG/CR-5078 V02: A REUABILITY PROGRAM FOR EMERGENCY DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Maintenance, Surveillance And Condition Monitonng. N RE /CR4625: THE POSTlRRADIATION EXAMINATION OF THE DC MELT DYNAMICS EXPERIMENTS- Dlttering Professional Opinions NUREG 1290 ADD DIFFERING PROFESSIONAL OPINIONS.1987 Spe-p,g gp,,,,, ciel Review Panet NUREG/CR-5146: DEBRIS DISPERSAL FROM REACTOR CAVITIES DURING HIGH. PRESSURE MELT EJECTION ACCIDENT SCENAR. Digest IOS. NUREG 0386 D04 R07: UNITED STATES NUCLEAR REGULATORY Decay Heat Removal COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July 2 reh 87 NUREG-1032: EVALUATION OF STATION BLACKOUT ACCIDENTS AT COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July ov n et a ue A 4 F opo

                                                                               ' 1972 - June 1987.

NUREG-1289: REGULATORY AND BACKFIT ANALYSIS UNRESOLVED SAFETY ISSUE A 45, SHUTDOWN DECAY HEAT REMOVAL RE. NUREG-0386 D04 R09: UNITED STATES NUCLEAR REGUtJTORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July OUIREMENTS. NUREG/CR 5072: DECAY HEAT REMOVAL USING FEED-AND-BLEED 1972 September 1987. FOR U.S. PRESSURIZED WATER REACTORS. NUREG 0386 004 R10. UNITED STATES NUCLEAR REGULATORY l- COMMISSION STAFF PRACTICE AND . PROCEDURE Decomm6eeloning DIGEST. Commission, Appeal Board And ucensinD Board Decisions, July NUREG-0586: FINAL GENERIC ENVIRONMENTAL IMPACT STATE- 1972 - December 1987. MENT ON DECOMMISSIONING OF NUCLEAR FACILITIES. NUREG 1221:

SUMMARY

, ANALYSIS AND RESPONSE TO PUBLIC D6 rect Conta6nment Heating COMMENTS ON PROPOSED AMENDMENTS TO 10 CFR PARTS NUREG/CR4917: DCH-2:RESULTS FROM THE SECOND EXPERIMENT. 30,40.50.51JO AND 72. DECOMMISSIONING CRITERIA FOR NUCLE. PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. AR FACILITIES. NUREG-1307: F.EPORT ON WASTE BURIAL CHARGES Escalation Of D6screte-Fracture Model Decommissioning Waste Disposal Costs At Low. Level Waste Bunal Fa- NUREG/CR.5240 COMPARATIVE EVALUATION OF SELECTED CON-c611 ties. TINUUM AND DISCRETE-FRACTURE MODELSEmphasis On Disper. NUREG/CR 0130 ADD 04: TECHNOLOGY, SAFETY AND COSTS OF sivity Calculations For Application To Fractured Geologic Media, Cros-DECOMMISSIONING A REFERENCE PRESSURIZED WATER REAC- ton Study Area. Eastern Washington.

r Subject index 91 Despersion Elasto Plastic NUREG/CR5277. THE TENSORIAL NATURE OF EFFECTIVE POROSI- NUREG/CR-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE TY AND LARGE SCALE DISPER$10N COEFFICIENTS. Application To l PREDICTION METHODS FOR PIPING. s The Creston Study Area. Eastern Washington. ] Elastomeric Seal 1 Disturbance NUREG/CR-5096: EVALUATION OF SEALS FOR MECHANICAL PENE- l" NUREG/CR-5159. PREDICTION OF CHECK VALVE PERFORMANCE TRATIONS OF CONTAINMENT BUILDINGS. AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. Elbow Effect Does Assessment NUREG/CR 5159 PREDICTION OF CHECK VALVE PERFORMANCE NUREG/CR 4000 V02: THE MESORAD DOSE ASSESSMENT AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. j MODEL Computer Code. Electr6 cal Cabinet NUREG/CR 5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABl. REG CR 038: OPTIMIZATION OF THE CONTROL OF CONTAMINA-NETS. TION AT NUCLEAR POWER PLANTS. Done Reduction Electrical DeWee NUREG/CR.5158 V01: WORLDWIDE ACTIVITIES ON THE REDUCTION NUREG/CR-5031: SIGNIFICANCE OF IN STRUCTURE GENERATED OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS MOTION IN SEISMIC QUALIFICATION TESTS OF CABINET MOUNT-ED ELECTRICAL DEVLCES. Dry Air Environment NUREG/CR 5018: URANIUM OXIDE 4RON OXIDE MIXED AEROSOL Electrochemistry EXPERIMENTS IN STEAM-AIR ATMOSPHERES NSPP Tests NUREG/CR-5166: ELECTROCHEMICAL EVALUATION OF SOLID 611.612.613 And 631. Data Record Report. STATE PH SENSORS FOR NUCLEAR WASTE CONTAINMENT. Ductile Embrittlement NUREG/CP-0064: SECOND CNSI WORKSHOP ON DUCTILE FRAC- NUREG/CR-5063: DEVELOPMENT OF A MECHANISTIC UNDER-TURE TEST METHODS. STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES- l SURE VESSEL STEELS. Final Report.  ! Dynamic Loading NUREG/CR-5129 EXPERIMENTAL ASSESSMENT OF THE INFLUENCE Emergency Planning OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF NUREG/CR-4874: THE SENSITIVITY OF EVACUATION TIME ESTi-DRIED CEMENT BOREHOLE SEALS MATES TO CHANGES IN INPUT PARAMETERS FOR THE l-DYNEV , NUREG/CR-5142: DUCTILE TO BRITTLE TOUGHNESS TRANSITION COMPUTER CODE.  ! CHARACTERl2ATION OF A533B STEEL NUREG/CR-5212: EMERGENCY ENVIRONMENTAL SAMPLING AND l ANALYSIS FOR RADIOACTIVE MATERIAL F ACILITIES. Dpamic Strain Aging NUREG/CR 5013: FATIGUE LIFE CHARACTERIZATION OF SMOOTH Emergency Preparedness . AND N"1TCHED PIPING STEEL SPECIMENS IN 2BB DEGREES C AIR NUREG-0654 S01 RO1: CRITERIA FOR PREPARATION AND EVALUA- I ENVIRONMENTS. TION OF FIADIOLOGICAL EMERGENCY RESPONSE PLANS AND EEBD PREPAREDNESS IN SUPPORT OF NUCLEAR POWER , NUREG/CH-5160. GUIDELINES FOR THE USE OF THE EEDB AT THE PLANTS Cntena For Utility Offsite Planning And Preparedness. l SUB COMPONENT AND SUBSYSTEM LEVEL. Final Report. NUREG 1140: A REGULATORY ANALYSIS ON EMERGENCY PRE- J PAREMESS FOR FUEL CYCLE AND OTHER RADIOACTIVE MATE- i EPICOR-Il RIAL LICENSEES Finat Report. I NUREG/CR-5137: BIODEGRADATION TESTING OF TMI-2 EPICOR-il NUREG/CR-4000 V02: THE MESORAD DOSE ASSESSMENT WASTE FORMS. MODELComputer Code. NUREG/CR-5229: ANNUAL REPORT OF THE TMI-2 EPICOR il FIESIN/ NUREG/CR-4B73. BENCHMARK STUDY OF THE l-DYNEV EVACU. , LINER INVESTIGATION Low-Level Waste Data Base Development ATION TIME ESTIMATE COMPUTER CODE. J Program For Fiscal Year 1988. Emergency Response Earthquake NUREG/CR 4000 V02. THE MESORAD DOSE ASSESSMENT NUREG/CR-3145 V06 GEOPHYSICAL INVESTIGATIONS OF THE MODEL. Computer Code. WESTERN OHIO-INDIANA REGION Annual Renort October 1986 - September 1987. Emergency Response Plan NURLG/CR-5012: SIMILARITY PRINCIPLES FOR EQUIPMENT OUAll- NUREG 0054 S01 RO1: CRITERIA FOR PREPARATION AND EVALUA-FICATION BY EXPERIENCE. TtON OF RADIOLOGICAL EMERGENCY FIESPONSE PLANS AND NUREG/CR-50B0: A STUDY OF NEW ENGLAND SEISMICITY WITH PREPAREDNESS IN SUPPORT OF NUCLEAR POWER EMPHASIS CN MASSACHUSETTS AND NEW HAMPSHIRE. Final PLANTS.Cntena For Utility Offsite Planning And Preparedness. Fleport Covenng The Penod 1976-1985 NURt.G/UR-5165. SEISMOLOGICAL INVESTIGATION OF EARTH- Enforcement Action OUAKES IN THE' NEW MADRID SEISMIC ZONE AND THE NORTH- NUREG 0940 V06 N04. ENFORCEMENT ACTIONS SIGNIFICANT AC-EASTERN EXTENT OF THE NEW MADRID SE1SMIC ZONE. Final TIONS FIESOLVED Ouarterly Progress Report October December Report. September 1981 December 1986. 1gg7 NUREG/CR 5250 V01: GEORG'A/ ALABAMA REGIONAL SEISMO- NUREG-0040 V07 Not. ENFORCEMENT ACTIONS.SIGNIFICANT AC-GRAPHIC NETWORK. Annual Report. August 1985 June 1986. TIONS RESOLVED Ouarterly Progress Report, January-March 1988. NUREG 0940 V07 NO2- ENFORCEMENT ACTIONS.SIGNIFICANT AC-N /R 811: THE ECONOMIC COSTS OF RADIATION-INDUCED NU EG 0 40 07 03 OR EN A OS Ni ANT AC-HEALTH EFFECTS. Estimation And Simulation. TIONS RESOLVED.Ouarterly Progress Report, July September 1988. Eddy Current Engineered S y Feature Actu g S stem NUFI /CR-5001: THREE-FREOUENCY EDDY CURRENT INSTRU-PROTECTION SYSTEMS. Effluent NUREG/CR-2007 V06 RADIOACTIVE MATERIALS FIELEASED FROM Engineered Safety System NUCLEAR POWER PLANTS Annual Report For 1965. NUREG/CR-4960 CONTROL ROOM HABITABILITY SUh/EY OF L1-NUREG/CR2907 V07. RADIOACTIVE MATERIALS FIELEASED FROM CENSED COMMERCIAL NUCLEAR POWER GENERATING STA-NUCLEAR POWER PLANTS. Annual Report For 1986. TIONS. Elaatsc Entropy NUREG/CR.5023 HIGH-LEVEL SEISMIC RESPONSE AND FAILURE NUREG/CR 4BS4 V01: THERMODYNAMIC TABLES FOR NUCLEAR PREDICTION METHODS FOR PIPING WASTE ISOL/ TION Aqueous Solutions Database

92 Subject index Entry / Emit Control Ertornal Hazard NUREG 1329: ENTRY / EXIT CONTROL AT FUEL FABRICATION FACill- NUREG/CR-$D42 S01: EVALUATION OF EXTERNAL HAZARDS TO NU-TIES L8 SING OR POSSESSING FORMULA QUANTITIES OF STRATE- CLEAR POWER PLANTS IN THE UNITED STATES. Seismic Hazard. GlC SPECIAL NUCLEAR MATERIAL. Flow Oscillat6on Environment NUREG/CFI-5082: SIMULATION EXPERIMENTS ON TWO PHASE NAT-NUREG/CR-5020:' A

SUMMARY

OF ENVIRONMENTALLY ASSISTED URAL CIRCULATION IN A FREON-113 FLOW VISUAll2ATION LOOP. CRACK GROWTH STUDIES PERFORMED AT WESTINGHOUSE ELECTRIC CORPORATION.Under Fundmg From The Heavy Section Failure Steel Technology Program. NUREG/CR4780 V01: PROCEDURES FOR TREATING COMMON NUREG/CR-5170: A REVIEW OF RESEARCH CONDUCTED BY LOS CAUSE FAILURES IN SAFETY AND FIELIABILITY ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHASIS STUDIES. Procedural Framework And Examples. ON THE MAXEY FLATS,KY, SHALLOW LAND WASTE BURIAL SITE. Failure Event Environmentalimpact Statement NUREG/CR-5044: ESTIMATION TECHNIQUES FOR COMMON CAUSE NUREG-0586, FINAL GENERIC ENVIRONMENTAL IMPACT STATE- FAILURE EVENTS, MENT ON DECOMMISSIONING OF NUCLEAR F ACluTIES. l NUREG 0683 S03 DRFT: PROGRAMMATIC ENVIRONMENTAL IMPACT Fat 6gue j STATEMENT RELAED TO DECONTAMINATION AND DISPOSAL OF NUREG/CR-5013: FATIGUE LIFE CHARACTERIZATION OF SMOOTH l RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 AND NOTCHED PIPING STEEL SPECIMENS IN 2BB DEGREES C AIR ACCIDENT.THREE MILE ISLAND NUCLEAR STATION, UNIT 2. Docket ENVIRONMENT S. No. 50-320.(GPU Nuclear, incorporated) NUREG/CR-5020: A

SUMMARY

OF ENVIRONMENTALLY ASSISTED CRACK-GROWTH STUDIES PERFORMED AT WESTINGHOUSE Env6ronmental Monitoring ELECTRIC CORPO. RATION.Under Fundmg From The Heavy-Section NUREG/CR 5054. RECOMMENDATIONS TO THE NRC FOR REVIEW Steel TechnoloDy Program. CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOAC- NUR2G/CR-5136: FATIGUE STRENGTH OF SMOOTH AND NOTCHED TlVE WASTE DISPOSAL Environmental fronitonng And Surveillance SPECIMENS OF ASME SA 106-B STEEL IN PWR ENVIRONMENTS. Programs NUREG/CR-5212: EMERGENCY ENVIRONMENTAL SAMPUNG AND Fat 6gue Crack Growth ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES. NUREG/CR-4828: FATIGUE CRACK GROWTH OF PART THROUGH CRACKS IN PRESSURE VESSEL AND PIPING STEELS. Air Erwron-Environmental Statement ment Results. NUREG/CR-5264: GUIDE FOR UCENSING EVALUATIONS USING CRAC2.A Computer Program For Calculating Reactor Acesdent Conse- Fat 6gue Damage quences. NUREG/CR 5159. PREDICTION OF CHECK VALVE PERFORMANCE p AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. NUREG-1320: NUCLEAR FUEL CYCLE FACluTY ACCIDENT ANALYSIS Fault HANDBOOK. NUREG/CR-5123: STUDIES OF THE PATTERN AND AGES OF POST. Equipment Quellficat6on METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND NORTH CAROLINA. NUREG/CR 4728: EQUIPMENT QUALIFICATION RESEARCH TEST OF A HIGH RANGE RADIATION MONITOR. Fault Tree NUREG/CR 5012: SIMILARITY PRINCIPLES FOR EOUlPMENT OUAll- NUREG/CR 5242: A FAST BOTTOM-UP ALGORITHM FOR COMPUT. NU EG/ 504 TAf IENT PENETRATION SYSTEM (CPS) H S NONCO6N N NS. TESTS UNDER ACCIDENT LOADS. Feed-And-Bleed NUREG/CR-5141: AGING AND QUALIFICATION RESEARCH ON SOLE-NOID OPERATED VALVES' NUREG/CR-5072: DECAY HEAT REMOVAL USING FEED-AND BLEED FOR U.S. PRESSURIZED WATER REACTORS. Equipment Survival Feedwater Line Break NUREG/CR 4763: SAFETY RELATED EQUIPMENT SURVIVAL IN HY' DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS. N REG /CR 4898: RESULTS OF SEMISCALE MOD-PC FEEDWATER AND STEAM LINE BREAK (S-FS) EXPERIMENT SERIES. Bottom Main Erosion Feedwater Line Break Accident Expenments. NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND

                                                                                                 ****'                         P'"8 FLOW VISUAUZATION STUDIES AND TECHNIQUES RELEVANT TO THE STUDY OF EROSION-CORROSION OF REACJOR PIPING SYS-                                     NUREG/CR-5149: EROSION-CORROSION OF PWR FEEDWATER TEMS                                                                                         PIPING SURVEY OF EXPERIENCE. DESIGN, WATER CHEMISTRY NUFIEG/CR 5149: EROSION CORROSION OF PWR FEEDWATER                                             AND MATERIALS.

PIPING SURVEY OF EXPERIENCE. DESIGN, WATER CHEMISTRY AND MATERIALS. Ferritic Material NUREG/CR-4996: A REPORT ON THE ROUND ROBIN PROGRAM Erosion Corrosion CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD NUREG/CR-5156 REVIEW OF EROSION CORROSION IN SINGLE- TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK PHASE FLOWS ARREST FRACTURE TOUGHNESS.K(IA).0F FERRITIC MATERIALS. Estimet6on Technique F6nancial Qualification NUREG/CR-5044: ESTIMAYlON TECHNIOUES FOR COMMON CAUSE NUREG/CR-5218: FINANCIAL QUAUFICATIONS FIEVIEW OF APPLt-F ALLURE EVENTS, CANTS FOR NUCLEAR POWER PLANT CONSTRUCTION PERMITS. Evacustoon Time Eat 6 mate F6re Protection NU6EG/CH-4873 BENCHMARK STUDY OF THE lDYNEV EVACU. NUREG/CR-5233: A COMPUTER CODE FOR FIRE PROTECTION AND l ATION TIME ESTIMATE COMPUTER CODE. RISK ANALYSIS OF NUCLEAR PLANTS ( NUREG/CR-4874: THE SENSITIVITY OF EVACUATION TIME EST1- l MATES TO CHANGES IN INPUT PARAMETERS FOR THE l-DYNEV Fke SaW I COMPUTER CODE. NUREG/CR-4527 V02: AN EXPERIMENTAL INVESTIGATION OF IN- I TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CONTROL Evente CABINETS Part II. Room Effects Tests. NUREG-0525 R14 SAFE GUARDS

SUMMARY

EVENT LIST (SSEL) Fiscal Year Estended Burnup Fuel NUREG/CH-4834 V02. RECOVERY ACTIONS IN PRA FOR THE RISK NUREG/CR 5009 ASSESSMENT OF THE USE OF EXTENDEO METHODS INTEGRATION AND EVALUATION PROGRAM BURNUP FUEL IN LIGHT WATER POWER REACTORS (RMIEP) Volume 2. Application Of The Data-Based Method j

                                                                                                                                                              -]

1 Subject index'- 93 Flee 4on Fractured Rock NUREG/CR-4778. PRELIMINARY STUDIES OF THE MORPHOLOGY OF NUREG/CR-5097: SIMULATION OF LIQUID AND VAPOR MOVEMENT THERMAL GRADIENT TUBE DEPOSITS FROM FISSION PRODUCT IN UNSATURATED FRACTURED ROCK AT .THE APACHE LEAP RELEASE EXPERIMENTS. TUFF S!1E.Models And Strategies NUREG/CR-5132. SEVERE ACCIDENT INSIGHTS REPORT. NUREG/CR 5240- COMPARATIVE EVALUATION OF FELECTED CON-1 F6eo6on Producg TINUUM AND DISCRETE-FRACTURE MODELSEmphasis On Disper- ! NUREG/CR 4777: STEAM OX1DATION OF ZlRCALOY CLADDING IN sivity Calculations For Application To Fractured Geologic Media, Cres-l THE ORNL FISSION PRODUCT RELEASE TESTS. ton Study Area. Eastern Washington. I NUREG/CR 4881: FISSION PRODUCT RELEASE CHARACTERISTICS NUREG/C45277: THE TENSORIAL NATURE OF EFFECTIVE POROSI- ' INTO CONTAINMENT UNDER DFSIGN BASIS AND SEVERE ACCf. TY AND LARGE SCALE DISPERSION COEFFICIENTS.Apphcation To DENT CONDITIONS. The Creston Study Area. Eastern Washington. Fitnese For Duty ' Fractured Tuff . NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREGICR-5007: SIMULATION OF LIQUID AND VAPOR MOVEMENT  : INDUSTRY.A Review Of Technicalissues. IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP ' - TUFF SITE.Models And Strategies. Flow DonenY NUREG/CR-5255 STABLE ISOTOPES OF AUTHIGENIC MINERALS IN l NUREG/CR-4860 Rot: FLAW DENSITY EXAMINATIONS OF A CLAD VARIABLY SATURATED FRACTURED TUFF. BOILING WATER REACTOR PRESSURE VESSEL SEGMENT. Free Fall Sp6fl Fiume ' NUREG/CR-4651 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA NUREG/CR-4997: METHODS FOR DESCRIBING AIRdORNE FRAC-BY RIPRAP TESTING IN FLUMES Phase ll. Followup Investigations. TIONS OF FREE FALL SPILLS OF POWDERS AND LIOUIDS. Foro6pn Maintenance Program Froon-113 L. cop NUFIEG-1333 DRFT FC: MAINTENANCE APPROACHES AND PRAC. NUREG/C45082: SIMULATION EXPERIMENTS ON TWO-PHASE NAT-  ; TICES IN SELECTED FOREIGN NUCLEAR POWER PROGRAWIS AND URAL CIRCULATION IN A FREON 113 FLOW VISUALIZATION LOOP, ' OTHER U.S. INDUSTRIES. Review And Lessons Leamed Draft Report - For Comment. Fuel Assembly NUREG/CR 5009: ASSESSMENT OF THE USE OF EXTENDED EG/C45210: TECHNICAL FINDINGS DOCUMENT FOR GENERIC ISSUE 51: IMPROVING THE RELIABILITY OF OPEN-CYCLE SERVICE- Fuel Cycle WATER SYSTEMS. NUREG/CR-5000: ASSESSMENT OF THE USE OF . EXTENDED BURNUP FUEL IN LIGHT WATER POWER REACTORS. Fracture NUREG/CP 0064: SECOND CNSI WORKSHOP ON DUCTILE FRAC- . NL E / 10 2 O EEDINGS OF THE SEMINAR ON LEAK. UE R-5119 METALLOGRAPHIC EXAMINATION OF THE SEVErt  ; BEFORE-BREAK. Progress in Regulatory Policies And Supportino Re- FUEL DAMAGE SCOPING TEST (SFD ST) FUEL ROD BUNDLE ] search. CROSS SECTIONS. Fracture Mechen6ce Fuel Performance NUREG/CP 0075: PROCEEDINGS OF CSNI/NRC WORKSHOP ON NUREG/CR-3950 V04: FUEL PERFORMANCE ANNUAL REPORT FOR DUCTILE PIPING FRACTURE MECHANICS 1986. NUREG/CR 4082 V06. DEGRADED PIPING PROGRAM PHASE ll. Sixth Program Report October 1986 - September 1987. Fuel Rod Bundle NUREG/C44219 V04 N2: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR-5119. METALLOGRAPHIC EXAMINATION OF THE SEVERE NU E / R 9 V05 AV CT N E TE$ OLOGY FUEL DAMAGE SCOPING TEST (SFD ST) FUEL ROD BUNDLE PROGRAM bemiannual Progress Report For October 1987 March bb 1988 Fuel Stored Energy 1 NUREG/C44888: PRESSURIZEDTHERMAL-SHOCK TEST OF 0-INCH l THICK PRESSURE VESSELS PTSE-2 Investigation Of Low Teanng Re- NUREG/CR 5232: UNCERTAINTIES IN MODELING AND SCALING IN sistance And Warm Prestressing THE PREDICTION OF FUEL STORED ENERGY AND THERMAL RE-NUREG/CR 4996: A REPORT ON THE ROUND ROBIN PROGRAM SPONSE. CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK Fuel Thermal Response ARREST FRACTURE TOUGHNESS,K(IA).0F FERFitTIC MATERIALS. NUREG/C45232: UNCERTAINTIES IN MODELING AND SCALING IN NUREG/CR-5020: A

SUMMARY

OF ENVIRONMENTALLY ASSISTED THE PREDICTION OF FUEL STORED ENERGY AND THERMAL RE-CRACK GROWTH STUDES PERFORMED AT WESTINGHOUSE SPONSE. ELECTRIC CORPORATION Under Funding From The Heavy-Section Steel Technology Program. Gamma Radiation NUREG/CR5198; INHALED (239)PUO(2) AND/OR TOTAL BODY UREG/ H 4 80 VD1: CHARACTERIZATION OF 1RRADIATED CUR-RENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR Gas Bubbling NURE /CR 488 2 C AR TER ON OF IRRADIATED CUS G B ' RENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUCtEAR PRES $URE VESSEL SERVICE. Generic Communication NUREG/CR-4888, PRESSURIZED THERMAL SHOCK TEST .0F 6 INCH NUREG/CR 4991: EVALUATION AND PROPOSED IMPROVEMENTS TO ' THICK PRESSURE VESSELS PTSE-2.Investigat:on Of Low Teanng Re-EFFECTIVENESS OF U.S. NUCLEAR REGULATORY COMMISSION N 96 N THE ROUND ROBIN PROGRAM GENERIC COMMUNICATIONS.  ; CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD i A REST FRA L ET UG NES IAI O RR TIC MA ER A S N G/C 51 AUDATION OF GENERIC COST ESTIMATES FOR l NUREG/CR 5024 TENSILE AND J-R RVE CHARACTERIZATION OF CONSTRUCTION-RELATED ACTIVITIES AT NUCLEAR POWER THERMALLY AGED CAST ST AINLESS STEELS. PLANTS Final Report. NUREG/CR 5201: EXPERIMENTAL ASSESSMENTS OF GUNDREM-MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS Gener6c feeve j STUDIES NUREG 1332: REGULATORY ANALYSIS FOR THE RESOLUTION OF l NUREG/CR-5?07. FRACTURE EVALUATION OF SURFACE CRACKS GENERIC ISSUE 125117. " REEVALUATE PROVISION TO AUTO-EMBEDDED IN REACTOR VESSEL CL ADDING Matenal Property Ed MATICALLY ISOLATE FEEDWATER FHOM STEAM GENERATOR l untions DURING A LINE DRMK."

94 Subject index Gener6c tasue 51 NUREG/CR-5095 V02. THERMODYNAMIC NONEOUILIBRIUWI IN POST. NUREG/CR4210. TECHNICAL FINDINGS DOCUMENT FOR GENERIC CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE Data For Stabb ISSUE $1 IMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE. 14 ed Quench Front Tests WATER SYSTEMS. NUREG/CR-5095 V03: THERMODYNAMIC NONEOUILIBRIUM IN POST-CRITICAL HE AT-FLUX BOILING IN A ROD BUNDLE. Data For Advanc-Gener6c Safety issues eng Quench Front Tests NUREG-0933 S07: A PRIORITIZATION OF GENERIC SAFETY ISSUES NUREG/CR 5095 V04- THERMODYNAMIC NONEOUILlBRIUM IN POST.

    ,     NUREG 0933 SON A PRIORITIZATION OF GENLRiC SAFETY ISSUES                        CRITICAL-HE AT FLUX BOILING IN A FIOD BUNDLE Data For Retreat-NUREG-1273 TECHNICAL FINDINGS AND REGULATORY ANALYSIS                           ang Quench Front Tests FOR GENERIC SAFETY ISSUE Il E 4.3, " CONTAINMENT INTEGRITY CHECK "                                                                  Heavy Sect 6on Steel Technology NUREG/CR 4219 V04 N2: HEAVY-SECTION STEEL TECHNOLOGY N
                                                                                                 ^"       '""        8       "     ^P       b G 1297 P .       EVIEW FOR HIGH-LEVEL NUCLEAR WASTE RE-NU EG/ R 4 9 V05 POSITORIES Genenc Technical Pos; tion.                                                                   AVY ECT N        E   TE    O      Y I                                                                                          PROGRAM Semiannual Progress Report For October 1987 March Y Groundwater                                                                       1000 NUREG 1308- RADIOACTIVE M ATE RIAL IN THE WEST LAKE High Energy Line Rupture NU E      -1  8 RO      4A ACTIVE MATERIAL IN THE WEST LAKE                  NUREG/CR-4662: CLOSEOUT OF IE BULLETIN B0-18 MAINTENANCE LANDFILLSummary Report                                                       OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING NUREG/CR-4706 V02: PROGRESS IN EVALUATION OF RADIONU.                          PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP.

CLIJE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- TURE. LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report For rol 1906 - September 1987. High Pressure inject 6on NU EG/CR 4764. INFLUENCE OF GROUNDWATER ON SOIL STRUC. NUREG/CR 4747 V02- AN AGING F AILURE SURVEY OF LIGHT f TURE INTERACTION WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. NUREG/CR.4879 VO1: DEMONSTRATION OF PERFORMANCE MOD-ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL SITE A High Pressure Melt Election Companson Of Predictive Radionuchde Transport Modeling Versus NUREG/CR 4917: DCH 2 RESULTS FROM THE SECOND EXPERIMENT Facid Ot*ervations At The Nitrate Disposal Pit Site. Chalk River Nuclear PERFORMED IN THE SURTSEY DIRECT HEATING TEST FACILITY. High-Level Nuclear Waste Repository Groundwater Modeling NUREG-1297 PEER REVIEW FOR HIGH-LEVEL NUCLEAR WASTE RE-NUREG-1249 V01 NRC MODEL SIMULATIONS IN SUPPORT OF THE POSITORIES Ger onc Technical Position. HYDROLOGIC CODE INTERCOMPARISON (HYDROCOIN) NUREG 1298- QUALIFICATION Oc EXISTING DATA FOR HIGH LEVEL STUDY. Level 1 - Code Venfication NUCLEAR W ASTE REPOSITORIES Genenc Technical Position. NUREG 1318: TECHNICAL POSITION ON ITEMS AND ACTIVITIES IN Gundremmingen KRD-A Reactor THE HIGH LEVEL WASTE GEOLOGIC REPOSITORY PROGRAM NUREG/CR4201- EXPERIMENTAL ASSESSMENTS OF GUNDRE M SUBJECT TO OUALITY ASSURANU REQUIREMENTS.

  /         MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS                                                                                         .

STUDIES High Level Waste l i NUREG/CR 4735 V03. EVALUATION AND COMPILATION OF DOE HEATING 6 W ASTE PACKAG.: TEST DATA. Biannual Report February-July 1967. j NUREG/CR-4777 STEAM OXIDATION OF ZlRCALOY CLADDING IN NUREG/CR 4735 V04: EVALUATION AND COMPILATION OF DOE THE ORNL FlSSION PRODUCT RELEASE TESTS. W ASTE PACKAGE TEST DATA. Biannual Report. August 1987 - Janu-HECTR "T ~ NUREG/CR 4916 HECTR ANALYSES OF THE NEVADA TEST SITE High-Level Waste Repository e (NT S) PREMNED COMBUSTION EXPERIMENTS. NUREG/CR 4708 V02: PROGRESS IN EVALUATION OF RADIONU. NURE G/CR.4993 A ST ANDARD PROBLEM FOR HECTR MAAP COMPARISON incompiete Burning CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Repo't For HSST Apni1986 September 1987. NURE G/CR-5010 NEUTRON EXPOSURE PARAMETERS FOR THE

                                                                    "     ^

lON EL TECH OGY AI NS lE PS ES F LG/ R-4 2 P NT OU L ATION RESEARCH TEST OF l HYDROCOIN l NUREG 1249 V01. NRC MODEL SIMULATIONS IN SUPPORT OF THE Histodcal Data Summary . HYDROLOGIC CODE INTERCOMPARISON (HYDROCO!N) NUREG-1214 R03. HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-STUDY. Level 1 - Code Venlication IC ASSESSMENT Or LICENSEE PERFORMANCE NUREG-1214 R04. HISTORICAL DAT A

SUMMARY

OF THE SYSTEMAT-Habitability IC ASSESSMENT OF LICENSEE PERFORMANCE. NUREG/CR-4900 CONTROL ROOM HADIT ABILITY SURVEY OF Lt. CENSED COMMERCIAL NUCLEAR POWER GENERATING ST A Hydraulic Conductivity TIONS NUREG/CR-5130: BENTONITE BOREHOLE PLUG FLOW TESTING WITH TlVE WATER TYPES 1 Health Effect NUREG/CR-4B11. THE E CONOMIC COSTS OF RADIATION-INDUCED Hydraulic $4mulation Model HLALTH UFE CTS Estimation And Simulation. NUREG/CR-5097: SIMULATION OF LIQUID AND VAPOR MOVEMENT IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP Heat Capacity TUFF SITE.Models And Strategies. NUREG/CR4864 VD1. THERMODYNAMIC T ABLES FOR NUCLEAR WASTE ISOLATION AQucous Solutions Database. Hydrodynarmc NUREG/CR 5084 IFCI AN INTEGRATED CODE FOR CALCULATION Heat Flux OF ALL PHASES OF FUEL COOLANT INTERACTIONS. NUREG/CR 5171 FLOW VISUALIZATION STUDY OF POST CRITICAL

      ?     HE AT F LUX REGION FOR INVERTED DUBBL Y, SLUG AND ANNULAR                Hydrogen Burn FLOW REGIMES                                                               NUREGICR-4763 SAFETY-RELATED EQUIPMENT SURVIVAL IN HY-DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS NUREG/CR 5095 VOI THE RMODYNAMIC NONEQUILIBRIUM IN POST.                    Hydrogen Event

• CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE Desenption Of NURE G/CR-4916 HECTR ANALYSES OF THE NEVADA TEST SITE Eqenments And Sumpic Reriutts (NTS) PREMlKED COMBUSTION EXPER!MENTS.

Subject Index 95 Hydrogen-Air Mixture Indes NUREG/CR 5086 PLATINUM CATALYTIC IGNITERS FOR LEAN HY- NUREG 0304 V12 N04: REGULATORY AND TECHNICAL REPORTS DROGEN AIR MIXTURES (ABSTRACT INDEX JOURNAL). Annual Compilation For 1987. NUREG-0304 V13 N01: REGULATORY AND TECHNICAL REPORTS Hydrogeologic (ABSTRACT INDEX JOURNAL). Compilacon For First Quarter NUREG/CR-5107. HYDROGEOLOGIC CHARACTERIZATION OF 19BB. January-March. BASALTS The Northern Rim Of The Columbia Plateau Physiographic NUREG-0304 V13 NO2: REGULATORY AND TECHNICAL REPORTS Province And Of The Creston Study Area. Eastern Washington (ABSTRACT INDEX JOURNAL). Compilation For Second Quarter 1988. April-June. Hycrologic NUREG-0304 V13 NO3: REGULATORY AND TECHNICAL REPORTF NUREG 1263 HYDROLOGIC DESIGN FOR RIPRAP ON EMBANKMENT (ABSTRACT INDEX JOURNAL) Compilation For Third Quarter j SLOPE S. 1988. July September. I Hydrologic Baseline Charactertration Indian Tribes NUREG/CR-5107: HYDROGEOLOGIC CHARACTERIZATION OF NUREG-1309 THE U.S NUCLEAR REGULATORY COMMISSION PRO-BASALTSThe Northesn Rim Of The Columbia Plateau Physiographic GRAM WITH STATE AND LOCAL GOVERNMENTS AND INDlAN Province And Of The Creston Study Area, Eastern Washington- TRIBES. Hydrologic Code inelastic NUREG 1249 VOI: NRC MODEL SIMULATIONS IN SUPPORT OF THE NUREGICR-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE HYDROLOGIC CODE INTERCOMPARISON (HYDROCOIN' PREDICTION METHODS FOR PIPING. STUDY. Level 1 Code Venfication. Inhalation NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED NUREG/CR 4873: BENCHMARK STUDY OF THE l-DYNEV EVACV- ALPHA AND DETA IRRADIATION OF THE LUNG Phase ll Report. ATION TIME ESTIMATE COMPUTER CODE. NUREG/CR-4874 THE SENSITIVITY OF EVACUATION TIME ESTI- I" " MATES TO CHANGES IN INPUT PARAMETERS FOR THE l-DYNEV EG R O2 O LING TIME TO RECOVERY AND INITIATING COMPUTER CODE. EVENT FREQUENCY FOR LOSS OF OFFSITE POWER INCIDENTS IE Bulletin 79-26 AT NUCLEAR POWER PLANTS. NUREG/CR 5189 CLOSEOUT OF IE DULLETIN 79-26 DORON LOSS h*' " " FROM BWR CONTROL BLADES. NU C STEAM GENERATOR GROUP PROJECT. Task 13 IE Bulietin 80-03 Fsnal Report. Nondestructive Examination (NDE) Validation. NUREG/CR-4932: CLOSEOUT OF IE BULLETIN 80 03 LOSS OF CHAR-AL FROM STANDARD TYP5 ll,TWO-INCH. TRAY ADSORDER ins NF 0040 VII N04. LICENSEE CONTRACTOR AND VENDOR IN-SPECTION STATUS REPORT- Quarterly Report. October-December IE Bulletin 8013 1987.(White Book) NUREG/CR-4523 CLOSEOUT OF IE BULLETIN 8013 CRACKING IN NUREG-0040 V12 N01: LICENSEE CONTRACTOR AND VENDOR IN-CORE SPRAY SPARGERS. SPECTION STATUS REPORT. Guarterly Report. January-March 1988 (White Book) IE Bulletin 80-14 NUREG 0040 V12 NO2: LICENSEE CONTRACTOR AND VENDOR IN-NUFmG/CR-5190 CLOSEOUT OF IE DULLETIN 80-14: DEGRADATION SPECTION STATUS REPORT- Quartorty Report. April June 1988.(White OF DWR SCRAM DISCHARGE VOLUME CAPADILITY. Book) NUREG-0040 V12 NO3. LICENSEE CONTRACTOR AND VENDOR IN-IE Bulletin 80-17 SPECTION STATUS REPORT. Quarterly Report. July-September NUHEG/CR-5191: CLOSEOUT OF IE DULLETIN 80-17 FAILURE OF 76 1988 (White Book) OF 185 CONTROL RODS TO FULLY INSERT DURING A SCRAM AT NUREG/CR 5058: PRA APPLICATIONS PROGRAM FOR INSPECTION A BWR. AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50 313(Arkansas Power And Light Company) IE Dulletin 8018 NUREG/CR-5151: PERFORMANCE-BASED INSPECTIONS. NUREG/CR 4662: CLOSEOUT OF lE BULLETIN 80-10 MAINTENANCE Or ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING Inspection System PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUP- NUREG/CR-5075 THE SAFT UT REAL TIME INSPECTION SYSTEM - TURE. OPERATIONAL PRINCIPLES AND IMPLEMENTATION. IE Dulletin 80-19 Installation Cost NUREG/CR 4933- CLOSEOUT OF IE BULLETIN 8019F AILURES OF NUREG/CR-5138. VALIDATION OF GENERIC COST ESTIMATES FOR MERCURY-WETTED MATR X RELAYS IN REACTOR PROTECTIVE CONSTRUCTION-RELATED ACTIVITIES AT NUCLEAR POWER SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED PLANTS. Final Report. BY COMUUSTION ENGINEERING. Integrated Plant Safety Assessment IE Dulletin 83-08 NUREG 0822 S01: INTEGRATED PLANT SAFETY ASSESSMENT SYS-NUREG/CR 4605. CLOSEOUT OF IE DULLETIN 83 08 ELECTRICAL TEMATIC EVALUATION PROGRAM OYSTER CREEK NUCLEAR CIRCULI BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN GENERATING STATION. Docket No. 50-219(General Pubisc Utahties USE IN SAFETY RELATED APPLICATIONS OTHER THAN THE RE- Corporation And Jersey Central Power And Light Company) ACTOR TRIP SYSTEM. Intergranular Corrosion IE Bulletm 85-02 NUREG-0313 A02. TECHNICAL REPORT ON MATERIAL SELECTION NUREG/CR 4935. CLOSEOUT OF IE BULLETIN 85 02:UNDERVOL- AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE l TAGE TRIP ATT ACHMENTS OF WESTINGHOUSE DD 50 TYPE RE- BOUNDARY PIPING Final Report ACTOR TRIP DRE AKERS Intergranular Stress Corrosion Crack IFCI NUREG/CR 5133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW NUREG/CR-5084 lFCI AN INTEGRATED CODE FOR CALCULATION THROUGH INTERGRANULAR STRESS CORROSION CRACKS OF ALL PHASES OF FUEL COOLANT INTE R ACTIONS Intergranular Stress Corrosion Cracking I In-Vivo NUREG/CR-5150 STEAM GENERATOR OPERATING NUREG/CR 5223 SCINTILLATION FIBER DETECTOR FOR IN V;VO EXPERIENCE. Update For 1984-1DBb. l l ENDOSCOPIC INTE RNAL DOSIMETRY. internal Dosimetry inc6 dent Investagston Manual NUREG/CR 6223 SCINTILLATION FIDER DETECTOR FOR IN VIVO NURE G 1303 INCIDE NT INVESTIC AllON MANUAL ENDOSCOPIC INTERNAL DOSIMETRY.

96 Subject index Internal Energy NUREG/CR-2907 V07: RADIOACTIVE MATERIALS RELEASED FROM NUFIEG/CR 5065: TIME AND VOLUME AVERAGED CONSERVATION NUCLEAR POWER PLANTS. Annual Report For 1986. EQUATIONS FOR MULTIPHASE FLOW USING MASS-WElGHTED VE. NUREG/CR-3444 V05; THE IMPACT OF LWR DECONTAMINATION LOCtTY AND INTERNAL ENERGY. ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU-PATIONAL EXPOSURE. Annual Report. FY 1987. Intrusion Alarm NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA-NUREG-1328: USE OF PERIMETER ALARMS AT FUEL FABRICATION TION SYSTEMS. FACILITIES USING OR POSSESSING FORMULA OUANTITIES OF NUREG/CR4242: SURVEY OF LIGHT WATER REACTOR CONTAIN-STRATEGIC SPECIAL NUCLEAR MATER &AL MENT SYSTEMS, DOMINANT FAILURE MODES AND MITIGATION OPPORTUNITIES Final Report. Inventory Difference Data NUREG/CR4243. VALUE/ IMPACT ANALYSIS FOR EVALUATING AL-NUREG 0430 V08 NO1: LICENSED FUEL FACILITY STATUS TERNATIVE MITIGATION SYSTEMS. REPORT. Inventory Difference Data January-June 1987.(Gray Book 11) NUREG/CR 4244. STRATEGIES FOR IMPLEMENTING A MITIGATION NUREG-0430 V08 NO2: LICENSED FUEL FACILITY STATUS POUCY FOR llGHT WATER REACTORS. REPORT. inventory Difference Data. July-December 1987.(Gray Book 11) NUREG/CR-4667 V04: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Senuannual Report, October 1986 - March inverter 1987. NUREG/CR-5051: DETECTING AND MITIGATING BATTERY CHARGER NUREG/CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE ANO INVERTER AGING. DAMAGE ACCIDENTS:1986,A STATUS REPORT. NUREG/CR 5192: TESTING OF A NATURALLY AGED NUCLEAR NUREG/CR-4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE POWER PLANT INVERTER AND BATTERY CHARGER. DAMAGE ACCIDENTS 1986.A STATUS REPORT. NUREG/CR 4688 V02: OUANTIFICATION AND UNCERTAINTY ANALY-Irradiated Reactor Fuel StS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT NUREG-0725 R06. PUBLIC INFORMATION CIRCULAR FOR SHlP- WATc.R REACTORS (OVASAR).Part IL Sensitwity Analysis Tech-MENTS OF IRRADIATED REACTOR FUEL rvques. Irradiation NUREGICR-4747 V02: AN AGING FAILURE SURVEY OF LIGHT WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. NUREG/CR 5019: NEUTRON EXPOSURE PARAMETERS FOR THE NUREG/CR-4813 RO1: ASSESSMENT OF LEAK DETECTION SYSTEMS 4 METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SEC* FOR LWRS. TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES  ! NUREGICR 4857: CADET;A DECISION w PORT SYSTEM FOR LIGHT l NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COMBINED WATER REACTOR SAFETY. i ALPHA AND BETA IRRADIATION OF THE LUNG. Phase il Report NUREG/CR-4881: FISSION PRODUCT RELEASE CHARACTERISTICS J-R Curve INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCI-NUREG/CR-5024: TENSILE AND J-R CURVE CHARACTERIZATION OF DENT CONDITIONS. THERMALLY AGED CAST ST AINLESS STEELS- NUREG/CR-5000: ASSESSMENT OF THE USE OF EXTENDED BURNUP FUEL IN LIGHT WATER POWER REACTORS. Jet Dis 6ntegrat6on NUREG/CR 5043: CONTAINMENT PENETRATION SYSTEM (CPS) NUREG/CR-5171: FLOW VISUAll2ATION STUDY OF POST CRITICAL HEAT FLUX REGION FOR INVERTED BUBBLY, SLUG AND ANNULAR TESTS UNDER NUREGICR-5061: ACCIDENT THREE-FREQUE LOADS'NCY EDDY CURRENT INSTRU-FLOW REGIMES- MENT. NUREG/CR-5083: DESIGN, CONSTRUCTION AND INSTRUMENTATION LER OF A 1/6-SCALE REINFORCED CONCRETE CONTAINMENT BUILD-ING. NUREGICR-2000 V06N12. LICENSEE EVENT REPORT (LER) COMPILATION For Month Of December 1987, NUREG/CR-5084: IFCI: AN INTEGRATED CODE FOR CALCULATION NUREG/CR-2000 V07 N1: LICENSEE EVENT REPORT (LER) OF ALL PHASES OF FUEL COOLANT INTERACTIONS. COMPILATION For Month Of January 1988. NUREG/CR-5109: RELOCATION OF METALLIC CONSTITUENTS IN NUREG/CR-2000 V07 N2: LICENSEE EVENT REPORT (LER) CORE DEBRIS BEDS. COMPILATION For Month Of February 1988. NUREG/CR-5134. APPLICATION OF ACOUSTIC LEAK DETECTION NUREG/CR 2000 V07 N3. LICENS12E EVENT REPORT (LER) TECHNOLOGY FOR THE DETECTION AND LOCATION OF LLAKS IN COMPILATION For Month Of March 1988. LIGHT WATER REACTORS. NUREG/CR-2000 V07 N4: LICENSEE EVENT REPORT (LER) NUREG/CR-5180: CHEMICAL DECONTAMINATION AND CHEMICAL COMPILATION For Month Of Apnl 1988 CLEANING OF LWR COMPONENTS AND POSSIBLE INTERACTIONS NUREG/CR 2000 V07 N5: LICENSEE EVENT REPORT (LER) WITH METALLURGICAL AGING EFFECTS. COMPILATION For Montt Of May 1988 NUREG/CR-2000 V07 N6: LICENSEE EVEN' REPORT (LER) Labor Productivity COMPILATION.For Month Of June 1988. NUREG/CR 5138: VALIDATION OF GENERIC COST ESTIMATES FOR NUREG/CR-2000 V07 N7. LICENSEE EVENT REPORT (LER) CONSTRUCTION-RELATED ACTIVITIES AT NUCLEAR POWER i< COMPILATION For Month Of July 1988. PLANTS. Final Report. ' NUREG/CR-2000 V07 N8 LICENSEE EVENT REPORT (LER) ~ COMPILATION For Month Of August 1988 Leak Detection NUREG/CR-2000 V07 N9 LICENSEE EVENT REPORT (LER) NUREG/CR 4513 P91: ASSESSMENT 'JF LEAK DETECTION SYSTEMS COMPIL ATION for Month Of September 1988 FOR LWRS. NUREG/CR-2000 V07N10: LICENSEE EVENT REPORT (LER) COMPILATION For Month Of October 1988. Leak Rate NUREG/CR-2000 V07N11: LICENSEE EVENT REPORT (LER) NUREG/CP-0092: PROCEEDINGS OF THE SEMINAR ON LEAK-COMPILATION For Month Of November 1988. BEFORE BREAK Progress in Regulatory Pohcies Antt Supporting Re-search. NUREG/CR-4728: EQUIPMENT QUALIFICATION RESEARCH TEST OF Leak-Before-Break A HIGH-RANGE RADIATION MONITOR _ ] NUREG/CP-0092. PROCEEDINGS OF THE SEMINAR ON LEAK- = LWR BEFORE-BREAK. Progress in Regulatory Policies And Supporting Re- l search. 1 NUREG-1217 DRFT FC: EVALUATION OF SAFETY IMPLICATIONS OF i CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS. Technical Legal issuances Findtngs Related To Unresolved Safety issue A-47. Draft Report For j NUREG 0750 V25102: INDEXES TO NUCLEAR REGULATORY COM-Comment. MISSION ISSUANCES. January-June 1087. NURIG-121B DRFT FC: REGULATORY ANALYSIS FOR PROPOSED NUREG-0750 V25 NOS: NUCLEAR REGULATORY COMMISSION IS-RESOLUTION OF US) A-4 7. Safety Imphcations Of Control SUANCES FOR MAY 1987.Pages 417-873 Systems Draft Rept For Comment. NUREG-0750 V25 N0ft NUCLEAR REGULATORY COMMISSION IS-NUI'IEG-1289 REGULATORY AND BACKFIT ANALYSIS. UNRESOLVED SUANCES FOR JUNE 1987.Papes 875-997. SAFETY ISSUE A-45, SHUTDOWN DECAY HEAT REMOVAL RE- l NUREG-0750 V26101: INDEXES TO flVCLEAR REGULATCiRY COM-OUIREMENTS MISSION ISSUANCES. July-September 1987.  ! NUREG/CR-2907 V06 RADIOACTIVE MATERIALS RELEASED FROM NUREG 0750 V26102: INDEXES TO NUCLEAR REGULATORY COM-NUCLEAR POWER PLANTS Annual Repor1 For 1985. MISSION ISSUANCES. July December 1987.  ! i i

o l Subject Index 97 NUREG 0760 V26 N01: NUCLEAR FIEGULATORY COMMISSION IS. NUREG/CR-2000 V07 N2: UCENSEE EVENT REPORT (LER) SUANCES FOR JULY 1987.Pages 170. COMPILATION.For Month Of February 1988. NUREG 0750 V26 NO2: NUCLEAR REGULATORY COMMISSION IS- NUREG/CR-2000 V07 N3: UCENSEE E' VENT REPORT (LER) l SUANCES FOR AUGUST 1987. Pages 71-107. COMPILATION;For Month Of March 1988. NUREG-0750 V26 NO3: NUCLEAR REGULATORY COMMISSION IS- NUREG/CR2000 V07 N4: LICENSEE EVENT REPORT (LER) SUANCES FOR SEPTEMBER 1987. Pages 109-248. COMPILATION.For Month Of April 1988. NUREG 0750 V26 N04: NUCLEAR REGULATORY COMMISSION IS- NU'1EG/CR-2000 V07 N5: UCENSEE EVENT FIEPORT (LER) SUANCES FOR OCTOBER 1987. Pages 249 381. COMPILATION For Month Of May 1988. NUREG 0750 V26 N05: NUCLEAR REGULATORY COMMISSION IS* NUREG/CR-2000 V07 N6. UCENSEE EVENT REPORT (LER) SUANCES FOR NOVEMBER 1987. Paper 383-447. COMPILATION:For Month Of June 1988. NUREG4750 V26 NDO: NUCLEAR REGULATORY COMMISSION IS- NUREG/CR-2000 V07 N7; LICENSEE EVENT REPORT (LER) SUANCES FOR DECEMBER 1987. Pages 4G536. COMPILATION:For Month Of July 1988 NUREG-0750 V27101: INDEXES TO NUCLEAR REGULATORY COM-MISSION ISSUANCES. January March 1988- NUREG/CR-2000 V07 NB: UCENSEE EVENT REPORT (LER) NUREG-0750 V27102. INDEXES TO NUCLEAR REGULATORY COM- COMPILATION.For NUREG/CR-2000 V07 Month Of August 1988' EVENT N9: UCENSEE REPORT (LER) Nt 50 V27  : U$ AR RE ULATORY COMMISSION IS- NU EG/ 0 V7 10: N E ENT REPORT (LER) NURE - 0 7 NO2 NU AR G TORY COMMISSION IS-SUANCES FOR FEBRUARY 1988. Pages 41-255. NU / 0 V 7N11: 1 SEE EVENT REPORT (LER) NUREG 0750 V27 NO3: NUCLEAR REGULATORY COMMISSION IS. COMPILATION.For Month Of November 1988. SUANCES FOR MARCH 1988. Pages 257 334-NUREG 0750 V27 N04; NUCLEAR REGULATORY COMMISSION IS- Ucensee Performance NUREG 1214 R03: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-SUANCES NUREG-0750 V27FOR N05: APRIL NUCLEA1988. Pa%a 335 REGULATORY 483. IS. COMMISSION IC ASSESSMENT OF LICENSEE PERFORMANCE. SUANCES FOR MAY 1988. Pages 485-626. NUREG-1214 R04: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT. i NUREG-0750 V27 N06: NUCLEAR REGULATORY COMMISSION IS- IC ASSESSMENT OF UCENSEE PERFORMANCE. 1 SUANCES FO9 JUNE 1988. Pages 627-665. NUREG-0750 V28 N01: NUCLEAR REGULATORY COMMISSION IS. Ught Water Reactor l SUANCES FOR JULY 1988 Pages 1-71. NUREG/CR-4244: STRATEGIES FOR IMPLEMENT'NG A MITIGATION l NUREG 0750 V2B NO2: NUCLEAR REGULATORY COMMISSION IS- POUCY FOR UGHT WATER REACTORS. SUANCES FOR AUGUST 1988. Pages 73-269. NUREG 1217 DRFT FC: EVALUATION OF SAFETY IMPLICATIONS OF NUREG 0750 V28 NO3- NUCLEAR REGULATORY COMMISSION IS- CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS. Technical SUANCES FO9 SEPTEMBER 1988 Pages 271-417. Findings Related To Unresolved Safety issue A-47. Draft Report For NUREG 0750 V28 ND4: NUCLEAR REGULATORY COMMISSION IS- Comment. SUANCES FOR OCTOBER 1988. Pagen 419-497- NUREG-1218 DRFT FC: REGULATORY ANALYSIS FOR PROPOSED Ucense Application RESOLUTION OF USl A-47. Safety Imphcations Of Control Systems. Draft Rept For Comment NUREG-1199 RO1: STANDARD FORMAT AND CONTENT OF A U* NUREG-1289: F.EGULATORY AND BACKFIT ANALYSIS: UNRESOLVED CENSE APPUCATION FOR A LOW-LEVEL RADIOACTIVE WASTE SAFETY ISSUE A-45, SHUTDOWN DECAY HEAT REMOVAL RE-OlSPOSAL FACluTY. QUIREMENTS NUREG 1200 RO1: STANDARD REVIEW PLAN FOR THE REVIEW OF A UCENSE APPLICATION FOR A LOW-LEVEL RADIOACTIVE WASTE NUREG/CR-2907 V06. RADIOACTIVE MATERIALS RELEASED FROM DISPOSAL FACILITY, NL CLEAR POWER PLANTS. Annual Report For 1985. NUHEG/CR-2907 V07: RADIOACTIVE MATERIALS RELEASED FROM Ucense Renewal NUCLEAR POWER PLANTS. Annual Report For 1986. NUREG 1317 DRFT FC: REGULATORY OPTIONS FOR NUCLEAR NUREG/CR-3444 VOS: THE IMPACT OF LWR DECONTAMINARONS PLANT LICENSE RENEWAL. Draft For Comment. ON SOUDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU-PATIONAL EXPOSURE. Annual Report. FY 1987. Ucensed Fuel facility NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA-NUREG 0430 V08 N01: LICENSED FUEL FACluTY STATUS TION SYSTEMS. REPORT. inventory Difference Data Januarydune 19871 Gray Book 11) NUREG/CR-4242: SURVEY OF UGHT WATER REACTOR CONTAIN-NUREG4430 V08 NO2 UCENSED FUEL FACILITY STATUS MENT SYSTEMS, DOMINANT FAILURE MODES AND MITIGATION REPORT. inventory Difference Data. July December 1987.(Gray Book ll3 OPPORTUNITIES.Fmal Report. NUREG/CR-4243. VALUE/lWACT ANALYSIS FOR EVALUATING AL-Licensed Operatir,g Reactors TERNATIVE MITIGATION SYSTEMS. NUREG-UO20 VII N12: LICENSED OPERATING REACTORS STATUS NUREG/CR-4667 V04: ENVIRONMENTALLY ASSISTED CRACKING IN

SUMMARY

REPORT. Data As Of November 30.1987.(Gray Book I) UGHT WATER REACTORS. Semiannual Report. October 1960 March NUREG-0020 V12 N01: LICENSED OPERATING REACTORS STATUS 1987, SUOMARY REPORT. Data As Of December 31,1987.(Gray Book 1) NUREG/CR 4674 V05 PRECURSORS TO POTENTIAL SEVERE CORE NUREG-0020 V12 NO2: UCENSED OPERATING REACTORS STATUS DAMAGE ACCIDENTS 1986,A STATUS REPORT.

SUMMARY

REPORT. Data As Of January 31,1988 (Gray Book l) NUREG/CR 4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE NUREG 0020 V12 NO3: UCENSED OPERATING REACTORS STATUS DAMAGE ACCIDENTS:1986.A GTATUS REPORT.

SUMMARY

REPORT. Data As Of February 29.1988 (Gray Dook 1) NUREG/CR-46BB V02: QUANTIFICATION AND UNCERTAINTY ANALY-NUREG-0020 V12 N04: LICENSED OPERATING REACTORS STATUS SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT

SUMMARY

REPORT. Data As Of March 31,1988.(Gray Book 1) WATER REACTORS (OVASAR).Part II: Sensitmty Analysis Tech- l NUREG-0020 V12 N05: UCENSED OPERATING REACTORS STATUS ' NURE G 002 V2N I SDO AC S STATUS W2: AN A M W RE SU N & M WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. NUREG0 V1 NO ICE D C PE AT kAC S STATUS NUREG/CR-4013 R01: ASSESSMENT OF LEAK DETECTION SYSTEMS NUREG 2 1 NOB ICEt SE OPE INC AC S STATUS NUREG H B57: CADET.A DECISION SUPPORT SYSTEM FOR LIGHT

SUMMARY

REPORT. Data As Of July 31,1988.(Gray Dook I) WATER REACTOR SAFETY. ' NUREG 0020 V12 NC9: UCENSED OPERATING REACTORS STATUS NUREG/C44BB1: FISSION PRODUCT RELEASE CHARACTERISTICS

SUMMARY

REPORT. Data As Of August 31,1988.(Gray Book f) INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCl-NUREG 0020 V12 N10: UCENSED OPERATING REACTORS STATUS DFNT CONDITIONS. i

SUMMARY

REPORT. Data As Of September 30.1988 (Gray Book 1) NUREG/CR-5009. ASSESSMENT OF THE USE OF EXTENDED i NUREG 0020 V12 N11 LICENSED OPERATlivG REACTORS STATUS BURNUP FUEL IN UGHT WATER POWER REACTORS.

SUMMARY

REPORT. Data As Of October 31,1988.(Gray Book l} NUREG/CR 5043. CONTAINMENT PENETRATION SYSTEM (CPS) TESTS UNDER ACCIDENT LOADS. Sicensee Event Report NUREG/CR-5061: THREE FREQUENCY EDDY CURRENT INSTRU-NUREG/CR 2000 V00N12; UCENSEE EVENT REPORT (LER) MENT. COMPILATION F of Month Of December 1987. NUREG/CR-5083. DESIGN CONSTRUCTION AND INSTRUMENTATION NUREG 'CR-2000 V07 N1: UCENSEE EVENT REPORT (LER) OF A 1/6-SCALE REINFORCED CONCRETE CONTAINMENT BUILD-COMPILATION For Month Of January 1988 ING { l

98 Subject index NUREG/CR-5084 IFCI. AN INTEGRATED CODE FOR CALCULATION NUREG-1200 R01: STANDARD REVIEW PLAN FOR THE REVIEW OF A OF ALL PHASES OF FUEL-COOLANT INTERACTIONS LICENSE APPLICATION FOR A LOW LEVEL RADIOACTIVE WAS1E NURE G/CH-5109 RELOCATION OF METALLIC CONSTITUENTS IN DISPOSAL FACILITY. CORE DEBRIS BEDS NUREG/CR-5134. APPLICATION OF ACOUSTIC LEAK DETECTION MAAP TECHNOLOGY FOR THE DETECTION AND LOCATION OF LEAKS IN NUREG/CR-4993: A STANDARD PROBLEM FOR HECTR MAAP LIGHT WATER REACTORS- COMPARISON. incomplete Buming NUREGICR-5180. CHEMICAL DECONTAMINATION AND CHEMICAL CLEANING OF LWR COMPONENTS AND POSSIDLE INTERACTIONS MELPROG WITH METALLURGICAL AGING EFFECTS. NUREG/CR-5029. MELT PROGRESSION IN SEVERELY DAMAGED RE-i ACTOR CORES. i Line Break NUREG/CR-5109. RELOCATION OF METALLIC CONSTFUENTS IN l NUREG-1332: REGULATORY ANALYSIS FOR THE RESOLUTION OF CORE DEBRIS BEDS. GENERIC ISSUE 12511.7, " REEVALUATE PROVISION TO AUTO-l MATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR MESORAD NUREG/CR 4000 V02: THE MESORAD DOSE ASSESSMENT DURING A LINE BRE AK." NUREG/CR-5179 EVALUATION OF GENERIC ISSUE MODEL Computer Code. 125117, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DURING A LINE DREAK. MiGRAT Code NUREG/CR-5183. A USER'S MANUAL FOR THE CONTAMINANT Loss Of Feedwater TRANSPORT MODULE OF THE MIGRAT CODE. NUREG-1266 501- SAFETY EVALUATION REPORT RELATED TO THE RESTART OF RANCHO SECO NUCLEAR GENERATING MIST STATION. UNIT 1.FOLLOWING THE EVENT OF DECEMBER NUREG-1236: THERMAL HYDRAULIC RESEARCH PLAN FOR BAB-26.1985 Docket No 50-312 (Sacramento Murucepal Ulikty Distnet COCK AND WILCOX PLANTS. Loss Of Offsite Power Maintenance NUREG/CR-5032. MODELING TIME TO RECOVERY AND INITIATING NUREG/CP-0099. PROCEEDirdGS OF THE PUBLIC V" OXSHOP FOR EVENT FREQUENCY FOR LOSS OF OFFSITE POWER INCIDENTS NRC RULEMAKING ON MAINTENANCE OF NUCLEAR POWER  ! AT NUCLEAR POWER PLANTS. PLANTS, Loss-Of-Coolant Accident Maintenance RulemakinD NUREG/CR-4728. EQUIPMENT QUALIFICATION RESEARCH TEST OF NUREG-1333 DRFT FC: MAINTENANCE APPROACHES AND PRAC-A HIGH-RANGE RADIATION MONITOR TICES IN SELECTED FOREIGN NUCLEAR POWER PROGRAMS AND ) NUREG/CR-5066: PLATINUM CATALYTIC IGNITERS FOR LEAN HY* OTHER U.S. INDUSTRIES. Review And Lessons Learned. Draft Report DROGEN AIR MIXTURES For Comment NUREG/CH 5135; THE THERMAL HYDRAULICS OF SDLOCAS RELA-TiVE TO PRESSURIZED THERMAL SHOCK' Management Performance Low Enriched Uranium NUREG/CR-5241: DEVELOPMENT OF FAOGRAMMATIC PERFORM-NUREG 1313: SAFETY EVALUATION REPORT RELATED TO THE ANCE INDICATORS. EVALUATION OF LOW. ENRICHED URANIUM SILICIDE-ALUMINUM Ma " DISPf RSION FUEL FOR USE IN NON. POWER REACTORS. G/ 52 AN OVERVIEW OF BWR MARK-1 CONTAINMENT Low Level Radioactive Waste VENTING RISK IMPLICATIONS. NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOAC. ""*'i9h.d TlVE WASTE DISPOSAL. Environrnental Monitonng And Surveihance NUREG/CRV'IOCY 5065: TIME. AND VOLUME-AVERAGED CONSERVATION Programs. EOUATIONS FOR MULTIPHASE FLOW USING MASS WEIGHTED VE-LOCITY AND INTERNAL ENERGY. Low Level Weste NUREG/CR-3444 V05: THE IMPACT OF LWR DECONTAMINATION Melt Progression ON SOLIDIFICATION.WASIE DISPOSAL AND ASSOCIATED OCCLp NUREG/CR-5029. MELT PROGRESSION IN SEVERELY DAMAGED RE-PATIONAL EXPOSURE. Annual Report. FY 1987 ACTOR CORES. NUREG/CR-4879 V01 DEMONSTRATION OF PERFORMANCE MOD-M* " ELING OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A j] g'019: NEUTRON EXPOSURE PARAMETERS FOR THE Compenson Of Predictive Radionuchde Transport Modehng Versus Field Observations At The Nitrate Disposal Pat Site, Chalk River Nuclear METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SEC-UEEL TECHNOLOGY IRRADIATION SERIES CAPSULES. NU EG 5137 BIODEGRADATION TESTING OF TMl,2 EPICOR-il y g

                     ~

NUREG/CR 5045: KANSAS-NEBRASKA SEISMICITY STUDIES USING Low Level Waste Data Base THE KANSAS NEBRASKA MICROEARTHOUAKE NETWORK. Final NUREG/CR-5D9 ANNUAL REPORT OF THE TMI-2 EPICOR.ll RESIN / Report LINER INVESTIGATION Low-Level Waste Data Base Developer.cnt M Program Fo. Fiscal Year 1908. U E /CR-4625: THE POSTIRRADIATION EXAMINATION OF THE DC Low-Level Radioactive Weste MELT DYNAMICS EXPERIMENTS. NUREG/CR-5204 LOW-LEVEL 9ADIOACTIVE WASTE SOURCE TERM MODEL DEVELOPMENT AND Tt' STING Mne Tamng NUREG 1263: HYOROLOGIC DESIGN FOR HIPRAP ON EMBANKMENT Low-Level Weste SLOPES. NUREG/CR 5041 V02 RECOMMENDATIONS TO THE NRC FOR REVtEW CRfTERIA f DR ALTERNATIVE METHODS OF LOW-LEVEL Mit'getion RADIOACTIVE WASTE DISPOSAL Task 2b Earth-Mounded Concrete NUREG-1318. TECHNICAL POSITION ON ITEMS AND ACTIVITIES IN Bunkers THE HIGH-LEVEL WASTE GEOLOGIC REPOSITORY PROGRAM SUBJECT TO QUALITY ASSURANCE REQUIREMENTS. Low 4evel Waste Iturtal Facility NUREG/CR-3900: SURVEY OF THE STA9 OF THE ART IN MITIGA-NUREG.1307. REPORT ON war T BURIAL CHARGES Escalation Of TlON SYSTEMS. Decommissioning Waste Disposai Cob's At Low Levet Weste Dunal Fa- NURE G/CR-4242: SURVEY OF LIGHT V ATER REACTOR CONTAIN-cihties MENT SYSTEMS, DOMINANT F AILURE MODES AND MITIGATION OPPORTUNITIES Final Report Low-Levet Weste D6sposal FacHy NUREG/CR-4243: VALUE/ IMPACT ANALYSIS FOR EVALUATING AL-NUREG-1199 RO1 STANDARD FORMAT AND CONTENT OF A Ll- TERNATIVE MITIGATION SYSTEMS. CENSE A,' PLICATION FOR A LOW-LEVEL RADIOACTIVE WASTE NUREG/CR 4244- STRATEGIES FOR IMPLEMENTING A MITIGATION DISPOSAL F ACILITY. POLICY FOR LIGHT WATER REACTORS.

i I Subject index 99 NUREG/CR429 V01; ASSESSMENT OF SEVERE ACCIDENT PhE- NRC Belletin 85-03 VENTION AND MITIGATION FEATURES.BWR. MARK I CONTAIN- NUREG/CR 5140: VALUE-lMPACT ANALYSIS FOR EXTENSION OF MENT DESIGN NRC BULLETIN 85-03 TO COVER ALL SAFETY-RELATED MOVS. NUREG/CR-4920 V02: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES BWR, MARK ll CONTAIN- NUC1.ARR j MENT DESIGN NUREG/CR-4639 V01: NUCLEAR COMPUTERIZED LIBRARY FOR AS-NUREG/CR 4920 V03; ASSESSMENT OF SEVERE ACCIDENT PRE- SESSING RE ACTOR RELIABILITY (NUCLARR) Volume i Summary De-VENTION AND MITIGATION FEATURES BWR. MARK lit CONTAIN- senption. MENT DESIGN- NUREG/CR 4639 V02: NUCLEAR COMPUTERIZED LIBRARY FOR AS-l NUREG/CR-4920 VD4: ASSESSMENT OF SEVERE ACCIDENT PRE- SECSING REACTOR RELIABILITY (NUCLARR) Programmer's Guide. i VENTIOlg AND MITIGATION FEATURES PWR.LARGE DRY CON- NUREG/CR 4639 V03 P1: NUCLEAH COMPUTERIZED LIBRARY FOR  ;

                                                                               ^ b NU           4 20 05: ASSESSMENT OF SEVERE ACCIDENT P9E-               Pr ess g A Reviss         Pa    T ch cat e VENTION AND MITIGATION FEATURES PWR. ICE CONDENSER                NUREG/CR-4639 V03 P2: NUCLEAR COMPUTERIZED LIBRARY FOR ASSES $!NG REACTOR RELIABILITY (NUCLARR) Guide To Data NU EG/         4    PRI ITl2ATION OF TIRGALEX-RECOMMENDED              Processing And Revision.Part 2: Human Error Probability Data Entry COMPONENTS FOR FURTHER AGING RESEARCH.

And Revision Procedures. I Mitigation Program NUREG/CR-4639 V03 P3: NUCLEAR COMPUTERIZED LIBRARY FOR l NUREG/CR-5156: REVIEW OF EROSION CORROSION IN SINGLE- ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data PHASE FLOWS. Processing And Revision. Part 3. Hardware Component Failure Data Entry And Revision Procedures. Modeling Time To Recovery NUREG/CR-4639 V04 P1: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5032: MODELING TIME TO RECOVERY AND INITIATING ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide,Part 1: FVENT FREOVENCY FOR LOSS OF OFFSITE POWER INCIDENTS Overview Of NUCLAP,R Data Retneval. AT NUCLEAR POWER PLANTS. NUREG/CR-4639 V04 P2: NUCLEAR COMPUTERIZED LIBRARY FOR ASSES $ LNG REACTOR RELIABILITY (NUCLARR). User's Guide.Part 2: Modification Guide To Operations NUREG/CR-4555 R01: GENERIC COST ESTIMATES FOR THE DIS- NUREG/CR-4630 V04 P3: NUCLEAR COMPUTERIZED LIBRARY FOR POSAL OF RADIOACTIVE WASTES ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide.Part 3: NUCLARR System Desenption. Molsture NUREG/CR 5183. A USER'S MANUAL FOR THE CONTAMINANT NUREG/CR-4639 VOS P1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Pa:t 1: TRANSPORT MODULE OF THE MiGRAT CODE. Summary Desenption. Molten Debris NUREG/CR 4639 V05 P2: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CH-5120. A MODEL FOR THE TRANSPORT AND CHEMICAL ASSESSING REACTOR RELIABILITY (NUCLARA. Data Manual Part 2: , l REACTION OF MOLTEN DEBRIS IN DIRECT CONTAINMENT HEAT. Human Error Probability (HEP) Estimates. l ING EXPERIMENTS NUREG/CR-4639 V05 P3. NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual,Part 3. Molten Jet Hardware Component Failure Data (HCFD). NUREG/CH-4500 Br_HAVIOR OF A CORIUM JET IN HIGH PRESSURE NUREG/CR 4639 V05 P4: NUCLEAR COMPUTERIZED LIBRARY FOR MELT EJECTION FROM A REACTOR PRESSURE VESSEL- ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part 4. j Summay Aggregations. j Morbidity NUREG/CR-5198. INHALED (2391PUO(2) AND/OR TOTAL BODY Natural circulation GAMMA RADIATION Early Mortahty And Morbidity In Rats And Dogs NUREG/CR-5048: REVIEW OF THE NATURAL CIRCULATION EFFECT Mortahty IN THE VERMONT YANKEE SPENT FUEL POOL. Docket No. 50-NUREG/CFt5198 INHALED (239)PUO(2) AND/OR TOTAL BODY 271.(Vermont Yankee Nuclear Power Corp) GAMMA RADIATION Early Mortahly And Morbedity in Rats And Dogs NUREG/CR-5082: SIMULATION EXPERIMENTS ON TWO PHASE b,AT-URAL CIRCULATION IN A FREON-113 FLOW VISUALIZATION LOOP. Motor Control Center NUREG/CR 5214. ANALYSES OF NATURAL CIRCULATION DURING A NUREG/CR.4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER SURRY ST ATION BLACKOUT USING SCDAP/RELAP5. PLANT COMPONENTS (PHASE ll) Motor Control Conter Switchboard.Panelboard And Power Supply Natural-Convection Phenomena NUREG/CH-5053. OPERATING EXPERIENCE AND AGING ASSESS- NUREG/CR 5070 ANALYSIS OF NATURAL-CONVECTION PHENOM-MENT OF MOTOR CONTROL CENTERS. ENA IN A 3-LOOP PWR DURING A TMLB' TRANSIENT USING THE Motor Rehability NUREG/CR-4939 V01. IMPROVING MOTOR RELIABILITY IN NUCLEAR Nephrotox6 city POWER PLANTS Volume 1. Performance Evaluation And Maintenance NUREG/CP-0093. PROCEEDINGS OF THE MEETING ON ULTRASEN-Practaces SITIVE TECHNIOUES FOR MEASUREMENT OF URANIUM IN BIO-l NUREG/CR-4939 V02 IMPROVING MOTOR RELIABILITY IN NUCLEAR LOGICAL SAMPLES AND THE NEPHROTOXICITY OF URANIUM. I l POWER PLANTS. Volume 2 Functional Indicator Tests On A Smah ! Doctnc Motor Sublected To Accelerated Aging Neutron Dosimetry l NUREG/CR-4939 V03: IMPROVING MOTOR RELIABILITY IN NUCLEAR NUREG/CR-5049- PRESSURE VESSEL FLUENCE ANALYSIS AND l POWER PLANTS Volume 3 Failure Analysis And Diagnostic Tests On NEUTRON DOSIMETRY. l A Naturally Aged large Electnc Motor. Neutron Exposure Parameter NUR 12 E MAL OVERLOAD PROTECTION FOR ELECTRIC NUREG/CR 5019: NEUTRON EXPOSURE PARAMETERS FOR THE MOTORS ON SAFETY RELATED MOTOR OPERATED VALVES - GE-METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SEC-NF AlC ISSUE 11 E.61. TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES-NUREG/CR 5140: VALUE lMPAC1 ANALYSIS FOR EXTENSION OF

• NRC DULLETIN 05 03 TO COVER ALL SAFETY-RELATED MOVS N RE / 4 0 V01: CHARACTERIZATION OF IRRADIATED CUR-Multiphase Flow RENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE TOR NUREG/CR 5065. TIME. AND VOLUME-AVERAGED CONSERVATION NUCLEAR PRESSURE VESSEL SERVICE.

EQUATIONS FOR MULTIPHASE FLOW USING MASS WEIGHTED VE. NUREGICR 4BB0 V02: CHARACTERl2ATION OF 1RRADIATED CUR-LOCITY AND INTERNAL ENERGY- RENT-PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUCLEAR PRESSURE VESSEL SERVICE. Muttostage Switch NUREG/CH-4992 V0t AGING AND SERVICL WEAR OF VULTIST AGE Nevada Test Site SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER NUREGICR-4916. HECTR ANALYSES OF THE NEVADA TEST SITE PLANTS Operating Expenence And Failure lderitihcation. (NTS) PRE MIXED COMBUSTION EXPER!MENTS

100 Subject index Nondestructive Examination Numerical Model NUREG/CR-5075: THE SAFT UT REAL TIME . INSPECTION SYSTEM - NUREG/CR-5099: EVALUATION OF MATERIALS OF CONSTRUCTION OPEFIATIONAL PRINCIPLES AND IMPLEMENTATION. FOR THE REINFORCED CONCRETE REACTOR CONTAINMENT NUREG/CR-5165: STEAM GENERATOR GROUP PROJECT Task 13 MODEL Final Report- Nondestructive Examinat on (NDE) Validation. Occupat6onal Exposure N FEG/CR4705: REVIEW AND EVALUATION OF DESIGN ANALYSIS NUREG/CR 5158 V01: WORLDWIDE ACTIVITIES ON THE REDUCTION METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS. BRANCH CONNECTIONS- Occupational Rad 6stion Cuclear Acc6 dent R6ak . NUREG-0713 V07: OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR-5000: METHODOLOGY FOR UNCERTAINTY ESTIMATION MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES IN NUREG 1150 (DRAFT) Conclusions Of A Revew Fanel. 1985 Eighteenth Annual Report. Nuclear Fuel Cycle Offsite Consequence - NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS NUREG/CR 5164: A SIMPLIFIED MODEL FOR CALCULATING EARLY i HANDBOOK. OFFSITE CONSEQUENCES FROM NUCLEAR REACTOR ACCl-Nuclear Material Managment NUREG/CR-4605: TRAINING MANUAL ON STATISTICAL METHODS Ottalte Emergency Plan FOR NUCLEAR MATERIAL MANAGEMENT.~ NUREG 0654 S01 R01: CFilTERIA FOR PREPARATION AND EVALUA-i' TION OF RADIOLOGICAL EMERGENCY RESPONSE PLANS AND UR 1 2 VO S FETY EVALUATION REPORT ON TENNESSEE PREPAREDNESS IN SUPPORT OF NUCLEAR POWER i VALLEY AUTHORITY. Sequoyah Nuclear Performance Plan. PLANTS Critena For Utility Offsite Planning And Preparedness. Nuclear Plant Aging Operating Experience NUREG/CR.5248: PRIORITIZATION OF TIRGALEX-RECOMMENDED NUREG 1272 V02 N0t REPORT TO THE U.S. NUCLEAR REGULA. COMPONENTS FOR FURTHER AGING RESEARCH. TORY COMMISSION ON ANALYSIS AND EVALUATION OF OPER-ATIONAL DATA 1987. Power Fleactors. Nuclear Plant-Aging Research NUREG 1272 V02 NO2: REPORT TO THE U.S. NUCLEAR REGULA- . NUREG/CR-4740: NUCLEAR PLANT-AGING RESEARCH ON REACTOR TORY COMMISSION ON ANALYSIS AND EVALUATION OF OPER-PROTECTION SYSTEMS. I ATIONAL DATA - 1987.Nonreac* ors. I

               ""'*                " ' V'                                                 NUREG/CR-4991: EVALUATION AND PROPOSED IMPROVEMENTS TO NUR               4B80 V   CHARACTER 12ATION OF IRRADIATED CUR-                EFFECTIVENESS OF U.S. NUCLEAR REGULATORY COMMISSION FIENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUR G/ - 9 V AG G A'ND SERVICE WEAR OF MULTISTAGE NL R / R48               ? CHAR TR         ON OF IRRADIATED CUR-             SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER RENT FRACTfCF WELDS AND A533 GRADE B CLASS 1 PLATE FOR                     PLANTS. Operating Expenence And Failure identification.

NUCLEAR FflESSURE VESSEL SERVICE' NUREG/CR-5052: OPERATING EXPERIENCE AND AGING ASSESS-MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR-Nuclear Reactor Accident IZED WATER REACTORS. NUREG/CR-5120: A MODEL FOR THE TRANSPORT AND CHEMICAL REACTION OF MOLTEN DEBRIS IN DIRECT CONTAINMENT HEAT, Operating Experience Feedback ING EXPERIMENTS. NUREG-1275 V03: OPERATING EXPERIENCE FEEDBACK REPORT a SERVICE WATER SYSTEM FAILURES AND - Nuclear Reactor Personnel DEGRADATIONS. Commercial Power Reactors. NUREG/CP.0089: PROCEEDINGS OF THE CSNI SPEC ALIST MEET. ING ON TRAINING OF NUCLEAR REACTOR PERSONNEL. Mold At Operating Procedure Orlando. Florida, April 2124,1987. NUREG/CR-4775: GUIDE FOR PREPARING OPERATING PROCE-Nuclear Safety Pilot Plant " ' NUREG/CR-5018 ORANIUM OX1DE lRON OXIDE MIXED AEROSOL Operational Event EXPERIMENTS IN STEAM. AIR ATMOSPHERES.NSPP Tests 611,612,613 And 631. Data Record Report- NL' REG /CR-4674 V05: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:198E.A STATUS REPORT. I Nuclear Safety Research NUREG/CR-4674 V06: PRECURSORS TO POTENTIAL SEVERE CORE l NUREG 1325: DISPOSITION OF RECOMMENDATIONS OF THE NA- DAMAGE ACCIDENTS-1986,A STATUS REPORT. TIONAL RESEARCH COUNCI CLEAR SAFETY RESEARCH,L IN THE REPORT " REVITALIZING NU- Operator Examination NUREG/CP 0001 V01: PROCEEDINGS OF THE FIFTEENTH WATER NWEG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PERFORM-REACTOR SAFETY INFORMATION MEETING. ANCE INDICATORS. NUREG/CP-0091 V02; PROCEEDINGS OF THE FIFTEENTH WATER REACTO A SAFETY INFORMATION MEETING Organtration Chart NUREG/C,r' 0091 V03: PROCEEDINGS OF THE FIFTEENTH WATER NUREG 0325 R11: U.S. NUCLEAR REGULATORY COMMISSION FUNC-REACTOR SAFETY INFORMATION MEETING. TIONAL ORGANIZATION CHARTS. NUREG/CP 0091 V04: PROCEEDINGS OF THE FIFTEENTH WATER REACTOR SAFETY INFORMATION MEETING. Oxide NUREG/CP 0091 VOS. PROCEEDINGS OF THE FIFTEENTH WATER NUREG/CR-5108. THERMODYNAMIC PROPERTIES OF TC(IV) NUREG CP O V EE NG T E FIFTEENTH WATER

                                                                                                          '"         ^"                                  " '

upl ' REACTOR SAFETY INFORMATION MEETING. NUREG/CP 0096: TRANSACTIONS OF THE SIXTEENTH WATER RE- P-Wave Res6 dual ACTOR SAFETY INFORMATION MEETING. NUREG/CR 3145 V06: GEOPHYSICAL INVESTIGATIONS OF THE C0uclear Waste WESTERN OHIO-INDIANA REGION. Annual Fleport. October 1986 - NUREG-1249 Vot: NRC MODEL S!MULATIONS iN SUPPORT OF THE September 1987. HYDROLOGIC CODE INTERCOMPARtSON (HYDROCOIN) STUDY 1evel 1 - Code Venfication. PH Sensor NUREG/CR416& ELECTROCHEMICAL EVALUATION OF SOLID NUREG/CR-5166. ELECTROCHEMICAL EVALUATION OF SOLID ST A1E PH SENSORS FOR NUCLE AR WASTE CONTAINMENT. STATE PH SENSORS FOR NUCLEAR WASTE CONTAINMENT. Nuclear Weste laolat6on PORV NUREG/CR 4864 V01: THERMODYNAMIC TABLES FOR NUCLEAR NUREG/CR-4999: ESTIMATION OF RISK REDUCTION FROM IM-W ASTE ISOLATION. Aqueous Solutions Database. PROVED PORV RELIABILITY IN PWRS. Final Report. l l l

1 Subject index 101-PRA . NUREG/CR-5149: EROSION CORROSION OF PWR FEEDWATER NUREG/CR-4834 V02: RECOVERY ACTIONS IN PAA FOR THE RISK PIPING SURVEY OF EXPERIENCE. DESIGN. WATER CHEMISTRY METHODS INTEGRATION AND EVALUATION PROGRAM ' AND MATERIALS. (RMIEP) Volume 2.Appiscation Of The Date-Based Method. NUREG/CR-5150: STEAM GENERATOR OPERATING NUREG/CR-4836: APPROACHES TO UNCERTAINTY ANALYSIS IN EXPERIENCE. Update For 1984-1986. PROBABILISTIC RISK ASSESSMENT. NUREG/CR4160: GUIDELINES FOR THE USE OF THE EEDB AT THE NUREG/CR-5058: PRA APPLICATIONS PROGRAM FOR INSPECTION SUB-COMFDNENT AND SUBSYSTEM LEVELFanal Report. AT ARKANSAS NUCLEAR ONE UNIT 1.Dochel No. 50-313.(Arkansas Power And Ligt$ Company) Panothoord NUREG/CR5076: AN APPROACH TO THE QUANTIFICATION OF SEIS* NUREG/CR-4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER MIC MARGINS IN NUCLEAR POWER PLANTS.The importance Of PLANT COMPONENTS (PHASE II). Motor Control NR / R5 8 E AL Tb b 7tNRIC ISSUE *"I** * * ^" "' 0 125.ll1. REEVALUATE PROMSION TO AUTOMATICALLY ISOLATE Peak Clad Temperature FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. NUREG/CR-5232; UNCERTAINTIES IN MODELING AND SCALING IN NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT AND ST! REQUIREMENTS AT THE ANO-1 NUCLEAR POWER THE PREDICTION OF FUEL STORED ENERGY AND THERMAL RE-PLANT. SPONSE. PmSIM - Peer Review Panel NUREG/CR-5021 V01: USER'S GUIDE FOR PRISIM ARKANSAS NU. NUREG/CRS113: FINDINGS OF THE PEER REVIEW PANEL ON THE CLEAR ONE - UNIT 1. Volume 1 Program For inspectors. DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG 1150. NUREG/CR5021 V02: USER'S 6UIDE FOR PRISIM ARKANSAS NU-CLEAR ONE UNIT 1. Volume 2. Program For Regulators. e p PVRC IC ASSESSMENT OF LICENSEE PERFORMANCE. - NUREG/CR-5105: RESPONSE MARGINS INVESTIGATION OF PIPING NUREG 1214 R04: HISTORICAL DATA

SUMMARY

OF THE SYSTEWAT-DYNAMIC ANALYSES USING THE INDEPENDENT SUPPORT IC ASSESSMENT OF LICENSEE PERFORMANCE. MOTION METHOD AND PVRC DAMPlNG. Performance indicator PWR-

                                                                                                 ' NUREG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PERFORM-NUREG 0844: NRC INTEGRATED PROGRAM FOR THE RESOLUTION                        ANCE INDICATORS.

OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-6 REGARDING STEAM GENERATOR TUBE INTEGRITY.Finnt Report. Personnel Barrter NUREG 1217 DRFT FC: E'. ALUATION OF SAFETY IMPLICATIONS OF NUREG-1330: PERSONNEL AND VEHICLE BARRIERS AT FUEL FABRI-CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTSTechnical Finrlings Related To Unresolved Safety issue A-47. Draft Report For CATION FACILITIES USING OR POSSESSING STRATEGIC QUANTI-TIES OF SPECIAL NUCLEAR MATERIAL, Comment. NUREG/CH 0130 ADD 04: TECHNOLOGY SAFETY AND COSTS OF Petit 6ons For Rutemaking DECOMMISSIONING A REFERENCE PRESSURIZED WATER REAC-TOR POWER STATION. Technical Support For Decommissioning Mat- NUREG 0936 - V06 N04: NRC REGULATORY AGENDA.Ouarterty

                                                                                                    - Report October-December 1987.

N /Y20 S7 N L OR (LER) NUREG-0936 V07 NO1: NRC REGULATORY AGENDA.Ouartnty NU EG/ 0 V7 9 LIE EVENT REPORT (LERg NURE -0936 07 NO2 NRC REGULATORY AGENDA.Ouarterly COMPILATION:For Month Of September 1988. Report, April-June 1988. NUREG/CR-2336: STEAM GENERATOR TUBE INTEGRITY NUREG 0936 V07 NO3. NRC REGULATORY AGENDA.Ouarterty PROGRAM. Phase 11 Final Report. Report. July September 1988 NUREG/CR4312 V01: RELAPs/ MOD 2 CODE MANUALVolume 1: Code Structure. Systems Models And SoluiLn Methods Phenomenological Research NUREG/CR-4312 V02 R1: RELAPS/ MOD 2 CODE MANUALVolume 2: NUREG/CR-5039 V01: REACTOR SAFETY RESEARCH SEMIANNUAL Users Guide And input Rettuirements. REPORT. January. June 1987. Volume 37. NUREG/CR-4597 V02: AGING AND SERVICE WEAR OF AUXILIARY NUREG/CR-5039 V02: REACTOR SAFETY RESEARCH SEMIANNUAL FEEDWATER PUMPS FOR PWR NUCLEAR PLANTS. Volume 2. Aging REPORT. July December 1987. Reactor Safety Research Program. Assessments And MonitonnqMethod Evaluations. NUREG/CR 4761 SAFETYJILLATED EQUIPMENT SURVIVAL IN HY- Physical Security DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS. NUREG/CR 4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE' NUREG 1178: VITAL EQUIPMENT / AREA GUIDELINES STUDY. Vital VENTION AND MITIGATION FEATURES.PWR,LARGE DRY CON- Area Committee Reportorial Repert. TAINMENT DESIGN. pipe NUREG/CR 4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES.PWR,1CE CONDENSER NUREG/CP 0075: PROCEEDINGS OF CSNI/NRC WORKSHOP ON CONTAINMENT DESIGN. DUCTILE PIPING FRACTURE MECHANICS. NUREG/CR4999: ESTIMATION OF RISK REDUCTION FROM IM- Pipe Fracture NURE /CR 50 3 AT G Li CHA h A QN OF SMOOTH NUREG 1222: PIPING RESEARCH PROGRAM PLAN. AND NOTCHED PIPING STEEL SPECIMENS IN 288 DEGREES C AIR Piping N RE / 5 IMPROVED RELIABILITY OF RESIDUAL HEAT RE- NUREG-1222: PlPING RESEARCH PROGRAM PLAN. . 1 l MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF NUREG/CR-4785: REVIEW AND EVALUATION OF DESIGN ANALYSIS j GENERIC ISSUE 99 METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND i NUREG/CR-5052. OPERATING EXPERIENCE AND AGING ASSESS. BRANCH CONNECTIONS. l MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR- NUREG/CR-5023: HIGH-LEVEL SEISMIC RESPONSE AND FAILURE , PREDICTION METHODS FOR PIPING.  : IZED WATER NUREG/CR4070: REACTORS'S OF NATURAL CONVECTION ANALYSl PHENOM. NUREG/CR-5073: OUANTIFICATION OF MARGINS IN PIPING SYSTEM { ENA IN A 3 LOOP PWR DURING A TMLB' TRANSIENT USING THE SEISMIC RESPONSE Methodologies And Damping. i COMMIX CODE. NUREG/CR-5105: RESPONSE MARGINS INVESTIGATION OF PIPING I NUREG/CR5072: DECAY HEAT REMOVAL USING FEED ANDBLEED DYNAMIC ANALYSES USING THE INDEPENDENT SUPPORT j FOR U S. PRESSURIZED WATER REACTORS. MOTION METHOD AND PVRC DAMPING. NUREG/CR 5135: THE THERMAL HYDRAULICS OF SBLOCAS REI A- NUREG/CR 5220 VO1: DIAGNOSIS OF CONDENSATION-INDUCED TIVE TO PRESSUR ZED THERMAL SHOCK. WATERHAMMERMethods And BacitDround. NUREG/CR-5130: FATIGUE STRENGTH OF SMOOTH AND NOTCHED NUREG/CR-5220 V02: DIAGNOSIS OF CONDENSATION-INDUCED SPECIMENS OF ASME SA 106 B STEEL IN PWR ENVIRONMENTS WATERHAMMER. Case Studies. '! NUREG/CR4146. DEBRIS DISPERSAL FROM REACTOR CAVITIES l DURING HIGH PRESSURE MELT EJECTION ACCIDENT SCENAR. Piping Research Program IOS. NUREG 1222: PIPING RESEARCH PROGRAM PLAN.

l l 102 Subject Index Piping Response Margin DIGEST. Commission. Appeal Board And bcensing Board Decisions.Juty l i NUREG/CR-5C73. OUANTIFICATION OF MARGINS IN PIPING SYSTEM 1972 December 1987. l SEISMIC RESPONSE Methodologies And Damping Pressure Boundary Piping l Piping Steel NUREG 0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION I NUREG/CR-4828. FATIGUE CRACK GROWTH Or PART THROUGH AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE CRACKS IN PRESSURE YESSEL AND PIPING STEELS Air Environ- DOUNDARY PIPING Final Report ment Results. NUREG/CR 5013: FATIGUE LIFE CHARACTERIZATION OF SMOOTH Pressure Vessel AND NOTCHED PIPING STEEL SPECIMENS IN 288 DEGREES C AIR NUREG/CR 4219 V04 N2. HEAVY SECTION STEEL TECHNOLOGY ENVIRONMENTS PROGRAM Semennual Progress Report For April-September 1987-i NUREG/CR-5136 FATIGUE STRENGTH OF SMOOTH AND NOTCHED NUREG/CR 4219 V05 N1: HEAVY-SECTION STEEL TECHNOLOGY l SPECIMENS OF ASME SA 106-0 STEEL IN PWR ENVIRONMENTS. PROGRAM Semaannual Progress Report For October 1987 - March 1988. P6PinD System NUREG/CR4860 R01: FLAW DENCITY EXAMINATIONS OF A CLAD NUREG/CR 5073 OUANTIFICATION OF MARGINS IN PIPING SYSTEM BOILING WATER REACTOR PRESSURE VESSEL SEGMENT. SEISMIC RESPONSE, Methodologies And Dampin9 NUREG/CR 4888: PRESSURIZED-THERMAL-SHOCK TEST OF 6-INCH THICK PRESSURE VESSELS PTSE-2: Investigation Of Low Teanng Re-U EG/ R 5204: LOW-LEVEL RADIOACTIVE WASTE SOURCE TERM NU G R

• 04 R SU ESSEL FLUENCE ANALYSIS AND MODEL DEVELOPMENT AND TESTING. NEUTRON DOSIMETRY.

NUREG/CR-6063: DEVELOPMENT OF A MECHANISTIC UNDER. NU EG/CR-5150. STEAM GENERATOR OPERATING STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES-EXPERIENCE Update for 1964-1986. U E M L NURE / 52 1 E PER L ASSESSMENTS OF GUNDREM- . Policy Statement MINGEN FIPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS l NUREG-1226- DEVELOPMENT AND UTILIZATION OF THE NRC STUDIES. POLICY STATEMENT ON THE REGULATION OF ADVANCED NU-CLE AR POWER PLANTS' Pressure Vessel Steel l NUREG/CR-4828. FATIGUE CRACK GROWTH OF PART.THROUGH j Population CRACKS IN PRESSURE VESSEL AND PIPING STEELS. Air Environ-NUREG/CR-2850 V06: POPULATION DOSE COMMITMENTS DUE TO rnent Results. J RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1984. _ Pressurized Thermal Shock ( l NUREG/CR4888: PRESSURIZED THERMAL SHOCK TEST OF 6-INCH l Population Dose THICK PRESSURE VESSELS.PTSE-2: investigation Of Low Teanng Re-NUREG/CR-2850 V07. POPULATION DOSE COMMITMENTS DUE TO sistance And Warm Prestressing. RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR-5135. THE THERMAL HYDRAULICS OF SBLOCAS RELA-IN 19B5, TIVE TO PRESSURIZED THERMAL SHOCK. Population Esposure Pressurized Water Reactor NURE G/CR4811: THE ECONOMIC COSTS OF RADIATION-INDUCED NUREG/CR-5016: COMPENDIUM AND COMPARISON OF INTERNA. HEALTH EFFECTS Estimation And Simulation. TIONAL PRACTICE FOR PLUGGING. REPAIR AND INSPECTION OF STEAM GENERATOR TUBING Porosity NUREG-0844. NRC INTEGRATED PROGRAM FOR THE RESOLUT!ON NUREG/CR 5277: THE TENSORIAL NATURE OF EFFECTIVE POROSI- OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A 5 FIEGARDING TY AND LARGE-SCALE DISPERSION COEFFICIENTS Application To STE AM GENEEATOR TUBE INTEGRITY. Final Report. The Creston Study Area. Eastern Washington. NUREG-1217 DAFT FC. EVALUATION OF SAFETY IMPLICATIONS OF CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS. Technical Fmdings Related To Unresolved Safety issue A 47. Draft Report For N E /C 3. STUDIES OF THE PATTERN AND AGES OF POST-METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND NU EG/ I 0130 ADD 04: TECHNOLOGY. SAFETY AND COSTS OF NORTH CAROLINA. DECOMMISSIONING A REFERENCE PRESSURIZED WATER REAC- l Poetirradiation Examination TOR POWER STATIONTechncal Support For Decommissioning Mat- 1 NUREG/CR4625: THE POSTIRrlADIATION EXAMINATION OF THE DC ters Related To Preparation Of The Final Decommissioning Rule. , MELT DYNAMICS EXPERIMENTS NUREG/CR-2000 V07 NB: LICENSEE EVENT REPORT (LER) i COMPILATION.For Month Of August 1986. Power Reactor NUREG/CR-2000 V07 N9: LICENSEE EVENT REPORT (LER) NUREG/CR4047. ANALYSIS OF THE A302B AND A533B STANDARD COMPILATION For Month Of September 1988. REFERENCE MATERIALS IN SURVEILLANCE CAPSULES OF COM- NUREG/CR 2336 STEAM GENERATOR TUBE INTEGRITY PROGR AM. Phase 11 Final Report. MERCIAL POWER REACTORS NUREG/CR 4312 V01: RELAPS/ MOD 2 CODE MANUALVolume 1: Code Power Spectral Density Structure. Systems Models And Solution Methods. NURE G/CR-3509 POWER SPECTRAL DENSITY FUNCTIONS COM. NUREG/CR4312 V02 R1: RELAPL/ MOD 2 CODE MANUALVolume 2: PATIBLE WITH NRC REGULATORY GUIDE 160 RESPONSE SPEC. Users Guide And input Requirements TRA NUREG/CR-4597 V02: AGING AND SERVICE WEAR OF AUXILIARY FEEDWATER PUMPS FOR PWR NUCLEAR PLANTSVolume 2. Aging Power Supply Assessments And MonitonnqMethod Evaluations. NUREGICR4659 V02 SEISMIC FRAGILITY OF NUCLEAR POWER NUREG/CR4763. SAFETY-RLLATED EQUIPMENT SURVIVAL IN HY-PLANT COMPONENTS (PHASE II) Motor Control DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILDINGS. , Center, Switchboard,Panolboard And Power Supply NUREG/CR 4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE- 1 VENTION AND MITIGATION FEATURES.PWR LARGE DRY CON-Practice And Procedure Digest T AINMENT DESIGN NUREG 0306 D04 R07: UNITED STATES NUCLEAR REGULATORY NUREG/CR4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE-COMMISSION STAFF PRACTICE AND PROCEDURE DIGESTJuly VENTION AND M!TIGATION FEATURES PWR,1CE CONDENSER 1972 - March 1967 CONTAINMENT DESIGN. NUREG 0386 D04 ROB. UNITED STATES NUCLEAR HEGULATORY NUREG/CR-4999- E STIM ATION OF RISK REDUCTION FROM IM-COMMISSION STAFF PRACTICE AND PROCEDURE DIGESTJuly PROVED PORV RELIABILITY IN PWAS Final Report. 1972 - June 1987. NUREG/CR 5013. FATIGUE LIFE CHARACTERIZATION OF SMOOTH NUREG-03H6 D04 ROD UNITED STATES NUCLEAR REGULATORY AND NOTCHED PIPlHG STEEL SPECIMENS IN 288 DEGREES C AIR COMM!$SION STAFF PRACTICE AND PROCEDURE DIGESTJuly ENVIRONMENTS 1972 September 1987 NUREG/CR-5015- lMPROVED RELIABILITY OF RESIDUAL HEAT RE. NUREG 0386 D04 R10 UNITED ST ATES NUCLEAR REGULATORY MOVAL CAPADILITY IN PWR$ AS RELATED TO RESOLUTION OF COMMISSION STAFF PRACTICE AND PROCEDURE GENERIC ISSUE 99

Subject index 103 NUREG/CR 5052. OFERATING EXPERIENCE AND AGING ASSESS. NUREG-129B: QUALIFICATION OF EXISTING DATA FOR HIGH-LEVEL MENT OF COMPOhENT COOLING WATER SYSTEMS IN PRESSUR- NUCLEAR WASTE REPOSITORIES Genenc Technical Position.

               '2ED WATER REACTORS.                                                NUREG-1318- TECHNICAL POSfTION ON ITEMS AND ACTIVITIES IN NUREG/CR-5070: ANALYSIS OF NATURAL CONVECTION PHENOM-                    THE HIGH LEVEL WASTE GEOLOGIC REPOSITORY PROGRAM ENA IN A 3 LC.OP PWR DURING A TMLB' TRANSIENT USING THE               SUBJECT TO QUALITY ASSURANCE REQUIREMENTS.                                             ;

COMMIX CODE - NUREG/CR-5151: PERFORMANCE-BASED INSPECTIONS. < NUREG/CR 5072: DECAY HEAT REMOVAL USING FEED-AND-BLEED 1 FOR U S PRESSURIZED WATER REACTORS Quality Assurance Program l NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS RELA' NUREG-1306: NRC SAFETY SIGNIFICANCE ASSESSMENT TEAM Nt G H $130 ATI E ST H F MOOTH AND NOTCHED

                                                                                                    ^       ^

PR T* S 2' I SPECIMENS OF ASME SA 106 B STEEL IN PWR ENVIRONMENTS. 1 NUREG/CR-5146. DEBRIS DISPERSAL FROM REACTOR CAVITIES l REIRS URING HIGH PRESSURE MELT EJECTION ACCIDENT SCENAR-NUREG-0713 V07; OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR-5149: EROSION-CORROSION OF PWR F EEDWATER MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES PIPING SURVEY OF EXPERIENCE. DESIGN. WATER CHEMISTRY 1985 Eighteenth Annus! Report. , AND MATERIALS. NUREG/CR-5150: STEAM GENERATOR OPERATING RELAPS/ MOD 2 EXPERIENCE Update For 1984-1986. NUREG/CR-4312 VOI: RELAP5/ MOD 2 CODE MANUALVolume 1: Code  ! NUREG/CR-5160. GUIDELINES FOR THE USE OF THE EEDB AT THE Structure. Systems Mocels And Solution Methods. I SUB COMPONENT AND SUBSYSTEM LEVELFmal Report. NUREG/CR-4312 V02 R1: RELAP5/ MOD 2 CODE MANUALVolume 2: I Users Guide And input Requ:mmants. Priorttlantion NUREG/CR-5194. RELAP5/ MOD 2 MODELS AND CORRELATIONS. NUREG 0933 S07: A PRIORITIZATION OF GENERIC SAFETY ISSUES. NUREG 0933 SOB: A PRIORITIZATION OF GENERIC SAFETY ISSUES- HMIEP Probabihstic Risk Analysis NUREG/CR-4834 V02: RECOVERY ACTIONS IN PRA FOR THE '41SK NUREG/CR 5076: AN APPROACH TO THE OUANTIFICATION OF SElS- METHODS INTEGRATION AND EVALUATION PROGRAM MIC MARGINS IN NUCLEAR POWER PLANTSThe importance Of (RMIEP) Volume 2. Application Ot The Data-Based Method. l BWR Plant Systems And Functions To Seismic Margins NUREG/CR-5140: VALUE lMPACT ANALYSIS FOR EXTENSION OF Radiation Dose l 3 NRC BULLETIN 85 03 TO COVER ALL SAFETY RELATED MOVS. NUREG/CR-2850 V06: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES Probabilistic Rtak Assessment IN 1984. l NUREG/CR-4834 V02: RECOVERY ACTIONS IN PRA FOR THE RISK NUREG/CR 5106: USER'S GUIDE FOR THE TACT 5 COMPUTER CODE. l METHODS INTE GRATION AND EVAL.UATION PROGRAM i (RMIEP) Volume 2. Apphcation Of The Data-Based Method. Radiation Embrittlement NUREG/CR-4836: APPROACHES TO UNCERTAINTY ANALYSIS IN NUREG/CR-5201: EXPERIMENTAL ASSESSMENTS OF GUNDREM-PROBABILISTIC RISK ASSESSMENT. MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS NUREG/CR-5021 V01. USER'S GUIDE FOR PRISIM ARKANSAS NU^ STUDIES. CLE AR ONE UNIT 1. Volume 1.Propam For inspectors NUREG/CR-bO21 V02: USER'S GUIDE FOR PRISlM ARKANSAS NU- Radiation Exposure NL C 5050 NOT TYD i L R PHY IABILITY AND

                                                                                             ##                                            b RISK DATA SOURCES                                                      PAREDNESS FOR FUEL CYCLE AND OTHER RADIOACTIVE MATE-NURE G/CR-5058: PRA APPLICATIONS PROGRAM FOR INSPECTION                   RIAL LICENSEES Final Report AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50-313 (Amansas NU E$/ 5 f8                      ALUATION    OF     GENERIC     ISSUE   N E /CR 4811: THE ECONOMIC COSTS OF RADATION-INDUCED 12511.7, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE                 HEALTH EFFECTS. Estimation And Simulation.

FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. NUREG/CR-5200 EVALUATION OF RISKS ASSOCIATED WITH AOT Radiation Monitoring Network AND STI REQUIREMENTS AT THE ANO 1 NUCLEAR POWER NUREG 0837 V08 N01: NRC TLD DIRECT RADIATION MONITORING PLANT. NETWORK. Progress Report. January-March 1988. NUREG 0837 V0B NO2. NRC TLD DIRECT RADIATION MONITORING Probable Maximum Flood NETWORK. Progress Report. Aprildune 1988. NUREG 0800 02 4 2 R3 STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Radiation Protection PLANTS. LWR Edition Proposed Reviwon 3 To SRP Sechon 2 4.2. NUREG/CR-5038. OPTIMl2ATION OF THE CONTROL OF CONTAMINA.

             " Floods " For Comment                                                 TION AT NUCLEAR POWER PLANTS.

NUREG 0800 02 4.3 R3. STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Radiation Source PLANTS. LWR Edition Proposed Revision 3 To SRP Section 2.4.3. NUREG 1310: NATURALLY OCCURRING AND ACCELERATOR-PRO-

             " Probable Maximum Flood (PMF) On Streams & Rivers " For Com-          DUCED RADIOACTIVE MATERIALS.1987 Review.

ment Project Descriptions NUREG 1260 V02 A REPORT TO CONGRESS ON NUCLE AR REGULA. NUREG 1140: A REGULATORY ANALYSIS ON EMERGENCY PRE-TORY RE SEACCKProject Desenphons For FYBB PAREDNESS FOR FUEL CYCLE AND OTHER RADIOACTIVE MATE-RIAL LICENSEES. Final Report. O List NUREG 1308 RADIOACTIVE MATERIAL IN THE WEST LAKE NUREG 1318 TECHalCAL POSITION ON ITEMS AND ACTIVITIES IN LANDFILLSummary Report. THE HIGH-LEVEL WASTE GEOLOGIC REPOSr10RY PROGRAM NUREG 1308 RO1: RADIOACTIVE MATERIAL IN THE WEST LAKE SUBJECT TO QUALITY ASSURANCE REQU!REMENTS LANDFILLSummary Report. NUREG 1310: NATURALLY OCCURRING AND ACCELERATOR PRO. OUASAR DUCED RADIOACTIVE MATERIALS 1967 Review NUREG/CR 46BB V02: OUANTIFICATION AND UNCERTAINTY ANALY. NUREG/CR 2907 V06- RADIOACTIVE MATERIALS RELEASED FROM SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT NUCLEAR POWER PLANTS. Annual Report For 1985. WATER REACTORS (OUASAR)Part it' Sensitety Analysis Tech. NUREGICR-2907 V07. RADIOACTIVE MATERIALS RELEASED FROM niques. NUCLEAR POWER PLANTS. Annual Report For 1986. NUREG/CR 4775 GUIDE FOR PREPARING OPERATING PROCE-Quality Assurance DURES FOR SHIPPING PACKAGES NUREG 1297. PEER REVIEW FOR HIGH-LEVEL NUCLE AR WASTE RE- NUREG/CR-6212: EMERGENCY ENVIRONMENTAL SAMPLING AND POSITORIES Genenc Technical Position ANALYSl5 4 09 RADIOACTIVE MATERi%L FACILITIES

104 Subject index Radioactive Release NUREG/CR-5099. EVALUATION OF MATERIALS OF CONSTRUCTION NUREG/CR-2850 V06: POPULATION DOSE COMMITMENTS DUE TO FOR THE REINFORCED CONCRETE REACTOR CONTAINMENT RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES MODEL-sN 1984 NUREG/CR-2650 V07: POPULATION DOSE COMMITMENTS DUE TO Reactor Coolant System RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR 5214. ANALYSES OF . NATURAL CIRCULATION DURING A IN 1985. SURRY STATION BLACKOUT USING SCDAP/RELAP5. Radioactive Waste Reactor Core NUREG 0683 S03 DRFT: PROGRAMMATIC ENVIRONMENTAL IMPACT NUREG/CR-5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE. STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL OF MOVAL CAPABILITY IN PWR$ AS RELATED TO RESOLUTION OF RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 GENERIC ISSUE 99. ACCIDENT,THREE MILE ISLAND NUCLEAR STATION.UtilT 2. Docket NUREG/CR-5106: USER'S GUIDE FOR THE TACT 5 COMPUTER CODE. No 50-320.(GPU Nuclear. incorporated) NUREG-1300. RADIOACTIVE MATERIAL IN THE WEST LAKE Reactor Piping System LANDFILLSumrnary Report- NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND NUREG-1308 RO1: RADIOACT!VE MATERIAL IN THE WEST LAKE FLOW VISUALIZATION STUDIES AND TECHNIQUES RELEVANT TO THE STUDY OF EROSION-CORROSION OF REACTOR PIPING SYS-NU E /R 3 VD Rt VALUATION OF NUCLEAR FACILITY DE-COMM:S$lONING PROJECTS. Summary Status Report,Three Mde TEMS. Island Unit 2,Radioactwe Waste And Laundry Shipments Reactor Pressure Boundary NUREG/CR4555 Rot: GENERIC COST EbTIMATES FOR THE DIS, NUREG'CR-5144; ACOUSTP EMISSION SYSTEM CALIBR4 TION AT POSAL OF RADIOACTIVE WASTES NUREG/CR4918 V02: CONTROL OF WATER INFILTRATION INTO WATTS BAR UNIT 1 NUCLEAR REACTOR. NEAR SURFACE LLW DISPOSAL UNITS. Task Report - A Discussion. Reactw Pressure Vessel Radiological Release NUREG/CR4508: BEHAVIOR OF A CORIUM JET IN HIGH PRESSURE NUREG/CR4881: FISSION PRODUCT RELEASE CHARACTERISTICS MELT EJECTION FROM A REACTOR PRESSURE VESSEL INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCI-DENT CONDITIONS. Reactor Protective System NUREG/CR4933: CLOSEOUT OF IE BULLETIN BO 19 FAILURES OF Radlonuchde MERCURY-WETTED MATRIX RELAYS IN REACTOR PROTECTIVE NUREG/CR-2850 V07: POPULATION DOSE COMMITMENTS DUE TO SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES BY COMBUSTION ENGINEERING. IN 1985. NUREG/CR4708 V02: PROGRESS IN EVALUATION OF RADIONU- Reactor Rehability CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- NUREG/CR4639 V02: NUCLEAR COMPUTERIZED LIBRARY FOR AS-i LEVEL NUCLEAR W ASTE REPOSITORY SITE PROJECTS. Report For SESSING REACTOR RELIABILITY (NUCLARR). Programmer's Guide. April 1986 September 1987.

  '      NUREG/CR-5047: RADIONUCLIDES ACCUMULATION BY AOUATIC                                   Reactor Hlsk                                                     l BIOTA EXPOSED TO CONTAMINATED WATER IN ARTIFICIAL ECO-                                NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE SYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH THE                                        DRAFT REACTOR RISK REFERENCE DOCUMENT,NUREG 1150.

GROUND. NUREG'CR 5170. A REVIEW OF HESEARCH CONDUCTED BY LOS Ructor Sam ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHASIS l ON THE MAXEY FLATS.KY, SHALLOW LAND WASTE BURIAL SITE' NUREG/CR-4805 V02: REACTOR SAFETY RESEARCH SEMIANNUAL ' REPORT July-December 1086. Volume 36. Radionuchde Rekase NUREG/CR-4914: THE INFLUENN OF SELECTED CONTAINMENT l NUREG 1228. SOURCE TERM ESTIMATION DURING INCIDENT RE. STRUCTURES ON DEBRIS DtSPtERSAL AND TRANSPORT FOLLOW-SPONSF TO SEVERE NUCLEAR POWE R PLANT ACCIDENTS. ING HIGH PRESSURE MELT EJECTION FROM THE REAC'OR l' NUREG/GR-5190 SUBMISSION FOR THE CSNI/GREST BENCHMARK VESSEL EXERCISE ON CHEMICAL THERMODYNAMIC MODELING IN CORE. NUREG/CR-5029: MELT PROGRESSION IN SEVERELY DAMAGED RE-CONCRETE INTERACTION FIELEASES OF RADIONUCLIDES ACTOR CORES. i NUREG/CR-5133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW l R2dsonuchde Transport Modeling THROUGH INTERGRANULAR STRESS CORROSION CRACKS. I NUREG/CR-4879 VD1. DEMONSTRATION OF PERFORMANCE MOD-ELING OF A L 3W-LEVEL WASTE SHALLOW-LAND BURIAL SITE.A Reactor Safety Research l Companson Of Tredictwe Radionucinde Transport Modehng Versus NUREG/CP 0091 V01: PROCEEDINGS OF THE FIFTEENTH WATER l Field Observations At The Nitrate Disposal Pit Site. Chalk Rwer Nuclear REACTOR SAFETY NFORMATION MEETING. l Le5s NUREG/CP-0091 V02. PROCEEDINGS OF THE FIFTEENTH WATER l REACTOR SAFETY INFORMATION MEETING. Reactor Accident NUREG/CP 0091 V03 PROCEEDINGS OF THE FIFTEENTH WATER NURE G-1228 SOURCE TERM ESTIMATION DURING INCIDENT RE- REACTOR SAFETY INFORMATION MEETING SPONSE TO SE VERE NUCLE AR POWER PLANT ACCIDENTS NUREG/CP-0091 V04 PROCEEDINGS OF THE FIFTEENTH WATER NUREG/CR-450B BEHAVIOR OF A CORIUM JET IN HIGH PRESSURE REACTOR SAFETY INFORMATION MEETING MELT EJECTION FROM A REACTOR PRESSURE VESSEL. NUREG /CP.0091 V04 ADD- PROCEEDINGS OF THE FIFTEENTH NUREG/CR 5029 MELT PROGRE SSION IN SEVERELY DAMAGED RE- WATER REACTOR SAFETY INFORMATION MEETING. l NUREG/CP-0091 VOS: PROCEEDINGS OF THE FIFTEENTH WATER l NURE /C 51 A SIMPLIFIED MODEL FOR CALCULATING EARLY REACTOR SAFTEY INFORMATION MEETING. l OF F SITE CONSEQUENCE S FROM NUCLEAR REACTOR ACCl-NUREG/CP-0091 V06 PROCEEDINGS OF THE FIFTEENTH WATER I A ^FET ^ N NUF G CR 5264- GUIDE FOR LICENSING EVALUATIONS USING g 03g 0 O S {EMy R SEARCH SEMlANNUAL CRAC2.A Computer Program For Calculating Reactor Accident Cons' ' REPORT. January-June 1987. Volume 37.

            #"""                                                                                  NUREG/CR-5039 V02: REACTOR SAFETY RESEARCH SEMIANNUAL Reactor Cavity                                                                              REPORT. July-December 1987. Reactor Safety Research Program.

NUREG/CR 4917. DCH-2 RESULTS FROM THE SECOND EXPERIMENT PERrOFIMED IN THE SURTSEY DIRECT HE ATING TES1 F ACILtTY. Reactor Safety Study NURE G/CR-St 46 DEBRIS DISFiRSAL FROM REACTOR CAVITIES NUREG/CR 4836. APPROACHES TO UNCERTAINTY ANALYSIS IN DURING HIGH PRESSURE MEL1 EJE CTION ACCIDENT SCENAR. PROBABILISTIC RISK ASSESSMENT, IOS Reactor Trip Bteaker Reactor Containment NUREG/CR-4935. CLOSEOUT OF IE BULLETIN 85-02.UNDERVOL-NUREG/CR-4917 DCH 2 RESULTS FROM THE SE COND EXPE RIMENT TAGE TRIP ATTACHMENTS OF WESTINGHOUSE DB-50 TYPE RE-P(RF ORMED IN THt. SURTSEY DIRE CT HEAllNG TEST F ACILITY ACTOR TRIP BRE AKERS ?

ia Subject index 10S Reactor Trip System NUREG 0090 V10 N04: REPORT TO CONGRESS ON ABNORMAL j NUREG/CR-4665- CLOSEOUT OF IE BULLETIN 83 08 ELECTRICAL OCCURRENCES October-December 1987, CIRCUlf BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE IN NUREG-0090 V11 N01: REPORT TO CONGRESS ON ABNORMAL USE IN SAFETY RELATED APPLICATIONS OTHER THAN THE RE- OCCURRENCES. January March 1988. i ACTOR TRIP SYSTEM. NUREG 0090 V11 NO2: REPORT TO CONGRESS ON ABNORMAL i NUREG/C44740: NUCLEAR PLANT-AGING FIESEARCH ON REACTOR OCCURRENCES AprlLJune 1988 PROTECTION SYSTEMS. NUREG 1260 V02: A REPORT TO CONGRESS ON NUCLEAR REGULA-TORY RESEARCH. Protect Desenptions For FY88. Reactor Vesnel NUREG/CR-4914: THE INFLUENCE OF SELECTED CONTAINMENT Repository Sealing STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW- NUREG/CR-5129: EXPERIMENTAL ASSESSMENT OF THE INFLUENCE ING HIGH PRESSURE MELT EJECTION FROM THE REACTOR OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF VESSEL DRIED CEMENT BOREHOLE SEALS. NUREG/C45207, FRACTURE EVALUATION OF SURFACE CRACKS NUREG/CR-5130: BENTONITE BOREHOLE PLUG FLOW TESTING i EMBEDDED IN REACTOR VESSEL CLADDING. Material Property Eval- WITH FIVE WATER TYPES. untions. Research Recovery Action NUREG-0975 vo6: COMPILATION OF CONTRACT RESEARCH FOR NUREG/CR 4834 V02: RECOVERY ACTIONS IN PRA FOR THE RISK THE MATERIALS ENGINEERING BRANCH, DIVISION OF  ; METHODS 8NTEGRATION AND EVALUATION PROGRAM ENGINEERING. Annual Report For FY 1987, (RMIEP) Volume 2. Application Of The Data-Based Method. Residual Heat Removal Redox Process NUREG/CR-5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE-NUREG/CR-5100. THERMODYNAMIC PROPERTIES OF TC(IV) MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF OXIDES. Solubilities And The Electrode Potential Of The Tc(Vil)/Tc(IV)- GENERIC ISSUE 99. Oxide Couple.  ; 1 Resin Waste Regulation NUREG/CR-5153: THE TEACHABILITY AND MECHANICAL INTEGRITY NUREG 1206 V02: NRC SAFETY RESEARCH IN SUPPORT OF REGU- OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED LATION - 1987-IN OFMENT AND VINYL ESTER-STYRENE. Regulatory Agenda Rwsin/ Liner investigation NUREG 0936 V06 N04: NRC REGJLATORY AGENDA.Ouarteny NUREG/CR 5229; ANNUAL REPORT OF THE TMI-2 EPICOR Il RESIN / NUR - 0 N01 NC REGULATORY AGENDA Ouartarty March 1988. og m For Fisca a 1988. Report0936 NURtG January v 07NO2: NRC REGULATORY AGENDA. Quar *,erly Resistive Layer NUR b V7 03: NRC REGULATORY AGENDA.Ouarterly W N 408 M M G M W M W W A W MO Report, July-September 1988. NEAR SURFACE LLW DISPOSAL UNITS. Task Report A Discussion. Regulatory Analysis Respirator Fit NUREG 1218 DRFT FC: REGULATORY ANALYSIS FOR PROPOSED NUREG/CR-5090: EFFECTS OF TEMPERATURE AND HUMIDITY ON RESOLUTION OF USl A 47- Safety implications Of Control RESPIRATOR FIT UNDER SIMULATED WORK CONDITIONS. Systems Draft Rept For Comment. Response Spectral Regulatory AndTechnicalReport NUREG/CR 3509: POWER SPECTRAL DENSITY FUNCTIONS COM-NUREG-0304 V12 N04. REGULATORY AND TECHNICAL REPORTS PATIBLE WITH NRC REGULATORY GUIDE 1.60 RESPONSE SPEC-(ABSTRACT INDEX JOURNAL) Annual Compilation For 19Br TRA. NUREG 0304 V13 N01: REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL) Compilation For First Quarter Restart i 1988, January-March NUREG 1286 S01: SAFETY EVALUATION REPORT RELATED TO THE l NUREG-0304 V13 NO2: REGULATORY AND TECHNICAL HEPORTS RESTART OF RANCHO SECO NUCLEAR GENERATING (ABSTRACT INDEX JOURNAL) Compilation For Second Quarter STATION. UNIT 1.FOLLOWING THC EVENT OF DECEMBER 1988, April-June. 26,1985 Docket No. 50-312.(Sacramento Municipal Utility Distnct) NUREG 0304 V13 NO3 REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL). Compilation For Third Quarte, Alprap 1988, July-Soptember. NUREG-1263. HYDROLOGIC DESIGN FOR RIPRAP ON EMBANKMENT SLOPES. Reinforced Concrete NUREG/CR 5083. DESIGN. CONSTRUCTION AND INSTRUMENTATION Riprap Design Criteria OF A 1/6-SCALE REINFORCED CONCRETE CONTAINMENT BUILD. NUREG/CR4E51 V02. DEVELOPMENT OF RIPRAP DESIGN CRITERIA ING BY RIPRAP TESTING IN FLUMES. Phase ll. Followup Investigations. NUREG/CR 6099 EVALUATION OF MATERIALS OF CONSTRUCTION FOR THE REINFORCED CONCRETE REACTOR CONTAINMENT Riprap Testing MODEL. NUREG/CR-4V.,1 V02: DEVELOPMENT OF RIPRAP DESIGN CRITERIA NUREG/CR-5182: THE SEISMIC CATEGORY l STRUCTURES BY RIPRAP TESTING IN FLUMES Phase ll Followup investigations. PROGRAM.Results For FY 1986 Risk Analysis Fielay NUREG/CR-5058: PRA APPLICATIONS FROGRAM FOR INSPECTION , NUREG/CR-4933. CLOSEOUT OF IE BULLETIN 8019. FAILURES OF AT ARKANSAS NUCLEAR ONE UNIT 1 Docket No. 50 313.(Arkansas l MERCURY-WETTED MATFilX RELAYS IN REACTOR PROTECTIVE Power And Light Company) i SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED l DY COMBUSTION ENGINEERING Ris,k Assessment NUREG/CR-5050: ANNOTATED BIBLIOGRAPHY OF RELIABILITY AND Reliability RISK DATA SOURCES. NUREG/CR4780 V01. PROCEDURES FOR TREATNG COMMON CAUSE F AILURES IN SAFETY AND RELIABILITY Risk Data Source STUDIE S Procedural Frameworit And Examples NUREG/CR 5050. ANNOTATED BIBLIOGRAPHY OF RELIABILITY AND NUREG/CR 6050. ANNOT ATED BIBLIOGRAPHY OF WETTABILITY AND RISK DATA SOURCES. RISK DATA SOURCES Flisk Reduct6on Report To Congress NUREG/CR4551 V5 DRF. EVALUATION OF SEVERE ACCIDENT NUREG.0090 V10 N01 REPORT TO CONGRESS ON ABNORMAL MSKS AND POTENTIAL FOR RISK REDUCTION ZION POWER OCCURRE NCES July September 1987 PLANT, Draft Report For Comment

106 Subject index WUREG/CR.4999: ESTIMATION OF RISK REDUCTION FROM IM- Safeguard PROVED PORV RELIABILITY IN PWRS F nel Report NUREG.1328: USE OF PERIMETER ALARMS AT FUEL FABRICATION FACILITIES USING OR POSSESSING FORMULA OUANTITIES OF Rod Buradle STRATEGIC SPECIAL NUCLEAR MATERIAL. NUREG/CR-5095 V01: THERMODYNAMIC NONEOUILIBRIUM IN POST

• NUREG-132p; ENTRY / EXIT CONTROL AT FUEL FABRICATION FACIL1-CRITICAL HEAT FLUX BOILING IN A ROD BUNDLE Desenption Of TIES USING OR POSSESSING FORMULA QUANTITIES OF STRATE-Expenments And Sample Results GIC SPECIAL NUCLEAR MATERIAL.

NUREG/CR b095 V02. THERMODYNAMIC NONEOUILIBRIUM IN POST- NUREG-1330: PERSONNEL AND VEHICLE BARRIERS AT FUEL FABn: CRITICAL-HEAT-FLUX DOILING IN A ROD BUNDLE. Data For Stat

• CATION FACILITIES USING OR POSSESSING STRATEGIC QUANTI-14 red Quench Front Tests. TIES OF SPECIAL NUCLEAR MATER 6AL.

NUREG/CR-5095 V03:1 THERMODYNAMIC NONEOUluBRIUM IN POST-CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Advanc- Safeguards Event sng Quench Front Tests. NUREG 1304: REPORTING OF SAFEGUARDS EVENTS. NUREG/CR 5095 V04. THERMODYNAMIC NONEOUILlBRIUM IN POST. CRITICAL-HEAT. FLUX BOILING IN A ROD BUNDLE. Data For Retreat- Safeguards Summary Event List ing Quench Front Tests NUREG-0525 R14: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL). Rulemaking Safety NUREG/CP-0099. PROCEEDINGS OF THE PUBLIC WORKSHOP FOR NUREG/CR-4780 V01: PROCEDURES FOR TREATING COMMON NRC RULEMAKING ON MAINTENANCE OF NUCLEAR POWER CAUSE FAILURES IN SAFETY AND RELIABILITY PLANTS. STUDIES. Procedural Framework And Examples. NUREG/CR-4805 V02: REACTOR SAFETY RESEARCH SEMIANNUAL Rules REPORT. July-December 1986 Volume 36. NUREG-0936 V06 N04. NRC REGULATORY AGENDA.Ouarterly Report,0ctober-December 1987. Safety Analysis NUREG4Ar36 V07 N01: NRC REGULATORY AGENDA.Ouarterly NUREG 0800 02.4.2 R3: STANDARD REVIEW PLAN FOR THE REVIEW Reper', January-March 1988. OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG-0936 V07 NOP: NRC REGULATORY AGENDA Ouarterly PLANTS. LWR Edition. Proposed Revision 3 To SRP Section 2 4.2, Report.ApnLJune 1988- " Floods." For Comment. NUREG 0036 V07 NO3: NRC REGULATORY AGENDA OuartertY NUREG 0800 02.4.3 R3: STANDARD REVIEW PLAN FOR THE REVIEW Report. July September 1988- OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR-0130 ADD 04: TECHNOLOGY, SAFETY AND COSTS OF PLANTS. LWR Edition. Proposed Revision 3 To SRP Section 2 4.3, DECOMMISSIONING A REFFRENCE PRESSURIZED WATER REAC* " Probable Maximum Flood (PMF) On Streams & Rivers." For Com-TOR POWER STATION Technical Support For Decommissionmg Mat- ment. ters Related To Preparation Of The Final Decommissic.iing Rule. NUREG/CR 0672 ADD 03 TECHNOLOGY. SAFETY AND COSTS OF Safety Assessment DECOMMISSIONING A REFERENCE DOILING WATER REACTOR NUREG/CR 5039 V01: REACTOR SAFETY RESEARCH SEMlANNUAL POWER STATION. Technical Support For Decommissioning Matters REPORT. January June 1987. Volume 37. Related To Preparation Of The Final Decommissioning Rule. Safety Evaluation Report Rules Of Practice NUREG-0781 S05: SAFETY EVALUATION REPORT RELATED TO THE  ! NUREG-0386 D04 R07. UNITED STATES NUCLEAR REGULATORY OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July Nos. 50-498 And 50-499. { Houston Lighting And Power Company) 1972 . March 1987. NUREG 0797 S14: SAFETY EVALUATION REPORT RELATED TO THE NUREG-0386 004 ROB. UNITED STATE 3 NUCLEAR REGULATORY OPERATION OF THE COMANCHE PEAK STEAM ELECTRIC COMMISSION STAFF PRArTICE AND PROCEDURE DIGEST. July STATION. UNITS 1 AND 2. Docket Nos 50 445 And 50-446,(Texas Utilo 1972 June 1987. ties Generating Company) NUREG 0386 D04 R09 UNITED STATES NUCLEAR REGULATORY NUREG 0797 S15: SAFETY EVALUATION REPORT RELATED TO THE COMMISSION ST AFF PRACTICE AND PROCEDURE DIGEST. July OPERATION OF COMANCHE PEAK STEAM ELECTRIC

                                                                                                                                                       ^         "              '

NU G 03 6 UNITED STATES NUCLEAR REGULATORY s e n mp y COMMISSION STAFF PRACTICE AND PROCEDURE NUREG 0797 $16. SAFETY EVALUATION REPORT RELATED TO THE DIGEST.Commmsion, Appeal Board And Licensing Board Decisions. July OPERATION OF COMANCHE PEAK STEAM ELECTRIC 1972 December 1987, STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-446.(Texas Utile bes Generating Company) SAFT-UT NUREG 0797 S17: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-5075. THE SAFT UT REAL-TIME INSPECTION SYSTEM ^ OPERATION OF COMANCHE PEAK STEAM ELECTRIC OPERATIONAL PRINCIPLES AND IMPLEMENTATION. STATION. UNITS 1 AND 2. Docket Nos 50-445 And 50-446.(Texas Utile ties Generating Company) SALP NUREG 0797 618; SAFETY EVALUATION REPORT RELATED TO THE NUREG 1214 R03 HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT. OPERATION OF COMANCHE PEAK STEAM ELECTRIC IC ASSESSMENT OF LICENSEE PERFORMANCE' STATION. UNITS 1 AND 2. Docket Nos 50-445 And 50-446.(Texas Utile SCALE ties Generating Company) NUREG-0797 S19: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR 5033

SUMMARY

DESCRIPTION OF THE SCALE MODU-LAR CODE SYSTEM OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2. Docket Nos. 50 445 And 50 446,(Texas Utils SCDAP/RELAP5 lies Generating Company) NUREG/CR.5214- ANALYSES OF NATURAL CIRCULATION DURING A NUREG 0797 S20: SAFETY EVALUATION REPORT RELATED TO THE SURRY STATION BLACKOUT USING SCDAP/RELAP5 OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION.UN!TS 1 AND 2. Docket Nos 50 445 And 50-446.(Texas Utili-SHORTCUT ties Generating Company) NUREG/CR-5242; A FAST BOTTOM.UP ALGORITHM FOR COMPUT- NUREG 1002 S06: SAFETY EVALUATION REPORT RELATED TO THE ING THE CUT SETS OF NONCOHERF NT F AULT TREES OPERATION OF BRAIDWOOD STATION. UNITS 1 AND 2. Docket Nos. 50 456 And 50 457.(Commonwealth Edison Company) SLITP NUREG-1137 S07. SAFETY EVALUATION REPORT RELATED TO THE NUREG-1309. THE U S NUCLEAR RE GULATORY COMMISSION PRO- OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 GRAM WITH STATE AND LOCAL GOVERNMENTS AND INDIAN AND 2. Docket Nos. 50 424 And 50-425.(Georgia Power Co.et al) TRIBES. NUREG-1231 S01: SAFETY EVALUATION REPORT RELATED TO BAB-COCK AND WILCCX OWNERS GROUP PLANT REASSESSMENT SabotaDe PROGRAM. NURE G.1178 VITAL L OUIPMENT/ AREA GUIDELINES STUDY. Vital NUREG 1232 V02. SAFETY EVALUATION REPORT ON TENNESSEE Area Committee ROPortfmal Report VALLEY AUTHORITY. Sequoyah Nuclear Performance Plan

t 1 i Subject index 107 NUREG 1283- SAFETY EVALUATION REPORT RELATED TO THE RE-

                         .                                                         NUREG/CR-5123; STUDIES OF THE PATTERN AND AGES OF POST-                   i NEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REAC-                    METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND                     j 10R AT PURDUE UNIVERSITY.                                                 NORTH CAROLINA.                                                       '

i NUREG 1286 S01: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR 5154: EXPERIMENTAL ASSESSMENT OF DAMPING IN RESTART OF RANCHO SECO NUCLEAR GENERATING LOW ASPECT RATIO, REINFORCED CONCRETE SHEAR WALL i' STATION. UNIT 1,FOLLOWING THE EVENT OF DECEMBER STRUCTURES. . 26.1985. Docket No. 50 312 (Sacramento Municipal Utility District) NUREG/CR-5182: THE SEISMIC CATEGORY I STRUCTURES 1 NUREG-1313. SAFETY EVALUATION REPORT RELATED TO THE PROGRAM.Results For FY 1986. EVALUATION OF LOW-ENRICHED URANIUM SILICIDE ALUMINUM DISPERSION FUEL FOR USE IN NON POWER REACTORS, Seismic Category i NUREG/CR-4924: SEISMIC CATEGORY I STRUCTURES Safety Goal PROGRAM.F nat Report. Fiscal Year 1983 1984. $ NUREG 1226. DEVELOPMENT AND UTILIZATION OF THE NRC NUREG/CR-4998. THE SEISMIC CATEGORY I STRUCTURES- J POLICY STATEMENT ON THE REGULATION OF ADVANCED NU- PROGRAM.Results For Fiscal Year 1985. i CLEAR POWER PLANTS. I l Seismic Design . l Safety leeue NUREG/CR-5209: DESIGN PROVISIONS FOR TANGENTIAL SHEAR IN NUREG/CR-5039 V01: REACTOR SAFETY RESEARCH SEMIANNUAL CONTAINMENT WALLS.- 1-REPORT. January-June 1987. Volume 37. NUREG/CR-5039 V02: REACTOR SAFETY RESEARCH SEMIANNUAL Seismic Design Criteria REPORT. July-December 1987. Reactor Safety Research Program. NUREG 1233 DRFT FC: REGULATORY ANALYSIS. FOR USl A 40J

                                                                                     " SEISMIC DESIGN CRITERIA." Draft Report For Comment NUREG 1266 V02: NRC SAFETY RESEARCH IN SUPPORT OF REGU-                Seismic Fragmty LATION 1987-                                                          NUREG/CR-4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER Safety Research Program                                                         PLANT        COMPONENTS           (PHASE       II). Motor    Control Center, Switchboard,Panelboard And Power Supply.

NUREG 1325: DISPOSITION OF RECOMMENDATIONS OF THE NA. TIONAL RESEARCH COUNCIL IN THE REPORT "REVITAUZING NU- Seismic Hazard CLEAR SAFETY RESEARCH." NUREG/CR-5042 Sot: EVALUATION OF EXTERNAL HAZARDS TO NU. iNUREG/CR-2331 V7N2 3: SAFETY RESEARCH PROGRAMS SPON- , SORED BY OFFICE OF - NUCLEAR CLEAR POWER PLANTS IN THE UNITED STATES Seismic Hazard. 1 REGULATORY ' RESEARCH. Progress Report, April September 1987. Seismic Margin NUREG/CR-2331 V8N12: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR NUREG/CR-5073: OUANTIFICATION OF MARGINS IN PIPING SYSTEM REGULATORY SEISMIC RESPONSE. Methodologies And Damping. RESEARCH. Progress Fleport, January June 1988. NUREG/CR-5076: AN APPROACH 10 THE QUANTIFICATION OF SEIS-MIC MARGINS IN NUCLEAR POWER PLANTS.The importance Of UE NRC F BWR Plant Systems And Functions To Seismic Margins. S IFICANCE ASSESSMENT TEAM REPORT ON ALLEGATIONS RELATED TO THE SOUTH TEXAS Setem6c Qualif6 cation PROJEOT UNITS 1 & 2. NUREG/CR-5031: SIGNIFICANCE OF IN STRUCTURE GENERATED MOTION IN SEISMIC QUALIFICATION TESTS OF CABINET MOUNT-UE 4747 V02: AN AGlNG FAILURE SURVEY OF LIGHT ED EmmM MCES. WATER REACTOR SAFETY SYSTEMS AND COMPONENTS Seism 6c Response l NUREG/CR 4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE

                                                                                , NUREG/CR 5023: HIGH-LEVEL SELSMIC RESPONSE AND FAILURE SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER PLANTS. Operating Exponence And Failure identification.                  PREDICTION METHODS FOR PIPING.

NUREG/CR 5147; FUNDAMENTAL ATTRIBUTES OF A PRACTICAL Seismicity s, + CONFIGURATION MANAGEMENT PROGRAM FOR NUCLEAR PLANT DESIGN CONTROL NUREG/CR 3145 V06: GEOPHYSICAL INVESTIGATIONS OF"THE WESTERN OHIO-INDIANA REGION. Annual Report.Octohar 1986 4 Safety-Related System September 1987. NUREG/CR 5053: OPERATING EXPERIENCE AND AGING ASSESS- NUREG/CR 5045: KANSAS NEBRASKA SEISMICITY STUDIES USING MENT CF MOTOR CONTROL CENTERS. THE KANSAS NEBRASKA ^ MICROEARTHOUAKE NETWORK. Final Report. Sampling NUREG/CR-5080: A STUDS OF NEW ENGLAND SEISMICITY WITH NUREG/CR-5212; EMERGENCY ENVIRONMENTAL SAMPUNG AND EMPHASIS ON MASSACHUSETTS AND NEW HAMPSHIRE. Final ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES R T Pe NUR C Qn g \985. 1 INVESTIGATION OF EARTH-Scintillation Fiber Detector OUAKES IN THE NEW MADRID SEISMIC ZONE AND THE NORTH-NUREG/CR-5223 SCINTILLATION FIBER DETECTOR FOR IN-VIVO EASTERN EXTENT OF THE NEW MADRID SEISMIC ZONE. Final ENDOSCOPIC INTERNAL DOSIMETRY, Report. September 1981 - December 1986. NUREG/CR-5258 V01: GEORGIA / ALABAMA REGIONAL SEISMO-Scoping Test GRAPHIC NETWORK. Annual Report August 1985 - June 1986. NUREG/CR 5119 METALLOGRAPHIC EXAMINATION OF THE SEVERE FUEL DAMAGE SCOPING TEST (SFD-ST) FUEL ROD BUNDLE Seismographic Network CROSS SECTIONS. NUREG/CR-5258 V01: GEORGIA / ALABAMA REGIONAL SEISMO. , GRAPHIC NETWORK. Annual Report August 1985 - June 1986. I Scram D6scharge NUREG/CR-5190: CWSEOUT OF IE BULLETIN 8014. DEGRADATION Semiscale Mod-2C OF BWR SCRAM DISCHARGE VOLUME CAPABILITV. NUREG/CR 4898 RESULTS OF SEMISCALE MOD-2C FEEDWATER l AND STEAM UNE BREAK (S-FS) EXPERIMENT SERIES. Bottom Main ' Seismic Feedwater Une Break Accident Expenments. NUREG/CR 4924 GEISMIC CATEGORY l STRUCTURES NUREG/CR-4971: RESULTS OF SEMISCALE MODPC FEEDWATER PROGRAM. Final Reportfiscal Year 1983 1984. AND STEAM UNE BREAK (S FS) EXPERIMENT SERIES Main Steam NUREG/CR 4998: THE SEISMIC CATEGORY I STRUCTURES Une Break Accident Expenments. j PROGRAM Results For Fiscal Year 1985. i NUREG/CR-5012: SIMILARITY PRINCIPLES FOR EQUIPMENT OUAu- Service Water System FICATION BY EXPERIENCE. l NUREG-1275 V03. OPERATING EXPERIENCE FEEDBACK REPORT - NUREGrCR.5031: SIGNIFICANCE OF IN-STRUCTURE GENERATED SERVICE WATER SYSTEM FAILURES AND MOTION IN SEISMIC QUAUFICATION TESTS OF CABINET MOUNT. DEGRADATIONS Commercial Power Reactors CD ELECTRICAL DEVICES. NUREG/CR-5210: TECHNICAL FINDINGS DOCUMENT FOR GENERIC NUREG/CR-5073' OUANTIFICATION OF MARGINS IN PIPING SYSTEM ISSUE 51: IMPROVING THE RELIABluTY OF OPENCYCLE SERVtCE-SEISMIC RESPONSE. Methodologies And Damping WATER SYSTEMS.

      . . ._ _ ________                                                                                                                                     J

i 108- Subject Index Severe Accident . NUREG/CR-2007 V07: RADIOACTIVE MATERIALS RELEASED FROM NUREG 1226. DEVELOPMENT AND UTILIZATION OF THE NRC NUCLEAR POWER PLANTS. Annual Report For 1986. POLICY ' STATEMENT ON THE REGULATION OF ADVANCED NU-- CLEAR POWER PLANTS Sorption . . NUREG 1252: NUCLEAR POWER PLANT THERMAL-HYDRAULIC PER- NUREG/CR-5108: THERMODYNAMIC PROPERTIES OF - TC(IV) FORMANCE RESEARCH PROGRAM PLAN OX1 DES. Solubilities And The Electrode Potential Of The Tc(Vil)/Tc(IV)- i l NUREG/CP 0095: PROCEEDINGS OF THE FOURTH WORKSHOP ON i Oxide Couple. 4 CONTAINMENT INTEGRITY l NUREG/CR 3908: SURVEY OF THE STATE OF THE ART IN MITIGA' Source Term NUREG 1228: SOURCE TERM ESTIMATION DURING INCIDENT REJ NL /CR 4 2 SURVEY OF LIGHT WATER REACTOR CONTAIN-SPONSE TO SEVERE NUCLEAR POWER PLANT ACCIDENTS. MENT SYSTEMS. DOMINANT FAILURE MODES AND MITlGATION NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS OPPORTUNITIES Fmal Report. NUREG/CR-4243: VALUE/ IMPACT ANALYSIS FOR EVALUATING AL. HANDBOOK. TERNATIVE MITIGATION SYSTEMS. NUREG/CR-4688 V02: QUANTIFICATION AND UNCERTAINTY ANALY-NUREG/CR 4244; STRATEGIES FOR IMPLEMENTING A MITIGATION SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT POLICY FOR LIGHT WATER REACTORS. WATER REACTORS (OVASAR).Part II: Sensitwity Analysis Tech-NUREG/CR 4688 V02: QUANTIFICATION AND UNCERTAINTY ANALY- nsques. SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT NUREG/CR-4997: METHODS FOR DESCRIBING AIRBORNE FRAC. WATER REACTORS (OUASAR).Part II: Sensitivity Analysis Tech- TIONS OF FREE FALL SPILLS OF POWDERS AND LIOUlDS. rwaves. . NUREG/CR-5204: LOW LEVEL RADIOACTIVE WASTE SOURCE TERM NUREG/CR 4881: F1SSION PRODUCT RELEASE CHARACTERISTICS MODEL DEVELOPMENT AND TESTING. INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCl-DENT CONDITIONS. Special Nuclear Material NUREG/CR-4920 V01: ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG-1328: USE OF PERIMETER ALARMS AT FUEL FABRICATION VENTION AND MITIGATION FEATURES:BWR, MARK i CONTAIN' FACILITIES USING OR POSSESSING FORMULA QUANTITIES OF 8T L L M NUREG/ 492 V02: ASSESSMENT OF SEVERE ACCIDENT PRE. NURE 13 E /EXI CO RO AT UEL FABRICATION FACILI-VEN 10 ND MITIGATION FEATURES:DWR. MARK 11 CONTAIN- TIES USING OR POSSESSING FORMULA QUANTITIES OF STRATE- t NUREG/GR-4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE. GIC SPECIAL NL CLEAR MATERIAL . VENTION AND MITIGATION FEATURES:BWR. MARK lil CONTAIN- Spent Fuel Pool MENT DESIGN. NUREG/CR-4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE. NUREG/CR 5048: REVIEW OF THE NATURAL CIRCULATION EFFECT VENTION AND MITIGATION FEATURES:PWR,LARGE DRY CON. IN THE VERMONT YANKEE SPENT FUEL POOLDocket.No. 50-TAINMENT DESIGN 271.(Vermont Yankee Nuclear Power Corp) NUREG/CR 4920 V05: ASSESSMENT OF SEVERE ACCIDENT PRE. . VENTION AND MITIGATION FEATURES.PWR,1CE CONDENSER Spent Fuel Shipmert Route  ! CONTAINMENT DESIGN NUREG 0725 R06: PUBLIC INFORMATION CIRCULAR FOR SHIP-NUREG/CR-5042 Sct: EVALUATION OF EXTERNAL HAZARDS TO NU- MENTS OF IRRADIATED REACTOR FUEL . CLEAR POWER PLANTS IN THE UNITED STATES.Seismac Hazard. NUREG/CR 5096: EVALUATION OF SEALS FOR MECHANICAL PENE- Spent Fuel Storage NURE / 5 2 EE A DE S TS REPORT, HANDBOOK. NUREG/CR-5157: THE DEVELOPMENT OF APRILMOD2. A COMPUT-ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL- g m g ,p, ING WATER NUCLEAR REACTORS. I UAEG/CR-5255: STABLE ISOTOPES OF AUTHIGENIC MINERALS IN Severe Accident Risk VARIABLY SATURATED FRACTURED TUFF. NUREG/CR-5000: METHODOLOGY FOR UNCERTAINTY ESTIMATION IN NUREG 1150 (DRAFT). Conclusions Of A Review Panet. Stainless Steel NUREG 0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION Shallow Land Burial AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE NUREG/CR-5204: LOW LEVEL RADIOACTIVE WASTE SOURCE TERM BOUNDARY PIPING. Final Report. MODEt DEVELOPMENT AND TESTING. NUREG/CR 5024: TENSILE AND J-R CURVE CHARACTERIZATION OF O bb b' Shear Wall Structure NUREG/CR 5154: EXPERIMENTAL ASSESSMENT OF DAMPING IN Stainless Steel Piping LOW ASPECT RATIO, REINFORCED CONCRETE SHEAR WALL NUREG/CR-5133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW STRUCTURES. THROUGH INTERGRANULAR STRESS CORROSION CRACKS. Shipping Package UR G 0800 2 2 R3: STANDARD REVIEW PLAN FOR THE REVIEW DURES F R S IPP G PACKAGES OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Silicide-Uranium Disperaion Fuel PLANTS. LWR Edition Proposed Revision 3 To SRP Section 2.4.2, NUREG 1313: SAFETY EVALUATION REPORT RELATED TO THE " Floods." For Comment. EVALUATION OF LOW-ENRICHED URANIUM SILtCIDE ALUMINUM NUREG-0000 02.4.3 R3. STANDARD REVIEW PLAN FOR THE REVIEW DISPERSION FUEL FOR USE IN NON-POWER REACTORS. OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS LWR Edition. Proposed Revision 3 To SRP Section 2.4.3, Simulated Decontamination " Probable Maximum Flood (PMF) On Streams & Rws." For Com- l NUREG/CR 5153: THE TEACHABILITY AND feECHANICAL INTEGRITY ment. . OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED NUREG 1200 RO1: STANDARD REVIEW PLAN FOR THE REVIEW OF A IN CEMENT AND VINYL ESTER-STYRENE. LICENSE APPLICATION FOR A LOW LEVEL RADIOACTIVE WASTE g,g, DISPOSAL FACILITY. NUREG/CR-5094; APPLICATION OF STOCHASTIC METHODS TO THE Stat 6c Eliminator MULA OF LARGE. SCALE UNSATURATED FLOW AND NUREG/CR-5145: FAILURE INVESTIGATION OF 3M SERIES 900 STATIC ELIMINATORS. Soll-Structure interaction NUREG/CR 4784: INFLUENCE OF GROUNDWATER ON SOIL STRUC. Station Blackout TURE INTERACTION. NUREG 1032: EVALUATION OF STATION BLACKOUT ACCIDENTS AT NUCLEAR POWER PLANTSTechrucal Findings Related To bare. Solid Weste D6sposal solved Safety issue A-44. Final Report. NUREG/CR-2907 V06: RADIOACTIVE MATERIALS RELEASED FROM NUREG-1109. REGULATORY /BACKFIT ANALYSIS FOR THE RESOLU-NUCLE AR POWER PLANTS Annual Report For 1985. TION OF UNRESOLVED SAFETY ISSUE A-44. STATION BLACKOUT.

Subject index 109 NUREG/CR-5078 V01: A RELIABILITY PROGRAM FOR EMERGENCY Structural Integrity DIESEL GENERATORS AT NUCLEAR POWER PLANTS. Program Structure NUREG/CR 4219 V05 N1: HEAVY SECTION STEEL TECHNOLOGY PROGRAM. Semiannual Progress Report For October 1987 - March NUREG/CR-5078 V02: A RELIABILITY PROGRAM FOR EMERGENCY 1983, DIESEL GENERATORS AT NUCLEAR POWER PLANTS Maintenance Surveillance And Condition Monsionng. Structural Steel Statistical Method NUREG/CP0064. SECOND CNSI WORKCHOP ON DUCTILE FRAC-TURE TEST METHODS. NUREG/CR 4605: TRAINING MANUAL ON STATISTICAL METHODS FOR NUCLEAR MATERIAL MANAGEMENT. Structural Stiffness Steam Air Environment NUREG/CR 492C SEISMIC CATEGORY t STRUCTURES PROGRAM Final Report, Fiscal Year 1983 - 1984. NUREG/CR-5018: URANIUM OXIDE-IRON OXIDE MIXED AEROSOL EXPERIMENTS IN STEAM AIR ATMOSPHERES.NSPP Tests NUREG/CR 4990: THE SE1MIC CATEGORY I STRUCTURES E11,612,613 And 631. Data Record Report PROGRAM.Results For Fiscal Year 1985-Structural integrtty NJREG 844 NRC INTEGRATED PROGRAM FOR THE RESOLUTION NUREG/CR 4219 V04 N2: HEAVY-SECTION STEEL TECHNOLOGY OF UNRESOLVED SAFETY ISSUES A-3 A 4 AND A 5 REGARDING PROGRAM Semiannual Progress Report For Apr61-September 1987, STEAM GENERATOR TUBE INTEGRITY. Final Report. NUREG-1236: THERMAL HYDRAULIC RESEARCH PLAN FOR BAB- Substance Abuse COCK AND WILCOX PLANTS NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG-1332: REGULATORY ANALYSIS FOR THE RESOLUTION OF INDUSTRY.A Review Of Technical lasues. GENERIC ISSUE 125.II.7 " REEVALUATE PROVISION TO AUTO-Subsurface-Complexation Modeling MATICALLY ISOLATE F,EEDWATER FROM STEAM GENERATOR NUREG/CR 4807: SURFACE-COMPLEXATION MODELING OF RADIO-NURE R 1 T E FREQUENCY EDDY-CURRENT INSTRU- NUCLIDE ADSORPTION IN SUBSURFACE ENVIRONMENTS. MENT EXP E CE Update For 1984-1986. N R G/C 4857: CADET:A DECISION SUPPORT SYSTEM FOR LIGHT NUREG/CR-5178. EVALUATION OF GENERIC ISSUE WATER REACTOR SAFETY. 125.II.7, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE l Surveillance Capsule FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. NUREG/CR-5185: STEAM GENERATOR GROUP PROJECT. Task 13 NUREG/CR-4947: ANALYSIS OF THE A3020 AND A5338 STANDARD Final Report: Nondestructive Examination (NDE) Validation. REFERENCE MATL' RIALS IN SURVEILLANCE CAPSULES OF COM-MERCIAL POWER REACTORS. Stesm Generator Tube NUREG/CR-2336: STEAM GENERATOR TUDE INTEGRITY Surveillance Program PROGRAMcPhase ll Final Report NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW Steam Generator Tubing CRITERLA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIOAC-TiVE VJASTE DISPOSAL Environmental Monitanng And Surveillance NUREG/CR 5016: COMPENDIUM AND COMPARISON OF INTERNA. Programs. TIONAL PRACTICE FOR PLUGGING, REPAIR AND INSPECTION OF STEAM GENERATOR TUBING. Surveillance Test Interval Steam Line Break NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT NUREG/CR-4898: RESULTS OF SEMISCALE MOD 2C FEEDWATER AND ST) REQUIREICNTS AT THE ANO 1 NUCLEAR POWER PLANT. AND STEAM LINE BREAK (S-FS) EXPERIMENT SERIES. Bottom Main Feedwater Line Break Accident Expenments. Switchboard NURE G/CR-4071: RESULTS OF SEMISCALE MOD-2C FEEDWATER AND STEAM LINE BREAK (S FS) EXPERIMENT SERIES. Main Steam NUREG/CR-4659 V02. SEISMIC FRAGILITY OF NUCLEAR POWER PLANT COMPONENTS (PHASE II) Motor Control Line Break Accident Expenments. Center, Switchboard,Panelboard And Power Supply. Steam Oxidation Systematic Assessment NUREG/CR-4777: STEAM OXIDATION OF ZlRCALOY CLADDING IN THE ORNL FISSION PRODUCT RELEASE TESTS NUREG 1214 R03: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-IC ASSESSMENT OF LICENSEE PERFORMANCE. NUREG-1214 R04: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-N REG /CR 5063: DEVELOPMENT OF A MECHANISTIC UNDER- IC ASSESSMENT OF LICENSEE PERFORMANCE. l STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES-SURE VESSEL STEELS. Final Report Systematic Evaluation Program NUREG-0822 S01: INTEGRATED PLANT SAFETY ASSESSMENT SYS-Stochastic Method TEMATIC EVALUATION PROGRAM OYSTER CREEK. NUCLEAR NUREG/CR 5094: APPLICATION OF STOCHASTIC METHODS TO THE GENERATING STATION. Docket No. 50-219(General Public Utilities SIMULATION OF LARGE-SCALE UNSATURATED FLOW AND Corporation And Jersey Central Power And Light Company) TRANSPORT. TAC 2D Stress Corrosion NUREG/CR-5126- TAC 2D STUDIES OF MARK 1 CONTAINMENT NUREG-0313 R02: TECHNICAL REPORT ON MATERIAL SELECTION DRYWELL SHELL MELT THROUGH. AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE BOUNDARY PIPING. Final Report TA NUREG/CR 5106: USER'S GUIDE FOR THE TACT 5 COMPUTER CODE. Stress Corrosion CracMnD NUREG-1222. PIPING RESEARCH PROGRAM PLAN. TIRGALEX NUREG/CR 2336: S1EAM GENERATOR TUBE INTEGRITY PROGRAM Phase il Final Report. NUREG/CR-5248. PRIORITIZATION OF TIRGALEX-RECOMMENDED COMPONENTS FOR FURTHER AGING RESEARCH. NUREG/CR-4667 V04 ENVIRC*1 MENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS Semiannual Repof1 October 1916 March TLD 1967 NUREG-0837 V07 N04 NRC TLD DIRECT RADIATION MONITORING NUREG/CR-5158 VOI- WORLDWIDE ACTIVITIES ON THE REDUCTION NETWORK Progress Report. October-December 1987. OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS NUREG 0837 V0B N01: NRC TLD DIRECT RADIATION MONITORING NUREG/CR 5180. CHEMICAL DECONTAMINATION AND CHEM! CAL NETWORK. Progress Report January-March 1988. CLEANING OF LWR COMPONENTS AND POSSIBLE INTERACTIONS WITH METALLURGICAL AGING EFFECTS NUREG-0837 V0B NO2. NRC TLD DlHECT RADIATION MONITORING NETWORK. Progress Report. Arni-Jur,e 1988.

110 Subject Index TORT . NUREG/CR-5095 V02: THERMODYNAMIC NONEOUILIBRIUM IN POST. NUREG/CR-4984: DEVELOPMENT OF A THREE DIMENSIONAL FLUX CRITICAL HFAT FLUX BOILtNG IN A ROD BUNDLE. Data For Stabi-SYNTHESIS PROGRAM AND COMPARISON WITH 3-D TRANSPORT lized Quench Front Tests. THEORY RESULTS, NUREG/CR 5005 V03. THERMODYNAMIC NONEOUILIBRIUM IN POST-

                                                                                                  ' CRITICAL HEAT. FLUX BOILING IN A ROD BUNDLE. Data For Advanc.

TRAC ang Quench Front Tests. NUREG/CR-5071: TRAC SUPPORT SOFTWARE. NUREG/CR-5095 V04: THERMODYNAMIC NONEOUILIBRIUM IN POST-CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE. Data For Retreat-N EG CR 5 9' DESIGN PROVISIONS FOR TANGENTIAL SHEAR IN N FE R5 RMODYNAMIC PROPERTIES OF ' TC(IV) CONTAINMENT WALLS. OXIDES. Solubilities And The Electrode Potential Of The Tc(Vil)/Tc(IV)- Technetium Oxde Couple. NUREG/CR-5108. THERMODYNAMIC PROPERTIES OF TC(IV) NUREG/CR$194: RELAPS/ MOD 2 MODELS AND CORRELATIONS. OXIDES.Solubaltties And The Electrode Potential Of The Tc(Vil)/Tc(IV). NUREG/CR-5196: SUBMISSION FOR THE CSNI/GREST BENCHMARK Oxide Couple' EXERCISE ON CHEMICAL THERMODYNAMIC MODELING IN CORE.

                                                                                                   ' CONCRETE INTERACTION RELEASES OF RADIONUCLIDES.

Techn6 cal Specification i NUREG-1305: TECHNICAL SPECIFICATIONS FOR SOUTH TEXAS Thermoluminescent PROJECT, UNIT 1. Docitet No. 50 498.(Houston LightinD And Power NUREG 0837 V07 N04: NRC TLD DIRECT RADIATION MONITORING Company) NETWORK. Progress Report. October-December 1987. Tectonics Thermoluminescent Doe 4 meter NUREG/CR 3145 V06. GEOPHYSICAL INVESTIGATIONS OF - THE NUREG 0837 V00 N01: NRC TLD DIRECT RADIATION MONITORING WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1986 - NETWORK. Progress Report. January-March 1988. September 1987. NUREG 0837 V08 NO2: NRC TLD DIRECT RADIATION MONITORING 1 NUREG/CR-5045: KANSAS-NEBRASKA SEISMICITY STUDIES USING NETWORK. Progress Report. April-June 1988. j THE KANSAS-NEBRASKA MICROEARTHOUAKE NETWORK Final Report. Three-D6mensional Fluu Synthesis Program NUREG/CR.5123: STUDIES OF THE PATTERN AND AGES OF POST- NUREG/CR 4984: DEVELOPMENT OF A THREE-DIMENSIONAL FLUX METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND SYNTHESIS PROGRAM AND COMPARISON WITH 3-D TRANSPORT NORTH CAROLINA. THEORY RESULTS. Tee Elbow Joint Through Wall Crack NUREG/CR5156: REVIEW OF EROSION CORROSION IN SINGLE . NUREG/CP 0075: PROCEEDINGS OF CSNt/NRC WORKSHOP ON PHASE FLOWS. DUCTILE PIPING FRACTURE MECHANICS. TeneNo NUREG/CR-5024. TENSILE AND J-R CURVE CHARACTERIZATION OF Title List THERMALLY AGED CAST STAINLESS STEELS NUREG 0540 V09 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. November 1-30.1987. Thermal Gradient Tube NUREG 0540 V09 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY l NUREG/CR 4778: PRELIMINARY STUDIES OF THE MORPHOLOGY OF AVAILABLE. December 1-31,1987 i THERMAL GRADIENT TUBE DEPOSITS FROM FISSION PRODUCT NUREG-0540 V10 No1: TITLE LIST OF DOCUMENTS MADE PUBLICLY RELEASE EXPERIMENTS' AVAILABLE. January 1 31.1988. NUREG 0540 V10 NO2: TITLE LIST OF DOCUMENTS MADE PUBLICLY Thermal Overload Protection . AVAILABLE. February 1-29.1988. s NUREG-1296: THERMAL OVERLOAD PROTECTION FOR ELECTRIC NUREG 0540 V10 Nc3: TITLE LIST OF DOCUMENTS MADE WBLICLY MOTORS ON SAFETY RELATED MOTOR OPERATED VALVES - GE. AVAILABLE. March 1 31.1988. NERIC ISSUE 11.E.6.1. NUREG-0540 V10 N04: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE April 1 30.1988. Thermal-Hydraulic Code NUREG-0540 V10 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY NUREG/CR-5095 VD1: THERMODYNAMIC NONEQUILIBRIUM IN POST- AVAILABLE. May 1 31.1988. CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE.Desenption Of NUREG-0540 V10 N06: TITLE LIST OF DOCUMENTS MADE PUBLICLY Expenments And Sample Results AVAILABLE. June 1 30.1968. NURLG/CR5095 V02: THERMODYNAMIC NONEOUILIBRIUM IN POST- NUREG-0540 V10 N07; TITLE LIST OF DOCUMENTS MADE PUBLICLY CRITICAL HEAT FLUX BOILING IN A ROD BUNDLE. Data For Statn- AVAILABLE. July 1-31.1988. lized Quench Front Tests NUREG 0540 V10 N08: TITLE LIST OF DOCUMENTS MADE PUBLICLY NUREG/CR-5095 V03: THERMODYNAMIC NONEQUILIBRIUM IN POST- AVAILABLE. August 1 31.1988. CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE. Data For Advanc- NUREG 0540 V10 N09. TITLE LIST OF DOCUMENTS MADE PUBLICLY eng Quench Front Tests AVAILABLE. September 1-30.1988. NUREG/CH 5095 V04 THERMODYNAMIC NONEOUILIBRIUM IN POST-CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Retreat- Toughness 6ng Quench Front Tests. NUREG/CP 0092: PROCEEDINGS OF THE SEMINAR ON LEAK- 1 EFORE-BREAK. Progress in Regulatory Policies And Supporting Re-Thermal Hydraul6c Research NUREG-1238: THERMAL HYDRAULIC RESEARCH PLAN FOR BAB-COCK AND WILCOX PLANTS Tra6ning And Regulatory Practices Thornal-HydrauHes 8% O E CW SWAW E-NUREG 1252: NUCLEAR POWER PLANT THERMAL-HYDRAULIC PER- 4NG ON TRAINING OF NUCLEAR RFACTOR PERSONNEL. Held At Orlando. Florida, April 2i-24,1987. FORMANCE RESEARCH PROGRAM PLAN NUREG/C44971: 9ESULTS OF SEMISCALE MOD'2C FEEDWATER T n AND STEAM LINL 9REAK (S.FS) EXPERIMENT SERLES Main Steam p L6ne Break Accident Eaenments FUtiDING THE NRC TRAINING PROGRAM FOR NUREG/CR 5135: THE THERMAL HYDRAULICS OF SOLOCAS RELA. STATES TIVE TO PRESSURIZED THERMAL SHOCK. NURE G/C45162: CHARM A MODEL FOR AEROSOL BEHAVIOR IN Transient 11ME VARYING THERMAL HYDRAULIC CONDITIONS. NUREG-1252: NUCLEAR POWER PLANT THERMAL-HYDRAULIC PER-FORMANCE RESEARCH PROGRAM PLAN. Thermodynam6c NUREG/CR-5071: TRAC SUPPORT SOFTWARE. NUREG/CR 4864 V01: THERMODYNAMIC TABLES FOR NUCLEAR WASTE ISOLATION. Aqueous Soluteons Database Trit 6um NUREG/CR-5095 V01. THERMODYNAMIC NONEOUILIBRIUM IN POST- NUREG/CR-5170. A REVIEW OF FIESEARCH CONDUCTED BY LOS CRillCAL-HEAT FLUX BOILING IN A FIOD BUNDLE.Desenption Of ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHASIS Expenments And Sample Results. ON THE MAXEY FLATS.KY, SHALLOW LAND WASTE BURIAL SITE.

Subject Index 111 Tube Degrad.ition Vehicle Barrier NUREG-0844. NRC INTEGRATED PROGRAM FOR THE RESOLUTION NUREG-1330: PERSONNEL AND VEHICLE BARRIERS AT FUEL FABRI-OF UNRESOLVED SAFETY ISSUES A-3.A 4 AND A-5 REGARDING CATION FACILITIES USING OR POSSESSING STnATEGIC QUANTI-STEAM GENERATOR TUBE INTEGRITY. Final Report. TIES OF SPECIAL NUCLEAR MATERIAL. Turbine Velocity Structure  ; NUREG/CR 4312 V02 R1: RELAPS/ MOO 2 CODE MANUAL. Volume 2: NUREG/CR-5080: A STUDY OF NEW ENGLAND SE!SMiGITY WITH j Users Guide And input Reqmrements. EMPHASIS ON MASSACHUSETTS AND NEW HAMPSHIRE. Final i Two-Phau Flow Report Covenng The Penod 1976 1985. NUREG/CR.5082 SIMULATION EXPERIMENTS ON TWO PHASE NAT" Ventilation System URAL CIRCULATION IN A FREON 113 FLOW VISUALIZATION LOOP' NUREG/CR-4932: CLOSEOUT OF IE BULLETIN 80-03: LOSS OF CHAR. USl A-40 COAL FROM STANDARD TYPE H,TWO-INCH, TRAY ADSORBER NUREG 1233 DRFT FC. REGULATORY ANALYSIS FOR USI A-40, CELLS.

          SEISMIC DESIGN CRITERIA." Draft Report For Comment                          NUREG/CR 4960 CONTROL ROOM HABITABILITY SURVEY OF Li-NUREG/CR 3509: POWER SPECTRAL DENSITY FUNCTIONS COM.                                    CE.NSED COMMERCIAL. NUCLEAR POWER GENERATING STA-PATIBLE WITH NRC REGULATORY GUIDE 1.60 RESPONSE SPEC.                             TIONS, TRA.

Venting System USl A-44 NUREG/CR-5225: AN OVERVIEW OF BWR MARK 1 CONTAINMENT NUREG-1032: EVALUATION OF STATION BLACKOUT ACCfDENTS AT VENTING RISK IMPLICATIONS-NUCLEAR POWER PLANTSTechnical Findings Related To Unre-solved Safety issue A-44 Final Report Vital Equipment NUREG 1109. REGULATORY /BACKHT ANALYSIS FOR THE RESOLU- NUREG 1178: VITAL EOulPMENT/ AREA GUIDELINES STUDY. Vital TION OF UNRESOLVED SAFETY ISSUE A 44, STATION DLACKOUT. Area Committee Report, Final Report, USI A-47 Volcanic Tuff NUREG 1218 DRFT FC: REGULATORY ANALYSIS FOR PROPOSED NUREG/CR-5097: SIMULATION OF LIQUlO AND VAPOR MOVEMENT RESOLUTION OF USI A-47. Safety Imphcations Of Control IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP Systems.Dratt Rereror Comment TUFF SITE.Models And Strategies. Uncertainty FGnat6on Volume-Averaged Velocity NUREu/GR 5000. METHODOLOGY FOR UNCERTAINTY E9TIMATION NUREG/CR-5065: TIME AND VOLUME-AVERAGED CONSERVATION IN NUREG-1150 (DRAFT) Conclusions Of A Review Panel. EQUATIONS FOR MULT1 PHASE FLOW USING MASS-WEIGHTED VE-Undervoltage LOCITY AND INTERNAL ENERGY, NUREG/CR-4935' CLOSEOUT OF IE BULLETIN 85 02:UNDERVOL.

                                                                                      **" P"      F       "

TAG TR A ACHMENTS OF WESTINGHOUSE DB-50 TYPE RE-NUREG/CR HYDROGEOLOGIC CHARACTER 12ATION OF BASALTS.The Northem Rim Of The Columbia Plateau Physiographic Unresolved Safety issue A 45 Province And Of The Creston Study Area. Eastern Washington. NUREG-12B9' REGULATORY AND BACKFIT ANALYSIS-UNRESOLVED SAFETY ISSUE A 45. SHUTDOWN DECAY HEAT REMOVAL RE, Warm Prestressing OUIREMENTS. NUREG/CR-4888. PRESSURIZED THERMAL-SHOCK TEST OF 6 INCH THICK PRESSURE VESSELS.PTSE 2: Investigation Of Low Teanng Re-Unsaturated Flow sistarr:e And Warm Prestressing. NUREG/CR-5094: APPLICATION OF STOCHASTIC METHODS TO THE SIMULATION OF LARGE SCALE UNSATURATED FLOW AND Waste Burlat ' TRANSPORT. NURLG/CR-5170. A REVIEW OF RESEARCH CONDUCTED BY LOS NUREG/CR 6097: SIMULATION OF LIOUlO AND VAPOR MOVEMENT ALAMOS NATIONAL LADORATORY FOR THE NRC WITH EMPHASIS IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP ON THE MAXEY FLATS,KY, SHALLOW LAND WASTE BURIAL SITE. TUFF SITE.Models And Strategies. Waste Burial Charges Unsolved Batety lesue NUREG-1307: REPORT ON WASTE BURIAL CHARGES Escalation Of NUREG-0844: NRC INTEGRATED PROGRAM FOR THE RESOLUTION Decommissioning Waste Disposal Costs At Low Level Waste Bunal Fa-OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 REGARDING c4tges. STEAM GENERATOR TUDE INTEGRITY. Final Report Waste Disposal Uranium NUREG/CR-4708 V02: PROGRESS IN EVALUATION OF RADIONU- NUREG/CR-5041 V02: RECOMMENDATIONS TO THE NRC FOR CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- REVIEW CRfTERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report For RADIOACTIVE WASTE DtSPOSAL Task 2ttEarth-Mounded Con: rete Apnl 1986 September 1987. Bunkers. Waste Form U G 5018. URANIUM OXIDE IRON OXfDE MIXED AEROSOL Mt N AD MW M2 WNH ' EXPERIMENTS IN STEAM-AIR A TMOSPHERE S.NSPP Tests WASTE FORMS. 611.612,613 And 631. Data Record Report-Waste Management l Uranium Toxicity NUREG/CR-4918 V02: CON 1ROL OF WATER INFILTRATION INTO NEAR SURFACE LLW DISPOSAL UNITS Task Report - A Discussion. l NUREG/CP 0093: PROCEEDINGS OF THE MEETING ON ULTRASEN-SITIVE TECHNIQUES FOR MEASUREMENT OF URANIUM IN BIO- ,,,,,p,eg,g, LOGICAL SAMPLES AND THE NEPHROTOKICITY OF URANIUM. NUREG/CR-4735 V03 EVALUATION AND COMPILATION OF DOE Value impact WASTE PACKAGE TEST DATA. Biannual Report: February July 1987. NUREG/CR 4555 RO1: GENERIC COST ESTIMATES FOR THE DIS- NUREG/CR-4735 V04: EVALUATION AND COMPILATION OF DOE POSAL OF RADIOACTIVE WASTES. WASTE PACKAGE TEST DATA. Biannual Report August 1987 - Janu-ary 198B. Value-impact Analysis NUREG/CR-5138. VALIDATION OF GENERIC COST ESTIMATES FOR Weste Repository CONSTRUCTION.RELATED ACTIVITIES AT NUCLEAR POWER NUREG/CR-5255: STABLE ISOTOPES OF AUTHIGENIC MINERALS IN PLANT %nal Report. VARIABLY-SATURATED F RACTURED TUFF. Valve Water Infiltration NUREG/CR-5141: AGING AND QUALIFICATION RESE ARCH ON SOLE- NUREG/CR 4918 V02: CONTROL OF WATER INFILTRATION INTO NOID OPERATED val.VES. NEAR SURFACE LLW DISPOSAL UNITS. Task Report A Discussion I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

l 112 Subject index stater Reactor Safety Weld NUREG/CP 0091 V04 ADO: PROCEEDINGS OF THE FIFTEENTH NUREG/CR 4600 V01: CHARACTERIZATION OF IRRADIATED CUR-WATER REACTOR SAFETY INFORMATION MEETING RENT PRACTLCE WELDS AND A533 GRADE B CLASS 1 PLATE FOR NUREG/CP 0090- TRANSACTIONS OF THE SIXTEENTH WATER RE-NUCLEAR PRESSURE VESSEL SERVICE. ACTOR SAFETY INFORMATION MEETING NUREG/CR 4880 V02: CHARACTERl2ATION OF IRRADIATED CUR-RENT-PRACTICE WELDS AND A533 GRADE O CLASS 1 PLATE FOR W lerhammer NUCLEAR PRESSURE VESSEL SERVICE. NUREG/CR 5220 VOI: DIAGNOSIS OF CONDENSATION-INDUCED WATERHAMMER Methods Atid Backgrovnd Zrceloy Cladding HUREG/CR-5220 V02- DIAGNOSIS OF CONDENSATION-INDUCED NUREG/CR-4777: STEAM OXIDATION OF ZlRCALOY CLADDING IN WATERHAMMER Case Studies THE ORNL FISSION PRODUCT RELEASE TESTS. I 1 3 f l l 1 i 1 I l

                                                                                                                                    )

1 l l l L__________._____________.._ _ _ _ _

1 NRC Originating Organization index (Staff Reports) This index lists those NRC organizations that have published staff reports. The index is ar-ranged alphabetically by major NRC organizations (e.g., program offices) and then by sub- .i sections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a i NUREG number and title of the report (s). If further information is needed, refer to the main  ! citation by NUREG number. j ADVISORY COMMITTEE (S) NUREG 1272 V02 N01: REPORT TO THE U.S. NUCLEAR REGULA. ACRS- ADVISORY COMMITTEE ON REACTOR SAFEGUARDS TORY COMM:SSION ON ANALYSIS AND EVALUATION OF OPER-NUREG-1125 V09. A COMPILATION OF REPORTS OF THE ADVISO- ATIONAL DATA.1987. Power Reactors. RY COMMITTEE ON REACTOR SAFEGUARDS.1987 Annual. NUREG 1272 V02 NO2: REPORT TO THE U S. NUCLEAR REGULA-TORY COMMISSION ON ANALYSIS AND EVALUATION OF OPER. OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) ATIONAL DATA .1987.Nonreactors. REGION 1. OFC OF THE DIRECTOR NUREG-1275 V03: OPERATING EXPERIENCE FEEDBACK REPORT - NUREG-0837 V07 N04: NRC TLD DIRECT RADIATION MONITORING SERVICE WATER SYSTEM NETWORK Progress Report October December 1987. FA! LURES AND DEGRADATIONS Commercial Power Reactors. l NUREG 0837 VOS N01: NRC TLD DIRECT RADIATION MONITORING NUREG 1303 INCIDENT INVESTIGATION MANUAL. l NETWORK. Progress Report. January-March 1983. DIVISION OF OPERATIONAL ASSESSMENT (POST B70413) j NUREG 0837 V08 NO2: NRC TLD DIRECT RADIATION MONITORING NUREG 1228: SOURCE TERM ESTIMATION DURING INCIDENT RE-NETWORK Pro 0ress Report. April-June 1988-SPONSE TO SEVERE NUCLEAR POWER PLANT ACCIDENTS. I OFC OF ENFORCEMENT (POST B70413) NUREG 0940 V06 N04: ENFORCEMENT ACTIONS SIGNIFICANT AC-OFFICE OF GOVERNMENTAL & PUBLIC AFFAIRS (POST 870413) TIONS RESOLVED.Ouarterly Progress Report. October DeLember STATE LOCAL & INDIAN TRIBE PROGRAMS 1987. NUREG-1309: THE U S NUCLEAR REGULATORY COMMISSION NUREG 0940 V07 NO1: ENFORCEMENT ACTIONS.SIGNIFICAN1 AC- PROGRAM WITH STATE AND LOCAL GOVERNMENTS AND , TIONS RESOLVED.Ouarterly Progress Report. January March 1988. INDIAN TRIBES. l NUREG-0940 V07 NO2: ENFORCEMENT ACTIONS.SIGNIFICANT AC- NUREG-1311: FUNDING THE NRC TRAINING PROGRAM FOR TlONS RESOLVED Quarterly Progress Report.Apnt-June 1988. STATES. NUREG 0940 V07 NO3: ENFORCEMENT ACTIONS SIGNIFICANT AC-TIO JS RESOLVED.Ouarterty Progress Report July-September 1988. EDO - OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT OFC OF SPECIAL PROJECTS (PRE 881231) OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT, DI-NUREG-0797 S14: SAFETY EVALUATION REPORT RELATED TO RECTOR (POST B70413) THE OPERATION OF THE COMANCHE PEAK STEAM ELECTRIC NUREG-0020 V12 NOB: LICENSED OPERATING REACTORS STATUS STATION. UNITS 1 AND 2. Docket Nos. 50-445 And 50 446(Texas

SUMMARY

REPORT. Data As Of July 31.1988.(Gray Book 1) f Utihties Generating Company) NUREG 0020 V12 N09: LICENSED OPERATING REACTORS STATUS NUREG 0797 SIS SAFETY EVALUATION REPORT RELATED TO

SUMMARY

REPORT. Data As Of August 31,1988.(Gray Book 1) THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC NUREG-0020 V12 NIO LICENSED OPERATING REACTORS STATUS STATION UNITS 1 AND 2. Docket Nos 50-445 And 50-446.(Texas

SUMMARY

REPORT. Data As Of September 30.1988 (Grey Book 1) Utthties Generahng Company) NUREG-0020 V12 N11: LICENSED OPERATING REACTORS STATUS NUREG 0797 S16: SAFETY EVALUATION REPORT RELATED TO

SUMMARY

REPORT. Data As Of October 31,1988.(Gray Book t) THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC NUREG 0325 R11: U S. NUCLEAR REGULATORY COMMISSION STATION UNITS 1 AND 2. Docket Nos 50 445 And 50 446.(Texas NU EG 14 V04 Utahties Generating Company) S N EAR REGULATORY COMMISSION NUREG-0797 S17: SAFETY El ALUATION REPORT RELATED TO DIVt OF E t O INFORMATION & PUBLICATIONS SERV. THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC ICES (POST BB0515 STATION. UNITS 1 AND 2. Docket Nos 50445 And 50 446.(Texas NUREG 0304 V13 N01: REGULATORY AND TECHNICAL REPORTS i NL 0797 1 SAF Y VALUATION REPORT RELATED TO I# ^ "W " " l I988 n ry h THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50 446.(Texas NUREG 0304 V13 NO2. REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL). Compilation For Second Quarter Utahtees Generating Company) 1988 April-June l NUREG-0797 S19: SAFETY EVALUATION REPORT RELATED TO NUREG 0304 V13 NO3: REGULATORY AND TECHNICAL REPORTS THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC (ABSTRACT INDEX JOURNAL). Compilation For Third Quarter STATION. UNITS 1 AND 2. Docket Nos. 50 445 And 50-446.(Texas 1988. July September Utihties Generating Company) l NUREG 0540 V10 N04: TITLE LIST OF DOCUMENTS MADE PUBLIC. NUREG 0797 S20. SAFETY EVALUATION REPORT FIELATED TO LY AVAILABLE. April 1-30,1988. THE OPERATION OF COMANCHE PEAK ST E AM ELECTRIC NUREG-0540 V10 N05: TITLE LIST OF DOCUMENTS MADE PUBLIC-STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-446(Texas LY AVAILABLE. May 1 31,1988. Utahties Generating Company) NUREG 0540 V10 N06. TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG-1232 V02. SAFETY EVALUATION REPORT ON TENNESSEE LY AVAILABLE. June 1-30,1988. VALLEY AUTHORITY. Sequoyah Nuclear Performance Plan. NUREG-0540 V10 N07: TITLE LIST OF DOCUMENTS MADE PUBLIC-LY AVAILABLE. July 1-31,19B8 EDO--OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG 0540 V10 NOB. TITLE L1ST OF DOCUMENTS MADE PUBLIC-DATA LY AVAILABLE. August 1 31,1988. ' OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, DI-NUREG 0540 V10 N09: TITLE LIST OF DOCUMENTS MADE PUBLIC-RECTOR LY AVAILABLE. September 1-30,1988. NUREG-0090 V10 NO3. REPORT TO CONGRESS ON ABNO9 MAL NUREG 0750 V27101: INDEXES TO NUCLEAR REGULATORY COM-OCCURRENCES. July September 1987. MIS $10N ISSUANCES. January-March 1988. NUREG 0090 V10 N04: REPORT TO CONGRESS ON ABNORJAL NUREG-0750 V27102. INDEXES TO NUCLEAR REGULATORY COM. OCCURRENCES October December 1987 MISSION ISSUANCES January June 1988. NOREG-0090 V11 N0t REPORT TO CONGRESS ON ABNORMAt NUREG.0750 V27 N04. NUCLEAR REGULATORY COMMISSION IS-OCCURRENCES January-Marct) 1988. SUANCES FOR APRIL 1988 Pages 335-483 NUREG 0090 Vit NO2. REPORT TO CONGRESS ON ABNORMAL NUREG.0750 V27 N05: NUCLEAR REGULATORY COMMISSION IS-OCCURRENCES Apnt-June 1988. SUANCES FOR MAY 1988 Pages 485 626. 113

114 NRC Originating Organization Index (Staff Reports) AUREG 0750 V27 N06 NL CLEAR REGULATORY COMMISSION IS. NUREG 0430 V0B NO2: LICENSED FUEL FACILITY STATUS SUANCES FOR JUNE 1910 Pages 627 665. REPORT. inventory Difference Data. July December 1987.(Grey Book NUREG-0750 V2B N01. N(CLE AR REGULATORY COMMISSION IS- II) SUANCES FOR JULY 19e B.Pages 171. NUREG-1310 NATURALLY OCCURRING AND ACCELERATOR-PRO-NJREG 0750 V2B NO2; NU"; LEAR REGULATORY COMMISSION IS- DUCED RADIOACTIVE MATERIALS 1987 Review. SUANCES FOR AUGUST 1988 Pages 73-269. DIVISION OF SAFEGUARDS & TRANSPORTATION (POST B70413) NUREG 0750 V2B NO3: NU7 LEAR REGULATORY COMMISSION IS- NUREG 0525 R14: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL). SUANCES FOR SEPTEMI ER 198B Pages 271417 NUREG 0725 R06: PUBLIC INFORMATION CIRCULAR FOR SHIP. NUREG 0750 V2B N04 NUOLEAR REGULATORY COMMISSION IS- MENTS OF IRRADIATED REACTOR FUEL. SUANCES FOR OCTOBER 1988 Pages 419 497- NUREG-1328 USE OF PERIMETER ALARMS AT FUEL FABRICA. NUREG 0936 V07 N01. N3C REGULATORY AGENDA Ouarterly TION FACILITIES USING OR POSSESSING FORMULA OUANTI-Fleport. January-March 1988 TIES OF STRATEGIC SPECIAL NUCLEAR MATERIAL NUREG.0936 V07 NO2. NI C REGULATORY ACENDA Quarterly NUREG-1329 ENTRY / EXIT CONTROL AT FUEL FABRICATION FA-NURE 09 7 NO NRd REGULATORY AGENDA.Ouarterly QMS USM M ESBSE NM WWWS & STRATEGIC SPECIAL NUCLEAR MATERIAL. Report,Jul September 1988 NUREG-1330 PERSONNEL AND VEHICLE BARRIERS AT FUEL FAB-NU EG 04 V1 GL ORY AN C AL REPORTS RICATION FACILITIES USING OR POSSESSING STRATEGIC (ABSTRACT INDEX JOURNA_) Annual Compilation For 1987. OUANTITIES OF SPECIAL NUCLEAR MATERIAL NUREG 0540 V09 N11: TITLr. LIST OF DOCUMENTS MADE PUBLIC. DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST , LY AVAILABLE. Novemtw 1,30,1987. 870729) l NUREG 0540 V09 N12: TITi E LIST OF DOCUMENTS MADE PUBLIC- NUREG 1308: RADIOACTIVE MATERIAL IN THE WEST LAKE LY AVAILABLE. December 1 31.1987. LANDFILLSummary Report. NUREG 0540 V10 N01: TITLE LIST OF DOCUMENTS MADE PUBLIC. NUREG.1308 Rot: RADIOACTIVE MATERIAL IN THE WEST LAKE LY AVAILABLE. January 1 31.1988. LANDFILL. Summary Report. NUREG 0540 V10 NO2: TITLE LIST OF DOCUMENTS MADE PUBUC- NUREG-1320. NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALY-LY AVAILABLE. February 1-29,1986 SIS HANDBOOK. NUREG 0540 VIO NO3. TITLE LIST OF DOCUMENTS MADE PUBLIC- DIVISION OF HIGH LEVEL WASTE MANAGEMENT (POST 870413) LY AVAILABLE March 1 31,1900 NUREG 1263. HYDROLOGIC DESIGN FOR RIPRAP ON EMBANK-NUREG 0750 V25107. INDEXES TO NUCLEAR REGULATORY COM- MENT SLOPES MISSION ISSUANCES. January-June 1987, NUREG-1297: PEER REVIEW FOR HIGH-LEVEL NUCLEAR WASTE NUREG 0750 V25 N05. NUCLEAR REGULATORY COMMISSION IS- REPOSITOR!ES Genene Technical Posit on. SUANCES FOR MAY 1987 Pages 417-873. NURE G-1298 QUALIFICATION OF EXISTING DATA FOR HIGH-NUREG-0750 V25 N06' NUCLEAR REGULATORY COMt4lSSION IS- LEVEL NUCLEAR WASTE REPOSITORIES Genonc Technical Poss l SUANCES FOR JUNE 1987.Pages B75-997- tion. NUREG 0750 V26101. INDEXES TO NUCLEAR REGULATORY COM- NUREG-1318: TECHNICAL POsthDN ON ITEMS AND ACTIVITIES IN MISSION ISSUANCE S. July-September 1987. THE HIGH-LEVEL WASTE GEOLOGIC REPOSITORY PROGRAM NUREG-07b0 V26102: INDEXES TO NUCLEAR REGULATORY COM- SUBJECT TO OUALITY ASSURANCE REQUIREMENTS.

                                                                                                            ^                 "

NUF 50 V Not t LEA ULATORY COMMISSION 15 POSTB NUREG-1199 F101: STANDARD FORMAT AND CONTENT OF A Lt. N E G- 0 26 2N LENE E ULATORY CUMMISSION IS- CENSE APPLICATION FOR A LOW LEVEL RADIOACTIVE WASTE SUANCE S FOR AUGUST 1987. Panes 71 107. NU EG 1200 0 T ANDARD REVIEW PLAN FOR THE REVIEW OF S AN S FOR SEP EMBER 987 Pa es 09 24 NUREG-0750 V26 N04. NUCLEAR REGULAIORY COMMISSION IS. A LICENSE APPLICATION FOR A LOW-LEVEL RADIOACTIVE SUANCES FOR OCTOBER 1987. Pages 249 381. AASTE DISPOSAL FACILITY. NUREG-0750 V26 N05: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR NOVEMBER 1987. Pages 383 447. U.S. NUCLEAR REGULATORY COMMISSION l NUREG-0754 V2t. N06: NUCLEAR REGULATORY COMMISSION IS. Of.FICE OF THE GENERAL COUNSEL ) NUREG-0386 D04 R07: UNITED STATES NUCLEAR REGULATORY SUANCES NUREG-0750 V27FOR N01:DECEMBER 19B7. Pahes NUCLEAR RE )LATORY 449 530 COMMISSION IS. COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST.JJy SUANCES FOR JANUARY 1988 Pages 1-39 1972 - March 1967. NUREG-0750 V27 N02 NUCLEAR REGULATORY COMMISSION IS. NUREG 0386 D04 ROB: UNITED STATES NUCLEAR REGULATORY SUANCES FOR IEBRUARY 1988 Pages 41-255. COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July NUREG 0750 V27 NO3 NUCLEAR REGULATORY COMMISSION 15 1972 June 1987. SUANCES FOR MARCH 19t!8 Pages 257 334. NUREG-0386 D04 R09. UNITED STATES NUCLEAR REGULATORY 1 DIVISION OF RULES & RECORDS (870413-0B0514) COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July NUREG 0936 UD6 N04. NRU REGlJLATORY AGENDA.Ouarterly 1972 - September 1987. Report. October December 1987 NUREG 0386 D04 R10: UNITED STATES NUCLEAR REGULATORY Dh/ISION OF BUDGET & ANALYSIS (POST B70413) { OMMISSION STAFF PRACTICE AND PROCEDURE NUREG-1100 V04 BUDGET ESTIMATES Fiscal Year 1989 CIGEST. Commission Appeal Board And Licensing Board DIVISION OF COMPUTE R & TELECOMMUNICATIONS SERVICES Decisions. July 1972 - December 1987. (POST B70413) NRC - NO DETAILED AFFluATION GIVEN NUREG-0020 V11 N12 LICENSED OPERATING RE ACTORS ST ATUS NUREG-0654 S01 ROI: CRITERIA FOR PREPARATION AND EVAL-UATION OF RADIOLOGICAL EMERGENCY RESPONFE PLANS NUFEG 20 ? o L ENS OP Fi N REA I$ A US # * #d#

SUMMARY

REPORT. Data As Of December 31.1987.(Gray Book I) PLANTS Cntena For Utihty Offsite Planning And Preparedness NUREG 0020 V12 NO2. LICENSED OPERATING RE ACTORS ST ATUS

SUMMARY

REPORT Data As Of January 31.1988 (Gray Book 11 NUREG-1290 ADD DIF F EFIINr PROFESSIONAL OPINIONS 1987 a Special Review Panet NUREG 11020 V12 NU3 LICENSED OPEFIATING REACTORS ST ATUS

SUMMARY

REPOFIT. Data As Of F ebruary 29.1988 (Gray Book I) NUREG/CR-3950 V04 FUEL PERFORMANCE ANNUAL FiEPORT ) NUREG 0020 V12 N04 LICENSED OPERATING REACTORS STATUS FOR 1986 l

SUMMARY

REPORT. Data As Of March 31.1988 (Gray Book 1) NUREG/CR-4918 V02: CONTROL OF WATER INFILTRATION INTO NUREG 0020 V12 NOS: LICENSED OPERATING RE ACTORS ST ATUS NEAR SURF ACE LLW DISPOSAL UNfTS Task Report A Discus- i

SUMMARY

REPORT Data As Of Apnl 30.1988 (Gray Book f) son. I NUREG-0020 V12 N06 LICENSED OPERATING RE ACTORS ST ATUS NUREG/CR 5151. PERFORMANCE-BASED INSPECTIONS.

SUMMARY

REPORT Data As Of May 31.19BB.(Gray Book 1) NUREG 0020 V12 ND7. LICE NSED OPERATING HE ACTORS ST ATUS EDO - OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

SUMMARY

REPORT Data As Of June 30.1988 (Gray Book 1) OFFICE OF NUCLEAR REGULATORY FIESEARCH, DIRECTOR (POST  ; B60720) l EDO -OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG 0506 FINAL GENERIC ENVIRONMENTAL IMPACT STATE-OFTICE OF NUCLE AR MATER 6AL SAFETY & MENT ON DECOMMISSIONING OF NUCLEAR FACILITIES SAf f GUARDS. DIRE CTOR NURE G-1032: EVALUATION OF STATION BLACKOUT ACCIDENTS NUREG 0430 V00 Not LICENSED FUEL F ACILITY ST ATUS AT NUCLEAR POWER PLANTS. Technical Findings Related To Un-REPORT. inventory Delference Data. January-June 1987.(Gray Book 11) resolved Safety lasue A 44 Final Report l l l

NRC Originating Organization index (Staff R'.eports) 115 NUREG-1109. REGULATORY /BACKFIT ANALYSIS FOR THE REEO- NUREG 1332: FIEGULATORY ANALYSIS FOR THE RESOLUTION OF LUTION OF UNRESOLVED SAFETY ISSUE A-44. STATION BLACK- GENERIC (SSUE 125.IL7. " REEVALUATE PROVISION TO AUTO-OUT MATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR NUREG-1226 DEVELOPMENT AND UTILIZATION OF THE N9C DURING A LINE BREAK." POLICY STATEMENT ON THE REGULATION OF ADVANCED NU-CLEAR POWER PLANTS EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80) NUREG-1260 V02. A REPORT TO CONGRESS ON NUCLEAR REGU. OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST LATORY RESEARCH Protect Desenptions For FY88 870411) NUREG-1266 VD2. NRC SAFETY RESEARCH IN SUPPORT OF REG- NUREG 06B3 S03 DRFT: PROGRAMMATIC ENVIRONMENTAL ULATION .1987. IMPACT STATEMENT RELATED TO DECONTAMINATION AND NUREG-1325 DISPOSITION OF RECOMMENDATIONS OF THE NA- DISPOSAL OF RADIOACTIVE WASTES RESULTING FROM TIONAL RESEARCH COUNCIL IN THE REPORT "REVITAUZING MARCH 28.1979 ACCICENTJ HREE MILE ISLAND NUCLEAR NUCLEAR SAFETY RESEARCH " STATION. UNIT 2. Docket No. 50-320.(GPU Nuclear, Incorporated) NUREG/CP 0096: TRANSACTIONS OF THE SIXTEENTH WATER RE- NUREG-1032: EVALUATION OF STATION BLACKOUT ACCIDENTS ACTOR SAFETY INFORMATION MEETING AT NUCLEAR POWER PLANTSTechncal Findings Related To Un-DIVISION OF ENGINEERING (POST B704131 resolved Safety issue A-44. Final Fleport NUREG-0800 02 4.2 R3: STANDARD REVIEW PLAN FOR THE NUREG 1109 REGULATORY /BACKFIT ANALYSIS FOR THE FIESO. REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR LUTION OF UNRESOLVED SAFETY ISSUE A-44,STAllON BLACK-POWER PLANTS LWR Edition. Proposed Reveon 3 To SRP Section OUT NUR G 00 0 4$ R3 T NDARD REVIEW PLAN FOR THE O "' REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR

                                                                                              ^       "

POWER PLANTS LWR Edition. Proposed Revision 3 To SAP Section NU EG 2d' 01:' AFET EVALUATION REPORT RELATED TO BABCOCK AND WILCOX OWNERS GROUP PLANT REASSESS-2 4 3. " Probable Maximum Flood (PMF) On Streams & Rivers " For MENT PROGRAM Comment-NUHEG-1217 DRFT FC: EVALUATION OF SAFETY IMPLICATIONS DIVISION OF REACTOR PROJECTS .1/ll (POST 870411) NUREG-0822 S01. INTEGRATED PLANT SAFETY ASSESSMENT OF CONTROL SYSTEMS IN LWR NUCLEAR POWER SYSTEMATIC EVALUATION PROGRAM OYSTER CREEK NUCLE-PLANTS Technical Findings Related To Unresolved Safety issue A-AR GENERATING STATION. Docket No. 50-219.(General Public Utili-NU EG 12 FT F tf LATORY ANALYSIS FOR PROPOSED # g g [7S07 SAF L ATIOtb RE R D TO RESOLUTION OF USl A 4 7. Safety Imphcations Of Control THE OPERATION OF VOGTLE ELECTRIC GENERATING NUl 1217 SUM AY NALY IS AND RESPONSE TO PUBLIC PLAr4T, UNITS 1 AND 2. Docket Nos. 50 424 And 50-425.(Georgia COMMENTS ON PROPOSED AMENDMENTS TO 10 CFR PARTS DIV! ON O E CTOR PROJECTS - lluV,V & SPECIAL PROJECTS 30.40.50.51.70 AND 72: DECOMMISSIONING CRITERIA FOR NU-(POST 870M1 N 1222 II RESEARCH PROGRAM PLAN NUREG-0781 SOS. SAFETY EVALUATION REPORT RELATED TO l NUREG-1233 DRFT FC: REGULATORY ANALYSIS FOR USl A40. THE OPERATION OF SOUTH TEXAS PROJECT. UNITS 1 AND l

                       " SEISMIC DESIGN CRITERIA." Draft Report For Comment                  P. Docket Nos. 50 49B And 50-499. (Houston Ughting And Power NUREG-1249 V01: NRC MODEL SIMULATIONS IN SUPPORT OF THE                  Company)

HVDROLOGIC CODE INTERCOMPARISON (HYDROCOIN) NUREG 1002 S06: SAFETY EVALUATION REPORT RELATED TO STUDY Level 1 - Code Ve ification. THE OPERATION OF BRAIDWOOD STATION. UNITS 1 AND NUREG 1296 THERMAL OVERLOAD PROTECTION FOR ELECTRIC 2 Docket Nos. 50-456 And 50-457.(Commonwealth Edison Company) MOTORS ON SAFETY RELATED MOTOR-OPERATED VALVES . NUREG 1286 S01: SAFETY EVALUATION REPORT RELATED TO GENERIC ISSUE il E 01 THE RESTART OF RANCHO SECO NUCLEAR GENERATING NUREG-1307. FIEPORT ON WASTE BURIAL CHARGES Escalation Of STATION, UNIT 1.FOLLOWING T'IE EVENT OF DECEMBER Decommissioning Waste Disposal Costs At Low-level Waste Bunal 26.1985 Docket No 50-312 (Sikoramento Municipal Utility Distnet) F acihties. NUREG-1305: TECHNICAL SPECIFICATIONS FOR SOUTH TEXAS MATE R6ALS ENGINEERING BR/sNCH PROJECT, UNIT 1, Docket No. 50 49B(Houston Ughting And Power NUREG-0975 V06. COMPILATION OF CONTRACT RESEARCH FOR Company) l THE MATERIALS ENGINEERING BRANCH. DIVISION OF NUREG 1306: NRC SAFETY SIGNIFICANCE ASSESSMENT TEAM ' ENGINEERING Annual Report For FY 1987. REPORT ON ALLEGATIONS RELATED TO THE SOUTH TEXAS DIVISION OF REACTOR ACCIDENT ANALYSIS (670413-880716) PROJECT. UNITS 1 & 2. NURE G 1140. A REGULATORY ANALYSIS ON EMERGENCY PRE- STANDARDl2ATION & NON-POWER REACTOR PROJECT DIRECTOR-PAREDNESS FOR FUEL CYCLE AND OTHER RADIOACTIVE MA- ATE TERIAL tlCENSEE S Final Report. NUREG-1283. SAFETY EVALUATION REPORT RELATED TO THE DIVISION OF REACTOR & PLANT SYSTEMS (870413 880716) RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH NUREG-1236: THERMAL-HYDRAULIC RESEARCH PLAN FOR BAB-REACTOR AT PURDUE UNIVERSITY. A THE NU G 125 UCL OWER PLANT THERMAL-HYDRAULIC PERFORMANCE RESEARCH PROGRAM PLAN EVALUATION OF LOW ENRICHED URANIUM SILICIDE ALUMINUM NUREG-1273. TECHNICAL FINDINGS AND REGULATORY ANALYSIS DISPERSION FUEL FOR USE IN NON POWER RE ACTORS. FOR GENERIC SAFETY ISSUE II E4 3, " CONTAINMENT 1NTEGRI- DIVISION OF ENGINEERING & SYSTEMS TECHNOLOGY (POST TYCHECK" B704 m DIVISION OF REGULATORY APPLICATIONS (POST 870413) NUREG 0313 R02. TECHNICAL REPORT ON MATERIAL SELECTION NUREG-0713 V07: OCCUPATIONAL RADIATION EXPOSbaE AT AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE COMMERCIAL NUCLEAR POWER REACTORS AND OTHER F A. BOUNDARY PIPING Fmal Report. CILITIES 1905 Eighteenth Annual Report. NUREG-0844. NRC INTEGRATED PROGRAM FOR THE RESOLU-NUREG 0933 S07. A PRIORITIZATION OF GENERIC SAFETY TlON OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 RE-ISSUES GARDING STEAM GENERATOR TUBE INTEGRITY Final Report. NURE G 0933 SDB A PRIORITIZATION Or GENERIC SAFfTy DIVISION OF REACTOR INSPECTION & SAFEGUARDS (POST i ISSUE S 8704111 l NUHEG 1333 DAFT FC MAINTENANCE APPROACHES AND PRAC. NUREG-0040 VII N01 LICENSEE CONTRACTOR AND VENDOR IN- ! TICES IN SELECTED FOREIGN NUCLEAR POWER PROGRAMS SPECTION STATUS REPORT. Quarterly Report. October-December AND OTHER U.S. INDUSTRIES Review And Lessons Learned Draft 1987 (White Book) Report For Comment. NUREG 0040 V12 N01: LICENSEE CONTRACTOR AND VENDOR IN-NURE G/CP 0099. PROCEEDINGS OF THE PUBLIC WORKSHOP FOR SPECTION STATUS REPORT. Quarterly Report, January-March NRC RULEMAKING ON MAINTENANCE OF NUCLEAR POWER 198B (white Book) PLANTS NUREG.0040 V12 NO2: LICENSEE CONTRACTOR AND VENDOR IN-DIVISION OF StJETY ISSUE RESOLUTION (POST 880717) SPECTION STATUS REPORT. Oertorly Report.Apni June NUHf G 1289 REGULATORY AND DACKFIT 1988 (White Book) ANALYSIS' UNRESOLVED SAFETY ISSUE A 45. SHUTDOWN NUREG-0040 V12 NO3 LICENSEE CONTRACTOR AND VENDOR IN-DECAY HE AT REMGVAL REQUIREMENTS SPECTION STATUS REPORT. Quarterly Report.Jaly-September NUREG-1317 DRFT FC. REGULATORY OPTIONS FOR NUCLEAR 1988.(Wnite Book)  ; PLANT LICENSE RENEWAL. Draft For Comment NUREG-1304 REPORTING OF SAFEGUARDS EVENTS i 8 1 _ _ _ _ _ _ _ _ J

116 NRC Originating Organization index (Staff Reports) DIVISION OF LICENSEE PERFORMANCE & OUALITY EVALUATION NUREG-121e R04: HISTORICAL DATA

SUMMARY

OF THE SYSTEM-(POST B70411) ATIC ASSESSMENT OF LICENSEE PERFORMANCE NUFIEG 1214 R03 Hr$TORICAL DATA

SUMMARY

OF THE SYSTEM. NUREG/CP 0089 PROCEEDINGS OF THE CSNI SPECLALIST MEET-ATIC ASSESSMENT OF LICENSEE PERFORMANCE. ING ON TRAINING OF NUCLEAR REACTOR PERSONNEL. Held At Orlando,Flonda. April 21-24,1987. l I S

NRC Originating Organization Index (International Agreements) l This index lists those NRC organizations that have published international agreement re-  ! ports. The index is arranged alphabetically by major NRC organizations (e.g., program of- I fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number. There were no NUREG/lA reports for 1988. 1 i l I 1 l i 117 j

W - l l l l l l l l l i i I 7

NRC Contract Sponsor index (Contractor Reports) This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major Nr2C organization (e.g., program office) and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza- I tion is followed by the NUREG/CR number and title of the report (s) prepared by that organi- ' zation. If further information is needed, refer to the main citation by the NUREG/CR number. i i OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS REGION t OFC OF THE DIRECTOR OFFICE OF NUCLEAR- MATERIAL SAFETY & NUREG/CR-5058: PRA APPLICATIONS PROGRAM FOR INSPEC. SAFEGUARDS, DIRECTOR TION AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50 NUREG/CR-4708 V02: PROGRESS IN EVALUATION OF RADIONU-313.(Arkansas Power And U;lht Cornpany) CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE NUREG/CR-5145: FAILURE INVESTIGATION OF 3M SERIES 900 AtlGH LEVEL NUCLEAR WASTE REPOSITORY SITE STATIC ELIMINATORS. PROJECTS. Report For April 1986 - September 1987 NUREG/CR-4997: METHODS FOR DESCRIBING AIRBORNE FRAC. EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL TlONS OF FREE FALL SPILLS OF POWDERS AND LIQUIDS. DATA DIVISION OF SAFEGUARDS & TRANSPORTATION (POST 870413) OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, Di- NUREG/CR-5033:

SUMMARY

DESCRIPTION OF THE SCALE MODU-RECTOR LAR CODE SYSTEM. HUREG/CH-2000 V06N12: LICENSEE EVENT REPORT (LER) DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST COMPILATION:For Month Of Decernber 1987. 870729) NUREG/CR.2000 V07 N1: LICENSEE EVENT REPORT (LER) NUREG/CR-5145: FAILURE INVESTIGATION OF 3M SERIES 900 i COMPILATION For Month Of January 1988. STATIC EUMINATORS. I NUREG/CR-2000 V07 N2: LICENSEE EVENT REPORT (LER) D' VISION OF HIGH LEVEL WASTE MANAGEMENT (POST 870413) COMPILATION For Month Of Fetwuary 1988. NUREG/CR 4735 V03: EVALUATION AND COMPILATION OF DOE NUREG/CR-2000 V07 N3: LICENSEE EVENT REPORT (LER) WASTE PACKAGE TEST DATA. Biannual ReportFebruary July COMPILATION.For Month Of March 1988. 1987-NUREG/CR 2000 V07 N4: LICENSEE EVENT REPORT (LER) NUREG/CR-4735 V04: EVALUATION ANO COMPILATION OF DOE i COMPluATION For Month Of April 1988. WASTE PACKAGE TEST DATA. Biannust ReportAugust 1987 - Jan- i uary 1988. NUREG/CR-2000 V07 NS: UCENSEE EVENT REPORT (LER) COMPILATION For Month Of May 1988. NUREG/CH-4807: SURFACE COMPLEXATION MOCELING OF RADI. NUREG/CR-2000 V07 N6. UCENSEE EVENT REPORT (LER) ONUCLiDE ADSORPTION IN SUBSURFACE ENVIRONMENTS. COMPILATION.For Month Of June 1988. NUREG/CR 4864 V01: THERMODYNAMIC TABLES FOR NUCLEAR NUREG/CR-2000 V07 N7; UCENSEE EVENT REPORT (LER) WASTE (SOLATION. Aqueous Solutions Datat,ase. COMPILATION.For Month Of July 1988. DIVISION OF LOW LEVEL WASTE MANAGEMENT & DECOMMISSION-MPILA For Mon Of A u t1988. N 651 VD : DEVELOPMENT OF SIPRAP DESIGN CRITE-NUREG/CR-2000 V07 N9' UCENSEE EVENT REPORT (LER) RIA BY RIPRAP TESTING IN FLUMES. Phase ILFollowup investiga-COMPILATION For Month Of September 1968 I'0"" NUREG/CR-2000 VG7N10. LICENSEE EVENT REPORT (LER) NUREG' /CFI-5041 V02: RECOMMENDATIONS TO THE NRC FOR COMPILATION For Month of October 1988 PEVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW. [ NUREG/CR-2000 V07N11: UCENSEE EVENT REPORT (LER) LEVEL RADIOACTIVE WASTE DISPOSAL Task 2b Earth-Mounded j COMPILATION.For Month Of November 19B8 Concrete Duh NUREG/CR 5054. RECOMMENDATIONS TO THE NRC FOR REVIEW ] NUREG/CR4674 VDS: PRECURSORS TO POTENTIAL SEVERE , CORE DAMACE ACCIDENTS:1986,A ST ATUS REPORT. CRITERIA FOR ALTERNATIVE METHODS OF LOW-LEVEL RADIO-A E WASTE DISPOSAL Enannental Mondonng And Suntee NUREG/CR 4674 V06. PRECURSORS TO POTENTIAL SEVERE OlVI N OF OPERA O AL S ESS E (POS 87 413') N RE /C 51 A USER'S MANUAL FOR THE CONT AMINANT NUREG/CR4000 V02: THE MESORAD DOSE ASSESSMENT TRANSPORT MODULE OF THE MiGRAT CODE. MODELComputer Code. DIVISION OF SAFETY PROCRAMS (POST 870413) EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405) OrrlCE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR (POST NUREG/CR 5050: ANNOTATED BIBUOGRAPHY OF REUABluTY AND RISK DATA SOURCES- 860720) NUREG/CR.2331 V7N2-3. SAFETY RESEARCH PROGRAMS SPON-EDO OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT OFFICE OF ADMIN 3TRATION & RESOURCES MANAGEMENT, DI- R SEARCH P ogress R .Apr Sept 1987. NUREG/CR-2331 VBN12: SAFETY RESEARCH PROGRAMS SPON-NUREG CR 2 5 V ULATION DOSE COMMITMENTS DUE TO SEARCH ens Ja uary J 1988 l RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR-4312 V02 Rt. RELAPS/ MOD 2 CODE MANUALVolume 2: N RE /CR 2907 V08. RADIOACTIVE MATERLALS RELEASED FROM NURN/C 47 7 MO ATl NUCLEAR POWER PLANTS. Annual F4 port For 1905 OF ZlRCALOY CLADDING IN THE ORML FtSSION PRODUCT RELEASE TESTS. NUREG/CR-2907 V07. RADIOAC'lVE MATERIALS RELEASED FROM NUCLE AR POWER PLANTS Annual Report For 1986 NUREG/CR4971: RESULTS OF SEMISCALE MOD-PC FEEDWATER AND STEAM LINE BREAK (S FS) EXPERIMENT SERIESMain DIVISION OF BUDGET & ANALYSIS (POST 870413) Steam Une Break Accident Expenments. NUREG/CR-2850 V07 POPULATION DOSE COMMITMENTS DUE TO NUREG/CR-5021 V01: USER'S GUIDE FOR PAISIM ARKANSAS NU. RADIOACTIVE RELE ASES FROM NUCLEAR POWER PLANT S!TES CLEAR ONE - UNIT t. Volume 1. Program For Inspectors IN 1985. NUREG/CR-5021 V02: USER'S GUIDE FOR PRISIM ARKANSAS NU. l DIVISION OF INFORMATION SUPPORT SEFIVICES (PRE 870413) CLEAR ONE - UNIT 1 Volume 2, Program For Regulators. i l NUREG/CR4264. GUIDE FOR LICENS'NG EVALUATIONS USING NUREG/CR-5082: SIMULATION EXPERIMENTS ON TWO. PHASE l CRAC2.A Computer Program for Calculatmg Reactor Accident Con-sequences. NATURAL CIRCULATION IN A FREON-113 FLOW VISUALIZATION ' LOOP. i l 119 l -

120 NRC Contract Sponsor index NUREG/CR-5120 A MODEL FOR THE TRANSPORT AND CHEMICAL NUREG/CR4808: PRESSURIZED THERMAL SHOCK TEST OF 6-REACTION OF MOLTEN DEBRIS IN DIRECT CONTAINMENT INCH TH!CK PRESSURE VESSELS.PTSE 2: Investigation Of Low HEATING EXPERIMENTS. Tearing Resistance And Warm Prestressing. PROGRAM MANAGEMENT, POLICY DEVELOPMENT & ANALYSIS NUREG/CR.4918 V02: CONTROL OF WATER INFILTRATION INTO STAFF (POST 870413) NEAR SURFACE LLW DISPOSAL UNfTS. Task Report - A Discus-NUREG/CR.4315 V09 R1: EVALUATION OF NUCLEAR FACILITY DE- sion. COMMISSIONING PROJECTS Summary Status Report.Three Mile NUREG/CR-4924: SEISMIC CATEGORY l STRUCTURES island Unit 2. Radioactive Waste ArJ Laundry Shipments. PROGRAM Final Report. Fiscal Year 1983 - 1984 NUREG/CR-5126: TAC 2D STUDIES OF MARK I CONTAINMENT NUREG/CR-4939 V01: IMPROVING MOTOR RELIABILITY IN NUCLE- ) DRYWELL SHELL MELT THROUGH. AR POWER PLANTS Volume 1. Performance Evaluation And Mainte-DIVISION OF ENGINEERING (POST 870413) nance Practices. l I NUREG/CR 0130 ADD 04: TECHNOLOGY. SAFETY AND COSTS OF NUREG/CR4939 V02: IMPROVING MOTOR FIELIABILITY IN NUCLE. DECOMMIS$10NING A REFERENCE PRESSURIZED WATER RE- AH POWER PLANTS. Volume 2 Functional indicator Tests On A ACTOR POWER STATIONTechnical Support For Decommissioning Small Electnc Motor Subpected To Accelerated Agin Matters Related To Preparation Of The Final Decommissioning Rule. NUREG/CR4939 V03: IMPROVING MOTOR RELIABk.ITY IN NUCL NUREG/CR-0672 ADD 03: TECHNOLOGY. SAFETY AND COSTS OF AR POWER PLANTS. Volume 3 Failure Analysis And Diagnostic DECOMMISSIONING A REFERENCE BOILING WATER REACTOR Tests On A Naturally Aged Large Electne Motor. POWER STATION.TechrFcal Support For Decommissioning Matters NUREG/CR4947: ANALYSIS OF THE A3028 AND A533B STAND. Related To Preparation Of The Fmal Decommissioning Rule. ARD REFERENCE MATERIALS IN SURVEILLANCE CAPSULES OF NUREG/CR-2336. STEAM GENERATOR TUBE INTEGRITY COMMERCIAL POWER FIEACTORS. PROGRAM. Phase 11 Final Report. NUREG/CR4984: DEVELOPMENT OF A THREE DIMENSIONAL NUREG/CR-3145 V06: GEOPHYSICAL INVESTIGATIONS OF THE FLUX SYNTHESIS PROGRAM AND COMPARISON WITH 3-D WESTERN OHIO-INDIANA REGION. Annual Report. October 1986 TRANSPORT THEORY RESULTS. September 1987. NUREG/CR-4992 V01: AGING AND SERVICE WEAR OF MULTIS-NUREG/CR-3444 V05: THE IMPACT OF LWR DECONTAMINATION TAGE SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED OG POWER PLANTS Operstmg Exponence And Failure identification. CUPATIONAL EXPOSURE. Annual Report. FY 1987. NUREG/CR4996. A REPORT ON THE ROUND ROBIN PROGRAM NUREG/CR-3899 S01: UTILITY FINANCIAL STABILITY AND THE CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD AVAILABILITY OF FUNDS FOR DECOMMISSIONING.An Analysis Of TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK internal And External Funding. ARREST FRACTURE TOUGHNESS,K(IA).0F FERRITIC MATERI-NUREG/CR-4082 V06. DEGRADED PIPING PROGRAM - PHASE ALS. Il Sixth Program Report. October 1986 September 1987. NUREG/CR-4998: THE SEISMIC CATEGORY I STRUCTURES NUREG/CR4219 V04 N2. HEAVY-SECTION STEEL TECHNOLOGY PROGRAM Results For Fiscal Year 1985. PROGRAM. Semiannual Progress Report For Apni-September 1987. NUREG/CR-5012: SIMILARITY PRINCIPLES FOR EQUIPMENT OUAL-NUREG/CR4210 VOS N1: HEAVY-SECTION STEEL TECHNOLOGY IFICATION BY EXPERIENCE PROGRAM. Semiannual Progress Report For October 1987 March NUREG/CR-5013. FATIGUE LIFE CHARACTERl2ATION OF SMOOTH 1988. AND NOTCHED PIPING STEEL SPECIMENS IN 288 DEGREES C NUREG/CR 4527 V02: AN EXPERIMENTAL INVESTIGATION OF IN. AIR ENVIRONMENTS. TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CON. NUREG/CR-5016: COMPENDIUM AND COMPARISON OF INTERNA-TROL CABINETS Part II: Room Effects Tests. TlONAL PRACTICE FOR PLUGGING. REPAIR AND INSPECTION OF NUREG/CR 45D7 V02: ACING AND SERVICE WEAR OF AUXILIARY STEAM GENERATOR TUBING. F EEDWATER PUMPS FOR PWR NUCLEAR PLANTS Volume NUREG/CR-5019: NEUTRON EXPOSURE PARAMETEFIS FOR THE 2 Aging Assessments And Monitonng Method Evaluat ons. METALLURGICAL TEST SPECIMENS IN THE FIFTH HEAVY SEC-NUREG/CR-4659 V02: SEISMIC FRAGILITY OF NUCLEAR POWER TION STEEL TECHNOLOGY IRRADIATION SERIES CAPSULES PLANT COMPONENTS (PHASE II). Motor Control NUREG/CR-5020. A

SUMMARY

OF ENVIRONMENTALLY ASSISTED Center. Switchboard.Panelboard And Power Supply. CRACK-GROWTH STUDIES PERFORMED AT WESTINGHOUSE NUREG/CR-4667 V04- ENVIRONMENTALLY ASSISTED CRACKWG ELECTRIC CORPORATION.Under Funding From The Heavy-Section IN LIGHT WATER REACTORS. Semiannual Report. October 1986 Steel Tectinology Program. Maech 1987. NUREG/CR-5023. HIGH LEVEL SEISMIC RESPONSE AND FAILURE NUREG/CH-4728 EQUIPMENT QUALIFICATION RESEARCH TEST PREDICTION METHODS FOR PIPING. OF A HIGH-RANGE RAD:ATION MONITOR. NUREG/CR-5024: TENSILE AND J R CURVE CHARACTERf2ATION NUREG/CR 4740: NUCLEAR PLANT AGING RESEARCH ON REAC. OF THERMALLY AGED CAST STAINLESS STEELS. TOR PROTECTION SYSTEMS. NUREG/CA 5031: SIGNIFICANCE OF IN-STRUCTURE GENERATED NUREG/CR 4747 V02. AN AGING FAILURE SURVEY OF LIGHT MOTION IN SEISMIC QUALIFICATION TESTS OF CABINET WATER RE ACTOR SAFETY SYSTEMS AND COMPONENTS MOUNTED ELECTRICAL DEVICES. NUREG/CR-4763: SAFETY RELATED EQUIPMENT SURVIVAL'IN HY. NUREG/CR-5043: CONTAINMENT PENETRATION SYSTEM (CPS) DROGEN BURNS IN LARGE DRY PWR CONTAINMENT BUILD- UD ACC NT ADS N A IN PERATING MO% NURE /CR DU ES R SHIPP G PA A S N S NEBRASKA SEISMICITY STUDIES USING THE KANSAS NEBRASKA MICROEARTHOUAKE NUREG/CR-4784 INFLUENCE OF ' GROUNDW ATER ON SOIL- NETWORK. Final Report. STRUCTURE INTE9 ACTION NUREG/CR 5047: RADIONUCLIDES ACCUMULATION BY AOUATIC NUREG/CR 4785 REVIFW AND EVALUATION OF DESIGN ANALY-BIOTA EXPOSED TO CONTAMINATED WATER IN ARTIFICIAL SIS METHODS FOR CALCULATING FLEXIBILITY OF NOZZLES AND BRANCH CONNECTIONS. ECOSYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH NUR G/ B1 RO1: ASSESSMENT OF LEAK DETECTION SYS- NUREG 5049: PRESSURE VESSEL FLUENCE ANALYSIS AND NUREG/CR 4828. FATIGUE CRACK GROWTH OF PART-THROUGH NUR G/CR5 1 F ING AND MITIGATING BATTERY CHARG-CRACKS IN PRESSURE VESSEL AND PIPING STEELS. Air Environ ^ ER AND INVERTER AGING. NU E C

                                                                                                                                                                ^                         ^   ^b    b 60 RO1: FLAW DENSITY EXAMINATIONS OF A CLAD DOILING WATER REACTOR PRESSURE VESSEL SEGMENT                                         URIZED AT A EACTO l                                               NUREG/CR 4079 V01 DEMONSTRATION OF PERFORMANCE MOD-                                    NUREG/CR-5053: OPERATING EXPiRIENCE AND AGING ASSESS-l                                                    ELING OF A LOW. LEVEL WASTE SHALLOW. LAND BURIAL SITE.A                             MENT OF MOTOR CONTROL CENTERS.

l Co-npanson Of Predictwe Radionuclides Triinsport Modeling Versus NUREG/CR 5061: THREE-FREQUENCY EDDY-CURRENT INSTRU-l Field Observations At The Nitrate Disposal Pit Site. Chalk River Nu- MENT. clear Late NUREG/CR-5063 DEVELOPMENT OF A MECHANISTIC UNDER. NUREG/CR-4880 V01: CHARACTER 12ATION OF IRRADIATED CUR- STANDING OF RADIATION EMBRITTLEMENT IN PIACTOR PRES-RENT. PRACTICE WELDS AND A533 GRADE D CLASS 1 PLATE SURE VESSEL STEELS. Final Report FOR NUCLEAR PRESSURE VESSEL SERVICE NUREG/CR-5073: OUANTIFICATION OF MARGINS IN PIPING NUREG'CR 4B00 V02 CHARACTERIZATION OF IRRAD'ATED CUR. SYSTEM SEISMIC RESPONSE Methodologies And Damping RENT-PRACTICE WELDS AND A533 GRADE B CLASS 1 PLATE NUREG/CR-6075. THE SAFT UT REAL-TIME INSPECTION SYSTEM - FOR NUCLEAR PRESSURE VESSEL SERVICE. OPERATIONAL PRINCIPLES AND IMPLEMENTATION.

l l NRC Contract Sponsor Index 121 NUREG/CR-5076: AN APPROACH TO THE QUANTIFICATION OF NUREG/CR-5185: STEAM GENERATOR GROUP PROJECT. Task 13 SEISMIC MARGINS IN NUCLEAR POWER PLANTS.The importance Final Report: Nondestructive Examination (NDE) Vahdation. Of BWR Plant Systems And Functions To Seismsc Margins NUREG/CR 5192: TESTING OF A NATURALLY AGED NUCLEAR I NUREG/CR 5080: A STUDY OF NEW ENGLAND SEISMICITY WITH POWER PLANT INVERTER AND BATTERY CHARGER. j EMPHASIS ON MASSACHUSETTS AND NEW HAMPSHIRE.Fenal NUREG/CR 5201: EXPERIMENTAL ASSESSMENTS OF GUNDREM- 1 Report Covenng The Penod 1976-1985 MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EF- j NUREG/CR 5083. DESIGN. CONSTRUCTION AND INSTRUMENTS- FECTS STUDIES. TION OF A 1/6 SCALE REINFORCED CONCRETE CONTAINMENT NUREG/CR 5203. DYNAMIC AMPLIFICATION OF ELECTRICAL CABi-BUILDING NETS. NUREG/CR-5086: PLATINUM CATALYTIC IGNITERS FOR LEAN HY* NUREG/CR-5204. LOW-LEVEL RADIOACTIVE WASTE SOURCE DROGEN-AIR MIXTURES NUREG/CR-5094: APPLICATION OF STOCHASTIC METHODS TO TERM MODEL DEVELOPMENT AND TEPING. NUREG/CR-5207; FRACTURE EVALUATivN OF SURFACE CRACKS THE SIMULATION OF LARGE-SCALE UNSATURATED FLOW AND TRANSPORT. EMBEDDED IN REACTOR VESSEL CLADDING Matenal Property g NUREG/CR-5096. EVALUATION OF SEALS FOR MECHANICAL PEN-ETRATIONS OF CONTAINMENT BUILDINGS. NUR / R 5'209: DESIGhl PROVISIONS FOR TANGENTIAL SHEAR NUREG/CR-5097: SIMULATION OF LIQUID AND VAPOR MOVE- IN CONTAINMENT WALLS. MENT IN UNSATURATED FRACTURED ROCK AT THE APACHE NilREG/CR-5210: TECHNICAL FINDINGS DOCUMENT FOR GENER. LEAP TURF SITE Models And Strategies. IC ISSUE $1: IMPROVING THE RELIABILITY OF OPEN-CYCLE NUREG/CR-5099: EVALUATION OF MATERIALS OF CONSTRUC. SERVICE-WATER SYSTEMS. TION FOR THE REINFORCED CONCRETE REACTOR CONTAIN. NUREG/CR 5229- ANNUAL REPORT OF THE TMI-2 EPICOR-ll MENT MODEL. RESIN / LINER INVESTIGATION. Low. Level Waste Data Base Devel-NUREG/CR-5105: RESPONSE MARGINS INVESTIGATION OF opment Program For Fiscal Year 1988. PlPING DYNAMIC ANALYSES USING THE INDEPENDENT SUP. NUREG/CR-5233: A COMPUTER CODE FOR FIRE PROTECTION PORT MOTION METHOD AND PVRC DAMPlNG. AND RISK ANALYSIS OF NUCLEAR PLANTS. NUREG/CR-5107: HYDROGEOLOGIC CHARACTERIZATION OF NUREG/CR-5240: COMPARATIVE EVALUATION OF SELECTED BASALTS.The Northern Ram Of The Columbia Plateau Physiographic CONTINUUM AND DISCRETE FRACTURE MODELS. Emphasis On Province And Of The Creston Study Area. Eastern Washington. Despersivity Calculations For Application To Fractured Geologic NUREG/CR-5108: THERMODYNAMIC PROPERTIES OF TC(IV) Media, Creston Study Area. Eastern Washington. OXIDES Solubilities And The Electrode Potentsal Of The Tc(Vil)/ NUREG/CR-5248: PRIORITIZATION OF TIRGALEX RECOMMENDED Tc(IV)-Oxide Couple. COMPONENTS FOR FURTHER AGING RESEARCH. NUREG/CR 5123: STUDIES OF THE PATTEF,N AND AGES OF NUREG/CR-5255: STABLE ISOTOPES OF AUTHIGENIC MINERALS POST METAMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA IN VARIABLY-SATURATED FRACTURED TUFF. AND NORTH CAROUNA. NUREG/CR-5258 V01: GEORGIA / ALABAMA REGIONAL SEISMO. NUREG/CR-5129: EXPERIMENTAL ASSESSMENT OF THE INFLU- GRAPHIC NETWORK. Annual Report August 1985 - June 1986. EfeCE OF DYNAMIC LOADING ON THE PERMEABILITY OF WET NUREG/CR-5277: THE TENSORIAL NATURE OF EFFECTIVE PO-AND OF DRIED CEMENT BOREHOLE SEALS. ROSITY AND LARGU-SCALE DISPERSION NUREG/CR.5130: BLNTONITE BOREHOLE PLUG FLOW TESTING COEFFICIENTS.Apphcation To The Creston Study Area. Eastern WITH FIVE WATER TYPES. NUREG/CR-5131: PRELIMINARY REVIEW OF MASS TRANSFER AND FLOW VISUAUZATION STUDIES AND TECHNIOUES RELE- DIVI NO REACTOR ACCIDENT ANALYSIS (870413-880716) VANT TO THE STUDY OF EROSION CORROSION OF REACTOR NUREG/C43509: POWER SPECTRAL DENSITY FUNCTIONS COM-PIPING SYSTEMS PATIBLE WITH NRC REGULATORY GUIDE 1.60 RESPONSE NUREG/CR 5134 APPLICATION OF ACOUSTIC LEAK DETECTION SPECTRA. TECHNOLOGY FOR THE DETECTION AND LOCATION OF' LEAKS NUREG/CR-4508: BEHAVIOR OF A CORIUM JET IN HIGH PRES-IN UGHT WATER RE ACTORS SURE MELT EJECTION FROM A REACTOR PRESSURE VESSEL. NUREG/CR-5136- FATIGUE STRENGTH OF SMOOTH AND NUREG/CR-4551 V5 DRF: EVALUATION OF SEVERE ACCIDENT NOTCHED SPECIMENS OF ASME SA 106-B STEEL IN PWR ENVl_ RISKS AND POTENTIAL FOR RISK REDUCTION.2 ION POWER RONMENTS PLANT. Draft Report For Comment. NUREG/CR-5137: BIODEGRADATION TESTING OF TMI-2 EPICOR il NUREG/CR4605: TRAINING MANUAL ON STATISTICAL METHODS WASTE FORMS. FOR NUCLEAR MATERIAL MANAGEMENT. NUREG/CR-5141: AGING AND QUALIFICATION RESEARCH ON SO. NUREG/CR-4625: THE POSTlRRADIATION EXAMINATION OF THE LENOID OPERATED VALVES. DC MELT DYNAMICS EXPERIMENTS. NUREG/CR-5142. DUCTILE TO BRITTLE TOUGHNESS TRANSITION NUREG/CR-4688 V02: QUANTIFICATION AND UNCERTAINTY ANAL-CHARACTERIZATION OF A533B STEEL. YSIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT NUREG/CR 5144: ACOUSTIC EMlSSION SYSTEM CALIBRATION AT WATER REACTORS (OVASAR).Part II: Sensitivity Analysis Tech- l WATTS BAR UNIT 1 NUCLEA*4 REACTOR. nmues. NUREG/CR-5149: EROSION-CORROSION OF PWR FEEDWATER NUREG/CR 4778: PREUMINARY STUDIES OF THE MORPHOLOGY PIPING SURVEY OF EXPERIENCE. DESIGN. WATER CHEMISTRY AND MATERnALS OF THERMAL GRADIENT TUBE DEPOSITS FROM FISSION PROD- l UCT RELEASE EXPERIMENTS  ! NUREG/CR-5153. THE TEACHABILITY AND MECHANfCAL INTEGRI' NUREG/CR4805 V02: REACTOR SAFETY RESEARCH SEMiANNU- l TY OF SIMULATED DECONTAMINATION RESIN WASTES SOUDI AL REPORT. July-December 1986 Volume 36. I NU EG/C 5 E ER EtTAL S ES MEN OF DAMPING IN G TER R O SF Y LOW ASPECT RATIO. REINFORCED CONCRETE SHEAR WALL NUREG/CR-4881: FISSION PRODUCT RELEASE CHARACTERISTICS NUl E I 5 REVIEW OF EROSION CORROSION IN SINGLE- INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACCI-PHASE FLOWS DENT CONDITIONS. NUREG/CR-5159. PREDICTION OF CHECK VALVE PERFORMANCE NUREG/CR4917: DCH-2.RESULTS FROM THE SECOND EXPERb AND DEGRADAPON IN NUCLEAR POWER PLANT SYSTEMS MENT PERFORMED IN THE SURTSEY DIRECT HEATING TEST NUREG/C45165: SEISMOLOGICAL INVESTIGATION OF EARTH. FACluTY. QUAKES IN THE NEW MADRID SEISMIC ZONE AND THE NORTH- NUREG/CR 4999: ESTIMATION OF RISK REDUCT10N FROM IM-EASTERN EXTENT OF THE NEW MADRID SEISMIC ZONE Final PROVED PORV REUABILITY IN PWRS Final Report. j Report. September 1981 December 1986. NUREG/CR 5000' METHODOLOGY FOR UNCERTAINTY ESTIMA- l NURE G/CR-5166. ELECTROCHEMICAL EVALUATION OF SOLID TION IN NUREG 1150 (DRAFT) Conclusions Of A Review Panel. 1 STATE PH SENSORS FOR NUCLEAR WASTE CONTAINMENT. NUREG/CR-5018 URANIUM OXIDE-IRON OXIDE MIXED AEROSOL ) NUREG/CR-5170. A REVIEW OF RESE ARCH CONDUCTED BY LOS EXPERIMENTS IN STEAM-AIR ATMOSPHERES.NSPP Tests j ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHA- 611.6t2.613 And 631. Data Record Report SIS ON THE MAXEY FLATS.KY, SHALLOW LAND WASTE BURIAL NUREG/CR-5029 MELT PROGRESSION IN SEVERELY DAMAGED , SITE REACTOR CORES-NUREG/CR.5180. CHEMICAL DECONTAMINATION AND CHENilCAL NUREG/CR-5032: MODEUNG TIME TO RECOVERY AND INITIATING CLEANING OF LWR COMPONENTS AND POSSIBLE INTERAC. EVENT FREQUENCY FOR LOSS OF OFFSfTE POWER INCIDENTS TiONS WITH METALLURGICAL AGING EFFECTS AT NUCLEAR POWER PLANTS. NUREG/CR 5182. THE SEISMIC CATEGORY t STRUCTURES NUREG/CR-5039 V01: REACTOR SAFETY RESEARCH SEMIANNU-PROGRAM Results For FY 1986 AL REPORT. January. June 1967. Volume 37.

122 NRC Contract Sponsor index < J NUREG/CR-5065: TIME- AND VOLUME AVERAGED CONSERVA- NUREG/CR-5095 V01: THERMODYNAMIC NONEOUILIBRtUM IN TION EOUATIONS FOR MULTIPHASE FLOW USING MASS- POST CRITICAL-HEAT FLUX BOILING IN A ROD WElGHTED VELOCITY AND INTERNAL ENERGY. BUNDLE.Desenption Of Expenments And Sample Results. s NUREG/C45070. ANALYSIS OF NATURAL CONVECTON PHENOM- NUREG/CR.5095 V02: THERMODYNAMIC NONEOUILIB91UM IN t ENA IN A 3-LOOP PWR DURING A TMLB' TRANSIENT USING THE POST-CRITICAL-HEAT FLUX BOILING IN A ROD BUNDLE. Data For COMMIX CODE. Stabilaed Quench Front Tests.

• NUREG/CR-5071: TRAC SUPPORT SOrTWARE. NUREG/C45095 V03: THERMODYNAMIC NONEOUILlBRIUM IN '

NUREG/CR 5084: IFCf. AN INTEGRATED CODE FOR CALCULATION POST-CRITICAL-HEAT. FLUX BOILING IN A ROD BUNDLE. Data For ' OF ALL PHASES OF FUEL-COOLANT INTERACTIONS. Advancing Quench Front Tests " NUREG/CR-5113: FINDINGS OF THE PEER REVIEW PANEL ON THE 9EG/CR-5095 V04: THERMODYNAMIC NONEQUtLIBRIUM IN

                                                                                                                                   >ST CRITICAL-HEAT-FLUX DOILING IN A ROD BUNDLE. Data For NU E /C 51 9 M             LL     APH         NA ON bF HE SEVERE FUEL DAMAGE SCOPING TEST (SFD ST) FUEL ROD                                                            N8RE / 5 2 SE E E                              OENT INSIGHTS REPORT.

NUREG/CR-5178: EVALUATION OF GENERIC ISSUE NURE 1 51 DEB i SPERSAL FROM REACTOR CAVITIES 125 tl.7, REEVALUATE PROVIScht TO AUTOMATICALLY ISOLATE { DURtNG HIGH-PRESSURE MELT EJECTION ACCOENT SCENAR. FEEDWATER FROM STEAM GENERATOR DURING A LINE lOS BREAK. NUREG/C45157: THE DEVELOPMENT OF APRfLMOD2 - A COM. NUREG/CR-5194. RELAP5/ MOD 2 MODELS AND CORRELATIONS. PUTER CODE FOR CORE MELTDOWN ACCfDENT ANALYSIS OF NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT  ; BOILING WATER NUCLEAR REACTORS AND STI REQUIREMENTS AT THE ANO-1 NUCLEAR POWER NUREG/CR 5164. A SIMPLIFIED MODEL F'OR CALCULATING EARLY ITE CONSEQUENCES FROM NUCLEAR REACTOR ACCl- DIVI ION OF REGULATORY APPLICATIONS (POST 870413) NUREG/CR-3908: SURVEY OF THE STATE OF THE ART IN MITIGA-l DIVISION OF REACTOR & PLANT SYSTEMS (870413 880716) Tf0N SYSTEMS. t l NUREG/CR 4639 V01: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR4242 SURVEY OF LIGHT WATER REACTOR CONTAIN-ASSESSING REACTOR RELIABILITY 4 (NUCLARR) Volume MENT SYSTEMS. DOMINANT FAILURE MODES AND MITIGATION I Summa De ion. OPPORTUNITIES. Final Report. NUREG/Ck463 04 P1: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR4243 VALUE/lMPACT ANALYSIS FOR EVALUATING AL. ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guede,Part TERNATIVE MITIGATION SYSTEMS. 1: Overview Of NUCLARR Data Retneval NUREG/CR 4244: STRATEGIES FOR IMPLEMENTING A MITIGA-NUREG/CR-4639 V04 P2: NUCLEAR COMPUTERIZED LidRARY f OR TION POLICY FOR LIGHT WATER REACTORS. ASSESSING REACTOR RELIABILITY (NUCLARR) User's Guide,Part 2: Guide To Operations NUREG/CR-4555 R01: GENERIC COST ESTIMATES FOR THE DIS- j NUREG/CR4639 V04 P3 NUCLEAR COMPUTERIZED LIBRARY FOR POSAL OF RADIOACTIVE WASTES NUREG/CR-5009: ASSESSMENT OF THE USE OF EXTENDED i ASSESSING REACTOR RELIABILITY (NUCLARR). User's Guide Part BURNUP FUEL IN LIGHT WATER POWER REACTORS.  ! 3 NUCLARR System Desenption NUREG/CR-5067: EARLY AND CONTINUING EFFECTS OF COM-NUREG/CR4639 V05 P1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part BINED ALPHA AND BETA (RRADIATION OF THE LUNG Phase il ~] N8P0rt I 1: Summary Desenption 4 NUREG/CR4639 V05 P2[ NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5090- EFFECTS OF TEMPERATURE AND HUMOtTY ON RESPIRATOR FlT UNDER SfMULATED WORK CONDITIONS. ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part NUREG/CR-5138: VALIDATON OF GENERIC COST ESTtMATES 2- Human Error Probability (HEP) Estimates NUREG/CR-4639 V05 P3: NUCLEAR COMPUTERIZED LIBRARY FOR FOR CONSTRUCTION RELATED ACTIVITIES AT NUCLEAR ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual,Part POWER PLANTS.Finaf Report. 3' Hardware Component Failure Data HCFD NUREG/CR-51SB V01: WORLOWIDE ACTIVITIES ON THE REDUC. ' NUREG/CR4639 V05 P4 NUCLEAR MPU ERIZED LIBRARY FOR TON OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS. a l ASSESSING REACTOR REUABILITY (NUCLARR) Data Manual,Part .I 4 Summary Aggregations NUREG/CRS160. GUIDELINES FOR THE USE OF THE EEDB AT THE SUB-COMPONENT AND SUBSYSTEM LEVEL.Fenal Report. NUREG/CR-4780 V01: PROCEDURES FOR TREATING COMMON ;1 CAUSE FAILURES IN SAFETY AND RELIABILITY NUREG/CR-5198. INHALED (239)PUO(2) AND/OR TOTAL-BODY STUDfES Procedural Framework And Examples. GAMMA RADIATION Early Mortality And Mortudity in Rats And Dogs. I1 NUREG/CR 5212. EMERGENCY ENVIRONMENTAL SAMPLING AND fJUREG/CR4834 V02 RECOVERY ACTONS IN PRA FOR THE RISK ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES. .j(1 METHODS !NTEGRATION AND EVALUATON PROGRAM NUREG/CR 5218: FINANCIAL QUALIFICATIONS REVIEW OF APPLI. (RMIEP) Volume 2 Apphcation Of The Data-Based Method CANTS FOR NUCLEAR POWER PLANT CONSTRUCTION PER. d, I l NUREG/CR4636: APPROACHES TO UNCERTAINTY ANALYSIS IN 1 PROBABILISTIC RISK ASSESSMENT. MITS NUREG/CR-5223. SCINTILLATION FIBER DETECTOR FOR IN-VIVO l NUREG/CR4898. RESULTS OF SEMISC ALE MOOL2C FEEDWATER ENDOSCOPIC INTERNAL DOSIMETRY. d AND STEAM LINE BREAK (S-FS) EXPERIMENT SERIES. Bottom Mam Feedwater Lane Breah Acetdont Expenments. DIVISION OF SAFETY ISSUE RESOLUTION (POST 880717) d NUREG/CR4920 VD1. ASSESSMENT OF SEVERE ACCIDENT PRE- NUREG/CR 5078 V02: A REUABILITY PROGRAM FOR EMERGENCY VENTION AND MITIGATION FEATURES.BWR. MARK I CONTAIN. OlESEL GENERATORS AT NUCLEAR POWER PLANTS Maintenance. Surveillance And Condition Monitonng HENT DESIGN. NUHEG/CR4920 V02. ASSESSMENT OF SEVERE ACCOENT PRE. NUREG/CR 5140. VALUE-lMPACT ANALYSIS FOR EXTENSION OF VENTON AND MlTIGATION FEATURE S BWR. MARK ll CONTAIN. NRC BULLETIN 85 03 TO COVER ALL SAFETY RELATED MOVS. NUREG/CR5225. AN OVERVIEW OF BWR MARK-1 CONTAINMENT MENT DESIGN NUREG/CR-4920 V03. ASSESSMENT OF SEVEPE ACCIDENT PRE- VENTING RISK IMPLICATIONS. DTISION OF SYSTEMS RESEARCH (POST 880717) VENTION AND MITIGATON FEATURES BWR. MARK Ill CONTAIN- NUREG/CR-4639 V02: NUCLEAR COMPUTERIZED LIBRARY FOR gggy ogg,gg ASSES $ LNG REACTOR RELIABILITY (NUCLARR). Programmer's NUREG/CR4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE. TON A D IGATION FEATURES.PW4.LARGE DRY CON- NUR CR4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR4920 V05: ASSESSUIENT OF SEVERE ACCIDENT PRE. ASSESSING REACTOR RELIABILITY (NUCLARR) Guide To Data Processing And Reviseon Part 1: Technical Overwew. VENTION AND MITIGATON FEATURES.PWR.lCE CONDENSER NUREG/CR4639 V03 P2. NUCLEAR COMPUTERIZED LIBRARY FOR CONTAINMENT DESIGN NUREG/CR$015 IMPROVED kEllABILITY OF RESOUAL HEAT RE. ASSESSING REACTOR RELIABILITY {NUCLARR)Gende To Data MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF Processing And RemssonPart 2: Hurr . Error Probatnitty Data Entry GENERIC ISSUE 99. And Revison Procedures. NUREG/CR5042 $01. EVALUATON OF EXTERNAL HAZARDS TO NUREG/CR4639 V03 P3: NUCLEAR COMPUTERIZED LIBRARY FOR NUCLEAR POWER PLANTS IN THE UNITED STATESScismic ASSESSING REACTOR RELIABILITY (NUCLARR)Gtude To Data Hazard Proceang And Rewsion. Part 3 Hardws e Component failure Data NUREG/CR 5072- DECAY HEAT REMOVAt USING FEED-AND- Entry And Remsion Procedures BLEED FOR U S PRESSURIZED W ATER LEACTORS NUREG/CR4914 THE INFLUENCE OF SELECTED CONTAINMENT NUREG/CR-5078 V01: A RELIABILITY PROCRAM FOR EMERGENCY STRUCTURES ON DEBRIS DISPERSAL AND TRANSPORT FOL-DIESEL GENERATORS AT NUCLEAR POWER PLANTSProDram LOWING HIGH PRESSURE MELT EJECTION FROM THE REAC-Structure TOR VESSEL

[ l l NRC Contract Sponsor index 123 NUREG/CR491t, HECTR ANALYSES OF THE NEVADA TEST SITE NUREG/CR-4665. CLOSEOUT OF IE BULLETIN 83 08 ELECTRICAL (NTS) PREMIXED COMBUSTICN E XPERIMENTS CIRCUll BREAKERS WITH AN UNDERVOLTAGE TRIP FEATURE NURE G/CR-4993. A STANDARD PROBLEM FOR HECTR-MAAP IN USE IN SAFETY RELATED APPLICATIONS OTHER THAN THE , COMPARISON incomplete Burnmg REACTOh TRIP SYSTEM l NUREG/CR 5039 V02. REACTOR SAFETY RESEARCH SEMIANNU- NUREG/CR 4932: CLOSEOUT OF IE DULLETIN BO 0310SS OF Al REPORT. July-Docemter 1987 Reactor Safety Research Program. ] CHARCOAL FROM STANDARD TYPE II,TWO-INCH. TRAY AD- ' NUREG/CR $109 RELOCATION OF METALLIC CONSTITUENTS IN SORBER CELLS j CORE DEDRis BEDS NUREG/CR 4933. CLOSEOUT OF IE DULLETIN B0-19 FAILURES OF NUREG/CR 5133 A COMPUTATIONAL MODEL FOR CRITICAL MERCURY-WETTED MATRlx RELAYS IN REACTOR PROTECTIVE FLOW THROUGH INTERGRANULAR STRESS CORROSION SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DE-CRACKS SIGNED BY COMBUSTION ENGINEERING NURE G/CR 5135. THE THERMAL HYDRAULICS OF SBLOCAS REL-NUREG/CR 4935: CLOSEOUT OF IE BULLETIN 85-02-UNDERVOL-ATIVE TO PRESSURIZED THERMAL SHOCK NUREG/CR-5162 CHARM.A MODEL FOR AEROSOL BEHAVIOR IN TAGE TRIP ATT ACHME NTS OF WESTINGHOUSE DB-50 TYPE RE- f ACTOR TRIP DREAKERS. TIME VARYING THERMAL HYDRAULIC CONDITIONS. NUREG/CR-5171 FLOW VISUALIZATION STUDY OF POST CRITI' NUREG/CR-4991: EVALUATION AND PROPOSED IMPROVEMENTS CAL HEAT FLUX REGION FOR INVERTED BUBBLY SLUG AND TO EFFECTIVENESS OF U.S. NUCLEAR REGULATORY COMMIS-SION GENERIC COMMUNICATIONS NUREG/CR 51 SUB S ON FOR THE CSNI/GREST BENCH' MARK EXERCISE ON CHEMICAL THERMODYNAMIC MODELING FR MB dNTROL DS NUREG/CR-5191: CLOSEOUT OF lE BULLETIN 60-17.F AllORE OF IN CORE CONCRETE INTERACTION RELEASES OF RADIONU' CUDE S 76 OF 185 CONTROL RODS TO FULLY INSERT DURING A NUREG/CR 5214. ANALYSES OF NATURAL CIRCULATION DURING SCRAM AT A DWR A SURRY STATION DLACKOUT USING SCDAP/RELAP5 DIVISION OF ENGINEERING & SYSTEMS TECHNOLOGY (POST NUREG/CR-5220 VOI: DIAGNOSIS OF CONDENSATION-INDUCED 870411) f WATERHAMMER Methods And Background NUREG/CR-3950 V04: FUEL PERFORMANCE ANNUAL REPORT NUREG/CR 5220 V02. DIAGNOSIS OF CONDENSATION INDUCED FOR 1986 NUREG/CRI4960- CONTROL ROOM HABITABILITY SURVEY OF LI-NU EG/C 52 U ERT TIES IN MODELING AND SCALING IN CENSED COMMERCIAL NUCLEAR POWER GENERATING STA. THE PREDICTION OF FUEL STORED ENERGY AND THERMAL

       "                                                                  NI E CR-5048 ' REVIEW OF THE NATURAL CIRCULATION NURE C                     41: DEVELOPMENT OF PROGRAMMATIC PERFORM. EFFECT IN THE VERMONT YANKEE SPENT FUEL POOL. Docket       j ANCE INDICATORS                                                      No. 50-271 (Verrnont Yankee Nuclear Power Corp) l NUREG/CR-5242: A FAST BOTTOM UP ALGORITHM FOR COMPUT.                 NUREG/CR 5150:          STEAM       GENERATOR     OPERATING ING THE CUT SETS OF NONCOHERENT FAULT TREES                          EXPERIENCE Update For 1984-1986.                          l DIVISION OF REACTOR SYSTEM SAFETY (860720-870413)'                    DIVISION OF REACTOR INSPECTION & SAFEGUARDS (POST NUREG/CR-5219 THE MixtNG OF tMMISCIBLE UOUlO LAYERS By                  870411)

GAS BUCDLINa NUREG/CR-5227. FITNESS FOR DUTY IN THE NUCLEAR POWER DIVISION OF ACCIDENT EVALUATION (POST 840101) INDUSTRY.A Review Of Technical Issues 2 NUREG/CA-4312 V01: RELAPS/ MOD 2 CODE MANUALVolume 1: DIVISION OF RADIATION PROTECTION & EMERGENCY PREPARED- I Code Structure, Systems Models And Solution Methods NESS (POST 87041t) NUREG/CR-4873. BENCHMARK STUDY OF THE l-DYNEY EVACU-i EDO OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80) ATION TIME ESTIMATE COMPUTER CODE. OUICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST NUREG/CR-4874 THE SENSITIVITY OF EVACUATION TIME ESTI-670411) MATES TO CHANGES IN INPUT PARAMETERS FOR THE l DYNEV , NUREG/CR 4811: THE ECONOMIC COSTS OF RADIATION-IN- COMPUTER CODE. i l DUCED HE AL'iH EFFECTS Estimation And Simulation NUREG/CR-5038. OPTIM:ZATION OF THE CONTROL OF CONTAMI. l l DIVISION OF REACTOR PROJECTS IlUV.V & SPECIAL PROJECTS NATION AT NUCLEAR POWER PLANTS. (POST B70411  ! NUREG/CR 5055 ATMOSPHE RIC DIF FUSION FOR CONTROL i NUREG/CR 5190 CLOSEOUT OF IE BULLETIN 80-14 DEGRADA- ROOM HABITABILITY ASSESSMENTS I TION OF BWR SCRAM DISCHARGE VOLUME CAFABILITY. NUREG/CR 5106 USER'S GUIDE FOR THE TACT 5 COMPUTER DIVISION OF OPERMIONAL EVENTS ASSESSMENT (POST B70411) CODE. NUREGICR 4523 CLOSEOUT OF IE BULLEllN 80-13 CRACKING IN DIVISION OF LICENSEE PERFORMANCE & OVALITY EVALUATION CORE SPRAY SPARGERS (POST B70411) NURE G /CR-4662 CLOSEOUT OF IE DULLETIN 801B M AINTE- NUREG/CR-5147. FUNDAMENTAL ATTRIBUTES OF A PRACTICAL NANCE OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL CONFIGURATION MANAGEMENT PROGRAM FOR NUCLE AR CHARGING PUMPS FOLLOWING SECONDARY SIDE HIGH PLANT DESIGN CONTROL ENERGY LINE RUPTURE NUREG/CR $151 PERFORMANCE-BASED INSPECTIONS l l

1 1 l 1 l l l l L

Contractor index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles oi their reports. If further information is needed, refer to the main citation by the NUREG/CR number. APPLIED RISK TECHNOLOGY CORP. NUREG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PERFORM-NUREG/CR 5076: AN APPROACH TO THE QUANTIFICATION OF SEIS. ANCE INDICATORS. MIC MARGINS IN NUCLEAR POWER PLANTS.The importance Of BWR Plant Systems And Functions To Seismic Margins. BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES NUREG/CP-0092. PROCEEDINGS OF THE SEMINAR ON LEAK-ARGONNE NATIONAL LABORATORY BEFORE-BREAK. Progress in Regulatory Policies And Supportmg Re-NUREG/CR4667 Vod: ENVIRONMENTALLY ASSISTED CRACKING IN - search. LIGHT WATER REACTORS. Semsannual Report. October 1986 - March NUREG/CR4082 V06: DEGRADED PIPING PROGRAM PHASE II.Smth 1987. Program Report. October 1986 - September 1987. NUREG/CR-4813 R01: ASSESSMENT OF LEAK DETECTION SYSTEMS NUREG/CR4857: CADET:A DECISION SUPPORT SYSTEM FOR LIGHT FOR LWRS. WATER REACTOR SAFETY. NUREG/CR 4960: CONTROL ROOM HABITABILITY SURVEY OF Lb CENSED COMMERCIAL NUCLEAR POWER GENERATING STA- BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST TIONS. LABORATORY NUREG/CR-5065: TIME AND VOLUME-AVERAGED CONSERVATION NUREG-1320: NUCLEAR FUEL CYCLE FACILITY ACCfDENT ANALYSIS EQUATIONS FOR MULTIPHASE FLOW USING MASS-WEIGHTED VE' HANDBOOK. LOCITY ANDINTERNAL ENERGY. NUREG/CP-0093: PROCEEDINGS OF THE MEETING ON ULTRASEN-NUREG/CR 5070: ANALYSIS OF NATURAL CONVECTION PHENOM-SITIVE TECHNIOUES FOR MEASUREMENT OF URANIUM IN BIO-ENA IN A 3-LOOP PWR DURING A TMLB' T9ANSIENT USING THE LOGICAL SAMPLES AND THE NEPHROTOxlCITY OF URANIUM. COMMIX CODE. NUREG/CR-5082: SfMULATION EXPERIMENTS ON TWO PHASE NAT* NUREG/CP 0099: PROCEEDINGS OF THE PUBLIC WORKSHCd /OR URAL CIRCULATION IN A FREON 113 FLOW VISUALIZATION LOOP. NRC RULEMAKING ON MAINTENANCE OF NUCLEAR '3WER PLANTS NUREG/CR-5131: PRELIMINARY . REVIEW OF MASS TRANSFER AND NUREG/CR 0130 ADD 04: TECHNOLOGY, SAFETY AND COLTS OF FLOW VISUALIZATION STUDIES AND TECHNIOUES RELEVANT TO THE STUDY OF EROSION CORROSlON OF REACTOR PIPING SYS- DECOMMISSIONING A REFERENCE PRESSURIZED WATER REAC-TOR POWER STATION. Technical Support For Decommissioning Mat-NU CR-5134: APPLICATION OF ACOUSTIC LEAK DETECTION NP / 672 D TECHNO O AFE C TS OF TECHNOL Y R HE DETECTION AND LOCATION OF LEAKS l.N DECOMMISSIONING A REFERENCF BOILING WATER REACTOR NUREG/CR-5149: EROSION-CORROSION OF PWR FEEDWATER POWER STATION. Technical Support For Decommissioning Matters PIP NG SU VEY OF EXPERIENCE. DESIGN. WATER CHEMISTRY NUR / 2 TE M N TNE "' INTEGRITY NURE /CR 5156 REVIEW OF EROSION-CORROSION IN SINGLE- PR ^M " P NUqEG/ R 28 'P PULAT ON DOSE COMMITMENTS DUE TO NUREG/CR-5171:' FLOW VISUALIZATION STUDY OF POST CRITICAL RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES AT FLUX REGION FOR INVERTED BUBDLY, SLUG AND ANNULAR N E C'R-2850 V07: POPULATION DOSE COMMITMENTS DUE TO NUREG/CR-5100: CHEMICAL DECONTAMINATION AND CHEMICAL RADIOACTIVE RELEASES FROM NJCLEAR POWER PLANT Si1ES CLEANING OF LWR COMPONENTS AND POSSIBLE INTERACTIONS IN 1985. WITH METALLURGICAL AGING EFFECTS. NUREG/CR-3950 V04: FUEL PERFORMANCE ANNUAL REPORT FOR NUREG/CR-5229: ANNUAL REPORT OF THE TMI-2 EPICOR Il RESIN / 1986. i LINER INVESTIGATION Low-Level Waste Data B+.se Development NUREG/CR-4000 V02: THE MESORAD DOSE ASSESSMENT l Program For Fiscal Year 1988. NUREGI R 60 T ING MANUAL ON STATISTICAL METHODS ARIZONA, UNIV. OF, TUCSON, AZ FOR NUCLEAR MATERIAL MANAGEMENT. NUREG/CR-5097: SIMULATION OF LIQUID AND VAPOR MOVEMENT NUREG/CR-4811: THE ECONOMIC COSTS OF RADIATION-lNDUCED IN UNSATURATED FRACTURED ROCK AT THE APACHE LEAP HEALTH EFFECTS Estimation And Simulaton. TUFF SITE.Models And Strat ies. NUREG/CR4873: BENCHMARK STUDY OF THE l-DYNEV EVACU. NUREG/CR-5129: EXPERIMEN AL ASSESSMENT OF THE INFLUENCE ATION TIME ESTIMATE COMPUTER CODE. OF DYNAMIC LOADING ON THE PERMEABILITY OF WET AND OF NUREG/CR 4874: THE SENSITIVITY OF EVACUATION TIME ESTi-DRIED CEMENT BOREHOLE SEALS. MATES TO CHANGES IN INPUT PARAMETERS FOR THE l-DYNEV NUREG/CR-5130. BENTONITE BOREHOLE PLUG FLOW TESTING COMPUTER CODE. WITH FIVE WATER TYPES. NUREG/CR 4879 V01: DEMONSTRATION OF PERFORMANCE MOD-NUREG/CR4255: STABLE ISOTOPES OF AUTHIGENIC MINERALS IN ELING OF A LOW-LEVEL WASTE SHALLOW LAND BURIAL SITE.A VARIABLY-SATURATED FRACTURED TUFF- Compannon Of Predictive Radionuclides Transport Modeling Versus Field Observations At The Nitrate Disposal Pat Site. Chalk River Nuclear ARMY, DEPT. OF, ARMY ENGINEER WATERWAYS EXPERIMENT Labs. STATION NUREG/CR 4991: EVALUATION AND PROPOSED IMPROVEMENTS TO NUREG/CR-5041 V02: FIECOMMENDATIONS TO THE NRC FOR EFFECTIVENESS Or U S. NUCLEAR REGULATORY COMMISSION REVIEW CRITERIA FOR ALTERNATIVE METHODS OF LOW LEVEL GENERIC COMMUNICATIONS. RADIOACTIVE WASTE DISPOSAL Task 2b Earth-Mounded Concrete NUREG/CR-4997: METHODS FOR DESCRIBING AIRBORNE FRAC. 1 Bunkers TIONS OF FREE FALL SPILLS OF POWDERS AND LIOUIDS. NUREG/CR-5009: ASSESSMENT OF THE USE OF EXTENDED SATTELLE HUMAN AFFAIRS RESEARCH CENTERS BURNUP FUEL IN LIGHT WATER POWER REACTORS. NUREG/CR 4991: EVALUATION AND PROPOSED IMPROVEMENTS TO NUREG/CR-5016: COMPENDIUM AND COMPARISON OF INTERNA-EFFECTIVENESS OF U S. NL8 CLEAR REGULATORY COMMISSION TIONAL PRACTICE FOR PLUGGING. REPAIR AND INSPECTION OF GENERIC COMMUNICATIONS STEAM GENERATOR TUBING NUREG/CR 5227: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/CR 5047: RADIONUCLIDES ACCUMULATION BY AOUATIC INDUSTRY.A Review Of Technical issues. BIOTA EXPOSED TO CONTAMINATED WATER t,4 ARTIFICIAL ECO. 125

126 Contractor Index SYSTEMS BEFORE AND AFTER ITS PASSAGE THROUGH THE NUREG/CR 4920 V03: ASSESSMENT OF SEVERE ACCIDENT PRE-GROUND VENTION AND MITIGATION FEATURES.BWR. MARK 111 CONTAIN-NUREG/CR 5048 REVIEW OF THE NATURAL CIRCULATION EFFECT MENT DESIGN. IN THE VERMONT YANKEE SPENT FUEL POOL. Docket No. 50- NUREG/CR-4920 V04: ASSESSMENT OF SEVERE ACCIDENT PRE-271.(Vermont Yankee Nuclear Power Corp) VENTION AND MITIGATION FEATURES.PWR,LARGE DRY CON. NUREG/CR-5054: RECOMMENDATIONS TO THE NRC FOR REVIEW TAINMENT DESIGN. CRITERIA FOR ALTERNATIVE METHODS OF LOW LEVEL RADIOAC- NUREG/CR-4920 VOS: ASSESSMENT OF SEVERE ACCIDENT PRE-TIVE WASTE DISPOSAL Environmental Morntonng And Surveillance VENTION AND MITIGATION FEATURES:PWR,1CE-CONDENSER Programs - CONTAINMENT DESIGN. NUREG/Ch-5055. ATMOSPHERIC DIFFUSION FOR CONTROL ROOM NUREG/CR 4939 V01: IM9 ROVING MOTOR RELIABILITY IN NUCLEAR HABITABILITY ASSESSMENTS POWER PLANTS. Volume performance Evaluation And Maintenance NURE G/045058. PRA APPLICATIONS PROGRAM FOR INSPECTION Practmes. AT ARKANSAS NUCLEAR ONE UNIT 1. Docket No. 50-313.(Arkansas NUREG/CR-4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR Power And Ught Company) POWER PLANTS. Volume 2. Functional Indicator Tests On A Small - I NUREG/CR-5075: THE SAFT UT REAL TIME INSPECTION SYSTEM - Electne Motor Sublected To Accelerated Aging. OPERATIONAL PRINCIPLES AND IMPLEMENTATION NUREG/CR-4939 V03: IMPROVING MOTOR RELIABILITY IN NUCLEAR NUREG/CR-5144. ACOUSTIC EMISSION SYSTEM CA'llBRATION AT POWER PLANTS. Volume 3. Failure Analysis And Diagnostc Tests On ' WATTS BAR UNIT 1 NUCLEAR REACTOR. A Naturally Aged Lar Electnc Motor NUREG/CR-5185: STEAM GENERATOR GROUP PROJECT. Task 13 NUREG/CR-4999: ES MATION OF ' RISK REDUCTION FROM IM-Onal Report: Nondestructree Examination (NDE) Vahdation. PROVED PORV RELIABILITY IN PWRS NUREG/CR-5198. INHALED (239)PUO(2) AND/OR TOTAL-BODY NUREG/CR-5000: METHODOLOGY FOR' Final Report UNCERTAINTY ESTIMATION - A Fi NG TFRGN IN NUREG 1150 (DRAFT) Conclusions Of A Renew Panet NUREG CR 5210 E I NUREG/CR-5015: IMPROVED RELIABILITY OF RESIDUAL HEAT RE-ISSUE 51 IMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE. MOVAL CAPABILITY IN PWRS AS RELATED TO RESOLUTION OF WATER SYSTEMS. GENERIC ISSUE 99. NUREG/CR-5212: EMERGENCY ENVIRONMENTAL SAMPLING AND NUREG/CR 5038: OPTIMIZATION OF THE CONTROL OF CONTAMINA-ANALYSIS FOR RADIOACTIVE MATERIAL FACILITIES. NUREG/CR 5218 FINANCIAL QUALIFICATIONS REVIEW OF APPLI. TION AT NUCLEAR POWER PLANTS. CANTS FOR NUCLEAR POWER PLANT CONSTRUCTION PERMITS NUREG/CR 5051: DETECTING AND MITIGATING BATTERY CHARGER NUREG/CR-5227: FITNESS FOR DUTY IN THE NUCLEAR POWER AND INVERTER AGING. INDUSTRY.A Remew Os Techncalissues NUREG/CR-5052: OPERATING EXPERIENCE AND AGING ASSESS-NUREG/CR-5241: DEVELOPMENT OF PROGRAMMATIC PERFORM- MENT OF COMPONENT COOLING WATER SYSTEMS IN PRESSUR-ANCE INDICATORS. 12ED WATER REACTORS.

             . NOREG/CR-5248: PRIORITIZATION OF TIRGALEX-RECOMMENDED                  NUREG/CR-5053: OPERATING EXPERIENCE AND AGING ASSESS-COMPONENTS FOR FURTHER AGING RESEARCH.                                 MENT OF MOTOR CONTROL CENTERS.

NUREG/CR-5105: RESPONSE MARGINS INVESTIGATION OF P& PING EROOKHAVEN NATIONAL LABORATORY DYNAMIC ANALYSES USING THE INDEPENDENT SUPPORT NUREG/CP-0091 Vot: PROCEEDINGS OF THE FIFTEENTH WATER MOTION METHOD AND PVRC DAMPlNG. REACTOR SAFETY INFORMA110N MEETING- NUREG/CR 5132: SEVERE ACCIDENT INSIGHTS REPORT. NUREG/CP.0091 V02: PROCEEDINGS OF THE FIFTEENTH WATER NUREG/CR-5140: VALUE-lMPACT ANALYSIS FOR EXTENSION OF REACTOR SAFETY INFORMATION MEETING- NRC BULLETIN 85 03 TO COVER ALL SAFETY-RELATED MOVS. NUREG/CP-0091 V03: PROCEEDINGS OF THE FIFTEENTH WATER NUREG/CR-5145; FAILURE INVESTIGATION OF 3M SERIES 900 REACTOR SAFETY INFORMATION MEETING. STATIC ELIMINATORS NUREG/CP 0091 V04 PROCEEDINGS OF THE FIFTEENTH WATER NUREG/CR-5146: DEBNS DISPERSAL FROM REACTOR CAVITIES CE Di OF THE FIFTEENTH S E E EMW A@N SOE NUF E /CPO VD D N G/CR-5153: THE TEACHABILITY AND MECHANICAL INTEGRITY NU EG/C 0091 5 D NGS T IFTEENTH WATER OF SIMULATED DECONTAMINATION RESIN WASTES SOLIDIFIED REACTOR SAFTEY INFORMATION MEETING. NUREG/CP-0091 V06. PROCEEDINGS OF THE FIFTEENTH WATER IN CEMENT ANO VINYL ESTER-STYRENE. NUREG/CR-5156: REVIEW OF EROSION-CORROSION IN SINGLE-REACTOR SAFETY INFORMATION MEETING. NUREG/CR-2331 V7N2 3 SAFETY RESEARCH PROGRAMS SPON- . PHASE FLOWS. SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-5158 V01: WORLDWIDE ACTIVITIES ON THE REDUCTION , OF OCCUPATIONAL EXPOSURE AT NUCLEAR POWER PLANTS - j RESEARCH Progress Report,Apnt September 1987. NURE G/CR2331 V8N1-2. SAFETY RESEARCH PROGRAMS SPON. NUREG/CR-5164: A SIMPLIFIED MODEL FOR CALCULATING EARLY a SORED BY OFFICE OF NUCLEAR REGULATORY OFFSITE CONSEQUENCES FROM NUCLEAR REACTOR ACCl- ') I RESE ARCH Progress ReportJanuary-June 1980. DENTS. NUREG/CR 2907 v06. RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR-5192: TESTING OF A NATURALLY AGED NUCLEAR I NUCLEAR POWER PLANTS Annual Report For 1985. POWER PLANT INVERTER AND BATTERY CHARGER l NUREG/CR 2907 V07 RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR-5200: EVALUATION OF RISKS ASSOCIATED WITH AOT NUCLEAR POrKR PL ANTS Annual Report For 1986. AND STI REQUIREMENTS AT THE ANO 1 NUCLEAR POWER NUREGICR-3444 V0k THE IMPACT OF LWR DECONTAMINATION PLANT- ) ON SOLIDIFICATION WASTE DISPOSAL AND ASSOCIATED OCCU- NUREG/CR 5203: DYNAMIC AMPLIFICATION OF ELECTRICAL CABi- i PAT?ONAL EXPOSURE Annual Report, FY 19B7. NETS.  ! NUREG/CR-4551 Y5 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR-5204: LOW LEVEL RADIOACTIVE WASTE SOURCE TERM l RISKS AND POTENTIAL FOR RISK REDUCTION. ZION POWER MODEL DEVELOPMENT AND TESTING. l PLANLDraft Report For Comrnent NUREG/CR 5232. UNCERTAINTIES IN MODELING AND SCALING IN l NUREG/CR-4659 V02: 3EISMIC FRAGILITY OF NUCLEAR POWER THE PREDICTION OF FUEL STORED ENERGY AND YHERMAL RE- j PLANT COMPONENT S (PHASE 10 Motor Control SPONSE. l Center. Switchboard.Panelboard And Power Supply ] NUREG/CRa688 V02 OUANTIFICATION AND UNCERTAINTY ANALY. CALIFORNIA, UNIV. OF, BERKELEY. CA i NUREG/CR-4918 V02: CONTROL OF WATER INFILTRATION INTO l SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT ' WATER REeCTORS (OUASAR)Part it Sensitiwty Ana9 sis Tech- NEAR SURFACE LLW DISPOSAL UNITS Tasit Report . A Discussion. nmues NUREG/CR-3133: A COMPUTATIONAL MODEL FOR CRITICAL FLOW NUREG/CH 4784 INFLUENCE OF GROUNDWATER ON SOIL.STRUC- THROUGH INTERGRANULAR STRESS CORROSION CRACKS. TURE INTERACTION. NUREG/CR4BB1: FISSION PRODUCT RELEASE CHARACTERISTICS CALIFORNIA, UNIV. OF, SA.WTA BARBARA, CA i l INTO CONTAINMENT UNDER DESIGN BASIS AND SEVERE ACC6- NUREG/CR-5135: THE THERMAL HYDRAULICS OF SBLOCAS RELA-

                  ' DENT CONDITIONS                                                      TIVE TO PRESSURIZED THERMAL SHOCK.

NUREG/CR4920 VDt. ASSESSMENT OF SEVERE ACCIDENT PRE-VENTION AND MITIGATION FEATURES BWR. MARK 1 CONTAIN- CALSPAN CORP. (SUBS. ARVIN INDUSTRIES / FRANKLIN RESEARCH MENT DESIGN CENTER) NURE G/CR-4920 V02 ASSESSMENT OF SEVERE ACCIDENT PHE- NUREG/CR 4992 V01: AGING AND SERVICE WEAR OF MULTISTAGE VENTION AND MITIGATION FEATURES BWR. MARK 11 CONTAIN- SWITCHES USED IN SAFETY SYSTEMS OF NUCLEAR POWER MENT DESIGN PLANTS Operating Expenence And Failure identif cation. w_____-________-_-_____________-_-_-_______-______-_________

Contractor index 127 , I CENTER FOR PLANNING & RESEARCH. INC. NUREG/CR-5031: SIGNIFICANCE OF IN-STRUCTURE GENERATED s NUREG/CR-5223: SCINTILLATION FIBER DETECTOR FOR IN-VIVO MOTION IN SEl5MIC QUALIFICATION TESTS OF CABINET MOUNT- J ENDOSCOPIC INTERNAL DOSIMETRY. ED ELECTRICAL DEVICES.  ? NUREG/CR-5043: CONTAINMENT PENETRATION SYSTEM (CPS) i CENTRAL HESEARCH INSTITUTE OF ELECTRIC POWER INDUSTRY TESTS UNDER ACCIDENT LOADS. NUREG/CP 0092: PROCEEDINGS OF THE SEMINAR ON LEAK. NUREG/CR-5050: ANNOTATED BIBuOGRAPHY OF RELIABILITY AND l BEFORE BREAK. Progress in Regulatory Pohcies And Supportmg Re- RISK DATA SOURCES. search. NUREG/CR-5072: DECAY HEAT REMOVAL USING FEED-ANC-BLEED FOR U.S. PRESSURIZED WATER REACTORS. NUR C 5233 COMPUTER CODE FOR FIRE PROTECTION AND NUREG/CR-5119: METALLOGRAPHIC EXAMINATION OF THE SEVERE RISK ANALYSIS OF NUCLEAR PLANTS. FUEL DAMAGE SCOPING TEST (SFD-ST) FUEL ROD BUNDLE CROSS SECTIONS. COLUM51A UNIV., NEW YORK, NY NUREG/CR-5137: BIODEGRADATION TESTING OF TMi-2 EPICOR-li NUREG/CR-3509: POWER SPECTRAL DENSITY FUNCTIONS COM- WASTE FORMS. PATIBLE WITH NRC REGULATORY GUIDE 1.60 RESPONSE SPEC. NUREG/CR-5178: EVALUATION OF GENERIC ISSUE i TRA. 125.II.7, REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE

                                                                                                                                                                                               ]

FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK. i CONSTRUCTION TECHNOLOGY LABORATORIES NUREG/CR-5191: RELAPS/ MOD 2 MODELS AND CORRELATIONS. l NUREG/CR-5209. DESIGN PROVISIONS FOR TANGENTIA'. SHEAR IN NUREG/CR 5214: ANALYSES OF NATURAL CIRCULATION DURING A - CONTAINMENT WALLS. SURRY STATION BLACKOUT USING SCDAP/RELAP5. CREARE, INC- NUREG/CR-5225: AN OVERVIEW OF BWR MARK 1 CONTAINMENT l VENTING FilSK IMPLICATIONS. i NUREG/CR-5220 V01: DIAGNOSIS OF CONDENSATION INDUCED NUREG/CR-5229: ANNUAL REPORT OF THE TMI-2 EPICOFull RESIN / ( NUR / 522 0 A O OF CONDENSATION-INDUCED LINER INVESTIGATION.Lc.w Level Waste Data Base Development j WATERHAMMER. Case Stud #es. arn 6 Rscal Year N -- DAVID W. TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER ENERGY, INC.  ! NUREG/CR-5142: DUCTILE TO BRITTLE TOUGHNESS TRANSITION NUREG/CR-5078 V02: A RELIABILITY PROGRAM FOR EMERGENCY CHARACTERIZATION OF A533B STEEL DIESEL GENERATORS AT NUCLEAR POWER PLANTS Maintenance. Surveillance And Condition Monitonng. EGSG IDAHO,INC. (SUBS OF EG&G. INC.) NUREG/CR 4312 V01: RELAPS/ MOD 2 CODE MANUALNolume 1: Code ENGINEERING & ECONOMICS RESEARCH, dNC. Structure. Systems Models And Soluteon Methods. NUREG/CR 3899 S01: UTILITY FINANCIAL STABILITY AND THE NUREG/CR-4312 V02 R1: RELAP5/ MOD 2 CODE MANUALVolume 2: AVAILABILITY OF FUNDS FOR DECOMMISSIONING An Analysis Of NUR$/ 63 VO N L C APUTERIZED LIBRARY FOR AS.

                                                                                                                       '"       ^"       " "" "E SESSING REACTOR RELIABILITY (NUCLARR) Volume LSummary Do"                 EOE, INC.

N R[ CFI4639 V02: NUCLEAR COMPUTERIZED LIBRARY FOR AS- 5073 OUANTIFICATION OF MARGINS IN PIPING SYSTEM

                                                                                                                                            . e        I gies And Dampmg.

SESSING REACTOR RELIABILITY (NUCLARR) Proarammer's Gude. NUREG/CR4639 V03 P1: NUCLEAR COMPUTERIZED LIBRARY FOR ERC INTERNATIONAL. lNC ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data NUREG/CR-5096: EVALUhTION OF SEALS FOR MECHANICAL PENE NU CR463 V03PY C CO PUT$ ZED LIBRARY FOR TRATIONS OF CONTAINMENT BUILDINGS. ASSESSING REACTOR RELIABILITY (NUCLARR). Guide To Data NUREG/CR-5147: FUNDAMENTAL ATTRIBUTES OF A PRACTICAL Processing And Revision.Part 2: Human Error Probatulity Data Entry CONFIGURATION MANAGEMENT PROGRAM FOR NUCLEAR PLANT Ard Revenson Procedures. DESIGN CONTROL NUREG/CR4639 V03 P3: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING FIEACTOR RELIABILITY (NUCLARR).Gude To Data FEDERAL EMERGENCY WANAGEMENT AGENC" Processing And Revissort Part 3: Hardware Component Failure Data NUREG-0654 S01 Rot: CRITERIA FOR PREPAMTiON AND EVALUA-Entry And Revision Procedures. TION OF RADIOLOGICAL EMERGENCY REbr)ONSE PLANS AND NUREG/CR-4639 V04 P1: NUCLEAR COMPUTERIZED LIBRARY FOR PREPAREDNESS IN SUPPORT OF NUCLEAR POWER ASSESSING REACTOR RELIABILITY (NUCLARR). User's Gude.Part 1: PLANTS Cntens For Utility Offsite Planning And Preparedness. Overview Of NUCLARR Data Retneval. NUREG/CR 4639 V04 P2: NUCLEAR COMPUTERIZED LIBRARY FOR FLORIDA. UNIV. OF, GAINESYlLLE, FL ASSESSING REACTOR RELIABILITY (NUCLARR). User's Gude.Part 2: NUREGICR-5063: DEVELOPMENT OF A MECHANISTIC UNDER-Gude To Operations. STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES-NUREG/CR4639 V04 P3: NUCLEAR COMPUTERIZED LIBRARY FOR SURE VESSEL STEELS.Fmal Report. ASSESSING REACTOR RELIABILITY (NUCLARR). User's Gusde. Par' 3: NUCLARR System Description. FRANKLIN INSTITUTE NUREG/CR 4639 VOS P1: NUCLEAR COMPUTERt2ED LIBRARY FOR NUREG/CR-5141: AGING AND OVALIFICATION RESEARCH ON SOLE-ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manuat.f' art 1: NOID OPERATED VALVES. Summary Desenpleon. NUREG/CR4639 VOS P2: NUCLEAR COMPUTERIZED LIBRARY FOR GEORGIA INSTITUTE OF TECHNOLOGY ATLANTA, GA ASSESSING RF4CTOR RELIABILITY (NUCLARR) Data Manual,Part 2: NUREG/CR 5258 V01: GEORGIA / ALABAMA REGIONAL SEISMO-N Rb/CR76 9 3N GRAPHIC NETWORK. Annual Report. August 1985 - June 1986. PUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual.Part 3: IDAHO NATIONAL ENGINEERING LABORATORY NU $4 V05 P4 CLE PUTERIZED LIBRARY FOR PR "E / 4740 NUCLEAR PLANT-AGING RESEARCH ON REACTOR ASSESSING REAC*OR RELIABILITY (NUCLARR). Data Manual.Part 4. NUREG/CR-5072: DECAY ' HEAT REMOVAL USING FEED-AND-BLEED NU GC ND BLEAR PLANT AGING RESEARCH ON REACTOR FOR U.S. PRESSURIZED WATER REACTORS. PROTECTION SYSTEMS NUREG/CR-5225: AN OVERVIEW OF BWR MARK-1 CONTAINMENT NUREG/CR-4747 V02: AN AGING FAILURE SURVEY OF LIGHT VENTING RISK IMPLICATIONS. WATER FIEACTOR SAFETY SYSTEMS AND COf.lPONENTS NUREG/CR-4898: RESULTS OF SEMISCALE MOD PC FEEDWATER WS% INC. AND STEAM LINE BREAK (S-FS) EXPERIMENT SERIES Bottom Main NUREG/CR-5107: HYDROGEOLOGIC CHARACTERIZATION OF Feedwater Line Break Acedent "imenments. BASALTSThe Northem Rim Of The Columtna Plateau Physiographic NUREG/CR-4971: RESULTS OF SEMISCALE MOD-2C FEEDWATER Provmce And Of The Creston Study Area. Eastern Washington AND STEAM LINE BREAK (S-FS) EXPERIMENT SERIES Ma n Steam NUhEG/CR-524& COMPARATIVE EVALUATION OF SELECTED CON-Line Break Accdent Experiments. TINUUM AND DISCRETE-FRACTURE MODELS Emphasis On Disper. NUREG/CR 5012: SIMILARITY PRINCIPLES FOR EQUIPMENT QUALi- swity Calculations For Apphcation To Fractured Geologic Me$a, Cres-FICATION BY EXPERIENCE. ton Study Area.Eastem Washington.

128 Contractor index NUREG/CR-5277. THE TENSORIAL NATURE OF EFFECTIVE POROSI- LOUISIANA STATE UNIV., BATON ROUGE, LA

      . TY AND LANGE-SCALE DISPERSION COEFFICIENTRApplication To           . NUREG/CR-4984: DEVELOPMENT OF A THREE DIMENSIONAL FLUX The Creston Study Area. Eastern Washington.                           SYNTHESIS PROGRAM AND COMPARISON WITH 3-D TRANSPORT JAPAN ATOMIC ENERGY RESEARCH INSTITUTE                                                                              .           .

NUREG/CR-4688 V02. QUANTIFICATION AND UNCERTAINTY ANALYa LOVELACE BIOMED & ENVIRONMENTAL RESEARCH INSTITUTE - SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT NUREG/CR 5067:EARLY AND CONTINUING EFFECTS OF COMBINED WATER REACTORS (OUASAR)Part IL Sensitivity Analysis Tech- ALPHA AND BETA 1RRADIATION OF THE LUNG. Phase il Report. MARYLAND, UNIV. OF, COLLEGE PARK, MD

 . JBF ASSOCIATES                                                     . NUREG/CR 4918 V02: CONTROL OF WATER INFILTRATION INTO NUREGICR-5021 V01; USER'S GUsDE FOR PRISIM ARKANSAS NU-                  NEAR SURFACE LLW DISPOSAL UNITS. Task Report . A Discuss 6on.

CLEAR ONE UNIT 1, Volume 1. Program For Inspectors NUREGICR-4996: A REPORT ON THE ROUND ROBIN PROGRAM NUREG/CR-5021 V02; USER'S GUIDE FOR PRISIM ARKANSAS NU- CONDUCTED TO EVALUATE THE PROPOSED ASTM STANDARD CLEAR ONE UNIT 1. Volume 2. Program For Regulators. TEST METHOD FOR DETERMINING THE PLANE STRAIN CRACK ARREST FRACTURE TOUGHNESS.K(IA),OF FERRITIC MATERIALS.. NUREG/CR 51A9. EROSION CORROSION OF PWR FEEDWATER ' MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE, MA ' PIPING SURVEY OF EXPERIENCE.DESf3N. WATER CHEMISTRY NUREG/CR-5080: A STUDY OF NEW ENGLAND SEISMICITY WITH AND MATEFliALS. EMPHASIS ON MASSACHUSETTS AND NEW . HAMPSHIRE. Final Report Covenng The Penod 1976-1985. QALSI ENGINEERING. lNC. NUREG/CR-5094: APPLICATION OF STOCHASTIC METHODS TO THE NUREG/CR 5159 PREDICTION OF CHECK VALVE PERFORMANCE SIMULATION OF LARGE-SCALE UNSATURATED FLOW AND AND DEGRADATION IN NUCLEAR POWER PLANT SYSTEMS. TRANSPORT. K ANSAS, UNIV. OF, LAWRENCE. KS MATERIALS ENGINEERING ASSOCIATES,INC. NUREG/CR-5045 KANSAS-NEBRASKA SEISMICITY STUDIES USING NUREG/CP-0064. SECOND CNSI WORKSHOP ON DUCTILE FRAC. THE KANSAS NEBRASKA MICROEARTHOUAKE NETWORK. Final TURE TEST METHODS. , Report NUREG/CR 4828: FATIGUE CRACK GROWTH OF PART THROUGH ' CRACKS IN PRESSURE VESSEL AND PIPING STEELS. Air Environ. LAWRENCE BERKELEY LABORATORY NUREG/CR 4864 V01: THERMODYNAMIC TABLES FOR NUCLEAR NUR 13: FATIGUE LIFE CHARACTERIZATION OF SMOOTH WASTE ISOLATION Acueous Solutions Database- AND NOTCHED PIPING STEEL SPECIMENS IN 288 DEGREES C AIR NURE / 52 TENS LE A R R CHARACTERIZATION OF NU EG 4775 G DE R P PAR NG OPERATING PROCE-NUREG/CR 5063: DEVELOPMENT OF A MECHANISTIC UNDER-  ; NUREG/CR-5042 SG1: EVALUATION OF EXTERNAL HAZARDS TO NU-STANDING OF RADIATION EMBRITTLEMENT IN REACTOR PRES- i CLE AR POWER PLANTS IN THE UNITED STATES. Seismic Hazard. SURE VESSEL STEELS. Final Peport. I NUREG/CR 5073 OUANTIFICATION OF MARGINS IN PlPING SYSTEM SFISMIC RESPONSE. Methodol ses And Dampin NUREG/CR.5136: FATIGUE STRENGTH OF SMOOTH AND NOTCHED NUREG/CR.5076. AN APPROACH O THE QUANTI ICATION OF SEIS- SPECIMENS OF ASME SA 106-B STEEL IN PWR ENVIRONMENTS. mig MARGINS IN NUCLEAR POWER PLANTSThe importance Of NUREG/CR-5201: EXPERIMENTAL ASSESSMENTS OF GUNDREM-BWR Plant Systems And Functions To Seismic Mar inn. MINGEN RPV ARCHIVE MATERIAL FOR FLUENCE RATE EFFECTS NUREG/CR 5113 FINDINGS OF THE PEER REVIE PANEL ON THE STUDIES-DRAf'T REACTOR RISK REFERENCE DOCUMENT.NUREG-1150 NUREG/CR-5207: FRACTURE EVALUATION OF SURFACE CRACKS NUREG/CR-5242 A FAST BOTTOM.UP ALGORITHM FOR COMPUT. EMBEDDED IN REACTOR VESSEL CLADDING. Material Property Eval-ING THE CUT SETS OF NONCOHERENT F AULT TREES uations. M ATHTECH. INC. j LEHIGH UNIV., BETHLEHEM #A  ! NUREG/CR 5095 VO1: THERMODYNAMIC NONEQUILIBRIUM IN POST. NUREG/CR-5138 VALIDATION OF GENERIC COST ESTIMATES FOR CRITICAL-HEAT-FLUX BOILING IN A ROD BUNDLE.Desenption Of CONSTRUCTION-RELATED ACTIVITIES AT NUCLEAR POWER , E rpenments And Sample Flesults PLANTS Final Report. l bUREG/CR.5095 V02. THERMODYNAMIC NONEOUILIBRIUM IN POST. NUREG/CR-5160: GUIDELINES FOR THE USE OF THE EEDB AT THE j CRITICAL HEAT-FLUX BOILING IN A ROD BUNDLE. Data For Stab,. SUB COMPONENT AND SUBSYSTEM LEVEL.Fnal Report. j hred Quench Front Tests MICHIGAN, UNIV. OF, ANN ARBOR, MI I NUREG/CR-5095 V03. THERMODYNAMIC NONEOUILIBRIUM IN POST, NUREG/CR 3145 V06- GEOPHYSICAL INVESTIGATIONS OF THE j CRITICAL-HE AT FLUX BOILING IN A ROD BUNDLE. Data For Advanc. sng Quench Front Tests WESTERN OHIO INDIANA REGION. Annual Report. October 1986 - NUREG/CR 5095 V04. THERMODYNAMIC NONEOUILIBRIUM IN POST. September 1987. j nch Front ss NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERLY NATIONAL BUREAU OF ) LOS ALAMOS NATIONAL LABORATORY NUREG/CR-4735 V03: EVALUATION AND COMPILATION OF DOE NUREG 13?0: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANALYSIS WASTE PACKAGE TEST DATA. Biannual Report February-July 1987. HANDBOOK NUREG/CR-4735 V04: EVALUATION AND COMPILATION OF DOE NUREG/CR-45G4 SEISMIC CATEGORY l STRUCTURES WASTE PACKAGE Tr.ST DATA. Biannual Report. August 1987- Janu-PROGR AM Final Report, Fiscal Year 1983 -1984 ary 1988. NUREG/CR 499D THE SEISMIC CATEGORY l STRUCTURES NUREG/CR-5166: ELECTROCHEMICAL EVALUATION OF SOLfD PROGRAM Results For Fmt at Year 1985 STATE PH SENSORS FOR NUCLEAR WASTE CONTAINMENT. NURE G/CR -5D44 ESTIMATION TECHNIOUES FOR COMMON CAUSE F AILURE EVENTS NUTECH ENGINEERS,INC. NURE G/CR 5071 TRAC SUPPORT SOFTWARE. NUREG/CR 4939 V01: IMPROVING MOTOR RELIABILITY IN NUCLEAR i NUREGICR 5090 EFFECTS OF TEMPERATURE AND HUMIDITY ON POWER PLANTS Volume 1. Performance Evaluation And Maintenance  ! RESPIRATOR FIT UNDER SIMULATED WORK CONDITIONS Practices I i NUREG/CR 5135 THE THERMAL HYDRAULICS OF SBLOCAS RELA- NUREG/CR 4939 V02: IMPROVING MOTOR RELIABILITY IN NUCLEAR TivL TO PRESSURllED THERMAL SHOCK. POWER PLANTS. Volume 2.Funchonal Indicator Tests On A Small NUHEGICR 5154 EXPERIMEN1 AL ASSESSMENT OF DAMPING IN Elecinc Motor Sublected To Accelerated Aging LOW ASPECT RATIO, REINFORCED CONCRETE SHEAR WALL NUREG/CR 4939 V03. IMPROVING MOTOR RELIABILITY IN NUCLEAR STRUCTURES POWER PLANTS. Volume 3. Failure Analysis And Diagnostic Tests On NURE G/CR 5 t 70 A REVIEW OF RESEARCH CONDUCTED BY LOS A Naturally Aged LarDe Electnc Motor. ALAMOS NATIONAL LABORATORY FOR THE NRC WITH EMPHASIS ON THE MAXEY FLATS KY. SHALLOW LAND WASTE BUR AL SITE OAK RIDGE NATIONAL LABORATORY NUREG/CR 5182 THE SEISMtC CATEGORY I STRUCTURES NUREG/CR-2000 V06N12; LICENSEE EVENT REPORT (LER) PROGRAM Results f or FY 1986 COMPILATION For Month Of December 1987.

Contractor index 129 NURE G/CR 2000 V07 N t. UCE NSEE EVENT REFORT (LER) NUREGICR 5033'

SUMMARY

DESCRIPTION OF THE SCALE MODU-COMPIL ATION F or Month Of Janustry 1988 LAR CODE SYSTE M NUni G/CR 2000 V07 N2. LIGENSEE EVENT REPORT (LE R) NURE G/CR4049: PRESSURE VESSEL FLUENCE ANALYSIS AS COMPIL ATION F or Month Of Fotwuary 1988 NEUTRON DOS! MET RY. NUREG/GR 2000 V07 N3 LIGLN$E E EVENT REPORT (LER) NUPEG/CR 5061; THRf'E FREQUENCY EDDY CURRENT INSTRU-COMPIL ATION For Month Of March 1988 MENT. NUREG'CR 2000 V07 N4 LICE NSEE EVENT REPORT (LER) NUREG/CR-5108- THERMODYNAMIC PROPERTIES OF TCOV)

                                                                                                                               ^"                " '

N EG/ 00 07 5 L N E EVENT REPORT (LE R) fC NUREG/CH-5157; THE DEVELOPMENT OF APRIL MOD 2 - A COMPUT. NU E / 07 6 tbE E EVENT REPORT (LE R) COMPIL ATION For Month Of June 196A ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL-NURLG/C42000 V07 N7. LICENSEE EM NT REPORT (LER) ING WATER NUCLEAR REACTORS 4 COMPILATION For Month Of Jul 1988 NUREG/CR-5183. A USEFFS MANUAL FOR THE CONTAMINANT l NURE G/CH-2000 V07 NU LIbENSEE EVENT REPORT (LER) TnANSPORT MODULE OF THE MIGRAT CODE. ) NUREG/CRS229 ANNUAL REPORT OF THE TMI-2 EPICOR Il RESlN/ COMPIL ATION NUREG'CR 2000 V07 For N9Month Of Aucbust Li ENSEE EVENT REPORT1968 (LER) LINER 14VESTIGATION Low-Level Waste Data Base Development COMhlATION for Month Of September 1988 Pro 9 tam For Fiscal Year 1988 NURE G/CR-2000 V07N t o: LICENSE E EVENT REPORT (LER) NUREG/CR 5264 GUIDE FOR LICENSING EVALUATIONS USING COMPIL ATION For Month Of October 1988 CRAC2 A Computer Program For Calculating Reactor Accutent Conse-NUREG/CR-2000 V07N11 LICENSEE EVENT REPORT (LER) Quences COMPIL ATION For Month Of Novembei 1988 , NUREG/CA 4219 V04 N2 HEAVY SECTION STEEL TECHNOLOGY PARAMETER,1NC. I PROGRAM Sammnnual Progress Repart For Apol Septemtier 1987 NUREG/CR 4523 CLOSEQUT OF IE BULLETIN 80-13 CRACKING IN NUREG/CR 4219 V05 N1; HEAVY-SECTION STELL TECHNOLOGY CORE SPRAY SPARGERS. PROGRAM Semiannual Progress Report For October 1987 March NUREG/CR 4662: CLOSEOUT OF IE BULLETIN 80-16 MAINTENANCE NL G/CR-4597 V02. AGING AND SERVICE WEAR Or AUXILIARY OF ADEOUATE MINIMUM FLOW THRU CENTRIFUGAL "HARGING FEEDWATER PUMPS FOR PWR NUCLE AR PLANTS Volume 2. Aging PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGV LINE RUP. T UU G/ 5 0 E OP L T R P DESIGN CRITERIA NUREldCR 4665. CLOSEOUT OF IE BULLETIN 83-08 ELECTRICAL BY RIPRAP TE STING IN FLUMES Phase il Followu inventi CIRCUIT BREAKERS WITH AN UNDERVOLTAGE TRIP FE ATURE IN NUREG/CR-4674 V05: PRECURSORS TO POTENTIAL SEVhationsAEUSE IN SAFETY RELATED APPLICATIONS OTHER THAN THE RE-CORE DAMAGE ACCIDENTS 1986 A STATUS REPORT ACTOR TRIP SYSTEM NUREG/CR-4674 V06 PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-4932: CLOSEOUT OF IE BULLETIN 80 03 LOSS OF CHAR-DAMAGE ACCIDENTS 19H6 A ST ATUS REPORT COAL FROM STANDARD TYPE II.TWO INCH.TR AY ADSORBER NUREC/CH 4708 V02 PROGRESS 4N EVALUATION OF RADIONU- CELLS CLIDC Gi OCHEMICAL INFORMATION DE.VELOPED BV DOE HIGH- NUREG/CR-4933 CLOSEOUT OF IE BULLETIN 8019 FAILURES OF LE VEL NUCLE AR W ASTE REPOSITORY SITE PROJECTS Report For MERCURY-WETTED MATRIX RELAYS IN REACTOR PROTECTIVE Apul t@6 - St ptemter 196' SYSTEMS OF OPERATING NUCLEAR POWER PLANTS DESIGNED NURf G/CR 4777 STE AM OXfDATION Of ZlRCALOY CLADDING IN BY COMBUSTION ENGINEERING. THE ORNL FISSION PRODUCT RE LEASE TESTS NUREG/CR 4935 CLOSLOUT OF IE BULLETIN 85 02.UNDEPVOL-NURI G/CR4778 PRELIM!NAR( STUDfES OF THE MORPHOLOGY OF TAGE TRIP ATTACHtAENTS OF WESTINGHOUSE DB SD TYPE RE-THERMAL GRADIE NT TUBE DEPOSITS FROM FISSION PRODUCT ACTOR TRIP BREAKERS. NL F 7 VI W AND EVALU."ON OF DESIGN ANALYSIS FR MB R N RO B DS METHODS FOR CALCUL ATING FLEXIBILITY OF NOZZLES AND NUREG/CR 5190 CLOSEOUT OF IE DULLETIN HO-14 DEGRADATION NL E 4f R F LAW DENSITY EXAMINATIONS OF A CLAD OF BWR SCRAM DISCHARGE VOLUME CAPADlLITY EsOILING WATER REACTOR PRESSURE VESSEL SEGMENT NUREG/CR 4680 VD1 CHARACTERIZAllON OF IRRADlATED CUR. OF 185 CONTROL RODS TO FULLY INSERT DURING A SCRAM AT RENT PRACTICE WELDS AND A533 GRADE B CLASS 1 PL ATE FOR ABWR. NUCLE AR PRESSURE VESSEL SERVICE NURE G/CR.4RBD V02 CHARACTERi2ATION OF IRRADIATED CUR. , RENT PRACTICE WELCS AND A533 GRADE B CLASS 1 PLATE FOR NUREG'!CR-4700 V01 - PR'OCEDURES FOR TREATING COMMON j NUCLE AR PRESSURE VESSEL SERVICE CAUSE FAILURES IN bAFETY /sND RELLAultlT Y ! NUREG/CR 4BHti PRESSURIZED THERMAL-SHOCK TE ST OF 6-INCH STUDIES Procedural Framework And Enamples THIGK PRESSURE VESSELS PTSE 2 investigation Of Low Tennng Re. tatant e And Warm Prestressing R&D ASSOCIATES NURE G/GR 4947 ANALYS:S OF THE A302D AND A5338 ST ANDARD NUREG/CR-3908 SURVEY OF THE STATE OF THE ART IN MITIGA-REFERENCE MATERIALS IN SURVEILLANCE CAPSULr:S OF COM TION SYSTEMS ME RCtAL POWER RE ACTORS NUREG/CR-4242 SUAVEY OF LIGHT WATER REACTOR CONTAIN-NURE G/CR.49M DEVE LOPMENT OF A THREE. DIME N$ TONAL FLUX MENT SYSTEMS, DCMINANT F AILURE MODES AND MlTIGATION S(NTHESIS PROGRAM AND COMPAR SON WITH 3-D TRANSPORT OPPORTUNITIES Fina! Report THf. OR f RE SUL T S NUREG/CR 4243 VALUE4MPACT ANALYSIS FOR EVALUAT!NG AL-NUREG/CR-4W2 VD1 AGING AND SEhVICE WE AR OF MULTISTAGE TERNAllVE MITIGATION SYSTEMS SWITCHLS USED IN SAFETY SYSTEMS OF NUCLE AR POWER NUREG/CR 424< STR nTEGIES FOR IMMEMENTING A MITIGATION Pt ANTS Operating E genence And F ailurt Irteniification POLICY FOR LOH1 AATER REACTORS NURE G /CR-4 %6 A REPORT ON THE HOUND ROBIN PROGRAM CONDUCTED TO E VALUATE THE PROPOSED ASTM ST/sNDARD RENSSELAER POLYTECHNIC INSTITUTE. TROY, NY Tl ST METHOD FOR DE TERMIN:NG THE PLANE STRAIN CRACK NUREG/CR-5157 THE DEVELOPMENT OF APRIL MOD 2 A COMPUT-ARRST F RACTURE TOUCHNi SS NIA)OF F E RRillC MATERIALS ER CODE FOR CORE MELTDOWN ACCIDENT ANALYSIS OF BOIL-NJP! G/CR %f 6 URANIUM OFIDE JRON OxlDE Mixt D ALROSOL ING WATER NUCLEAR REACTORS E) PE R!MI N T S IN STE AM AIR AT MOSPHERE S NSPP Tests fa t 1 M P 611 And t+31 Data Record Report SANDIA NATIONAL LABORATORIES NURE GICR W19 NE UTRON E FPOSURE PARAMETERS FOR THE NUREG/CP-0095 PROCEEDINGS OF THE FOURTH WORKSHOP ON Ml T AllURGICAL TE ST SPECIMENS IN THE FIF TH HE AVY SEC- CONT AINMENT INTEGRIT Y TION STIi L TECHNOL OGY IRR ADIATION SERIE S CAPSUL E 5 NUREG/CR-4508 BEHAvlOR OF A CORIUM JET IN HIGH PRESSURE NUHt G/CR W20 A

SUMMARY

Of E NVIRONME NT AL L Y ASSISTED MELT EJECTION FROM A REACTOR PREbSURE VESSEL CHAC0 GROWTH STUDil S PE RFORME D AT WESTINGHOUSE NUREG/CR.4527 V02 AN EXPERIMENT AL INVESTIGATION OF IN-E L E C T MC CORPORATION Under F unding F tom The Heavy Section TERNALLY IGNITED FIRES IN NUCLEAR POWER PLANT CONTROL WI Tm.nnalogy Program CABINETS Part it Room Effects Tests NURL G'CR 5021 V01 USE R S GUIDE FOR PR!SIM ARKANSAS NU- NUREGICR-4625 THE POSTlRRADIATION E XAMINATION OF THE DC CL 1. AR ONI UNIT 1 VMume t Praam For InNwetors MELT DVNAMICS EXPERIME NTS NURE G/CR W21 V02 USt R'S Guldf FOR PRISIM ARKANSAS NU- NUREG/CR-4728 E QUIPMENT QUALIFICATION RESE ARCH TEST OF r'. FAN ONI UNIT 1 Volume 2.Petyam f cr Regulatork A HIGH.RANCE RADIATION MONilOR

130 Contractor index NUREG/CR 4763 SAFE TY RELATED EOUtPMENT SURVIVAL IN HY. SCIENCE & ENGINEERING ASSOCIATES,INC. DROGE N BURNS IN L ARGE DRY PAR CONT AJLIENT BUILDINGS NUREG/CR-4555 RO1' GENERIC COST ESTNATES FOR THE DIS. NUREG/CR-4805 V02 RE ACTOR SAF ETY RESEARCH SEMIANNUAL POSAL OF RADIOACTIVE WASTES REPORT Juiv-December 1986 Volume 36 NUREG/CR 5138 VALIDATION OF GENERIC COST ESTIMATES FOR NURE G /CR-4807 SURFACE COMPLE XAHON MODELING OF RADIO- CONSTRUCTION-RELATED ACTivlTIES Al NUCL E AR POWER NUCLIDE ADSORPTION IN SUBSURF ACE ENVIRONMENTS PL ANTS Final Report NUREG/CR-4834 V02. RECOVL RV ACTIONS IN PHA FOR THE RISK NUREG/CRA160 GUIDELINES FOR THE USE OF THE EEDB AT THE METHODS INTEGRATION AND EVAL UATION PROGRAM SUB COMPONENT AND SUBSYSTEM LEVEL Final Report (RMIEP) Volume 2 Apphcation Of The Data $ased Method NURE G/CR-4836 APPROACHES TO UNCERTA!NTY ANALYSIS IN SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY PROBABILISTIC RISK ASSESSMENT SCIENCE APPLICATIONS, NUREG/CR 4864 V01, THERMODYNAMIC TABLES FOR NUCLEAR NUREG/CR-5078 VD1 A RELIABILITY PROGRAM EOR EMERGENCY WASTE ISOLATION Aqueous Solutions Database- DIESEL GENERATORS AT NUCLEAR POWER PLANTS Program NUREG/CR-4914 THE INFLUENCE OF SELECTED CONTAINMENT Structure STRUCIURES ON DEBRIS DISPERSAL AND TRANSPORT FOLLOW- NUREG/CR-5078 V02. A RELIABILITY PROGRAM FOR EMERGENCY ING HIGH PRESSURE MELT EJECTION FROM THE REAC' TOR DIESEL GENERATORS AT NUCLEAR POWER VESSEL. PLANTS Maintenance,$urvedlance And Condition Monitor:ng NUREG/GR.4916 HECTR ANALYSES OF THE NEVADA TEST SITE NUREG/CR-5106' USER'S GUIDE FOR THE T ACT5 COMPUTER CODE. (NTS) PREMWED COMBUSTION EXPERIMENTS NUREG/CR 5135. THE THERMAL HYDRAULICS OF SBLOCAS RELA-NUREG/CR 4917. DCH 2 RESULTS f ROM THE SECOND EXPER! MENT TlVE TO PRESSURIZED THERMAL SHOCK. PE RFORMED IN THE SURTSEY DIRECT HE.ATING TEST FACILITY. NUREG/CR-5150 STEAM GENERATOR OPE RAtlNG NUREG/CR-4993 A ST ANDARD PROBLE M FOR HECTR-MAAP EXPERIENCE. Update For 1984-1986 COMPARISON Incomplete Burning NUREG/CR-5151: PERFORMANCE BASED INSPECTIONS NUREG/CR 5029 MELT PROGRESSION IN SEVERELY DAMAGED RE- NURE G /CR-5248. PRIORITIZATION OF TIRGALELRE COYMENDED ACTOR CORES COMPONENTS FOR FUFtTHER AGING RESEARCH NUREG/CR 5032: MODELING 1:ME TO RECOVERY AND INITIATING EVENT FREQUENCY FOR LOSS OF OFFSiTE POWER INCIDENTS f9UTHWCST RESEARCH INSTITUTE AT NUCLEAR POWFR PLANTS INREG/CP-0075. PROCEEDINGS OF CSNI/NRC WORKSHOP ON NUREG/CR 5039 V01- RE ACTOR SAFETY RESEARCH SEMIANNUAL WCTILE PIPING FRACTURE MECHANICS. REPORT January June 1987 Volume 37. NURt.9/CR-50l? *QlLARITY PR!NCIPLES FOR EQUIPMFNT OUAll-NUREG/CR 5039 V02 RE ACTOR SAFETY RESEARCH SEMIANNUAL FICATION BY EXPERIENCE. REPORT. July December 1987 Reactor Safety Research Program NUREG/CR 5078 V01. A RELIABilliV PROGRAM FOR EMERGENCY ST. LOUIS UNIV., ST. LOUIS. MO DIESEL GENERATORS AT NUCLE AR POWER PLANTS Program NUREG/CR-5165. SEISMOLOGICAL INVESTIGATION OF E ART H-Structure OUAKES IN THE NEW MADR D SEISMIC ZONE AND THE NORTH-NUREG/CR-5078 V02. A RELIABILITY PROGRAM FOR EMERGENCY EASTERN EXTENT OF THE NEW MADRID SEISMIC ZONEfinal DlESEL GENERATORS AT NUCLEAR POWER Report, September 1981. December 1985. PL ANTS Maintenance Survedlance And Condition Monttonng NUREG/CR-5063 DESIGN. CONSTRUCTION AND INSTRUMENTATION STANFORD UNIV., STANFORD CA OF A 1/6-SCALE REINFORCED CONCRETE CONTAINMENT BUILD NUREG/CR-4807: SURFACDCOMPLEXATION MODELING OF RADIO-ING NUCLIDE ADSORPTION IN SUBSURF ACE ENVIRONMENTS. I NUREG/CR-5084 IFCI AN INTEGRATED CODE FOR CALCUL ATION OF ALL PHASES OF FUEL COOLANT INTERACTIONS TRIDENT ENGINEERS. INC. NUREG/CR 5086 PLATINUM CATALYTIC !GNITERS FOR LEAN HY. NUREG/CR 5078 V02 A RELIABILITY PROGRAM FOR EMERGENCY DROGEN-AIR M;XTURES. DIESEL GENERATOR 3 AT NUCLEAR POWER NUREG/CR 5096 EVALUATION OF SEALS FOR MECHANICAL PENE. PLANTS Maintenance. Surveillance And Condition Momtonng TRATIONS OF CONTAINMENT BUILDINGS NUREG/CR-5099 EVALUATION OF MATER lALS OF CONSTRUCTION VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSBURG, FOR THE REINFORCED CONCRETE RE AC'OR CONTAINMENT VA MODEL NUREG/CR-5123 STUDIES OF THE PATTEFIN AND AGES OF POST. NUREGICR $109 RELOCATION OF MET ALLIC CONSTITUENTS IN MET AMORPHIC FAULTS IN THE PIEDMONT OF VIRGINIA AND CORE DEBRIS BEDS. NORTH CAROLINA. NUREG/CR 5120 A MODEL FOR THE TRANSPORT AND CHEMICAL REACTION OF MOLTEN DEBRIS IN DIRECT CONTAINMENT HEAT- WElRICH & ASSOCIATES ING F kPERfMENTS NUREG/CR !'51 PERFORMANCE-BASED INSPECTIONS NURE G/CR 5126 TAC 2D STUDIE S OF MARK 1 CONTAINMENT DRYWELL SHELL MELT-THROUGH WESTINGHOUSE ELECTRIC CORP. NUREG/CR 5162 CHARM A MODEL FOR AEROSOL BEHAVIOR IN NUREG/CR'5020. A

SUMMARY

OF ENVIRONMENTAL.LY ASSISTED TIME VARYING THERMAL-HYDRAtiLIC CONDITIONS CRACK-GROWTH STUDIES PERFORMED AT WESTINGHOUSE NUREG/CR 5196 SUBM!SSION FOR THE CSNt/GREST BENCHMARK ELECTRIC CORPORATION Under Funding From The Heavy-Section EXERCISE ON CHEMICAL THERMODYNAMIC MODELING IN CORE. heel Technolcgy Program CONCRETE INTE RACTION FIELEASES OF RADIONUCLIDES NURE G /CR 4.,219 THE M;XING OF IMMISCIBLE LIQUlO LAYERS BY WESTINGHOUSE HANFORD CO. GAS BUBBLING NUREG/CA-4315 V09 R1 EVALUATION OF NUCLEAR F ACILITY DE-COMMISSIONING PROJECTS Summaey Status Report.Three Mele SC& A. INC. Island Unit 2 Radioactive Waste And Laundry Shipments NUREG/CA-4555 R01 GENERIC COST ESTIMATES FOR THE DIS- NURE G/CR-5023 HIGH-LEVEL SEISMIC RESPONSE AND F AILURE POSAL OF RADIOACTIVE A ASTE5 PREDICTION METHODS FOR PIPING.

international Organization Index This index lists, in alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed below each country and per-forming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number.

                                                                                                             \

There were no NUREG/lA reports for 1988. l l l i 131

  ~ ~ ~ - - _ _ _ _

l \ 3

Licensed Facility index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is needed, refer to the main citation by the NUREG number. 50 424 Alvin W Voglie Nuclear Plant Une' t Georpa NUHlG 1137 507 M 446 ComandK h'an 5:en f wterc Stahon Une 2 NUREG G/97 53 Power to Texas Ohles fic:tr 50-425 Aunn W Vogt e Nuclea Plant Uret 2. Gwwg.a NUREG t137 S07 50 371 LaSane County Staban Umt 1 Cowe*eanh N AEG Cl4 4r f5 PoweiCo Iorson C0 50 313 Arkaasas Naclear One Umt 1. Amansas Power NUAEG/CR 5021 V01 taSane County Staten Urvr 2 Commen*ea"h 50 374 NUM Go 43t.

                  & lght Co                                                                      I0$of; CG 50-313       Arkansas Nucipar Orie Um11. Arkansas Power Nunf G!CH S021 V02                  Oyster (mr4 Nades Power Ph"! Jewy brb219                                                         NUAEG 90 vt
                  & Lg*11 Co Centra' Poner & L$1 CO 50 313       Arkansas Nuciear One. Umt t Arka,sas Power NUREG'CR 505H           Sg ;82      PueJue Univ Research Pea' &                       NJRiG 12t 3
                  & lyhl Co                                                       50 312      Ra9the Sect Ndder Gervrata: Stahx.                NJHEG t2% Set 50 313       Arka%as bclear One. Umt 1 Arkansas Power NURL G/CH 5200                           Sat.ramento Marvupa! Uhnty
                  & LQ'11 Co                                                      % 3/7       Seadoyah Nacivar Pant Urv 1 Tennewe               NU6f 4 ??32 V02 STN 50456 Braowood Stahon Umt 1 Commonaeaitt' Esson NUREG 1002 SM                              vg,y y%

Co 50 3pg Spy, nob Nxh'ar Punt Ur nt 2 Tormeum STr. 51457 NJM G O2 V^? Brad cod Staten Un.t 2 Common *estn E$ son NJRE G 1002 Sa6 yag a7ts Co 3;N W 496

  $3445                                                                                      Sduth icus Po.ed Unit i Hasbn LWng & NJREGC W $?

Comanche Peak Swam Electnc Stabon. Umt 1. NUREG-0797 514 p,, Co Texas U!Ahes trecer s:N ya9s 50 445 Carnanche Fea6 Steam Iwcine Stahon. Urnt t. NU4!G4797 St$ soais re,as nn eci omi t tioowx,: 9ney a NUnEG tm p 50445 Co a h S am E~lertnc Stahon, Uml 1, NJREG 079/ Sib

                                                                                         ##                           "d" # ' "'"" #'l                 '

Texas Utrbhes Electr 54445 Coma,cne Peak Steam Electoc Stacan. Umi t. Tenas uttihes tiectr NURE G 07F $17 O #d" E'# h'* ' " I"?D U'dI *" " #3' A #U# # 5Su5 Comanc* ri ak Steam Elennc Stanoq umt t. i nan G 017 5tB Sm D e 5,atb 5 es Pv % "wr WuqE NUN C im Tenas Ubhhes Ew:tr ""* " k 5D 4J5 C3manche Pep $1 cam [lectg Stah00, Ury: 1. N$EG 0?!C S P) 50 E k"' h*"' MAD" ' % D 'I'f & T"K' ' Ienas Ut+htes Elecir %rk 50-M5 Comanche Pen Sicam Eectn: Stahan Umi i Nunf G 077 529 5t' 281 Sd"Y N M35

• bl ? W N f ^ l'W & b' N is i b 1 Tekas Chl$es [iectr IC'*'

50 446 Comanche f ce Steam Ente Statum. U9:t 2, NUAEG DNP Sid E no h w esaMN+4ar, b uw 2 Nuni -:, >-+ ' Temas Uhbhes [lectr GNd I N Utd hG 6

  % M6        Comd9Che Peh Steam EleclT StahDn Uret 2.       NUREG 0797 515      60 UO       '

F"' U" W"d b} if D ' " ' NNU *VI'"' Texas Ut6te. Elect Ge""'N % W t 50 u6 Cornanc* Pe',6 Sicam Electre State Umt 2. NURf G 079/ 54  % 2M Th'ee Mac A'a"'ta' No'e Mavn Uu , OHi O Tenas utAt.es Electr Gewa. % Ur^nes 5SM6 Co'nanr*ie Pr A Steam [lectnc State UM 2 NJP! G OT 517 4 't Ve'm:nt av m r M kr %m SaW Vmmom N 'M C/ ; r, p Tevas Uthbes [icctr iawe % tim Poem 5106 Coma @c Fe9 bream liectre brabon Umt ? NJRE5 0747 518 '/ ;9c v.a':s ba. NAr Fst fy i ? . , va N3 ; (is f f Teias Uhbles Ee:Ir V A v Aathar!!v 50 44f f,Drra'Enf Pcin Steam [leN4; $iahn Uqrt 2 NUR[G.C77 b $ 4. O k Not i gnW N vN.s + W NJM  ? FF 7enas Urikties Electr 45NCt '.seit:ah la"JM h eN 4 t'. N Hi l1J? ' 133 l l

NRC FOsw 335 U S. NUCLE AR HEGUL ATORY COMMISSION ) 5' f 6' r .. -..i'

  *2 g9                                                                                                                                       ( Asstyned tsy NRC And Vol , $wp; Rev gre ;y t gc?,

said Adr>6 dom Numtete, d any r neum BIBLIOGRAPHIC DATA SHEET

                                                <3,, ,ns,,va a,,,s o,, g,e ,,,,,so NUREG-0304 t m LE AND SUBMLi                                                                                                                          Vol. 13, No. 4 Regulatory and Technical Reports (Abstract Index Journal)                                                                          ,            g ,, , n y , ,,            y Annual Compilation for 1988                                                                                                                   " "

l May 1989 4 f IN OH CHANT NUYBE H

 *s AUT HOR (Sp                                                                                                                         f; T Y PE OF REPORT Reference i l't Hs0D Covt HL D         o...   .D, 1989 8 PL RF ORMING ORG ANfZAT ION - NAME AND AVDH L 5s ter Nnc. proer+ 0,vwn. on or k~               n goon. U 5 huctose N earsrory Conu rouaw wo mastow aaarra orwv a u r a io. . ..o name end motiong adoresU Division of Freedom of Information and Publications Services Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 9 SPONSORING ORGANIZATION - NAVE AND ADD'< LS5 or NAC ,vro. 's e u as stw orcomesu sw provute NRC Dwwo** Otru r or Rnnen, V S Nw le.n kwta,,as Co                                 e and marrono addren I Same as 8, above.

10 SUPPLEMENT ARY NOTE S 11 ADST R ACT upo wora> or eue This journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceedings of conferences and workshops, as well as international agreement reports. The entries in this compilation are indexed for access by title and abstract, secondary reoort number, personal author, subject, MRC organization for staff and international agreements, contractor, international organization, and licensed facility. u n t y won Ds Di sc s c os s ro., -ro. ... ,,,,,. ... , ~.. .,,,,,,,, ,.m.,. . , ...,-.,,~,,~,.4.~ . . . , , Unlinited ' compilation mvo~ ,- .n abstract index o ,a. . Unclassified Unclassified is Novem o, , uu s it. H il t NRC 6 Ont/ 3J'y t/ 69

A NUCLEAR K : 70AY COMMISSION WASHINGTON, D.C. 20666 m=E'0"JEldie. .= Main Citations and Abstracts 120555139531 1 1 ANI AC19L190 US NRC-0 ADM DIV FOIA & PUBLICATIONS SVCS TPS PDR-NUREG Secondary Report P-209 Number index WASHINGTON DC 20555 Personal Author index Subject index NRC Originating Organization index (Staff Reports) NRC Originating Organization Index (International Agreements) l NRC Contractor  ! SponsorIndex  ; i i Contractor index i i l International Organization Index Licensed Facility Index . . . . _ .}}