ML20237K506

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Second Quarter 1987,April-June

ML20237K506
Person / Time
Issue date:	08/31/1987
From:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
To:
References
NUREG-0304, NUREG-0304-V12-N02, NUREG-304, NUREG-304-V12-N2, NUDOCS 8709040390
Download: ML20237K506 (60)

Text

_--

L NUREG-0304 Vol.12, No. 2 Regulatory and Technical Reports (Abstract Index Journal}

Compilation for Second Quarter 1987 April - June U.S. Nuclear Regulatory Commission Office of Administration and Resources Management i

p s" "'%

(bk((')

I i

i fh90 n,0933 0304 R pan

k

)

i l

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 i

l l

l 1

l l

l j

i l

NUREG-0304 Vol.12, No. 2 Regulatory and Technical Reports

' 4 Abstract Index Journal?

)

i j

l

\\

I i

Compilation for Second Quarter 1987 April - June Date Published: August 1987 i

l Policy and Publications Management Branch Division of Publications Services l

Office of Administration and Resources Management U.S. Nuclear Regulatory Commission Weshington, DC 20555 fa asaug

/

..__-____m--__

1 I

l l

l CONTENTS Preface....

... v i

Index j

Tab 1

1 j

Main Citations and Abstracts.

• Staff Reports j

• Conference Proceedings

• Contractor Reports

• International Agreernent Reports Secondary Report Number index.

... 2 Personal Author index 3

Subject Index

.. 4 NRC Ori<pr.ating Organization index (Staff Reports).

5 NRC Originating Organization index (International Agreements).

....... 6 i

NRC Contract Sponsor index (Contractor Reports).

7

)

Contractor index...

... 8 1

International Organization index.

.9 Licensed Facility lndex.

... 10 l

4 l

l i

III

1 l

PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to:

)

Division of Publications Services Policy and Publications Management Branch Publishing and Translations Section l

Woodmont 537 l

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 l

NRC Originating Organization Index (Staff Reports)

)

NRC Originating Organization Index (International Agreements) 4 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 i

NUREG-0808: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

AND'ERSON, 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 RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070.

)

l Where the entries are (1) repott number, (2) report title, (3) report author, (4) organization that compiled I

the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use).

Contractor Report NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L. BENNETT, P.R.

Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242.

Where the entries are (1) report nur'nber, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use).

)

v i

1

____-_-.___a

International 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 staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the National Technical information Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the following address:

Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 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/lA is used for international 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.

vi

I Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is an NRC staff-onginated report, NUREG/CP XXXX is an NRC-sponsored conference report, NUREG/CR-XXXX is an NRC contractor-prepared report, and NUREG/lA-XXXX is an inter-national agreement re aort, The bibliographic information (see Preface for details) is followed by a brief abstract of t1is report, NUREG-0020 V10 N11: LICENSED OPERATING REACTORS There was one abnormal occurrence reported by an Agreement STATUS

SUMMARY

REPORT. Data As Of October State; it involved a therapeutic medical misadministration. The 31,1986.(Gray Book 1) ROSS,P.A.; BEEDE M.R. Division of report also contains information updating some previously re.

Computer & Telecommunications Services (Post 870413). April ported abnormal occurrences.

1987. 471pp. 8705120010. 40901:348.

The OPERATING UNITS STATUS REPORT - LICENSED OP-NUREG-0304 V12 N01: REGULATORY AND TECHNICAL RE-ERATING REACTORS provides data on the operation of nucle-PORTS (ABSTRACT INDEX JOURNAL). Compilation For First ar units as timdy and accurately as possible. This information is Quarter 1987, January March.

• Division of Publication Services -

coilected by the Office of Resource Management from the (Post 870413). May 1987. 62pp. 8706030076. 41166:112.

Headquarters staff of NRC's Office of Inspection and Enforce-This journal includes all formal reports in the NUREG series ment, from NRC's Regional Offices, and from utihties. The three prepared by the NRC staff and contractors; proceedings of con-sections of the report are: monthly highhghts and statistics for ferences and workshops; as well as international agreement re-ports. The entries in this compilation are indexed for access by da c pi t of de aled n orm t n on ea h unit ov ed by NRC's Regional Offices, IE Headquarters and the utilities; title and abstract, secondary report number, personal author, and an appendix for miscellaneous information such as spent subject, NRC organization for staff and intemational agree-i fuel storage capability, reactor. years of experience and non.

ments, contractor, international organization, and licensed facih-I power reactors in the U.S. It is hoped the report is helpful to all ty.

l agencies and individuals interested in maintaining an awareness of the U.S. energy situation as a whole.

NUREG 0332: POTENTIAL HEALTH AND ENVIRONMENTAL IM-PACTS ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL

• j NUREG-0020 V10 N12: LICENSED OPERATING REACTORS CYCLES. Final Report. GOTCHY,R.L. Office of Nuclear Reactor STATUS

SUMMARY

REPORT. Data As Of November Regulation. Director (Post 870411). June 1987, 73pp.

30,1986.(Gray Book I) ROSS.P.A. Division of Computer & Tele.

8707070538. 41600:001 Estimates of mortality and morbidity are presented based on 8 7 42 4 58 5

See NUREG-0020,V10,N11 abstract.

present-day knowledge of health effects resulting from current i

I component designs and operations of the Wear and coal fuel NUREG-0040 V11 N01: LICENSEE CONTRACTOR AND cycles, and anticipated emission rates ard occupational expo-VENDOR INSPECTION STATUS REPORT. Quarterly sure for the various fuel cycle facilities ex9ected to go into oper-Report, January. March 1987.(White Book)

• Division of Reactor ation during the next decade. The author concluded that, al-m wgh e are large unmahties b he estimams of potenbal 87061 0 06 4 11 03 health effects, the coal fuel cycle nternative has a greater i

This penodical covers the results of inspections performed by health impact on man than the uranir n fuel cycle. However, the the NRC's Vendor Inspection Branch that have been distributed to the inspected organizations dunng the period from January increased nsk of health effocts for u.her fuel cycle represents a 1987 throucfi March 1987. Als3, included in this issue are the very small incremental risk to 'he average individual in the results of certain inspections performed prior to January 1987 public for the balance of this century. The potential for large im-that were not included in previous issues of NUREG 0040.

pacts exists in both fuel cycles, but the potential impacts asso-NUREG-0090 V09 NO3: REPORT TO CONGRESS ON ABNOR.

fossil fuels, such as coal, cnnot yet be reasonably quantified.

MAL OCCURRENCES. July. September 1986.

• Office for Analy-Some f po n at esonmeMal spacts of me coal fuel sis & Evaluation of Operational Data, Director. April 1987.60pp.

Cycle cannot currently be realistically estimated, but those that 870529030f 41115:296.

Section 208 of the Energy Reorganization Act of 1974 identi-can appear greater than those from the nuclear fuel cycle.

fies an abnormal occurrence as an unscheduled incident or NUREG-0386 D04 R05: UNITED STATES NUCLEAR REGULA.

event which the Nuclear Regufatory Commission determines to TORY COMMISSION STAFF PRACTICE AND PROCEDURE be significant from the standpoint of pubhc health and safety DIGEST. July 1972 - September 1986.

• Office of the General and requires a quarterly report of such events to be made to Counsel. June 1987, 300pp. 8707130184. 41682:228.

Congress. This report covers the penod July 1 to September This Rovision 5 of the fourth edition of the NRC Staff Practice 30, 1986. During the report period, there were four abnormal occurrences at the nuclear power plants licensed to operate.

and Procedure Digest contains a digest of a number of Com-The events were (1) a dfferent'al pressure switch problem in mission, Atomic Safety and Ucensing Appeal Board, and Atomic safety systems at LaSalle facaty, (2) abnormal cooldown and Safety and Licensing Board decisions issued during the period depressunzation transient at Catawba Unit 2, (3) significant July 1,1972 to September 30,1986, interpreting the NRC Rules safeguards deficiencies at Wolf Creek and Fort St. Vrain, and of Practice in 10 CFR Part 2. This Revision 5 replaces in part (4) significant deficiencies in access controls at River Bend Sta-eartier editions and supplements and includes appropriate tion. There was one abnormal occurrence at the other NRC h-changes reflecting the amendments to the Rules of Practice ef-censees; it involved a therapeutic medical misadministration.

fective September 30,1986.

1

2 Main Citations and Abstr:cto NUREG-0540 V09 NO2: TITLE LIST OF DOCUMENTS MADE and Licensing Board Panel, the Administrative Law Judge, the PUBLICLY AVAILABLE. February 1-28,1987.* Duision of Publi-Director's Decisions, and the Denials of Petitions for Rulemak-cation Services (Post 870413). April 1987. 400pp. 8706240149.

ing are presented.

41440:175.

This document is a monthly pubhcation containing desenp-NUREG-0750 V24 N04: NUCLEAR REGULATORY COMMISSION tions of information recewed and generated by the U.S. NRC.

ISSUANCES FOR OCTOBER 1986.Pages 489-679.

• Division This information includes (1) docketed material associated with of Publication Services (Post 870413). April 1987. 202pp.

civihan nuclear power plants and other uses of radioactwo ma-8 05120123 408 2 257.

tenals, and (2) nondocketed matenal received and generated by Legalissuances of the Commission, the Atom,c Safety and Li-i NRC pertinent to its role as a regulatory agency. The following censing Appeal Panel, the Atom:c Safety and Licensing Board indexes are included: Personal Author Index, Corporate Source Panel, the Administrative Law Judge, and NRC Program Offices Index, Report Number index and Cross Reference to Principal are presented.

Documents Index.

NUREG-0750 V24 N05: NUCLEAR REGULATORY COMMISSION NUREG-0540 V09 NO3: TITLE LIST OF DOCUMENTS MADE ISSUANCES FOR NOVEMBER 1986.Pages 681768.

• Division PUBLICLY AVAILABLE. March 1-31,1987.

• Division of Pubhca, of Pubhcation Services (Post 870413). May 1987. 99pp.

tion Services (Post 870413). May 1987. 469pp 8706160119.

8706040313. 41180:238.

41315:256.

See NUREG-0750,V24,N04 abstract.

See NUREG 0540,V09,N02 abstract.

NUREG-0750 V24 N06: NUCLEAR REGULATORY COMMISSION NUREG-0540 V09 N04: TITLE LIST OF DOCUMENTS MADE ISSUANCES FOR DECEMBER 1986.Pages 769 930.

• Division PUBLICLY AVAfLABLE. April 1-30,1987.

• Dwision of Publica.

of Publication Services (Post 870413). June 1987. 175pp.

tion Services (Post 870413). June 1987. 450pp. 8706230495.

8707090378. 41633:182.

41434:117.

See NUREG-0750,V24,N04 abstract.

See NUREG-0540,V09,N02 abstract.

NUREG-0781 S03: SAFETY EVALUATION REPORT RELATED NUREG-0683 S02: PROGRAMMATIC ENVIRONMENTAL IMPACT TO THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 STATEMENT RELATED TO DECONTAM! NATION AND DIS-AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And POSAL OF RADIOACTIVE WASTES RESULTING FROM Power Company)

• Division of Reactor Projects - til.IV,V & Spe.

MARCH 28,1979 ACCIDENT AT THREE MILE ISLAND NUCLE.

cial Projects (Post 870411). May 1987.172pp. 8706240260.

AR STATION. UNIT 2. Final Supplement Deahng With Disposal 41441:226.

Of....

• TMI-2 Cleanup Project Directorate. June 1987. 300pp.

The Safety Evaluation Report issued in April 1986 provided 8707080318. 41616:213.

the results of the NRC staff's review of the Houston Lighting in accordance with the National Environmental Pohey Act, the and Power Company's apphcation for heenses to operats the Programmatic Environmental impact Statement Related to De.

South Texas Project. The facihty consists of two pressun2ed contamination and Disposal of Radioactwe Waste for the 1979 water nuclear reactors located in Matagorda County, Texas.

Accident at Three Mile Island Nuclear Station, Unit 2 (PElS) has Supplement No.1 issued in September 1986 updated the infor-been supplemented. This supplement updates the environmen-mation Contained in the Safety Evaluation Report and ad-tal evaluation of accident. generated water disposal alternatives dressed the ACRS Report issued on June 10, 1986. Supple-pubbshed in the PElS, utikzing more complete and current infor.

ment No. 2 issued in January 1987 addressed and resolved mation, and covering the hcensee's proposal to dispose of the some of the outstanding issues remaining after issuance of the water by evaporation to the atmosphere. The staff concludes Safety Evaluation Report and Supplement No.1. This Supple-that the water can be disposed of without incurring significant ment No. 3 also addresses and resolves some of the outstand-environmental impact. The staff's evaluation of a number of dis.

irig issues remaining after issuance of the SER and Supple-posal alternatwes indicates that no alternative is clearly prefera.

ments 1 and 2.

ble to the others, and that the hconsee's proposal method is satisfactory.1he nsks to the general public from exposure to ra-NUREG-0800 06.5.2 R2: STANDARD REVIEW PLAN FOR THE dioaclwe effluents from any alternatwe have been quantitatively REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR estimated and are very small fractions of the estimated normal POWER PLANTS. LWR Edition. Proposed Revision 2 To Section sne:Jonce of cancer fatalities and genetic disorders. The most 6.5.2, " Containment Spray As A Fission Product Cleanup significant potential impact associated with any disposal alterna.

System." For Comment.

• Office of Nuclear Reactor Regulation, Director (Post 870411). April 1987. 80pp 8704270036.

tNe is the nsk of physical injury associated with transportation 40682:016 accidents. Additionally, no significant impacts to aquatic or ter.

restnal biotic from any disposal attema!Ne are expected.

Proposed revision 2 to SRP Section 6.5.2 would incorporate changes in the requirements for containment spray chemical NUREG-0728 R02: NRC INCIDENT RESPONSE PLAN.

• Office additrve systems, and explicitly states computational models for Analysis & Evaluation of Operational Data, Director. June which had only appeared in references in previous revisions.

1987. 29pp 8706240202. 41448:219 The requirement for immediate initiation of caustic addition to The Nuclear Regulatory Commission (NRC) regulates civilian the spray would be deleted, and the minimum pH to be nuclear actwitnes to protect the public health and safety and to achieved would be reduced from 8.5 to 7. If adopted, this revi-preserve environmental quakty. An incident Response Plan had sion would be required to be used for future plants, and would been developed and has now been revised for the second time be optiortal for present heensees. The proposed revision is ac-to reflect current Commission pokcy, NUREG-0728. Rev 2 as-companied by a regulatory analysis and two supporting techiti-signs responsibilities for responding to any potentially threaten-cal documents.

ing incident involving NRC licensed activities and for assunng that tre NRC will fulfill its statutory mission. Revision 2 was nec-NUREG-0800 06.5.5 RO: STANDARD REVIEW PLAN FOR THE essa y to reflect organizational changes' REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Proposed Revision 0 To New NUREG-0750 V24101: INDEXES TO NUCLEAR REGULATORY SRP Section 6.5.5, " Pressure Suppression Pools As Fission COMMISSION ISSUANCES. July-September 1986.

• Dwision of Product Clean-Up Systems." For Comment.

• Office of Nuclear Pubhcation Services (Post 870413). June 1987. 61pp.

Reactor Regulation, Director (Post 870411). April 1987. 51pp.

8707020110.41569 166.

8704280166. 40716:350.

Digests and indexes for issuances of the Commission, the Proposed new SRP Section 6.5.5 would provide acceptance Atomic Safety and Licensing Appeal Panel, the Atomic Safety entena and review procedures to be used in assessing the role

1 l

l Main Citttisns cnd Abstracts 3

l of pressure suppression pools as fission product cleanup sys-of every Congress. Contents of NUREG 0980 include: The tems following potential reactor accidents. A calculational model Atomic Energy Act of 1954. as amended; Energy Reorganiza-to account for drywell bypass is given, and minimum fission tion Act of 1974, as amended; Uranium Mill Taihngs Radiaton product decontamination factors are listed for use in instances Control Act of 1978; Low-Level Radioactive Waste Policy Act; l

in which no detailed calculations of pool scrubbing have been Nuclear Waste Policy Act of 1982; and NRC Authorization and perforrned. The proposed section is accompanied by a regula-Appropriations Acts. Other materials included are statutes and tory analysis and a supporting technical report.

treaties on export licensing, nuclear non-proliferation, and envi-ronmental protection.

NUREG-0837 V06 N04: NRC TLD DIRECT RADIATION MONI-TORING NETWORK. Progress Report, October December 1986.

NUREG 1002 S03: SAFETY EVALUATION REPORT RELATED l

JANG,J.: COHEN.L Region 1, Office of Director. April 1987.

TO THE OPERATION OF BRAIDWOOD STATION, UNITS 1 l

324pp. 8704280116. 40712.118.

AND 2. Docket Nos. 50-456 And 50-457.(Commonwealth Edison This report provides the status and results of the NRC Ther-Company)

• Division of Reactor Projects - litIV,V & Special l

moluminescent Dosimeter (TLD) Direct Radiation Monitoring Projects (Post 870411). May 1987. 35pp. 8706150208.

Network. It presents the radiation levels measured in the vicinity 41300 212.

of NRC licensed facility sites throughout the country for the in November 1983, the staff of the Nuclear Regulatory Com-l fourth quarter of 1986.

mission issued its Safety Evaluation Report (NUREG-1002) re.

NUREG-0904 S01: DRAFT SUPPLEMENT TO THE FINAL ENVI-garding the application filed by the Commonwealth Edison Com-RONMENTAL STATEMENT RELATED TO THE DECOMMIS-pany, as applicant and owner, for a license to operate Braid-SiONING OF THE RARE EARTHS FACILITY, WEST wood Station, Units 1 and 2 (Docket No. 50-456 and 50-457).

CHICAGO.lLLINOIS. Docket No. 40 2061.(Kerr McGee)

• Divi.

The first supplement to NUREG-1002 was issued in September sion of Fuel Cycle, Medical, Academic & Commercial Use 1986; the second supplement to NUREG 1002 was issued in Safety (Post 870413). June 1987. 475po. 8707060327.

October 1986. This third supplement to NUREG 1002 reports the status of certain items that remained unresolved at the time 41584:239.

This Draft Supplement to the Final Environmental Statement Supplement 2 was published. The facility is located in Reed is issued by the U.S. Nuclear Regulatory Commission in to-Township, Will County, Illinois.

sponse to the Atomic Safety and Licensing Board's ruling that NUREG 1021 R04: OPERATOR LICENSING EXAMINER STAND-the staff must supplement the Final Environmental Statement in ARDS.

• Operator Licensing Branch. May 1987, 200pp.

order to evaluate the impact of permanent disposal of the Kerr.

8706030301.41165:001.

McCae Rare Earths Facility wastes located at West Chicago,11, The Operator Licensing Examiners Standards provide policy linois. The statement considers the Kerr McGee preferred plan and guidance to NRC examiners and establishes the proce-and various alternatives to the plan. The action proposed by the dures and practices for examining and licensing of applicants Commission is the renewal of the Kerr-McGee license to allow for NRC operator licenses pursuant to Part 55 of Title 10 of the disposal of wastes onsite and for possession of the wastes Code of Federal Regulations (10 CFR 55). It is intended to under license for an indeterminate time. The license could be assist NRC examiners and facility licensees to understand the terminated at a later date if certain specified requirements were examination process better and to provide for equitable and met.

consistent administration of examinations to all apphcants by NUREG-0936 V05 N04: NRC REGULATORY AGENDA. Quarterly NRC examiners. This standard is not a substitute for the Opera-Report. October. December 1986.

• Division of Rules & Records tor Licensing Regulations. As appropnate, this standard will be (Post 870413). May 1987.160pp. 8706150112. 41300:247.

periodically revised to accommodate comments and reflect new The NRC Regulatory Agenda is a compilation of all rules on information or expenence, which the NRC has proposed or is considenng action and all petitions for rulemaking which have been received by the Com-NUREG-1057 S05: SAFETY EVALUATION REPORT RELATED mission and are pending disposition by the Commission. The TO THE OPERATION OF BEAVER VALLEY POWER l

Regulatory Agenda is updated and issued each quarter.

STATION, UNIT

2. Docket No.

50-412.(Duquesne Light

• pa a

NUREG-0936 V06 N01: NRC R,EGULATORY AGENDA.Ouarterty 43 987 7 p 87 6 2 9 411 104.

Report, January-March 1987.

Division of Rules & Records Supplement No. 5 to the Safety Evaluation Report for the ap-S URE O

6. 05 N abba plicaton filed by Duquesne Light Company, et al., for kcense to PP operate the Beaver Valley Power Station, Unit 2 (Docket No.

NUREG-0940 V06 N01: ENFORCEMENT ACTIONS:SIGNIFICANT 50-412), located in Beaver County, Pennsylvania, has been pre-ACTIONS RESOLVED.Ouarte ly Progress Report, January. March pared by the Office of Nuclear Reactor Regulation of the Nucle-1987.

• Office of Enforcement (Post 870413). June 1987, ar Regulatory Commission. The purpose of this supplement is to 400pp. 8706240320. 41450:266.

update the Safety Evaluation of (1) additional information sut>

This compilation summanzes signifcant enforcement actions mitted by the applicants since Supplement No. 4 was issued.

that have been resolved dunng one quarteriy penod (January.

and (2) matters that the staff had under review when Supple-March 1987) and includes copies of letters, Notees, and Orders rnent No. 4 was issued.

sent by the Nuclear Regulatory Commission to Icensees with NUREG 1100 V03 ADD: BUDGET ESTIMATES. Fiscal Years respect to these enforcement actions. It is antcipa'ed that the 1988-1989.* Division of Budget & Analysis (Post 870413). May information in this publication will be widely disseminated to 1987. 44pp. 8706300181. 41491:060.

managers and employees engaged in activities licensed by the This report contains the fiscal year budget justifications to NRC, so that actions can be taken to improve safety by avoid-Congress. The budget provides estimates for salanes and ex-ing future violations similar to those described in this pubica.

penses for fiscal years 1988 1989. This addendum is required tion.

due to the NRC reorganization of April 12,1987.

NUREG-0980 R03: NUCLEAR REGULATORY LEGISLATION.

HOSPODOR,S. Offee of the General Counsel. Apnl 1987.

NUREG.1122 901: KNOWLEDGES AND ABILITIES CATALOG FOR i

115pp. 8705290061. 41116.266.

NUCLEAR POWER PLANT OPERATORS. Pressurized Water NUREG-09B0 is a compilation of nuclear regulatory legislation Reactors.

• Offee of Nuclear Reactor Regulation, Director (Post and other relevant material through sne 99th Congress, 2nd 870411). April 1987. 240pp. 8705290304. 41116:001.

Session. This compilation has been prepared for use as a re-This document catalogs roughly 5300 knowledges and abili-source document, whch the twRC intends to update at the end ties of reactor operators and senior reactor operators. It resuits I

1

4 Miln Citati:ns end Abstr: cts from a reanalys:a of a much larger job-task analysis data base NUREG 1184 DRFT: INTEGRATED SAFETY ASSESSMENT compiled by the institute of Nuclear Power Operations (INPO).

REPORT,1NTEGRATED SAFETY ASSESSMENT PROGRAM -

Knowledges and abilities are cataloged for 45 major power MILLSTONE NUCLEAR POWER STATION, UNIT 1. Docket No.

plant systems and 38 emergency evolutions, grouped according 50-245.(Northeast Nuclear Energy Co). Draft Report.

• Office of to 11 fundamental safety functions (e.g., reactivity control and Nuclear Reactor Regulation, Director (Post 870411). April 1987.

reactor coolant system inventory control). Supplemental pages

$24pp. 8704270155. 40632:334.

have been added to conform to NUREG-1123, "Knowledges The Integrated Safety Assessment Program (ISAP) was initiat-and Abilities Catalog for Nuclear Power Plant Operators: Boiling ed in November 1984, by the U.S. Nuclear Regulatory Commis.

Water Reactors," September,1986. A structured sampling pro-sion to conduct integrated assessments for operating nuclear cedure for both catalogs is under development by the Nuclear power reactors. The integrated assessment is conducted on a Regulatory Commission (NRC) and will be published as a com-plant-specific basis to evaluate all licensing actions, licensee ini-panton document, " Examiners' Handbook for Developing Oper-tiated plant improvements and selected unresolved generic /

ator Licensing Examinations" (NUREG 1121). With appropriate safety issues to establish implementation schedules for each sampling from these catalogs, operator hcensing examinations item. In addition, procedures will be established to allow for a having content validtty can be developed. The examinations de-periodic updating of the schedules to account for licensing i

veloped by using the catalogs and handbook will cover those issues that arise in the future. This report documents the review topics listed under Title 10, " Code of Federal Regulations," Part of Millstone Nuclear Power Station, Unit No.1, operated by 1

55.

Northeast Nuclear Energy Company, which is one of two plants NUREG 1125 V08: A COMPILATION OF REPORTS OF THE AD-being reviewed under the pilot program for ISAP. This report in-VISORY COMMITTEE ON REACTOR SAFEGUARDS 1986.

• dicates how 85 topics selected for review were addressed and ACRS - Advisory Committee on Reactor Safeguarda. April 1987' presents the staks recommendations regarding the corrective 217pp. 8705200003. 40987:013.

actions to resolve the 85 topics and other actions to enhance This compilation contains 58 ACRS reports submitted to the plant safety. The report is being issued in draft form to obtain Commission or to the Executive Director for Operations during comments from the licensee, nuclear safety experta, and the the calendar year 1986. All reports have been made available to Advisory Committee for Reactor Safeguards. Once those com-the public through the NRC Public Document Room and the ments have been resolved, the staff will present its positions, U.S. Library of Congress. The reports are divided into two along with a long-term implementation schedule from the licens.

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

ee, in the final version of this report.

ACAS Reports on Genenc Subjects. Part 1 contains ACRS re.

NUREG 1214 R01: HISTORICAL DATA

SUMMARY

OF THE SYS-ports alphabetized by project name and within project name by TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE.

chronological order. Part 2 categudzes the reports by the most appropriate genonc subect area and within subject area by JOHNSON.M R.; WATERMAN,D.K. Division of Inspection Pro-l chronological order.

grams (850212-87c411) April 4, 1987. 69pp. 6704270378.

40704.218.

NUREG-1145 V03: U S. NUCLEAR REGULATORY COMMISSION This report presents a history of Systematic Assessment of 1986 ANNUAL REPORT.

• Office of Administration & Re.

Licensee Per'ormance (SALP) ratings for nuclear power plant sources Management. Director (Post 870413). June 1987.

facilities in operation and under construction. The SALP results 266pp. 8707020359. 41567.001, are listed by NRC region in three sections: the most recent This report covers the major activities, events, decisions and report, operating facihtees, and facilities under constructiort The planning that took place dunng fiscal year 1986 within the NRC historical data raimmary report has been prepared by the NRC or involving the NRC.

Office of inspection and Enforcement (IE). Information con-tained in this report has been updated to include those pub-NUREG-1147 R01: SEISMIC SAFETY RESEARCH PROGRAM hshed S ALP reports received before March 19,1987.

PLAN.

• Division of Er'gineering (Post 870413). May 1987.

200pp. 8706240291. 41438:334.

NUREG 1230 DRFT FC: COMPENDIUM OF ECCS RESEARCH This document presents a plan for seismic research to be FOR REALISTIC LOCA ANALYSIS. Draft Report For Comment.

performed by the Structural and Seismic Engineering Branch in

• Division of Reactor & Plant Systems (Post 870413). April the Office of Nuclear Regulatory Research. The plan desenbos 1987 1348pp. 8705290033. 41 t01:238.

the regulatory needs and related research necessary to address Emergency Core Coohng Systems (ECCS) are required on all the following issues: uncertainties in seismic hazard, earth.

light water reactors (LWRs) in the United States to provide cool-quakes larger than the design bass, seismic vulnerabilities, ing of the reactor core in the event of a break in the reactor shifts in building frequency, piping design, and the adequacy of piping. These accidents are called loss-of Coolant accidents current critena and methods, in addition to presenting current (LOCA), and they range from small leaks to a postulated full and proposed research within the NRC, the plan discusses re-break of the largest pipe in the reactor cooling system. Federal search sponsored by other domestic and foreign sources.

government regulations require that calculations of the LOCA be performed to show that the ECCS will maintain fuel rod clad-NUREG-1166: FINAL E.WIRONMENTAL STATEMENT FOR DE-ding temperatures, cladding oxidation, and hydrogen production COMMISSIONING HUMBOLDT BAY POWER PLANT. UNIT within certain limits. The Nuclear Regulatory Commission (NRC)

3. Docket No. 50-133 (Pacific Gas And Electric Company)

• Divi-and others have completed an extensive investigation of fuel sion of Reactor Projects - lit.IV,V & Special Projects (Post rod behavior and ECCS performance. The technology has been 870411). April 1987.164pp. 8705190499. 40978.319.

advanced to the point that is now possible to make a realistic The Final Environmental Statement contains the assessment estimate of ECCS performance dunng a LOCA and to quantify of the environmental impact associated with decommissioning the uncertainty of this calculation. This report serves as a gen-the Humboldt Bay Power Plant Unit 3 pursuant to the National erat reference for ECCS research. The report (1) summanzes Environmental Policy Act of 1960 (NEPA) and Title 10 of the the understanding of LOCA phenomena in 1974, (2) reviews ex-Code of Federal Regulations, Part 51, as amended, of the Nu-penmental and analytical programs developed to address the clear Regulatory Commission regulations. The proposed decom-phenomena, (3) describes best estimate computer codes devel-missioning would involve safe storage of the facility for about 30 oped by the NRC, (4) discusses the sahent technical aspects of years, after which the residual radioactivity would be removed LOCA phenomena and our current understanding of them, (5) so that the facihty would be at levels of radioactivity acceptable discusses probabilistic nsk studes, and (6) examines the impact for release of the facility to unrestncted access.

of research on the ECCS regulations.

i Miln Cit:ti:ns and Ab:trzets 5

NUREG-1235: TECHNICAL SPECIFICATIONS FOR CLINTON site phase, the staff will work with the facility licensee to con-POWER STATION, UNIT 1. Docket No. 50 461.(lllinois Power duct a review of the four technical areas, and to evaluate the Company)

• Office of Nuclear Reactor Regulation, Director results of tests that are conducted. Findings will be based upon (Post 870411). April 1987. 562pp. 8705120069. 40897:147.

the staff's judgment of the degree of compliance of the simula-The Clinton Power Station, Unit No.1 Technical Specifica-tion facility with 10CFR55 in terms of its suitability for the con-3 tions were prepared by the U.S. Nuclear Regulatory Commis-duct of operating examinations.

/

sion to set forth hmits, operating conditions, and other require-NUREG 1259: TECHNICAL SPECIFICATIONS FOR BEAVER ments applicable to a nuclear reactor facility as set forth in Sec.

VALLEY POWER STATION, UNIT 2. Docket No. 50-412.(Du-J ton 50.36 of 10 CFR 50 for the protection of the health and quesne Light Company)

• Division of Reactor Projects 1/II l

safety of the public.

(Post 870411). May 1987. 429pp. 2706120083. 41282:007.

NUREG-1239: ENVIRONMENTAL ASSESSMENT FOR RENEW-The Beaver Valley Power Station, Unit 2, Technical Specifica-AL OF SOURCE MATERIAL LICENSE NO. STB-401. Docket No.

tions were prepared by the U.S. Nuclear Regulatory Commis-40-6563.(Columbium-Tantalum Dwison, Mallinkrodt,Inc.)

• Divi-sion to set forth the limits, operating conditions, and other re-I sion of Fuel Cycle, Medical, Academic & Commercial Use quirements applicable to a nuclear reactor facility as set forth in Safety (Post 870413). April 1987. 69pp. 8705190502.

Section 50.36 of 10 CFR 50 for the protection of the health and 40980:267.

safety of the public.

l l

In response to an application for renewal of Materials License No. STB-401 for the Columbium-Tantalum Division of Mallinck-NUREG 1261: TECHNICAL SPECIFICATIONS FOR BPAIDWOOD l

rodt, Inc., St. Louis, Missouri, the NRC staff prepared this Envi-STATION, UNITS 1 AND 2. Docket Nos. 50-456 And 50-l ronmental Assessment. The Environmental Assessment in-457.(Commonwealth Edison Company)

• Dwision of Reactor

{

cludes discussions of the need for the proposed renewal, alter-Projects Ill.IV.V & Special Projects (Post 870411). May 1987.

)

natives to the action, and the environmental impacts of pro-225pp. 8707140211. 41692:120.

/

posed action.

The Braidwood Station, Unit Nos.1 and 2. Technical Specifi-cations were prepared by the U.S. Nuclear Regulatory Commis-NUREG-1244: PLAN FOR INTEGRATING TECHNICAL ACTIVl-si n to set forth the limits, operating conditions, and other re-TIES WITHIN THE U.S. NRC AND ITS CONTRACTORS IN THE querements applicable to a nuclear reactor facility as set forth in AREA OF THERMAL HT. RAULICS.

• Office of Nuclear Regu-Section 50.36 of 10 CFR Part 50 for the protection of the health l

latory Research, Rector (Post 860720). April 1987. 42pp.

and safety of the public.

8705120074. 40896:B2.

The Executwe Director for Operations (EDO) directed the NUREG-1265: UNCERTAINTY PAPERS ON SEVERE ACCIDENT NRC staff to prepare a coordinated plan for the integration of SOURCE TERMS.

• Office of Nuclear Regulatory Research, Di-technical activities within the agency and specified a number of rector (Post 860720). May 19f'7, 180pp. 8707140200.

issues to be addressed. This report summantes the status of 41691:216 agency programs involved in thermal hydraulic research and An assessment of the severe accident source term technolo-proposes management methods to accomplish the EDO's direc-gy was recently published by the NRC in NUREG-0956. State-twes.

of-the-art methods desenbed in NUREG-0956 are now being NUREG 1245 V01: RADIOACTIVE WASTE MANAGEMENT RE.

used in risk assessments and as the basis for implementing the SEARCH PROGRAM PLAN FOR HIGH-LEVEL WASTE - 1987.

NRC's Severe Accident Policy Statement and its Safety Goal.

• Dwision of Engineering (Post 870413). May 1987. 62pp.

Notwithstanding major advances in source term technology re-8706120346.41280:026.

sutting from recent severe accident research programs, The program of research desenbed in this plan is intended to NUREG-0956 identified eight technical areas where uncertain-identify and resolve technical and scientific issues invoNad in ties remain large and where our near term research efforts the NRC's licensing and regulation of disposal systems intend-should be focused. Iridividual programs within the severe acci-ed to isolate high-level hazardous radioactwo wastes (HLW) dent research program are being adjusted to address these i

from the human environment. The Plan describes the program eight areas of uncertainty with a concentrated effort. To plan for goals, discusses the research approach to be used, lays out these program changes, NRC research program managers have peer review procedures, discusses the history and development revtewed the nature of the uncertainties in their respectwe sub-of the high-level radioactive waste problem and the research ject areas and prepared background papers. These background effort to date and desenbes study objectwes and research pro-papers (or uncertainty papers) are presented in this report.

grams in the areas of: (a) materials and engineering (b) hydrol-NUREG 1269: LOSS OF RESIDUAL HEAT REMOVAL ogy and geochemistry, and (c) compliance with international SYSTEM.Diablo Canyon Unit 2, April 10.1987. CREWS,J.L.;

waste management research programs. In addition, a proposed TRAMMELL,C.M.; LYON W.C.; et al. Region 5, Office of Direc.

Earth Science Seismotectonic Research Program plan for radio-tor. June 1987.105pp. 8707060049. 41584:040.

actwe waste facilities is appended This report presents the findings of an NRC Augmented in.

NUREG-1258 DRFT: EVALUATION PROCEDURE FOR SIMULA.

spection Team (AIT) investigation into the circumstances asso-TlON FACILITIES CERTIFIED UNDER 10CFR55 Draft Report.

ciated with the loss of residual heat removal (RHR) system ca-LAUGHERY,K.R.; PLOTT.C.; WACHTEL,J. Dwision of Human pability for a penod of approximately one and one-half hours at Factors Technology (851125-870411). March 1987, 141pp.

the Diablo Canyon, Unit 2 reactor facility on April 10,1987. This i

8704270073, 40682:096.

event occurred while the Diablo Canyon, Unit 2, a pressurized This document desenbes toe process to be followed by the water reactor, was shutdowa with the eactor coo! ant system NRC for the inspecton of simulation facilities certJfied by facility (RCS) water level drained to approximately mid level of the hot licensees in accordance with 10CFR55. Such inspections are leg piping. The reactor containment building equipment hatch dwided into four major technical areas: performance testing; was removed at the time of the event, and plant personnel were physical fidelity / human factors; control capabilities; and design, in the process of removing the pnmary side manways to gain updating modification and testing. Inspections will be performed access into the steam generator channel head areas. Thus, two by NRC staff with interdisciplinary skills including license exam-fisson product bamers were breached throughout the event.

~,

iner, operations specialist and human factors expert. Inspec.

The RCS temperature increased from approximately 87 degrees tions may consist of off-site and/or on-site phases. The off-site F to bulk boiling conditions without RCS temperature indication j

phase consists of an examinaton of simulation facility doCU-available to the plant operators. The RCS was subsequently mentaten, and the identification of those operations that may pressun2ed to approximately 7-10 psig. The NRC AIT members be Considered for use in on-stte performance testing. In the on-concluded that the Diablo Canyon, Unit 2 plant was, the time of i

l

6 M:In Cit:tions End Abstracts the event, in a condition not previously analyzed by the NRC this information to the staff, the ESRP is intended to assure staff. The AIT findings from this event appear significant and ge-quahty and uniformity of approach in individual reviews as well nene to other pressunzed water reactor facilities licensed by the as compliance with the National Environmental Policy Act of NRC.

1969. In addition, the ESRP will make information about the en-NUREG-1270 V01: INTERNATIONAL CODE ASSESSMENT AND vironmental component of the licensing process more readily APPLICATIONS PROGRAM. Annual Report.

TING,P.;

available and thereby improve the understanding of this process HANSON.R.; JENKS.R. Office of Nuclear Regulatory Research, among the public, States and Regional Compacts and the regu-Director (Post 860720). March 1987. 298pp. 8704270543.

lated community.

40713 223.

NUREG/CP-0078: PROCEEDINGS OF THE SYMPOSIUM ON The first ICAP Annual Report is devoted to coverage of pro-CHEMICAL PHENOMENA ASSOCIATED WITH RADIOACTIV-gram activities and accomplishments dunng the penod between ITY RELEASES DURING SEVERE NUCLEAR PLANT ACCI-April 1905 and March '987. The ICAP was organized by the DENTS. NIEMCZYK,S.J. Amencan Chemical Society. June Office of Nuclear Regulatory Research, United States Nuclear 1987 700pp. 8706250118. 41465:079.

Regulatory Commission in 1985. The ICAP is an intemational The Symposium on Chemical Phenomena Associated with cooperative reactor safety research program planned to provide independent assessment of the NRC computer codes devet-Radioactivity Releases Dunng Severe Nuclear Plant Accidents oped for analysis of reactor transients and loss of-coolant acci' was held during the Amencan Chemical Society National Meet-dents.

ing in Anaheim, California, September 9-12, 1986. The purpose of the symposium was to provide a forum for discussion of NUREG 1271: GUIDELINES AND PROCEDURES FOR THE chemical processes and phenomena potentially occurring during INTERNATIONAL CODE ASSESSMENT AND APPLICATIONS severe nuclear reactor accidents. The symposium included an PROGRAM. TING P.; BESSETTE.D.; HANSON.R. Division of overview setsion designed to help place chemical issues in Reactor & Plant Systems (Pest 870413). April 1987. 80pp.

context in a severe accident perspective, as well as six ses-8705120086.40896:073.

sions devoted to a vanety of severe accident chemistry topics.

This document presents the guidelines and procedures by Those topics included releases of radioactive and nonradioac-which the International Code Assessment and Applications Pro-tive species from core materials in the reactor vessel; transport gram (ICAP) will be conducted. The document summarizes the and behavior of those species in the reactor coolant system; management structure of the program and the relationships be-transport and behavior of released species within the contain-tween and responsibilities of the United States Nuclear Regula-ment; releases dunng core-concrete interactions, as well as tory Commission (USNRC) and the international participants.

other aspects of such interactions; effects of engineered safety The procedures for code maintenance and necessary documen-features and other plant systems; effects of extreme in-plant en-tation are desenbed. Guidelines for the performance and docu-vironments (such as high radiation fields and combustion mentation of code assessment studies are presented. An over-zones); and potentially disruptive phenomena (such as hig,-

view of an effort to quantify code uncertainty, which the ICAP pressure ejection of the melt from the vessel). The proceedings supports, is included.

reprnent the compilation of all the papers presented at the NUREG 1272: REPORT TO THE U.S. NUCLEAR REGULATORY symposium.

COMMISSION ON ANALYSIS AND EVALUATION OF OPER-NUREG/CP-0086 V01: PROCEEDINGS OF THE 10TH DOE /NRC ATIONAL DATA 1986. ' Office for Analysis & Evaluation of NUCLEAR AIR CLEANING CONFERENCE. Held in Operational Data, Director. May 1987. 249pp. 8706160095.

41315:007.

Seattle, Washington, August 18 21,1986. FIRST,M.W. Harvard School of Public Heaith, Boston, MA. May 1987. 650pp.

This annual report of the U S. Nuclear Regulatory Commis-8706250268. CONF-860820. 41463:073.

sion's Office for Analysis and Evaluation of Operational Data (AEOD) is devuted to the activities performed during the calen-This document contains the papers and the associated dis-dar year 1986. Comments and observations are provid9d on op-cussions of the 19th DOE /NRC Nuclear Air Cleaning Confer-eratsng exponence at nuclear power plants and other NRC h-ence. Sessions were devoted to (1) fire, explosion and accident consees, including results from selected AEOD studies; summa-analysis, (2) adsorption and iodine retention, (3) fdters and fdter nes of abnormal occurrences involving U.S. nuclear plants; re-testing, (4) standards and regulation, (5) treatment of radon, views of hcensee event reports and their quality, reactor scram krypton, tntium and carbon-14 (6) ventilation and as cleaning exponence from 1984 to 1986, engineered safety features actu-in reactor operations, (7) disolver off-gas cleaning, (B) adsorber ations, and the trends and patterns analysis program; and as-fires, (9) nuclear grade carbon testing, (10) sampling and moni-sessments of nonreactor and medical misadministration events.

toring, and (11) field test experience.

In addition, the report provides the year end status of all recom-NUREG/CP-0086 V02: PROCEEDINGS OF THE 19TH DOE /NRC mondations included in AEOD studies, and listings of all AEOD NUCLEAR AIR CLEANING CONFERENCE. Held in reports issued from 1p80 through 1986.

Seattle, Washington, August 18-21,1986. FIRST,M W, Harvard NUREG 1300: ENVIRONMENTAL STANDARD REVIEW PLAN F )R THE REVIEW OF A LICENSE APPLICATION FOR A 06 0 14 F 860 2'0. 4 442 03 See NUREG/CP-0086,V01 abstract.

LOW LEVEL RADIOACTIVE WASTE DISPOSAL FACILITY, PANGBURN.G C.

Office of Nuclear Matenal Safety &

NUREG/CP-0087:

SUMMARY

REPORT OF THE SYMPOSIUM Safeguards. Director.

April 1987.

251pp.

8705120121.

ON SEtSMIC AND GEOLOGIC SITING CRITERIA FOR NUCLE-

'i 40899:001.

AR POWER PLANTS CUMMINGS.G E.: BERNREUTER,0.L.;

The Environmental Standard Review Plan (ESRP) (NUREG-MURRAY,R C.: et at Lawrence Livermore National Laboratory.

1300) provides guidance to staff reviewers in the Office of Nu-June 1987. 49pp. 8706300026. UCID 21039. 41515.242.

clear Matenal Safety and Sateguards who perform environmen-This is a summary report of the Symposium on Seismic and tal reviews of Applicant's environmental reports prepared in Geologic Siting Cnteria for Nuclear Power Plants held in Rock-support of heense apphcations to construct and operate new vdle, Maryland, on October 7 9, 1986. The purpose of the sym.

Iow-level radioactive waste disposal facilities. The individual posium was to provide a forum to delineate and examine the ESRP's which make up this document dentify the information issues relating to a proposed revision tc Appendix A of 10 CFR considered necessary to conduct the review, the purpose and Part 100. Appendix A to Part 100 is the basic U S. Nuclear Reg-scope of the review, the analysis procedure and evaluation, the ulatory Commission regulation conceming seismic siting entena formal inputs made to the Environmental Statement and the rel.

for nuclear power plants. Some of the issues discussed at the erences considered appropnate for each review. By providing symposium pertained to incorporating current probabihstic con-

)

Main Cit:tions and Abstracts 7

cepts into the regulation, decoupling the OBE and SSE, and re.

Identification of Age Related Failure Modes; Application of vising the definition of basic concepts and parameters. The HRA/PRA Results to Support Resolution of Generic Safety extent of any revision is yet to be determined.

Issues involving Human Performance, Protective Action Deci-sionmabng. Masehning of % for kn, MainM %

NUREG/CR 2000 V06 N2: LICENSEE EVENT REPORT (LER) fo ance Design Objective, and Operational Safety Reliability COMPILATION For Month Of February 1987.

• Oak Ridge Na' R

tional Laboratory. March 1987. 123pp. 8704270093. ORNL/

NSIC-2000. 40678 297.

NUREG/CR-2850 V05: POPULATION DOSE COMMITMENTS This monthly report contains Licensee Event Report (LER)

DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER operational information that was processed into the LER data PLANT SITES IN 1983. BAKER,D.A.; PELOQUIN,R.A. Battelle file of the Nuclear St.fety Information Center (NSIC) during the Memorial Institute, Pacific Northwest Laboratones. April 1987.

one month penod identified on the cover of the document. The 141pp. 8704280126. PNL-4221. 40713:082.

LERs, from which this information is derived, are submitted to Population radiation dose commitments have been estimated the Nuclear Regulato y Commission (NRC) by nuclear power from reported radionuclides releases from commercial power re-plant licensees in accordance with federal regulations Proce-actors operating dunng 1983. Fifty-year dose commitments from dures for LER reporting for revisions to those events occumng a one-year exposure were calculated from both liquid and at-prior to 1984 are desenbed in NRC Regulatory Guide 1.16 and mosphenc releases for four population groups (infant, child, NUREG-1061, " Instructions for Preparation of Data Entry teen-ager and adult) residing between 2 and 80 km from each Sheets for Licensee Event Reports." For those events occurnng of 52 sites. This report tabulates the results of these calcula-on and after January 1,1984, LERs are being submitted in ac-tions, showing the dose commitments for both liquid and air-cordance with the revised rule contained in Title 10 Part 50.73 borne pathways for each age group and organ. Also included of the Code of Federal Regulations (10 CFR 50.73 - Licensee for each of the sites is a histogram showing the fraction of the Event Report System) which was published in the Federal Reg-total population within 2 to 80 km around each site receiving ister (Vol. 40, No.144) on JJIy 26,1983. NUREG 1022. "Li-vanous average dose commitments from the airbome pathways.

consee Event Report System Desenption of Systems and The total dose commitments (from both liquid and airborne Guidelines for Reporting." provides supporting guidance and in-pathways) for each site ranged from a high of 45 person-rem to formation on the revised LER rule. The LER summanes in this a low of 0.002 person-rem for the sites with plants operating report are arranged alphabetically by facility name and then throughout the year with an anthmetic mean of 3 person-tem.

chronologically by event date for each facility. Component-The total population dose for all sites was estimated at 170 system, keyword, and component vendor indexes follow the person-rem for the 100 million people considered at risk. The summaries. Vendors are those idenhfied by the utility when the site average individual dose commitment from all pathways L

LER form is initiated-the keywords for the component, system, ranged from a low of 1 x 10( 6) mrem to a high of 0.06 mrem.

j and general keyword indexes are assigned by the computer No attempt was made in this study to determine the maximum

)

using correlation tables from the Sequence Coding and Search dose commitment received by any one individual from the ra-System-dionuchdes released at any of the sites.

NUREG/CR-2000 V06 N3: LICENSEE EVENT REPORT (LER)

COMPILATION For Month Of March 1987.

• Oak Ridge Nation.

NUREG/CR 3231:

PIPE TO-PIPE IMPACT PROGRAM.

al Laboratory Apnl 1987.107pp. 8705190343. ORNL/NSIC-ALZHEIMER,J.M.; BAMPTON.M C.; FRILEY,J.R.; et al. Battella 2000. 40975 324.

Memonal Institute, Pacific Northwest Laboratories. May 1987.

See NUREG/CR 2000,V06,N02 abstract 104pp. 8706240152, PNL-5779. 41452:059.

The objective of this research was to determine the extent of NUREG/CR 2000 V06 N4: LICENSEE EVENT REPORT (LER) damage that occurs when two pipes experience an impact COMPILATION For Month Of April 1987.

• Oak Ridge National event due to one whipping against the other. The research was Laboratory. May 1987.150pp 8706230059. ORNL/NSIC 2000.

conducted through expenmental and analytical approaches. The 41421:297.

former required the development of a specialized impact ma-See NUREGiCR.2000,V06.N02 abstract-chine that could accelerate a whipping pipe with sufficient NUREG/CR 2000 V06 NS: LICENSEE EVENT REPORT (LER) energy to cause failure of a target pipe that was heated and COMPILATION For Month Of May 1987.

• Oak Ridge Nstional pressurized to Pressunzed Water Reactor (PWR) conditions.

Laboratoty. June 1987,110pp.8707090381. ORNL/NSIC 2000.

Damage was measured in terms of crushing, bending, and fail-41633 074.

ure. The results of the tests permitted the correlation between See NUREG/C42000,V06,N02 abstract.

pipes of a certain site and the damage they could cause when impacting with a certain amount of known energy. These results i

NUREG/CR 2331 V06 N3: SAFETY RESEARCH PROGRAMS were used to evaluate the pipe whip cnteria in the Standard SPONSORED BY OFFICE OF NUCLEAR REGULATORY Review Plan 3 6 2-4. It was estabhshed that the criteria condi-RESEARCH Ouarterly Progress Report, July September 1986.

tions did not fully represent the results obtained expenmentally.

WEISS.A J Brookhaven National Laboratory. March 1987.

An analysis procedure to model the pipe whip event was devel-114pp.8706120037. BNL NUREG-51454 41187:087-oped and used to establish the test matnx for the expenmental l

l This progress report will desenbe current activities and techna-program. This analytical procedure can also be used to predict cal progress in the programs at Brookhaven National Laboratory deformation and rupture for postulated pipe whip scenanos.

sponsored by the Division of Accident Evaluation, Division of Engineenng Technology, and Division of Risk Analysis & Oper-NUREG/CR-3319 R01: LWR PRESSURE VESSEL SURVElL-ations of the U S Nuclear Regulatory Comrmssion, Office of Nu-LANCE DOSNETRY IMPROVEMENT PROGRAM. LWR Power l

clear Regulatory Research. The projects reported are the fol-Reactor Surveillance Physics-Dosimetry Data Base Compendi-l lowing High Temperature Reactor Research, SSC Code im-um. MCELROY,W.N. Hanford Engineenng Development Labora-provements, Thermal Hydraulic Reactor Safety Expenments, tory. May 1987. 86pp. 8706160066. HEDL TME 85-3.

Thermodynamic Core-Concreto Interaction Expenments and 41310.239.

Analysis. Plant Analyzer, Code Assessment and Application, This NRC physics dosimetry compendium (Sections 1.0 Code Maintenance (RAMONA 3B). MELCOR Venfication and through 4.0) is a coltation of information and data developed Benchmarking, Source Term Code Package Venfication and from available research and commercial fight water reactor Benchmarking. Uncertainty Analysis of the Source Term, Stress vessel surveillance program (RVSP) documents and related sur.

Corrosion Cracking of PWR Steam Generator Tubing. Soil-veillance capsule reports. The Section 4.0 data represents the Structure Interaction Evaluation and Structural Benchmarks.

results of the HEDL least squares FERRET-SAND 11 Code re-

8 Main Cit:ti:ns cnd Abstr: cts evaluation of exposure units and values for 47 PWR and BWR cause of high resistance paths, with most of the problems surveillance capsules. Using a consistent set of auxiliary data caused by corrosion. Results indicate a correlation between the and dcsimetry-adjusted reactor physics results, the revised chronological age of circuits and circuit degradation.

fluence values for E greater than 1 MeV averaged 25% higher than the onginally reported values. The range of fluence values NUREG/CR-4082 V05: DEGRADED PIPING PROGRAM PHASE (new/old) was from a low of 0.80 to a high of 2.38. These ll. Semiannual Report, April-September 1986. WILKOWSKI,G.M.;

HEDL denved FERRET-SAND 11 exposure parameter values AHMAD,J.; BARNES.C.R.; et al. Battelle Memorial institute, Co-are being used for NRC-supported HEDL and other PWR arid lumbus Laboratories. April 1987. 238pp. 8704270130. BMI.

BWR trend curve data development and testing Studies, which 2120. 40675:326.

support Revision 2 of Regulatory Guide 1.99. These trend Presented herein is the Fifth Semiannual Report of the U.S.

curves are used by the utilities and by the NRC to account for NRC's Degraded Piping Program - Phase 11. The intent of this neutron radiation damage in setting pressure / temperature limits, program is to experimentally validate and enhance available en-in analyzing fractures, and in predicting neutron-induced afytical methods for evaluating the mechanical behavior of nu-changes in reactor PV steel fracture toughness and embrittle-clear power plant piping containing circumferentially-oriented ment dunng the vessel's service life. The status of the develop-defects. Fifty-one pipe experiments have been conducted to ment and application of new advancements in LWR reactor sur-date. These and approximately 42 additional pipe expenments veillance programs is discussed, such as cavity physics-dosime-from other programs have been analyzed. In the analytical try for improving the reliability of current and end of life (EOL) effort, a screening criterion has been developed to show when predictions on the metallurgical conditions of pressure vessels the nel-section. collapse analysis is valid. This shows that even and their support structures.

tough materials such as stainless steel can fail at less than net-NUREG/CR-3925 REV-SWIFT 11 SELF TEACHING section-collapse loads if the pipe diameter is sufficiently large.

CURRICULUM.filustrative Problems For The Sandia Waste-Iso-Numerous predictive J-estimation schemes have been evaluat.

lation Flow And Transport Model For Fractured Media.

ed and modified. A finite length surface cracked pipe estimation REEVES,M.; WARD.D.S.; DAVIS,P.A.; et al Sandia National scheme has also been developed. Finite element analyses of Laboratories. January 1987. 794pp. 8704300036. SAND 84-specimens with welds suggest that the size of the welo relative 1586. 40749:199 to the specimen or structure size can affect the deformation J Several documents have been wntten describing SWIFT 11, values. Supporting research efforts involve investigating geome-the most current version of the SWIFT (Sandia Waste Isolation try effects on J.R curves, as well as characterizing the material Flow and Transport) Model. Reeves et al. (1986a), descnbes properties for each pipe tested.

the theory and implementation, and Reves et al. [1986b], de-NUREG/CR-4098: SEISMIC-FRAGILITY TESTS OF NEW AND scribes the required input of data and parameters. Ward et al.

ACCELERATED AGED CLASS 1E BATTERY CELLS.

(1984a), and (1984b], describe the companson of the results BONZON,L.L. Sandia National Laboratories. JANIS,W.J.;

from the SWIFT code with field data and other existing codes.

This document is devoted to assisting the analyst who desires BLACK,D.A.; et al. Ontario Hydro. January 1987. 135pp.

to use the SWIFT 11 code. The analyst is referred to the User's 8705200231. SAND 84 2631. 40984:326.

Manual for SWIFT 11 (Reeves et al. [1986b]) for detailed data The seismic fragility response of naturally aged nuclear sta.

input instruc tions. Eight examples are presented to illustrate the tion safety related batteries is of interest for two reasons: (1) to use of SWIF T 11. The implementation of the numerical simulation determine actual failure modes and thresholds and (2) to deter-of the physical prcblem is desenbed for each example. For mine the validity of using the ekictrical capacity of individual cells as an indicator of the potential survivability of a battery he t ut are r v de given a sesmic event. Pnor reports in this senes discussed the seismic fragility tests and results for three specific naturally-NUREG/CR 3956: IN SITU TESTING OF THE SHIPPINGPORT aged cell types: 12-year old NCX 2250,10-year old LCU-13, ATOMIC POWER STATION ELECTRICAL CIRCUITS.

and 10-year old FHC-19. This report focuses on the comple-DiNSEL,M R.; DONALDSON M.R.;

SOBERANO,F.T. EG&G mentary approach, namely the seismic-fragility response of ac-Idaho, Inc. (subs. of EG&G, Inc.). April 1987. 49pp.

celerated-aged batteries. Of particular interest is the degree to 8706120172. EGG-2443. 41281:001.

which such approaches accurately reproduce the actual failure This report discusses the results of electncal in situ testing of modes and thresholds. In these tests the significant aging ef-selected circuits and components at the Shippingport Atomic fects observed, in terms of seismic survivabinty, were: embrittle-Power Station in Shippingport, Pennsylvania. Testing was per-ment of cell cases, positive bus matenal and positive plate formed by EG&G Idaho in support of the United States Nuclear active matenal causing hardening and expansion of positive Regulatory Commiss>on (USNRC) Nuclear Plant Aging Research plates. The IEEE Standard 535 accelerated aging method suc-(NPAR) Program. The goal was to determine the extent of aging cessfully reproduced seismically significant aging effects in new or degradation of vanous circuits from the original plant, and the cells but accelerated grid embnttlement an estimated five years two major coreplant upgrades (representing three distinct age beyond the conditional age of other components.

groups), as well as to evaluate previously developed surveil-lance technology. The electrical tasting was performed using NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH-the Electncal Circurt Charactenzation and Diagnostic (ECCAD)

LEVEL WASTE REPOSITORY. Volume 2: Tuff. DIDWALL,E.M.

system developed by EG&G for the U.S. Department of Energy Lawrence Livermore National Laboratory. BENSON,S.M.;

to use at TMI-2. Testing included measurements of voltage, ef-BINNALL,E.P.; et al. Lawrence Berkeley Laboratory. May 1987, fective senes inductance, impedance, effective series resist-106pp. 8706030120. UCID-20092. 41144:242.

ance, de resistance, insulation resistance and time domain re.

This report addresses entical parameters specific to a reposi-flectometry (TDR) parameters. The circuits evaluated included tory in tu", using the Yucca Mountain tuffs of Nevada as the pressurizer heaters, control rod position indicator cables, and pnncipal example. For the purposes of this report, a parameter safety injection system motor operated valves. It is to be noted is considered to be a physical property whose value helps de-that the operability of these circuits was tested after several termine the charactenstics or behavior of a repository system.

years had elapsed because plant operations had concluded at Parameters which are defined as entical are those essential to Shippingport. There was no need following plant shutdown to evaluate and/or monitor leakage of radionuclides from the re-j retain the circuits in working condition, so no effort was expend.

pository and to evaluate the need for retrieval. The parameters i

ed for that purpose. The in situ measurements and analysis of are considered with respect to the disciplines of geomechanics, the data confirmed the effectiveness of the ECCAD system for geology, hydrology, and geochemistry and are rank ordered in i

detecting degradation of circun connections and splices be-terms of importance. The specific role of each parameter, spe-

M;in Cit;tians End Abstr cts 9

cific factors affecting the measurement of each parameter, and and strong cooperative links between the U.S. NRC-supported the interrelationships between the parameters are considered in activities at HEDL, ORNL, NBS, and MEA and those supported detail.

by CEN/SCK (Moi, Belgium). EPRI (Palo Alto, USA), KFA NUREG/CR 4165: SEVERE ACCIDENT SEQUENCE ANALYSIS (Julich, Germany), and several United Kingdom Laboratories have been extended to a number of other countries and labora.

PROGRAM ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULATIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT tories. These cooperative links are strengthened by the active 1.

DALLMAN R.J.; GOTTULA.R.C.; HOLCOMB,E.E.; et al.

membership of the scientific staff from many participating coun-EG&G idaho, 'nc. (subs. of EG&G, Inc.). May 1987. 92pp.

tries and laboratones in the ASTM E10 Committee on Nuclear 8707060336. EGG-2379. 41561:147.

Technology and Applications. Several subcommittees of ASTM An analysis of five antrespated transients without scram E10 are responsible for the preparation of LWR surveillance (ATWS) was conducted at the Idaho National Engineenng Labo-standards. Results of FY 86 research by a number of LWR-PV.

I ratory (INEL) The five detailed deterministic simulations of pos.

SDlP participants are reported in this progress report.

tulated ATWS sequences were initiated from a main steam line isolation valve (MSIV) closure. The subject of the analysis was NUREG/CR-4330 V03: REVIEW OF LIGHT WATER REACTOR the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor REGULATORY REQUIREMENTS. Assessment Of Selected I

(BWR) of the BWR/4 product line with a Mark I containment.

Regulatory Requirements That May Have Marginal Importance j

The simulations yielded insights to the possibie consequences To Risk:Postaccident Sampling System.

Turbine J

resulting from a MSIV closure ATWS. An evaluation of the ef.

Missiles, Combustible Gas Control, Charcoal Filters. SCOTT W.B.:

tects of plant safety systems and operator actions on accident JAMISON.J.D.; STOETZEL,G.A.; et al. Battelle Memorial Insti-progression and mitigation is presented.

tute, Pacific Northwest Laboratories. May 1987. 134pp.

8706150009. PNL 5809. 41301:193.

NUREG/CR-4219 V03 N2: HEAVY SECTION STEEL TECHNOL-OGY PROGRAM. Semiannual Progress Report For April-Sep-In a study Commissioned by the Nuclear Regulatory Commis-tember 1986. PUGH,C E. Oak Ridge NatioM Laboratory. De.

sion (NRC), Pacific Northwest Laboratory (PNL) evaluated the cember 1986.234pp.8706030097. ORNL/TM 9593. 41151:228.

costs and benefits of streamlining regulatory requirements in the The Heavy Section Steel Technology (HSST) Program is an areas of postaccident sampling systems, turbine missiles, com-engineenng research activity conducted by the Oak Ridge Na-bustible gas control, and impregnated charcoal filters. The basic tional Laboratory for the Nuclear Regulatory Commission. The framework of the analyses was that presented in the Regulatory program compnses studies related to all areas of the technolo-Analysis Guidelines (NUREG/BR-0058) and in the Handbook gy of material fabncated into thick-section primary-coolant con-for Value-Impact Assessment (NUREG/CR-3568). Thu effects tainment systems of light-water-cooled nuclear power reactors.

of streamlined regulations were evaluated in terms of such fac-The investigation focuses on the behavior and structural integri-tors as population dose and costs to industry and NRC. The re-ty of steel pressure vessels containing cracklike flaws. Current sults indicate that streamlining regulatory requirements in three work is organized into ten tasks: (1) program management (2) of the four areas, i.e., pobtaccident sampling systems, turbine fracture-methodology and analysis, (3) material characterization missiles, and combustible gas control, would have little impact and properties, (4) environmentally assisted crack growth stud-on public risk. Streamlined regulatory requirements in the fourth ies, (5) crack-arrest technology, (6) irradiation effects studies.

area, impregnated charcoal fitters, might increase public nsk.

(7) cladding evaluations, (8) intermediate vessel tests and anal' Cost evaluations indicate substantial savings by lengthening the i

ysis, (9) thermal-shock technology, and (10) pressunzed ther-inspection interval for low pressure turbine rotors. Small-to-l mal-shock technology.

moderate saving may be realized through postulated modifica-l NUREG/CR-4300 V04 N1: ACOUSTIC EMISSION / FLAW RELA.

tions to the postaccident sampling system requirements and to l

TIONSHIP FOR IN SERVICE MONITORING OF NUCLEAR the combustible gas control requirements for hydrogen recom-PRESSURE VESSELS Progress Report, October 1986 - March biners in inerted BWR Mark I and ll containments. The results 1987. HUTTON,P.H.; FRIESEL,M.A. Battelle Memorial Institute, indicate that the use of impregnated charcoal filters is the most Pacific Northwest Laboratories. June 1987. 20pp. 8707100289.

cost-effective method of radioiodine removal from building venti-PNL 5511. 41663.077.

lation systems.

This report discusses technical progress for the penod Octo-ber 1986 to March 1987 for the NRC sponsored research pro-NUREG/CR-4469 V05: NONDESTRUCTIVE EXAMINATION gram concerned with " Acoustic Emission / Flaw Relationships for (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT Inservice Monitonng of Nuclear Reactor Pressure Boundaries "

WATER REACTORS. Semiannual Report, April-September 1986.

Topics discussed included AE monitonng of pnmary piping DOCTOR.S R.; BATES,0.J.; DEFFENBAUGH.J.; et al. Battelle dunng reactor operaSon, development of AE/lGSCC relation-Memorial institute, Pacific Northwest Laboratories. April 1987.

ships, evaluation of the effects of crack growth rate on detec.

39pp. 8705280412. PNL-5711,41097:183.

tion of the associated AE, validation of the AE signal identifica-Evaluation and improvement of NDE Reliability for Inservice tion method, and progress in developing an ASME Section XI Inspection of Light Water Resctors (NDE Reliability) Program at Code appendix for continuous AE monstonng of pressure the Pacific Northwest Laboratory was established by the Nucle-boundary components-ar Regulatory Commission to determine the reliability of current NUREG/CR-4307 V03: LWR PRESSURE VESSEL SURVEIL-inservice inspection (ISI) techniques and to develop recommen-LANCE DO31 METRY IMPROVEMENT PROGRAM.1986 Annual dations that will ensure a suitably high inspection reliability. The Report. October 1985 September 1986. MCELROY,W.N Han.

objectives of this program include determining the reliability of ford Engineenng Development Laboratory. April 1987. 232pp.

ISI performed on the pnmary systems of commercial light-water 8705120103. HEDL-TME 86-2. 40896.2f 6.

reactors (LWRs); using probabilistic fracture mechanics analysis The Light Water Reactor Pressure Vessel Surveillance Dosim.

to determine the impact of NDE unreliability on system safety; etry improvement Program (LWR-PV SDlP) has been estab-and evaluating reliability improvements that can be achieved fished by the U S Nuclear Regulatory Commission (NRC) to im-with improved and advanced technology. A final objective is to prove, test, venty, and standardize the physics-dosimetry-metal.

formulate recommended revisions to ASME Code and regula-turgy, damage correlation, and t.ssociated reactor analysis tory requirements, based on material properties, service condi-methds. procedures and data us,ed to predict the integrated tions, and NDE uncertainties. The program scope is limited to effect of neutron exposure to LWR pressure vessels and their ISI of the pomary systems including piping, vessel, and other irk support structures A vigorous research effort attacking the spected components. This is a progress report covering the same measurement and analyms problems exists worldwide, programmatic work from April 1986 through September 1986.

10 Miln Cititi:ns tnd Ab:trects NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION year. Small LOCAs with failure of emergency ccount recircula.

(NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT tion account for over half of core damns frequency. Loss of WATER REACTORS. Semiannual Report, October 1986 - March component cooling water, leading to a reactor coolant pump 1987. DOCTOR.S.R ; DEFFENBAUGH,J.; GOOD,M.S.; et al.

seal LOCA is the next largest contributor, accounting for nearly Battelle Memonal Institute, Pacific Northwart Laboratories. June one third of core damage frequency. Station blackout se-1987. 63pp. 8707080329. PNL-5711. 416/1:250, quences account for 5% of core damage frequency. No other The Evaluation and improvement of NDE Reliability for in-sequence types account for more than 5% of core damsge fre-service inspection of Light Water Reactors (NDE Reliability) quency. The numerical results are influenced by modeling as-Program at the Pacific Northwest Laboratory was established by sumptions and data selection for issues such as coolant pump the Nuclear Regulatory Commission to determine the reliability seal LOCA, common cause failure probabilities, and operator re-of current inservice inspection (ISI) techniques and to develop sponse to emergency conditions such as small LOCA and sta-recommendations that will ensure a suitably high inspection reli-fon blackout. The sensitivity studies explore the impact of alter-abihty. The objectives of this program include determining the nate theories and data on these issues The results of the un-reliability of 151 performed on the prtmary systems of commer-certainty and sensitivity analyses should be considered before ciel light-water reactors (LWRs); using probabilistic fracture me-any future actions are taken based on this analysis, chanics analysis to determine the impact of NDE unrehability on system safety; and evaluating reliability improvements that can NUREG/CR-4650 V06PTt: ANALYSIS OF CORE DAMAGE FRE-be achieved with improved and advanced technology. A final OUENCY FROM INTERNAL EVENTS: GRAND GULF, UNIT objective is to formulate recommended revisions to ASME Code

1. Main Report. DROUIN,M.T.; LACHANCE,J.L; SHAPIRO.B.J.;

and Regulatory requirements, based on matenal properties, et al. Science Applications Intematml Corp. (formerly S*nce service conditions, and NDE uncertainties. The program scope Applications, Inc.). April 1987. 500p 8707080284. WNO86-is hmited to iSI of the primary systems including the piping.

2084. 41618:163.

vessel, and other inspected components. This is a progress This document contains the accident sequence analysos for report covering the programmatic work from October 1986 Grand Gulf Unit 1, one of the reference plants being examined through March 1987.

as part of the NUREG-1150 effort by the Nuclear Regulatory NUREG/CR-4527 V01: AN EXPERIMENTAL INVESTIGATION OF Commission. NUREG-1150 will document the risk of a selected INTERNALLY IGNITIED FIRES IN NUCLEAR POWER PLANT group of nuclear power plants. As part of that work, this report CONTROL CABINETS.Part 1: Cabinet Effects Tests.

contains the overall core damage frequency estimate for Grand CHAVEZ,J.M. Sandia National Laboratories. Apnl 1987.117pp, Gulf Unit 1 and the accompanying plant damage state frequen.

8707060099. SAND 86-0336. 41583.285.

cies. Sensitivity and uncertainty analyses provide additional in-A senes of full-scale cabinet fire tests was conducted by sights regarding the dominant contributors to the Grand Gulf Sandia National Laboratories for the U.S. Nuclear Pegulatory core damage frequency estimate. The mean core damage fre-Commission. The cabinet fire tests were prompted by the po.

quency at Grand Gulf was calculated to be 2.9E-5. Station tential threat to the safety of J nuclear power plant by a cabinet blackout type accidents (loss of all AC power accidents) were fire in either the control room or in a switchgear type room. The found to dominate the overall results. Anticipated transient with-purpose of these cabinet fire tests was to characterize the de.

out scram accidents wers also found to be contributors..ne nu-velopment and effects of internally ignited cabinet fires as a merical results are largely driven by common mode failure prob-function of several parameters believed to most influence the ability estimates and, to some extent, human error. Because of burning process. A primary goal of this test program was to test significant data uncertainties in these two areas, it is recom-representative and credible configurations and materials. This mended that the results of the uncertainty and sensitivity analy-senes of 22 cabinet fire tests demonstrated that fires in either ses be considered before any future actions are taken based on benchboard or vertical cabinets with either IEEE183 qualified this analysis. In particular, the single most dominant scenario cable or unqualified cable can be ignited and propagate. How.

may require a more detailed data search and analysis before ever, fires with IEEE 383 qualified cable do not propagate as actions are implemented on the basis of this scenario.

rapidly nor to the extent that unqualified Cable does. Futher-more, the results showed that the thermal environment in the NUREG/CR-4550 V06PT2,: ANALYSIS OF CORE DAMAGE FRE-test enclosure and adacent cabinets is not severe enough to OUENCY FROM INT U NAL EVENTS: GRAND GULFUNIT l

result in autoignition of other combustibles; although in some of

1. Appendices. DROUIN,u.7.: LACHANCE.JL; SHAPIRO.B.J.;

the larger fires melting of plastic materials may occur. Smoke et al. Science Applications inwriaGonal Corp. (formerly Science accumulation in the room appeared to be the m^st significant Applications, Inc.). Apnl 1987. 600pp. 8707080295. SAND 86-problem, as smoke obscured the view in the enclosure within 2084. 41619.317.

minutes after ignition. Essentially, a cabinet fire can propagate See NUREG/CR-4550,V06,PT01 abstract.

within a single cabinet; however, for the conditions tested it NUREG/CR d551 V2 DRF: EVALUATION OF SEVERE ACCl-does not appear that the fire poses a threat outside the burning DENT. RISKS AND THE POTENTIAL FOR RISK cabinet except the resulting smoke.

REDUCTION:SGOUOYAH POWER STATION UNIT 1. Draft For NUREG/CR 4550 V05: ANALYSIS OF CORE DAMAGE FRE.

Comment. BENJAMIN,A.S.; KUNSMAN,0.M.; LEWIS,S.R.; et al.

QUENCY FROM INTERNAL EVENTS: SEQUOYAH UNIT 1.

Sandia National Laboratories. February 1987. 516pp.

BERTUCIO.R C_; MOORE,DL; HELD.J.T.: et al. Sandia Nation-8706120125. SAND 86-1309. 41283:076.

af Laboratories. February 1987. 424pp. 8705200221. SAND 86 The Severe Accident Risk Reduction Pr9 gram (SARRP) has 2084. 40985:101.

completed a rebaselining of the nsks to the INW irom a par.

This document contsna the accident sequence analyses for ticular pressurized water reactor with an ice condenser contain-Sequoyah, Unit I; one of the reference plants being examined ment (Sequoyah, Unit 1). Emphasis was placed on determining as part of the NUREG 1150 effort by the Nuclear Regulatory the magnitude and character of the uncertainties, rather than fo-Commission (NRC). NUREG 1150 will document the nsk of a cusing on a point estimate. The risk-reduction potential of a set selected group of nuclear power plants. As part of that work, of proposed safety option backfits was also studied, and their this report contains the overall core damage frequency estimate costs and benefits were also evaluated. It was found that the for Sequoyah, Unit 1, and the accompanying plant damage nsks from internal events are generally comparable to those state frequencies. Sensitivity and uncertainty analyses provide evatuated in the Reactor Safety Study for a diffsrent pressurized additional insights regarding the dominant contributors to the wst6f reactor, although the calculated uncertainty bands indi-Sequoyah core damage frequency estimate. The mean core

<ste that the nsk could be higher or lower by as much as an damage frequency at Sequoyah was calculated to be 1.0E-4 per order of magnitude. Pnnciple sources of uncertainty include the I

Main Citations and Abstracts 11 modeling of ccmroco-cause failure in the component cooling option backfits was also studied, and their costs and benefits water system, the characteristics of hydrogen generation and were also evaluated It was found that the risks from internal burning, and the possibility of fission product releases that events are generally low relative to previous studies; for exam-bypass the ice condenser. Most of the postulated safety options

• , most of the uncer;ainty range is lower than the point esti-do not appear to be cost offective for the Sequoyah plant; how-rnate of nsk for the Peach Bottom plant in the Reactor Safety ever, certain relatively nexpensive hardware and procedural Study (RSS). However, certain unresolved issues cause the top changes to prevent core damage appear to be marginally cost of the uncertainty band to appear at a level that is comparable effective. It should be noted that this work is based on a draft with the RSS point estimate. These issues include the diesel report of the ASEP study of core-damage frequency for Se-generator failure rate, iodine and cesium reevolution after vessel quoyah. The final ASEP results have lower frequencies for sta-breach, and the possibility of reactor vessel pedestal failure tion-blackout sequences. These finar NJEP results will be incor-caused by core debris attack. Some of the postulated safety op-porated in the final version of this stddj The differences are not tions appear to be potentially cost effective for the Grand Gulf expected to cause significant changes in the conclusions of the power plant, particularly when onsite accident costs are includ-report because risk is influenced by a vanoty of accident se-ed in the evaluation of benefits. Principally these include proce-quences, not just station blackout. This work supports the Nu-dural modifications and relatively inexpensive hardware addi-clear Regulatory Commission's assessment of severe accidents tions to insure core cooling in the event of a station blackout.

in NUREG-1150.

This work supports the Nuclear Regulatory Commission's as-sessment of severe accidents in NUREG-1150.

NUREG/CR-4551 V3 PT1: EVALUATION OF SEVERE ACCl-DENT RISKS AND THE POTENTIAL FOR RISK NUREG/CR-4583 V02: DEVELOPMENT AND VALIDATION OF A REDUCTION. PEACH BOTTOM,0 NIT 2. Main Report Draft For REAL-TIME SAFT-UT SYSTEM FOR THE INSPECTION OF Comment. AMOS.C.N.; BENJAMIN,A.S.; BOYD.G.J.; et al.

LIGHT WATER REACTOR COMPONENTS Annual Sandia National Laboratones April 1987. 200pp. 8707060116.

Report. October 1984 - September 1985. DOCTOR,S.R.;

SAND 86-1309. 41593.078.

HALL,T.E.; REID,L.D.; et al Battelle Memonal Institute Pacific The Severe Accident Risk Reduction Program (SARRP) has Northwest Laboratories. June 1987, 85pp. 8707080345. PNL-completed a rebaselining of the nsks to the public from a boiling 5822. 41621:314.

water reactor with a Mark I containment (Peach Bottom, Unit 2).

The Pacific Northwest Laboratory is working to design, fabri-Emphasis was placed on determining the magnitude and char" cate, and evaluate a real time flaw detection and charactenza-acter of the uncetenties, rather than focusing on a point esti-tion system based on the synthetic aperture focusing technique mate. The risk teduction potential of a set of proposed safety for ultrasonic testing (SAFT-UT). The system is ocsigned to per-option backfits was also studied, #7d their costs and benefits form inservice inspection of light-water reactor components. in-were also evaluated. It was found that the nsks from internal cluded objectives of this program for the Nuclear Regulatory events are generally low relative to previous studies, for exam-Commission are to develop procedures for system calibration pie, most of the uncertainty range is lower than the point esti-and field operation, to validate the system through laboratory mate of nsk for the Peach Bottom plant in the Reactor Safety and field inspections, and to generate an engineering data base Study (RSS). However, certain unresolved issues cause the top to Support ASME Code acceptance of the technology. This

' 3 of the uncertainty band to appear at a level that is comparcble progress report covers the programtr.atic work from October with the RSS point estimate. These issues include the modeling 1984 through September 1985.

of the common-mode failures for the de power system, the like-labood of offsite power recovery versus t!me dunng a station NUREG/CR 4615 V02: MODELING STUDY OF SOLUTE TRANS-I blackout, the probability of drywell failure resultirc from meltth-PORT IN 1HE UNSATURATED ZONE. Workshop Proceedings.

rough of the drywell shell, the magnitude oiM fssion product SPRINGER.E.P.; FUENTES,H.R. Los Alamos National Laborato-releases dunng core-concrete interactions pi tho 6contami-ry. Apn11987. 250pp. 8705190513. LA-10730-MS. 40976:191.

nation effectiveness of the reactor enclosure buildii( Most of These proceedings include the technical papers, a panel the postulated safety options do not appear to be cost effective, summary report, and discussions held at the workshop on Mod-although some based on changes to procedures or inexpensive oling of Solute Transport in the Unsaturated Zone held June 19-hardware additions may be marginally cost effective. This draft 20,1986, at Los Alamos, New Mexico. The central focus of the for comment of the SARRP report for Peach Bottom does not workshop was the analysis of data collected by Los Alamos l

include detailed technical appendices, which are still in prepara-under agreement with the U.S. Nuclear Regulatory Commission tion. The appendices will be issued under separate cover when on intermediate-scale caisson experiments. Five different mod-completed. This work supports the Nuclear f4gulatory Commis-eling approaches were used. The purpose was to evaluate sion's assessment of severe accidents in NUREG-Y F0 models for near surface waste disposal of low level radioactive NUREG/CR 4551 V3 PT2: EVALUATION Di SEVERE ACCl.

wastes. The workshop was part of a larger study being conduct-DENT RISKS AND THE POTENTIAL FOR RISK ed by Los Alamos on transport in the unsaturated zone under REDUCTION. PEACH BOTTOM, UNIT 2 AWndices Draft For agreement with the U S. Nuclear Regulatory Commission.

Comment. AMOS.C.N.; BENJAMIN,A.S ; BOYD G.J.; et al.

NUREG/CR-4617: ONSITE ASSESSMENTS OF THE EFFEC-Sandca National Laboratones May 1987. 500po. 87070601'35~

TlVENESS AND IMPACTS OF UPGRADED EMERGENCY OP-Se EG 455, 03.PT01 abstract' FO%R Ea et al. EG8G Idaho, Inc. (sws. of EG&G, Inc.).

NUREG/CR 4551 V4 DRF: EVALUATION OF SEVERE ACCl-March 1987.157pp. 8706220014. EGG 2456. 41415 226.

DENT RISKS AND THE POTENTIAL FOR RISK The implementation of upgraded emergency operating proce-REDUCTION GRAND GULF, UNIT 1 Draft ' cnr Comment.

dures (EOPS) by the nuclear utilities supports Three Mile Island

(

AMOS,C.N.; BENJAMIN,A S.; BOYD.G.J.; of ar. Sandia National (TMI) Action Plan items 4.C.1 and l C.9. This is the final report of j

Laboratones. Apnl 1987. 750pp 8707060161. SAND 86-1309 a research protect directed at assessing the costs and benefits l

41589:226 of the resultant EOPs. A dual methodology was used to assess The Severe Accident Risk Reduction Program (SARPP has effectiveness; expenmental and onsite data collection. A simula-j completed a rebaselining of the nsks to the public from a boiling tor based, laboratory expenment was conducted, which was de-water reactor with e Mark 111 containment (Grand Gulf, Unit 1).

signed to be sensrtive to changes in operator effectiveness Emphasis was placed on determining the magn tude and nar-using function oriented EOPs versus event based EOPs. The l

acter of the uncertainties, rather than focusing on a point esti-onsite data collected were related to the effectiveness of up.

rnate. The nsk. reduction potential of a set of proposed safety graded EOPs as implemented with all attributes (e g., function

12 Miln Citations and Ab2trt. cts oriented, human factored, etc.). The acceptance of the upgrad-considerations, physical and geochemical charactenzation, com-ed EOPs by the control room operators was also measured. A puter modehng techniques, and parameter estimation proce-cost / benefit ratso for upgraded EOPs was estimated per regula-dures. Radionuclides transport pathways are as solutes in tory analysis guidelines. Summary conclusions are disclosed as groundwater and as vapor through air-filled voids. The latter to the ultimate benefits of the upgraded EOP orogram for the may be important near a hoat source. Water flow and solute regulation of the tafety of commercia! nuclear power plants.

transport properties of a rock matnx may be quantified using NUREG/CR-4623: IN. SITU STRESS MEASUREMENTS IN THE rock core analyses. Natural spatial variation dictates many sam-ples. Observed fractures can be characterized and combined to EARTH'S CRUST IN THE EASTERN UNITED STATES, 1 rm a fracture network for hydraulic and transport assess-RUNDLE,T.A.; SINGH,M.M.; BAKER,C.H. Engineers intemation-ments. Unresolved problems include the relation of network hy-al, Inc. April 1987. 609pp. 8705120101. 40900:099.

draulic conductivity to fluid pressure and to scale. Once charac-l In-situ stress measurements were made in thtee seismic tenzed, the mainx and fracture network can be coupled. Reh-areas in the Eastem United States, using the hydrauhc fractur-ing technique. The areas covered were (i) Moodus, Connecticut, able performance assessment requires additional studies.

(ii) the Ramapo fault system, and (ist) the Central Virginia seis-NUREG/CR-4663: CLOSEOUT OF IE BULLETIN 03-01: FAILURE mic zone. At each location, one borehole was dnlled within the OF REACTOR TRIP BREAKERS (WESTINGHOUSE DB-50) TO seismic zone and the second outside it, so as to compare the OPEN ON AUTOMATIC TRIP SIGNAL, FOLEY,W.J.;

results obtained. No geologic interpretation of the data was DEAN,R S.; HENNICK,A. Parameter, Inc. May 1987. 34pp.

made during this project.

8706240093. IEB-83-01. 41450:074.

NUREG/CR-4651: DEM.DPMENT OF RIPRAP DESIGN CRITE-Dunng startup of Salem Unit 1 on February 25,1983, the un-RIA BY RIPRAP TESTING IN FLUMES Fhase 1. ABT,S.R.;

dervoltage inp attachments (UVTAs) of both Westinghouse DB-KHATTAK,M S.; NELSON.J D.; et al. Colorado State Univ., Fort 50 circuit breakers failed to open automatically upon receipt of Collins, CO. May 1987,120pp. 8705280379. ORNL/TM 10100.

a valid trip signal from the Reactor Protection System (RPS) on 41100:230.

Iow-low steam generator level. The reactor was tripped manual.

Fiume studies were conducted in which riprap embankments fy about 30 seconds later. The manual inp used shunt relays in-were subjected to overtopping flows. Embankment slopes of 1, stalled in the DB-50 breakers. Similar failures of only one of a 2,8,10 and 20% were protected with nprap layers with median pair of DB-50 breakers in senes had been reported to the NRC/

stone sizes of 1,2,4,5 and/or 6 inches. Riprap design critena IE and Westinghouse. Because of concern about the event at for overtopping flowc were developed in terms of unit discharge Salem Unit 1 and previous single failures, the NRC/IE issued at failure, interstitial velocities and discharges through the nprap Bulletin 83-01 on February 25, 1983. Licensees of operating layer, resistance to flow over the nprap surface, potential im-pressurized water reactors with Westinghouse Type DB break-pacts of the filter blanket on the riprap layer stability, and the ers having UVTAs were required to take specific actions. The effects of flow concentration on the reprap stabihty. The result-bulletin was issued for information to all other nuc! car power fa-ing nprap design cnteria were compared to the Stephenson, the cilities. Evaluation of utility responses and NRC/IE inspection j

U S. Army Corps of Engineers, the U.S. Bureau of Reclamation, reports shows that the bulletin can be closed out per specific I

and the Safety Factors methods for riprap stone design; the enteria for all of the 50 facilities to which it was issued for

)

Leps relation for interstitial velocities through riprap; and the An-action. Malfunctions of the UVTAs were reported for only two I

l derson et al. and Corps of Engineers relationships for estimat-sites other than Salem. There are no remaining areas of con-ing Manning's n values for resistance to flow.

corn for this bulletin.

NUREG/CR-4653: GASPAR 11 TECHNICAL REFERENCE AND USER GUIDE STRENGE,D 1.; BANDER,T.J.; SOLDAT J.K. Bat.

NUREG/CR 4664: CLOSEOUT OF IE BULLETIN 83-04: FAILURE tolle Memorial Inctitute, Pacific Northwest Laboratones. March OF THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR 1987. 600pp. 8707090389 PNL-5907. 41631:179.

TRIP BREAKERS. FOLEY,W.J ; DEAN,R.S.; HENNICK,A. Pa-The Nuclear Regulatory Commission's GASPAR ll computer rameter, Inc. May 1987. 31pp. 8706240323. IEB-83-04.

program performs environmental dose analyses from releases 41431216.

of radioactive effluents from nuclear power plants into the at-During shutdown of San Onofre units 2 and 3 on March 3 and l

4 mosphere. The analyses estimate radiation dose to individuals 8,1983, four General Electric (GE) Type AK-2 circuit breakers and population g*oups from inhalation, ingestion, and external in the reactor protection systems (RPSs) failed to open on acti-exposure pathways. The estimated doses provide information vation of the undervoltage tnp coil during testing. S_nce issu-i for National Environmental Policy Act evaluations and for deter-ance of IE Bulletin 79-09 April 17,1979 on failures of GE Type mining compliance with Appendix I of 10 CFR 50. This report AK-2 breakers, additional failures had been reported before the descnbos the mathematical models used in the GASPAR ll tests at San Onofre. Because of concern about continued fail-computer program, instructs the user in prepanng input to the ures of the subject breakers in RPSs, the NRC/IE issued IE Bulletin 83-04 on March 11, 1983. All licensees of operating program, and supplies detailed information on program structure and parameters used to modify the program.

pressunzed water power reactors, except those with Westing-house Type DB breakers, were required to take five specific ac-NUREG/s ' 4655: UNSATURATED FLOW AND TRANSPORT tions. The bulletin was issued for information to all other nuclear THROUGH FRACTURED ROCK RELATED TO HIGH-LEVEL power facilities. Evaluation of utility responses and NRC/IE in-WASTE REPOSITORIES Final Report - Phase ll.

spection reports indicates that the bulletin can be closed out for RASMUSSEN,T.C.; EVANS.D D Anzona, Univ. of, Tucson, AZ.

all of the 50 facilities to which it was issued for action. Sir May 1987. 499pp. 8706150101. 41298:131.

piants had breakers which failed to operate satisfactorily dunng in response to high-level radioactive waste repository licens' tests for bulletin requirements. There are no remaining areas of f

ing needs of the U.S. Nuclear Regulatory Commission, this concern for this bulletin.

report examines and provides insights into physical charactens-tics and methodologies for performance assessment of Candi-NUREG/CR-4674 V03: PRECURSORS TO POTENTIAL SEVERE date sites in unsaturated fractured rock. The focus is on the CORE DAMAGE ACCIDENTS:1984,A STATUS REPORT. Main ability of the geologic medium surrounding an underground re.

Report And Appendixes A And B.

MINARICK,J.W.;

pository to isolate radionuclides from the accessible environ.

HARRIS,J.D.; AUSTIN,P.N.; et al. Oak Ridge National Laborato-ment. Media of interest are consolidated rocks with vanable ry.

May 1987. 133pp. 6706160050. ORNL/NOAC-242.

fractunng, rock matnx permeabilities, contained water urider 41309.298.

negative pressure, and air filled voids Temperature gradients Forty-eight operational events, reported in Licensee Event are also of interest Stdes present conceptual and theoretical Reports (LERs) and occumng at commercial light water reac-

Main Citations and Abstracts 13 Lors dunng '?% are considered to be p'ecursors to potential Data and analyses hora a large number of NRC and industry-severs core Gm.ge. Tnese are descnbeu along with associst.

scwored programs have been reviewed and used as a basis ed si(Nficance est<rr,ates categorization, and subse tuent analy-for quantifying the event tree, ie., determining the likelihood of

$1b Th6 study is a cont'nuation of the work presented in earter each pathway for a vanety of accident sequence in,tiators. A volun:es in this senes which evaluated the 1969-i98 f and generalized containment event tree code, called ENVTRE has 198b 9 vents. W ieport wquentially discusses (1) the general been devtdoped to facihtate the cuantification. The uncertenty rationale for th 8 (*u y, (2) the program rnethods for runr w and in the results has been exammed b, performing the Qqntifca-a s

09Cumentat' er. of operational events as precursors (3) the use tios tnree times, using a different sN of input each time to rep-of the conditional probabihty of subsequent severe core damage Wtat the vanation of opinion in the reuctor safety community.

estimates to rank precursor events, and (4) initial conclusions in the so called " central" estimate the likelihood of early con-from the assessment of 1984 W;nts.

tainmen, 'a. lure (occurnng before er at the time of reactor NUREG/CR-4674 VOC PRECURSORS TO 8 OTENTIAL SEVERE vessei trt.ach) was found to be h# for station blackout se-CORE DAMAC6 ACCIDENTc 1984.A STATUS quences but very low for other accident sea'once initiators. Un-REPORT.Appendues C,D And E. M!NAHICK.J W.; HARRi3,J D.;

awmity of igniters and air return fans was the pnncipal AUSTIN.P.N ; et al Oak Ridge National Laboratory May 1987.

mon for the high failure probability for station blackouts. Tho 534pp.8706160086. ORNL/NOAC-232. 41312:196.

ana!ysis also showed that me: ting or bypass of the ice before or See NUREG/C4 4674.V03 abstract wohin a short time after vessel breach can be expocted to o cur mth moderate to high likelihood dunng station blackouts NUREG/CR 4679: OUAi4TITATlVE DATA ON THE FIR 8:: DEHAV-

' aad dming sequences initiated by very small LOCAs with failcre IOR OF COMBUSUE1.E MATEAIALS FOUND IN NUCLEAR of emergency core coohng in the recirculation phase after suc-POWER PLANTS A Literature F.eview. NOWLEN.6 F San 9 cess in the injection phase. This work sinpvts NRC's asses.

National Laboratotes. Fouruary 1987. 166pp. 8706120N 3 i

ment of severe acciden' nsks to be pubt N in NUREG 1G.

SAND 86-0311, 41285 21S.

This report presents the findings of a task in which currently NUREG/CR-4700 V4 DRh CONTAINMENT EVENT ANALYSIS available fire research literature was reviewed for quantitative FOR POSTULATED SEVERE ACCIDENT & GRAND GULF NU-data on the burn:ng charactenstics of combustible materials that CLEAR STATION, UNIT

1. Draft For Corim t.

AMOS.C.N.;

are found in nuclear power plants The matenals considered for KOLACZKOWSKl.A. Sandia National Laboratones. April 1987.

which quantitative data were avulable include cable insulation 448pp,8706120033. SAND 861135 41287:115.

matenals, flammable hquids, furn,ture, tTh and general refuse, A studv has been performed as part of the Sevare Accident and wood and wood products A total v 10 figures and tables, Risk Reduction Program (SARRP) to irnestigate the response taken pnmanly from me referenced wort e wh>ch summanze the of a paricular boiling water reactor with a Mark lli containM *rf available quantitat:ye fire cra.ractenration information for these (Grand Gulf Unit 1) to postulated severe accidents. A det&hed matenals is presented conta:nment event tree for the Grand Gulf plant has been de-NUREG/CR-4681: dNCLOSUAE ENVIRONMENT CHARACTER.

veloped to desenbe the vanoA possible accident pathways that IZATION TESTING FOR THE BASE UNG VAllDATION OF can lead to radioactive releases from containment. Data and COMPUTER FIRE SlMULAllON COOFS NOWLEN S P Sandia analyses from a large numt;er of NRC and industry-sponsored Nabonal Laboratonns. March 198'/.

82pp. 8706150161.

programs have been reviewed and used as a basis for quantify-FAND86-1296. 4130. 04L

'ng the event tree, ie., determining the like;ihood of the path-This report descnbes a serns of fire tt:sts conducted under ways at each branch point for a vanety of accident sequence

'he direction of Sandia National Laboratones for the U.S Nucio-inittators. A generalized containment event trce code, called at Regulatory Cammission. The pnmar/ purpose cf these tests EVNTRF. Nis been developed to facilitate the quetiication.

was to pr wd Gata against which to validate computer fire en.

The unctrteintt n the results has been exancnM 1:y performing i

virorwern urvlation models to be used in the analysis of nucle.

the quarb4.ation three times, using a different sei of input each ar pow!r plant encloture fire situations. Examples of the data time to represent the varia%on of opinion in the reactor safety gathered donng tMo of the tests are presented, thoJgh the pri-community in the so-called " central" etteate, the likelihood mary ob ective of this report is to provide a timely desenption of of early containment failure (occumng before or within a short l

the test effort itseU Those tests were conducted in an encio.

time atte reactor vessel brea-h) was found to be significant be-sure measunng 60M0x;.0 feet. All of the tests utilized forced cause oi th9 possibikty of hydrogen deflagrations or detonations vent <labon conditions typical of nuclear power plant installations.

that can threaten contarment integnty. However, uncertainties A total M 22 lcsts using simple gas burner, heptane pool, meth-surrounding these issues cWe cause the early failure hkelihood anol pool, and MfMA sokd fires was conducted Four of these to be significantly lower than in the central estimate. Further, re-tests wers conducted with a full-scale control room mockup in dioactive releases following containment failure would most place. Parameters venoa dunng testing were fire intensity, en-l'kely be scrubbed by water, which would lower the overall closure ventilation race, and fire location. Data gathered includ.

Source term. This work supports NRC's assessment of severe ed air temperatures, air velocities, radiative and convective heat accident nsks to be pubhshed in NUREG-1150.

flux levels, optical smoke densities. inner and outer enclosure surface temperatures, enclosure surface heat flux levels, and NUREG/CR-4719: COOLABluTY OF STRATIFIED 002 DEBRIS gas concentrations within the enclosure in the exhaust stream.

IN SODIUM WITH DOWNWARD HEAT REMOVALThe D13 Ex-penrnnnt. OTTINGER,C.A.; MITCHELL,G.W.; REED A.W. Sandia NUREG/CR 4700 V2 DRF: CONTAINMENT EVENT ANALYS'S Nationat Laboratones. March 1987. 67pp. 8706150126.

FOR POSTULATED SEVERE ACCIDENTer SEQUOYAH SAND 861043. 41301:127.

POWER STATION, UNIT

1. Draft Report For Comment.

The LMFBR Debris Coolability Program at Sandia National BEHR,V1 ; BENJAMIN.A.S.; KUNSMAN,D.M.; et at Sandia Na-Laboratones investigates the coolabikty of particle beds that tional Laboratones February 1987. 208pp. 8/05200197, may form following a severe accident involving core disassem-SAND 86-1135. 40986.165.

bly in a nuclear reactor. The D senes expenments utilize fission A My has been pe formed as part of the Severe Accident herng of fully ennched UO2 particles submergeo iq sodium tc Risk Reeuction Program (SARRP) to investigate the response reahst'cally almulate decay teating. The D13 experiment is the of a farticular pressunzed water teactot wdh an ice-condenser first in the senos to study the effects of bottom coohng of strati-contai 9nent (Sequeyah Unit 1) to postulated severe accidents.

fied debns, which could be provided in an actual accident condi-A detaHd contanment event tree for the Sequoyah plant has tion bv structural matenals onto which the debn3 might settle.

been di vised to desent;e the variock possible accident path.

Addrtiv a!!y me D13 expenment was designed to achieve maxi-n ways At ca1 lead tu radioactive releasts fro 6 ptainment.

mum amperatures in the debris approaching the melting point

14 Main Citations and Abstracts of UO2. The expenment was operated for over 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and to a specific plant. The major investment in time and manpower

,A irsestigated downward heat removal at specific powers of 0.22 occurs in setting up the base case; changes are comparatively and 2 58 W/g Channeled dryout in the debns was achieved at easy to implement powers from 0 94 to 2.58 W/g Maximum temperatures ap-proaching 2700 degrees C were attained. Bottom heat removal NUREG/CR-4765: MXS CROSS-SECTION PREPROCESSOR was up to 750 kW/m2 as compared to 450 kW/m2 in the D10 USER'S MANUAL. PARKER,F.; LUCNL. Los Alamos National

~

expenment.

Laboratory. ISHIKAWA M FBR Safety Laboratory, Ibaraks Pre-HUREG/CR-4726: EVALUATION OF PROTEC TIVE ACTION tecture, Japan. March 1987. 58pp. 8705190577. LA 10856-M.

40966.217.

RISKS WITZlG.W.F.; SHILLENN.J.K. Pennsylvania State Univ.,

University Park, PA.

June 1987. 000pp. 8707070524.

The MXS preprocessor has been designed to reduce the exe-41600.074 cution time of programs using isotopic cross-section data and to Risk of death and injury due to evacuation was estimated by both reduce the execution time and improve the accuracy of studying 902 evacuahons that occurred in the United States be-shielding-factor interpolation in the SIMMER Il accident anafysis tween January 1,1973 and Apol 30, 1986. The risk of death program. MXS is a duai-purpose preprocessing code to (1) mix due to evacuation is quito small. It was estimated to be equal to isotopes into matenals and (2) fit analytic functions to the self-the nsk of exposure to radiation doses of several hundred to shielding data. The program uses the isotope microscopic neu-about a thousand millirems Key factors for a successful evacu-tron cross section data fram the CCCC standard interface file ation were found to include: an emergency plan, good commu-ISOTXS and the isotope Bondarenko self-shielding data from nications and coordination, practice dolls, and defined authenty-the CCCC standard interkce file BRKOXS to generate cross-Few evacuations used the emergency broadcasting system or section and self-shielding data for materials. The matenals may warning sireris. Reports of panic ar,. traffic jams were very few be a mixture of several isotopes The self-shielding data for the Traffic flow was usually desenbed as light to moder6te. The matenals may be the actual sl,ielding factors or a set of coeffs speed of vehicles at the height of the evacuation was most cients for functions representing the background dependence of commonly reported to be in the 25 to 40 mph range.

the shieldinf, factors. A set of additional data is given to de-NUREG/CR 4739: RAMONA 3B CALCULATIONS FOR BROWNS scnbe the functions necessary to interpolate the shielding fac-FERRY ATWS STUDY. SAHA P.; SLOVIK G.C ; NEW") TIN.L.Y.

tors over temperature.

Brookhaven National Laboratory. February 1987. 120pp.

8705190595. BNL NUREG-52021. 40976:071.

NUREG/CR-4772: ACCIDENT SEQUENCE EVALUATION PRO.

Several aspects of the Anticipated Transient Without Scram GRAM HUMAN RELIABILITY ANALYSIS PROCEDURE.

(ATWS) initiated by an inadvertent closure of all Main Steam SWAIN,A D.

Sandia National Laboratones. February 1987.

Isolation Valves (MSIV) in a typical BWR/4 are analJed in the 167pp. 8706120044. SAND 86-1996. 41186.280.

report. The analysis is performed using the Brookhaven Nation-This document presents a shortened version of the proce-al Laboratory code, RAMONA-3B, which employs a three di-dure, models, and data for human reliability analysis (HRA) monsional neutron kinetics model coupled with a parallel-chan-which are presented in the " Handbook of Human Reliability nel thermal-hydraulics in representing a Boiling Water Reactor Analysis With Emphasis on Nuclear Power Plant Applications" (BWR) Core Four different transient scenanos have been inves-(NUREG/CR 1278, August 1983). This shortened version was tigated: a) downcomer water level and reactor pressure control, prepared and tned out as part of the Accident Soquence Eval-b) manual control rod insertion transient, c) high pressure boil ^

uation Program (ASEP) funded by the U.S Nuclear Regulatory off, and d) recirculation pump top failure. Results of these calcu.

Commission aod managed by Sandia National Laboratones.

lations should provide better understanding of mitigative effects The intent of this new HRA procedure, celled the "ASEP HRA of operator actions dunng ATWS, thus helping in the develop-ment of adequate Emergency Procedure Guidelines (EPG) re-Procedure," is to enable systems analysts, with minimal support quired for the BWR plant safety. A few unresolved questions from experts in human reliability analysis, to make estimates of sub ect to future investigations are also discussed the human error probabilities and other human performance t

charactenstics that are sufficiently accurate for many probabilis-NUREG/CR-4758: A RETRAN MODEL OF THE CALVERT tic nsk assessments. The ASEP HRA Procedure consists of a CLIFFS-1 PRESSURIZED WATER REACTOR FOR ASSESS-Pre-Accident Screening HRA, a Pre. Accident Nominal HRA, a ING THE SAFETY IMPLICATIONS OF CONTROL SYSTEMS.

Post Accident Screening HRA, and a Post Accident Nominal RENIER,J A.; SMITH,0 L.

Oak Ridge National Laboratory HRA The procedJre in this document includes changes made March 1987.116pp. 8706030092 ORNL/TM-10236. 41161:126.

after tryout and evaluation of the procedure in four nucicar The failure modo and effects analysis of Calvert Cliffs 1 iden-power plants by four diffarent systems analysts and related per.

tified sequences of events judged sufficiently complex to ment further analysis in detailed dyNmic simulations. This report de.

sonnel, including human reliability specialists. The changes con-scnbes the RETRAN model developed for this purpose and the sist of some additiona! explanatory matenal (including exam-results obtained. The mathematical tool was RETRAN/ Mod 3, ples), and more detailed definitions of some of the terms.

the latest version of widely used and extensively validated ther-NUREG/CR-4773: DESIGN FEATURES TO FACIUTATE INTER-mal-hydraulic production code RETRAN2 is based on a first.

NATIONAL SAFEGUARDS AT MIXED OXIDE CONVERSION pnnciples methodology that treats two phase flow with slip.

FACILITIES. HARMS.N.L; ROBERTS.F.P. Battelle Memonal in-Thermal equilibnum of phases is assumed except in the pres-stitute, Pacific Northwest Laboratones June 1987, 44pp sunzer, where non.equilibnum processes are important and spe-8706300252. PNL 5894. 41491.001.

cial methodology is usect Heat transfer in solids is obtained from the conventional conduction equation. Point or 1 D kinetics This study for the Nuclear Regulatory Commission identifies is available for the reactor core The fundamental methodology and analyres facility designs that can facilitate International is supplemented with a broad list of process submodels that Atomic Energy Agency safeguards for mixed plutonium uranium calculate heat transfer coefficients, fluid and metal state proper.

oxide (MOX) conversion plants A baseline facility is defined and ties, choked flow, form and wall fnction losses, and other pa.

the implementation of safeguards is analyzed Areas are identi-rameters Also supplied are component submodels for vanous fred for which special facility design considerations can facilitate types of valves and pumps, the latter of which incorporate four-IAEA inspections for timely detection of possible diversion of quadrant charactenstics for components in which two-phase or nuclear matenal. Design features are proposed to enhance in-reverse flow may be expected, and heat versus flow curves for spection capabilities for venfication of nuclear matenal flows others Extensive input allows the code to be highly particultzed and inventones

Main Citations and Abstracts 15 NUREG/CR-4779: NEW DATA FOR AEROSOLS GENERATED NUREG/CR-4802: AN EVALUATION OF TRAC-PF1/ MODI COM-BY RELEASES OF PRESSURIZED POWDERS AND SOLU-PUTER CODE PERFORMANCE DURING POSTTEST SIMULA-TIONS IN ST ATIC AIR BALUNGER.M Y.; SUTTER,S L.;

TIONS CF SEMISCALE MOD-2C FEEDWATER LINE BREAK HODGSON.W H. Battelle Memonal inst;tute, Pacif,c Northwest TRANSIENTS. HALL.D.G ; WATKINS.J C. EG&G Idaho, Inc.

Laboratones. May 1987. 51pp 870624004S. PNL-6065 (subs. of EG8G, Inc ) February 1987. 204pp. 8704270202.

41448.248 EGG-2486 40681:173 Safety assessments and environmental irnpact statements for This report documents an evaluation of the TRAC-PF1/ MOD 1 nuclear fuel cycle facilities require an estimate of potential air-reactor safety analysis computer code dunng computer simula-tions of feedwater line break transients. The expenmental data borne releases. Aerosols genersited by accrdents are being in-base for the evaluation included the results of three bottom vestigated by Pacific Nortnwest Laboratory to develop radioac-feedwater line break tests performed in the Semiscale Mod-2C tive source-term estimation methods. Expenmer ts rneasunng test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11),

the mass airborne and particle size distnbution of aerosols pro-and 100% (S-FS-68) breaks. The test facihty and the TRAC-duced by pressunzed releases were run Carbon dioxide was PF1/ MODI model used in the calculations are desenbed. Eval-used to pressunze uranine solutions to 50, 250, and 500 psig uations of the accuracy of the calculations are presented in the before release. The mass airborne from these expenments was form of compansons of measured and calculated histones of higher than for comparable air-pressunzed systems, but not as selected parameters assaiated with the pnmary and secondary great us expected based on the amount of gas dissolved in the systems. In addition to evaluating accuracy of the code calcula-hquid and the volume of liquid elected from the release equip-tions, the computational performance of the code dunng the ment. Flashing sprays of uranine at 60,125, and 240 psig pro-s,mu!ations was assessed. A conclusion was reached that the duced a much larger source term than all other pressunzed re-code is capable of making feedwater line break transient calcu-leases performed under th's program low-pressure releases of lations efficiently, but there is room for significant improvements depleted uranium dioxide at 9,17.5, and 24 5 psig provided in the simulations that were performed. Recommendations are data in the energy regiun bewoon 3.m spills and 50-psig pres-made for follow. on investigations to determine how to improve sunzod releases.

future feedwater line break calculations and for code improve-ments to make the code easier to use.

NUREG/CR 4783: ANALYSIS OF BALANCE-OF-PLANT REGU-LATORY ISSUES Final Report.

LAY,R.;

ETTLINGER,L; NUREG/CR-4814: SOURCES OF CORRELATION BETWEEN SETH,S.; et al Mitre Corp June 1987. 157pp 8706160138.

EXPERTS Empincal Results From Two Extremes. MEYER M.A.;

MTR 66 WOO 213,41310f71.

BOOL ER.J M. Los Alamos National Laboratory. Apol 1987.

61pp 8705190636. LA-10918-MS. 40977:256.

The MITRE Corporation, under contract to the U.S Nuclear Through two studas, this report seeks to identify the sources Regulatory Commission (NRC), has examined certain regulatory and performance aspects of the conventional, or power conver-Expert estimates are relied upon as sources of data whenover sion, side of nuclear plants, often referred to as the Balance-of-expenmental data is lacking such as in nsk analysis and rehabil-Plant. This report includes MITRE's charactenzauon and analy.

sty assessments Correlation between experts is a problem in sis of the Balance-of Plant failures, a perspective on the safety the ehcitation and subsequent use of subjective estimates. Until significance of these failures, a descnption of current NRC ac-now, there has been no data confirming sources of correlation, tivities and industry initiatives that have the potential to reduce although the experts' background is commonly speculated to be these failures; and the formulation of a set of recommendations one. Two different populations of experts were administered for NRC's consideration in addressing the safety is ues raised questions in their arcas of expertise. Data on their professional t'y the Balance of-Plant. In general, the Balance-of-Plant prob-backgrounds and means of solving the questions were elicited lems represent the most frequent reason for unanticipated plant using techniques from educational psychology and ethnography.

shutdowns, and they compromise safety. The NRC needs to The results from both studies indicate that the way in which an look at the total power plant facihty as a system, and provide expert snives the problem is the major SJurce of Correlation.

the same level of attention to reducing challenges to the plant The experts' background can not be shown to be an important safety systems as it does to responding and mitigating those source of correlation nor to influence his choice of method for challenges MITRE believes that this can be accomphshed problem solution. From these results, some recommendations whhin the context cf NRC's present regulatory posture am given for the ekcitation and use of expert opinion.

NUREG/CR-4800: SIGPI A USER'S MANUAL FOR FAST COM-NUREG/CR-4815: DEMONSTRATION TESTING OF A SURVEIL-LANCE ROBOT AT BROWNS FERRY NUCLEAR PUTATION OF THE PROBABILISTIC PERFORMANCE OF AEAnahsis Of Costs And Benefits.

WHITE.J.R.;

COMPLEX SYSTEMS PATEN AUDE.C.J Lawrence Livermore H

A et at how hcWogy National Laboratory May 1987 75pp. 8706150086. UCID-Corp. March 1987. 82pp. 8704270352. 40704.117.

20679 41304 154 This report presents the results of an NRC project to deter.

The SIGPI program computes the probabihty of complex sys-mnie whether robotics equipment can be cost effective in per.

tems as defined by cut sets or other binary product sets. The forming surveillance and inspection work at existing r'uclear SIGPl program uses two fast complementary methods of com-power plants. A mobile surveillance robot, called SURBOT, was puting the probacihstic performance of complex systems: the !!

developed by the Remote Technology Corporation (REMOTEC) method and the sigma method The li algonthm exploits the fact to perform visual, sound, and radiation surveillance within rooms that carefully defined system components are often statistically designated as radiologically hazardous. SURBOT was tested in independent conditional to the environment in which they are the turbine building of the Browns Ferry Nuclear Plant (BFNP) embedded The sigma algonthm computes the probabikty of by TVA personnel for a five-month penod. The results showed combinations of components produced by the il algonthm by that SURDOT obtains higher quakty data and can perfoon more disjointing and partitioning such components, thereby allowing thorough surveillance within radiation areas than workers wear-the exact computation of performance. The program assumes ing protective clothing. SURBOT can be transferred between input of up to three data types cut set data in disjoint normal rooms without releasing contamination in the hallways using a form, basic compormnt probabiktres for independent basic com, portable enclosure. TVA has estimated that over 100 person-ponents, and/or moan and covanance data for statistically de.

rem exposure and $100,000 operating costs can be saved an-pendent basic components nually at the BFNP using SURBOT for surveillance in 54 turbine and reactor building rooms TVA recornmendations for improv-

16 Main Citations and Abstracts ing the function, rehabihty, and maintainability have been incor-NUREG/CR-4841: FRACTURE EVALUATION OF SURFACE potated into a production model of SURBOT which is now com.

CRACKS EMBEDDED IN REACTOR VESSEL mercially available from HEMOTEC aiong with other types of CLADDING Unirradiated Bend Specimen Results. MCCABE.D E.

mobsle robots and manipulators Materials Engineenng Associates, Inc. May 1987. 70pp.

N'JREG/CR-4819 V01: AGING AND SERVICE WEAR OF SOLE-8705W063L MEA-2200 409W8L NOID-OPERATED VALVES USED IN SMETY SYSTEMS OF The surface crack embedded in the clad layer of a reactor NUCLEAR POWER PLANTS Volume 1 Operating Exponence vessel has been identified as a cntical fracture safety assess.

And Failure identification BAC/ NSKAS,V P.; ROBERTS.G C.;

ment condition relative to the pressurized thermal shock acci-l TOMAN,G J ; et al Oak Ridge Na'ional Laboratory. March 1987.

dent scenano This project was initiated to determine the severi-69pp 8706030204. 41147 287.

ty f s;ch cracks expenmentally, using irradiated matenal, and An assessment of the types and uses of solenoid operated to identify the matenal property and stress conditions in the valves (50V) in nuclear power plant safety-related service is local region of the crack that are significant to the analysis.

provided. Through a desenption of each SOV's operation, com.

Bend bar tests provide the experimental simulation of the sub-bened with knowtudge of nuclear power plant apphcations and ject RPV ;'urface crack. This report desenbes the initial investi-operational occurrences, the significant stressors responsible gation using unirradiated matenal, addresses the analysis tech-for degradation of SOV performance are identified A review of naques, and presents the findings indicated by the expent" ental actual operating expenence (failure data) leads to identification results. Charactenzation of the irradiated material will be pre-of potential nondestructive in-situ testing which, if property de-sented in a subsequent report.

veloped, could provide the methodology for deterioration moni-tonng of SOVs. Recommendations are provided for continuation NUREG/CR-4842: A STUDY OF NATURAL GLASS ANALOGUES of the study into the test methodology development phase.

AS APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS.

BYERS.C D.; JERCINOVIC.M.J.; EWING,R.C. Argonne National NUREG/CR-4821: REACTOR COOLANT PUMP SHAFT SEAL Laboratory. February 1987. 165pp. 8705200253 ANL-86-46.

"TABILITY DURING STATION BLACKOUT. RHODES,D B.;

40984:161.

HILL R C ; WENSEL,R G AECL, Chalk River Nuclear Laborato-A two-part study was undertaken on the alteration of natural nes May 1987. 60pp 8706120189 AECL 9342. 41187.201.

basalt and nuclear waste glasses. In the first part, the University Results are presented from an investigation into the behavior of New Mexico charactenzed a wide variety of natural basaltic of Reactor Coolant Pump shaft seals dunng a potential ttation glasses with respect to.aaction products, reaction kinetics, and blackout (loss of all ac power) at a nuclear power plant. The in-geologic history. The important outcome of this study was a de-vestigation assumes loss of coohng to the seals and focuses on senption of the process whereby natural glass alters to palagon-the effect of high temperature on polymer seals located in the ite and authigenic materials, including clays and teolites. In the shaft seal assembhes. and the identification of parameters second part ANL performed laboratory tests to simulate the having the most influence on overall hydraulic seal perform-natural alteration process with a synthetic basaltic glass. For ance Predicted seal failure thresholds are presented for a laboratory tests in which the glass samples were exposed to range of station blackout conditions and shaft s(!al geometnes-water vapor at high temperature (> 90 degrees C), a close sim-NUREG/CR-4822:

BROAD BAND SEISMIC DATA ilanty was found in the alteration of natural and synthetic glass-ANALYSIS September 1984 September 1986 CARTER,J.A.;

es The same alteration process was ivund for a nuclear waste BARSTOW,N ; SUTTON.G H.; et al. Rondout Associates, Inc.

glass (SRL 165). It was concluded that both the naturaf alter.

Apol 1987.100pp. 8704270184 40682:234.

ation of basaltic glass and water-vapor laboratory tests can be This report contains a detailed desenption of the SRNY sta.

used as an analogue to assess the performance of nuclear tion including its response function, desenption of the tech-waste g; ass under potential repository conditions.

niques and software used to analyze the data, and evaluations of both the station and processing methods. The station was NUAEG/CR-4845: AN ANALYSIS OF THE SEMISCALE MOD-2C evaluated through noise studies under both quiet and no:sy con.

S-NH-3 TEST USING THE TRAC-PF1 COMPUTER PROGRAM.

dations, its detection and location thresholds, and the frequency DRISKELL.W E.; KULLBERG,C.M. EG&G Idaho, Inc. (subs of duration, and causes of data loss The report has been divided EG&G, Inc ).

March 1987. 45pp. 8706030180. EGG-2496.

Into three main sections The first section gives a desenption of 41144.203.

the broad band digital seesmic station SRNY installed near the A calculation was performed using the TRAC-PF1/ MODI Rondout Associates, incorporated (RAI) offices in Stone Ridge, computer prog am to simulate a small break, loss-of coolant ex-NY. Included in the discussion are the system response func.

penment where the high-pressure injection was not used to miti-tions and an analysis of the causes of data loss The cocond gate the fuel rod temperature excursion. This expenment, desig-section gives details of the data analysis methods used in nated the S-NH-3 Test, simulated a 0.5% cold-leg break in a studying broad band and array data Although several methods PWR and was one of a senes of tests conducted in the Semis-have been studied, much of this section is devoted to the cale Mod-2C test facility. The pnmary purpose for doing the cal-adaptive polanzation method which has shown promise as a culation was to evaluate the capability of the TRAC PF1 code to single station incation tool The final section examines the noise calculate the thermal-hydrauhc response observed in the experi-charactenstics at SRNY dunng both quiet and noisy conditions ment. The evaluation employs the companson of selected code-and compares these leveis to the Regional Seismic Test Net-calculated system responses with the test data. Conclusions work station RSNY in the Adirondack Mountain region of north-and recommendations on improving the quakty of the calcula-ern New York. The detection and location capabilities of SRNY tion are included.

are also desenbod and single-station locations at SRNY, as well as RSNY, are compared to network locat:ons NUREG/CR-4848:

STEAM GENERATOR GROUP PROJECT. Annual Report 1985. KURTZ,R.J.; LEWIS M.;

NUREG/CR-4830: MELCOR VALIDATION AND VERIFICATION CLARK.R.A. Battelle Memonal Institute Pacific Northwest Lab-1986 PAPERS. LEIGH.C.D Sandia National Laboratones.

oratones April 1987.

122pp. 8704270138.

PNL-5771.

March 1987. 223pp. 8706120166. SAND 86 2689 41281:049 40681:051.

MELCOR validation and venfication resuits from 1986 are This report is a summary of the Steam Generator Group presented. Results of compansons to analytic solutions and ex.

Project progress for 1985. Statistical analyses of data fron non-penments are included. The major areas tested in these com-destructive examination (NDE) round robins performed on the pansons are the control volume hydrodynamics and thermody-Surry generator are presented. Cntena are hsted for selection of namics, the heat transfer and the aerosol behavior in MELCOR.

tube specimens to be removed from the generator for vahdation A set of nine standard tests is included of the NDE round robin results. A samphng plan is desenbed

Main Citations and Abstracts 17 along with the initial steps taken to implement this plan. Special results show that the SC.TNP method tends to overestimate the tooling fabncated for specimen removals is discussed. Removal maximum loads by 15 percent on the average whereas the of three 9-tutie sections of tube sheet is presented. Preliminary SC.TKP method tends to underpredict the maxirr:um loads, as results from destructive analysis of one section are reported.

desired, by 32 percent.

Validation activities were initiated by metallurgical evaluation of two tubes pulled from the hot leg tube sheet Results from de-NUREG/CR-4875: CHARACTER!ZATION OF CRUSHED TUFF structive measurement of maximum defect depth from these FOR THE EVALUATION OF THE FATE OF TRACERS IN two tubes indicated that in situ bobbin coil eddy current meas.

TRANSPORT STUDIES IN THE UNSATURATED ZONE.

urements consistently undersized the defects.

POLZER W.L.;

FUENTES.H R.; RAYMOND,R.; et al. Los Alamos National Laboratory. March 1987, 46pp. 8704280288.

NUREG/CH-4866: AN ASSESSMENT OF HYDROGEN GENERA-LA-10962-MS. 40693:324.

TION FOR THE PBF SEVERE FUEL DAMAGE SCOPING AND Results of field-scale (caisson) transport studies under un-11 TESTS CRONENBERG.A W; MILLER.R W.; OSETEK.DJ saturated moisture and steady and nonsteady flow conditions EG&G Idaho, Inc. (subs of EG&G. Inc ). April 1987. 95pp' 8706120168. EGG 2499. 41281:272.

indicate vanability and a lack of conservation of mass in solute transport. The tutt matenals used in that study were analyzed An evaluation of zircaloy oxidation and hydrogen generation for the presence of tracers and of freshly precipitated matenal data es presented for the first two severe fuel damage (SFD) tests SFD-ST and SFD 11, conducted in tho PBF at the INEL.

to help explain the variability and lack of conservation of mass.

The report presents an assessment of data in terms of the influ-Selectt,J tuff samples were charactenred by neutron activation ence of zircaloy melting on oxidation behavior and fuel bundle analysis for tracer identification, by x-ray diffraction for mineral reconfiguration effects which may alter steam flow and hydro-identification, by petrographic analysis for identificaen of fresh-gen generation charactenstics. A comparison of the H(2) gen-ly precipitated matenal, and by x ray fluorescence analysis for cration and cladding thermocouple data indsCates that a signifi-identification of major and trace elements. The results of these cant amount of hydrogen was produced after the initiation of zir-analyses indicate no obvious presence of freshly precipitated caloy melt-induced fuel d:ssolution (greater than or equal t matenal that would retard tracer movement. The presence of 2150 K) Posttest metallographic observations corroborate the the nonsorbing tracers (bromide and iodide) suggests the reten-trend of the on-line data. Analyses hiso indicate that essentially lion of these tracers in immobile water. The presence of sorbing complete flow area blockage (> 98%) would be required to de and nonsorbing tracers on the tuff at some locations (even at minish steam flow through the degraded test bundle, reducing the 415-cm depth) and not at others suggests variability in hvdwogen production. Neither on-line data nor posttest examina, transport.

tion of the SFD-ST and SFD 11 fuel bundles indicates that NUREG/CR-4877: ASSESSMENT OF DESIGN BASIS FOR such extreme flow area blockages occurred. For the steam-nch LOAD CARRYING CAPACITY OF WELD-OVERLAY REPAIRS.

SFD-ST expenment, UO(2) fuel oxidation was also observed' possibly accounting for approximately 20% of the total hydro-SCOTT,P.M. Battelle Memonal Institute, Columbus Laboratones.

gen production. Fuel oxidation has also been noted from re.

Apnl 1987. 81pp. 8704280032. BMI-2150. 40709.171.

tneved TMI-2 core debns samples. Thus, oxidation of UO(2) t This study was conducted to assess the current load carrying a hypostoichiometnc condition may add to the total hydrogen capacity design basis for weld-overlay repairs (WORs). Although burden for severe accic " 's.

not specifically addressed in it, the design of WOHs is in the s int of the ASME Boiler and Pressure Vessel Code,Section XI, NUREG/CR-4872: EXPEH! MENTAL AND ANALYTICAL ASSESS-Article IWB-3640. NUREG-0313 Revision 2 provides guidance MENT OF CIRCUMFERENTIALLY SURFACE-CRACKED PIPES for the implementation of the procedures outhned in IWB-3640.

UNDER BENDING. SCOTT,P M.; AHMAD J Battelle Memonal How9ver, neither of these documents specifies the values for Institute, Columbus Laboratones. April 1987.

167pp.

diameter or thickness which are to be used in the WOR design 8705190623. BMI 2149. 40978.152 analysis. Throughout this report we have used the combined This study was performed to assess the validity of varioes thickness and diameter of the repaired cross section to calcu-techniques to predict maximum loads for circumferentially sur-late the membrane and bending stresses The maximum stress face. cracked pipes under bending Expenmental data were de-from each of the four WOR pipe expenments conducted was veloped for both carbon steel and stainless steel pipes. Predic-significantly higher than that predicted by tha IWB-3640 analysis tions of maximum loads were made using the net-section-col-for the design guidelines set forth in NUREG-0313 Revision 2 lapse method. the IWB-3640 analysis procedures, and a newly for a " Standard" overlay; the average failure stress was ap-developed finite-length surface-cracked pipe J-estimation proximately 30 percent higher than the IWB-3640 predicted fail-method. The net-section collapse method gave good maximum-ure stresses. These values do not include the Code safety fac-load predictions for certain types of pipe. However, for pipes tors on stress. For a " Standard" overlay, the flaw size consid-with large radius to thickness (R(m)/t) ratios and/or low tough-ered in the analysis is completely through the original pipe wall ness, this analysis method tended to overpredict the expemen-for the entire circumference of the pipe. Such an overlay would tal maximum load. % plastic-zone screening entenon was devel-be suitable for long. term plant opntion. If actual flaw dimen-oped to show when this method was valid and when elastic-sions were used in the design analysis, then in two of the four plastic fracture machanics should be used. The limit-load proce-experiments the maximum stress was less than that predicted dures embodied in lWB 3640 provide the desired underpredic-by the IWB-3640 Source Equations. The greatest difference tion of the failure stress. The average failure stress for the nine was 8 percent. For the flaw sizes evaluated in this study, an stainless steel base metal expenments was 61 percent higher overlay design based on actual flaw dimensions would be con-than predicted by Table IWB-3641 1 and 23 percent higher than sidered a " Limited Service" overlay, suitable only for short-term predicted by the Source Equations. For the three stainless steel plant operation, not to exceed one fuel cycle, flux weld expenments the predicted failure stresses were adjust-ed by a stress multiplier to account for the lower toughness of NUREG/CR-4878: ANALYSIS OF EXPERIMENTS ON STAIN-the flux welds. The average failure stress for the flux weld ex-LESS STEEL FLUX WELDS Topical Report. WILKOWSKI,G.;

penments was 78 percent higher than predicted by Table IWB-AHMAD J.; BRUST,F.; et al. Battelle Memona! Institute, Colum-36415 and 39 percent higher than predicted by the Source bus Laboratories. Apnl 1987.187pp. 8705190355. BMI-2151.

Equations. Predictions from two versions of the new finite-length 40975:137.

surface-cracked pipe J-estimation method were compared to This report describes experimental and analytical efforts to expenmental results. One version is for pipes with large R(m)/t evaluate fracture of stainless steel flux-welded pipe. Seven pipe estios (SC.TNP) while the other is a more general approach fracture expenments (four with through-wall circumferential (SC TKP) where the large R(m)/t ratic restnction is relaxed. The cracks and three with circumferential intemal surface cracks)

18 Main Citations and Abstracts were conducted at 550 degrees F (288 degrees C) Matunal mol% had no effect on oxidation kinetics The rate-controlkng charactenzation efforts involved laboratory specimen tests to factor appears to be diffusion through the oxide layer. Finally, assess specimen stze effects, effects of solution-annealing. and the ondation kinetics of prereacted Zircaloy 15 mol% UO(2) cracl-growth behavior in the HAZ, along the fusion line, and in were measured at 1400 to 2150 degrees C. The rates were the weld metal. Efforts involved assessir>g the net-section col-comparable to those obtained for Zirceloy above 1500 degrees lapse analysis, the plastic-zone screen rig cntenon, inherent C.

safety margins in the IWB-3640 flux weld analysis, through-wall-NUREG/CR-4890: HEAT OF REACTION OF MOLTEN ZlRCONI-cracked pipe predictive J-estimation schemes for LBB analyses, UM WITH 002. BRIMHALL.) L.; PRATER,J T. Battelle Memonal n-factor J-R curves calculated from the pipe expenments for institute. Pacific Northwest Laboratones. April 1987. 32pp.

companson to C(T) specimen results, and finite element analy-8705120091. PNL-6165. 40896.225.

sis of C(T) specimens and one pipe expenment. This report also The heat of reaction for the dissolution of UO(2) by motten evaluates the technical significance of these results and their C q me significance relative to licensing decisions I

taal thermal analysis The dissolution of 25 wt% UO(2) by zirco-NUREG/CR 4883: REVIEW OF RESEARCH ON UNCERTAIN-nium was determined to be an exothermic reaction with a heat TIES IN ESTIMATES OF SOURCE TERMS FROM SEVERE release of approximately 20 kcal/mol UO(2). Expenmental diffi-ACCIDENTS IN NUCLEAR POWER PLANTS. KOUTS.H. Brook-cutties eruounterr;d at these high temperatures precluded a pre-haven National Laboratory April 1987. 106pp. 8705190524-cise determination of the heat of react;on.

BNL-NUREG-52061. 40977:081.

A review has been undertaken by four panels of experts, og NUREG/CR-4b94: A USER'S GUIDE TO THE NRC'S PIPING the sources of uncertainty in source terms from accidents to nu-FRACTURE MECHANICS DATA BASE (PlFRAC). HISER,A.L.;

clear power plants as presented by the document NUREG.

CALLAHAN.G M. Matenals Engineenng Associates, Inc. May 0956. These panels contained eminent scientists from the 1987. 82pp. 8706030146 MEA-2210. 41164.278.

I I

United States, the United Kingdom and the Federal Republic of This Guide is the reference for use of the NRC's Piping Frac-ture Mechanics Data Base (PIFRAC), a computenzed data base Germany. Separate reports by the panels provide detailed d5 cussions and conclusions regarding the uncertainties and the containing the material s property data for steels used in nucle-s NRC research programs for their resolution. An overati summary ar power plant piping. These data are for use by NRC regulators of the results of panel deliberation is also given.

as required for structural integnty assessments of safety mar-gins in nuclear piping. The data of pnmary utility in such assess-NUREG/CR-4885: SEISMIC HAZARD CHARACTERIZATION OF ments are fracture toughness (pnncipally J R curve) and tensile THE EASTERN UNITED STATES Comparative Evaluation Of (stress-strain) data, but other charactenstic data such as chemi-The LLNL And EPRI Studies BERNREUTER.D L.; SAVY,J B.;

cal composition and Charpy V data are also provided where MENSING.R W. Lawrence Livermore National Laboratory. May available This Guido desenbes PIFRAC as presently config-1987. 289pp. 8706160074. UCID-20696. 41311'267 ured. Revisions, updates, and coi,ections to PIFRAC will require in 1982, the Lawrence Livermore National Laboratory (LLNL) revisions to this Guide.

was funded by the U S. Nuclear Regulatory Commission (NRC) to develop a methodology to charactenze the se:smic hazard NUREG/CR-4899: COMPONENT FRAGILITY RESEARCH for all sites of the eastern United States (EUS) nast of the PROGRAM. Phase l Component Pnontization. HOLMAN G.S.;

Rocky mountains The utility sponsored Electnc Power Re.

CHOU C K.

Lawrence Livermore National Laboratory. June search Institute (EPRI) followed suit in late 1983 with a similar 1987.177pp. 8706300007. UCID 21003. 41491:118.

study The LLNL methodology was applied at 10 test sites of Current probabilistic nsk assessment (PRA) methods for nu-the EUS and the results reported in 1985 (LLNL Report UCIO.

clear power plants utilize seismic " fragilities" - probabilities of 20421) The EPRI study was presented in a senes of draft re.

failure conditioned on the seventy of seismic input motion - that ports in 1985 under the project number P101-29. The purpose are based largely on limited test data and on engineenng judg-of this study was to help in understanding the reasons for differ.

ment Under the NRC Component Fragihty Research Program ences in results between the LLNL and EPRI study We first in-(CFRP), the Lawrence Livermore National Laboratory (LLNL) vestigated possible differences in the theones and assumptions has developed and demonstrated procedures for using test data used to develop the hazard models and concluded that all to denve probabilistic fragility desenptions for mechanical and inpu* poing equal, the two methods were essentially equiva-electncal components As part of its CFRP activities, LLNL sys-lent We analyzed the vanous input parameters, their values and thematically identified and categorized components influencing the way they were collected, and finally we performed sensitivity plant safety in order to identify " candidate" components for analysis. The three main differences were found to be (1) the future NRC testing Plant systems relevant to safety were first lower bound magnitudes of integration, (2) the ground motion identified, within each system components were then ranked ac-models, and (3) the fact that LLNL accounted for a local site cording to their importance to overall system function and their correction and EPRI did nest.

anticipated seismic capacity. Highest prionty for future testing

• • • assigned to those "very important" components having NUREG/CR-4889: ZlRCALOY-4 OxtDATION AT 1300 TO 2400 DEGREES C. PRATER.J T.; COURTRIGHT,E L. Battelle Memo-I w.. seismic capacity. This report desenbes the LLNL pnonti; zation effort. which also included application of high-level nal Institute, Pacific Northwest Laboratones. Apnl 1987.41pp quahfication data as an alternate means of developing probabi-8705120087. PNL-6166. 40896.034 listic fragility desenptions for PRA applications.

The oxidation kinetics of Zircaloy-4 in steam have been ex.

tended to 2400 degrees C. The ZrO(2) and -Zr layers display NUREG/CR-4901: EFFECTS FROM INFLUENT BOUNDARY parabolic growth behavior over the entire temperature range CONDITIONS ON TRACER MIGRATION AND SPATIAL VARIA.

studied. A discontinuity in the oxidation kinetics at 1510 degrees BILITY FEATURES IN INTERMEDIATE-SCALE EXPERIMENTS.

C causes rates to increase above those previously established FUENTES.H R.; POLZER.W.L.; SPRINGER,E.P. Los Alamos by the Baker Just relationship. This increase coincides with the National Laboratory. Apnl 1987.131pp. 8706030186_ LA-10981-tetragonal-to-cubic phase transformation in ZrO(2 x) No addr MS 41144 072.

tional discontinuity in the oxide growth rate was observed when in previous unsaturated transport studies at Los Alamos, dis-l the metal phase metted The effects of ten perature gradients persion coefficients were estimated to be higher close to the were taken into account, and corrected values representative of tracer source than at greater distances from the source InjeC-near esothermal conditions were computed. Oxide growth was t.on of tracers through dtscrete influent outlets could have ac-also measured in vanous steam-hydrogen mixtures at 1565 de-counted for those higher dispersioi s. Also, a lack of conserva-grees C and 1815 degrees C. Hydrogen concentrations up to 90 tion of mass of the tracers was observed and suspected to be O

Main Citations and Abstracts 19 due to spatial vanability in transport. In the present study, ex-NUREG/CR-4903 V03: SELECTION OF EARTHOUAKE RESIST-penments were performed under uniform influent (ponded) con.

ANT DESIGN CRITERIA FOR NUCLEAR POWER PLANTS -

ditions in which breakthrough of tracers were monitored at four METHODOLOGY AND TECHNICAL CASES. Dislocation Models locations at each of four depths. All other conditions were simi-Of Near Source Earthquake Giound Motion. A Review.

• Struc-lar to those of the unsaturated transport expenments. A com.

tural & Earthquake Engineering Consultants. LUCO JE. Califor-panson of results from these two sets of expenments indicates naa. Univ. of, San Diego, CA. May 1987,177pp. 8706120158.

differences in the parameter estimates Estimates were made 41288:203.

for the dispersion coeffic$nt and the retardation factor by the The solutions available for a number of theoretical fault one-dimensional steady flow computer code, CFITIM. Estimates models are examined in an attempt at establishing some of the were also made for mass and for velocity and the dispersion expected charactenstics of earthquake ground motion in the coefficient by the method of moments. The dispersion coeffi.

near-source region, in particular, solutions for two dimensional cient oecreased with depth under discrete influent application anti-plane shear and plane-strain models as well as for three-and er. creased with depth under ponded snfluent application. Dif, dimensional fault models in full space, uniform half-space and ferences in breakthroughs and in estimated parameters among layered half-space media are reviewed.

locations at the same depth were observed under ponded influ-NUREG/CR-4912: DATING GROUND WATER AND Tl:E EVAL.

ent application. Those differences indicate that there is a lack of UATION OF REPOSITORIES FOR RADIOACTIVE WASTE.

conservation of mass as well as significant spatial vanability DAVIS,S.N.; MURPHY,E. Arizona, Univ. of, Tucson, AZ. April across the expenmental domain-1987.197pp. 8/05190537. 40977:315.

NUREG/CR 4903 V01: SELECTION OF EARTHOUAKE RESIST-The age of ground water is the length of time that the water ANT DESIGN CRITERIA FOR NUCLEAR POWER PLANTS has been isolated from the atmosphenc portion of the hydrolog.

METHODOLOGY AND TECHNICAL CASES. Direct Empincal ic cycle. It is a theoretical concept only, because all ground Scaling Of Response Spectral Arrplitudes From Vanous Site water to some extent is a mixture of waters of different ages. In And Earthquake Parameters. LEE.V W.; TRIFUNAC,M.D. Struc-simple systems, however, relative ages of ground water from different portions of an aquifer can be determined by different tural & Earthquake Engineenng Consultants. May 1987.343pp.

8706120186. 41284.232.

methods, and the dates obtained are commonly in accord with each other and reflect systematic increases of water ages in New frequency dependent attenuation function of Fourier am-downgradient directions. At least nine independent methods can plitude spectra of recorded strong earthquake ground accelera-be used to approximate ground-water ages. The methods vary tion has been developed. The iterative regression analyses widely in precision but all give useful information. In complex assume simple functional forms to model the trends of the data ground-water systems, as many dating methods as possible and have sufficient flexibility to detect dependence of attenu-should be used. Discordant " dates

• will result which, when ation on source dimensions, depth and frequency of wave properly interpreted, will not give a single water age but will give motion. It has been found that for distances less than about valuable information concerning the hydrodynamics of the 100 km there is clear frequency dependent vanation of attenu-ground-water system. Dating methods which use isotopic and atson functions, with high frequency amplitudes attenuating other geochemical techniques will read the actual history of the faster with distance. Our previous empincal scaling model for water and will give direct information on average ground-water Pseudo Relative Velocity (PSV) spectrum amplitudes has been conditions over long penods of time. If these penods exceed refined by introducing this new frequency dependent attenuation several hundred years, geochemical methods which use past of amplitudes with distance. The new model also considers the conditions to predict the future are superior to hydrodynamic depth of earthquake focus and the approximate charactenzation methods which use an extrapolation of short-term data to pre-of the source site to compute the " representative" source to dict long-term hydrogeologic conditions.

station distances in addition to all other scaling parameters used previously.

NUREG/CR-4918 V01: CONTROL OF WATER INFILTRATION INTO NEAR SURFACE LLW DISPOSAL UNITS. Annual NUREG/CR-4903 V02: SELECTION OF EARTHOUAKE RESIST.

Report, October 1985 - September 1986. SCHULZ,R.K.;

ANT DESIGN CRITERIA FOR NUCLEAR POWER PLANTS -

RIDKY,R.W.; O'DONNELL,E.; et al. Califomia, Univ. of, Berke-METHODOLOGY AND TECHNICAL CASES, Methods For intro.

ley, CA. April 1987. 30pp. 8705120092. 40896:196.

duction Of Geological Data into Characterization Of Active in the humid eastern part of the United States, trench covers Faults And Seismicity And...

ANDERSON J G.; LEE,V.W.;

have, in general, failed to prevent some of the incident precipi-TRIFUNAC M.D Structural & Earthquake Engineenng Consult.

tation from percolating downward to buried wastes. It is the pur-ants. May 1987.197pp. 8706120163. 41289:020.

pose of ths present work to investigate and demonstrate a pro-This report reviews the physical and expenmental bases for a cedure or technique that will control water infiltration to buried

• O quantitative relationship between earthquake occurrence rates and geological deformation rates. These relationships are well to date show the proposed procedure to be very promising and founded on mathematical statements of the elastic rebound are applicable to shallow land burial as well as above ground disposal (e.g. Tumulus). In essence, the technique combines theory, and well supported by observations. The concept of uni-engineered or positive control of run-off, along with a vegetative form nsk spectra of Anderson and Tnfunac (1977) is general-cover, and is named " bioengineering management". To investi-12ed to include (1) more refined desenption of earthquake gate control of infiltration, lysimeters are being used to make source zones, (2) the uncertainties in estimating setsmicity pa-complete water balance measurements. The studies have been rameters a and b in log (10)N = a-bM, (3) the uncertainties in underway at the MMey Flats, Kentucky, low level waste dispos.

estimation of maximum earthquake size in each source zone, ai facility for the past three seasonal years. When the original and (4) the most recent results on empincal scallhg of strong Maxey Flata site closure procedure is followed, it is necessary motion amplitudes at a site. Examples of using the new NEO.

to pump large emounts of water out of the lysimeters to prevent RISK program are presented and compared with the corre-the water table from nsino cincar than 9 motore from the sur.

sponding case studies of Anderson and Tnfunac (1977). The or.

face. Using the bioengineering management procedure, no ganization of the computer program NEORISK is also bnefly de-pumping is required. As a result of the encouraging initial find-scobed.

ings in the rather small-scale fysimeters at Maxey Flats, a large-scale facility for demonstration of the bioengineenng manage-ment technique has been constructed at Beltsville, Maryland.

20 Main Citations and Abstracts This faciiity is now operational with the demonstration and data significant data This data es needed to establish meaningful collection underway testing or replacement intervals for safety system RTDs. An im-NUREG/CR 4921: ENGINEERING AND OUALITY ASSURANCE p rtant corollary benefit from this expanded program will be a better definition of achievable accuracies in RTO calibration.

COST FACTORS ASSOCIATED WITH NUCLEAR PLANT This report concludes a six monti Phase I pro}ect performed for MODIFICATION SMITH,M.H.; ZIEGLER E.J United Engineers the Nuclear Regulatory Commission under the SBIR program.

& Constructors, Inc. (subs of Raytheon Co ) Apnl 1987.59pp.

8706P40192 4 t452 001.

NUREG/CR-4936:

AN INTEGRATED This study provides genenc estimates of engineenng and GEOLOGICAL, GEOPHYSICAL,AND GEOCHEMICAL INVESTI-Quality Assurance (OA) costs based upon the development and GATION OF THE MAJOR FRACTURES ON THE EAST SIDE analysis of new and existing data The estimates are cost fac-OF THE NEW MADRID EARTHOUAKE ZONE. STEARNS.R.G.;

tors, not absolute dollar values, expressed as a percentage of REESMAN.A L Vanderbilt Univ., Nashville, TN. May 1987.

the direct cost of implemer ting the plant modification. These 36pp 8705290297,41115:261.

factors vary significantly depending upon the work environment The eastern edge of the Mississippi Valley graben (Reelfoot at the time of the modification Generally the work environment nft) is a senes of offset segments marked by offset "ndges" of refers to two groupings of plants: the first relates to require-gravity anomaly. The most prominent effsetting fault is the monts affecting structures and systems already in place, while Dyersburg line that is at least 60 miles long and cuts completely the second relates to new construction requirements which maV through the graben. The Dyersburg line is traced by earth-be applicable to future plants, plants under construction, and/or quakes, the pattern of the geothermal gradient, offset gravity operating plants The types of modifications this study address' anomahes, and surf ace linears Composition of ground water es are the physical modifications to the structures / systems of from the Cretaceous McNairy Formamon, the Eccene Wilcox nuclear power plants as opposed to analytical or procedura!

and Claiborne formations, and Holocene alluvium are all signife-changes The derived estimates, when multiplied by the direct cant for the structure of the nft. Trends for some chemical spe-cost 0 e., equipment, material, and installation labor) or instalhng cies are parallel to the nft or the Dyersburg line (and probably specific structures, systems, and pieces of equipment at a nu-the pre-Cretaceous Pascola arch). Banum and lithium show par-clear power plant will generate a reasonable order of magnitude allebsm to both; strontium isotope ratios trend parallel to the nft estimate of the enginconng O/A costs associated with a physi-whereas carbon isotope ratios trend parallel to the Pascola cal modificatiori.

arch. Upward leakage along faults of mineralized and high pres.

NUREG/CR-4922: STEAM SEPARATOR MODELING FOR VARI.

sure water from the McNairy is the likely explanation for local-OUS NUCLEAR REACTOR TRANSidNTS.

PAIK,C.Y.;

ired occurrence of mineralized water in younger aquifers. The MULLEN Ga KNOESS.C.; et al. Massachusetts Institute of best example is chloride water in Holocene alluvium on the Technoloay, Cambndge, MA. June 1987. 227pp. 8706240352.

Dyersburg line.

EPRI NP-E272. 41439 289 Expenments were performed using air and water on three dif-NUREG/CR-4938: OCCUPATIONAL RADIATION EXPOSURES ferent types of centnfugal sepemtors: a cyclone as a genenc ASSOCIATED WITH ALTERNATIVE METHODS OF LOW-separator, a Combustion Engineenng type stationary swirl vane LEVEL WASTE DISPOSAL HERRINGTON,W.N.; HARTY,R.;

separator, and a Westinghouse type separator. The cyclone MERWIN,S E. Battelle Memonal Institute, Pacific Northwest separator system has three stages of separation: first the cy.

Laboratones MaY 1987. 110pp. 8706240313. PNL-6217, clone, then a gravity separator, and finally a chevron plate sep.

4 t 450:176 arator. The other systems have only a centnfugal separator t isolate tho effect of the pnmary separator. Expenments wer (LLRWPA) Act of 1985 mandates that the U S. Nuclear Regula-also done in MIT blowdown ng. with and without a separator, tofY Commission $NRC), in consultation with states and other in-using steam and water. The separators appear to perform well terested parties, identify disposal methods other than shallow at flow rates well above the design values as long as the down-land bunal (SLB), the method currently used at the three low-comer water level is not high. High downcomer water level level waste (LLW) disposal sites operating in the United States.

rather than high flow rates appear to be the pnmary cause of We compared projected occupational exposures associated degraded performance. Appreciable carry.over from the separa-with the SLB method and five alternative disposal methods, in-tor section of a steam generator occurs when the drain lines cluding below ground vaults (BGV), above ground vaults (AGV),

from three stage of separation are unable to carry off the liould earth mounded concrete bunkers (EMCB), augured holes (AH),

flow _

and mined cavities (MC). MC facilities were studied in less detail because this disposal method is not being actively considered-NUREG/CR-4928: DEGRADATION OF NUCLEAR PLANT TEM-Reference facility designs and a list of probatJe tasks required PERATURE SENSORS. HASHEMIAN.H Ma PETERSEN.K.M.;

at each site for receiving and disposing of the 3w-level waste KERUN T W.; et al Analysis & Measurement Services Corp.

were developed. Each task was analyzed by worker require.

June 1987. 85pp 8707020108. 41568.281.

ments, time requirements, distances between workers and the A program was estabbshed and initial tests were performed to waste, and exposure rates at those distances. This information evaluate long term performance of resistance temperature de-was used to estimate the dose received by workers during dis-tectors (RTDs) of the type used in U.S. nuclear power plants.

posal of four types of waste packaging: drums, wood boxes, The effort addressed the effect of aging on RTD calibration ac-resin liners, and dumpsters. The results of this study suggest curacy and response time This Phase i effort included expo-that, of the methods studied in detail, occupational dose equiva-sure of thirteen nuclear safety system grade RTD elements to lents would be highest for the EMCB method (1.81 person-simulated LWR temperatures. Full calibration) were performed mrem /m(3) of waste disposed). The lowest occupational dose initially and monthly, sensors were monitored and cross equwalents would occur for the AH method (1.29 person-checked continuously dunng exposure, and responso time tests mrem /m(3)). Projected occupational dose equivalents for SLB, were performed before and after exposure. Short term cahbra-BGV, and AGV disposal methods are 1.38, 1.47, f nd 1.61 tion dnfts of as much as 1.8 degrees F (1 degree C) were ob-person-mrem /m(3),respectively.

served. Hesponse times were essentially unaffected by this testing. This program shows that there is a sourd reason for NUREG/CR-4950 V01: THE SHORELINE ENVIRONMENT AT-concem about the accuracy of temperature measurements in MOSPHERIC DISPERSION EXPERIMENT nuclear power plants These limitad tests should be expanded (SEADEX) Experiment Desenption.

CANTRELL.B.K.;

in a Phase il program to involve more sensors and longer expo.

JOHNSON.W.B.; MORLEY,B.M.; et al. SRI international. June sures to simulated LWR conditions in order to obtain statistically 1987. 40pp. 8707130162. 416A4.136.

Main Citations and Abstracto 21 The SEADEX atmospheric dispersion field study was conduct-mendations to improve the biases include providing calibrations l

ed dunng the penod May. June 8,1982, in ryytheastern Wiscon-(using appropriate irradiation standards and phantoms) followed I

sin, in the vicinity of the Kewaunee Power Plant on the western by another set of performance tests, as well as visiting proces-shore of Lake Michigan. The specific objectives of SEADEX sors to identify other possible sources of error. It is further rec-were to charactenze (1) the atmosphenc dispersion and (2) the ornmended that the draft standard be re-evaluated to ensure meteorological conditions influencing this dispersion as com-that it is appropriate for the performance testing of extremity do-pletely as possible dunng the test penod. This field study includ-simeters.

ed a senes of controlled tracer tests utikzing state of the-art tracer measurement technology to determine honzontal and ver-NUREG/CR-4964: UPDATE OF TABLE S-3 NONRADIOLOGICAL tical dispersion over both land and water. Extensive meteorolog-ENVIRONMENTAL PARAMETERS FOR A REFERENCE ical measurements were obtained to thoroughly characterize the LIGHT-WATER REACTOR. Uranium Mining Milling And Enrich-three. dimensional structure of the atmosphenc boundary layer ment. HABEGGER,L.J.; CARSTEA.D.D.; OPELKA,J.H. Argonre controlkng the dispersion process. This volume desenbos the National Laboratory. June 1987, 68pp. 8707010655. ANL/EES-expenmental design for, and conduct of, the study.

TM-332. 41545.268.

NUREG/CR-4959: PERFORMANCE TESTING OF EXTREMITY in 1974, Table S-3 of the report " Environmental Survey of the DOSIMETERS. HARTY,R.; REECE,W.D.; HOOKER,C.D. Battelle Uranium Fuel Cycle" was pubhshed as a technical basis for Memorial Institute, Pacific Northwest Laboratones. June 1987.

consideration of the environmental effects of the uraniurn fuel 65pp. 8707020364. PNL-6218. 41566.202.

cycle supporting operation of light-water reactors. A ref6rence The Health Physics Society Standing Committee (HPSSC) reactor cooled with light, or ordinary, water was established to Working Group on Performance Testing of Extremity Dosimeters reduce the burden on the Nuclear Regulatory Comminsion has issued a draft of a proposed standard for extremity dosi.

(NRC) staff, reactor license applicants, and other interested per-meters. The draft standard proposes methods to be used for sons by removing the necessity to relatigate the environmental testing extremity dosimetry systems and the performance ente.

effects attnbutable to the fuel cycle, effects that are not within non used to determine comphance. This study evaluates the an applicant's control, in every individual reactor licensing pro-draft standard's proposed performance entenon (absolute value ceeding. In a 1984 evaluation of a license application, it was of B + S less than or equ

• 0.35, where B is the bias and S is demonstrated that the Table S-3 estimate of annual effluent of the standard deviation) ao. ' the performance of extremity do-coal particulate is larger, possibly by as much as a factor of simeter processors. Twei.. ane types of extremity dosimeters 100, than actual current values. Partially as a result of this eval-from 11 processors were irradiated by the Pacific Northwest uation, the NRC initiated a study to update all of the major non.

Laboratory (PNL) to specific dose levels in one or more of radiological values in Table S-3. The results of the study are seven categones. The processors evaluated the doses and re-documented in this update. The report evaluates only the turned the results to PNL for analysis. Approximately 60% of mining. milling, and isotopic-enrichment components of the fuel the dosimeters met the performance cntenon. Two-thirds of the cycle's environmental parameters since these are the areas in remaining dosimeters had large biases (ranging from 0.25 to which the greatest changes from the original study could be an-0.80) but small standard deviations (less than 0.15). Recom-ticipated.

s

L' Secondary Report Number index This index lists, in alphabetical order, the performing organization-issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross-referenced to the NUREG numoer for the report and to the 10 digit NRC Document Control System accession number.

SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER AECL 9342 NUREG/CR4821 ORNL/TM 10100 NUREG/CR-4651 ANL-8646 NUREG/CR-4842 ORNL/TM 10236 NUREG/CR-4758 ANL/EES-TM-332 NUREG/CR-4964 ORNL/TM 9593 NUREG/CR 4219 V03 N2 BMI-2120 NUREG/CR-4082 VOS PNL-4221 NUREG/CR-2850 VOS PNL-5511 NUREG/CR-4300 V04 N1 BMI-2149 NUREG/CR-4872 PNL-5711 NUREG/CR-4469 V05 BMI-2150 NUREG/CH4877 R-4 W

NL BMi-2151 NUREG/CR4878 p

,q PN(5 779 BNL NUREG 51454 NUREG/CR-2331 V06 N3 PNL5 NUREG/CR-3231 BNL NUREG-52021 NUREG/CR-4739 PNL 5809 NUREG/CR 4330 V03 BNL NUREG-52001 NUREG/CR-4883 PNL 5822 NUREG/CR-4583 V02 CONF-860820 NUREG/CP 0086 V02 PNL-5894 NUREG/CR-4773 CONF 880820 NUREG/CP-0086 V01 PNL 5907 NUREG/CR-4653 EGG-2379 NUREG/CR4165 PNL-6065 NUREG/CR-4779 EGG-2443 NUREG/CR-3956 PNL-6165 NUREG/CR4890 EGG-2456 NUREG/CR-4617 PNb6166 NUREG/CR4889 EGO-2486 NUREG/CR-4802 PNL-6217 NUREG/CR-4938 EGG-2492 NUREG/CR-4821 PNL-6218 NUREG/CR-4959 EGG-2496 NUREG/CR-4845 SAND 841586 NUREG/CR 3925 REV EGG-2499 NUREG/CR 4868 SAND 84 2631 NUREG/CR-4098 EPRI NP-5272 NUREG/CR-4922 SAND 86-0311 NUREG/CR-4879 HEDL TME 85 3 NUREG/CR-3319 R01 SAND 86-0336 NUREG/CR4527 VOI HEDL TME 86-2 NUREG/CR-4307 V03 SAND 86 '043 NUREG/CR4719 IEB-83-01 NUREG/CR-4663 SAND 86-1135 NUREG/CR-4700 V2 DRF SAND 86-1135 NUREG/CR-4700 V4 DRF IE B-83-04 NURE"/CR-4664 SAND 86-1296 NUREG/CR-4681 LA 10730-MS NUREC/CR-4615 V02 SAND 86-1309 NUREG/CR-4551 V3 PT1 LA 10856 M NUREG/CR-4765 S

309 MEGW551 V3 M2 LA 10918-MS NUREG/CR-4814 SAND 861309 NUREG/CR-4551 V2 DRF LA 10962 MS NUREG/CR4875 SAND 86-1309 NUREG/CR4551 V4 DRF LA 10981 MS NUREG/CR4901 SAND 861996 NUREG/CR-4772 MEA 2200 NUREG/CR4841 SAND 86-2084 NUREG/CR-4550 VOS MEA 2210 NUREG/CR-4894 SAND 86-2084 NUREG/CR-4550 V06PT1 MTR-86 WOO 213 NUREG/CR-4783 SAND 86-2084 NUREG/CR-4550 V06PT2 ORNL/NOAC-232 NUREG/CR 4874 V04 SAND 86-2689 NUREG/CR 4830 ORNL/NOAC-242 NUREG/CR-4674 V03 UCID-20092 NUREG/CR 4161 V02 ORNL/NSIC-2000 NUREG/CR-2000 V06 N2 UCID 20679 KREG/CR 4800 ORNL/NSIC-2000 NUREG/CR-2000 V06 N3 UCID 20696 NUREG/CR-4885 ORNL/NSIC-2000 NUREG/CR-2000 V06 N4 UCID'21003 NUREG/CR-4899 ORNL/NSIC 2000 NUREG/CR 2000 V06 NS UCID-21039 NUREG/CP-0087 23

l Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by the author. If further information is needed, refer to the main cita-tion by the NUREG number.

A B T.S.R.

BANDER,T.J.

NUREG/CR4651: DEVELOPMENT OF RIPRAP DESIGN CAfTERIA BY NUREG/CR-4653: GASPAR 11 TECHNICAL REFERENCE AND USER RIPRAP TESTING IN FLUMES. Phase 1.

GUiOE.

AHMAD,J.

BARNES,C.R.

NUREG/CR4082 V05' DEGRADED. PIPING PROGRAM PHASE NUREG/CR-4082 VOS: DEGRADED PIPING FROGRAM PHASE NRG 87 X

N ANALYTICAL ASSESSMENT 11 Semiannual Report, April-September 1986.

j OF CIRCUMFERENTIALLY SURFACE CRACKED PIPES UNDER BARSTOW,N.

NUF E 4878. ANALYSIS OF EXPERIMENTS ON STAINLESS NUREG/CR 4822: BROAD BAND SEISMIC DATA ANALYSIS. September STEEL FLUX WELDS Topical Report.

1984 - September 1986.

AL2HEIMER.J.M.

BATES,D.J.

NUREG/CR-3231: PIPE.TO. PIPE IMPACT PROGRAM.

NUREG/CR 4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RELi-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUR'E /CR-4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION PEACH BEEBE,M.R.

N E C 4 1V 2

L Tl OF S RE ACCIDENT RISKS

-0020 m m WEEED ONM ONS mM A

PORLData As Of October 3W86Say M O AND THE POTENTIAL FOR RISK REDUCTION PEACH BOTTOM, UNIT 2. Appendices Draft For Comment' SEVERE ACCIDENT NUREG/CR-4551 V4 DRF: EVALUATION OF BEHR'V.L RISKS AND THE POTENTIAL FOR RISK REDUCTION: GRAND NUREG/CR4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR GULF, UNIT 1. Draft For Comment.

POSTULATED SEVERE. ACCIDENTS: SEQUOYAH POWER NUREG/CR 4700 V4 DAF: CONTAINMENT EVENT ANALYSIS FOR STATION UNIT 1. Draft Report For Comment.

1 POSTULATED SEVERE ACCIDENTS: GRAND GULF NUCLEAR STATION,0 NIT 1. Draft For Comment.

BENJAMIN,A.S.

NUREG/CR-4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT ANDERSON.J.G.

RISKS AND THE POTENTIAL FOR RISK REDUCTION.SEQUOYAH NUREG/CR-4903 V02: SELECTION OF EARTHOUAKE RESISTANT POWER STATION. UNIT 1. Draft For Comment.

DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO-NUREG/CR.4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS GY AND TECHNICAL CASES. Methods For Introduction Of Geolograt AND THF POTENTIAL FOR RISK REDUCTION. PEACH l

Data into Characteri2allon Of Actwe Faults And Seismicity And..

BOTTO:..//*J.. (Main Report. Draft For Comte 1t.

NUREG/CR-4551 V3 FT2: EVALUATION OF SIVERE ACCIDENT RISKS N RE / R 4928 DEGRADATION OF NUCLEAR PLANT TEMPERA-TURE SENSORS.

TTOM. UNIT 2 es Draft For Co nt.

NUREG/CR4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT AUSTIN,P.N.

RISKS AND THE POTENTIAL FOR RISK REDUCTION: GRAND I

NUREG/CR4674 V03. PRECURSs MS TO POTENTIAL SEVERE CORE GULF, UNIT 1. Draft For Comment.

DAMAGE ACCIDENTS.1984,A SMTUS REPORT. Main Report And Ap-NUREG/CR-4700 Y2 DRF: CONTAINMENT EVENT ANALYSIS FOR pendixes A And B.

POSTULATED SEVERE ACCIDENTS: SEOUOYAH POWER NUREG/CR4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE STATION. UNIT 1. Draft Report For Comment.

DAMAGE ACCIDENTS 1984,A STATUS REPORT. Appendixes C,0 And OM NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL BACANSKAS,V.P.

WASTE REPOSITORY. Volume 2: Tuff.

NUREG/CR 4819 V01: AGING AND SERVICE WEAR OF SOLENOID-OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR BERNREUTER,D.L POWER PLANTS Volume 1 Or rating Experience And Failure identitL NUREG/CP 0087:

SUMMARY

REPORT OF THE SYMPOSIUM ON SEIS-cation.

MIC AND GEOLOGIC SITING CRJERIA FOR NUCLEAR POWER PLANTS.

CAKER,C.H.

NUREG/CR4885: SEISMIC HAZARD CHARACTER 12ATION OF THE NUREG/CR4623: IN-SITU STRESS MEASUREMENTS IN THE EASTERN UNITED STATES.Comparatwo Evaluation Of The LLNL And EARTH'S CRUST IN THE EASTERN UNITED STATES.

EPRI Studies.

NUR G'/CR-2850 V05: POPULATION DOSE COMMITMENTS DUE TONUREG/C' R-4550. V05: ANALYSIS OF CORE DAMAGE FREQUENCY RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1983.

FROM INTERNAL EVENTS: SEQUOYAH. UNIT 1.

SALLINGER.M.Y.

BESSETTE.D.

NUREG/CR-4779: NEW DATA FOR AEROSOLS GENERATED BY RE.

NUREG 1271: GUIDELINES AND PROCEDURES FOR THE INTERNA.

LEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC TIONAL CODE ASSESSMENT AND APPLICATIONS PROGRAM.

AIR.

BINNALL,E.P.

BAMPTON,M.C.

NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL NUREG/CR-3231: PIPE.TO. PIPE IMPACT PROGRAM.

WASTE REPOSITORY. Volume 2. Tuff.

25 w

26 P:tsonal Author Ind;x BISH.D.L CLETCHER.J.W.

NUREG/CR-4875. CHARACTERIZATION OF CRUSHED TUFF FOR THE NUREG/CR4674 V03: PRECURSORS TO POTENTIAL SEVERE CORE EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES DAMAGE ACCIDENTS 1984.A STATUS REPORT. Main Report And Ap.

IN THE UNSATURATED ZONE pendmes A And B.

NUREG/CR4674 V04. PRECURSORS TO POTEN1. \\l. SEVERE CORE CLACK,0.A.

DAMAGE ACCIDENTS.1984,\\ STATUS REPORT.A e idns C.D And NUREG/CR-4098. SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL-E.

ERATED-AGED CLASS 1E BATTERY CELLS COHEN,L ILACKMAN,H.S.

NUREG-0837 V06 N04 NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4617; ONSITE ASSESSMENTS OF THE EFFECTIVENESS AND IMPACTS OF UPGRADED EMERGENCY OPERATING PROCE.

NETWORK. Progress Report, October-December 1986.

DURES.

COURTRIGHT,E.L

^

BONANO,E.J.

REES '

NUREG/CR-3925 REV:

SWIFT ti SELF TEACHING l

CURRICULUM illustrative Problems For The Sandia Waste-Isolation CRAWFORD,S.L Flow And Transport Model For Fractured Media-NUREG/CR4583 V02: DEVELOPMEN' AND VALIDATION OF A REAL-TIME SAFT-UT SYSTEM FOR THL INSPECTION OF LIGHT WATER Nt EG C 4098: SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL-REACTOR COMPONENTS Annual Report, October 1984 September ERATED-AGED CLASS IE BATTERY CELLS.

1985 BOCXER,J.M.

CREWS.J.L NUREG/CR4814: SOURCES OF CORRELATION BETWEEN NUREG 1269. LOSS OF RESIDUAL HEAT REMOVAL SYSTEM.Diablo EXPERTS Empencal Results From Two Extremes.

Canyon Unit 2, Apr410.1987.

BOYD.G.J CRONENBERG,A.W.

NUREG/CR4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS NUREG/CR4866. AN ASSESSMENT OF HYDROGEN GENERATION AND THE POTENTIAL FOR RISK REDUCTION PEACH FOR THE PBF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

BOTTOM, UNIT 2 Main R Draft For Comment.

rD E

E TIAL FR IK REDU T ON PEACH NUREG/C 7:

SUMMARY

REPORT OF THE SYMPOSIUM ON SEIS.

BOTTOM. UNIT 2 Appendices Draft For Comment.

MIC AND GEOLOGIC SITING CRITERIA FOR NUCLEAR POWER NUREG/CR 4551 V4 DRF. EVALUATION OF SEVERE ACLIDENT PLANTS.

J RISKS AND THE PO7NTIAL FOR RISK REDUCTION GRAND GULF, UNIT 1 Draft For Comment.

M '

SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-REG /C 4165:

ERIZHALL,J.L GRAM ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA.

NUREG/CR-4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM TIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

WITH U02.

DAVIS,P.A.

BRUST,F.

NUREG/CR-3925 REV:

SWIFT 11 SELF-TEACHING NUBEG/CR-4082 V05. DEGRADED PIPING PROCRAM - PHASE CURRICULUM.lliustrative Problems For The Sandia Waste-Isolation 11 Semiannual Report Apnt-September 1986.

Flow And Transport Model For Fractured Media.

I NUREG/CR4876: ' ANALYSIS OF EXPERIMENTS ON STAINLESS STEEL FLUX WELDS Topical Report.

DAVIS,S.N.

NUREG/CR-4912: DATING GROUND WATER AND THE EVALUATION BYERS.C.D.

OF REPOSITORIES FOR RADIOACTIVE WASTE.

NUREG/CR4842: A ST!'DY OF NATURAL GLASS ANALOGUES AS APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS DEAN.R.S.

NUREG/CR4663: CLOSEOUT OF lE BULLETIN 83-01: FAILURE OF RE-CALLAHAN,0.M.

ACTOR TRIP BREAKERS (WESTINGHOUSE DB-50) TO OFEN ON NUREG/CR-4894: A USER'S GUIDE TO THE NRC'S PIPING FRAC-AUTOMATIC TRIP SIGNAL, TURE MECHANICS DATA BASE (PIFRAC).

NUREG/CR4664: CLOSEOUT OF IE BULLETIN 83-04: FAILURE OF CAMP.A.L THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP BREAK-NUREGrCR-4550 VOS ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS: SEOUOYAH. UNIT 1-DEFFENDAU1H,J.

CANTRELL,B.K NUREG/CR 4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RELI-ABILITY COR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR 950 V01: THE SHORELINE ENVIRONMENT ATMOS-PHERIC DISPERSION EXPERIMENT (SEADEX)Expenment Desenp.

NUREG CR44 06 N D TRU i AT N (NDE) RELI-I ABILITY FOh INSERVICE INSPECTION OF LIGHT WATER CARSTEA.D.D.

REACTORS Semennual Report, October 1986 March 1987.

NUREG/CR4964. UPDATE OF TABLE S.3 NONRADIOLOGICAL ENVI-RONMENTAL PARAMETERS FOR A REFERENCE LIGHT WATER N REG /CR 161 V02: CD'TICAL PARAMETERS FOR A HIGH LEVEL REACTOR. Uran um Mining. Milling And Ennchment.

WASTE REPOSITORY. Volume 2. Tuff CARTER.J.A.

N EG/CR4822 OAD BAND SEISMIC DATA ANALYSIS September EL'G CR 3956 IN SITU TESTING OF THE SHIPPINGPORT ATO RE POWER STATION ELECTRICAL CIRCulTS.

CHAVEZ,J M.

NUREG/CR4527 V01: AN EXPERIMENTAL INVESTIGATION OF IN_

DOCTOR,S.R.

TERNALLY IGNITIED FIRES IN NUCLEAR POWER PLANT CONTROL NUREG/CR4469 VOS: NONDESTRUCTIVE EXAMINATION (NDE) REll-CABINETS.Part 1: Cabinet Effects Tests.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER l

REACTORS Semiannual Heport, Aprd-September 1986.

CHOU,C.K.

NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RELl-NUREG/CR4899.

COMPONENT FRAGILITY RESEARCH ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER PROGRAM Phase I Component Pnontization.

RE ACTORS. Semiannual Repert. October 1986 March 1987.

NUREG/CR4583 V02: DEVELOPMENT AND VALIDATION OF A REAL-CLARK,R.A.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CR-4848: STEAM GENERATOR GROUP PROJECT. Annual REACTOR COMPONENTS. Annual Report, October 1984 - September Report 1985.

1986.

l

PCreonCl Author index 27 DONALDSON,M.R.

GOOC.M.S.

NUREG/CR-3956. IN SITU TESTING OF THE SHIPPINGPORT ATOMIC NUREG/CR4469 V05: NONDESTRUCTIVE EXAMINATON (NDE) RELi-POWER STATION ELECTRICAL CIRCUITS.

ABILITY FOR INSERVICE ' INSPECTON OF LIGHT WATER REACTORS. Semiannual Report April September 1986.

DRISKELL,W.E.

NUREG/CR4469 V06: NONDESTR'.;CilVE EXAMINATION (NDE) REU-NUREG/CR4845: AN ANALYSIS OF THE SEMISCALE MOD-2C S-NH-3 ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER TEST USING THE TRAC-PF1 COMPUTER PROGRAM-REACTORS.Semannual Report October 1986 March 1987.

DROUIN,M.T.

GOTCHY,R.L NUREG/CR4550 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG-0332: POTENTIAL HEAL"H AND ENVIRONMENTAL IMPACTS ATTRIBUTABLE TO THE NUCl. EAR AND COAL FUEL CYCLES. Final NUREG/ -

V T2 ANAL IS F DAMA R

EN-CY F ROM INTERNAL EVENTS GRANO GULF UNIT 1. Appendices.

Report ETTLINGER,L GOTTULA,R.C.

NUREG/CR 4783: ANALYSIS OF BALANC.E OF PLANT REGULATORY NUREG /CR-4185: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-ISSUES Final Report GRAM ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA.

j TONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

EVANS,D.D.

NUREG/CA 4655 UNSATURATED FLOW AND TRANSPORT GREEN E.R.

THROUGH FRACTURED ROCK FtELATED TO HIGH-LEVEL WASTE NUREG/CR-4469 V06. NONDESTRUCTIVE EXAMINATON (NDE) REll-REPOSITORIES Final Report. Phase ll ABluTY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS. Semiannual Report October 1986 March 1987.

NUREG/CR-4842: A STUDY OF NATURAL GLASS ANALOGUES AS ORIESMEYER,J.M.

I APPUED TO ALTERATION OF NUCLEAR WASTE GLASS.

NUREG/CR4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION: PEACH FARNSTROM.K.A.

BOTTOM, UNIT 2. Main ReportDraft For Comrnent NUREG/CR4815: DEMONSTRATION TESTING OF A SURVEILLANCE NUREC/CR-4551 V3 PT2: EVALUATON OF SEVERE ACCIDENT RISKS ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs AND THE P0TENTIAL FOR RISK REDUCTION.FEACH And Benehts.

BOTTOM. UNIT 2. Appendices. Draft For Comment.

FELTMAN,A.

NUREG/CR4815: DEMONSTRATION TESTING OF A SURVEILLANCE CR-4922: STEAM SEPARATOR MODELING FOR VARIOUS ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs NUG EAR REACTOR TRANSIENTS.

And Benehts.

FIRST,M.W.

GUERRIERl,0 NUREG/CP-0006 V01: PROCEEDINGS OF THE 19TH DOE /NRC NU.

nUREG/CR-4082 VOS: DEGRADED PIPING PROGRAM PHASE CLEAR AIR CLEANING CONFERENCE. Held in ll. Semiannual Report, April-September 1986.

I SeaWWashington. August 18-21,1986.

NUREG/CP-0086 V02: PROCEEDINGS OF THE 19TH DOE /NRC NO.

HA8 EGGER.LJ.

CLEAR AIR CLEANING CONFERENCE. Held in NUREG/CR4964: UPDATE OF TABLE S 3 NONRADIOLOGICAL ENVb Seattle. Washington, August 18-21,1986.

RONMENTAL PARAMETERS FOR A REFERENCE LIGHT WATER REACTOR. Uranium Mining. Milling And Ennchment.

FOLEY,W.J.

NUREG/CR4663: CLOSEOUT OF IE BULLETIN 83-01: FAILURE OF RE-HAGEN,E.W ACTOR TRIP BREAKEN, (WESTINGHOUSE DB-50) TO OPEN ON NUREG/CR-4674 V03: PRECURSORS TO POTENTIAL SEVERE CORE AUTOMATIC TRIP SIGNAL.

DAMAGE ACCIDENTS.1984,A STATUS REPORT. Main Report And Ap-NUREG/CR-4664: CLOSEOUT OF lE BULLETIN 83-04. FAILURE OF pendixes A And B.

THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP BREAK.

NUREG/CR-4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE ERS.

DAMAGE ACCIDENTS:1984,A STATUS REPORT. Appendixes C,D And FORD,R.E.

NUREG/CR-4617: ONSITE ASSESSMENTS OF THE EFFECTIVENESS HALL,D.G, AND IMPACTS OF UPGRADED EMERGENCY OPERATING PROCE.

NUREG/CR4802: AN EVALUATION OF TRAC-PFt/ MOD 1 COMPUTER DURES CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE-MISCALE MOD-2C FEEDWATER LINE BREAK TRANSIENTS.

FRIESEL,M.A.

NUREG/CR4300 V04 N1: ACOUSTIC EMISSON/ FLAW RELATION.

HALL,T.E SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR4583 V02: DEVELOPMENT AND VAUDATION OF A REAL.

VESSELS. Progress Fleport, October 1986 March 1987.

TIME SAFT-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER FRILEY,J.R.

REACTOR COMPONENTS. Annual Report,0ctober 1984 September NUREG/CR-3231: PIPE TO-PIPE IMPACT PROGRAM.

1985-FUENTES,H.R.

HANEY,LN.

NUREG/CR4615 V02: MODELING STUDY OF SOLUTE TRANSPORT NUREG/CR4617: ONSITE ASSESSMENTS OF THE EFFECTIVENESS IN THE UNSATURATED ZONE. Workshop Proceedings.

AND IMPACTS OF UPGRADED EMERGENCY OPERATING PROCE.

NUREG/CR-4875: CHARACTERIZATION OF CRUSHED TUFF FOR THE DURES.

EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES IN THE UNSATURATED ZONE.

HANSON,R.

NUREG/CR 4901: EFFECTS FROM INFLUENT BOUNDARY CONDI.

NURF.G-1270 V01: INTERNATIONAL CODE ASSESSMENT AND APPLl.

TIONS ON TRACER MIGRATION AND SPATIAL VARIABluTY FEA.

CATIONS PROGRAM. Annual Report.

TURES IN INTERMEDIATE SCALE EXPERIMENTS.

NUREG-1271: GUIDEUNES AND PROCEDURES FOR THE INTERNA-TIONAL CODE ASSESSMENT AND APPUCATIONS PROGRAM.

GILBERT,R.W.

NUREG/CR-4583 V02: DEVELOPMENT AND VAllDATION OF A REAL.

HARMS,N.L TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CR-4773: DESIGN FEATURES TO FACluTATE INTERNATION-REACTOR COMPONENTS Annual Report,0ctober 1984. Septembe AL SAFFGUARDS AT MIXED-OXIDE 90NVERSION FACIUTIES.

1985.

HARPER F.T.

GLADNEY.E.S.

NUREG/CR-4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY NUREG/CR4875: CHARACTERIZATION OF CRUSHED TUFF FOR THE FROM INTERNAL EVENTS: SEQUOYAH, UNIT 1.

EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES NUREG/CR4550 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN-IN THE UNSATURATED ZONE.

CY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1. Main Report.

l 28 Pcroonal Auth:r Ind:x NUREG/CH4550 V06PT2: ANALYSIS OF CORE DAMAGE FREQUEN.

HOLBERT,K.E.

CY FROM INTERNAL EVENTS GRAND GULF, UNIT 1. Appendices.

NUREG/CR4928: DEGRADATION OF NUCLEAR PLANT TEMPERA.

HARRIS,J.D.

NUREG/CR 4674 V03 PRECURSORS TO POTENTIAL SEVERE CORE HOLCOMS.EE.

DAMAGE ACCOENTS.1984,A STATUS REPORT. Main Report And Ap-NUREG/CR-4165: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-pendises A And B.

qAM. ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA-NUREG/CR4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE TONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

DAMAGE ACCOENTS.1984.A STATUS REPORLAppendizes C.D And E.

HOLMAN,G.S.

NUREG/CR 4899:

COMPONENT FRAGILITY RESEARCH

'"P "*"I " " "

• NUREG/CR4938: OCCUPATIONAL RAD 1ATON EXPOSURES ASSOCI-ATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE DIS-HOOKER,C.D.

NU E /CR4959. PERFORMANCE TESTING OF EXTREMITY DOSI-M EA METERS.

HOSPODOR,S.

HARVEY,H.W I

NUREG-0980 R03: NUCLEAR REGULATORY LEGISLATION.

NUREG/CR-48& DEMONSTRATION TESTING OF A SURVEILLANCE ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs HOWARD,R.

And BeneMs NUREG/CR4815: DEMONSTRATION TESTING OF A SURVEILLANCE

. ASHEMtAN,H.M.

ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analyses Of Costs NUREG/CR 4928: DEGRADATION OF NUCLEAR PLANT TEMPERA.

And Benefas.

TURE SENSORS' HUTTON,P.H.

i HASKIN,F.E.

NUREG/CR4300 V04 N1: ACOUSTIC EMISSION / FLAW RELATON-(

NUREG/CR4551 V3 PT1:EVALUAPON OF SEVERE ACCIDENT RISKS SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE AND THE POTENTIAL FOR RISK REDUCTION. PEACH VESSELS. Progress Report. October 1986. March 1987.

BOTTOM. UNIT 2 Main Report Draft For Comment.

NUREG/CR 4551 V3 PT2: EVALUATON OF SEVERE ACCOENT RISKS 18HIKAWA M-AND THF POTENTIAL FOR RISK REDUCTION PEACH NUREG/CR.4765: MXS CROSS SECTION PREPROCESSOR USER'S BOTTOM. UNIT 2. Appendices. Draft For Comment.

MANUAL HEASLER,P.G.

JAMISON.J.D.

NUREG/CR4469 V05 NONDESTRUCTIVE EXAMINATION (NDE) RELI.

NUREG/CR4330 V03: REVIEW OF LIGHT WATER REACTOR REGU-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-i i

REACTORS.Semaannual Report. April. September 1986.

querements That May Have Marginal importance To Hask:Postaccident l

NUREG/CR-4469 V06 NONDESTRUCTIVE EXAMINATON (NDE) RELI-Sampling System, Turbine Missiles Combustible Gas Control, Charcoal ABILITY FOR INSERVICE INSPECTION OF UGHT WATER Filters.

REACTORS Semiannual Report, October 1986 March 1987.

JANG,J.

HELD.J.T.

NUREG-0837 V06 N04. NRC TLD DIRECT RADIATION MONITORING NUREG/CR4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY NETWORK. Progress Report, October December 1986.

FROM INTERNAL EVENTS: SEQUOYAH. UNIT 1.

JANIS,W.J.

HELTON J.C.

NUREG/CR4098: SEISMIC FRAGILITY TESTS OF MW AND ACCEL.

NUREG/CR4551 V3 PT1: EVALUATION OF SEVERE ACCIDFNT RISKS ERATED-AGED CLASS 1E BATTERY CELLS.

AND THE POTENTIAL FOR RISK REDUCTON PEACH BOTTOM. UNIT 2. Main Report Draft For Comment.

JENKS,R.

NUREG/CR 4551 V3 PT2: EVALUATON OF SEVERE ACCIDENT RISKS NUREG 1270 V01: INTERNATIONAL CODE ASSESSMENT AND APPLi-AND THE POTENTIAL FOR RISK REDUCTION. PEACH CATIONS PROGRAM. Annual Report.

BOTTOM. UNIT 2. Appendices. Draft For Comment.

JERCINOYlC,M.J.

HENNICK A.

NUREG/CR-4842; A STUDY OF NATURAL GLASS ANALOGUES AS NUREG/CR-4663. CLOSEOUT OF IE BULLETIN 83-01. FAILURE OF RE-APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS.

ACTOR TRIP BREAKERS (WESTINGHOUSE DB-50) TO OPEN ON AUTOMATIC TRIP SIGNAL JOHNSON,M.R.

NUREG/CR 4664: CLOSEOUT OF IE BULLETIN 83-04. FAILURE OF NUREG 1214 R01: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

THE UNDERVOLTAGE TRIP FUNCTON OF REACTOR TRIP BREAK-IC ASSESSMENT OF LICENSEE PERFORMANCE.

ERS.

JOHNSON,W.8, HEf RINGTON'4938 OCCUPATIONAL RADIATION EXPOSURES ASSOCl-W.N NUREG/CR-NUREG/CR4950 VOI: THE SHORELINE ENVIRONMENT ATMOS-ATED WITH ALTERNATIVE METHODS OF LOW LEVEL WASTE D't-PHERIC DISPERSION EXPERIMENT (SEADEX).Expenment Desenp-POSAL D0"-

i HILL R.C.

JOUSE,W.C.

l NUREG/CR4821: REACTOR COOLANT PUMP SHAFT SEAL STABlu, NUREG/O'4-4165: SEVERE ACCIDENT. SEQUENCE ANALYS!S PRO-TY DURING STATION BLACKOUT.

GRAM. ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA.

TIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

HINKLE N.E.

NUREG/CR4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY KERLIN,T.W.

- RIPRAP TESTING IN FLUMES Phase L NUREG/CR4928. DEGRADATION OF NUCLEAR PLANT TEMPERA.

TURE SENSORS.

HISER.A.L.

NUREG/CR4894: A USER'S GUOE TO THE NRC'S PIPING FRAC.

KHATTAK,M.S.

TURE MECHANICS DATA BASE (PIFRAC)

NUREG/CR4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY RIPRAP TESTING IN FLUMES. Phase L HODGSON,W.H.

NUREGiCR 4779: NEW DATA FOR AEROSOLS GENERATED BY RE-KlEFNER,J.

LEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC NUREG/CR 4082 V05: DEGRADED PIPING PROGRAM - PHASE AIR.

tLSemiannual Report, Apnl. September 1986.

l PIrsonil Authtr Indzx 29 KNOESS.C.

LEWIS.M.

j NUREG/CR 4922: STEAM SEPARATOR MODtELING FOR VARIOUS NUREG/CR 4848: STEAM GENERATOR GROUP PROJECT. Annual NUCLEAR REACTOR TRANSIENTS.

Report 1985.

KOLACZKOWSKI,A Lgwig,gjt NUREG/C44700 V.* DAF: CONTAINMENT EVENT ANALYSIS FOR NUREG/CR 4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT POSTULATED SEVERE ACCIDENTS. GRAND GULF NUCLEAR RISKS AND THE POTENTIAL FOR RISK REDUCTION.SEQUOYAH STATON. UNIT 1. Draft For Comment.

POWER STATON, UNIT 1. Draft For Comment.

NUREG/CR-4551 V3 PT1: EVALUATION OF SEVERE ACCOENT RISKS KOUTS.H' C44883:

AND THE POTENTIAL FOR RISK REDUCTION. PEACH NUREG/

REVIEW OF RESEARCH ON UNCERTAINTIES IN M

T ESTIMATES OF SOURCE TERMS FROM SEVERE ACCIDENTS IN NURE /CR 1

P2 AT OS RE ACCIDENT RISKS NUCLEAR POWER PLANTS.

AND THE POTENTIAL FOR RISK REDUCTION. PEACH BOTTOM UNIT 2.Apperdcas. Draft For Comment.

KRAMER,G.

NUREG/CR4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR-4082 VOS: DEGRADED PIPING PROGRAM - PHASE RISKS AND THE POTENTIAL FOR RISK - REDUCTION GRAND ll Semiannual Report, AprSSeptember 1986.

L GULF, UNIT 1. Draft For Comment.

KULHOWVICK,0.

NUREG/CR4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR l

NUREG/CR-4082 VOS: DEGRADED PIPING PROGR/JA PHASE POSTULATED SEVERE ACCIDENTS: SEQUOYAH POWER II.Semiennual Report, Apr4 September 1986.

STATON, UNIT 1. Draft Report For Comment.

KULLSERQ,C.M.

LITTLEFIELD,R.

NUREG/CR-4845: AN ANALYSIS OF THE SEMISCALE MOD-2C S-NH-3 NUREG/CR-4583 V02: DEVELOPMENT AND VALIDATION OF A REAL.

TEST USING THE TRAC-PF1 COMPUTER PROGRAM.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER CTOR COMPONENTS. Annual Report, October 1984 September NUREG/CR-4551 V2 DRF: EVALUATON OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION.SEQUOYAH LLEWELLYN,P.

POWER STATION,0 NIT 1. Draft For Comment NUREG/CR4815: DEMONSTRATION TESilNG OF A SURYLILLANCE NUREG/CR-4651 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs AND THE POTENTIAL FOR RISK REDUCTION: PEACH And Benefits' BOTTOM. UNIT 2 Main Report. Draft For Comment.

NUREG/CR4551 V3 PT2: EVALUATION OF SEVERE ACCIDENT RISKS LOPEZ,E.A.

AND THE POTENTIAL FOR RISK REDUCTION. PEACH NUREG/CR 4875: CHARACTERIZATION OF CRUSHED TUFF FOR THE EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES NURE R4 1 4 D E

UA I F SEVERE ACCIDENT IN THE UNSATURATED ZONE.

RISKS AND THE POTENTIAL FOR RISK REDUCTION: GRAND GULF, UNIT 1. Draft For Comment LUCK,L NUREG/CR4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR NUREG/CR 4765: MXS CROSS-SECTION PREPROCESSOR USER'S POSTULATED SEVERE ACCIDENTS: SEQUOYAH POWER MANUAL STATION. UNIT 1. Draft Report For Comment.

KURTZ,R.J.

LUCO,J.E.

NUREG/CR4848: STEAM GENERATOR GROUP PROJECT. Annual NUREG/CR-4903 V03: SELECTION OF EARTHOUAKE RESISTANT DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-Report - 1985.

GY AND TECHNICAL CASES. Dislocation Models Of Near Source LACHANCE J.L, Earthqualte Ground Motion A Review.

NUREG/CR 4550 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN-CY FROM INTERNAL EVENTS GRAND GULF. UNIT 1. Main Report.

LYON,W.C.

NUREG/Cb4550 V06PT2. ANALYSIS OF CORE DAMAGE FREQUEN-NUREG 1269: LOSS OF RESIDUAL HEAT REMOVAL SYSTEM.Diablo CY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1. Appendices.

Canyon Unit 2, April 10.1987.

LANDOW,M.

MALONE,G.

NUREGICR.4082 V05: DEGRADED PlPING PROGRAM PHASE NUREG/CR4783: ANALYSIS OF BALANCE OF-PLANT REGULATORY 4

!! Semiannual Report, Apni-September 1986.

ISSUES. Final Report.

LAUGHERY,K.R.

MARSCHALL C.W, NUREG.1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION NUREG/CR-4082 V05: DEGRADED PIPING PROGRAM - PHASE FACILIHES CERTIFIED UNDER 10CFR55 Draft Report.

Bl. Semiannual Report, April-September 1986.

LAY,R.

MART,G.A.

NUREG/CR4783: ANALYSIS OF BALANCE OF-PLANT REGULATORY NUREG/CR-4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RELI-ISSUES. Final Report.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS. Semiannual Report, April. September 1986.

LEAHY T.J~

NUREG/CR 4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RELi-NUREG/CR-4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER FROM INTERNAL EVENTS: SEQUOYAH, UNIT 1.

REACTORS Semiannual Report, October 1986 March 1987.

LEE,0.W.

NUREG/CR 4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY MAKEY,W.

RIPRAP TESTING IN FLUMES. Phase 1.

NUREG/CR.4062 V05: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report. April September 1986.

LEE V.W.

NUREG/CR-4903 V01: SELECTION OF EARTHOUAKE RESISTANT MCCABE,0.E.

DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO.

NUREG/CR4841: FRACTURE EVALUATION OF SURFACE CRACKS GY AND TECHNICAL CASES. Direct Empincal Scaling Of Response EMBEDDED W REACTOR VESSEL CLADDING Unirramated Bend Socctral Amplitudes From Vanous Site And Earthouake Parameters.

Specimen Results.

NUHEG/CR-4903 V02: SELECTION OF EARTHQUAKE RESISTANT DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO.

MCELROY,W.N.

GY AND TECHNICAL CASES. Methods For introduction Of Geological NUREG/CR 3319 ROI: LWR PRESSURE VESSEL SURVEILLANCE DO-Data Irito Characterization Of Active Faults And Seismicity And-.

SIMETRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveil-lance Physics-Dosimetry Data Base Compendium.

LEIGH.C.D.

NUREG/CR-4307 V03 LWR PRESSURE VESSEL SURVEILLANCE DO-NUREG/CR4830; MELCOR VALIDATION AND VERIFICATION 1986 SIMETRY IMPROVEMENT PROGRAM.1986 Annual Report, October PAPERS.

1985 September 1986.

30 Personil Author Indzx EEISTER,H.

NIEMCZYK,S.J.

NUREG/CR-4719: COOLABILITY OF STRATIFIED 002 DEBAIS IN NUREG/CP 0078: PROCEEDINGS OF THE SYMPOSIUM ON CHEMI-SODIUM WITH DOWNWARD HEAT REMOVALThe D13 Expenment CAL PHENOMENA ASSOCIATED WITH RADIOACTIVITY RELEASES DURING SEVERE NUCLEAR PLANT ACCIDENTS.

KENSING,RW.

NUREG/CR 4885: SEISMIC HAZARD CHARACTERIZATION OF THE NITZ,K.C.

EASTERN UNITED STATES. Comparative Evaluation Of The LLNL And NUREG/CR4950 V01: THE SHOREUNE ENVIRONMENT ATMOS-EPRI Stuses PHERIC DISPERSION EXPERIMENT (SEADEX)Expenment Desenp-MERWIN,S.E.

NUREG/CR-4938: OCCUPATIONAL RADIATION EXPOSURES ASSOCl-NOWLEN,S.P.

ATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE DIS-NUREG/CR4679: QUANTITATIVE DATA ON THE FIRE BEHAVIOR OF POSAL COMBUSTIBLE MATERIALS FOUND IN NUCLEAR POWER PLANTS.A uterature Review.

MEYER,MA NUREG/CR-4681: ENCLOSURE ENVIRONMENT CHARACTERl2ATION NUREG/CR4814: SOURCES OF CORRELATION BETWEEN TESTING FOR THE BASE LINE VALIDATION OF COMPUTER FIRE l

EXPERTS Empincel Results From Two Extremes-SIMULATION CODES.

ME YER,0.R.

o DONNELL,E.

NUREG/CR-4617-ONSITE ASSESSMENTS OF THE EFFECTIVENESS NUREG/CR4918 V01: CONTROL OF WATER INFILTRATION INTO AND IMPACTS OF UPGRADED EMERGENCY OPERATING PROCE*

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report,0ctober 1985 DURES September 1986.

MILLER,R.W.

OPELKA,J.H.

NUREG/CR4866: AN ASSESSMENT OF HYDROGEN GENERATION NUREG/CR4964 UPDATE OF TABLE S-3 NONRADIOLOGICAL ENVb FOR THE PDF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

RONMENTAL PARAMETERS FOR A REFERENCE UGHT WATER MINARICK,J.W.

REACTOR Uranium Mining. Milling And Ennchment NUREG/CR4674 V03: PRECURSORS TO POTENTIAL SEVERE CORF OSETEK,D.J.

DAMAGE ACCIDENTS.1984.A STATUS REPORT Main Report And Ap.

NUREG/CR 4866: AN ASSESSMENT OF HYDROGEN GENERATION NU

/CR4 04: PRECURSORS TO POTENTIAL SEVERE CORE FOR THE PBF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

DAMAGE ACCIDENTS:1984,A STATUS REPORT. Appendixes C D And OTTINGER C.A.

I NUREG/CR 4719-COOLABluTY OF STRATIFIED 002 DEBRIS IN EITCHELL,0.W.

SODIUM WITH DOWNWARD HEAT REMOVAL.The 013 Expenment.

NUREG/CR-4719: COOLABILITY OF STRATIFIED LO2 DEBRIS IN PAIK C.Y SODIUM WITH DOWNWARD HEAT REMOVALThe D:3 Expenment-NUREG/CR4922: STEAM SEPARATOR MODELING FOR VARIOUS MOORE.D.L NUCLEAR REACTOR TRANSIENTS.

NUREG/CR-4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS. SEQUOYAH, UNIT 1.

PANGBURN,0.C' NUREG 1300- ENVIRONMENTAL STANDARD REVIEW PLAN FOR THE MORLEY,B.M.

REVIEW OF A LICENSE APPUCATION FOR A LOW LEVEL RADIO.

NUREG/CR-4950 VO1: THE SHORELINE ENVIRONMENT ATMOS.

ACTIVE WASTE DISPOSAL FACILITY.

PHERIC DISPERSION EXPERIMENT (SEADEX)Expenment Desenp-PAPASPYROPOULCS NUREG/CR-4082 V05. DEGRADED PIPING PROGRAM - PHASE MULLEN,G.

11 Semiannual Report, Apnl. September 1986.

NUREG/CR4922: STEAM SEPARATOR MODEUNG FOR VARIOUS NUCLEAR REACTOR TRANSIENTS.

PARKER.

MURFIN.W.B.

MANUAL NUREG/CR-4551 V2 DAF: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION SEQUOYAH PASUPATHI,V.

POWER STATION. UNIT 1. Draft For Comment.

NUREG/CR 4082 V05: DEGRADED PIPING PROGRAM PHASE NUREG/CR-4700 V2 DAF: CONTAINMENT EVENT ANALYSIS FOR ll.Somsannual Report, Apni-September 1986.

POSTULATED SEVERE ACCIDENTS: SEOUOYAH POWER STATION. UNIT 1. Draft Report For Comment.

PATENAUDE C MURPHY,E, TION OF THE PROBABILISTIC PERFORMANCE OF COMPLEX SYS-NUREG/CR-4912: DATING GROUND WATER AND THE EVALUATION TEMS.

OF REPOSITORIES FOR RADIOACTIVE WASTE.

PAULSEN.G.A.

MURRAY,R.C.

NUREG/CR-4098: SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL-NUREG/CP 0087:

SUMMARY

REPORT OF THE SYMPOSIUM ON OlS.

ERATED. AGED CLASS 1E BATTERY CELLS.

MIC AND GFOLOGIC SITING CRITERIA FOR NUCLt.2, K)WER PMNTS PELOQUIN,R.A.

NUREG/CR 2850 V05: POPULATION DOSE COMMITMENTS DUE TO NAKAGAKI,M.

RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR4082 V05: DEGRADED PIPING PROGRAM PHASE IN 1983.

11 Semiannual Report. Apnt September 1986.

p NARBUT,P.P.

NUREG/CR-4928: DEGRADATION OF NUCLEAR PLANT TEMPERA-NUREG-1269: LOSS OF RESIDUAL HEAT REMOVAL SYSTEM.Diablo TURE SENSORS.

Canyon Unit 2, April 10.1987.

NELSON,J.D.

NUREG-1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION NUREG/CR-4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY FACILITES CERTIFIED UNDER 10CFR55. Draft Report.

RIPRAP TESTING IN FLUMES. Phase L POLZER,W.L NEYMOTIN,LY, NUREG/CR-4875. CHARACTER!2ATION OF CRUSHED TUFF FOR THE NUREG/CR 4739 RAMONA-38 CALCULATIONS FOR BROWNS EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES FERRY ATWS STUDY.

IN THE UNSATURATED ZONE.

Personal Author index 31 NUREG/CR-4901: EFFECTS FROM INFLUENT BOUNDARY CONOl-RUFF,J.F.

TIONS ON TRACER MIGRATION AND GPATIAL VARIABILITY FEA-NUREG/CR-4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY TURES IN INTERMEDIATE SCALE EXPERIMENTS.

RIPRAP TESTING IN FLUMES Phase L POMEROY,P.W.

RUNOLE T.A.

NUREG/CR4822: BROAD BAND SEISMIC DATA ANALYSIS.Septernber NUREG/CR-4623. IN-SITU STRESS MEASUREMENTS IN THE 1984 September 1986.

EARTH *S CRUST IN THE EASTERN UNITED STATES.

PRATER.J.T.

SAHA,p, NUREG/CR4889: ZlRCALOY 4 OXIDATON AT 1300 TO 2400 DE-NUREG/CR-4739: RAMONA-38 CALCULATIONS FOR BROWNS NUREG/C -4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM WITH U02.

SAVY,J.B.

NUREG/CP 0087:

SUMMARY

REPORT OF THE SYMPOSlVM ON SEIS-NUR G 9

SS OF RESIDUAL HEAT REHOVAL SYSTEM.Diablo N

Canyon Urut 2, Ap tl 10,1987.

NUREG/CR 4885: SEISMIC HAZARD CHARACTERl2ATION OF THE EASTERN UNITED STATES. Comparative Evaluation Of T.e LLNL And PUGH,C.E.

EPRI Studies.

NUREG/CR-4219 V03 N2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM.Semaannual Progress Report For April-September 1986.

SCHAEFER,P.

NUREG/CR-4783: ANALYSIS OF BALANCE-OF. PLANT REGULATORY RASMUSSEN,T.C.

NUREG/CR4655: UNSATURATED FLOW AND TRANSPORT ISSUES Final Report.

THROUGH FRACTURED ROCt; RELATED TO HIGH LEVEL WASTE SCHULZ.R K REPOSITORIES Final Report - P5ase 11.

NUREG/CR-4918 V01: CONTROL OF WATER INFILTRATION INTO NEAR SURFACE LLW DISPOSAL UNITS. Annual Report,0ctober 1985 RAYMOND,R.

NUREG/CR-4875: CHARACTER 12ATION OF CRUSHED TUFF FOR THE

. September 1986.

EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES IN THE UNSATURATED ZONE.

NU d/CR-4082 V05: DEGRADED PIPING PROGRAM - PHASE Il Semiannual Report April-September 1986.

REECE,W.D.

NUREG/CR-4959: PERFORMANCE TESTING OF EXTREMITY DOSI-SC M PR l

NUREG/CR-4872: EXPERIMENTAL AND ANALYTICAL ASSESSMENT I

REED,A.W.

OF CIRCUMFERENTIALLY SURFACE CRACKED PIPES UNDER l

NUREG/CR-4719: COOLABILITY OF STRATIFIED 002 DEBRIS IN BENDING.

NUREG/CR4877: ASSESSMENT OF DESIGN BASIS FOR LOAD CAR-SODIUM WITH DOWNWARD HEAT REMOVAL.The D13 Expenment.

l RYING CAPACITY OF WELD-OVERLAY REPAIRS.

REESMAN.A.L NUREG/CR4936: AN INTEGRATED GEOLOGICAL. GEOPHYSICAL,AND SCOTT,W.8.

GEOCHEMICAL INVESTIGATION OF THE MAJOR FRACTURES ON NUREG/CR 4330 V03: REVIEW OF LIGHT WATER REACTOR REGU-I THE EAST SIDE OF THE NEW MADRID EARTHOUAKE ZONE.

LATORY REQUIREMENTS. Asessment Of Selected Reguietory Re-quirements That May Have Marginal importance To Risk:Postaccident REEVES.M.

Samphng System, Turbine Missiles, Combustible Gas ControlCharcoal NUREG/CA.3925 REV:

SWIFT 11 SELF. TEACHING Filters.

CURRICULUM.lliustrative Problems For The Sandia WasteIsolation J

Flow And Transport Model For Fractured Media.

SETH,S.

j NUREG/CR4783: ANALYSIS OF BALANCE.0F PLANT REGULATORY f

REID,LO.

ISSUES Final Report.

NUREG/CR4583 V02: DEVELOPMENT AND VALIDATION OF A REAL-TIME SAFT-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER SHAIKH.A.

REACTOR COMPONENTS. Annual Report. October 1984 September NUREG/CR 4851: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY 1985.

RIPRAP TESTING IN FLUMES. Phase 1.

RENIER,J.A.

SHAPIRO,8.J.

NUREG/CR 4758. A RETRAN MODEL OF THE CALVERT CLIFFS 1 NUREG/CR4550 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN-j PRESSURIZED WATER REACTOR FOR ASSESSING THE SAFETY CY FROM INTERNAL EVENTS: GRAND GULF UNIT 1. Main Report.

1 IMPLICATIONS OF CONTROI SYSTEMS.

NUREG/CR-4550 V06PT2 ANALYSIS OF CORE DAMAGE FREQUEN.

CY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1. Appendices.

RHODES,D.B.

NUREG/CR-4821: REACTOR COOLANT PUMP SHAFT SEAL STABILI.

SHILLENN,J.K.

TY DURING STATION BLACKOUT, NUREG/CR-4726: EVALUATION OF PROTECTIVE ACTION RISKS.

RfDKY,R.W.

SIMONEN,F.A.

NUREG/CR4918 V01: CONTROL OF WATER INFILTRATION INTO NUREG/CR 3231: PIPE-TO-PIPE IMPACT PROGRAM.

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report,0ctober 1985 NUREG/CR 4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) REll.

- September 1986.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS. Semiannual Report. April-September 1986.

ROSERTS,F.P.

NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RELl-NUREG/CH-4773; DESIGN NEATURES TO FACILITATE INTERNATION-ABILITY FOR INSERVICE INSPECTION OF Ur2HT WATER AL SAFEGUARDS AT MIXED-OxlDE CONVERSION FACILITIES.

REACTORS. Semiannual Report, October 1986 - Maxr 17.

ROBERTS,G.C.

SINGH,M.M.

NUREG/CR-4819 V01: AGING AND SERVICE WEAR OF SOLENOID-NUREG/CR 4623: IN SITU STRESS MEASUREMENTS IN THE OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR EARTH'S CRUST IN THE EASTERN UNITED STATES.

POWER PLANTS. Volume 1. Operating Exponence And Failure identifi-

cation, SKINNER,L NUREG/CR-4783. ANALYSIS OF BALANCE OF PLANT REGULATORY ROSS,P A.

ISSUES. Final Report.

NUREG 0020 V10 N11: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT.Deta As Of October 31,1986 (Gray Book f)

SLOVlK,G.C.

NUREG-0020 V10 N12: LICENSED OPERATING REACTORS STATUS NUREG/CR-4739: RAMONA 38 CALCULATIONS FOR BROWNS

SUMMARY

REPORT. Data As Of November 30,1986.(Gray Book I)

FERRY ATWS STUDY.

~

1 32 P:rson:1 Authtr index SMITH.LN.

NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) REL' NUREG/CR 4551 V3 PT1: EVALUATION OF SEVERE ACCOENT RISKS ABluTY - FOR INSERVICE 1NSPECTION OF LIGHT WATER AND THE POTENTIAL FOR RISK REDUCTION. PEACH REACTORS. Semiannual Report October 1986. March 1987.

BOTTOM,0 NIT 2 Main Report Draft For Comment.

NUREG/CR.4551 V3 PT2: EVALUATON OF SEVERE ACCOENT RISKS T1PQ,P.

AND THE POTENTIAL FOR RISK REDUCTON-PEACH NUREG 1270 V01: INTERNATIONAL CODE ASSESSMENT AND APPU-VAiO SEVERE ACCIDENT CATIONS PROGRAM. Annual Report.

NU E /CF 1 V DR E NUREG 1271: GUIDEUNES AND PROCEDURES FOR THE INTERNA-RISKS AND THE POTENTIAL FOR RISK HEDUCTON: GRAND GULF,0 NIT 1. Draft For Comment.

TONAL CODE ASSESSMENT AND APPLICATIONS PROGRAM.

SMITH.M.H.

TOKUNAGA.T.K.

NUREG/CR-4921: ENGINEERING AND QUALITY ASSURANCE COST NUREG/CR4161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL FACTORS ASSOCIATED WITH NUCLEAR PLANT MOOlFICATION.

WASTE REPOSITORY. Volume 2: Tuff.

SMITH.O.L TOMAN G.J.

NUREG/CR 4758. A RETRAN MODEL OF THE CALVERT CLIFFS.1 NUREG/CR-4819 V01: AGING AND SERVICE WEAR OF SOLENOO-PRESSURIZED WATER REACTOR FOR ASSESSING THE SAFETY OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR IMPUTATIONS OF CONTROL SYSTEMS.

POWER PLANTS. Volume 1. Operating Exponence And Failure idenbfi-SOSERANO,F.T.

caten.

NUREG/CR.3956: IN SITU TESTING OF THE SHIPPINGPORT ATOMIC POWER STATON ELECTRICAL CIRCulTS.

TRAMMELL.C.M.

NUREG 1269: LOSS OF RESIDUAL HEAT REMOVAL SYSTEM.Diablo SOLDAT,J.K.

Canyon Unit 2, April 10,1987.

NUREG/CR 4653. GASPAR 11 - TECHNICAL REFERENCE AND USER GUIDE.

TRIFUNAC M.D.

NUREG/CR 4903 V01: SELECTION OF EARTHOUAKE RESISTANT SPANNER,J.C.

DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-NUREG/CR4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) REU-ABluTY FOR INSERVICE INSPECTON OF UGHT WATER GY AND TECHNICAL CASES. Direct Empincal Scahno Of Response REACTORS. Semiannual Report April-September 1986.

Spectral Amphtudes From Varcus Site And Earthqualte Parameters.

NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RELi.

NUREG/CR4903 V02: SELECTION OF EARTHOUAKE RESISTANT ABluTY FOR INSERVICE INSPECTION OF UGHT WATER DESON CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO.

REACTORS. Semiannual Report, October 1986 March 1987.

GY AND TECHNICAL CASES. Methods For Introduction Of Geological Data into Characternation Of Active Faults And Seismicity And.

NUREG/CR-4615 V02: MODEUNG STUDY OF SOLUTE TRANSPORT TSAO,L i

N

/C.4 FF F Mi LU N OMDARYCONDI.

^

STE RE Si RY' e2T '

TlONS ON TRACER MIGRATION AND SPATIAL VARIABILITY FEA-TURES IN INTERMEDIATE SCALE EXPERIMENTS.

UTHE,E.E.

STEARNS.R.G.

NUREG/CR-4950 V01: THE SHOREUNE ENVIRONMENT ATMOS-j NUREG/CR-4936. AN INTEGRATED GEOLOGICAL.GEOPHYSICALAND PHERIC DISPERSION EXPERIMENT (SEADEX). Experiment Descrip-

]

GEOCHEMICAL INVESTIGATION OF THE MAJOR FRACTURES ON ton.

THE EAST SIDE OF THE NEW MADRID EARTHOUAKE ZONE.

VAN FLEET,LO.

STOETZEL,0.A.

NUREG/CR-4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) REU-NUREG/CR4330 V03: REVIEW OF UGHT WATER REACTOR REGU*

ABluTY FOR INSERVICE INSPECTION OF LIGHT WATER LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-REACTORS. Semiannual Report. April. September 1986.

querements That May Have Marginal importance To Risk.Pustaccusent NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) REU.

Sampling System, Turtune Missiles,Combushbie Gas Control. Charcoal ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Fitters.

REACTORS. Semiannual Report, October 1986. March 1987.

STRENGE,D.L YO T.V NUR CR-4651 GASPAR 11 TECHNICAL REFERENCE AND USER LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-SULLIVAN,R.

Quirements That May Have Marginal importance To Risk:Postaccident NUHEG/CR-4783. ANALYSIS OF BALANCE OF-PLANT REGULATORY Sampling System Turbine Missues. Combustible Gas Control. Charcoal ISSUES Final Report.

Fitters.

SUTTER,8.L WACHTEL.J.

NUREG/CR4779: NEW DATA FOR AEROSOLS GENERATED BY RE' NUREG.1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION LEASES OF PRESo lRIZED POWDERS AND SOLUTIONS IN STATIC FACluTIES CERTIFIED UNDER 10CFR55. Draft Report.

AIR.

SUTTON.O.H.

WAGONER,S.R.

NUREG/CR-4822; BROAD BAND SEISMIC DATA ANALYSIS September NUREG/CR4185: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO.

1984 - September 1986.

GRAM - ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA-TIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

ST;AIN A.D.

NUREG/CR 4772: ACCOENT SEQUENCE EVALUATON PROGRAM WARD.D.S.

HUMAN REUABluTY ANALYSIS PROCEDURE.

NUREG/CR.3925 REV:

SWIFT ll SELF. TEACHING CURRICULUM. Illustrative Problems For The Sandia Waste-Isolation URE /C 30 V03: REVIEW OF UGHT WATER REACTOR REGU.

Flow And Transport Model For Fractured Media.

LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-WATERMAN,D.K NUREG 12 4 R0 : H STOR L ATA U MARY F THE SYSTEMAT-ng S m no M s le as trol. h i

Filters.

TAYLOR,T.T.

WATKINS,J.C.

NUREG/CR4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RELI.

NUREG/CR-4802: AN EVALUATION OF TRAC PF1/ MOD 1 COMPUTER ADILITY FOR INSERVICE INSPECTION OF UGHT WATER CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE.

REACTORS. Semiannual Report, April-September 1986.

MISCALE MOD 2C FEEDWATER UNE BREAK TRANSIENTS.

P rsorti Author index 33 WEISS A.J.

W1LKQWSK),0.M.

NUREG/CR 2331 V06 N1 SAFETY RESEARCH PROGRAMS SPON.

NUREG/CR4082 V05: DEGRADED P1 PING PROGRAM PHASE SORED BY OFFICE OF NUCLEAR REGULATORY ll. Semiannual Report, April.Septemt>er 1986.

l RESEARCH Ouarterty Progress Report, July Septemtsor 1986.

k WENSEL,RG NUREG/CFi4821: REACTOR COOLANT PUMP SHAFT SEAL STABILI.

NUREG/CA-4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION SEOUOYAH TV DURING STATION BLACKOUT ^

FOWER STATION UNIT 1. Draft For Comment.

WHEATLEY,P.D.

NUREG/CR4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR.4165: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO.

RISKS AND THE POTENTIAL FOR RISK REDUCTION. GRAND GRAM ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA.

GULF, UNIT 1.Drett For Comment.

110NS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

WITTLER,R.J.

WHEELER,T.A.

NUREG/CR-4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY NUREG/CR4550 VO6PT1: ANALYSIS OF CORE DAMAGE FREQUEN-RIPRAP TESTING IN FLUMES. Phase 1.

CY FROM INTERNAL EVENTS GRAND GULF. UNIT 1 Main Report.

NUREG/CR-4550 V06PT2 ANALYSIS OF CORE DAMAGE FREQUEN.

WITZlG.W.F.

CY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1. Appendices.

NUREG/CR4726: EVALUATION OF PROTECTIVE ACTION RISKS.

WHITE,J.R NUREG/CR-4815-DEMONSTRATION TESTING OF A SURVEILLANCE WOLLENBERG.H.A.

ROOOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs NUREG/CRJ161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL And Donefits WASTE REPOSITORY. Volume 2. Tuff.

WILKOWSKl.G.

ZIEGLER,E.J.

NUREG/CR4870 ANALYSIS OF EXPERIMENTS ON STAINLESS NUREG/CR-4921: ENGINEERING AND QUALITY ASSURANCE COST STEEL FLUX WELDS. Topical Report.

FACTO 9S ASSOCtATED WITH NUCLEAR PLANT MODIFICATION I

l l

Subject index This index was developed from keywords and word strings in titles and abstracts. During this development period, there will be some redundancy, which will be removed later when a rea-sonable thesaurus has been developed through experience. Suggestions for improvements

{

are welcome.

1 i

i 10CFR55 Aged Colle Nf jREG 1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION NUREG/CR-4098: SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL.

FACILITIES CERTIFIED UNDER 10CFR55. Draft Report.

ERATED AGED CLASS 1E BATTERY CELLS.

ACRS Reporte gging NUREG 1125 V08: A COMPILATION OF REPORTS OF THE ADVISORY NUREG/CR4819 V01: AGING AND SERVICE WEAR OF SOLENOID-l COMMITTEE ON REACTOR SAFEGUARDS,1986.

OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR 1

AEOO POWER PLANTS Volume 1.Operatino Expenence And Failure identifi-NUREG-1272: REPORT TO THE U.S NUCLEAR REGULATORY COM-cation.

MISSION ON ANALYSIS AND EVALUATION OF OPERATIONAL DA(A 1986.

Air Cleaning NUREG/CP-0086 VO1: PROCEEDINGS OF THE 19TH DOE /NRC NU-ATWS CLEAR AIR CLEANING CONFERENCE. Held in l

NUREG/CR-4165: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-Seattle Washington. August 18-21,1986.

)

GRAM - ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA-NUREG/CP 0086 V02: PROCEEDINGS OF THE 19TH DOE /NRC NU-1 NI 73 : RAMONA B C

S OI BROWNS att,Washe ton,Augus 21 1986.

FERRY ATWS STUDY.

Airborne Releases l

Abnormel Occurrences NUREG 0090 V09 NO3 REPORT TO CONGRESS ON ABNORMAL NUREG/CR-4779. NEW DATA FOR AEROSOLS GENERATED BY RE-OCCURRENCES. July September 1986.

LEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC AIR.

Abstract l

NUREG-0304 V12 N01: REGULATORY AND TECHNICAL REPORTS Analysis And Evaluation Of Operational Data (ABSTRACT INDEX JOURNAL). Compilation For First Ouarter NUREG 1272: REPORT TO THE U S. NUCLEAR REGULATORY COM-1967, January-March.

MISSION ON ANALYSIS AND EVALUATION OF OPERATIONAL DATA %86.

Accident Analysis NUREG/CR-4765: MXS CROSS SECTION PREPROCESSOR USER'S Annual Report NUREG-1145 V03: U.S. NUCLEAR REGULATORY COMMISSION 1986 Acc6 dent Scenerlo ANNUAL REPORT.

NUREG/CR-4841: FRACTURE EVALUATION OF SURFACE CRACKS I

EMBEDDED IN REACTOR VESSEL CLADDING Untrradiated Bend Anticipated Transient Without Scram i

Specimen Results.

NUREG/CR-4165: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-GRAM - ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA.

Accident Sequence TIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

NUREG/CR-4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS: SEOUOYAH. UNIT 1-Atmospheric Dispersion NUREG/CR 4550 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR-4950 V01: THE SHORELINE ENVIRONMENT ATMOS-PHERIC DISPERSION EXPERIMENT (SEADEX).Expenment Desenp-NU EG VO T2 A A IS F A

Al A UEN-D"-

CY FROM INTERNAL EVENTS: GRAND GULF. UNIT 1.

ndices.

NUREG/CR-4874 V03: PRECURSORS TO POTENTIAL VERE CORE GE A DENTS 1984,A STATUS REPORT. Main Report And Ap-NUREG/CR-4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS NUREG/CR-4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE AND THE POTENTIAL FOR RISK REDUCTION. PEACH DAMAGE ACCIDENTS:1984,A STATUS REPORT. Appendixes C,0 And BOTTOM. UNIT 2. Main Report. Draft For Comment.

E.

NUREG/CR-4551 V3 PT2: EVALUATION OF SEVERE ACCIDENT RISKS NUREG/CR-4772: ACCIDENT SEQUENCE EVALUATION PROGRAM AND THE POTENTIAL FOR RISK REDUCTION PEACH HUMAN RELIABILITY ANALYSIS PROCEDURE.

BOTTOM. UNIT 2. Appendices. Draft For Comment.

NUREG/CR4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT UE SO2 PROGRAMMATIC ENVIRONMENTAL IMPACT GULF UNI. Draft For mm t STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL OF NUREG/CR-4700 V4 DRF: CONTAINMENT EVENT ANALYSIS FOR RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 ACCl-POSTULATED SEVERE ACCIDENTS: GRAND GULF NUCLEAR DENT AT THREE MILE ISLAND NUCLEAR STATION, UNIT 2.Fenal STATION, UNIT 1. Draft For Comment.

Supplement Dealing With Disposal Of...

NUREG/CR4739: RAMONA38 CALCULATIONS FOR BROWNS FERRY ATWS STUDY.

Acoustic Emmiselon/ Flow Relationship NUREG/CR4300 V04 N1: ACOUSTIC EMISSION / FLAW RELATION-Bolence-Obmant SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR4781. ANALYSIS OF BALANCE OF-PLANT REGULATORY VESSELS. Progress Report, October 1986 March 1987.

ISSUES. Final Report.

Aerosol NUREG/CR-4779. NEW DATA FOR AEROSOLS GENERATED BY RE.

Bettery Cell LEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC NUREG/CR-4098: SEISMIC FRAGILITY TESTS OF NEW AND ACCEL-AIR.

ERATED-AGED CLASS 1E BATTERY CELLS.

35

36 Subj:CtInd:x pendMg Sampling System. Turtwne Missdes. Combustible Gas Control, Charcoal NUREG/CR4872 EXPERIMENTAL AND ANAL,YTICAL ASSESSMENT Fdters.

OF CIRCUMFERENTIALLY SURFACE-CRACKED PIPCS UNDER BE NDING Combustible Material NUREG/CR-4679: QUANTITATIVE DATA ON THE FIRE BEHAVIOR OF Bolling water Reactor COMBUSTIBLE MATERIALS FOUND IN NUCLEAR POWER NUREG/CR-4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS PLANTS.A Literature Rewsw AND THE POTENTIAL FOR RISK REDUCTION PEACH BOTTOM, UNIT 2 Main Refort Draft For Comment Component Fragutty I

NUREG/CR 4551 V3 PT2. cVALUATION OF SEVERE ACCIDENT RISKS NUREG/CR4899.

COMPONENT FRAGILITY RESEARCH AND THE POTENTIAL FOR RISK REDUCTION PEACH BOTTOM. UNIT 2 Appendeces Draft For Cornment.

PROGRAM Phase l Component Pnontastion.

NUREG/CR 4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT Conta6nment Event Analysle RISKS AND THE POTENTIAL FOR RISK REDUCTION GRAND NUREG/CR4700 V2 DRF CONTAINMENT EVENT ANALYSIS FOR NURE 700 V4 R TAINMENT EVENT ANALYSIS FOR POSTULATED SEVERE ACCIDENTS. SEOUOYAH POWER b

S.

GRAND GULF NUCLEAR SA,0 1 an

• g DRF T EVENT ANALYSIS FOR iO T1 or E

NUREG/CR 4739 RAMONA-38 CALCULATIONS FOR BROWNS POSTULATED SEVERE ACCIDENTS. GRAND GULF NUCLEAR FERRY ATWS STUDY.

STATION. UNIT 1. Draft For Comment.

Eudget Containment Spray NUREG 1100 V03 ADO: BUDGET ESTIMATES. Fiscal Years 1988-1989 NUREG 0800 06 5 2 R2: STANDARD REVIEW PLAN FOR THE REVIEW I

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER I

Curled Waste PLANTS LWR Edition. Proposed Revision 2 To Section 6.5.2, "Contain-NUREG/CR-4918 V01: CONTROL OF WATER INFILTRATION INTO ment Spray As A Fission Product Cleanup System." For Comment.

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report,0ctober 1985 September 1986.

Control Systems NUREG/CR 4758: A RETRAN MODEL OF THE CALVERT CLIFFS.1

)

UR G/CR4679 QUANTITATIVE DATA ON THE FIRE BEHAVIOR OF AbN b l

COMBUSTIBLE MATERIALS FOUND IN NUCLEAR POWER IMPLICAT!ONS OF CONTROL SYSTEMS.

J PLANTS A Literature Review Core Damage I

Ctbenet Fire Test NUREG/CR 4550 V05: ANALYSIS OF CORE Di MAGE FREQUENCY I

j NUREG/CR-4527 V01: AN EXPERIMENTAL INVESTIGATION OF IN.

FROM INTERNAL EVENTS: SEOUOYAH. UNIT 1 i

TERNALLY IGNITIED FIRES IN NUCLEAR POWER PLANT CONTROL NUREG/CR-4550 V00PT1: ANALYSIS OF CORE DAMAGE FREQUEN-l CABINETS Part 1 Cabinet Effects Tests CY FROM INTERNAL EVENTS GRAND GULF.L' elT 1. Main Report.

NUREG/CR4550 V06PT2: ANALYSIS OF CORE DAMAGE FREQUEN-Carbon Steel CY FROM INTERNAL EVENTS GRAND GULF. UNIT 1. Appendices NUREG/CR-4872. EXPERIMENTAL AND ANALYTICAL ASSESSMENT NUREG/CR-4674 V03: PRECURSORS TO POTENTIAL SEVERE CORE OF CIRCUMFERENTIALLY SURFACE CRACKED PIPES UNDER DAMAGE ACCIDENTS.1984 A STATUS REPORT. Main Report And Ap-BENDING-pendmes A And B.

NUREG/CR4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE NU E /CP 8 F ROCEEDINGS OF THE SYMPOSIUM ON CHEMI.

l CAL PHENOMENA ASSOCIATED WITH RADIOACTIVITY RELEASES DURING SEVERE NUCLEAR PLANT ACCIDENTS Correlation Services Circuits NUREG/CR 4814:

SOURCES OF CORRELATION BETWEEN NUREG/CR 3956. IN SITU TESTING OF THE SHIPPINGPORT ATOMIC EMMEmpincel Nsda Rom M Emmes.

POWER STATION ELECT AICAL CIRCUITS.

Cost CI:dding NUREG/CR4815: DEMONSTRATION TESTING OF A SURVEILLANCE NUREG/CR-4841: FRACTURE EVALUATION OF SURFACE CRACKS ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs EMBEDDED IN REACTOR VESSEL CLADDING Unirradiated Bend And Benefits.

Specimen Results.

Cla:s IE NURES/CR-4921: ENGINEERING AND OUALITY ASSURANCE COST NUREG/CR-4098: SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL.

FACTORS ASSOCIATED WITH NUCLEAR PLANT MODIFICATION.

ERATED-AGED CLASS 1E BATTERY CELLS.

Cracked Pipe Cloieout NUREG/CR4082 VO5: DEGRADED PIPING PROGRAM - PHASE NUREG/CR 4663. CLOSEOUT OF IE BULLETIN 83 01 FAILURE OF RE.

Il Semiannual Report, Apni September 1986.

ACTOR TRIP BREAKERS (WESTINGHOUSE DB 50) TO OPEN ON NUREG/CR-4872: EXPERIMENTAL AND ANALYTICAL ASSESSMENT AUTOMATIC TRIP SIGNAL OF CIRCUMFERENTIALLY SURFACE-CRACKED PIPES UNDER NURE G/CR4664. CLOSEOUT OF IE BULLETIN 83-04 FAILURE OF BENDING.

THE UNDERVOLTAGE TR@ UNCTION OF REACTOR TRIP BREAK.

ERS.

Debria Cootability Co ( Fuel Cycle NUREG/CR-4719: COOLABILITY OF STRATIFIED UO2 DEBRIS IN NUREG-0332: POTENTIAL HEALTH AND ENVIRONMENTAL lMPACTS SODIUM WITH DOWNWARD HEAT REMOVALThe D13 Expenment.

ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL CYCLES Final Rep rt.

Decay Heat Removel System NUREG-1269. LOSS OF RESIDUAL HEAT FEMOVAL SYSTEM Diablo Code Assessment Canyon Unit 2, April 10,1987.

NUREG-1270 V01: INTERNATIONAL CODE ASSESSMENT AND APPLl-CATIONS PROGRAM Annual Report.

Decommissioning NUREG-1271: GUIDELINES AND PROCEDURES FOR THE INTERNA.

NUREG-0904 S01: DRAFT SUPPLEMENT TO THE FINAL ENVIRON-TlONAL CODE ASSESSMENT AND APPLICATIONS PROGRAM.

MENTAL STATEMENT RELATED TO THE DECOMMISSIONING OF THE RARE EARTHS FACILfTY, WEST CHICAGO,lLLINOIS Docket No.

Combustible Gas Control 40 2061.(Kerr-McGee)

NUREG/CR4330 V03: REVIEW OF LIGHT WATER REACTOR REGU-NUREG-1166: FINAL ENVIRONMENTAL STATEMENT FOR DECOM.

LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-MISSIONING HUMBOLDT BAY POWER PLANT, UNIT 3. Docket No.

Querements That May Have Marginal Importance To Assk Postaccident 50-133 (Pacific Gas And Electnc Company)

q j

l i

i Subject index 37

)

Decontamination Electrical in Situ Testing NUREG-0683 SO2: PROGRAMMATIC ENVIRONMENTAL IMPACT NUREG/CR-3950: IN SITU TESTING OF THE SHIPPINGPORT ATOMIC STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL OF POWER STATION ELECTRICAL CIRCUITS.

RADIOACTIVE WASTES RESULitNG FROM MARCH 28.1979 ACCl-DENT AT THREE MILE ISLAND NUCLEAR STATION. UNIT 2 Final Emergency Core Cooling System Supplement Doah;)g With Disposal Of...

NUREG 1230 DRFT FC: COMPENDIUM OF ECCS RESEARCH FOR RE-

1. U3

^

80 Degradauon l

NUREG/CR4819 V01: AGING AND SERVICE WEAR OF SOLENOID-Emergency Operating Procedures OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCL EAR NUREG/CR 4617: ONSITE ASSESSMENTS OF THE EFFECTIVENESS POWER PLANTS Volume 1. Operating Exponence And Failure identifi-AND IMPACTS OF UPGRADED EMERGENCY OPERATING PROCE-NU E /CH-4928. DEGRADATION OF NUCLEAR PLANT TEMPERA-DURES.

TURE SENSORS.

Emergency Respense Degraded Piping NUREG/CR4726. EVALUATION OF PRoldCTIVE ACTION RISKS.

NUREG/CR4082 V05. DEGRADED PIPtNG PROGRAM PHASE II. Semiannual Report, Aprd September 1986.

Enforcement Action J

NUREG 0940 V06 NO1: ENFORCEMENT ACTIONS:SIGNIFICANT AC.

1 Design Basle TlONS RESOLVED.Ouarterty Progress Report. January-March 1987.

I NUREG/CR-4877: ASSESSMENT OF DESIGN BASIS FOR LOAD CAR-RYING CAPACITY OF WELD-OVERLAY REPAIRS.

Engineering Cost NUREG/CR4921: ENGINEERING AND OUALITY ASSURANCE COST U EG-0386 D04 ROS: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July Environmental Assessment 1972 September 1988.

NUREG-1239: ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF Dispersion SOURCE MATERIAL LICENSE NO. STB 401. Docket No. 40-65631Co-NUREG/CR-4950 VOI: THE SHORELINE ENVIRONMENT ATMOS.

lun:biurn-Tantalum Division.Malknckrodt,Inc.)

RIC DISPERSION EXPERIMENT (SEADEX)Expenment Dwre Environmental impact NUREG 0332: POTENTIAL HEALTH AND ENVIRONMENTAL IMPACTS Disposal Of Radioactive Waste ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL CYCLES. Final NUREG4683 S02. PROGRAMMATIC ENVIRONMENTAL IMPACT Report.

l STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL OF l

RADIOACTIVE WASTES RESULTING FROM MARCH 28.1979 ACCl-Enytronmental impact Statement DENT AT THREE MILE ISLAND NUCLEAR STATION. UNIT 2. Final NUREG 0683 S02: PROGRAMMATIC ENVIRONMENTAL IMPACT Supplement Deakng with Disposal Of....

STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL OF RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 ACCl-NR 124 1: RADIOACTIVE WASTE MANAGEMENT RESEARCH PROGRAM PLAN FOR HIGH-LEVEL WASTE 1987.

Supplement Deahng With Disposal Of....

Doolmeters Environmental Paramotare NUREG/CR 4959. PERFORMANCE TESTING OF EXTREMITY DOSI-NUREG/CR4964: UPDATE OF TABLE S-3 NONRADIOLOGICAL ENVI-1 METERS RONMENTAL PARAMETERS FOR A REFERENCE LIGHT WATER

]

REACTOR Uranium Mining Milhng And Ennehment.

Dyersburg Line j

NUREG/CR-4936. AN INTEGRATED GEOLOGICAL. GEOPHYSICAL,AND Enytronmental Qualification GEOCHEMICAL INVESTIGATION OF THE MAJOR FRACTURES ON NUREG/CR4819 VOI: AGING AND SERVICE WEAR OF SOLENOID-THE EAST SIDE OF THE NEW MADRID EARTHOUAKE ZONE.

OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR ECCAD System c

n' NUREG/CR-3956: IN SITU TESTING OF THE SHIPPINGPORT ATOMIC POWER STATION ELECTRICAL CIRCUITS.

Environmental Standard Review Plan ECCS NUREG-1300- ENVIRONMENTAL STANDARD REVIEW PLAN FOR THE NUREG 1230 DRFT FC COMPENDIUM OF ECCS RESEARCH FOR RE-REVIEW OF A LICENSE APPLICATION FOR A LOW LEVEL RADIO-ALISTIC LOCA ANALYSIS. Draft Report For Comment.

ACTIVE WASTE DISPOSAL FACILITY.

EYNTRE Environmental Statement NUREG/CR4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR NUREG 1300. ENVIRONMENTAL STANDARD REVIEW PLAN FOR THE POSTULATED SEVERE ACCIDENTS: SEQUOYAH POWER REVIEW OF A LICENSE APPLICATION FOR A LOW LEVEL RADIO-STATION. UNIT 1. Draft Report For Comment.

ACTIVE WASTE DISPOSAL FACILITY.

NUREG/CR-4700 V4 DRF: CONTAINMENT EVENT ANALYSIS FOR POSTULATED SEVERE ACCIDENTS: GRAND GULF NUCLEAR Evacuation Risk STATION, UNIT 1. Draft For Comment.

NUREG/CR-4726: EVALUATION OF PROTECTIVE ACTION RISKS.

Earthquake Engineering Evaluation Procedure NUREG/CR-4903 V01: SELECTION OF EARTHOUAKE RESISTANT NUREG-1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-FACILITIES CERTIFIED UNDER 10CFR55. Draft Report.

GY AND TECHNICAL CASES. Direct Empincal Scahng Of Response Spectral Amplitudes From Vanous Site And Earthquake Parameters-Experts NUREG/CR-4903 V02: SELECTION OF EARTHOUAKE RESISTANT NUREG/CR4814'-

SOURCES OF CORRELATION BETWEEN DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-GY AND TECHNICAL CASES. Methods For introduction Of Geological EXPERTS Empincal Results from Two Extremes.

Data into Charactertration Of Active Faults And Seismicrty And....

NUREG/CR-4903 V03: SELECTION OF EARTHOUAKE RESISTANT ExtremitI Doelmeters DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO-NUREG/CR-4959: PERFORMANCE TESTING OF EXTREMITY DOSI-GY AND TECHNICAL CASES. Dislocation Models Of Near-Source METERS.

Earthquake Ground Motion. A Review.

Failure Identification Earthquake Zone NUREG/CR-4819 VOI: AGING AND SERVICE WEAR OF SOLENOID-NUREG/CR4936. AN INTEGR ATED GEOLOGICAL. GEOPHYSICAL,AND OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR GEOCHEMICAL INVESTIGATION OF THE M.*JOR FRACTURES ON POWER PLANTS. Volume 1. Operating Expenence And Failure identifi-THE EAST SIDE OF THE NEW MADRID EARTHOUAKE ZONE.

cation.

38 Subj:ct Indix Feedwater Line Brook NUREG/CR4964: UPDATE OF TABLE S-3 NONRADIOLOGICAL ENVl-NUREG/CR 4802: AN EVALUATION OF TRAC PF1/ MOD 1 COMPUTER RONMENTAL PARAMETERS FOR A REFERENCE LIGHT WATER CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE-REACTOR Urarnum Mining. Milling And Ennchment.

MISCALE MOD 2C FEEDWATER UNE BREAK TRANSIENTS.

Fuel Damage Final Environmental Statement NUREG/CR4866: AN ASSESSMENT OF HYDROGEN GENERATION NUREG-0904 S01: DRAFT SUPPLEMENT TO THE FINAL ENVIRON-FOR THE PBF SEVERf FUEL DAMAGE SCOPING AND 11 TESTS.

MENTAL STATEMENT RELATED TO THE DECOMMISSIONING OF THE RARE EARTHS FACILITY, WEST CHICAGO.lLLINOIS. Docket No.

GASPARll 40 2061.(Kerr.McGee)

NUREG/CR-4653: GASPAR 11 - TECHNICAL REFERENCE AND USER NUREG-1166; FINAL ENVIRONMENTAL STATEMENT FOR DECOM-GUIDE.

MISSIONING HUMBOLDT BAY POWER PLANT, UNIT 3. Docket No.

50133 (Pacific Gas And Electnc Company)

Geological Deformation NUREG/CR-4903 V02: SELECTION OF EARTHOUAKE RESISTANT Fire DESIGN CRITERIA FOR NUCLEAR POWER PLANTS. METHODOLO-NUREG/CR-4527 V01: AN EXPERIMENTAL INVESTIGATION OF IN-GY AND TECHNICAL CASES. Methods For introduction Of Geological TERNALLY IGNITIED FIRES IN NUCLEAR POWER PLANT CONTROL Data into Charactenration Of Active Faults And Seis:nicity And...

CABINETS Part 1 Cabinet Effects Tests.

NUREG/CR-4679. QUANTITATIVE DATA ON THE FIRE BEHAVIOR OF Ground Motion COMBUSTIBLE MATERIALS FOUND IN NUCLEAR POWER NUREG/CRJ903 V03: SELECTION OF EARTHOUAKE RESISTANT PLANTS A Lnerature Review.

DESIGN CRITERIA FOR NUCLEAR POWER PLANTS. METHODOLO-NUREG/CR-4681: ENCLOSURE ENVIRONMENT CHARACTERIZATION GY AND TECHNICAL CASES. Dislocation Models Of Near Source TESTING FOR THE RASE LINE VALIDATION OF COMPUTER FIRE Earthquake Ground Motion. A Revow.

SIMULATION CODES.

Ground Water Dating Fiscal Year NUREG/CR-4912. DATING GROUND WATER AND THE EVALUATION h0 REG-1100 V03 ADD: BUDGET ESTIMATES Fiscal Years 1988-1989.

OF REPOSITORIES FOR RADIOACTIVE WASTE.

I Fission Product Cleanup System Health Fffecte i

NUREG-0800 06.5.2 R2. STANDARD REVIEW PLAN FOR THE REVIEW NUREG 0332: POTENTIAL HEALTH AND ENVIRONMENTAL IMPACTS l

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL CYCLES. Final PLANTS LWR Editon. Proposed Revision 2 To Section 6.5.2, "Contain-ment Sprav As A Fisson Product Cleanup System." For Comment.

NUREG-oa06 06.5.5 RO: STANDARD REVIEW PLAN FOR THE REVIEW Heat Decay I

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR-4719: COOLABILITY OF STRATIFIED 002 DEBRIS IN PLANTS. LWR Editon Proposed Rev: won 0 To New SRP Secten 6.5.5, SOOlUM WITH DOWNWARD HEAT REMOVALThe D13 Experiment.

i Pressure Suppresson Pools As Fission Product Clean Up Systems."

l For Comment Heat Of Reaction NUREG/CR-4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM Fiume NUREG/CR 4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY RIPRAP TESTING IN FLUMES Phase I.

Heat Removal NUREG/CR 4719: COOLABILITY OF STRATIFIED 002 DEBRIS IN Flux-Welded Pipe SODIUM WITH DOWNWARD HEAT REMOVALThe D13 Experiment.

NUREG/CR 4878: ANALYSIS OF EXPERIMENTS ON STAINLESS STEEL FLUX WELDS. Topical Report.

Heat Transfer Fracture Mechanica NUREG/CR-4758: A RETRAN MODEL OF THE CALVERT CLIFFS-1 PRESSURIZED WATER REACTOR FOR ASSESSING THE SAFETY NUREG/CR-4082 VOS: DEGRADED PIPING PROGRAM PHASE IMPLICATIONS OF CONTROL SYSTEMS.

ll. Semiannual Report. Apni-September 1986'N NUREG/CR 4219 V03 N2: HEAVY SECTIO STEEL TECHNOLOGY NUREG/CR4830: MELCOR VALIDATION AND VERIFICATION 1986 PAPERS.

PROGRAM Semiannual Progress Report For April-September 1986.

NUREG/CR-4872: EXPERIMENTAL AND ANALYTICAL ASSESSMENT j

H'a -

on St ITech y

F RCUMFERENTIALLY SURFACE CRACKED PIPES UNDER N

NUREG/CR-4877: ASSESSMENT OF DESIGN BASIS FOR LOAD-CAR.

PROGRAM. Semiannual Progress Report For April-September 1986.

RYING CAPACITY OF WELD OVERLAY REPAIRS, NUREG/CR-4878: ANALYSIS OF EXPERIMENTS ON STAINLESS High-Level Weste STEEL FLUX WELDS Topical Report.

NUREG-1245 V01: RADIOACTIVE WASTE MANAGEMENT RESEARCH NUREG/CR-4894. A USER'S GUIDE TO THE NRC'S PIPING FRAC.

PROGRAM PLAN FOR HIGH LEVEL WASTE 1987.

TURE MECHANICS DATA BASE (PlFRAC).

Fractured Media NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH LEVEL NUREG/C43925 REV:

SWIFT ll SELF TEACHING WASTE REPOSITORY. Volume 2: Tuff.

CURRICULUM.lliustrative Problems For The Sandia Waste-Isolaton NUREG/CR-4655: UNSATURATED FLOW AND TRANSPORT Flow And Transport Model For Fractured Media.

THROUGH FRACTURED ROCK RELATED TO HIGH-LEVEL WASTE REPOSITORIES. Final Report Phase II.

Fractured Rock NUREG/CR-4655.

UNSATURATED FLOW AND TRANSPORT High-Temperature Oxidation NROUGH FRACTURED ROCK RELATED TO HIGH-LEVEL WASTE NUREG/CR-4889: ZlRCALOY-4 OXIDATION AT 1300 TO 2400 DE.

l REPOSITORIES. Final Report Phase ll.

GREES C.

NUREG/CR-4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM Fractures WITH UO2.

NUREG/CR-4936: AN INTEGRATED GEOLOGICAL. GEOPHYSICAL,AND GEOCHEMICAL INVESTIGATION OF THE MAJOR FRACTURES ON History THE EAST SIDE OF THE NEW MADRIO EARTHOUAKE ZONE.

NUREG 1214 R01: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-IC ASSESSMENT OF LICENSEE PERFORMANCE.

Fragility NUREG/CR-4899.

COMPONENT FRAGILITY RESEARCH Human Error Probability PROGRAM. Phase l Component Prontaaton.

NUREG/CR d772: ACCIDENT SEQUENCE EVALUATION PROGRAM HUMAN RELIABILITY ANALYSIS PROCEDURE.

Fuel Cycle NUREG-0332: POTENTIAL HEALTH AND ENVIRONMENTAL IMPACTS Human Rollability ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL CYCLES Final NUREG/CR-4772: ACCIDENT SEQUENCE EVALUATION PROGRAM Report.

HUMAN RELIABILITY ANALYSIS PROCEDURE.

Subj:ct Ind3x 39 i

Hydrogen Generation Knowledges And Atselittee Catalog NUREG/CR-4866: AN ASSESSMENT OF HYDROGEN GENERATION NUREG-1122 S01: KNOWLEDGES AND SKILLS CATALOG FOR NU-FOR THE PSF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

CLEAR POWER PLANT OPERATIONS.Pressunred Water Reactors.

ICAP LER NUREG-1270 V01: INTERNATIONAL CODE ASSESSMENT AND APPU-NUREG/CR-2000 V06 N2. UCENSEE EVENT REPORT (LER)

CATIONS PROGRAM. Annual Report COMPILATION For Month Of February 1987.

NUREG-1271: GUIDELINES AND PROCEDURES FOR THE INTERNA-NUREG/CR-2000 V06 N3-LICENSEE EVENT REPORT (LER)

TIONAL CODE ASSESSMENT AND APPLICATIONS PROGRAM.

COMPILATION.For Month Of March 1987.

NUREG/CR-2000 V06 N4: UCENSEE EVENT REPORT (LER)

IE Bulletin 83-01 COMPILATION For Month Of April 1987.

NUREGICR-4663: CLOSEOUT OF IE BULLETIN 8341: FAILURE OF RE ;

NUREG/CR.2000 V06 N5: UCENSEE EVENT REPORT (LER)

ACTOR TRIP BREAKERS (WESTINGHOUSE DB-50) TO OPEN ON '

COMPILATION For Month of May 1987.

AUTOMATIC TRIP SIGNAL LMFBR NUREG/CR-4719: COOLABILITY OF STRATIFIED 002 DEBRtS IN N

/ R 664: CLOSEOUT OF IE BULLETIN 83-04. FAILURE OF THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP BREAK-SOM WH Mm@ W MWho W3 Egenment ERS.

LOCA in-88tu Stroes NUREG-1230 DRFT FC: COMPENDIUM OF ECCS RESEARCH FOR RE-NUREG/CR-4623: IN-SITU STRESS MEASUREMENTS IN THE AUSTIC LOCA ANALYSIS. Draft Report For Comment.

EARTH'S CRUST IN THE EASTERN UNITED STATES-LWR Incident Response NUREG 1230 DRFT FC: COMPENDIUM OF ECCS RESEARCH FOR RE-NUREG 0728 R02: NRC INCIDENT RESPONSE PLAN' AUSTIC LOCA ANALYSIS. Draft Report For Comment.

NUREG/CR-3319 R01: LWR PRESSURE VESSEL SURVEILLANCE DO-Inden SIMETRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveel.

NUREG 0304 V12 N01: REGULATORY AND TECHNICAL REPORTS lance Physics-Dosimetry Data Base Compendium.

(ABSTRACT INDEX JOURNAL). Compilation For Frst Quarter NUREG/CR-4307 V03: LWR PRESSURE VESSEL SURVEILLANCE DO-1987, January-March.

SIMETRY IMPROVEMENT PROGRAM.1986 Annual Report, October 1985 September 1986.

Influent Soundary Condit6one NUREG/CR-4330 V03: REVIEW OF UGHT WATER REACTOR REGU-NUREG/CR-4901: EFFECTS FROM INFLUENT BOUNDARY CONDI-LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re.

TIONS ON TRACER MIGRATION AND SPATIAL VARIABluTY FEA-Qurements That May Have Marginal importance To Risk:Postaccident TURES IN INTERMEDIATE-SCALE EXPERIMENTS.

Sampling System. Turbine Missiles. Combustible Gas Controf. Charcoal Filters.

Ineervice inspect 6on NUREG/CR-4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RELi-NUREG/CR 4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) REU" ABluTY FOR INSERVICE INSPECTION OF UGHT WATER ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS. Semiannual Report. April-September 1986.

REACTORS. Semiannual Report. April-tomber 1986-NUREG/CR 4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RELi-I NUREG/CR-4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) REU' ABILITY FOR INSERVICE INSPECTION OF UGHT WATER l

ABluTY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report. October 1986 March 1987.

REACTORS. Semiannual Report, October 1986 - March 1987.

NUREG/CR4583 V02: DEVELOPMENT AND VAUDATION OF A REAL-Ineervice Monitoring TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CR4300 V04 Nt: ACOUSTIC EMISSION / FLAW RELATION-REAMOR CONENTS. Annual Repod,0ctober 1984 Septenh SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE N

/CR-4674 V03: PRECURSORS TO POTENTIAL SEVERE CORE VESSELS Progress Report. October 1986 March 1987.

DAMAGE ACCIDENTS:1984,A STATUS REPORT. Main Report And Ap-l inspection pendmes A And B.

l i

NUREG 0040 VII NOI: UCENSEE CONTRACTOR AND VENDOR IN-NUREG/CR 4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE i

SPECTION STATUS REPORT. Quarterly Report. January-March DAMAGE ACCIDENTS:1984,A STATUS REPORT. Appendixes C,D And 1987.(White Book)

E.

NUREG 1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION NUREG/CR-4964. UPDATE OF TABLE S-3 NONRADIOLOGICAL ENVI-FACluTIES CERTIFIED UNDER 10CFR55 Draft Report.

RONMENTAL PARAMETERS FOR A REFERENCE UGHT-WATER NUREG/CR-4583 V02: DEVELOPMENT AND VALIDATION OF A REAL.

REACTOR.Uramum Mining, Milling And Ennchment.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF UGHT WATER I

• REACTOR COMPONENTS. Annual Report. October 1984 - September R

V24 101: INDEXES TO NUCLEAR REGULATORY COM.

MISSION ISSUANCES. July September 1986.

Integrated Safety Assessment NUREG-0750 V24 N04: NUCLEAR REGULATORY COMMISSION IS-NUREG-1184 DAFT: INTEGRATED SAFETY ASSESSMENT SUANCES FOR OCTOBER 1986.Pages 489-679.

REPORT. INTEGRATED SAFFTY ASSESSMENT PROGRAM - MILL.

NUREG 0750 V24 N05: NUCLEAR REGULATORY COMMISSION IS-STONE NUCLEAR POWER STATION. UNIT

1. Docket No. 50 SUANCES FOR NOVEMBER 1986.Pages 681768.

245 (Northeast Nuclear Energy Co). Draft Report.

NUREG-0750 V24 N06: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR DECEMBER 1986.Pages 769 930.

Internal Evente NUREG/CR4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY Legisladon FROM INTERNAL EVENTS: SEQUOYAH. UNIT 1.

NUREG-0980 R03: NUCLEAR REGULATORY LEGISLATION.

NUREG/CR4550 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN-CY FROM INTERNAL EVENTS GRAND GULF. UNIT 1. Main Report.

License Application NUREG/CR4550 V06PT2: ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG-1300: ENVIRONMENTAL STANDARD REVIEW PLAN FOR THE CY FROM INTERNAL EVENTS: GRAND GULF. UNIT 1. Appendices.

REVIEW OF A LICENSE APPUCATION FOR A LOW LEVEL RADIO-ACTIVE WASTE DISPOSAL FACluTY.

Internat6onal Code Assosoment And Appl 6 cation NUREG 1270 V01: INTERNATIONAL CODE ASSESSMENT AND APPU.

Licensed Operating Reactors CATIONS PROGRAM. Annual Report.

NUREG 0020 V10 Nti: UCENSED OPERATING REACTORG STATUS NUREG 1271: GUIDELINES AND PROCEDURES FOR THE INTERNA-

SUMMARY

REPORT. Data As Of October 31.1986.(Gray Book 1)

TIONAL CODE ASSESbMENT AND APPLICATIONS PROGRAM.

NUREG 0020 V10 N12: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of Novertber 30,1986.(Gray Book 1)

NUREG/CR4758: A RETRAN MODEL OF THE CALVERT CUFFS 1 Ucensee Event Report PRESSURIZED WATER REACIOR FOR ASSESSING THE SAFETY NUREG/CR-2000 V06 N2: UCENSEE EVENT REPORT (LER)

IMPLICATIONS OF CONTROL SYSTEMS.

COMPILATION For Month Of February 1987.

.I 40 Subject Index NUREG/CR 2000 V06 N3 LICENSE E EVENT '9EPORT (LE R)

Meteorological Measurements COMPILATION For Month Of March 1987 NUREG/CR4950 V01: THE SHORELINE ENVIRONMENT ATMOS-NUREG/CH 2000 VO6 N4 LICENSEE EVENT REPORT (LER)

PHERIC DISPERSION EXPERIMENT (SEADEX)Expenment Desenp-COMPILATION For Month Of Apnl 1987 ton.

N9 REG /CR 2000 V06 N5 LICENSEE EVENT REPDAT (LER)

COMP 8LATION For Month Of May 1987.

Mixed-Oxide Conversion Facilities NUREG/CR-4773. DESIGN FEATURES TO FACILITATE INTERNATION-12 0 OF Fh FC: COMPENDIUM OF ECCS RESEARCH FOR RE-RE All3 TIC LOCA ANALYSIS Draft Report For Comment NDE Rollability NUREG/CR 3319 RO1 LWR PRESSURE VESSEL SURVEILLANCE DO-f4UREG/CR4469 V05 NONDESTRUCTIVE EXAMINATION (NDE) RELI.

S: METRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveil-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER lance Physics Dotometry Data Base Compendium.

REACTORS Semiannual Report Apnt-September 1986 NUREG/Cfl4307 V03 LWR PRESSURE VESSE L SURVEILLANCE 00-NUREG/CR-4469 V06 NONDESTRUCTIVE EXAMINATION (NDE) REll-SIMETRY IMPROVEMENT PROGRAM.1986 Annual Report. October ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report. October 1986 - March 1987.

NUF G/CR 0 V03 EVIEW OF LIGHT WATER REACTOR REGU-LATORY REQUIREMENTS. Assessment Of Sciected Regulatory Re~

NEORISK querements That May Have Margeal importance To RisM Postaccident NUREGICR4903 V02. SELECTION OF EARTHQUAKE RESISTANT Samphng System. Turbine Missdes. Combustible Gas Control. Charcoal DESIGN CRITER6A FOR NUCLEAR POWER PLANTS - METHODOLO.

GY AND TECHNICAL CASES. Methods For introduction Of Geological NU 1E /CR 4469 VOS NONDE STRUCTIVE EXAMINATION (NDE) RELI-Data Into Charactenzation Of Actwe Faults And Seismicity And...

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER RE ACTORS Somiannual Report. Apnl September 1986 National EnviromMntal PolicI Act NUREG/CR4469 V06 NONDESTRUCTIVE EXAMINATION (NDE) RELl-NUREG-1300. ENVIRONMENTAL STANDARD REVIEW PLAN FOR THE ABILITY FOR INSE RVICE INSPECTION OF LIGHT WATER REV!EW OF A LICENSE APPLICATION FOR A LOW-LEVEL RADIO-REACTORS Semiannual Report. October 1986 March 1987.

NUREG/CR4583 V02: DEVELOPMENT AND VALIE ATION OF A REAL.

ACTIVE WASTE DISPOSAL FACILITY.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER REACTOR COMPONENTS Annual Report. October 1984 Esptember g

C STUDY OF NATURAL GLASS ANALOGUES AS l

NUREG/CR4674 V03: PRECURSORS TO POTENTIAL SEVERE CORE APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS l

DAMAGE ACCIDENTS 1984.A STATUS REPORT Main Report And Ap-New Madr:d pendixes A And B NUREG/CR4674 V04. PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4936 AN INTEGRATED GEOLOGICALGEOPHYSICAL.AND l

DAMAGE ACCIDENTS 1984.A STATUS REPORT. Appendixes C.D And GEOCHEMICAL INVESTIGATION OF THE MAJOR FRACTUFsES ON E.

THE EAST SIDE OF THE NEW MADRID EARTHOUAKE ZONE.

NUREG/CR4964 UPDATE OF TABLE S-3 NONRADIOLOGICAL ENVI.

RONMENTAL PARAMETERS FOR A REFERENCE LIGHT-WATER Nolas Detection REACTOR Uranium Mining Milling And Ennchment NUREG/CRJ922: BROAD BAND SEISMIC DATA ANALYSIS September 1984 September 1986.

Liquid Metal Fast Breeder Reactor NUREG/CR4719 COOLABILITY OF STRATIFIED UO2 DEBRIS IN Nondestructive Examination SOOlUM WITH DOWNWARD HE AT REMOVAL.The D13 Expenment.

NUREG/CR4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RELi-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Lord-Carrying Capacity RE ACTORS Semiannual Report. Aprii. September 1986.

NUREG/CR4877 ASSESSMENT OF DESIGN BASIS FOR LOAD CAR-NUREG/CR4469 V06-NONDESTRUCTIVE EXAMINATION (NDE) RELl-RYING CAPACITY OF WELD-OVERLAY REPAIRS ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report. October 1986 March 1987.

Lots-Of Coolant Accident NUREG/CR 4583 V02: DEVELOPMENT AND VALIDATION OF A REAL-NUREG-1230 DRFT FC. COMPENDIUM OF ECCS RESEARCH FOR RE-TIME SAFT-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER ALISTIC LOCA ANALYSIS. Draft Report For Comment REACTOR COMPONENTS Annual Report. October 1984 - September Low Level Radioactive Waste NUREG/CR4615 V02 MODELING STUDY OF SOLUTE TRANSPORT Nuclear Fuel Cycle IN THE UNSATURATED ZONE Workshop Proceedings NUREG-0332: POTENTIAL HEALTH AND ENVIRONMENTAL IMPACTS NUREG/CR4938 OCCUPATIONAL RADIATION EXPOSURES ASSOCl-ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL CYCLES. Final ATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE DIS-POSAL Nuclear Plant Modification NUREG/CR4921: ENGfNEER!NG AND OUALITY ASSURANCE COST N REG 100 EN ON EN L STANDARD REVIEW PLAN FOR THE REVIEW OF A LICENSE APPLICATION FOR A LOW-LEVEL RADIO-FACTORS ASSOCIATED WITH NUCLEAR PLANT MOOlFICATION ACTIVE WASTE DISPOSAL FACILITY NUREG/CR4918 VOI: CONTROL OF WATER INFILTRATION INTO N REG 0980 R0 NU E R AEGULATORY LEGISLATION NEAR SURFACE LLW DISPOSAL UNITS Annual Report. October 1985

- September 1986 Nuclear Waste Glass MELCOR NUREG/CR4B42: A STUDY OF NATURAL GLASS ANALOGUES AS NUREG/CR4830 MELCOR VALIDATION AND VERIFICATION 1986 APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS.

b Occupational Radiation MKS Preprocessor NUREG/CR4938 OCCUPATIONAL RADIATION EXPOSURES ASSOCl-NUREG/CR4765. MXS CROSS-SECTION PREPROCESSOR USER'S ATED WITH ALTERNATIVE METHODS OF LOW LEVEL WASTE DIS-MANUAL POSAL

)

Luin Steam leolation Valve Operating Esportence NUREG/CR4739 RAMONA 3B CALCULATIONS FOR BROWNS NUREG 1272 REPORT TO THE U S NUCLEAR REGULATORY COM-FERRY ATWS STUDY.

MISSION ON ANALYSIS AND EVALUATION OF OPERATIONAL DATA 1986 Mark Ill Containment NUREG/CR4819 V01: AGING AND SERVICE WEAR OF SOLENOID-NUREG/CR4700 V4 DAF CONTAINMENT EVENT ANALYSIS FOR OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR POSTULATED SEVERE ACCIDENTS. GRAND GULF NUCLEAR POWER PLANTS Votume 1. Operating Expenence And Failure identife-STATION. UNIT 1 Draft For Comment cation

l

)

Subject Index 41 Operational Event Pressure Suppression Pool NUREG/CR-4674 V03 PRECURSORS TO POTENTIAL SEVERE CORE NUREG 0000 06.5.5 RO. STANDARD REVIEW PLAN FOR THE REVIEW DAMAGE ACCIDENTS 1984,A ST ATUS REPORT Main Report And Ap-OF SAFETY ANALYSIS REPORTS W W8 CLEAR POWER pendines A And B.

PLANTS LWR Edition. Proposed Revism 0 To New ST P Section 6.5.5.

NUREG/CR 4674 V04 PRECURSORS TO POTENTIAL SEVERE CORE

" Pressure Suppression Poots As Fission Product Cleanup Systems" DAMAGE ACCIDENTS 1984,A STATUS REPORT. Appendixes C,0 And For Comment.

E.

Pressure Vessel NUREG/CR 3319 RO1: LWR PRESSURE VESSEL SURVEILLAtiCE DO-RE 2 0 0 E ATOR LICENSING EXAMINER STANDARDS.

SIMETRY IMPROVEMENT PROGRAM LWR Power Reactor Su; veil-Oxidation lance Physics-Dosametry Data Base Compendium.

NUREG/CR-4866: AN ASSESSMENT OF HYDROGEN GENERATION NUREG/CR4219 V03 N2: HEAVY SECTION STEEL TECHNOLOGY FOR THE PDF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

PROGRAM. Semiannual Progress Report For Apni-September 1986.

NUREG/CR4300 VO4 N1: ACOUSTIC EMISSION / FLAW RELATION-Oxide Growth Rate SHIP FOR IN-SERVICE VONITORING OF NUCLEAR PRESSURE NUREG/CR4889 ZlRCALOY4 OXIDATICN AT 1300 TO 2400 DE-VESSELS Progress Repon, October 1980 March 1987.

GREES C.

NUREG/CR4307 V03: LWR PRESSURE VESSEL SURVEILLANCE DO-NUHEG/Cr 4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM 1985 September 1986 { PROGRAM.1986 Annual Report,0ctobe SIMETRY IMPROVEMfi WITH UG2.

PIFRAC Pressurized Releases NUREG/CR4894. A USER'S GUIDE TO THE NRC'S PIPING FRAC-NUREG/CR4779. NEW DATA FOR AEROSOLS GENERATED BY RE-TURE MECHANICS DATA BASE (PlFRAC)

LEASES OF PRE.MiURIZED POWDERS AND SOLUTIONS IN STATIC PWR AlR.

NUREG-1122 501: KNOWLEDGES AND SKILLS CATALOG FOR NU-CLEAR POWER PLANT OPERATIONS Pressurized Water Reactors.

Pressurized Thermal Shock NURE G/CR-3231: PIPE TO-PIPE IMPACT PROGRAM NUREG/CR4841: FRACTURE EVALUATION OF SURFACE CRACKS NUREG/CR-4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT EMBEDDED IN REACTOR VESSEL CLADDING Unirradiated Bend RISKS AND THE POTENTIAL FOR RISK REDUCTION SEQUOYAH Specimen Results.

POWER STATION. UNIT 1. Draft For Comment.

NUREG/CR-4700 V2 DRF. CONTAINMENT EVENT ANALYSIS FOR Pressurized Water Reactor POSTULATED SEVERE ACCIDENTS SEOUOYAH POWER NUREG-1122 SOI: KNOWLEDGES AND SKILLS CATALOG FOR NU.

i STATION, UNIT 1. Draft Report For Comment-CLEAR POWER PLANT OPERATIONS.Pressunred Water Reactors.

{

NUREG/CR4758 A RETRAN MODEL OF THE CALVERT CLIFFS 1 NUREG/CR-3231: PIPE TO-PIPE IMPACT PROGRAM.

a PRESSURIZED WATER REACTOR FOR ASSESSING THE SAFETY NUREG/CR-4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT NU CF B4 AN ANAL S S THE SEMISCALE MOD 2C S-NH-3 RSTAT N.U t

ft For Com nt TEST USING THE TRAC PF1 COMPUTER PROGRAM.

NUREG/CR4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR

{

Performance Testing POSTULATED SEVERE ACCIDENTS: SEQUOYAH POWER j

NUREG/CR 4959 PERFORMANCE TESTING OF EXTREMITY DOSI.

STATION. UNIT 1. Draft Report For Comment.

METERS.

NUREG/CR4758: A RETRAN MODEL OF THE CALVERT CLIFFS-1

)

PRESSURIZED WATER REACTOR FOR ASSES $1NG THE SAFETY Petitions For Rulemaking IMPLICATIONS OF CONTROL SYSTEW.

NUREG.0936 V05 N04 NRC REGULATORY AGENDA.Ouarterly NUREG/CR4845: AN ANALYSIS OF THD SEMISCALE MOD 2C S NH-3 Report. October December 1986 TEST USING THE TRAC-PF1 COMPUTER PROGRAM.

NUREG-0936 V06 Not NRC REGULATORY AGENDA Quarterly Repor January-March 1987.

Probabilistic Performance NUREG/CR-4800: SIGPl.A USER'S MANUAL FOR FAST COMPUTA-fEMS N

G/C 33 RO1: LWR PRESSURE VESSEL SURVEILLANCE DO-SIMETRY IMPROVEMENT PROGRAM LWR Power Reactor Surveil-lance Physics-Dos: metry Data Base Compondium, Probabillstic Risk Assessment Pipe Whlp NUREG/CR4800: SIGPI:A USER'S MANUAL FOR FAST COMPUTA-NUREG/CR-3231: PIPE TO-PIPE IMPACT PROGRAM.

TION OF THE PROBABILISTIC PERFORMANCE OF COMPLEX SYS-TEMS.

Pipe-To-Pipe impact NUREG/CR4899-COMPONENT FRAGILITY RESEARCH NUREG/CR 3231: PIPE TO-PIPE IMPACT PROGRAM.

PROGRAM. Phase 1 Component Pnoritization.

Piping Fracture Mechanics Data Base Quality Assurance Cost NUREG/CR4694. A USER'S GUIDE TO THE NRC'S PIPING FRAC-NUREG/CR-4921: ENGINEERING AND OUAllTY ASSURANCE COST TURE MECHANICS DATA BASE (PIFRAC)-

FACTORS ASSOCIATED WITH NUCLEAR PLANT MODIFICATION.

Population RAMONA 3P NUREG/CR-2850 V05 POPULATION DOSE COMMITMENTS DUE TO M^WG/CR 4739: RAMONA-38 CALCULATIONS FOR BROWNS RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES FERRY ATWS STUDY, IN 1983.

RETRAN Postaccident Sampling System NUREG/CR 4330 V03 REVIEW OF LIGHT WATER REACTOR REGU.

NUREG/CR 4758: A RETRAN MODEL OF THE CAUIRT CL/FS1 LATORY REQUIREMENTS Assessment Of Selected Regulatory Re-PRESSURIZED WATER REACTOR FOR ASSESSINC THE SAFETY quirements That May Have Marginal importance To RisitPostaccident IMPLICATIONS OF CONTROL SYSTEMS.

phng System, Turbine Missdes, Combustible Gas Control, Charcoal RETRAN2/ Mod 3 NUREG/CR4758: A RETRAN MODEL OF THE CALVERT CLIFFS-1 Practice And Procedure Digest PRESSURIZED WATER REACTOR FOR ASSESSING THE SAFETY NUREG-0386 004 ROS: UNITED STATES NUCLEAR REGULATORY IMPLICATIONS OF CONTROL SYSTEMS.

COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July 1972 - September 1986.

Radiation Dose NUREG/CR-2850 V05. POPULATION DOSE COMMITMENTS DUE TO Pressure Boundary RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR-4300 V04 N1: ACOUSTIC EMISSION / FLAW RELATION-IN 1983.

SHIP FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR4653: GASPAR 11 - TECHNICAL REFERENCE AND USER VESSELS Progress Report. October 1986 March 1987.

GUIDE.

42 SubgetInd2x Radiation Esp 9 auro Repositories NUREG/04415 DEMONSTRATION TESTING OF A SURVEILLANCE NUREG/CR4912: DATING GROUND WATER AND THE EVALUATION ROBOT AT DROWNS FERRY NUCLEAR PLANT.Analysss Of Costs OF REPOSITORIES FOR RADIOACTIVE WASTE.

AW 3enefits NUREG/CR.4938. OCCUPATIONAL RADIATION EXPOSURES ASSOCl-Ree6 dual Heat Removat System ATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE DIS-NUREG 1269: LOSS OF RESIDUAL HEAT REMOVAL SYSTEM.Diablo AEAL Canyon Unit 2. April 10.1987.

Ridletion Monhoring Network Reelstance Temperature Detectors NUREG4837 V06 N04. NRC TLD DIRECT RADIATION MONITORING NUREG/CR 4928: DEGRADATION OF NUCLEAR PLANT TEMPERA.

NETWORK. Progress Report, October-December 1986.

TURE SENSORS.

Radioactive Effluents NUR G/CR4653: GASPAR 11 TECHNICAL REFERENCE \\ND USER NU

/CR4 Vot: SELECTION OF EARTHOUAKE RESISTANT DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO-Radioactive Release GY AND TECHNICAL CASES Direct Empirical Scaling Of Response NUREG/CP.0078. PROCEEDINGS OF THE SYMPOSIUM ON CHEMI.

Spectral Amplitudes From Vanous Site And Earthqualte Parameters.

CAL PHENOMENA ASSOCIATED WITH RADIOACTIVITY RELEASES DUFUNG SEVERE NUCLEAR PLANT ACCIDENTS Riprap Design Criterta NUREG/CR.2850 V05: POPULAT!Ov4 DOSE COMMITMENTS DUE TO NUREG/CR 4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY RADIOACTIVE RELEASES FROM NvCLEAR POWER PLANT SITES RIPRAP TESTING IN FLUMES. Phase 1.

IN Vm3.

Riprap Testing R:dlos,tethe Waste NUREG/CR-4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY NUREG 0653 G02: PROGRAMMA flC ENVIRONMENTAL IMPACT RIPRAP TESTING IN FLUMES. Phase 1.

STATEMW RELATED TO DECONTAMINATION AND DISPOSAL OF RADIOAC7M WASTES RESULTING FROM MARCH 28.1979 ACCI-Risk DENT AT TN rE MILE ISLAND NUCLEAR STATION. UNIT 2. Final NUREG/CR4330 V03: REVIEW OF UGHT WATER REACTOR REGU-Supplerre OnaFng With Disposa! Of....

LATORY REQU:REMENTS. Assessment Of Selected Regulatory Re-NUREG-1245 VOL RADIOACTWE WASTE MANAGEMENT RESEARCH Quirements That May Have Marginal importance To RiskPostacesdent PROGRAM PLAf4 FOf t NGH LEVEL WASTE ud7.

Samptng System, Turune Missites. Combustible Gas Control, Charcoal NUHtG/CH4912: DATING GROUND WATER AND THE EVALUATION Filters.

OF REPOSITORIES FOR RADIOACTIVE WASTE.

NUREG/CR-4726: EVALUATION OF PROTECTIVE ACTION RISKS.

NUREG/CR4918 V01: CONTROL OF WATER INFILTRATION INTO NUREG/CR 4772: ACCIDENT SEQUENCE EVALUATION PROGRAM NEAR SURFACE LLW DISPOSAL UNITS.Arinual Report,0ctober 1985 HUMAN RELIABluTY ANALYSIS PROCEDURE.

- September 1986-NUREG/CR4814: SOURCES OF CORRELATION BETWEEN l

F m al NsuHs %n ho Memes.

RIactor Coolant Pump NUREG/CR-4821: REACTOR COOLANT PUMP SHAFT SEAL STABlu-Risk Reduct6on TY DURING STATION BLACKOUT.

NUREG/CR-4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT Rxetor Operator RISKS AND THE POTENTIAL FOR RISK REDUCTION SEQUOYAH NUREG 1122 Sot KNOWLEDGES AND SKILLS CATALOG FOR NU-POWER STATION. UNIT 1. Draft For Comment.

CLEAR POWER PLANT OPERATIONS Pressortzed Water Reactors' NUREG/CR-4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION PEACH Ructor Trip Dreaker BOTTOM. UNIT 2. Main ReporLDraft For Comment.

1 NUREG/CR-4663: CLOSEOUT OF lE BULLETIN 83 01. FAILURE OF RE.

NUREG/CR4551 V3 PT2: EVALUATION OF SEVERE ACCIDENT RISKS ACTOR TRIP BREAKERS (WESTINGHOUSE DB-50) TO OPEN ON AND THE POTENTIAL FOR RISK REDUCTION: PEACH AUTOMATIC TRIP SIGNAL.

BOTTOM, UNIT 2.Appendicesbratt For Comment.

NUREG/CR-4664 CLOSEOUT OF IE BULLETIN 83-04 FAILURE OF NUREG/CR-4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP BREAK-RISKS AND THE POTENTIAL FOR RISK REDUCTION: GRAND ERS.

GULF. UNIT 1. Draft For Comment RJactor Vesset Robotics NUREG/CR-4841: FRACTURE EVALUATION OF SURFACE CRACKS NUREG/CR-4815: DEMONSTRATION TESTING OF A SURVEILLANCE EMBEDDED IN REACTOR VESSEL CLADDING Unirradiated Bend ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs Specimen Results And Benefits.

R1gulatory Agenda Rules NUREG-0936 V05 N04 N9C REGULATORY AGENDA.Ouarterty NUREG 0936 VOS N04: NRC REGULATORY AGENDA.Quarterty Report. October-December '936 Report. October-December 1986 NUREG.0936 V06 N01: MIC REGULATORY AGENDA.Ouarterly NUREG-0936 V06 N01: NRC REGULATORY AGENDA.Ouarterty Report. January-March 1987.

Report. January March 1987.

R gulatory And Techn6 cal Report Rules Of Practice 4

NUREG 0304 V12 N01: REGULATORY AND TECHNICAL REPORTS NUREG-0386 D04 R05: UNITED STATES NUCLEAR REGULATORY (ADSTRACT INDEX JOURNAL). Compilation For First Quarter COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July 1997, January March.

1972 September 1986.

RJgulatory lesues N REG /CR4783 ALYSIS OF BALANCE-OF PLANT REGULATORY N REG / R4845: AN ANALYSIS OF THE SEM! SCALE MOD-2C S-NH-3 TEST USING THE TRAC-PF1 COMPUTER PROGRAM.

R gulatory Raquiremente NUREG/CR4330 V03. REVIJ.F OF CGHT WATER REACTOR REGU.

SAFT-UT System LATORY REQUIREMENTS. Assewsent Of Selected Regulatory Re.

NUREG/CR 4583 V02: DEVELOPMENT AND VALIDATION OF A REAL-querements That May Hays WrW Importance To Risk:Postaccident TIME SAFT UT SYSTEM FOR THE INSPECTION OF UGHT WATER Sampling System. Tur:w VistlhCombustible Gas Control. Charcoal REACTOR COMPONENTS. Annual Report. October 1964 September Filters.

1985.

Report To Congress SALP NUREG 0090 V09 NO3. REPORT TO CONGRESS ON ABNORMAL NUREG 1214 RO1: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-OCCURRENCESJuly September 1986.

IC ASSESSMENT OF LICENSEE PERFORMANCE.

1 SubpctInd2x 43 SEADEX Seismicity NUHE G/CR.4950 V01. THE SHORELINE ENVIRONMENT ATMOS-NUREG/CR4623 IN-SITU STRESS MEASUREMENTS IN THE PHERIC DISPERSION EXPERIMENT (SEADEX)Expenment Desenp-EARTH'S CRUST IN THE EASTERN UNITED STATES.

Ison Seismology StGP1 NUREG/CR4903 V01: SELECTION OF EARTHOUAKE RESISTANT NUREG/CH4800 SIGPl A USER'S MANUAL FOR FAST COMPUTA-Ot! SIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-TION OF THE PROBABILISTic PERF ORMANCE OF COMPLEX SYS-TEMS GY AND TECHNICAL CASES Direct Empincal Scahng Of Response Spectral Amphtudes From Varous Site And Earthquake Parameters SIMMER H NUR'EG/CR4903 V02: SELECTION OF EARTHOUAKE RESISTANT NUREG/CR 4765 MxS CROSS-SECT 6ON PREPROCESSOR USER'S DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-MANUAL GY AND TECHNICAL CASES Methods For introduction of Geological Data into Charactertration Of Active Faults And Seismscity And..

SRNY Station NUREG/CR4903 V03-SELECTION OF EARTHOUAKE RESISTANT NUREG/CR4822. BROAD BAND SEISMiG DATA ANALYSIS September DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO-1G84 - September 1986.

GY AND TECHNICAL CASES. Dislocation Models Of Near-Source Earthquake Ground Moton. A Revew.

SURBOT NUREG/CR4815 DEMONSTRATION TESTING OF A SURVEILLANCE Self TeacNng Curriculum

(

ROBOT AT BROWNS FERRY NUCLEAR PLANT Analysis Of Costs NUREG/CR-3925 REV:

SWIFT 11 SELF-TEACHING

{

And Benefits CURRICULUM.lliustrhtsve Problems For The Sandse Waste Isolaton i

Flow And Transport Model For Fractured Media.

l SWIFT tl NUREG/CR 3925 REV:

SWIFT 11 SELF TE ACHING Serniscale Mod-?C CURRICULUM lilustratwe Problems For The Sendia Waste-Isolation NUREG/CR 4802. AN EVALUATION OF TRAC-PFI/ MODI COMPUTER Flow And Transport Model For Fractured Media CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE-Safeguards MISCALE MOD 2C FEEDWATER LINE BREAK TRANSIENTS.

NUREG/CR4845: AN ANALYSIS OF THE SEMISCALE MOD 2C S-NH-3 NUREG/CR-4773 DESIGN FEATURES TO FACILITATE INTERNATION' TEST USING THE TRAC-PF1 COMPUTER PROGRAM.

l AL SAFEGUARDS AT mixed-OxlDE CONVERSION FACILITIES Safety Analysis b*

C'***'

NUREG-0800 06.5.2 R2. STANDARD REVIEW PLAN FOR THE REVIEW EN484 M AM AND SEME MAR & M6 i

OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR

]

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Editon. Proposed Revison 2 To Secton 6 5 2. "Contain-POWER PLANTS Volume 1. Operating Exponence And Folute identih-ment S As A Fission Product Cleanup System " For Comment caton.

NUREG-Oe 06 5 5 R0 STANDARD REVIEW PLAN FOR THE REVIEW Severe Accident OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS LWR Edition Proposed Revison 0 To Now SRP Section 6.5 5, NUREG-1265: UNCERTAINTY PAPERS ON SEVERE ACCIDENT "F e sure S ppresson Pools As Fisson Product Clean Up Systems."

NUREG/CP 8 PROCEEDINGS OF THE SYMPOSIUM ON CHEML CAL PHENOMENA ASSOCIATED WITH RADIOACTIVITY RELEASES I

Safety Evaiustion Report DURING SEVERE NUCLEAR PLANT ACCIDENTS NUREG-0781 503. SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR 4165: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket GRAM. ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA-Nos 50498 And 50499 (Houston Lighting And Power Companj TIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

NUREG-1002 S03. SAFETY EVALUATION REPORT RELATED w THE NUREG/C44551 V2 DRF: EVALUATION OF SEVERE ACCIDENT OPERATION OF BRAIDWOOD STATION. UNITS 1 AND 2 Docket Nos.

RISKS A lO THE POTENTIAL FOR RISK REDUCTION.SEQUOYAH 50450 And 50457 (Commonwealth Edison Company)

POWER STATION, UNIT 1. Draft For Comment NUREG-1057 505 SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-4551 V3 PT1' EVALUATION OF SEVERE ACCIDENT RISKS OPERATION OF BEAVER VALLEY POWER STATION. UNIT 2 Docket AND THE POTENTIAL FOR RISK REDUCTION. PEACH No 50-412 (Duquesne Light Company.et al)

BOTTOM. UNIT 2 Main Report. Draft For Comment.

NUREG/CR-4551 V3 PT2: EVALUATION OF SEVERE ACCIDENT RISKS Safety Research Program AND THE POTENTIAL FOR RISK REDUCTION PEACH NUREG/CR 2331 V06 N3 SAFETY RESEARCH PROGRAMS SPON-BOTTOM. UNIT 2. Appendices. Draft For Comment SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4551 V4 DRF. EVALUATION OF SEVERE eCCIDENT RESEARCH Ouarterty Progress Report. July September 1986.

RISKS AND THE POTENTIAL FOR AlSK REDUCTION: GRAND Safety Systems GULF, UNIT 1. Draft For Comment.

NUREG/CR 4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR NUREG/CR-4819 VOI: AGING AND SERVICE WEAR OF SOLENOID-POSTULATED SEVERE ACCIDENTS SEQUOYAH POWER OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR STATION. UNIT 1. Draft Repor1 For Comment POWER PLANTS. Volume 1 Operating Exponence And Failure identif'-

NUREG/CR-4700 V4 DRF: CONTAINMENT EVENT ANALYSIS FOR caton-POSTULATED SEVERE ACCIDENTS: GRAND GULF NUCLEAR STATION. UNIT 1. Draft For Comment UR G/CP 87 UM ARY REPORT OF THE SYMPOSIUM ON SEIS-MIC AND GEOLOGIC SITING CRITERIA FOR NUCLEAR POWER SODIUM WITH DOWNWARD HEAT FIEMOVAL.The D13 Expenment TS.

NUREG/CR4866: AN ASSESSMENT OF HYDROGEN GENERATION FOR THE PBF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS Selamic Data NUREGICR-4883. REVIEW OF RESEARCH ON UNCERTAINTIES IN NUREG/CR-4822 BROAD BAND SEISMIC DATA ANALYSIS September ESTIMATES OF SOURCE TERMS F90M SEVERE ACCIDENTS IN 1984 - September 1986 NUCLEAR POWER PLANTS.

Selsmic Hazard Shaft Seal NUREG/CR-4885 SEISMIC HAZARD CHARACTER 12ATION OF THE NUREG/CR-4821: REACTOR COOLANT PUMP SHAFT SEAL STABILI-EASTERN UNITED STATES Comparatwe Evaluation Of The LLNL And TY DURING STATION BLACKOUT.

EPRI Studies.

Simulation Facilities Seismic Safety Research NUREG-1258 DRFT; EVALUATION PROCEDURE FOH SIMULATION NUREG-1147 ROI: SEtSMIC SAFETY RESEARCH PROGRAM PLAN FACILITIES CERTIFIED UNDER 10CFR55. Draft Report.

Seismic-Fragility Response Small Dreak LOCA NUREG/CR4098 SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL-NUREG/CR4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY ERATED AGED CLASS 1E BATTERY CELLS FROM INTEHNAL EVENTS SEQUOYAH. UNIT 1.

r r-I e4 LubpCt Ind3x Solenoid Operated Valve Surveillance Robot NUREG/CR-4819 V01: AGING AND SERVICE WEAR OF SOLENOlD-NUREG/CR4815: DEMONSTRAflON TESTING OF A SURVEILLANCE

}

OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs J

POWEa PLANTE. Volume 1 Operating Expenence And Failure Identifi-And Benefits.

cation Systematic Assessment of f1censee Performance RE

^ ^^

i /CR 15 V02: MODELING STUDY OF SOLUTE TRANSPORT A

F NS PRO' (N THE UNSATURATED ZONE. Workshop Proceedings.

NUREG/CR-4655.

UNSATURATED F LOW AND TRANSPORT.

NUREG 0837 V06 N04: NRC TLD D'r TCT * \\DIMON LONITORING l

R POSI R ES na R Phase NUREG/CR4875: CHARA AlZATION OF CRUEv.D TUFF FOR THE NETWORK. Progress Report, Octobor-Dwmer "n6.

EVAL'JATION OF THE FATE OF TRACERS IN TRANSPORT STUDIFS TRAh-PF1/ MOD 1 sN THE UNSATURA'.4 ZONF NUREG/CR4802: AN EVALUATION @ TRAC PF1/H0D1 COMPUTER i

Sourco Term Uncertain 6 CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE-flUREG-1265: UNCERTAINTY PAPERS ON SEVERE A M ENT MISCALE MOD 2C FEEDWATER LINE BREAK TRANSIENTS.

SOURCE TERMS.

NUREC/CR 4845. AN ANALYblS OF THE SEMISCALE MOD 2C S-NH-3 NUREG/CR4M3: flEVIEW OF RESEARCH ON UNCERTAINT:t3 IN TEST USING THE TRAC PF1 COMPU~ER FROGRAM.

ESTIMATES 'OF SOU9CE TERMS FROM SEVERE ACCIDETS IN NUCLEAR POWER PLANTS.

T echnical Specification NUREG-1235: TECHNICAL SPECIFICATIONS FOR CLINTON POWER

'patial Variabillt STATION. UNIT 1. Docket No. 50-461.(llirois Power Cornpany)

NUREG/CR4 1: WFECTS FROM INFLUF'dT BCON%RY CONDI-WW2m TEGM S6WMS M B6 WM TIONS ON TRACER MIGR/ TION AND EM.TIAL VAKARill'Y FEA-POWER STATION, UNIT 2. Docket No. 50412TDuquesne Light Com-TURES IN INTf'RMEDIATE SCALE EXPERIMENTS.

pany)

Stainless Steel NUREG 1261: TECHNICAL SPECIFICATIONS FOR BRAIDWOOD NUREG/CR4872: EXPERIMENTAL AND ANALYTICNL ASSESSMENT STATON, UNITS 1 AND 2. Docket Nos. 50-456 And 50-457.(Common.

OF CIRCUMFERENTIALLi SURFAC'E CRACKET F4 PES t/iDE R wealth Edison Company)

BENDING NUREG/CR4878. ANAL" SIS OF EXPErCMEN'S ON STAINLESS Temperature Sensor STEEL FLUX WELDS. M pica' hJr art.

NUREG/CR4928 DEGRADATION OF NUCLEAR PLANT TEMPERA-TURE SENSORS.

Standard Review Plan NUREG-0800 06 5.2 R2. STANDARD REVIEW PLAN FOR THE REVIEW Thermal-Hydraulic Production Code l

OF SAFETY ANALY SIS REPORTS FOR NUCLEAR POWER NUREG/CR-4758. A RETRAN MODEL OF THE CALVERT CLIFFS 1 PLANTS LWR Edition. Proposed Rewsion 2 'o Section 6.5 2. 'Contain.

VAF.SSURIZED WATER REACTOR FOR ASSESSING THE SAFETY NURE Ob 06 5 5 b. TA D R E

R F IEW /

OF SAFETY ANALYSIS REPORTS FOR 4UCLEAR oOWER Thermal-Hydraulics PLMTS LWR Edruon Proposed Revision 0 To N /* SHP SeNon 6.5 5, NUREG-1244. PLAN FOP ifITEGRATING TECHNICAL ACTIVITIES l

"Prusure Supmessr>n Faois As Fission Product Cean-Up 3ystems."

W(THIN THE U.S NRC ANO ITS CONTRACTORS IN THE AREA OF l

NU E 1 ENVikiWENTAL STANDARD REVIEW PLAN FOR THE NUREG/C -48 AN ALYSIS OF THE SEMISCALE MOD-2C S-NH-3 REVIEW f f A LICENM APPLICATION FOR A LOW-LEVEL RADIO-TEST USING THE TRAC PF1 COMPMER PROGRAM.

l ACTIVE WASTE DISPOS:21. FACil lTY.

j Static Air Thermodynamic f

NUREG/CR4779. NEW DAT' FOR AEROSOLS L.ENERATED BY RE.

NUREG/CR-4830. MELCOR VALIDATION AND VERIFICATIOtt 1986 LEASES OF PRESSUAIZE'f0WDERS AND SOLUTIONS IN STATIC PAPERS.

j AIR Thermoluminescent Dasitnoter Station Blackout NUREG 0837 \\ 06 N04: NRC TLD D RECT RECIATION MONITORING NUREGICA 4821: REA0 TOR COOLANT PUMP SHAFT SEAL STABill-NETWORK.Progrees Report, October-DecerrAn 1086.

Tv OURiNG STATION BLACKOUT.

Title List i

N FG/

16 8.

STEAM GENERATOR GROUP PROJECT. Annual AA B F Regwt 1985

/

NUREG-05c0 V09 NN f TirLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLT March 1-31,1987.

Steam Separator Modeling NUREGW A09 N04 TITLE LIST OF DOCUMENTS MADE PUBLICLY NUREG/CR4922. STEAM EEPARATOR MODELING FOh VARIOUS '

NUCLEAR REACTOR TRANSIENTS AVAll ABLE. Ar6130,19t7,.

i

/

Steel Yracer IMgration NUREG/CR42ts V03 N2 HEAVY SECTION 4 TEEL TECHNOLOGY NUREG/CR.4pn t : EFFECTS FROM INFLUENT BOUNDARY CONDI-PitOGRAM Semiannual Progress Repor1 For Apnl-September 1980 TONS ON MACER MIGRATION AND SPATIAL VARIABILITY FEA-TUP.ES IN INTERMEDIATE-SCALE EXPERIMENTS.

Stren Measurement l'

%.lREG/CR4623 IN-SITU STRESS MEASUREMENTS IN THE Tracers l

F.ARTH'S CRUST IN THE EASTERN UNITED STATES NUREG/CR-4875 CHARACTERIZAT;ON OF CRUSHED TUFF FOR THE EVALUATION OF THE F ATE OF TRACERS IN TRANSPORT STUDIES l

$<uctutel Integrity IN THE UNSATURATED 2ONE.

NUREG CR4219 V03 N2 HEAVY.SECTION STEEL TECHNCLOGY TROGRAY Somianut Progress Report For Apnt-September 1996.

Transient NUREG/CR-4802: AN EVALUATION OF TRAC PF1/ MOO 1 COMPUTER Surf ace Creek NURK./CR4841: FAACTURE EVALUATION OF SU9 FACE CRACKS CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE-EMBEDDED IN REACTOR VESSEL CLADDING Uvrad p:f Dend MISCALE MOD-2C FEEDWkTER LINE BREAK TRANSIENTS.

NUREG/CR4022. STEAM SEPARATC3t MODELING FOR VARIOUS Specimen Resufis i

NUCLE /9 RE ACTOR TRANSIENTS tWr a tente Dosimety frorciement Program NL4EG 'CR43'07 V(3 LWR PRESSURE VESSEL SURVEILLANCE DO-Tuff SIMETRY IMPROVEMENT PROGRAM IM6 Annual Report. October NUREG/CR4181 V02' CRITICAL pap.aMETERS FOR A HIGH LEVEL 1985 Septernber 1986.

WASTE REPOOITORY. Volume 2: Tuff.

l

Subbetindsx 45 NUREG/CR 4875: CHARACTERIZATION OF CRUSHED TUFF FOR THE Waste Disposal EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES NUREG/CR 4615 V02: MODELING STUDY OF SOLUTE TRANSPORT IN THE UNSATURATED ZONE.

IN THE UNSATURATED ZONE. Workshop Proceedings.

NUREG/CR4938: OCCUPATIONAL RADIATION EXPOSURES ASSOCl-Turbine Missiles ATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE DIS-NUREG/CR4330 V03. REVIEW OF LIGHT WATER REACTOR REGU-POSAL.

LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-quirements That May Have Marginal importarice To Risk:Postaccident Waste Management Samphng System. Turbine Missdes. Combustible Gas Control, Charcoal NUREG/CR4918 V01: CONTROL OF WATER INFILTRAT:ON INTO Filters.

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report. October 1985 September 1986.

NUREG/CR4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM Water infiltrators WITH UO2.

NUREG/CR-4918 Vit: CONTROL OF WATER INFILTRATION INTO NEAR SURFACE LLW DISPOSAL UNITS. Annual Report. October 1985 Uttrasonic Testing September 1986.

NUREG/CR-4583 V02: DEVELOPMENT AND VAllDATION OF A REAL.

TIME SAFT-(TT SYSTEM FOR THE INSPECTION OF LIGHT WATER Weld-Overlay Repaire REACTOR COMPONENTS. Annual Report October 1984 - September NUREG/CR4877: ASSESSMENT OF DESIGN BASIS FOR LOAD-CA9 1985.

RYING CAPACITY OF WELD-OVERLAY REPAIRS.

Undervoltage Trip Function Welds NUREG/CR4664. CLOSEOUT OF IE BULLETIN s3-04. FAILURE OF NUREG/CR-4878. ANALYSIS OF EXPERIMENTS ON STAINLESS THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP BREAK-STEEL FLUX WELDS. Topical Report.

ERS.

Westinghouse DB 50 Uniform Risk Spectra NUREG/CR 4663: CLOSEOUT OF IE BULLETIN 83 01: FAILURE OF RE-NUREG/CR4903 V02. SELECTION OF EARTHOUAKE RESISTANT ACTOR TRIP BREAKERS (WESTINGHOUSE DB 50) TO OPEN ON DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO-AUTOMATIC TRIP SIGNAL.

GY AND TECHNICAL CASES. Methods For introduction Of Geological Data into Charactenzation Of Active Faults And Seismicity And...

Zircaloy NUREG/CR4866: AN ASSESSMENT OF HYDROGEN GENERATION Unsaturated Zone FOR THE PBF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

NUREG/CR-4615 V02: MODELING STUDY OF SOLUTE TRANSPORT IN THE UNSATURATED ZONE. Werkshop Proceedings.

Zircaloy 4 Oxidation NUREG/CR 4655: UNSATURATED FLOW AND TRANSPORT NUREG/CR 4889. ZlRCALOY-4 OXIDATION AT 1300 TO 2400 DE-l THROUGH FRACTURED ROCK RELATED TO HIGH-LEVEL WASTE GREES C.

REPOSITORIES Final Report Phase ll.

NUREG/CR-4875: CHARACTERIZATION OF CRUSHED TUFF FOR THE Zirconium EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES NUREG/CR 4890: HEAT OF REACTION OF "OLTEN ZlRCONIUM IN THE UNSATURATED ZONE.

WITH UO2.

i

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-sections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

ADVISORY COMMITTEE (S)

OFFICE OF NUCLEAR MATERIAL SAFEfY & SAFEGUARDS ACRS - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS OFFICE OF NUCLEAR MATERIAL SAFETY NUREG 1125 V08. A COMPILATION OF REPORTS OF THE ADVISO-SAFEGUARDS. DIRECTOR RY COMMITTEE ON REACTOR SAFEGUARDS,1986.

NUREG 1300: ENVIRONMENTAL STANDARD REVIEW PLAN FOR THE REVIEW OF A LICENSE APPLICATION FOR A LOW LEVEL OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

RADIOACTIVE WASTE DISPOSAL FACILITY.

REGION 1. OFFICE OF DIRECTOR DIVISION OF FUEL CYCLF, MEDICAL, ACADEMIC & COMMERCIAL i

NUREG-0837 V06 N04 NRC TLD DIRECT RADIATION MONITORING USE SAFETY (POST l

NETWORK.Pr ress Report. October-December 1986.

NUREG 0904 S01: ORA 9UPPLEMENT TO THE FINAL ENVIRON.

I REG 5.OFFI F

E MENTAL STATEMENT RELATED TO THE DECOMMISSIONING OF O RES UAL HEAT REMOVM SYSTEMDablo Canyon Unit 2. April 10.1987.

THE RARE EARTHS FACILITY, WEST CHICAGO,lLLINO'S. Docket OFFICE OF ENFORCEMENT (POST 870413)

No. 40-2061.(Kerr-McGee)

NUREG-0940 V06 N01: ENFORCEMENT ACTIONS:SIGNIFICANT AC.

NUREG 1239. ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF TIONS RESOLVED.Ouarterly Progress Report. January-March 1987.

SOURCE MATERIAL LICENSE NO. STB 401. Docket No. 40-6563.(Columbeum Tantalum Division. Mallinkrodt.inc.)

EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA U.S. NUCLEAR REGULATORY COMMISSION OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, DI.

OFFICE OF THE GENERAL COUNSEL RECTOR NUREG-0386 D04 R05: UNITED STATES NUCLEAR REGULATORY NUREG4090 V09 NO3: REPORT TO CONGRESS ON ABNORMAL COMMISSION STA"F PRACTICE AND PROCEDURE DIGEST. July OCCURRENCES. July-Soptember 1986.

1972. September 1986.

NUREG-0728 R0e NRC INCIDENT RESPONSE PLAN NUREG-0980 R03: NUCLEAR REGULATORY LEGISLATION.

NUREG-1272: REPORT TO THE U.S. NUCLEAR REGULATORY COM-MISSION ON ANALYSIS AND EVALUATION OF OPERATIONAL OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/41)

DATA - 1986.

OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR (POST 860720)

OFFICE OF INSPECTJON & ENFORCEMENT (POST 12/11/80)

NUREG 1244: PLAN FOR INTEGRATING TECHNICAL ACTIVITIES DIVISION OF INSPECTION PROGRAM 3 (850212-870411)

WITHIN THE U.S. NRC AND ITS CONTRACTORS IN THE AREA OF NUREG-1214 R0t HISTORICAL JATA

SUMMARY

OF THE SYSTEM-THERMAL HYDRAULICS.

j ATIC ASSESSMENT OF LICENSEE PERFORMANCE.

NUREG 1265: UNCERTAINTY PAPERS ON SEVERE ACCIDENT j

SOURCE TERMS.

1 OFFICE OF INFORMATION RESOURCES MANAGEMENT NUREG 1270 V01: INTERNATIONAL CODE ASSESSMENT AND AP.

l OFFICE OF ADM:N!STRATION & RESOURCES MANAGEMENT, DI-PLICATIONS PROGRAM. Annual Report.

l RECTOR (POST 8704t3)

DIVISION OF ENGINEERING (POST 870413)

NUREG 1145 V03; U S. NUCLEAR REGULATORY COMMISSION NUREG 1147 RO1: SEISMIC SAFETY RESEARCH PROGRAM PLAN.

1986 ANNUAL REPORT-NUREG 1245 V01: RADIOACTIVE WASTE MANAGEMENT RE.

OlVISION OF PUBLICATION SERVICES (POST 870413)

SEARCH PROGRAM PLAN FOR HIGH LEVEL WASTE - 1987.

NUREG-0304 V12 N01: REGULATORY AND TECHNICAL REPORTS DIVISION OF REACTOR & PLANT SYSTEMS (POST 870413)

(ABSTRACT INDEX JOURNAL) Compdation For First Quarter NUREG-1230 DRFT FC: COMPENDIUM OF ECCS RESEARCH FOR 1987 January March-NUREG 0540 v:09 NO2: TITLE LIST OF DOCUMENTS MADE PUBLIC-REALISTIC LOCA ANALYSIS. Draft Report For Comrnent.

NUREG-1271: GUIDELINES AND PROCEDURES FOR THE INTERNA-NU EG-0 0V NO TLE i F DOCUMENTS MADE PUBLIC-LY AVAILABLE. March 1 31,1487-OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

NUREG 0540 V09 N04' TITLE LIST OF DOCUMENTS MADE PUBLIC-OFFICE OF NUCLEAR RE.CTOR REGULATION, DIRECTOR (POST LY AVAILABLE. Apol 1-30.1987.

870411)

NUREG-0750 V24101: INDEXES TO NUCLEAR REGULATORY COM-NUREG 0332: POTENTIAL HEALTH AND ENV RONMENTAL IM.

MISSION ISSUANCES. July September 1986-PACTS ATTRIBUTABLE TO THE NUCLEAR AND COAL FUEL NUREG 0750 V24 N04. NUCLEAR REGULATORY COMMISSION IS-CYCLES Final Report.

SUANCES FOR OCTOBER 1986 Pages 489-679 NUREG-0800 06.5.2 R2: STANDARD REVIEW PLAN FOR THE NUREG-0750 V24 N05: NUCLEAR REGULATORY COMMISSION IS*

REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR NUREG 4 N06 U LEA REYhL TORY COMMISSION IS-son 2 % S g n

SUANCES FOR DECEMBER 1986.P s 769-930.

"""'"*"I

'Y ^' ^ * * "

• ""E Y

"**"I DIVISION OF RULES & RECORDS (POS 870413)

NUREG-0936 V05 N04: NRC REGULATORY AGENDA.Ouarterly NUREG-0800 06.5.5 R0 STANDARD REVIEW PLAN FOR THE Report October-December 1986 REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR NURE G-0936 V06 N01: NRC REGULATORY AGENDA. Quarterly POWER PLANTS. LWR EditsortProposed Revision 0 To New SAP Report. January-March 1997.

Section 6.5.5. " Pressure Suppression Pools As Fission Product DIVISION OF BUDGET & ANALYSIS (POST 870413)

Clean-Up Systems." For Comment.

NUREG 1100 V03 ADD: BUDGET ESTIMATES Fiscal Years 1988 NUREG 1122 S01: KNOWLEDGES AND SKILLS CATALOG FOR NU.

1989 CLEAR POWER PLANT OPERATIONS.Pressunred Water Reactors.

DivlSION OF COMPUTER & TELECOMMUNICATIONS SERVICES NUREG 1164 ORFT: INTEGRATED SAFETY ASSESSMENT (POST 870413)

REPORT. INTEGRATED SAFETY ASSESSMENT PROGRAM MILL.

NUREG-0020 V10 N11: LICENSED OPERATING REACTORS STATUS STONE NUCLEAR POWER STATION. UNIT 1. Docket No. 50-

SUMMARY

REPORT. Data As Of October 31.1986 (Gray Book 1) 245 (Northeast Nuclear Energy Co). Draft Report NUREG-0020 V10 N12. LICENSED OPERATING REACTORS STATUS NUREG 1235: TECHNICAL SPECIFICATIONS FOR CLINTON POWER

SUMMARY

REPORT. Data As Of November 30,1986(Gray Book 1)

STATION. UNIT 1. Docket No. 50-461.(Ilkros Power Company) 47

48 NRC Origin: ting Org::niz:ti:n index (St:ff R:p rts)

DIVISION OF REACTOR PROJECTS.1/ll (POST 8704t1)

NUREG 1261: TECHNICAL SPECIFICATIONS FOR BRAIDWOOD NUREG 1057 505: SAFETY EVALUATION REPORT RELATED TO STATION. UNITS 1 AND 2. Docket Nos. 50-456 And 50 457.(Com-THE OPERAT'ON OF BEAVER VALLEY POWER STATION, UNIT monwealth Edison Companyl

2. Docket No.60-412 (Duquesne Light Company,et al)

TMI-2 CLEANUP PROJECT DIRECTORATE NUREG-1259: TECHNICAL SPECIFICATIONS FOR BEAVER VALLEY NUREG-0683 SO2: PROGRAMMATIC ENVIRONMENTAL IMPACT POWER STATION, UNIT 2.Docliet No. 50-412.(Duquesne Light Com.

STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL pany)

OF RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 DIVISION OF REACTOR PROJECTS. Ill,lV,V & SPECIAL PROJECTS ACCIDENT AT THREE MILE ISLAND NUCLEAR STATION, UNIT (POST 870411

2. Final Supplement Dealing With Disposal OL..

DIVISION OF REACTOR INSPECTION & SAFEGUARDS (POST NUREG-0781 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SOUTH TEXAS PROJECT. UNITS 1 AND NU G 40 VII Noi: LICENSEE CONTRACTOR AND VENDOR IN-et Nos. 50-498 And 50-499. (Houston Lsghting And Power SPECTION STATUS REPORT. Quarterly Report. January-March NUREG-1002 S03: SAFETY EVALUATION REPORT RELATED TO OPERATO L EN i G BRANCH THE OPERATION OF BRAIDWOOO STATION, UNITS 1 AND NUREG 1021 R04: OPERATOR LICENSING EXAMINER STAND-

2. Docket Nos. 50-456 And 50-457.(Commonwealth Edison Compar f)

ARDS.

NUREG 1166: FINAL ENVIRONMENTAL STATEMENT FOR DECOM-DIVISION OF HUMAN FACTORS TECHNOLOGY (851125-870411)

MISSION!NG HUMBOLDT BAY POWER PLANT, UNIT 3. Docket No.

NUREG-1258 DRFT: EVALUATION PROCEDURE FOR SIMULATION 50-133.(Pacific Gas And Electnc Company)

FACILITIES CERTIFIED UNDER 10CFR65. Draft Report.

j i

NRC Originating Organization index (International Agreements)

This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., pr,ogram of-fices) and then by subsections of these (e.g., divisions, branches) where appropnate. 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 this quarter, i

i 49

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organizuns that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program offi,ce) and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) prepared by that organi-zation. If further information is needed, refer to the main citation by the NUREG/CR number.

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

NUREG/CR4875: CHARACTERIZATION OF CRUSHED TUFF FOR URANIUM RECOVERY FIELD OFFICE THE EVALUATION OF THE FATE OF TRACERS IN TRANSPORT NUREG/CR-4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA STUDIES IN THE UNSATURATED ZONE.

{

BY RIPRAP TESTING IN FLUMES. Phase 1.

EDO. OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

DATA OFFICE OF NUCLEAR REGULATORY RESEAF.OH DIRECTOR (POST 860720)

OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA Dl-RECTOR NUREG/CR 2331 V06 N3: SAFETY RESEARCH PROGRAMS SPON.

SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-2000 V06 N2: LICENSEE EVENT REPORT (LER)

RESEARCH.Ouarterly Progress Report. July-September 1986.

COMPILATION For Month Of February 1987.

NUREG/CR4082 VOS: DEGRADED PIPING PROGRAM - PHASE NUREG/CR-2000 V06 N3 LICENSEE EVENT REPORT (LER)

II. Semiannual Report, April-September 1986.

COMPILATION For Month Of March 1987.

NUREG/CR-4098. SEISMIC-FRAGILITY TESTS OF NEW AND AC-NUREG/CR 2000 V06 N4: LICENSEE EVENT REPORT (LER)

CELERATED-AGED CLASS 1E BATTERY CELLS.

COMPILATION For Month Of April 1987-NUREG/CR-4219 V03 N2: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR 2000 V06 N5: LICENSEE EVENT REPORT (LER)

COMPILATION For Month Of May 1987.

PROGRAM. Semiannual Progress Report For Apnt-September 1986.

NUREG/CR-4307 V03: LWR PRESSURE VESSEL SURVEILLANCE NUREG/CR4674 V03: PRECURSORS TO POTENTIAL SEVERE DOSIMETRY IMPROVEMENT PROGRAM 1986 Annual CORE DAMAGE ACCIDENTS 1984,A STATUS REPORT. Main Report And Appendines A And B.

Report,0ctober 1985 - September 1986.

NUREG/CR-4617: ONSITE ASSESSMENTS OF THE LFFECTIVE-NUREG/CR-4674 V04. PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS 1984,A STATUS REPORT. Appendixes NESS AND IMPACTS OF UPGRADED EMERGENCY OPERATING C.D And E.

PROCEDURES.

NUREG/CR-4623: IN-SfTU STRESS MEASUREMENTS IN THE OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)

EARTH'S CRUST IN THE EASTERN UNITED STATES.

DIVISI OF EME EN PREPAREDNESS & ENGINEERING RE-NUREG/CR-4679: QUANTITATIVE DATA ON THE FIRE BEHAVIOR OF COMBUSTIBLE MATERIALS FOUND IN NUCLEAR POWER NUREG/CR-4663 CLOSEOUT OF IE BULLETIN 83-01. FAILURE OF

^

EA RS (WESTINGHOUSE DB-50) TO OPEN NUR

-4681 E CL RE ENVIRONMENT CHARACTERIZE-ON A O TRI TlON TESTING FOR THE BASE LINE VAllDATION OF COMPUTER NUREG/CR-4664. CLOSEOUT OF IE BULuETIN 83 04 FAILURE OF FIRE SIMULATION CODES.

THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP NUREG/CR-4700 V4 DRF: CONTAINMENT EVENT ANALYSIS FOR BREAKERS ~

POSTULATED SEVERE ACCIDENTS: GRAND GULF NUCLEAR STATION, UNIT 1 Draft For Comment OFFICE OF INFORMATION RESOURCES MANAGEMENT NUREG/CR 47t9. COOLABILITY OF STRATIFIED UO2 DEBRIS IN OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT, DI.

SODIUM WITH DOWNWARD HEAT REMOVALThe D13 Expenment.

RECTOR (POST 870413)

NUREG/CR-4739: RAMONA 38 CALCULATIONS FOR BROWNS NUREG/CR4921: ENGINEERING AND OUALITY ASSURANCE COST FERRY ATWS STUDY.

FACTORS ASSOCIATED WITH NUCLEAR PLANT MODIFICATION.

NUREG/CR 4758: A RETRAN MODEL OF THE CALVERT CLIFFS-1 OlVtSION OF COMPUTER & TELECOMMUNICATIONS SERVICES PRESSURIZED WATER REACTOR FOR ASSESSING THE SAFETY (POST 870413)

IMPLICATIONS OF CONTROL SYSTEMS.

NUREG/CR-2850 VOS POPULATION DOSE COMMITMENTS DUE TO NUREG/CR-4765: MXS CROSS-SECTION PREPROCESSOR USER'S RADI CTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/C -4772: ACCIDENT SEQUENCE EVALUATION PROGRAM HUMAN RELIABILITY ANALYSIS PROCEDURE.

OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CR4802: AN EVALUATION OF TRAC-PF1/ MOD 1 COMPUT-OFFICE OF NUCLEAR MATERIAL SAFETY ER CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SAFEGUARDS. DIRECTOR i

SEMISCALE MOD-2C rEEDWATER LINE BREAK TRANSIENTS.

I NUREG/CR-4615 V02: MODELING STUDY OF SOLUTE TRANSPORT NUREG/CR-4814: SOURCES OF CORRELATION BETWEEN 1

IN THE UNSATURATED ZONE. Workshop Proceedogs.

EXPERTS Empirical Results From Two Extremes.

NUREG/CR4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA NUREG/CR-4819 V01: AGING AND SERVICE WEAR OF SOLENOID-BY RIPRAP TESTING IN FLUMES Phase L NUREG/CR4773 DESIGN FEATURES TO FACILITATE INTERNA-OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR POWER PLANTS Volume 1. Operating Expenence And Failure ident>-

TlONAL SAFEGUARDS AT MIXED-OXIDE CONVERSION FACILl-fication.

TIES.

NUREG/CR 4822:

BROAD BAND SEISMIC DATA NUREG/CR 4901: EFFECTS FROM INFLUENT BOUNDARY CONDl-ANALYSIS. September 1984 September 1986.

TIONS ON TRACER MIGRATION AND SPATIAL VARIABILITY FEA-TURES IN INTERMEDIATE-SCALE EXPERIMENTS.

NUREG/CR 4830: MELCOR VALIDATION AND VERIFICATION 1986 PAPERS.

DIVISION OF FUEL CYCLE. MEDICAL, ACADEMIC & COMMERCIAL NUREG/CR-4845: AN ANALYSIS OF THE SEMISCALE MOD-2C S-USE SAFETY (POST NW3 TEST USING THE TRAC-PF1 COMPUTER PROGRAM.

NUREG/CR4779. NEW DATA FOR AEROSOLS GENERATED BY NUREG/CR4848. STEAM GENERATOR GROUP PROJECT. Annual RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN Report 1985.

STATIC AIR.

DIVISION OF WASTE MANAGEMENT (PRE 870413)

NUREG/CR-4872: EXPERIMENTAL AND ANALYTICAL ASSESSMENT OF CIRCUMFERENTIALLY SURFACE-CRACKED PIPES UNDER NUREG/CR 3925 REV:

SWIFT 11 SELF-TEACHING BENDING.

CURRICULUMlitustrative Problems For The Sandia Waste-Isolation NUREG/CR4877: ASSESSMENT OF DESIGN BASIS FOR LOAD-l i

Flow And Transport Model For Fractured Media.

CARRYING CAPACITY OF WCLD-OVERLAY REPAIRS.

)

m l

52 NRC C:ntr:ct Sp:nnr IndIx NUREG/CR4878. ANALYSIS OF EXPERIMENTS ON STAINLESS NUREG/CR-4550 V06PT1: ANALYSIS OF CORE DAMAGE FRE-STEEL FLUX WELDS. Topical Report.

QUENCY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1. Main NUREG/CR4889: ZlRCALOY-4 OXIDATION AT 1300 TO 2400 DE.

Report.

GREES C.

NUREG/CR 4550 V06PT2: ANALYSIS OF CORE DAMAGE FRE-OUENCY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1.Appen.

DIVISION OF ENGINEERING (POST 870412)ROGRAM.

NUREG/CR 323t: PIPE-TO PIPE IMPACT P dices NUREG/CR 3319 RO1: LWR PRESSURE VESSEL SURVEILLANCE NUREC/CR4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT DOSIMETRY IMPROVEMENT PROGRAM. LWR Power Reactor Sur.

RISKS AND THE POTENTIAL FOR RISK REDUCTION: PEACH veillance Physics-Dosimetry Data Base Compendium.

BOTTOM. UNIT 2 Main Report. Draft For Comment.

NUREG/CR-3956: IN SITU TESTING OF THE SHIPPINGPORT NUREG/CR4551 V3 PT2: EVALUATION OF SEVERE ACCIDENT ATOMIC POWER STATION ELECTRICAL CIRCUfTS.

RISKS AND THE POTENTIAL FOR RISK REDUCTION: PEACH NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL BOTTOM. UNIT 2. Appendices. Draft For Comment.

WASTE REPOSITORY. Volume 2: Tuff.

NUREG/CR-4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR-4300 V04 N1: ACOUSTIC EMISSION / FLAW RELATION.

RISKS AND THE POTENTIAL FOR RISK REDUCTION: GRAND SHtP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE GULF. UNIT 1. Draft For Comment.

VESSELS Progress Report. October 1966 lOrch 1987.

NUREG/CR-4726: EVALUATION OF PROTECTIVE ACTION RISKS.

HUREG/CR-4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) RE-NUREG/CR4800: SIGPl:A USER'S MANUAL FOR FAST COMPUTA-LIABILITY FOR INSERVICE INSPECTION OF LIGHT WATER TON OF THE PROBABILISTIC PERFORMANCE OF COMPLEX REACTORS. Semiannual Report. Apni-September 1986.

SYSTEMS NUREG/CR4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RE-NUREG/CR4866: AN ASSESSMENT OF HYDROGEN GENERATON LIABILITY FOR INSERVICE INSPECTION OF LIGHT WATER FOR THE PBF SEVERE FUEL DAMAGE SCOP!NG AND 11 TESTS.

REACTORS Semiannual Report. October 1986 March 1987.

NUREGICR4883. REVIEW OF RESEARCH ON UNCERTAINTIES IN NUREG/CR4527 V01: AN EXPERIMENTAL INVESTIGATION OF IN-ESTIMATES OF SOURCE TERMS FROM SEVERE ACCIDENTS IN TERNALLY IGNITIED FIRES IN NUCLEAR POWER PLANT CON-NUCLEAR POWER PLANTS.

NR 8R HEM & ECW W WN DEM NU

/

83 02 D PME VALIDATON OF A REAL TIME SAFT-UT SYSTEM FOR THE INSPECTION OF LIGHT DIVISION OF REACTOR & PLANT SYSTEMS (POST 870413)

WATER REACTOR COMPONENTS. Annual Report. October 1984 -

NUREG/CR-4922: STEAM SEPARATOR MODELING FOR VARIOUS September 1985-NUCLEAR REACTOR TRANSIENTS.

NUREG/CR4655: UNSATURATED FLOW AND TRANSPORT DIVISION OF REGULATORY APPLICATIONS (POST 870413)

THROUGH FRACTURED ROCK RELATED TO HIGH-LEVEL WASTE NUREG/CR-4330 V03: REVIEW OF LIGHT WATER REACTOR REGO.

REPOSITORIES. Final Report - Phase 11-LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-NUREG/CR4821: REACTOR COOLANT PUMP SHAFT SEAL STABIL-guirements That May Have Marginal importance To 8

C N E C 4841 F CTU E UATION OF SURFACE CRACKS g,

NUREG/CR-4938. OCCUPATI' NAL RADIATON EXPOSURES ASSO-EMBEDOED IN REACTOR VESSEL CLADDING.Unstradiated Bend O

CIATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE NU 8

A USER'S GUIDE TO THE NRC'S PIPING FRAC-NU E 959: PERFORMANCE TESTING OF EXTREMITY DOSI-NU E /

4 9 NEN F AGILITY RESEARCH

!903 h ON FE THOUAKE RESISTANT A U Y OF A G

NALOGUES AS D

E E

NURE /

1 DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHOD-APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS.

OLOGY AND TECHNICAL CASES. Onect Empincal Scaling Of Re-DIVISION OF REGULATORY APPLICATIONS (860720-870413) sponse Spectral Amplitudes From Vanous Site And Earthoualce Pa-NUREG/CR4815: DEMONSTRATION TESTING OF A SURVEtL-NR R-4903 V02: SELECTION OF EARTHOUAKE PESISTANT Co t A Be ft DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHOD-DIVISION OF 1EACTOR SYSTEM SAFETY (860720-870413)

OLOGY AND TECHNICAL CASES. Methods For introduction Of Go-NUREG/CR-4550 V05: ANALYSIS OF CORE DAMAGE FREQUENCY j

ological Data into Characterization Of Active Faults And Seismicity FROM INTERNAL EVENTS SEQUOYAH, UNIT 1.

And...

NUREG/CR 4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR4903 V03. SELECTION Of EARTHOUAKE RESISTANT RISKS AND THE POTENTIAL FOR RISK REDUCTION:SEQUOYAH DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHOD-POWER STATION. UNIT 1. Draft For Comment.

OLOGY AND TECHNICAL CASES. Dtalocation Models Of Near-NUREG/CR4700 V2 DAF: CONTAINMENT EVENT ANALYSIS FOR Source Earthquake Ground Motx>n. A Review POSTULATED SEVERE ACCIDENTS: SEQUOYAH POWER NUREG/CR4912: DATING GROUND WATER AND THE EVALUA-STATION. UNIT 1. Draft Report For Comment.

TlON OF REPOSITORIES FOR RADIOACTIVE WASTE.

NUREG/CR4918 V01: CONTROL OF WATER INFILTRATION INTO OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report. October OFFICE OF NUCLEAR REACTOR REGULATION. DIRECTOR (POST 1985 September 1986.

870411)

NUREG/CR4928 DEGRADATION Or NUCLEAR PLANT TEMPERA-NUREG/CR 4783: ANALYSIS OF BALANCE OF-PLANT REGULA-TURE SENSORS.

TORY ISSUES Final Report.

NUREG/CR4936' AN INTEGRATED NUREG/CR4964: UPDATE OF TABLE S-3 NONRADIOLOGICAL EN-GEOLOGICALGEOPHYSICAL.AND GEOCHEMICAL INVESTIGA-VIRONMENTAL PARAMETERS FOR A REFERENCE LIGHT-TION OF THE MAJOR FRACTURES ON THE EAST SIDE OF THE WATER REACTOR. Uranium Mnng. Milling And Ennchrnent.

8 NEW MADRfD EARTHOUAKE ZONE.

DIVISION OF ENGINEERING & SYSTEMS TECHNOLOGY (POST NUREG/CR4950 V01: THE SHORELINE ENVIRONMENT ATMOS-870411)

PHERIC DISPERSION EXPERIMENT (SEADExtExpenment Desenp.

NUREG/CR4885: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED STATES. Comparative Evaluation Of The LLNL tion.

OfVISION OF REACTOR ACCIDENT ANALYSIS (POST 870413) S PRO-DIVISION OF SAFETY REVIEW & OVERSIGHT (85112E 870411)

And EPRI Studies.

NUREG/CR4165: SEVERE ACCOENT SEQUENCE ANALYSI GRAM ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA-NUREG/CR-4653: GASPAR 11 TECHNICAL REFERENCE AND USER TIONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

GUIDE.

Contractor Index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports. If further information is needed, refer to the main citation by the NUREG/CR number.

AECL, CHALK RIVER NUCLEAR LABORATORIES NUREG/CR4848: STEAM GENERATOR GROUP PROJECT. Annual NUREG/CRJ821: REACTOR COOLANT PUMP SHAFT SEAL STABlu-Report - 1985.

TY DURING STATION BLACKOUT.

NUREG/CR4889: ZlRCALOY-4 OXIDATION AT 1300 TO 2400 DE-GREES C.

AMERICAN CHEMICAL SOCIETY NUREG/CR4890: HEAT OF REACTION OF MOLTEN ZlRCONIUM NUREG/CP 0076: PROCEEDINGS OF THE SYMPOSIUM ON CHEMI-WtTH 002.

CAL PHENOMENA ASSOCIATED WITH RADIOACTIVITY RELEASES DURING SEVERE NUCLEAR PLANT ACCIDENTS.

NUREG/CR-4938: OCCUPATIONAL RADIATION EXPOSURES ASSOCl-t ATED WITH ALTERNATIVE METHODS OF LOW-LEVEL WASTE DIS.

POSAL NURE CR 98 D A AT N OF N EAR PLANT TEMPERA.

NUR G/CR4959 PERFORMANCE TESTING OF EXTREMITY DOSI.

TURE SENSORS.

ARGONNE NATIONAL MBORATORY BROOKHAVEN NATIONAL LABORATORY NUREG/CR4842: A STUDY OF NATURAL GLASS ANALOGUES AS NUREG/CR-2331 V06 N3: SAFETY RESEARCH PROGRAMS SPON-i APPLIED TO ALTERATION OF NUCLEAR WASTE GLASS.

SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4964 UPDATE OF TABLE S-3 NONRADIOLOGICAL ENVI.

RESEARCH.Ouarterly Progress Report, July-September 1986.

RONMENTAL PARAMETERS FOR A REFERENCE LIGHT-WATER NUREG/CR-4739: RAMONA-3B CALCUMTIONS FOR BROWNS REACTOR. Uranium Mirung Mdhng And Ennchment.

FERRY ATWS STUDY.

NUREG/CR-4883: REVIEW OF RESEARCH ON UNCERTAINTIES IN t.RIZON A, UNIV. OF, TUCSON, AZ ESTIMATES OF SOURCE TERMS FROM SEVERE ACCIDENTS IN NUREG/CR-4655: UNSATURATED FLOW AND TRANSPORT NUCLEAR POWER PLANTS.

THROUGH FRACTURED ROCK RELATED TO HIGH-LEVEL WASTE REPOSITORIES. Final Report. Phase 11.

CALIFORNIA, UNIV. OF, BERKELEY, CA NUREG/CR-4912: DATING GROUND WATER AND THE EVALUATION NUREG/CR4918 V01: CONTROL OF WATER INFILTRATION INTO OF REPOSITORIES FOR RADIOACTIVE WASTE.

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report, October 1985 EATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES

- Septee M86.

NUREG/CR4082 V05: DEGRADED PIPING PROGRAM PHASE CALIFORNIA, UNIV. OF, SAN DIEGO, CA PElk NRG 87 :

TA ALYTICAL ASSESSMENT NUREG/CR-4903 V03: SELECTION OF EARTHOUAKE RESISTANT OF CIRCUMFERENTIALLY SURFACE-CRACKED PIPES UNDER DESIGN CRITERIA FOR NUCLEAR POWER PLANTS - METHODOLO-

BENDING, GY AND TECHNICAL CASES. Dislocation Models Of Near-Source NUREG/CR 4877: ASSESSMENT OF DESIGN BASIS FOR LOAD-CAR.

Earthquake Ground Motion. A Review.

RYJNG CAPACITY OF WELD OVERLAY REPAIRS.

NUREG/CR4878: ANALYSIS OF EXPERIMENTS ON STAINLESS CALSPAN CORP,(SUBS. ARVIN INDUSTRIES / FRANKLIN RESEARCH j

STEEL FLUX WELDS. Topical Report.

CENTER)

NUREG/CR 4819 V01: AGING AND SERVICE WEAR OF SOLENOlD-BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR LABORATORIES NUREG/CR-2850 V05: POPULATION DOSE COMMITMENTS, OUE TO POWER PLANTS Volume 1.Operaling Experience And Failure Identifi-cation.

RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1983 NUREG/CR-3231. PIPE.TO-PIPE IMPACT PROGRAM-COLORADO STATE UNIV., FORT COLLINS, CO NUREG/CR4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY NUREG/CR4300 V04 N1: ACOUSTIC EMISSION / FLAW RELATION-SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE RIPRAP TESTING IN FLUMES. Phase L VESSELS Progress Report. October 1986 March 1987-N'.9EG/CR4330 V03: REVIEW OF LIGHT WATER REACTOR REGU-EG40 IDAHO, INC,(SUBS. OF EG&G INC.)

LATORY REQUIREMENTS. Assessment Of Selected Regulatory Re-NUREG/CR 3956: IN SITU TESTINb OF THE SHIPPINGPORT ATOMIC i

quirements That May Have Marginal Irnportance To RisitPostaccident POWER STATION ELECTRICAL CIRCUlTS.

Samphng System. Turtune Missiles. Combustible Gas Control, Charcoal NUREG/CR4165: SEVERE ACCIDENT SEQUENCE ANALYSIS PRO-GRAM - ANTICIPATED TRANSIENT WITHOUT SCRAM SIMULA-NU E CR-4469 V05: NONDESTRUCTIVE EXAMINATION (NDE) REll.

TlONS FOR BROWNS FERRY NUCLEAR PLANT UNIT 1.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR4617: ONSITE ASSESSMENTS OF THE EFFECTIVENESS REACTORS Semannual Report. April-September 1986.

AND IMPACTS OF UPGRADED EMERGENCY OPERATING PROCE-NUREG/CR4469 V06: NONDESTRUCTIVE EXAMINATION (NDE) RELI, OURES.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR4802: AN EVALUATION OF TRAC-PF1/ MODI COMPUTER REACTORS. Semiannual Report. October 1986. March 1987.

CODE PERFORMANCE DURING POSTTEST SIMULATIONS OF SE-NUREG/CR4583 V02: DEVELOPMENT AND VALIDATION OF A REAL.

MISCALE MOD 2C FEEDWATER LINE BREAK TRANSIENTS.

TIME SAFT-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CR-4821: REACTOR COOLANT PUMP SHAFT SEAL STABILi-REACTOR COMPONENTS Anneaf Report. October 1984 Septembe, TV DURING STATION BLACKOUT.

1985.

NUREG/CR4845: AN ANALYSIS OF THE SEMISCALE MOD-2C S-NH-3 NUREG/CR46$3 GASPAR ll. TECHNICAL REFERENCE AND USER TEST USING THE TRAC-PFI COMPUTER PROGRAM.

GUIDE.

NUREG/CR4866: AN ASSESSMENT OF HYDROGEN GENERATION NUREG/CR-4773: DESIGN FEATURES TO FACILITATE INTERNAllON-FOR THE PBF SEVERE FUEL DAMAGE SCOPING AND 11 TESTS.

AL SAFEGUARDS AT MIXED OXIDE CONVERSION FACILITIES, NUREG/CR-4779. NEW DATA TOR AEROSOLS GENERATED BY RE.

ENGINEERS INTERNRION AL, INC.

LEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC NUREG/CR 4623. IN-SITU STRESS MEASUREMENTS IN THE AIR.

EARTH'S CRUST IN THE EASTERN UNITED STATES.

53

54 Contractor Index FBR SAFETY LABORATORY,IBARAKI PREFECTURE, JAPAN NUREG/CR-4674 V03: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-4765: MXS CROSS-SECTION PREPROCESSOR USER'S DAMAGE ACCIDENTS:1984,A STATUS REPORT. Main Wport And Ap-MANUAL.

pendixes A And B.

NUREG/CR4674 V04: PRECURSORS TO POTENTIAL SEVERE CORE HANFORD ENGINEERING DEVELOPMENT LABORATORY DAMAGE ACCIDENTS:1984,A STATUS REPORT. Appendixes C,D And NUREG/CR-3319 R01: LWR PRESSURE VESSEL SURVEILLANCE DO-E.

SIMETRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveil-NUREG/CR4758: A RETRAN MODEL OF THE CALVERT CLIFFS 1 lance Phymes-Dosimetry Data Base Compendium.

PRESSURIZED WATER RE/CTOR FOR ASSESSING THE SAFETY NUREG/CR-4307 V03: LWR PRESSURE VESSEL SURVEILLANCE 00-IMPLICATIONS OF CONTROL SYSTEMS.

SIMETRY IMPROVEMENT PROGRAM.1986 Annual Report, October NUREG/CR-4819 V01: AGING AND SERVICE WEAR OF SOLENOID-1985 - September 1986 OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR POWER PLANTS. Volume 1. Operating Expenence And Failure identifu HARVARD SCHOOL OF PUBLIC HEALTH, BOSTON, MA cation.

NUREG/CP-0086 V01: PROCEEDINGS OF THE 19TH DOE /NRC NU-CLEAR AIR CLEANING CONFERENCE. Held in ONTARIO HYDRO NUREG/CR4098: SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL-b Etl OF THE 19TH DOE /NRC NU.

ERATED-AGED CLASS 1E BATTERY CELLS.

NU C

8 CLEAR AIR CLEANING CONFERENCE. Held in Seattle,WasNngton, August 18 21,1986.

PARAMETER, INC.

NUREG/CR-4663: CLOSEOUT OF lE BULLETIN tl3 01: FAILURE OF RE-LAWRENCE BERKELEY LABORATORY ACTOR TRIP BREAKERS (WESTINGHOUSE DB-50) TO OPEN ON NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL AUTOMATIC TRIP SIGNAL.

WASTE REPOSITORY. Volume 2: Tuff.

NUREG/CR 4664: CLOSEOUT OF IE BULLETIN 83-04. FAILURE OF LAWRENCE LIVERMORE NATIONAL LABORATORY THE UNDERVOLTAGE TRIP FUNCTION OF REACTOR TRIP BREAK.

NUREG/CP 0087:

SUMMARY

REPORT OF THE SYMPOSIUM ON SEIS-ERS.

MIC AND GEOLOGIC SITING CRITERIA FOR NUCLEAR POWER PENNSYLVANIA STATE UNIV., UNIVERSITY PARK PA PLANTS.

NUREG/CR-4161 V02: CRITICAL PARAMETERS FOR A HIGH-LEVEL NUREG/CR-4726: EVALUATION OF PROTECTIVE ACTION RISKS.

I WASTE REPOSITORY. Volume 2: Tuff.

NUREG/CR4800: SIGPl.A USER'S MANUAL FOR FAST COMPUTA-REMOTE TECHNOLOGY CORP.

TION OF THE PROBABILISTIC PERFORMANCE OF COMPLEX SYS-NUREG/CR4815: DEMONSTRATION TESTING OF A SURVEILLANCE TEMS.

ROBOT AT BROWNS FERRY NUCLEAR PLANT. Analysis Of Costs NUREG/CR-4885: SEISMIC HAZARD CHARACTERIZATION OF THE And Benefits.

EASTERN UNITED STATES. Comparative Evaluation Of The LLNL And EPRI Studies.

RONDOUT ASSOCIATES,INC.

NUREG/CR-4899:

COMPONENT FRAGILITY RESEARCH NUREG/CR4822: BROAD BAND SEISMIC DATA ANALYSIS. September PROGRAM Phase ! Component Pnontization.

1984 September 1986.

LOS ALAMOS NATIONAL LABORATORY SANDIA NATIONAL LABORATORIES NUREG/CR 4615 V02: MODELING STUDY OF SOLUTE TRANSPORT NUREG/CR-3925 REV:

SWIFT 11 SELF TEACHING IbSOR USER'S CURRICULUM.litustratrve Problems For The Sandia Waste-Isolation N E /C 476 MX COSE R

Flow And Transport Model For Fractured Media.

MANUAL NUREG/CR-4098: SEISMIC-FRAGILITY TESTS OF NEW AND ACCEL-NUREG/CR4814: SOURCES OF CORRELATION BETWEEN ERATED-AGED CLASS 1E BATTERY CELLS.

EXPERTS.Empincal Results From Two Ertremes NUREG/CR-4527 V01: AN EXPERIMENTAL INVESTIGATION OF IN-NUREG/CR4875: CHARACTER (ZATION OF CRUSHED TUFF FOR THE TERNALLY IGNITIED FIRES IN NUCLEAR POWER PLANT CONTROL EVALUATION OF THE FATE OF TRACERS IN TRANSPORT STUDIES CABINETS.Part 1: Cabinet Ef'ects Tests.

IN THE UNSATURATED ZONE.

NUREG/Cd4901: EFFECTS FROM INFLUENT BOUNDARY CONDI.

NUREG/CR4550 V05: ANALYSIS OF ("RE DAMAGE FREQUENCY FROM INTERNAL EVENTS: SEOUOYAH, UNIT f.

TIONS ON TRACER MIGRATION AND SPATIAL VARIABILITY FEA.

NUREG/CR-4650 V06PT1: ANALYSIS OF CORE DAMAGE FREQUEN.

TURES IN INTERMEDIATE-SCALE EXPERIMENTS.

CY FROM INTERNAL EVENTS: GRAND GULF. UNIT 1. Main Report.

M ARYLAND UNIV. 0F, COLLEGE PARK. MD NUREG/CR 4550 V06PT2: ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG/CR4918 V01: CONTROL OF WATER INFILTRATION INTO CY FROM INTERNAL EVENTS: GRAND GULF, UNIT 1. Appendices.

NEAR SURFACE LLW DISPOSAL UNITS. Annual Report. October 1985 NUREG/CR-4551 V2 DRF: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION:SEQUOYAH September 1988.

l POWE9 STATON, UNIT 1. Draft For Comment MASSACHUSETTS INSTITU'E OF TECHNOLOGY, CA84 BRIDGE, MA NUREG/CR4551 V3 PT1: EVALUATION OF SEVERE ACCIDENT RISKS

)

NUREG/CR4922: STEAM SEPARATOR MODELING FOR VARIOUS AND THE POTENTIAL FOR RISK REDUCTION. PEACH I

NUCLEAR REACTOR TRANSIENTS.

BOTTOM. UNIT 2. Main ReportDraft For Comment.

NUREG/CR-4551 V3 PT2; EVALUATON OF SEVERE ACCICENT RISKS MATERIALS ENGINEERING ASSOCIATES,INC.

AND THE POTENTIAL FOR RISK REDUCTION. PEACH NUREG/CR4841: FRACTURE EVALUATON OF SURFACE CRACKS BOTTOM. UNIT 2. Appendices. Draft For Comment.

EMBEDDED IN REACTOR VESSEL CLADDING.Unirradiated Bend NUREG/CR4551 V4 DRF: EVALUATION OF SEVERE ACCIDENT Specm i Results.

RISKS AND THE POTENTIAL FOR RISK REDUCTION GRAND NUHEG/CH4894. A USER'S GUIDE TO THE NRC'S PIPING FRAC.

GULF, UNIT 1. Draft For Comment TURE MECHANICS DATA BASE (PIFRAC).

NUREG/CR 4679-QUANTITATIVE DATA ON THE FIRE BEHAVIOR OF COMBUSTIBLE MATERIALS FOUND IN NUCLEAR POWER WITRE CORP NUREG/CR 4783-ANALYSIS OF BALANCE OF PLANT REGULATORY NUR 4

ENCL U E ENVIRONMENT CHARACTERIZATION ISSUES Final Report' TESTING FOR THE BASE LINE VALIDATION OF COMPUTER FIRE SIMULATION CODES.

OAK RIDGE NATIONAL LABORATORY NUREG/CR4700 V2 DRF: CONTAINMENT EVENT ANALYSIS FOR NUREG/CR-2000 V06 N2.

LICENSEE EVENT REPORT (LER)

POSTULATED SEVERE ACCIDENTS: SEQUOYAH POWER r

COMPILATION For Month Of February 1987.

NUREG/CR-2000 V06 N3: LICENSEE EVENT REPORT (LER)

STATION. UNIT 1. Draft Report For Comment NUREG/CR4700 V4 DAF: CONTAINMENT EVENT ANALYSIS FOR COMPILATION For Month Of March 1987, NUREG/CR-2000 V06 N4.

LICENSEE EVENT REPORT (LER)

POSTULATED SEVERE ACCIDENTS: GRAND GULF NUCLEAR STATION.UNfT 1. Draft For Comment COMPILATION For Month Of Apnl 1987 NUREG/CR-2000 V06 N5: LICENSEE EVENT REPORT (LER)

NUREG/CR-4719. COOLABILITY OF STRATIFIED UO2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL.The D13 Expenment COMPILATION For Month Of May 1987.

NUREG/CR 4219 V03 N2: HEAVY-SECTON STEEL TECHNOLOGY NUREG/CR4772: ACCIDENT SEQUENCE EVALUATION PROGRAM PROGRAM Semiannual Progress Report For Apnt September 1986.

HUMAN RELIABILITY ANALYSIS PROCEDURE.

NURE G/CR-4651: DEVELOPMENT OF RIPRAP DESIGN CRITERIA BY NUREG/CR4830: MELCOR VAllDATION AND VERIFICATION 1986 RIPRAP TESTING IN FLUMES Phase 1.

PAPERS.

Contractor inde 55 SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY NUREG/CR-4903 V02: SELECTION OF EARTHOUAKE RESISTANT SCIENCE APPLICATIONS.

DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO.

NUREG/CR-4550 V06PT1: ANALYS!S OF CORE DAMAGE FREQUEN-GY AND TECHNICAL CASEb. Methods For introduction Of Geological CY FROM INTERNAL EVENTS: GRAND GULF,0 NIT 1. Main Report.

Data into Charactenzation Of Actsve Faults And Seismicity And....

NUREG/CR-4550 V06PT2: ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG/CR 4903 V03: SELECTION OF EARTHOUAKE RESISTANT CY FROM INTERNAL EVENTS GRAND GULF. UNIT Lapper' dices.

DESIGN CRITERIA FOR NUCLEAR POWER PLANTS METHODOLO-GY AND TECHNICAL CASES. Dislocation Models Of Near Source SRI INTERNATIONAL Earthquake Ground Motion. A Revie*

NUREG/CR-4950 V01: THE SHORELINE ENVIRONMENT ATMOS-PHERIC DISPERSION EXPERIMENT (SEADEX).Expenment Desenp-UNITED ENGINEERS & CONSTRUCTORS, INC. (SUBS. OF RAYTHEON tion' CO.)

NUREG/CR-4921: ENGINEERING AND OUALITY ASSURANCE COST STRUCTURAL & EARTHOUAKE ENGINEERING CONSULTANTS ACMS ASEARD WW EM MM MODRCATM

+

NUREG/CR-4903 V01: SELECTION OF EARTHOUAKE RESISTANT VANDERBILT UNIV., NASHVILLE, TN DESIGN CR;TERIA FOR NUCLEAR POWER PLANTS - METHODOLO-NUREG/CR-4936: AN INTEGRATED GEOLOGICALGEOPHYSICAL,AND GY AND TECHNICAL CASES. Direct Empincal Scahng Of Response GEOCHEMICAL INVESTIGATION OF THE MAJOR FRACTURES ON Spectral Amphtudes From Vanous Site And Eamquake Parameters.

THE EAST SIDE OF THE NEW MADRID EARTHOUAKE ZONE.

4

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

l There were no NUREG/f A reports for this quarter.

1 57 l

1

{

Licensed Facility Index 1

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.

l S 412 Beaver Valley Power Statoi, Unt 2. Duquesne NUREG1057 S05 50 416 Grand Gulf Nuclear Stabon, Unt 1, Messspp NUREG/CR4550 V06FT1

)

Ught Co.

Power & Lght Co.

54412 Beant Vaney Power Stabon, Urut 2, Duquesne NUREGt259 50416 Grand Gulf Nuclear Staten. Umt 1, Misssspp NUREG/CR4550 V06PT2 bght Co.

Power & bght Co.

STN 50-456 Bradwood Stabon, Urut 1 Commonwealth Edson NUREG1002 S03 50-416 Grand Gulf Nuclear Stabon, Umt 1. Massspp NUREG/CR4551 V4 DAF Co.

Power & bght Co.

ST450456 Bradwood Staton, Unt 1. Commonwealth Edson NUREG-1261 50416 Grand Gull Nuclear Staton, Umt 1. Messsppi NUREG/CR-4700 V4 DRF Co.

STN 50457 Bradwood Stabon, Urut 2. Commonweauh Edson NUREG1002 S03 Power & bght Co-S 133 Humboldt Bay Power Plant Urut 3, Pacdc Gas & NUREG-1166 Ca Electnc Co ST450457 Bradecod Siabon, Unt 2, Commonwealth Edson NUREG1261 40 2061 Kerr-McGee Chemcal Corp., West Chcan fL, NUREG4904 S01 Ca 40-6563 Malknchrodt Chemeal Works, St Loue,,JO, NUREG1239 S 259 Browns Ferry Nuclear Power Stabon, Umt 1, NUREG/CR-4165 50-245 Mdistone Nucles Power Stabon, Urut 1, NUREG1184 DRFT 50 259 uclear e Stabon, Urul 1 NUREG/CR4739 50 277 Pe ottorn t Power tabon, Urut 2, NUREG/CR 4551 V3 PT1 50 259 Browns Feny Nuclear war Stabon, Umt 1, NUREG/CR4815 S 277 P

er Stabon, dat 2, NUREG/CR-4551 v3 PT2 54260 Browns any Power Station Umt 2.

NUREG/CR-4739 50 227 Nuclear n Urvt 1, Tennessee NUREG/CR-4550 V05 S 260 Browns any Nucle; Stabon, Und 2.

NUREG/CR48t5 54327 sh le Plant, Umt 1, Tennessee NUREG/CR4551 V2 DAF S 296 Browns Feny Nuclear Power Statot Unt 3.

NUREG/CR4739 Valley Authonty Tennessee Vaney Authon 50-327 Sequoyah Nucbar Plant, Urvt 1, Tennessee NUREG/CR4700 V2 DAF 50 296 Browns Feny Nuclear Power Stabon, Unt 3.

NUREG/CR4815 Valley Authenty Tennessee Valley Authon STN-50 a98 South Texas Project Unt 1, Houston bghbng & NUREG4781 S03 50 317 Calvert Chfts Nuclear Power Plant Urvt 1, NUREG/CR4758 Power Co Baltimore Gas & Electnc STN 50-499 South Texas Protect Urut 2 Houston Ughbng & NURE40781 S03 50461 Ointon Power Stabon, Umt 1, Ilhnom Power Co. NUREG 1235 Power Co.

54323 Diatio Canyon Nuclear Power Plant, Urvt 2, NUREG1269 54320 Three Mde Island Nuclear Station, Urut 2, NUREG4683 S02 Pacdc Gas & Electnc Co General Pubhc Ubirbes 59

3 I

=

NAC FOZM 24 U S. NUCLE AR K80VLATO3V COMMIS&.;g 9 R E*OR Y NUMSE R (Agg.pn,g gy T,,C, edst yet p, if any, f2 54i

',y O',"38 BIBLIOGRAPHIC DATA SHEET NUREG-0304 Vol.

No. 2 SEE INSTRUCTION 5 0N REVERSE

/

3 TIT LE AND SU9 f tT LE 3 LE AVE 9 LANK Regulatory a Technical Reports (Abstract Index Journal)

Compilation fo Second Quarter 1987 g,,,,,co,,,,,,,

April-June ooN T,,

,,AR 6 AUTMOR(5p

// 6 DAf f REPORT ISSUED j

3 N T.,

veAR Augu 1987 7 #ERFORMING ORGANIZAYlON NAME ANO M ING ADDR E85 f tec/vae le Codel 8 PRO T T ASn%ORK UNIT NUM9ER Division of Publicatio Services Office of Administration nd Resources Management

//rN oa oa^N' NuMaia U.S. Nuclear Regulatory C ission Washington, DC 20555 f

f

..eN,_Re.N., A T eN N A.E.Ne MA L.Ne.. s,,_,. C,

x

... ~. - E eRT Reference Same as 7, above.

/

,,,,,,,co,,,,,,,,,,,,,,,,,,

d'

(

f April-June 1987

,2 su,,uME N T AR v ~oT E.

j i

/

,,,,,,AAC,,....,

y

^ This journal includes all formal report -

the NUREG series prepared by the NRC staff j

and contractors; proceedings of confere s and workshops; as well as international r.greement reports.

The entries in thi pilation are indexed for access by title and abstract, secondary report numbe per nal author, subject, NRC organization for staff and international agreements, ontrac r, international organization, and j

' licensed facility.

}

J

/

s

/

4 DOCvMENT AN ALv5l3 - a EE v

• D5 DE$CRWTORS 5 5 A V AILASt LIT Y STATEMENT compilation abstract inde Unlimited 16 SECURITY CL A$$4FICATION IThan 00,09

. ioE NT,,ie Rs,o E ~.otoTEaMs Unclassified IThus recoet)

Unclassified I? NUMBER OF PAGE8 16 PReCE l

• u. 5. COVE AhmLhf PAlbf thc orr Ict 1987 1st.602663201

UNrTED STATES

$PECIAL FOURTM-CLASS AATE NUCLEAR REGULATORY COMMISSION POSTA 8Idc us WASHINGTON, D.C. 20666 s..

PENA F

Rfv TE E 432 120555078877 1 1ANIAC19L1901 US N R C - 0 A R M - A ') M DIV 0F PUH SVCS l

Main Citations

[_0 j Y

8.

pus MGT BR-PCR NUREG and Abstracts WASHINGTON DC 205S5

.i I

Secondary Report Number Index i

Personal Author index Subject index i

t I

NRC Originating Organization Index (Staff Reports) l T <I NRP Originating Organization 8'" l Iridex (International Agreements) c; C<

Zi m

co '

• t NRC Contractor SponsorIndex

(

Contractor Index l

international Organization Index Licensed Facility Index I

(

(

f f

. _ - _ _ _ - _