ML20100D616
| ML20100D616 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1984 |
| From: | NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V09-N03, NUREG-304, NUREG-304-V9-N3, NUDOCS 8412050837 | |
| Download: ML20100D616 (149) | |
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NUREG-0304 Vol. 9, No. 3 4
Regulatory and Technical Reports (Abstract Index Journal)
Compilation for Third Quarter 1984 July - September l
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Available from NRC/GPO Sales Program Superintendent cf Documents Government Printing Office Washington, D. C. 20402 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 A,
Microfiche of single copies are available from NRC/GPO Sales Program Washington, D. C. 20555
NUREG-0304 Vol. 9, No. 3 Regulatory and Technical Reports (Abstract Index Journal)
Com ailation for Thirc Quarter 1984 July - September D;te Published: November 1964 Division of Technical Information and Document Control Office of Administration U.S. Nuclear Regulatory Commission Washington, D.C. 206Ei5 l
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CONTENTS Preface.........................'...-""""'"'"'*""'""*"**
Index Tab Main Citation and Abstracts............................................................. 1
. S taff R eports.......................................................................
C onference Proceedings.............................................................
Contractor R eports..................................................................
Contractor Report Number index......................................................... 2 Personal Author index.................................................................. 3 S u bject Inde x......................................................................... 4 NRC Originating Organization index (Staff Reports)......................................... 5 NRC Contract Sponsor index ( Contractor Reports).......................................... 6 Contractor Index....................................................................... 7 Licensed Facility index.................................................................. 8 c-iii
Ca N
F L PREFACE I
This compilation consists of tu' bis ~ographic 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 compelation quarterly and to cumulate it annually. Your comments will be ap-s preciated. Pleseo send them to:
Division of' Technical Information and Document Control Policy and Publications Yr inction
. ia Branch Publishing and Translations S Woodmont 501.
U.S. Nuclear Regu Commiselon Washington, D.C.
t The main citations and abstracts in this comp 61stion are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, and NUREG/CR-XXXX. These precede the following indexes:
Contractor Report Number Index Personal Author index
' Subs'ect Index NRC Origmatm* g Organization index (Staff Reports) -
2 NRC Contract Sponsor Index (Contractor Reports) 3 Contractor Index Licensed Facility Index
- A detailed explanation of the entries precedes each index..
The bibliographic elements of the main citations are the following:
Staff Report NUREG-CliO8: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.
ANDERSON,'C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048. 09670:200.
I' Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for intomal 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 1
Laboratory. May 1981.141 pp. 810 man 9en. ANL-81-3. 00832:070.
' Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC intomal use).
Co a..cior Report NUREG/CR-15ti6: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R.
Sandia Laboratorise. May 1981.100 pp. 8107010449. SAND 80-0029. 08912:242.
Where the entries are (1) report nutnber, (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 intomal use).
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.. Availability of NRC Publications m
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Copies of NRC staff and contractor reports may be purchased either from th6 NRC-GPO Sales Office or
?s from the National Technical Information Earvice, Springfield, Virginia 22161.' To purchase documents 3 from the NRC-GPO Sales Office send a check or money order, ptyable to the Superintendent of.
! Documents, to t*,s following addreesj
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2, 8 3-L U.S. Nuclear Regulatory Commission S ATTN: Seles *.tm
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- You may charge'any purchase to your GPO Deposit Account, Master Charge card, or VISA charge card -
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i by calling the NRC GPO Sales Office on (301) 482-9630. Non-U.S. customers must make payment in advance either by intomational Postal Money Order, payable to the Superintendent of Documents, or.
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.by draft on c United Stajas or Canadien bank, payable to the Superintandent of Documents.
NRC ibport Codes -
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~ The NUREG designation,-NUREG-XXXN,ind: cates' 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 r
. identification replaces contrictor 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 repo.rted.
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3 in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-spoux, red conference proceedings.
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Main Citations i nd Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is an NRC staff originated report, NUREG/CP-XXXX is an NRC sponsorec. conference report, and NUREG/CR-XXXX is an NRC contractor-prepared report. The bibliographic information (see Preface for details) is followed by a brief abstract of the report.
NUREG-0020..V08 N06 LICENSED OPERATING REACTORS STATUS
SUMMARY
REPT. Data As Of May 31,1984.(Grey Book)
- Division of Budget &
Analysis.
July 1984 3Y4pp.
8408160191.
26123:001 The OPERATING UNITS STATUS REPORT - LICENSED OPERATING HEACTORS provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office of Resource Management from the Headquarters staff of NRC's Office of Inspection and Enf orcer4ent, from NRC's Regional Offices, and from utilities.
The three sections of the report ares monthly highlights ano statistics for cummercial operating units, and errata from previously reported catal a compilation of detailed information on each unit, provided oy NHC's Regional Offices, IE Headquarters and the utilities; and an appendix for miscellaneous information suen as spent fuel storage capability, reactorayears of experience and non-power reactors in the U.
3 It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U. S. energy situation as a whole.
NUREG=0020 V08 NOT: LICENSED OPEHATING REACTORS STATUS
SUMMARY
REPORT. Data As Of June 30,1964.(Grey Book) e Division of Budget &
Analysis.
August 1964 400pp.
8409200282.
26605:001.
See NUREG-0020,v00,N06 abstract.
NUREG-0020 VOS N08: LICENSLD OPERATING REACTORS STATUS
SUMMARY
REPORT.Dete As of July 31,1984.(Grey Book)
- Division of Buoget &
Analysis.
September 1984 410pp.
8410180141.
27181:143 See NUREG=0020,V06,H06 abstract.
NUMEG=0040 V08 N02: LICENSLE CONTRACTCR AND VENDUR INSPECTION STATUS
(
REPORT. Quarterly Report, April-June 1984,(nhite Book).
- Division of QA, Safeguards & Inspection Programs (Post 830103).
July 1984 433pp.
8408130186.
200473001.
This periodical covers the results of inspections performed by the NRC's Vendor Program Branch that have been distributed to the inspected organizations during the perico from April 1984 througn June 1984 Also included in this issue are the results of certain 1
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PAGE 2
inspections performed. prior to April 1984 that were not included in previous issues of NUREG=0040 1
NUREG=0090 V07~N01: REPORT Tu CONGRESS ON ABNORMAL 4 /
Director's Of fice. / July 1984 OCCURRENCES. January = March 1984 I
52pp.
8408290353. 26299:305 identifies Section 208 of the Energy. Reorganization Act of 1.974 an abnormal.cccurrence as an unscheduled-incid?9t or event wnich tne Nuclear Regulatory Commission determines to be signiflcant from the standpoint of public necith or safety and requires a quarterly report of such events to be made to Congress.
Thls report c' overs tne period January.1 to March 31, 1984 During the report period, there were threecabnormal occurrences at.the nuclear power plants licensed by the NRC to operate.
Tne first
. involved an inoperable containment spray system # the second' involved a through well crack in a vent header inside a BnR containment torus; and the third involved a serious degradation of a reactor depressurization system.
There were two abnormal occurrences for the other NRC licenmees.
The first involved an overexposure to a member of the public7.and tne second involved a therapeutic medical misedministration.
There was one abnormal occurrence reported by the t
i l
Agreement Statest the event involved an overexposure of a radiographer l.
and assistant.
l
.The report also contains information updating some previously reported abnormal occurrences.
NUREG=0304 V09 N02: REGULATORY AND TECHNICAL REPORTS. Compilation For
'Second Guarter 1984
- Division of Technical Information & Document Control.
August 1984, 180pp.
8408300279 26331:346..
reports This compilation lists all NRC regulatory and technical
' published under the NUMEG serles during the second quarter of 1984 NUREG=0386 003: UNITED STATE 3 NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDUNE DIGLST.
- Office of the Executive Legal Director.
- . Aspen bystems, Inc.
July 1984 569pp.
8408230125.
l 26232:033 This third edition of the NRC Staff Practice and Procedure-l '
. Digest, prepared by-Aspen Systems Corporation under contract w(tn;the l
NRC, contains. digests of board decisions issued during the perico from July'1, 1972 to December 31, 1981 interpreting tne NRC's rules of
(
practice in 10 CFR Part 2 This third edition replaces the second i
L edition ano its three supple'ments and contains additional material on F
decisions issued tnrough the end of 1981.
Tnis third edition also l
.contains multiple inoices not included in previous editions of tne
-digest.
L LThe third edition of the' digest will oe supplementea periodically j
l-with updated replacement page supplements.
NUREG=0420 SG6 SAFETY EVALUATION REPORT RELATED TO THE OPERATION UF SHOREHAM NUCLEAR POWER STATION, UNIT h0
- 1. Docket No. 50-322 (Long Island Lighting Company)
- Division of Licensing.
July 1984 33pp.
8408220107.
26199 304 l
Supplement No. 6 (SSER 6) to the Safety Evaluation Report on Long l
. Island Lighting Company's application for a license to. operate tne Shoreham Nuclear Power. Station, Unit 1, located in Suffolk County, New York, has been prepared Dy tne Office of Nuclece Reactor Regulation of L
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k d
4 the U. 3. Nuclear Regulatory Commission.
This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued.
NUREG-0420 307: SAFETY EVALUATION REPORT RELATED TO.THE OPERATION OF SHOREHAM NUCLEAR POWhR STATION UNIT No.1. Docket N'.
50-322. (Long o
Division of Licensing.
September 1984 Island Lighting Company)
- 161pp.
8410100158 26904: 073 Supplement 7 (SSER 7) to the safety Eyaluation. Report on Long Island Lighting Company's Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regule*ory Commission.
This supplement addresses several items that have been reviewed by the staff since the previous supplement was issueo.
-NUREG-0430.V04 N02: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference. Data, July 1983-December 1983.(Buff Book)
- Director's Office, Uffice of Inspection and Enforcement.
August 1984 16pp.
8409170410, 26498:278.
NRC is commi,tted to the periodic publication of licensed fuel facilities inventory difference data, following anency review of the information and completion of any related NRC investigations.
Information>in thls. report includes inventory difference data for active fuel fabrication facilities possessing more than one effective i
kilogram of high enricned uranium, low enriched uranium, plutonium, or uranium-233.
NUREG=0540 V06 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILAbLL.May
!=31, 1984 Division of Technical Information & Document Control,
, July.1984 649pp.
8408130049, 260358001.
This document.is a monthly publication containina descriptions of information received and generated by the U S. NRC.
This information includes.(1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketeu material received and generated by NRC pertinent to its role as a regulatory agency.
The following indexes are included:
Personal 4
i~
Author. Indow, Corporate Source Index,. Report Number-Index, and Cross I
Reference to Principal Documerts Index.
NUREG-0540 V06 N06: TITLE LIST OF DOCUMENTS MADE PUBLICLY l
-AVAILA8LE. JUNE 1=30, 1984.
- Division of Technical Information &
Document Control.
August 1984 710pp.
8409200392.
26603:001.
.See NUREG-0540,V06,H05 abstract, t
'NUREG=0540-V06 N07: TITLE-LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE l'
July 1=31, 1984 Division of-Technical Information & D'cument o
Control.
September 1984 600pp.
8410110527 26943:001 j-See NUREG-0540,v06,N05 abstract.
I l
NUREG=0606 V06 NO3 UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As of August l
17,1984. (Aqua Book)
- Division of Safety Technology.
August 1984
(
57pp.
8409260639 26730:201 Provides an overview of the status of the progress and plans for l
3 l
L
resolution of the generic tasks addressing " Unresolved Safety Issues" as reported to Congress.
NUREG=0649 R018. TASK ACTION PLANS FOR UNRESOLVED SAFETY ISSUES RELATLD TO NUCLEAR POWER PLANTS.
- Division of Safety Technology.
Septemoer
-1984 200pp.
8410170313 270268081.
This document contains Task Action Plans for generic tasks addressing Unresolved Safety Issues (USIs) related to nuclear power plants.
Progress on USIs is reported to Congress each year in the NRC Annual Report pursuant to the requirements of Section 210 of the Energy Reorganization Act of 1974. as amended.
In addition, the NRR issues NOREG-0606, " unresolved Safety Issues Summary, Aous Book" on a quarterly basis 8 this report provides current schedule information for each USI.
.The Task Action Plans in this document include a description of the issue, a description of the NRC staff's approach to resolving the issue, a general discussion of the basis for continued operation and licensing pending resolution of the issue, a discussion of the technical organizations involved in the task, the requirements of manpower and program support funding, interactions with outsidt organizations and potential problems.
This document does not include Task Action Plans for generic tasks addressing USIs for which reports providing the NRC staff resolution of the issue have been puc ' hed.
Those tasks for whicn reports have been published are ide
.. and the reports are referenced.
The Task Action Plans for active USIs are revised on a yearly basis.
This report contains the 1984 revisions to the Task Action Plans.
NUREG-0675 S24: SAFETY EVALUATION REPONT RELATED TU THE OPERATION OF DIABLO CANYON NUCLEAR POAER PLANT, UNITS 1 & 2. Docket Nos. 50-275 &
50-323.(Pacific Gas 6 Electric Company)
- Division of Licensing.
July 1984 16pp.
8408130002 26038:295 Supplement 24 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323),
has been prepared by tne Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission.
This supplement reports on tne 1
independent design verification program (IDVP) for Diablo Canyon Unit 1 that was performed between November 1981 and May 1984 in response to Commission Order CLI-81-30 and an NRC letter and its application by PGhE in the Internal Technical Program (ITP).
Specifically, Supplement 25 presents tne final resolution of the remaining issues and other matters identified in Supplements,18, 19 and 20 This SER Supplement applies only to Diaolo Canyon Unit 1.
NUREG-0675 525: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)
- Division of Licensing.
July 1984 122pp.
8408160080 26124:033.
Supplement 25 to the safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared Dy the Office of Nuclear Reactor Regulation of the U. S.
Nuclear Regulatory Commission.
This supplement reports on the staff's inspection and evaluation efforts on the matter of piping and piping supports as reflected oy the seven technical license conditions in our 4
. " Order Modifying License" issued by the Office of Nuclear Reactor Regulation on Apr 18, 1984. 1 NUREG-0675 326: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF l 'DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2.
- Division of l
Licensing. July 1984. 204pp. 8408220346, 26200:001. Supplement 26 to.the Safety Evaluation Report for Pacific Gas and Electric Company's application for'11 censes to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Cocket,Nos. 50-275 and 50-323), has-been prepared jointly by the Office of Nuclear Reactor Regulation and the Hegion V Office of the U. S. Nuclear Regulatory Commission. This supplement reports on tne status of the staff's investigation, inspection and evaluation o'1 those allegations or concerns that nave been identified to the NRC as of July 1,-1984 The report specifically addresses those allegations which the staff determinec must be satisfactorily resolved prior to full power operation of 4 Diablo Canyon Unit 1. NUREG-0675 S27: SAFE 1Y EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR P0nEH PLANT, UNITS 1.AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company) Division of Licensing. July 1984 31pp. 8408160250 26122:235 . Supplement No. 27 to the Safety Evaluation Report for pacific Gas and Electric company's application for licenses to operate Diaolo Canyon Nuclear Power Plant, Units 1 and 2 (Dockets 50-275 and 50-323), has been prepared:oy the Office of Nuclear Reactor Regulation of the 'U. S. Nuclear Regulatory Commission. This supplement reports on tne independent design verification program (IDVP) for Diablo Canyon and conditions contained in Amendment No. 10 to the Operating License. NUREG-0680 3058-TMI-1 NESTART.An Evaluation Of The Licensee's Management Integrity As It Affects Restart Of Three Mile Island Unit , 1. Docket 50-289.
- Division of Licensing.
July 1984, 179pp. 8408060413 '25937:102. Supplement 5 to the Safety Evaluation Report (SEP) on THI-1 Restart documents the review by the Nuclear Regulatory Commission 4 (NRC) staff of nine investigations conducted by the NRC Office of Investigations into matters identified as relevant and material to an evaluation of the licensee's " management integrity." The staff has included, as part of its evaluation,' materials from its review of the GPU v.'B&W 1awsuit record (NUREG-1020LD, "GPU v. B&W Lawsuit Review and Its Effect on TMI-1") as well as other relevant materials developed since the close of the record in tne THI-1 Restart proceeding. In-developing its position on General Public utilities Nuclear Corporation's character (i.e., management integrity), the staff evaluated matters that cast doubt on the licensee's character, individually and collectively; consioereo the remedial actions taken by the licenseel and balanced past improper conduct of the licensee i - against its subsequent record of. remedial actions ano performance and . record of current senior management of the licensee. Tne staff concluded that, while the past improper conduct was grave, the -remedial actions taken, the subsequent record of performance, and the record of current senior management support of finding that GPUN can and will operate TMI-1 witnout undue risk to the nealth and safety of the public. 5 t
.NUREG=0748 V04 N058 OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data Asl0f May 31,1984.(Orange dook)
- Office of Resource Management, Director.
July 1984. -200pp.. 8407170557, 25630:001 The Operating Reactors Licensing Actions. Summary is designed to provide the Management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing,with operating power and nonpower reactors. NUREG-0748 V04 N06: UPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of June 30,1984 (Orange Book)
- Of'fice of Resource Management, Director.
July 1984 400pp. 8408080363 25981:129 See NUREG=0748,V04,N05 abstract. NUREG=0748 V04 N078 QPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of July 31,1984.(Orange Book)
- Office of Resource Management, Director.
August 1984 150pp. 8409170276. 264978001. See NUREG=0748,v04,tt05 abstract. NUREG=0750 V19 101: INDEAES TO NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JANUARY-MARCH 1984 Division of Technical Information & Document Control., September 1984 73pp. 8410100166 26906:247. Digests and indexes for issuances of the Commission, the Atomic Safety and Licensing Appeal Panel, and the Atom c Safety and Licensing Board Panel, the Administrative Law Judge, the Directors' Decisions, and the Denials of T'titions for Rulemaking. NUREG=0750 V19 N03 NUCLEAR REGULATORY COMMISSION ISSUANCES FOR MANCH 1984 Pages 555-936 Uivi.sion of Technical Information & Document Control. August 1984. 386pp. 8408240187 262503001. Legal issuances of the Commission, the Atomic Safety and Licensing Appeal Board, the Atomic Safety and Licensing Boara, tne Administrative Law Judge, and NRC Program Offices. NUREG=0750 V19 N043 NUCLEAN NEGULATORY COMMISSION ISSUANCES FOR APRIL 1984.Pages 937-1,149 e Division of Technical Information & Document Control. August 1984, dOOpp. 8409260625 267053114 See NUREG=0750,V19,N03 abstract. 1 NUREG-0750 V19 N058 NUGLEAN NEGULATORY COMMISSION ISSUANCES _FOR MAY 1984.Pages 1,151-1,321.
- Division of Technical Information &
Document Cor. trol. September 1984 180pp. 8410150096 26999:144 See NUREG-0750,V19,N03 abstract. j NUREG=0787 307: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF -WATERFORD NUCLEAR POWER PLANTguNIT 3. Docket No. 50=382. (Louisiana Power & Light Company)
- Office of Nuclear Reactor Regulation, Director.
September 1984 300pp. 8410110526 26941:001. Supplement'7 to the Safety Evaluation Report for Louisiana Power & Light applicati'n for a license to operate Waterford Steam Electric j o Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been Jointly prepared by the Office of Reactor ' Regulation and the Region IV Office of the U.S. Nuclear Regulatory Commission. This supplement providts the results of the staff's 6
evaluation of approximataly 350 allegations and concerns of poor construction practices at the Waterford-3 facility. NUREG=0798 304: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF ENRICO FERMI ATOMIC P0nEN PLANT, UNIT NO. 2. Docket No. 50-341. .(Detroit Edison Company) *. Division of Licensing. September 1984 91pp. 8410120046, 26983 121. -Supplement No. 4 to the Safety Evaluation Rsport related to the operation of the Enrico Fermi Atomic Power Planta, Unit 2, provides the staff's evaluation of additional information submitted by the application regarding outstanding review issues identified in Supplement,No. 3 to the Safety Evaluation Report, dated January 1983. ~ NUREG-0800 06.2.1 R6 STANDARD REVIEW PLAN FOR THE REVIEW OF-SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition Revision 6 to Section 6.2.1.1.C, " pressure-Suppression Type BWR Containments."
- Office of Nuclear Reactor Regulation, Director.
August 1984 9pp. 8410020474 26791:275 Revision 6 to 8"P Section 6.2.1,1.C of the Stancard Review Plan incorporates the resolution of Generic Issue B-10, " Behavior of BWR/ Mark III Containments." NUREG-0831 305: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF i GRAND GULF NUCLEAR, STATION, UNITS 1 AND 2. Docket Nos. 50-416 And 50-417.(Mississippi Power And Light Company, Middle South Energy,Inc Ano South Mississippi Electric Power Association) e' Division of ~ Licensing. August 1984 122pp. 8409270154, 26719:164 .Suoplement 5,to tne Safety Evaluation Report for,Mississ'ppi Power & Light Company, et al, joint application for licenses to operate-the Grand' Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississipp) River near Port Gibs'on in Claibo,rne County, Mississippi, has been prepared by the Office of Reactor Regulation of the U.S. Nuclear _ Regulatory Commission. This supplement . reports the status on the resolution of those issues that require 'further evaluation before authorizing operation of Unit 1 above 5% of rated power. NUREG=0831 S06: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF GRAND GULF NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-416 And 50-417 (Mississippi Power And Light Company)
- Division of Licensing.
August 1984
- 204pp, 8409260656, 26701:181.
Supplement hv. 6 to the Safety Evaluation Report for Mississippi Power & Light. Company et al joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulator.y Commission. The supplement reports the NRC staff's evaluation of open items from previous l supplements and Technical Specification changes required before . authorizing operation of Unit 1 above 5% of rated power. NUREG-0926 R01: TECHNICAL SPECIFICATIONS FOR GRAND GULF NUCLLAN -STATION, UNIT 1. Docket No. 50-416 (Mississippi Power And Light -Company) HOFFMAN,D.R. Division of Licensing. August 1984 540pp. 8409260633, 26691 001. 7
=- Tho Grcnd Gulf Nucloor Stotien, Unit 1 Technical Specificottons were prepared by the U.S. Nuclecr Regulatory Commission to set forth the limits, operating. conditions-and other requirements applicable to j e nuclear reactor facility as set forth in Section 50.36 of 10 CFR 50 .for;the protection of the health and safety of the public. 1 NUREG-0933 3013.A PRIONITI4ATION OF GENERIC SAFETY ISSUES. EMRIT,R.3 MINNERS,N.; VANDER MOLEN,H.; et al. Division of Safety Technology. July 1984 120pp. 8408220265. 26212:062 The report presentt the priority rankings for generic safety issues related to nuclear power plants. The purpose of these_ rankings is'to assist in the timely and efficient a)1ocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankingt are HIGH, MEDIUM, LOh, and DROP and nave been assigned on the bas.is. risk significance estimates, the ratio of risk to costs and ot tr. impacts estimatea.to result if resolutions of the. safety issues w,'re implemented, and the consiceration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. NUMEG-0935: ACOUSTIC WAVL PROPAGATION IN FLUIDS WITH COUPLED CHEMICAL REACTIONS. MARGULIES,T.S.; SCHWARZ,W.H. Division of Risk Analysis & ) Operations (post 840429). August 1984 44pp. 8409200389 26608:310 This report presents a hydroscoustic theory which accounts for -sound absorption and dispersion in a multicomponent mixture of reacting fluids (assuming a set of first-order acoustic equations without offfusion) such that several coupled reactions can occur i simultaneously. General results are obtained in the foem of a b1 quadratic characteristic equation (called the Kirchhoff-Langevin equation) for the complex propagation variable chi * - (alpha + w/c) in which alpha is the attenuation coeff'cients c is the phase speed of i the progressive wave and w is the angular frequency. Computer nimulations of sound aosorption spectra have been made for enree different chemical systems each comprised of two-step chemical reactions using physico-chemical data available in the literature. The relative chemical reaction and classical viscothermal contributions to the sound absorption are also presented. Several j discrepancies that can arise when interpreting ultrasonic measurements ) for~ estimating thermodynamic data (chemical reaction heats or volume changes) for multistep chemeial reaction systems are discussed. NUREG-0936 V03 N02: NRC REGULATORY AGENDA.Guarterly Report, April-June 1984 Division of Rules and Records. July 1984 201pp. 8408100149 25998:112. .The NRC Regulatory Agenda is a compilation of all rules on which the NRC has proposed or is considering action and all. petitions for rulemaking which have Deen received by the Commission and are pending disposition by the Commission. ~The Regulatory Agenda is updated and issued each quarter. The Agendas for April and October are puolished in their entirety in the Federal Register while a notice of availability.is publisnea in the Federal Register for the January ano July Agendas. 8
i NUREG=0940 V03 N02: ENFORCEMENT ACTIONS SIGNIFICANT ACTIONS RESOLVED. Quarterly Progress Report, April-June 1984 Enforcement Staff. July 1984 363pp. 8408220308 26202:001. This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (April - June 1984) and includes copies of letters, notices, and orders sent by the Nuclear Regulatory Commission to licensees with respect to these enforcement actions and the licensees' responses. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, in the interest of promoting public health and safety as well as common defense and security. NUREG-0954 803: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF CATAn8A NUCLEAR STATION, UNITS 1 AND 2. DOCKET Nos. 50-413 And 50-414.(Duke Power Company,et al)
- Division of Licensing.
July 1984 140pp. 8408070009, 25954 334 The report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 oy the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission with respect to the apolication filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc., as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and SC-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Aylie. This supplement provides additional information supporting tne license for fuel loading and precriticality testing for Unit 1. NUREG-0978: MARK III LUCA-NELATED HYOR0 DYNAMIC LOAD DEFINITION. Generic Technical Activity 8-10. Final Report. FIELDS,M.B.; KUDRICK,J.A. Division of Systems integratton (post 811005). August 1984, 100pp. 8409200450 26628:031. This report, prepared by the statt of the Office-of Nuclear Reactor Regulation and its consultants at the Brookhaven National Lacoratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in coiling water reactor (BhR) facilities witn the Mark III pressureasuppression containment design. Its issuance completes NRC Generic Technical Activity 8-10, " Behavior of BWH Mark III Containmer.t." On the casis of certain large-scale tests conoucted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases and conclude that, with
- few specified changes, the proposed load definitions provide conser/ative loading conditions.
The staff approved acceptance criteria for LOCA-related hydrodynamic loaos are provided in Appendix C of this report. NUREG-0985 R01: U.S. NUCLEAR REGULATORY COMMISSION HUMAN FACTORS 1 PROGRAM PLAN.
- Division of Human Factors Safety.
September 1984 62pp. 8409280095 2613b 253. This document is the First Annual Revision to the NRC Human Factors Program Plan originally published August 1983 The purpose of this document is to ensure that proper 9 i - - ~.
consideration is given to human-factors in the planning, design, construction, operation ano maintenance of nuclear facilities. The plan represents a systematic and comprehensive approach for adoressing human factors concerns important to nuclear power plant safety in the j FY-84 througn FY-86 time frame. The plan adoresses the planning of seven major program elements: 1.0 Staffing and Qualifications, 2.0 Training, 3.0 Licensing i Examinations, 4.0 Procedures, 5.0 Man-Machine Interface, 6.0 Management and Organization, and 7.0 Human Reliacility. Appencix (A) Program Element Schedules. NUREG-1029 A COMPUTER CODE FOR GENERAL ANALYSIS OF HADON RISKS (GARR). GINEVAN,M. Division of Hadiation Programs & Earth Sciences (o'ost 840429). September 1984 96pp. 8410120006. 26986 006 1 I Evaluating t'he level of lung cancer risk associated with a given level-of radon-daughter exposure is a c'omplex matter. There is the question of whether one's elsk assessment should apply absolute risk models or relative risk models and, even when a general model form has been selected, there are decisions as to the exact form of risk projection, the appropriate method of accounting exposure over time, and how much a personal ha6lt such as smoking can modify rism. This document presents a computer model for general analysis of redon risks that allows the user to specify a large number of possible mode.s with a small number of simple commands. The model is written in a version of BASIC which conforms closely to the American National Standaros Institute.(ANSI) definition for minimal BASIC and thus is readily modified for use on a wide variety of computers and, in particular, microcomputers. NUREG-1031 SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MILLSTONE NUCLEAR P0aEN STATION, UNIT NO. 3. Docket No. Division of Licensing. 50-423.(Nortneast' Nuclear Energy Company) July 1984 591pp. 8408160262 26121:001 This report provides the results of the NRC staff review of Northeast Nuclear Energy Company's application for a license to operate the Millstone Nuclear Power Plant, Unit No. 3 The facility is located in nat'erforo Township, New London County, Connecticut. Subject to favorable resolution of the items discussed in the Safety Evaluation Report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public. NUREG=1050 PROBABILISTIC RISK ASSESSMENT (PRA) REFERENCE Division of Risk Analysis & Operations (post DOCUMENT. Final Rept.
- 840429).
September 1984 230pp. 8410120005 26984:001. The Commission's Safety Goal PoIIcy Statement in NUREG-0880, Rev. 1,' directs the st'aff "...to collect available information on PRA studies and prepare a reference document that describes the current status of knowledge concerning'the risk of plants licensed in the U.S." The document discusses the purpose and content of a PRA and identifies tne PRAs and other probabilistic studies performed,to date. It then discusses the level of development and uncertainties associated with the various elements of PRA methodology as well as -those generic. insights derived from studies performed.
- Finally, potential uses of PRA in regulation are evaluated.
10 i .. - - - -. - - -.. _ _ _ ~ _ _.. _, _, _ _ _. _ _ _
4 JUREG-1054 SIMPLIFIED ANALYSIS FOR LIQUID PATHWAY STUDIES. CODELL,R.B. Division of Engineering. August 1984 106pp. 8408300292 26331:001 The analysis of the potential contamination of surface water via groundwater contamination from severe nuclear accldents is routinely calculated during licensing reviews. This analysis is fact.11tated by the methods described in this report, which is codified into a BASIC language computer p ogram, SCREENLP..This program performs simplified calculates population doses to potential users of the contaminated wat9r irrespective of possible mitigation methods. The results are then comparea to similar analyses performeo using cata for.the generic sites in NUREG-0400, " Liquid Pathway Generic Study", to determine if the site being investigatea would pose any unusual liquid pathway hazards. NUREG-1061 V01: REPORT OF THE U.S. NUCLEAR REGULATORY COMMISSION PIPING REVIEn COMMITTEE. Volume 1 Investigation And Evaluation Of Stress Corrosion Cracking In Piping Of Boilint Water Reactor Plants.
- Piping Review Committee.
August 1984 400pp. 8409260636. 26700:045 Severe intergranular stress corrosion cracking (IGSCC) of the recirculation piping system in several boiling water reactors occurred during 1982-1983. A Task Group on Pipe Cracks was established by the U.S. Nuclear Regulatory Commission with the broad charter of developing an integrated program to deal witn the ent'irety of the stress corrosion cracking problem. This report presents specific conclusions and recommendations that are tieo closely to relevant regulatory documents so that necessary changes can be implemented. This report covers aspects such as the causes and descriptions of IGSCC phenomenal current status of pipe cracking in BhR's; nondestructive evaluations of piping welds; inspection of piping for IGSCC; decisions and criteria for replacement, review of continued operation without repairi risks related to the presence of IGSCC; and a value-impact assessment of IGSCC. NUREG-1064: DRAFT ENVINDHMLNTAL STATEMENT RELATED TO THE OPERATIuN OF MILLSTONE. NUCLEAR P0aEN STATION, UNIT 3. DOCKET NO. 50-423 (NORTHEAST NUCLEAR ENERGY COMPANY,et al)
- Division of Licensing.
July 1984 335pp. 8408010148 258/1:012. The information in this statement is the second assessment of tne environmental impact associated with the construction and operation of the Millstone Nuclear Power Station, Unit No. 3, located in naterford Township, New London County, Connecticut. The first assessment was the Final Environmental btatement related to construction issued in February 1974 prior to issuance of the Millstone Construction Permit. The present assessment is the result of the NRC staff review of the activities associated with the proposed operction of the plant. NUREG-1068: REVIEW INSIGnTs UN THE PHOBABILISTIC RISK ASSESSMENTS FON THE LIMERICK GENERATINb STATION, UNIT 1 AND 2. CHELLIAH,E. Division of Safety Technology. August 1984 126pp. 8408290158 26309:151. In recognition of tne high population density around the Limerick Generating Station site and the proposed power level, the Philadelphia Electric Company, in response to NRC staff requests, conoucted and submitted Detween Maren 1961 and November 1983 a probabilistic rism assessment (PRA) on internal event contributors ano a severe accident risk assessment on external event contributors to assess risks posed 11
by operation of the plant. The applicant has developed perspectives using PRA models on the risk profile of the Limerick plant and has altered the plant design to reduce accident vulnerabilities identified in these PRAs. The staft's review of the Limerick PNA has particularly emphasizeo the dominant accident sequences and the i resulting insights into demonstration of compliance with regulatory requirements, unique design features and major plant vulnersoilities to assess the need for any additional measures to further improve the safety of the LGS. The staff's review insights and PRA safety review conclusions are presented in this report. NUREG-10698 SAFETY EVALUATION REPORT.RELATED TO THE RENEhAL OF THE OPERATING LICENSE FOR THE bENERAL ELECTRIC-NUCLEAR TEST REACTOR (GE-NTR).DOCKtT NU 50-73.(General Electric Company) Division of Licensing. September 1984 89pp. 8410180222. 27045:201 This safety Evaluation Report for the application filed by the General Electric Corporation for a renewal of operating license R-33 to continue to operate a research reactor. has been prepared oy tne Office of Nuclear Reactor Negulation of the U.S. Nuclear Regulatory Commission. The facility is owned and operated by the General Electric Corporation and is located in Pleasanton, California. Ihe i staff concludes that tne reactor facility can continue to be operated by GE without endangering the health and safety of the public. NUHEG-10728 TECHNICAL SPECIFICATIONS FOR CATAn8A NUCLEAR STATION, Unit
- 1. Docket No. 50-413. ANDERSON,F.D.
Division of Licensing. July 1984. 525pp. 8408130011. 260393001. The Catawba Nuclear Station, Unit 1, Technical Specifications were prepared by the U. S. Nuclear Regulatory Ccmmission to set forth the limits, operating conditions and other reatirements applicable to a nuctsN' reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. NUREG-10738 DRAFT ENVINONMENTAL STATEMENT RELATED TO THE OPERAi!ON OF RIVER BEND STATION. Docket No. 50-458.(Gulf States Utilities Company Division of Licensing. July Cajun Electric Power Cooperative) 1984 249pp. 8408220325 26199 001. This Draft Environmental Statement contains the second assessment of the environmental impact associated with the operation of River Send Station, pursuant to the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51, as amended, of the Nuclear Regulatory Commission regulations. Tnis l statement examines tne environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs. Comments on tnis statement should be filed no later than 45 days after the date on which the Environmental Protection Agency notice of availability of this statement is pub 11shed in the Federal Register. i l l NUREG-1075 DECENTRALIZATION OF OPENATING REACTOR LICENSING REVIEWS.NRR Pilot Program. HANNON,J.L. Division of Licensing. July 1984 26pp. l 8408080370. 25981:102. This report, which has incorporated comments received from the Commission and ACRS, describes the program for decentralization of selected operating reactor licensing technical review activities. The safety is 2-year pilot program will be reviewed to verify that 12
l enhanced as anticipated by the incorporation of prescribed management techniques and application of resources. If the program fails to operate as designede it will be terminated. The 2-year pilot program will be limited to two operating power plants in each of three regions and will be implemented tos (1) test the method of selecting licensing actions for technical review in the regions, (2) evaluate predictad improvements in the effectiveness of licensing and inspection programs, and (3) verify that safety is enhanced (as anticipated) by incorporating prescribed management techniques and applying regional resources to this technical review function. NUREG-1080 V01: LONG-RANGE RESEARCH PLAN FY 1985-1989.
- Office of Nuclear Regulatory Research, Director.
September 1984 199po. 8410100090 2o903:001. The Long-Range Research Plan (LRRP) was prepared by the Office of Nuclear Regulatory Research (RES) to assist the NRC in coordinating its long-range research planning with the short-range budget cycles. The LRRP lays out programmatic approaches for research to help resolve regulatory issues. Ihe plan will be updated annually. NUREG-1081 POST-ACCIDENT GAS GENERATION FROM RADIALYSIS OF ORGANIC MATERIALS. WING,J. Division of Engineering. September 1984 40pp. 8410190332, 27083:157. This report presents a methodology for estimating the pas generation rates resulting from radiolysis of organic materfats in 1 paints and electrical caole insulation inside a nuclear reactor containment building.uncer design basis accident conditions. Tne methodology was based on absorption of the radiation energies from the post-accident fission products and the assumed gas yields of the irradiated materials. A sample calculation was made using conservative assumptions, plant-specific data of a nuclear power plant, and a radiation source term which took into account the time-dependent release and physico-chemical behavior of the fission products. NUREG-1083: SAFETY EVALUATION REPORT RELATED TO THE HENEWAL OF THE OPERATING LICENSE FOR THE nESTINGHOUSE RESEARCH REACTOR AT ZION,ILLINDIS. DOCKET NO. 50-87.(Westinghouse Electric Company)
- Division of Licensing.
September 1984 74pp. 8410170290 27029:208 This Safety Evaluation Report, prepared by the Office of Nuclear l Reactor Regulation of the U.S. Nuclear 9egulatory Commission, is for en application filed by the Westinghouse Electric Corporation (WLC) for renewal of operating license R-119 The facility is owned and operated by the destinghouse Electric Corporation and is located in the City of Zion. Illinois. The staff concludes that the reactor facility can continue to be operated by 6EC without endanaecing the health and safety of the public. NUREG-10848 SAFETY EVALUATION REPORT RELATED TO THE RENEnAL OF THE OPERATING LICENSE FOR THE HESEARCH REACTOR AT MICHIGAN STATE UNIVERSITY. Docket No. 50-294.
- Division of Licensing.
August 1984 89pp. 8409260645. 20701:086 This Safety Evaluation Report for the application filed by the Michigan State University (MSU) for a renewal of operating license 13
number R=114 to continue to operate the TRIGA research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is owned and operated by the Michigan Stat'e University,and is located on the campus of Michigan State University in East Lansing, Ingham County, Michigan. The staff concludes that the TdIGA reactor facility can continue to be operateo by MSU without endangering the health and safety of the public. i NUREG-1085: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF NINE MILE POINT NUCLEAR STATION, UNIT NO.2. Docket No. 50-410 (niagara Mohawk Power Corpore ion, Rochester Gas & Electric Corporation And 1 Central Hudson Gas & Electric Corporation)
- Division of Licensing.
July-1984 313pp. 6408420318 26201:037. This Draft Environmental Statement contains the assessment of the 1 environmental impact associattJ with the operation of the N)ne Mile Point huclear Station, Unit 2, pursuant to the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51, as amended, of the Nuclear Regulatory Commission regulations. This statement examines the environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs, i NUREG-1092 ENVIRONMENTAL ASSESSMENT FOR 10 CFR PART 72, " LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT FUEL AND HIGH-LEVEL RADI0 ACTIVE WASTE." SCHULTEN,C.S. Division of Engineering Technology. August 1964. 75pp. 8409200397. 26607:253 The Nuclear Waste Policy Act of 1982 (NNPA) addresses the need for development of monitored retrievable storage for spent fuel ana high-level radioactive waste. The Commission has examined its regulations and determined that much of existing 10 CFR part 72 regulations can be used uuring initial design development for a monitoreo retrievable storage installation (MRS), however, changes are needed to 10 CFR Part 72 to clarify specific issues which have been raised by the NWPA. The proposed revisions to 10 CFR,Part 72 establish licensing requirements for a monitored retrievable storage installation. However, unless Congress authorizes construction of an MRS promulgation of these requirements would not result in construction or operation of such an installation. The issues identified as requiring resolution by the proposal amendments are (1) establishing license criteria for the long-term storage of spent fuel and high-level radioactive waste in an MRS, (2) inclusion of license requirements for the long-term storage of spent fuel and hign= level rao.oactive waste in an HMS under 10 CFR Part 72, and (3) elimination of the current restrictions placed on fuel cladding integrity in the present Part 72 which require the fuei cladding be protec.ted against degradetion and gross ruptures, and substitution of restrictions on radioactive releases to the environment. NUREG/CP=0051: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON LEAK-SEFORE-BREAK IN NUCLEAR REACTOR PIPING.
- Division of Engineering Technology.
August 1984 561pp. 8409070237 82. 26410:001 On September 1 and de 1983, the CSNI Subcommittee on Primary System Integrity held a special meet'ing in Monterey, California, on the subject of Leak =defore-Break in Nuclear Reactor Piping Systems. The purpose of the meeting was to provide an international forum for the exchange of ideas, positions, and research results; to identify 14
a o crcco roquiring.cdditional research and developmentf and to determine the general attitude towara acceptance of the leak =before-creak concept. This report documents the presentations made at the meeting in the areas of (1) application of piping fracture mechanics to leak-before-breaks (2) leak rate and leak detection) (3) leak-before-break studies, methods, and results; and (4) current and ) proposed positions on leak =before-break. J NUREG/CP=0053: PROCEEDINGS OF THE NINTH ANNUAL STATISTICS ~ SYMPOSIUM ON NATIONAL ENERGY ISSUES,0ctober 19-21,1983 BRYSON,M.C. Los Alamos Scientific Laboratory. August 1984 200pp. 8408240249 LA=10127=C. 26251:027. The Ninth Annual Statistics Symposium on National Energy Issues was held in Rockville, Maryland, at the Holiday Inn C,rowne Plaza, October 19-21, 1983, under the Joint sponsorship.~ Los Alamos National Laboratory and the Nuclear Regulatory Commission. Sessions jincluded.two cont,r'buted-paper sessions, two tutorial sessions, and i one discussion group. Included in these proceedings are those papers for which final c'py was previoed by the authors, together with a list o of papers present'ed and a IIst of attendees. NUREG/CR=0130 A0003: TECHNULUGY, SAFETY ANL COSTS OF DECOMMISSIONING A 4 REFERENCE PRESSURIZE 0 nATER REACTOR P0nER STATION. MURPHY,E.S. Battelle Memorial Institute, Pacific Northwe'st Laboratories. September 1984 50pp. 8410170293 27026:290. The radioactive wastes expected to result from decommisioning of the reference pressurized water reactor power station are rev}ewed and classified in accordance with 10 CFR 61. The 17,885 cubic meters of waste from DECON are classified as follows: Class A, 98.C%; Class o, 1.2%; Class C, 0.1%. About 0.7% (133 cubic meters) of the waste would be generally unacceptaole for disposal using near-surface disposal methods. NUREG/CR=0169 V17 LOFT EXPEHIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recordeo On DAVDS. EVANS,H.P.; MCKNIGHT,K.D. EGtG, Inc. September 1984 55po. -8410120037 EGG-2037..26977:159.- Uncertainty analyses are presented tt quantify the uncertainty !~ bounds for the Loss-of= Fluid Test (LOFT) process measurements. Ihe C process instruments are.those used to control the plant operation safety. The uncertainties presented are of two types objectlve uncertainties (basically random) which can be duplicated in the I isboratory.and for which data are available, and subjectiva -uncertainties (basically systematic) for which no specific data are l available. l l NUREG/CR=0072 ADD 02: TECHNULUGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATEN REACTOR POWER STATION. Classification of l Decommissioning Wastes. MUNPHY,E.S. Battelle Memorial Institute, Pacific Northwest Laboratories. September 1984 50pp. 8410170289, j -27029:283 i The radioactive wastes expected to result from decommissioning of the reference boiling water reactor pcwer station are reviewed and classified in accordance with 10 CFR 61. The 18,949 cubic meters of waste from DECON are classified as follows: Class Ar 97.5%7 Class or 15_
2.0%; Class C, 0.3%. About 0.2% (47 cubic meters) of the waste would be generally unacceptable for disposal using near-surface methods. t NUREG/CR-1740 R01: DATA SUMMARIES OF LICLNSEE EVENT REPORTS OF SLLLCTED INSTRUMENTATION AND C0ivTHOL COMPONENTS Al U.S. COMMERCIAL NUCLEAR P0nER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. TROJOVSAY,M.; ~ BROWN,S.R.- EG&G, Inc. July 1984. 344pp. 8408240382. EGG-2307. 26249:001 This report describes a computer-based data file developed from Licensee Event Reports (Leks) of instrumentation and control (IEC) components in United States commercial nuclear power plants for the period January 1, 1976, to December 31, 1981. In addition to the 1 creation of the file, summaries of data contained in the file were made to obtain data for risk assessment and statistical purposes, i I Gross constant fault (failure ano command fault) rates were estimated for major components and cnannels that provide a direct reactor trip. Explanations, figures, and summary tables of the results are provided. This report updates ano supersedes the original May 1981 edition of 3 NUREG/CR-1740. NUREG/CR-2000 v03 N6 LICENSEE EVENT REPORT (LER) COMPILATION For Month of June 1984 Cak Ridge National Lacoratory. July 1984. 87pp. 8408070007 ORNL/NSIC-200 25955:112. Tnis monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file of the Nuclear Safety Information Center (NSIC) during the one month period identified on the cover of this document. The LERs, from which this information is derived, are submitted to the feuclear Regulatory Commission (NRC) by nuclear power plant licensees in accordance with federal regulations. Procedures for LER reporting are described in detail in NRC Regulatory Guide 1.16 and NUREG-0161, Instructions for Preparation of Data Entry Sheets for Licensee Event Reports. The LER summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facil?ty. Component, system keywords, and component vencor indexes follow the summaries. The components, systems, and vendors are those identified by tne utility when the LER form is initiated; the keywords are assigned by the computer using correlation tables from the Sequence Coding and Search System. i NUREG/CR-2000 V03 N7: LICENSEE EVENT REPORT (LER) Compilation For Month of July 1984.
- Oak Ridge National Laboratory.
August 1984 10!pp. 8408300287. ORNL/NSIC-200 26331:123. See NUREG/CR-2000,V03,N6 abstract. NUREG/CR-2000 V03 N8 LICENSEE EVENT REPORT (LER) Compilation For Month Of August 1984 Oak Ridge Nati. s1 Laboratory. September 1984. 68pp. 8410100171 URNL/NSIC-200. 26904:001. See NUREG/CR-2000,V03,N6 abstract. NUREG/CR-2015 V08: PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. NELLS,J.E.7 GEORGE,L.L.; CUMMINGS,G.E. Lawrence Livermore National Laboratory. September 1984 188pp. 6409280110. UCRL-53021 V08 26764 001 This document reports on the Phase 1 efforts of the Systems 16 I l
Anclycio projoct to dovolop tho tools cnd mothoda for computing the 4 probability of radioactive release from a commercial nuclear power plant in the' event of an earthquake, j NUREG/CR-2331 V03 N3: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report, July-September 1983. WEISS,A.J. Brookhaven National Laboratory. July 1984 154pp. 8408010179 8NL-NUREG-bl454 25870 181. The Aovanced and nater Reactor Safety Research Programs Querterly Progress Reports have oeen combined and are included in this report entitled, " Safety Research Programs Sponsored by tne Office of Nuclear Regulatory Research - Quarterly Progress Report." This progress report will describe current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Accident Evaluation, Division of Engineering Technology, and Division of Facility Operations of the U. S. Nuclear Regulatory Commission, Office.of Nuclear Regulatory Research. The projects reported see tne following: HTGH Safety Evaluation, SSC Development, Validation and Application, CRBR 6alance of Plant Modeling, Thermal-hydraulic Reactor Safety Experiments, LWR Plant Analyzer Development, LWR Cooe Assessment and Application, Tnermal Reactor Code Development (HAMONA-36); Stress Corrosion Cracking of Pep Steam Generator Tubing, dolting Failure Analysis, Probability,dased Load ComDinations for Design of Category I Structures, Mechanical Piping Benchmark Problemsi Human Error Data for Nuclear Power Plant i Safety-Related Events, and Human Factors in Nuclear Power Plant Safeguards. The previous reports have covered the period October 1, 1976 through June 30, 19d3 l NUREG/CR-2331 V03 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY THE OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report,0ctober 1 -Decemoer 31,1983 WEISS,A.J. Brookhaven National Laboratory. September 1984 129pp. 8410120059. BNL-huREG-51454, 26983:212. The Advanced and hater Reactor Safety Research Programs Quarterly progress reports have been combined and are included in this report entitled, " Safety Research Programs Sponsored by (ne Office of Nuclear l Regulatory Research - uuarterly Progress Report." This progress report will describe current activities ano technical progress in the programs at drookhaven National Laboratory sbonsored by the utvision of Accident Evaluation, Division of Engineering Technology, and Division of Facility Operations of the U.S. Nuclear Regulatory l Commission, Office of Nuclear Regulatory Research. The projects reported are the following: High Temperature Reactor Research, SSC vevelopment, Validation and Application, CRBR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety l Experiments, Development of Plant Analyzer, Code Assessment ano Application (Transient and LOCA Analyses), Thermal Reactor Code Development (RAMONA-38), Calculational Quality Assurance in bupport of PTS; Stress Corrosion Cracking of PWR Steam Generator Tubina, oolting Failure Analysis, Probability. Based Load Combinations for Design of Category I Structures, Meenanical Piping Benchmarking Problems, Identification of Age-deleted Failure Modes; Analysis of Human Error l Data for Nuclear Power Plant Safety-Related Events, Human Factors in Nuclear Power Plant Safeguards, Emergency Action Levels, and Protective Action Decision Making. The previous reports have covered the period October 1, 1976 through September 30, 1983 17
'NUREG/CR-2482 V05: REVIEn-UF DOE WASTE PAChAGE PROGRAM. Subtask 1.1 National Weste Package Program, April 1983 - September 1983. 300,P. Brookhaven National Lacoratory. August 1984 122pp. 8409170270. BNL-NUREG-51494. 26498 04S. Tnis report addresses part of an ongoing task to review tne national-high-level waste package effort. It includes evaluations of reference waste form, container and packing material components with respect to determining how tney.may contribute to the containment and controlled release of racionuclides after waste packages have oeen emplaced in salt, basalt, ano tuff repositories. A section on caroon steel container corrosion is includeo to complement prior work on TiCode-12 and Type 34 stainless steel. Use of crushed tuff as a packing material is otscussed, and waste package c omp'o nen t interaction test data are included. Licensing data requirements are specified. NUREG/CR-2499 REVIEn uF ENEHGENCY RADIOLOGICAL INSTRUMENTATION AND ANALYTICAL METHODS AI NMSS-LICENSEE SITES. HERRINGTON,n.N.; KATHREN,R.L.; nEN0 YEN,J.L.; et al. Battelle Memorial Institute, Pacific Northwest Lacoratories. August 1984. 60pp. 8409200301 PNL-4163. 26609:001. This report provides a brief review of emergency radiolonical monitoring instrumentation capabilities based on visits to Nuclear Material Safety and safeguards (NMSS) licensees ano on a review of the open literature. Recommenaations based on findings are made witn regard to instrument design and operation, training, calibration, testing, analytical methods, sampling procedures, and quality assurance. An assessment of currently available instrumentation is made with respect to types of instruments, instrument specifications, ano the future needs of NRC/NMSS licensees as seen by instrument manufacturers and to what extent those needs will be met. NUREG/CR-2576: BAR FULL INTEGRAL SIMULATION TEST (FIST)== Facility Description Report. STEPHENS,A.G. General Electrlc Co. September 1984 267pp. 8410120002 GEAP-22054 26955:100 A new boiling water reactor safety test facility (FIST, Full Integral Simulation Test) is described. It w111 be used to investigate small breans and operational transients and to tie results from such tests to earlier large break test results determined in the TLTA. Tne new facility's full height and prototypical comp'onents constitute a major scaling improvement over earlier test facilities. A heated feedwater system, permitting steady state operation,,ano a large-increase in the number of measurements are other significant improvements. Program background is outlined and program objectives defined. Design basis is presented together with a detailed, complate description of the facility and measurements to be made. An extensive component scaling analysis and prediction of performance are presented. The report is intended to serve as a reference document for those needing detailed information about the facility. NUREG/CR-2996: SENSITIVITY OF DETECTING IN-CORE VIBRATIONS AND BUILING IN PRESSURIZED WATER REACTORS USING EX-CURE NEUTRON NOISE. SWEENEY,F.J.; RENIER,J.P. Oak Ridge National Laboratory. July 1984 98pp. 8409110086 ONNL/TM-8549 26447:003 Neutron transport and diffusion theory space-and energy-dependent reactor kinetics calculations were performed in the frequency domain to determine the sensitivity of an ex-core neutron detector to in-core vioretions and coolant boiling in a pressurized 18
wotor rocctor (PhR). The rocults of those calculations indicoto thot the ex-core, detectors,are sensitive to neutron sources, to vibrations, and.to boiling occurring over large regjons of the core. Calculations were also performed to determine the effects of fuel burnup, boron concentration, and xenon poisoning on the spatial sensitivity of the ex-core detector. Tnese calculated results show that fuel assemoly vibrations would be expected to produce "60% greater ex-core detector response at the end of the first fuel cycle at Sequoyah-1 compared to the beginning of the fuel cycle for a constant amplitude of vibration. The results were compared uith experimental ex core neutron noise data obtained from Sequoyah-1 during the first fuel cycle. The predicted increase in ex-core neutron noise was experimentally observed in the 2.5= to 4.0=Hz frequency range (the range of frequencies associated with fuel assembly vibration), indicating that the vibrational amplitude of the fuel assemblies did not increase significantly during the first fuel cycle. NUREG/CR-3053: CLOSEQUT.uF IE BULLETIN 80-08: EXAMINATION OF CONTAINMENT LINER PENETRATION nELOS. DEAN,R.S.; FOLEY,W.J.; HENNICK,A. Parameter, Inc. July 1984 31pp. 8408130265 IEB-80-08 260383074 During an NRC inspection at Nine Mile Point 2, examination ey radiography of prjmary containment liner penetration sleeve-to-process pipe (flued head fitting) welds revealed rejectable defects not originally found by ultrasonic examination. Apparently,. ultrasonic signals from the welo backing bar masked signals from defects. Further investigation founo similar problems at Beaver Valley 2 ano North Anna 3 and 4 It nulletin 80-08 was issued to acquire information from all facilities to determine the generic nature of the problem. It was found that, because of evolution of the ASME duclear Code, plants under construction aesigned to that Code since about 1974 are required to volumetrically examine these welds, and so, in general, do not have the problem. Operating plants, built to earlier codes not requiring such design and examination for the containment welds, present a concern for the quality of this type of weld and for the integrity of the primary containment bounoary. Bulletin status is closed for all but 11 tacilities. Recommendations are made for resolution of the problem for these facilities. These include meaningful radiograpnic examination of welds of concern, if possible, and if not, licensee justification for not making a radiographic examination. NUREG/CR-3139 SCENARIUS AND ANALYTICAL METHODS FOR UF6 RELEASES AT NRC-LICENSED FUEL CYCLE FACILITIES. SIMAN-TOV,M.; DYKSTRA,J.; HOLT,D.D.; et al. Oak Ridge National Laboratory. July 1984 97pp. 8408130007 .ORNL/ENG/TM-25, 26038:203 Tnis report identifies and discusses potential scenarios for the accidental release of UF(6) at NRC-licensed UF(6) production and fuel fabrication facilities based on a literature review, site visits, and DOE enrichment plant experience. Calculational methods needed for analyzing such releases are also reviewed. Accident scenarios are { presented under the headinus of cylinder failcres, process system failures, criticality events, and operator errors and are categorized by location, release source, UF(6) phase prior to release, release flow characteristics, release causes, initiating events, and UF(c) inventory at risk. Releases identified for further examination includes (1) a release from a cylinder outdoors, (2) a release from a l pista11 or cylinder in a steam chest, and (3) an indoor release from i 19
oithor (c) o pigtail or cylindor or (b) other indoor sourco doponding on facility design'and operating procedures. Indoor release pnenomena may be analyzed using a time-dependent homogeneous compartment model or a more. complex hyorodynamic model if time-dependent, spatial variations in concentrations, temperature, and pressure are important. Analytical tools for modeling directed jets ano explosive releases are discussed as we11 as some of the complex phenomena to be considered in analyzing UF(6) releases both indoors and outdoors. NUREG/CR-3169: SUPER SYSTEM CODE (SSC,REV. 0).AN ADVANCED THERM 0 HYDRAULIC SIMULATION FOR TRANSIENTS IN LMFBRS. GUPPY,J.G. Brookhaven National Lauoratory. September 1984 460pp. 8409260630. BNL-NUREG-51650. 26702:163. Tne Super System Code (SSC) calculates the response of nuclear reactor systems during operational, incidental and accidental, transients, especially natural circulation events. Modules simulated and parameters calculateo includes core flow rates and temperatures, loop flow rates and temperatures, pump performance, ano heat exchanger operation. Additionallyt all plant protection systems and plant control systems are accounted for. All calculations are done in SI units.SSC is a general system transient code. It is highly flexiole, with complete variable dimensioning, allowing any number of user specified loops, pipes and nodes. Single phase and two phase thermal hydraulics are used in a multi-channel core representation. Interessembly flow redistribution is accounted for; a detailed fuel pin model is used. The heat transport system geometry is user specified. The code has both transient and steady state options. Restart capability is provided. SSC is available in either a CDC UPDATE format or as FONTRAN source. The customary transmittal package also includes the input files for the three standard benchmark problems, as well as 48x microfiche which contain tne SSC support documentation and sample output for each of tne benchmark problems. SSC is currently available as a draft release from brookhaven National Lacoratory with NRC consent. NUREG/CR-3190: PLUGM A COUPLED THERHAL-HYDRAULIC COMPUTER MODEL FOR FREEZING MELT FLOW IN A CHANNEL. PILCH,M. Sandia Laboratories. MAST,P.K. Science Applications, Inc. September 1984 140pp. 8410120042 SAND 82-1580. 269773001 PLUGM models the flow and freezing of molten material in a nonmelting channel. PLUGM is being developed for applications in Sandia's Ex= Vessel Core Hetention Materials Assessment Program and in Sandia's LMFBR Transition = Phase Program. PLUGM mocols time-dependent flow from a reservoir, tnrough a channel and possibly into a catch tank. Three user-specified geometry options enable realistic modeling of melt flow and freezing in tubes, thin slits, and particle beds. Axial variation of relevant channel parameters is possible. Sample problems, pertaining to ex-vessel core retention and LMFBR transition phase, illustrate features and capabilities of the code. NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REACTOR PRESSURE SOUNDARY COMPONENTS. Annual Report For 1983.
- Materials Engineering Associates, Inc.
Septemoer 1984 135pp. 8410170224 MEP-2051, 27031:213 The objective of this research program is to characterize 20
cotoriolo bohovicr in colotion to structural sofoty and rolichility of pressure boundary components for light water reactors.. Specific objectives include developing an understanding of elastic-plastic fracture and environmentally-assisted crack propagation phenomena in terms of continuum mechanics, metallurglcal variables, and neutron irradiation. Emphasis is placed on identifying metallurgical factors responsible for radiation embrittlement of steels and on developing procedures for embrittlement relief, including guidelines for radiation = resistant steels. The underlying goal is,the interpretation of material properties performance to establish engineering criteria for structural reliaoility and long-term operation. Current work is organized into three major tasks (1) fracture mechanics investigations,-(2) environmentally-assisted crack growth in high temperature, primary reactor water and (3) radiation sensitivity and postieraciation properties recovery. Research progress in these tasks for 1983 is summarized here. i NUREG/CR-3273: COM6uSTION UF HYDROGEN AIR MIXTURES IN THE VGES CYLINDRICAL TANK. BENEGICK,W.B.; CUMMINGS,J.C.; PRASSINOS,P.G. Sandia Laboratories. July 1984 165pp. 8408100155 SAND 83-1022, 25997:308 Sandia National Laboratories is currently involved in a numoer of experimental projects to provide data that wil.1 help quantify the threat of hydrogen comoustion curing nuclear plant accidents. Several experimental facilities are part of the Variable Geometry Experimental System (VGES). The purpose of this report is to document the experimental results from the first round of combustion tests performed at one of these facilities a 5-m(3) cylindrical tank. The j data provided by tests at this facility can be used to guide further testing ano for the development and assessment of analytical models to predict nyorogen comoustion behavior. NUREG/CR-3318: lwr PRESSURE WESSEL SURVEILLANCE DOSIMETRY IMPROVEMENI PROGRAMPCA Experiments,ulind Test,And Physics-Dosimetry Suoport For The PSF Experiments. McELROY,W.N. Hanford Engineering Development l Laboratory. September 1984 150pp. 8410190172 HEDL-TME 84-1, 27059:001, This report was prepared tot 1) Serve as a general reference document containing Dencnmarked experimental and tneoretical data and information required to determine and certify the accuracy of the experimental and analytical methods and data that are recommenced in a series of ASTM LnR pressure vessel surveillance standards; 2) Provide detailed experimental ano theoretical results to determine tne l limiting accuracy of transport theory calculations for preoicting l dosimetry sensor reaction rates and derived values of neutron exposure parameters (total fluence, fluence greater than 0.1 ano 1.0 Mev, and dpa) for LhR pressure vessel benchmark fields simulating steel-water configurations of commercial power reactors; 3) Assess the accuracy of the methodology used to translate measured pressure vessel steel damage and exposure data (and the corresponding uncertainties) obtained at surveillance locations to the pressure vessel Deltline l region; 4) Provide PGA 4/14 and 4/12 SSC configurations' experimental I and tneoretical physics = dosimetry results in support of the " PSF l Experiments and Blind rest." After an executive summary, a description of the PCA experimental test facility is provided in Section 1 followed by the presentation and discussion of experimental measurements and cata in Sections 2, 3, I ano 4 The results of neutronic calculations by participants are l 21
given end roforenced in Section 6., The cosperison and evoluotion of measured and derived data are considered in Section 7 NUREG/CR-3346: B10 ASSAY OATA AND A RETENTION-EXCRETION MODEL FOR SYSTEMIC PLUTONIUM. LEwGLTT,R.W. Oak Rioge National Laboratory. July 1984 .96pp. 8408130009 ORNL/TM=8795. 26038:108 The estimation of systemic burdens from urinalyses has neon the most common and useful method of quantifying occupational exposures to plutonium. Problems arise in using this technique, however, because l of inadequate modeling of human retentfon, translocation, and excretion of this element. Present methods for estimating tne systemic burden from urinalyses were derived to a large extent from patterns ooserved in tne first few months after exposure, but there is now evidence that these same patterns do not persist over long periods. In fact, recent comparisons of autopsy data with urinalyses suggest that extrapolation to extended periods based on these observed patterns usually leads to a large overestimate of the systemic burden at times more than a few years after exposure. In this report we collect and discuss human and animal data for Pu together with general . physiological properties neeaed for the interpretation of oioassay results. This information is used to develop a mechanistic model of the movement, retention, and excretion of systemic pu. This model appears to be a reasonably accurate predictor of excretion for times ranging from one day to several decades after contamination of blood. NUREG/CR-3369: AN UNCERTAINTY STUDY OF PnR STEAM EXPLOSIONS. BERMAN,M.3 SWENSON,0.V.; nICMETT,A.J. Sandia Laboratories. July 1984 91pp. 8408130006 SAND 83-1438. 26040:238 of containment Some previous, assessments of the probability failure caused by.in-vessel steam explosions in a PhR have recognized large uncertainties and assigned broad ranges to the probability, while others have concluded that the probacility is small or zero. In i this report we st'udy the uncertainty in,the probability of containment failure ny comnining tne uncertainties in the component physical processes using a Monte Carlo method. We conclude that, despite substantial research, the combined uncertainty is still large. Some areas are identified in which improvements in our understanding may lead to large reductions in the overall uncertainty. I i NUREG/CR-3418: SCREENING TESTS OF TERMINAL 8 LOCK PERFORMANCE IN A SIMULATED LOCA ENVIRONMENT. CRAFT,C.M. Sandia Laboratories. August 1984 275pp. 8410120045 SAND 83-1617 26981:001. Twenty-four terminal blocks were tested in simulated Design Basis Event (DBE), Loss of Coolant Accident (LOCA) environments. The terminal blocks were powered at voltages of 4 Vdc, 45 Vde, and 125 Vdc. Resulting currents associated with these voltage levels were 1.8 i mA,'20 mA, and 1 A, respectively. Terminal-to-terminal and terminal-to-ground leakage currents were monitored on a discrete time i -basis throughout the test. Based on these measurements, insulation resistance were calculated. During exposure to the LOCA steam environment insulation resistance was observed to decrease fron. initial values of 10(8) to 10(10) ohms et 10(2) to 10(5) ohms. These t decreases in IR are interpreted as being caused by conduction in surface moisture films rather than bulk conduction through the insulation material. Insulation resistance for.all applied voltage l 1evels appear to be approximately the same. Sporadic breakdowns lasting from fractions of a second to several minutes were ooserved. 22
Further, rapid increases In applied voltage caused large decreases in insulation resistance. The measured IR wasHelso dependent upon temperature. Subsequent to the test, terminal block insulatio'n resistance returned to acceptable levels (10(6) to 10(8) ohms), though not to pre-test levels. Tne comparison of spray and no-spray results shows that no discernable difference in irs existed between the periods with and witnout chemical spray. NUREG/CR-3459: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT) BUNDLE B-5. CHAPMAN,R.H.; CROWLEY,J.L.; LONGEST,A.n. Oak Ridge National Laboratory. August 1984 173pp. 8409110096 ORNL/TM-8889 26444: 157. B-5 test data are presented and interpreted to the extent necessary for understanding pertinent features of the 8 x 8 test. Objectives of the test were to investigate.the effects of array size and rod-to-rod interactions on cladding deformation in the high-alpha-Zircaloy temperature range under conditions that simulated the adiabatic heatup (reneat) phase of a light-water-reactor loss-of-coolant accident. Test conditions, nominally the same as used in an earlier 4 x 4 (B-3) test, were~ conducive to large deformation. The fuel pin simulators were electrically heated (3.0 kW/m) anc were slightly cooled with a very low flow (He
- 140) of low-pressu,re superheated steam.
Cladding temperature increased at a rate of 9.6 l degrees C/s. The. simulators burst in a very narrow temperature range, with an average of 768 degrees C. Cladding burst strain ranged from 32-to 95%, with an average of 61%. Heated length volumetric expansion ranged from 35 to 79%, with an average of 52%. Average burst strain was slightly greater for the interior than for the exterior simulators; average volumetric expansion was significantly greater. Maximum coolant channel flow area reduction was 69% for the entire 8 x 8 array, 83% for tne interior 6 x 6 array, and 91% for the central 4 x 4_ array. The results show deformation was greater in the Dundle interior and suggest rod-to-rod mechanical interactions caused axial propagation of the deformation. NUREG/CR-3460: EXPERIMENT DATA REPORT FOR MULTIR0D BURST TEST (MRBT) i SUNDLE B-6 CH APM A N, d.it. ; LONGEST,A.W.; CROWLEY,J.L. Oak Ridge National Laboratory. July 1984 157pp. 8409110104 ORNL/TM-8890 i 264443004 A reference source of MRBT bundle B-6 test data is presented with i minimum interpreta. tion. The primary objective of this 8 x 8 multirod burst test was to investigate cladding deformation in the alpha-plus-beta-ZIrcaloy temperature range under simulated light-water-reactor (LnR) loss-of-coolant accident (LOCA) conditions. I B-6 test conditions simulateo the adiabatic heatup (reheat) phase of a l LOCA and produced very uniform temperature distributions. Tne fuel pin simulators were electrically heated (average linear power generation of 1.42 kW/m) and were slightly cooled with a very low flow (Re " 140) of low-pressure superheated steam. The cladding temperature increased from the initial temperature (330 degrees C) to the burst temperature at a rate of 3.5 degrees C/s. Tne simulators l burst in a very narrow temperature range, with an average of 930 l degrees C. Cladoing burst strain ranged from 21 to 56%, witn an average of 31%. Volumetric expansion over the heated length of the cladding ranged from to to 32%, with an average of 23%. The average burst strain and the average volumetric expansion for the interior simulators were only slightly greater than tne averages for the exterior simulators. The coolant channel flow area reduction was 23
( r v modest, with a maximum of 39% for the entire 8 x 8 array, 43% if based on the interior 6 x 6 array, and 45% if based on the central 4 x 4 array. As expect'ed, no evidence of rod-to-rod mechanical interaction i effects was observed. i NUREG/CR=3469 V01: OCCUPATIONAL DOSE REOUCTION AT NUCLEAR POWEH PLANTS ANNOTATED SIBLIOGRAPHY OF SELECTED READINGS IN RADIATION PROTECTION l AND ALARA. 8AUM,J.W.; SCHULT,0.A. Brookhaven National Laboratory. September 1984 125pp. 8410120021. BNL=NUREG=51708 26984:216. This report contains selected abstracts on dose and dose reduction at nuclear power plants. Abstracts were derived primarily from APPLIED HEALTH PHYSICS ABSTRACTS AND NOTES, Volume e, No. 1, 1980 through Volume 9, No. 1, January 1983. Subsequent reports will contain additional sostracts from earlier and more recent literature. NUREG/CR=3470s AThS AT BR0nNS FERRY UNIT ONE = ACCIDENT SEuuENCE ANALYSIS. HARRINGTON,R.M.; HODGE,S.A. Oak Ridge National Laooratory. July 1984, 234pp. 8409110093. ORNL/TM-8902. 26446 001. This study describes the predicted response of Unit One at browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mecnanisms by which voids are formed in the moderator / coolant play a dominant rola in the propression of the accident. Actions taken by the operator greatly influence tne quantity of voids in the coolant and the effect is analyzed in tnis report. The progression of the accident sequence under existing andunder recommended prococures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented. NUREG/CR=3474: LONG= LIVER ACTIVATION PRODUCTS IN REACTOR MATERIALS. EVANS,J.C.; LEPEL,E.L.; SANDERS,R.W.; et al. Battelle Memorial Institute, Pacific Northwest Laboratories. August 1984 179pp. 8409200285 PNL-4824 26600: 190. The purpose of this program was to assess the problems posed to reactor decommissioning by long-lived activation products in reactor construction materials. Samples of stainless steel, vessel steel, concrete, and concrete ingredients were analyzed for up,to 52 elements in order to develop a data base of activatable major, minor, and trace elements. Large compositional variations were noted for some thor'ugn evaluation was made of all possible nuclear elements. A o reactions that could lead to long lived activation products. It was concluded that all major activation products have been satisfactority accounteo for in decommissioning planning studies completed to date. A comparison is made between calculated activation levels and regulatory guidelines for shallow land disposal according to 10 CFR 61. Most of the massive components were found to qualify as eitner ~ Class A or Class B waste with the exception of PWR and BWR shroud material which clearly exceeds Class C limits. Selected samples of activated steel and concrete were subjected to a limited radiochemical analysis program as a verification of the computer model. Reasonsoly good agreement with the calculations was obtained where comparison was 24
possible. In particular, the presence'or 94Nb in activated stainless steel at or somewhat above expected levels was confirmed. NUREG/CR-3480: VALUE/ IMPACT ASSESSMENT FOR SEISMIC DESIGN CRITERIA USI Aa40. COATS,0.W.; LAPPA,D.A. Lawrence Livermore National Laboratory. August 1984 138pp. 8409200279 UCRL-53489. 26606:053 In October, 1981, tne Nuclear Regulatory Commission approved a reorganization that resulted in the establishment of the Committee to Review Generic Requirements (CRGR). The charter for the CRGR requires that written justification accompany all proposed new regulatory requirements submitted to the CRGR for review. At the request of the Nuclear. Regulatory Commission's Generic Issues Branen, Lawrence Livermore National Laboratory has provided the required written justification to accompany proposed new requirements to SRP Sections 3.7.1., 3.7.2, and 3.7.3. These proposed new requirements are the result of technical studies performed, as part of the Unresolved Safety Issues (USI) A-40 program, by LLNL and others. NUREG/CR-1161, " Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria," by LLNL, provided the technical resolution to USI A-40 and was the Desis for the proposed new recommendations. The report contained herein present a technical evaluation and value/ impact assessment of the proposed new requirements. 1 NUREG/CR-3492 V04: MIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION QUARTERLY PROGRESS REPORT, October-December 1963 BALL,S.J.; CLEVELAND,J.C.; HARRINGTON,R.M.; et al. Oak Ridge National Lacoratory. July 1984 27pp. 8408130001 ORNL/TM-8921/V4 26011:225 Development worm continued on models ano codes for predicting source terms in both tne Fort St. Vrain (FSV) and 2240-Mn(t) lead plant reartors. Experimental work on fission-product vapor pressures and diffusion rates through graphite continued on temperatures up to 2775 K, and a mathematical model of the experimental system was i. developeo to aid analysis of the results and to guide improvements in the system and experiment design. Benchmarking of the BLAST steam generator code continued using FSV data, and more support work was done for proposed FSV core bypass flow model verification. Progress was made in setting up cooperative high-temperature gas-cooled reactor 1 (HTGR) safety research witn the Federal.Repuolic of Germany. A review of a FSV technical specification on limiting maximum core temperature was begun. l i l NUREG/CR-3493: A REVIEn UF THE LIMERICK GENERATING STATION SEVERE i ACCIDENT RISK ASSESS.4EnT. Review of Core Melt Frequency. AZARM,M.A.; bARI,R.A..; BOCCIO,J.L.; et al. Brookhaven National Laboratory. July 1984 192pp. 6408220333 BNL-NUREG-51711. 26200:205, l A limited review is performed of the Severe Accident R%sk Assessment for the Limerick Generating Station. The review considers the impact on the core-melt frequency of seismic-and fire-initiating events. An evaluation is performeo of methodologies usea for l determining the event frequencies and their impacts on the plant components and structuret. Particular attention is given to uncertainties and critical assumptions. Limited requentification is performed for selected core-melt accident sequences in order to 11'ustrate sensitivities of the results to the underlying assumptiuns. 1 25
-NUREG/CR-35133 MECHANICAL HELIABILITY EVALUATION OF ALTERNATE MOTORS FOR USE IN A RADIDIODINE AIR SAMPLER. BIRD,S.K.; HUCHTON,R.L.; MOTES,8.G.; et al. EG6G, Inc. July 1984 41pp. 8409110069 n!NCO-1006 26437:264 The purpose of the study was to, evaluate the mechanical, reliability of mo, tors for use in a prototype system. designed,for post accident collection end measurement of radiciodine in the environs of a nuclear reactor..The two types of motors were tested for lifetimes and operational performance. characteristics.under extremes of 1 temperature, relative numioity and/or in dusty air and rainfall. The 12 volt airect current motors exhibited satisfactory performance under i all environmental conditions and demonstrated lifetimes of 47 hours, o97 hours and 188 hours. Tne 12 volt direct current voltage and satisfactory operation on. alternating current. voltage;.at failure the AC/DC voltage motors demonstrated lifetimes of nominally 6 hours, 3-1 hours and 2 hours. The ofrect current voltage only motors are the better candidates for incorporation into the air sampler. NUREG/CR-3518 V01: SLIM-MAUD:AN APPROACH TO ASSESSING HUMAN ERROR .PROBASILITIES USING STHUCTURED EXPERT JUDGEMENT. Volume I Overview of SLIM-MAUD. EMBREY,0.E.; HUMPHREYS,P.; ROSA,E.A.; et al. Brooknaven National Laboratory. July 1984 36pp. 8408010166. 8NL-NUREG-51716 25667 253 .This two-volume report presents the procedures and analyses in developing an approach for structuring exoert judgments to. estimate human error probabilities. Volume I presents an overview of work performed in developing the approach: SLIM-MAUD (Success Likelinood Index Methodology, implemented through the use of an interactive computer program calleo MAUD--Multi-Attribute Utility Decomposition). Volume _II provides a more aetailed analysis of the technical issues underlying the approach. NUNEG/CR-3520 V01: LONG-TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR PONER. PLANTS. Volume I Summary And Users Guide. 'O'BRIEN,J.N.; FAINSERG,A. Brookhaven National Laboratory. August 1984 42pp. 8409280071. 8NL-NUREG-51718, 26750:228 The first task was to identify and rank human factors affecting .the quality of nuclear power plant safeguards in terms of their importance. The opinions of over 85 experts were solicitea and, responses were received. These responses were rigorously analyzed to ascertain what human factors could be considered important to power plant safeguards. In addition, the Safeguards Summary List (NUREG-0525) was systematically analyzed for human factors influences. Also, relevant government and industry l}terature was reviewed. These data sources were then aggregated and an overall importance ranking of human factors issues was developed.. Tnis part of the research effort is fully documented and described in Chapter 2 of Volume II. The second part of this effort involved determining the feasibility of conducting research in the areas found to be important to power plant safeguards. A determination of research feasibility l was based on the practicality, usefulness, and acceptability of conducting research and using the results in a regulatory context. 4 This part of the effort is fully documented in Chapter 3 of Volume II. Research efforts addressing human factors in safeguards were then developed and prioritized according to the importance of human factors areas derived in the first part of the study and the feasibility of research determined in tne second part. Research was also grouped to 26
[ take advantage.of common research approaches and data sources where appropriate. Chapter 4 of Volume II details the development of methodological groupings for.cotimizing resource use. i NUREG/CR-3520 V02: LONG-TENM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR POAEH PLANTS. Volume II Development Of Jetailed Analyses. O'BRIEN,J.N.; FAINSERG,A. Brookhaven National Laboratory. August 1984
- 204pp, 8409280062.
BNL-NUREG-51718 267623001. See NUREG/CR-3520,V01 abstract. NUREG/CR-3524: ORGANIZATIONAL. INTERFACE IN REACTOR EMERGENCY PLANNING AND RESPONSE. SORENSEN,J.H.; COPENHAVER,E.D.; MILETI,0.S.; et al. I Oak Ridge National Laboratory. July 1984 52pp. 8409110100 ORNL-6010 26446:230 The purpose of this research was to determine if existjng regulations have ~ led to effective interfaces between utilities and offsite organizations in emergency planning and response. Findings suggest-that regulations have provided the necessary framework for achieving adeauste interfaces. That interface has been achieved is demonstrated by comprenensive response plans and good coheslveness among organizations involved in emergency.resoonse. Interface problems identified in the research,can be reduced by better implementation of existing regulations rather than by revision of existing ones. NUREG/CR-3544: BETA PARTICLE MEASUREMENT AND 00SIMETRY AT NRC-LICENSED FACILITIES. RATHbuN,L.A.; ENDRES,G.W. FOX,R.A.; et al. Battelle Memorial Institute, Pacific Northwest LaDoratories. August 1984 58po. 8409070234, PNL-48a6 20419: 203. Researchers from Pacific Northwest Laboratory (PNL) have conducteo beta radiation measurements under laboratory and field conditions to assess tne degree of tne measurement problem and offer suggestions for possible remedies. The primary measurement systems selected for use in this study were the sillcon (Si) surface barrier spectrometer system and the multielement beta dosimeter. Three boiling water reactors (dWRs), two pressurized water reactors (PnRs), and one fuel fabrication facility were visited during the course of the study. Although beta fields from cobalt-60 were the most common type found at locations associated with spent fuel handling, liquio radioactive waste, and BnR turbine components, Commercially-available dosimeters and survey instruments were used to measure the same [ laboratory and licensee facility beta fields characterized witn PNL's l active and passive spectrometers. A prototype survey. meter was also used in the laboratory measurements. The commercial instruments and I dosimeters used in tnis study typically resoonded low to the beta l fields measured, especially where maximum beta energies were less than l approximately 500 key. A single calibration factor is usually not l adequate for either peta dosimeters or instruments. There is a need for more refinement In beta measurement devices and training for tne users of such devices. l l NUREG/CR-3569: SPECIAL AND 00SIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITcS. 60BERSON,P.L.; FOX,R.A.; HOLBROOK,K.L.; et al. Battelle Memorial Institute, Pacific Northwest Laboratories. August 1984 182p0 6408290171. PNL=a915. 26299:119 Spectral and dosimetric measurements of photon fleids were l
i~ 1 s x e performe? at sevon comasrcio) nuclear recctor cites. Revisions to 10 CFH 20 thuta specify exposure-to-dose conversion factors (Cx) muen greater than unity for photons between 40 kev and 200 kev could impact personnel. monitoring practices. Monitoring at effective depths of 1 cm of tissue and shallower could underestimate doses received from high-energy photon fields (>3 HeV), No locations with large C(x) factors (approximately 1.5 rad /R) were found. The most s19nificant production of low-energy photons was found to be due to photon scettering. The scatter continuum has an ] offec.tive.Cx factor of approximately 1.2 rad /R.,One location was found>with a nearly pure scatter spectrum. Othe.5 locations contained sign'ificant contributions from medium-energy photo,s due primarily to radioactive decay of cobalt and cesium isotopess ~ Monitoring requirements at 0.00/=cm and 1.0-cm cepths in tissue were found to be adequate for estimating dose c.eceived in radiation fields containing l I high-energy photons, tnnanced surfaced doses. attributed to nign=erergy knock-on electrons were measured in all locations monitored.. Personnel monitoring techniques may provide inaccurate results in high* energy fields. NUREG/CR-3589 V01: REACTUR SAFETY RESEARCH QUARTERLY REPORTgJanuary-Harch 1983 Sandia Laboratories. July 1984, 179pp, 16409180335. SAND 83-2425. 26568 001. Sar.aia National Laboratories is conducting phenomenological research related to the safety of commercial nuclear power reactors. The overcll objective of tnis work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2)underatanding risk-significant accident sequences,-(3) developing and veritying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the united States without appropriate consideration being given to their ef f ects on heal th saf ety.,This report describes progress in a number of activities dealing with current safety issues relevant to :both light watte and breeder reactors. The work includes'a bread range of experiment to simulate accidental conditions to provide the data base required to understand important accident sequences and to serve as a basis for development and verification of the complex. computer simulation models and codes used in accident analysis and. licensing reviews. Current major emphasis is focused on providing information to NRC relevant to (1) its deliDerations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor. NUREG/CR-3589 V02: REACTOR SAFETY RESEARCH QUARTERLY REPORT. April-June 1983.
- Sandia Laboratories.
July 1984 166pp. 8409180443. SAN 083-2425. 26588 180. See NOREG/CR-35u9,V01 abstract. NUNEG/CR-3590: EVALUATION OF ISOTOPE DILUTION MASS SPECTROMETRY FOR BI0 ASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. DYER,F.F.; MAY,M.P.; WALKER,R.L.; et al. Oak Ridge National Laboratory. August 1984 79pp. 8408300272. ORNL/TM-9006 26330:224 A study was made to evaluate the sensitivity, precision and and practicality of isotope dilution mass spectrometry
- accuracy, (IDMS) for bioassay of uranium, plutonium, and thorium in human 28
urino. Tho' study showed that uranium at a concentration of 0.06 mg/L -(0.04 pC1/L natural uranium), plutonium at 3 pg/L (0.2 p Ci/L Pu-239, and thorium at 0.1 mg/L (0.01 pC1/L Th-232) could be measured eith an ' uncertainty (RSD) of ten percent using 10 m1 samples. The. lower limitsaDf detection for uranium and thorium were set.by background contamination, whereas tne detection limit for plutonium was determined by. chemical yield and intrinsic instrumental sensitivity , factors._ Precision and accuracy is excellent (~1-31, HS0) at l concentration levels wnere background contamination is insignificant and instrumental sensitivity is adequate. Comparison of IDMS with othee' methods shows the tecnnique is more sensitive than conventional fluorometric methods but is similar in sensitivity to alpha-radioactivity measurement methods that utilize .large sample volumes (1 L). Costs for urine analysis by IDMS (560-5100 per_ sample) are estimated to be considerably higher than cost ~for fluorometric analysis and approximately the same as the cost for alpha-radioactivity methods. Other methods that have Deen useu or are currently under development are discusseo. NUREG/CR-3591 V01: PRECUNSURS'TO POTENTIAL SLVERE CORE DAMAGE ACCIDENTS - 1980-1981 A Status Heport. COTTRELL,W.B.; MINARICK,J.A.; AUSTIN,P.N.; et al.- Oak Ridge National LaDoratory. July 1984. 262pp. 8408290190 26312:001. Descriptions of fifty-eight operational events reported as Licensee Event Reports, which occurred at commercial light water j reactors during 1980-1981 and which-are considered to be precursors to i potential severe core uamage, are presented, along with associated event-trees, categorization, and subsequent analyses. This study is a continuation of t'he work presented in AUREG/CR-2497 whien some. hat similarly evaluated the 19o9-1979 events. The current study incorporates improvements which evolved,from an assessment of the _ comments on the earlier report and applies these in the assessment of the LERs which occurreo during 1980 and 1981. Tne report sequentially discusses (1).the general rationale for this study, (2) the program methods for LER review and documentation, (3) the calculation of function failure prooaoilities and initiating event' frequencies.oased upon precursor data, (4) the use of the conditional probability of subsequent severe core damage estimates to rank precursor events and estimate an average inaustrywide risk of severe core damage, and (5) the conduct of sensitivity analyses on these results. There was some apparent decrease in most initiating event frequencies and function failure probabilities in the 1980 and 1981 period, as compared to the earlier report. .Altnough it was not possible to conclude that tnese decreases were statistically significant, they did result in a reduction in_the industry average estimated severe core damage frequency ~for 1980-1981 as compared to-the 1969-1979 period. 4 NUMEG/CR-3591 V02: PHECURSuRS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status-Report. COTTHELL,W.B.; MINARICK,J.n.; AUSTIN,P.N.; et al. Oak Ridge National Laboratory. July 1984 266pp. 6408290198 26311:001. See NUREG/CR-3591,VU1 austract. NUREG/CR-3593 v01: SYSIEMS INTERACTIUN RESULTS FROM THE DIGRAPH HATRIX ANALYSIS OF A NUCLEAR P0nER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. SACKS,I.J.; ASnMURE,d.C.; ALESSO,H.P. Lawrence Livermore National Laboratory. July 1984, 184pp. 8409280090 UCRL-53467. 29
26749:181 The report describes the demonstration of the Digraph-Matrix Analysis on a Nuclear Power Plant's High Pressure Safety Injection System. The demonstration work was beyond tne scope of ooth the n thods and the criteria used by the NRC to license nuclear power i 1 plants. The analysis discovered components whose failure could jeopardize the High Pressure Injection System given the postulated accident. All these components had been previously considered both in the safety analysis and in the licensing review. The results demonstrate the capaoility of Digraph-Matrix Analysis to modet en accident sequence (including front =11ne' systems, support systems, and operator actions) as a continuously integrated model to discover functional systems interactions. Also, the method is scrutable ano can be used on a complex system which contains botn a large number of components and depenaent loops. Volume 1 is the main report and the description of the metnod. Volume 2 contains the digraphs, adj acency listings, and data base. NUREG/CR-3S93 V02: SYSIEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR PonER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2. SACAS,I.J,7 ASHMORE,B.C.; ALESSO,H.P. Lawrence Livermore National Laboratory. July 1984 16bpp. 8409280123. UCRL-53467. 26735:096 See NUREG/CR-3593,V01 abstract. NUREG/CR-3599 SOURCES OF UNCERTAINTY IN THE CALCULATIONS OF LOADS ON SUPPORTS OF PIPING SYSIEMS.
- Dak Ridge National Laboratory.
RODABAUGH,E.C. E.C. Rodabaugh Associates,'Inc. July 1984 73pp. I 8408130041, 26040:166 Loads on piping systems are obtained from'an, analysis of the piping system. The piping system analysis, involves uncertainties from various sources. These sources of uncertainties are discussed and ranges of uncertainties are illustrated by simple examples. The sources of uncertainties are summarized and assigned s Judgmental i ranking of the typical relative significance of the uncertainty. NUREG/CR-3610: NEUTRON 00SIMETRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of Subtask C 3he Neutron Spectrometer. BRACKEN 80SH,L.; REECE,W.D.; TANNEF,J.E. Battelle Memorial Institute, Pacific Northwest Laboratories, September 1984 120pp. 6410120008 PNL-4943. 269838001. In commercial nuclear power plants, personnel routinely enter containment for maintenance and inspections while the reactor is operating and can be exposed to intense neutron ficids. The low-energy neutron fields found in reactor. containment cause problems in proper interpretation of TLD-albedo dosimeters and survey j instrument readings. Described is a technique that can aid plant health pnysicists to improve the accuracy of personnel neutron dosimetry programs. .A (3)He neutron spectrometer can,be used to neoasure neutron energy spectra and deter:nine dose equivalent rates at work locations inside containment. Energy correction factors for .TLD-albedo dosimeters can be determined from the measured spectra if the dosimeter energy response is known, or from direct measurements with dosimeters placed on phantoms at locations where the dose equivalent rate has been measured. This report descr}bes how to assemble a spectr'ometer system using only commercially available components, how to use it for reactor energy spectrum measurements, 30
IBsth (3)Ho and cnd h3w to cnolymo-tho dato cnd interprot tho roculto. ~ multisphere-spectrometers were used to measure neutron energy spectra -and dose equivalent at three PWRs end.one BWR. In_ general, the (3)He spectrometer measures.nigher dose equivalent rates than the multisphere spect'rometer. In the energy range from 10 kev to 1 MeV, the dose equivalents measured by the (3)He spectrometer and multisphere spect'rometer agree within about 35% for the spectra measured. NUREG/CR-3617: NOBLE GAS,100lNE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT. BROWNS FERRY. WICHNER,R.P.; 4 WEBER,C.F.; WRIGHT,A.L.; et al. Oak Ridge National Laboratory. Septemoer 1984, 199pp. 8410120013. ORNL/TM-9028 26978:001. This report presents an analysis of the movement of noble gas, iodine, and cesium l fission products within the Mark-I containment bWR reactor system represented by Browns Ferry Unit 1 during a postuinted accident sequence initiated by a loss of decay heat removal capapility following a scram. This accident could be brought under control by various means, but the sequence with no operator action ultimately leads to failure followed by loss of water from the reactor vessel, i core degrecation due to overheating, and reactor vessel failure with attendant movement of core debris onto the drywell floor. Tne fission product transport' analysis is. based on the no-operator-action sequence and provides an-estimate of fission product inventories, as a function of time, within 14 control volumes outsiae the core, with the atmosphere considered as the final control volume in the transport sequence. We find small barrier - f or noble gas ejection to air, these gases being effectively purged from the drywell and reactor ouilding by steam and concrete degradation gases. In contrast, large degrees l of.holoup for iodine and cesium are projected due to the chemjcal l reactivity of these elements. Only about 2 x 10(-4%) of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. NUREG/CR-3618: OCA-P,A DETERMINISTIC AND PROBABILISTIC FRACTURE-MECH 4NICS CODE FOR APPLICATION TO PRESSURE VESSELS. CHEVERTON,R.D.; BALL,0.G.. Oak Ridge National Laborat'ory. July 1964 105pp. 8406060415. ORNL-5991. 25977:338 OCA-P is a probabilistic fracture-meenanics code that was prepared specifically for the purpose of evaluating the integrity of PWR pressure vessels when suojected to overcooling-accident loading conditions. The code nas two-dimensional and some three-dimensional-flaw capability; it is based on linear elastic fracture mechanics; and it can treat cladding as a discrete region. Both deterministic and propabilir. tic analyses can be performed, and for the former analysis it is possible to conduct a search for critical values of the fluence and the nil ductility reference temperature corresponding to incipient initiation of the initial flaw. The probabilistic portion of OCA-P is based on Monte Carlo techniques, and simulated parameters include fluence, flaw depth, fracture toughness, nil ductility reference temperature, and concentrations of copper, nickel and phosphorous. Plotting capabilities include tne construction of critical-crack-depth diagrams (determinstic analysis) and various histograms (probabilistic analysis). 31 m
~ NUREG/CR-36033. HETEROGENEOUS OXIDATIVE DEGRADATION IN IRRADIATED ' POLYMERS. CLOUGH,R.L.; GILLER,A.T.; QUINTANA,C.A. Sandia -Lacoratories.. July 1984 43pp. 8408080359. SAND 83-2493. 25978:153 When polymeric materials are irradiated in the presence of air, oxygen-diffusion effects can, depending upon dose rate, lead to oxidative degradation which occurs only near the edges. This report describes the use of several recently developed techniques which are of general use for studying heterogeneous degradation in commercial polymeric materials. The techniques discussed ares optical evaluation of cross-sectioned, polished samples; cross-sectional profiling of changes in relative hardness; and profi1Ing of density j changes. Oxidation penetration. depths are given for a number of major polymer types as a function of dose rate. A detailed example is given graphically illustrating tne effects of differing oxidative penetration depths on the radiation-degradation Dehavior of a Viton 0-ring material; this particular material becomes hard and brittle when irradiated at-hign dose rate, but soft ano stretchable when irradiated at low dose rates. I l~ NUREG/CR-3654: PhR FLECHT 5EASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14. HOCHREITER,L.E.; RUPPHECHT,S.O.; DEDERER,J.T.; et al. nestinghouse Electric Corp. September 1984. i
- 584pp, 8410100382.
EPRI NP-3497. 26899:001. A series of natural circulation tests were conducted at a FLECHT SEASET facility whien is scaled 1/307 by volume to a full size par. l The purpose of these tests was to identify hydraulic and heat transfer phenomena during natural circulation cooling modes. The resulting data, evaluation, and analysis are to be used for PWR codes and model assessments as well as to provide a comparison to similar experiments L in other scaled systems. Steady-state single-phase, two-phase, ano j reflux condensation moces of natural circulation cooling were established in the FLECHT SEASET systems effects facility and the flow 1 and neat transfer characteristics of the different cooling modes were ) j identified. This report presents the test data; data reduction, l analysis, ano evaluation; and resulting model development and analysis. The models which have been developea include a reflux tube condensation mooel as well as a single-and two-pnase model for the i overall system. NUREG/CR-3655: A METHOU FOR ANALYTICAL EVALUATION OF COMPUTER-BASED DECISION AIDS. ROUSE,W.B.; FREY,P.R.; et al. Search Technology, Inc. KISNER,R.A. Oak Ridge National Laboratory. July 1984 202pp. 8409110215 ORNL/TM-9068 26445 001. This report presents a proposed methodology that involves a two-stage process of' classification and analytical evaluation of i decision aids for nuclear power plant operators. The classification scheme relates any particular aid to one or more general decision-making tasks. Lvaluation proceeds using a normative top-down design process based on the classification scheme and involves determining how various design issues associated with this pr' cess o were resolved b:t the designer. The result is an assessment of the "understandecility" of the aid as well as the identification of training and display requirements necessary to ensure understandability. The methodology is Illustrated by applying it to the evaluation of an aid designed to support operators in recovery of critical safety functions at a pressurized-water reactor. 32 L
Tuo cppondicoo cro includod. Appendix A contoins information collected from manufacturers, developers, and users of operational aid i systems. Appendix B is a review of NRC documents and guidelines that might apply to operational sids. NUREG/CR-3660 V02: PR06Ah!LITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF hESTINGHOUSE PWR PLANTS. Volume 2 Pipe Failure Induced By Crack Growth. WOD,H.H.; MENSING,R.W.; dENDA,8.J. Lawrence Livermore National Laboratory. August 1984 71pp. 8409200445 UCID=19968 26628:132 This report assesses the probability of reactor coolant loop (RCL) piping failures resulting from a crack growth mechanism. The Westinghouse pressurized water reactor (PWR) plants in the United -States east of-the Rocky Mountains are considered. After the introduction (S'ection 1), the assessment is presented in five parts _(Sections 2-6). Section 2 describes the characteristics of RCL piping in these~ Westinghouse PWR plants. Section 3 describes the metnodology used in the analysis. Sections 4 and 5 present the best-estimate ano uncertainty analyses, respectively. Our conclusions are presented in Section 6, along witn recommended items for consideration in future licensing regulations. NUREG/CR-3662: FUEL-DISRUPTION EXPERIMENTS UNDER HIGH-RAMP-RATE HEATING CONDITIONS. nRIGHT,S.A.; WORLEDGE,D.H.; CANO,G.L.; et al. Sandia Labocatories. August 1984 8bpp. 8409280084 SAND 81-0413, 267643340 This topical report presents the preliminary results and analysis of the High Ramp Rate fuel-disruption experiment series. These experiments were performed in the Annular Core Research Reactor at Sandic National Laboratories to investigate the timing and mode of fuel disruption during tne prompt-burst phase of a loss-of-flow accident. High-speed cinematography was used to oaserve tne timing and mode of the fuel disruption in a stack of five fuel pellets. Of the four experiments discussed, one used fresh mixed-oxide fuel, and three used irradiated mixou-oxide. fuel. Analysis of the experiments indicates that in all cases, the observed disruption occurred well before fuel-vapor pressure was high enough to cause the disruption. The disruption appeared as a. rap 10 ~ spray-like expansion and occurred near the onset of fuel melting in the irradiated-fuel experiments and near the time of complete fuel melting in the fresh-fuel experiment. This early occurrence of fuel disruption is significant Decause it can potentially lower the work-energy release resulting from a prompt-ourst disassemoly accident. i NUREG/CR-3663 V02: PR08AulLITY OF PIPE FAILURE IN THE REACTON COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS.Vol 2: Pipe Failure induced by Crack Growth. LO,f.Y.; MENSING,R.W.; h00,H.H.; et al. Lawrence Livermore hational Laboratory. September 1984 94PP. 8410120004 UCRL-53500 26985:177 The U.S. Nuclear,Hegulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to conduct a study to determine if the procapility of occurrence of a double-ended j guillotine break (deb 8) in the primary coolant piping warrants the current design reautrements that safeguards against the effect of DEGB. Tnis report describes the results of an assessment of reactor coolant loop piping systems designed by Combustion Engineering, Inc. l 33
A probcbiliotic fracturo machenics approcch.was unod to. estimate the crack growth and.to assess the crack stability in the piping -throughout the lifetime of the plant. The results of the assessment indicate that the probability of occurrence of DEGd due to crack growth-and instability is extremely small, which supports the argument that the postulation of UEGB in design shoula be eliminated and replaced with more reasonsole criteria. NUREG/CR=3665: OPTIMIZATION OF PUBLIC AND, OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonEd PLANTS. Executive Summary. COHEN,J.J. Science Applications,'Inc. September 1984 7pp. 8410100781 SAIC=84/1317. 26903:248. -An area of growing concern in recent years has been the apparent increase in levels of collective radiation dose to workers at nuclear power plants in the USA. U.S. Nuclear Regulatory Commission (NRC) l decisions and rulings related to in-service inspection, retrofits, and plant upgrades have oeen primarily int.ndea to reduce the risk of puolic radiation exposure resulti.ig from either routine release of radioactivity or potential accident situations.
- However, implementation of the required control measures ana procedures can often result in increasea levels of occupational radiation exposure.
Recognizing ~the need to incorporate occupational dose into probabilistic risk assessments (PRA), value-impact, and cost-benefit . analyses, the NRC has sponsored this study with the objective of developing an appropriate methodology to factor potential worker exposures into safety assessments. This report on the study is presented in three volumes. The following are subtitles for Volumes 1-3 Volume 1, "A Review of Occupational Dose Assessment Considerations in Current Probabilistic Risk Assessments and Cost-Benefit Analyses, Volume 2, " Consicerations in Factoring Occupational Dose Into=Value= Impact and Cost =8enefit Analyses, " and Volume 3, " A Calculation Method." NUREG/CR=3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RA01ATIUN PROTECTION AT NUCLEAR POWER PLANTS.A Review Of Occupational Dose Assessment Considerations In Current Probabilistic Risk Assessment And Cost = Benefit Analyses. LOBNER,P. Science Applications, Inc. September 1984 5Spp. 64101006S6 SAI-83/1125 26900:232 This report reviews current value=fmpact analysis and ~ probabilistic risk assessment methods and discusses the manner and 2 degree-to which these methods consider occupational radiation exposure that may form a variety of in= plant activities, includingt (a) normal operation and maintenance, (b) repair, (c) retrofit, (d) minoe incidents and cleanup, (e) major accidents, and (f) decommissioning. Value= Impact analysis methods which include occupational exposure as an element of the va)ue-impact equation have been developed, however, no standard approach to such analysis has been adopted. Comparison of the results of_value-impact analysis must, therefore, be done with I caution because different value-laden assumptions made by the analyst can have strong effects on the outcome. Such assumptions include the monstray equivalent of a person = rem, and the relative valus of occupational and public exposure. Probabilistic methods have oeen used in value-impact evaluations to quantify incremental or averted occupational exposure from reactor accidents, however, occupational i exposure has not been addressed in probabilistic risk assessments (PRAs) of nuclear power plants. Consideration of occupational exposure in a PRA would greatly increase the complexity of the plant i model and the benefits from such an analysis are uncertain. In lieu 3 34 1 l
, of oxpcnding th3 ccepo of PRAD to cddroco occupational rick, the separate,<Iimited-scope probabi11stic evaluations. developed f'r o value-impact analysis should provide a more practical analytical capability to support the evaluation and optimization of occupational and public radiat' ion exposure. NUREG/CR-3665 V02: OPTIMAZATION OF PUBLIC AND OCCUPATIONAL RADIAIION PROTECTION AT NUCLEAR POMER PLANTS. Considerations In Factoring Occupational Dose Into Value-Impact And Cost-Benefit Analyses. . COHEN,J.J. Science Applications, Inc. September 1984 46pp. 8410100785 ~SAI-84/1010 V02.- 26903:200. Many NRC decisions intended.for the improvement of public health and safety invoIve concomitant increases in occupational radiation exposure. Previous study (Volume 1) indicates that occupational dose consequences generally have not been considered in cost-benefit and value-impact analyses supportirg decisions related to puolic safety. Such consideration, nowever, would be consistent witn ALARA guidance. This study derives a methodology for factoring occupational vs. puolic radiation exposure, stocnastic vs. non-stochastic effects, probabilistic risk considerations, uncertainty, and de minimus levels. NUREG/CR-3665 V03: OPTIM1ZATION OF PUBLIC AND OCCUPAIIONAL RADIATION PROTECTION AT NUCLEAR POWER PLANTS.A Calculation Method. HORTON,n.H. Science Applications, Inc. September 1984 87pp. 84101007SS. - SAI-84/3037 V03. 26904: 001. The methodology presented in this report formulates an approach for.the optimizat' ion of Denefits resulting from NRC decision making processes.. Recent increases in occupational exposures in nuclear power plants resulting from NRC regulatory practices have leo to the - questioning by NRC of the overall benefit of specific regulations. The'optimizztion methodology in this report provides a tool for the determination of the cost-oenefit of proposed NRC regulations. Detailed methods are presented for the modeling of plant safety systems undergoing inspection, testing, and/or repair. This methodology utilizes dynamic Markov modeling techniques with extensive additional model development associatec.with operator errors involved in the inspection, test, and repair activities of the plant. Closed form solutions to the Markov models are provided. The report appendix presents the Markov model solution process in detail sufficient for model verification. Other methods necessary for the optimization process are discusseo in lesser cetail. An application of the methodology' dealing with steam generator inspection frequency and steam generator tube rupture events is presented. The example determines the steam generator inspection intervals which minimize expected costs and total expected occupational and public cose. - NUREG/CR-3671: ASSESSMENT OF RADIATION EFFECTS RELATING TO REACTOR PRESSURE VESSEL CLAD 0!NG. CORWIN,W.R. Oak Ridge National Laooratory. July 1984 70pp. 6408080419 ORNL-6047, 25977:271. l .Because the weld overlay cladding on the interior of light water reactor pressure vessels was applied for corrosion resistance and not . for structure, little attention has been given to the potential of mechanical property degradation due to radiation exposure. In lignt of the concerns recently raised regarding overcooling transients in nuclear power reactors, it has been suggested that any such degradation coulo adversely affect the serviceability ana/or integrity of the vesse!. 35 d w-e,------,-n.-w s.-- .e--. ,,e-,.,-.,-w--m.e.m........ee-,e-~w-w.-. .evwf,.,----w-a-r .w--,y.--,---,ww,wre,y ww s, w - w ,3 y
A litoreturo curvoy accooses the curront knowledes regcrding tho effects of neutron radiation on the mechanical fracture properties of stainless steel weld overlay cladding under conditions relevant to light-water reactor operation. In particular, effects on the material's microstructure and tensile, fatigue, impact, and fracture properties are examined. Although information is lacking on the specific materials under tne exact irradiation conditions of interest, a wealth of information is available on irraciated stainless steel weldments in general, from which basic behavioral trends can be obtained. Some irradiation emorittlement apparently does occur it.
- sinless steel weldments at tne relatively low temperatures and fluence, typical of light-water reactors.
Tensile strength increases and j ductility decreases. Low-cycle fatigue behavior is degraded somewhat, but high= cycle fatigue and fatigue crack growth seem largely l l l unaffected, Effects of ferrite on fracture resistance are small in ooth irradiated and unieradiated materials. Notch impact and fracture toughness are both reduced by irradiation, and a dependence of toughness on testing rate, not seen in wrougnt material, is inoicated. NUREG/CR-3678: ESTIMATION METHODS FOR PROCESS HOLOUp 0F SPECIAL NUCLEAR MATERIALS. PILLAY,K.S.; PICARD,R.R.; MARSHALL,R.S. Los Alamos Scientific Laboratory. July 1984 121pp. 8408080409 LA-10038. 25980:028 Los Alamos National Laboratory studied the use of statistical estimation methods for materials holdup at highly enriched uranfom (MEU)-processing facilities. Use of historical holdup data from processing facilities and selected holdup measurements at two operating facilities confirm the need for high-quality data and reasonable control over process parameters in developing these models. Large-scale experiments were conducteo to demonstrate the value of the models from good-quality experimental data. Using these data, we developed statistical models to estimate residual inventories of uranium in large process equipment and facilities. Some important findings are the following Holdup in some equipment at HEU-processing facilities, such as air filters, ductwork, calciners, dissolvers, pumps, pipes, and pipetittings can be readily modelea. Holdup profiles of process equipment such as glove boxes, precipitators, and rotary drum filters can change with time, necessitating several measurements at the time of inventory. Reasonable estimation of hidden inventories of holdup to meet regulatory requirements can De accomplished through good measurements and statistical modeling. .NUREG/CR-3679: CALIBRATIuN AND QUALIFICATION OF THE LOS ALAMOS FAILURE MODEL (LAFM). BAARS,R.E. Los Alamos Scientific Laboratory. July 1984 64pp. 8408080357. LA-10041-MS. 25980:329 The analysis procedure is described in detail for use of the LAFM computer code to predict LMFBR fuel pin performance under transient overpower conditions; miso, 5 tests for catioration and 13 tests for qualification are analyzed., The times of cladding breach (molten fuel expulsion) were predicteu with an average relative error of 5 per cent. An enthalpy of 1112 kJ/kg correlated the peak fuel enthalpies at the time of failure with a standard deviation of 98 kJ/kg. We conclude with a discussion that many varied tests must be analyzed for adequate evaluation of a fuel pin performance code. 36
' NUREG/CR-3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT-WATER = REACTOR SAFETY RESEARCH PROGRAM Quarterly Progress Report, January-Maren 1963 SHACKen.J. Argonne National Laboratory. July 1984.. 169pp. d408100151. ANL-83-85 25998:313. This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1983 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Produce Release, Clad Properties 2 for Code Verification, and Long-Term Embrittlement of Cast Duplex Stainless Steels in LWH bystems. NUREG/CR-3689 V02: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT-WATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress Report, Apri1* June 1983. SHACK,W.J. Argonne National Laboratory. July 1984 141pp. 8408310083. ANL-83-85 26348:215. See NUREG/CR-3669,V01 abstract. NUREG/CR-3689 V03: MATtRIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT-WATER REACTOR SAFETY RESEARCH PROGWAM.Guarterly Progress Report, July-September 1983. REST,J. Argonne National Laboratory. July 1984 40pp. 8406310077 ANL-83-85 26347:277 See NUREG/CR-3609,V01 abstract. NUREG/CR-3690 RELAP5 ASSESSMENT SEMISCALE NATURAL CIRCULATION TESTS S-NC-3,S-NC-4,AND S-NC-8 n0NG,C.C.; KMETY,K.L. Sandia LaDoratories. July 1984 11Spp. 8400310089 SANO-0402. 26348:099 The RLLAPS/ MOD 1 code is being assessed against test data from a number of integral and separate effects test facilities. As part of this assessment matrix, we have analyzed a number of natural circulation tests performed at the Semiscale facility. Our results for the single-loop ano two-loop steady state basecase tests S-NC-2 and'S*NC-7 have previously been documented; this report gives the results of calculations for two single-loop degraded heat transfer tests, S-NC-3 and S-NC-4, and for the two-loop ultra-small break i transient test S-NC-d. For tests S-NC-3 and S-NC-4, our analyses snow that RELAP5/M001 describes correctly the qualitative influence of steam generator secondary side heat transfer degradation on uoth two-phase and reflux natural circulation. The agreement between calculated and measureo two-phase mass flow rates in test S=uC-3 is i better with a primary mass inventory of 85% (where the peak two phase mass flow rate is calculated to occur) instead of 92% (where the peak mass flow rate occurred in S-NC-2). Flow oscillations are calculated for both tests, and were seen during S=NC-3, but were not reported in the S-NC-4 experiment. Some of these predicted oscillations are real, but others are nonphysical and can be inhibited by reducing the time step being useo (indicating problems in the time step control algorithm). The results for test S-NC-8, an ultra-small (0.4%) cold leg break, also compare reasonably well with the outcome of that j experiment. The overall conclusions and tneir possiole relevance to future RdLAPS code app!ication and development are discussed. NUREG/CR-3692: POSSIBLt MODES OF STEAM GENERATOR OVEHFILL HESULTING FROM CONTROL SYSIEM MALFUNLTIONS AT UCONEE-1 NUCLEAR PLANT. CLARK,F.H.; CLAPP,N.L. Van Ridge National Laboratory, oRUAUWATER,R. Tennessee Tech. Univ., Cookeville, TN. July 1984 50pp. 37
8409180488 ORNL/TM-9061. 26589 213. A study has been made of control system failures which might lead to overfill of the steam generator in Babcock and Hilcox nuclear plants. The steam generator and its control system are descriued. Only one sequence.has oeen found in which a single fa'ilure would lead to overfill, and in that case the final stages of the overfill would proceed rather slowly, uecause of high level protective features all other failure sequences we have examined require at least two fattures to produce overfill oeyond the point of high level protection. Several sequences are described in which high level protection features can be placed in an undetec'ted failed state by a control j system failure; a suosequent aoditional failure, occurring prior to the detection and correction of the first failure, could then produce system overfill. Mechanical damage is identified which might be consequent upon steam generator overfill and water entry into the main steam line, Several ways of reducing the probability of steam generator overfill are suggested. No assessment has been made of 3he probability of occurrence of any of the sequence. NUREG/CR=3708: LNR SPENT FUEL DRY STORAGE UEHAVIOR AT 229 C. EINZIGER,R.E. Westingnouse Electric Corp. COOK,J.A. EG&G, Inc. July 1984 134pp. 8408240379 2o255 001. A whole rod test was conducted at 229 degrees centigrade to investigate the long-term stability of spent fuel rods under a variet/ of poss1Dle dry storage conditions. All combinations of BWR or PWR rods, inert or air atmospheres, and intact or defect d rods were tested. After 2235 nours, visual observations, diametral measurements and radiograhic smears were used to assess the degree of cladding deformation and particulate release. The same examinations plus metallography and x-ray analysis were conducted after 5962 hours. None of the intact rods, the rods tested in inert atmosphere, or the defected PhR rod ttsted in unlimited air showed any measurable change from the pretest condition. The upper cefect on the bWK rod tested in unlimited air had split open *0.5 in. efter 2235 hours and had "10% cladding deformation. The crack grew to ~2.5 in, after 7962 hours. X-ray analysis indicated that the U0(2) had oxidized to U(3)0(8). The difference in behavior of tne upper and lower defects is attributed to the air's accessibility to the fuel because of the deflect's position with respect to the pellet = pellet interface. The oxidized fuel appeared to form a powdery compact that remained for the most part in the split cladding. Only a fraction o' the fuel out of the cladding became airborne. Some crud spalled frSt the rods but appeared to have no airborne particulate in the 2-to 15-mean-respirable range. Tnis report discusses the details and meaning of the data from this test. 38
NUREG/CR-3711: SnR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST RESULTS. HWANG,W.S.; ALAMGIRJ SUTHERLAND,W.A.; et al. General Electric Co. Septemper 1984 300pp. 8410100085 EPRI NP-3602 i l 26901:001 A new full heignt BWR system simulator has been built under the Full Integral Simulation Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary,. discussions, highlights, and conclusions of the FAST Phase I Tests. Eight. matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCAs, as well as natural circulation and power transients. Results and government phenomena of each test have been evaluated and discussed in detail in this report. a Two of there tests tie back to tests in the earlier TLTA facility. Comparisons between.the FIST and TLTA tests have been made. The similarities and differences be. tween counterpart tests are identified. Effects of the facility scaling compromises on the test results are i den t'i f i ed. One of the FIST program oojectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made witn IRACdO2 are presented and compared with test data in this report. These predictions agree very well with the test results. TRAC's capability to correctly predict the systes responses during the transient is demonstrated. NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONMENTAL RADIATION STANDARDS FOR GEOLOGIC DISPOSAL UF_HIGH-LEVEL. RADIOACTIVE HASTES. KOCHER,0.C. Oak Ridge National Laboratory. July 1984 80pp. 8408080422. ORNL-6006 25977:192. This report discusses the different technical issues that must be considered in developing an environmental standard for geologlc disposal of high-level rautoactive wastes. These issues jnclude (1) defining acceptable risk, (2) specifying acceptable risk in the standard, (3) formulating the standard so that reasonable demonstrations of compliance can be obtained, (4) applying the standard to protection of individuals or tne population, (5) applying the stanaard to expected occurrences only or to unexpected processes as well, (6) determining a time limit for the standard, and (7) specifying condit' ions to be assumed for demonstrating compliarce. It is concluded that many issues are not resolvable on technical grounds alone, but that an effective standard will allow flexibility and tne exercice of subjective scientific judgements is reaching licensing decisions. NUREG/CR-3724: ULTIMATE STHENGTH ANALYSES OF THE WATTS BAR, MAINE i YANKEE,AND BELLEFONTE CONTAINMENTS. JUNG,J. Sandia Laboratories. 1 July 1984 82pp. 8409260031. SAND 84-0660. 26699:240. As part of Sandia National Laboratories' Severe Accident Sequence Analysis (SASA) Program, structural analyses of the Watts Bar, Maine Yankee, ano Bellefonte containment structures were performed with the I objective of obtaining realistic estimates of their ultimate static pressure capacities. The Watts Bar investigation included analyses of the containment shell, equipment hatch, anchorage systems, and personnel lock. The ultimate pressure capability is estimated to ce between 120 and 137 psig, corresponding to shell yielding and equipment hatch buckling, respectively. The Main Yankee investigation consisted of an analysis of the containment shell and estimated its failure pressure to be between 96 ano 118 psig. For the Bellefonte containment, analyses of the containment shell and equipment hatch 39
uoro parformed. Tho proscuro capacity of the Bollofonto contoinmont is estimated to be between 130 and-139 psig, corresponding to dome tendon yielding and cylinder wall tendon yielding, respectively. NUREG/CR-3734: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982'- Maren 1983. BERMAN,M. Sandia Laboratories. July 1984 276pp. d408100152 SAND 84-0688, 25996 001. This report describes the investigations and analyses conducted at Sandia National Laboratories, Albuquerque, in support of the Light Water Reactor Safety Research Program from October 1982 througn March 1983. The Molten Fuel / Concrete Interactions (MFCI) Study investigates the mechanism of concrete erosion'by molten core materials, the nature ) and rate of generation of evolved gases, and the effects of fission-product release. The Core Melt / Coolant Interactions (CMCI) Study investigates tne characteristics of explosive and nonexplosive interactions between molten core materials and concrete, and the 'nteractions. In the Hydrogen probabilities and consequences of such i Program, the HECTR code for modelling hydrogen deflagration is being developed, experiments (including those in the FITS facility) are being conducted, and the Grand Gulf Hydrogen Igniter System II is ceing reviewed. All activities are continuing. NUREG/CR-3735: ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES-=A LITERATUNE REVIEW. BOLSTAD,J.W.; GREGORY,W.S.; MARTIN,R.A.; et al. Los Alamos Scientific Laboratory. July 1984 45pp. 8408160146 LA-100079-MS. 26122:266. The reported investigation is part of a program that was established for deriving radiological source terms at a euclear facility's atmospheric boundaries unoer postulated accident conditions. The overall program consists of three parts: (1) accident delineation and survey, (2) internal source term characterization and refease, and (3) induced, flow and material transport. This report is an outline of pertinent induced-flow ano material transport literature. Our objectives are to develop analytical techniques ano cata that will permit prediction of accident-induced transport of airborne material to a plant's atmospheric boundaries. the Prediction of material transport requires investigation of ares of flow dynamics and reentrainment/ deposition. A review of material transport, fluid dynamics, and reentrainment/ deposition literature is discussed. In particular, those references dealing with model development are oiscussea with special emphasis on application to a facility's interconnected ventilation system. NUNEG/CR-3739: THE OPEHATON FEEDSACK WORKSHOP:A TECHNIQUE FOR USTAINING FEE 08ACK FROM UPERATIONS PERSONNEL. MCGUIRE,M.V.; WALSH,M.E.; BOEGEL,A.J. Battelle Human Affairs Research Centers. September 1984 246pp. 8409280100 PNL-5214 26750 001. This report presents the results of three workshops that were designed, conducted, and assessed for the Nuclear Regulatory Commission. The purposes of the workshops were to (1) examine tne effectiveness of worksnops and other teenniques as mechanisms-for obtaining feedback from_ utility personnel, including comparison of several different workshop procedures 7 and (2) obtain feedback for the NRC on topics of interest and concern. The workshops were held in NRC Regions I, II, and III between December 1981 and May 1982. A total of 60 utility personnel attended the workshops and offered comments and 40
Cugg30tiCnD concorning Otoffing, onQinGOring support in tho, control room, training tools, training programs, and licensing examinations. Workshop participants ano observers evaluated the workshops favorably. Further assessment of the workshop process and content i f suggested that the wormanops were effective in obtaining useful i feedback 1for the NRC. NUREG/CR-3742: BUCKLING OF STEEL CONTAINMENT SHELLS UNDER TIME-DEPENDENT _ LOADING. dABC0CK,C.D.; BAKER,n.E.; FLY,J.; et al. Los Alamos Scientific Leooratory. July 1984 38pp. 8408100153. LA-10087-MS.- 25997:268. The problem of dynamic effects for steel containment shs11s subjected to time-dependent loadings that produce large compressive mesurene stresses in tne shell well is considered. Loadings on typical containment structures are reviewed, along with a description of the complete dynamic-buckling problem. Simplifications and the assumptions-that are currently used are critically examined and 3 reviewed with respect to buckling analysis. Based on these reviews, } three program cbjectives are defined and the tasks that can accomplish these objectives within a 2-year effort at level funding are outlined in detail. NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCI'0WER 1983 - MARCH 1984 PUGH,C.E. Oak Ridne National Laboratory. July 1984 200pp. 8408080355. ORNL/TM-9154/V1. 25979:146 The Heavy =Section Steel Technology (HSST) Program is an engineering research activity conducted by the Oak Ridge National Laboratory for the Nuclear Regulatory Commission. The program I comprises studies related to all areas of the technolcuy of materials i faDricated into thien-section primary-coolant containment systems of light-water-cocled nuclear power reactors. The investigation focuses on the behavior and structural integrity of steel pressure vessels containing cracklike flaws. Current work is organized into ten tasks (1) program management, (2) fracture-methodology and analysis, (3) material characterization and properties, (4) environmentally assisted crack growen studies, (5) crack arrest technology, (6) i irradiation effects studies, (7) cladding evaluations, (8) intermediate vessel tests and analysis, (9) thermal-shock technology, i and (10) pressurized tnermal-snock technology, I i NUREG/CR-3750: JOB ANALYb!3 OF NUCLEAR PUWER REACTOR HEALTH PHYSICS I TECHtJICIANd. DAVIS,L.T.; MAZ0UR,T.J.; CLARn,P.V.; et al. Analysis & Tecnnology, Inc. August 1984 200pp. 8410120030 Btvl=NUREG-51769. l 269803001. ( This report describes a project, an industry wide Job Analysis of ? Nuclear Power Reactor Health Physics Technicians (hPTs), sponsored by the Nuclear Regulatory Commission and conducted by Brookhaven National i Laboratory and Analysis
- Technology, Inc., to provide the industry l
with job performance data that can be used in systematically defining l training programs in ter,ms of required job functions, responsibilities, and performance standards. The job-analysis methodology is consistent with tnat used by the Institute of Nuclear l Power Operations (INPOJ in similar irdustry-wide projects and includes l administration of over 850 job task questionnaires to utility and i contractor Health Physics Technicians throughout the country. Date collected includes task performance (difficulty, importance, and j 41 -.~ -
frequency) and industry-wide demographics (Job levels, experience, education, and training). The results of this project discussed herein include model job descriptions for HPT positions, summaries of l HPT experience, education, and training, industry = wide listings with task = performance characteristics, and recommendations of selecteo I tasks as a basis for HPT training development. Finally, potential future applications of tne data base by utility and contractor organizations in training program development and evaluation and personnel qualifications are discussed. NUREG/CR-3751: EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FLOUD l PROTECTION COSTS FOR UHANIUM TAILINGS IMPOUNDMENTS. ECKER,R.M. Battelle Memorial Institute, Pacific Northwest Laboratories. July 1984 98pp. 8408130043. PNL-5068 26011:254 This report examines the costs of rock riprap flood protection for design flood events at two uranium tailings impoundments 'in western Colorado. Tne two sites are the Grand Junction impoundment located along the Coloraco River and the Slickrock impoundment located along the Dolores River. The sensitivity of rock type, embankment side slope, and various safety factors is evaluated for six design flood events at Grand Junction and one flood event at Slickrock. The safety factor method of riprap design is used for the cost comparison. NUREG/CR=3758: CROSSHOLE GLOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL WASTE REPOSITORIES. RAMIREZ,A.L.; LYTLE,R.J.; HARBEN,P. Lawrence Livermore National Laboratory. August 1984 76pp._ 8409110090. UCID=20060, 26443:290 An evaluation is_given of remote = probing geophysical techniques likely to be used to investigate the near vicinity of geologic repositories for nuclear waste. The sensors to be used would oe placed inside the borenoles, shafts and tunnels of the repository to provide nigh resolution information of the rock near the repository. The geophysical methods evaluated are known as active methods because they make use of artificial seismic, electric or electromagnetic fields to probe rock mass. Techniques involving through transmission measurements are emphasized. These techniques show merit for remote detection of geological heterogeneities such as fracture zones which influence the containment capacity of repository sites. The report discusses the results obtained with exploration methods used at a site near Oracle, Arizona. NUREG/CR=3761: RELAPS THERMAL = HYDRAULIC ANALYSES OF PREJSURIZED THERMAL SHOCK SEQUENCES FOR THE OCONEE-1 PRESSURIZED WATER REACTOR. FLETCHER,C.D.; BOLANDER,M.A.; STITT,B.D.; et al. EG&G,lInc. July 1984 165pp. 8408160248 EGG-2310 26125 001. Using the RELAPS computer code, engineers at the Idaho National Engineering Laboratory (INLL) performed thermal = hydraulic analyses of pressurized thermal shock sequences for the Oconee=1 pressurized water reactor. This report summarizes the results of previously reporteo calculations and presents the results of more recently completed calculations. Comparisons of two counterpart calculations performed, using the RELAPS code at the INEL and the TRAC code at Los Alamos National Laboratory (LANL), are included as appendices. The results of these thermal-hydraulic analyses will serve as boundary conditions for fracture-mechanics calculations which are to be performed at Oak Ridge National Laboratory. 42
i NUREG/CR=3763: REVIEO AND ASSESSMENT OF RADIONUCLIOE SORPTION INFORMATION FOR THE WASALT WASTE ISOLATION PROJECT SITE (1979 Through i May,1983). MELMERS,A.D. Oak Ridge National Laboratory. Sootember 1984 47pp. 8410180132. ORNL/TM-9157 27045:041. This document presents a scientific review and technological 4 assessment of the radionuclIde sorption information repor,ted by the Basalt Weste Isolation Project. (BWIP) for the candidate high-level waste repository in the Columbia River basalt flows in the Hanford Reservation. Quantified radionuclide sorption data are necessary for repository performance assessment to model expected radioactivity release rates in groundwater = intrusion-groundwater-migration scenarios. Three key SWIP reports were identified which contain most of the sorption information for a number of radionuclides with basalt, secondary minerals, or interbed materials. An extended review of these data is presented in this document. The technological i assessment identified seven potentially significant deficiencies in the radionuclide sorption information published by BWIP that could lead to questionable or nonconservative radioactivity release -calculations. These deficiencies are discussed in the document in detail. Specific additional information needs were also defined and reported. ] NUREG/CR=3765: MINETSIMULATIONOFAHELiCALCOILSODIUM/hATERSTEAM i GENERATOR, INCLUDING STRUCTURAL EFFECTS. VAN TUYLE,G.J. Brookhaven l National Laboratory, beptember 1984 27pp. 8410120010 BNL=hUREG=51766 26985 338 A test transient performed at a helical coil sodium =to= water ^ steam generatcr test facility was simulated using the MINET code. It was determined that correct calculation of the sodium outlet temperaturs requires representation of heat capacitance of the structure. NUREG/CR=3766: TESTING OF NULLEAR GRADE LUBRICANTS AND THEIR EFFECT UN A540 AND A193 87 BOLTING MATERIALS. CZAJK0hSKI,C.J. Brookhaven National Lsboratory. deptemoer 1984 83pp. 8409260637 n i BNL=NUREG=51767. 266993322 An investigation was performed on eleven commonly used, lubricants l by the nuclear power industry which included EDS analysis of the i lubricants, notched-tensile constant extension rate testing of bolting l materials with tt + lubricants, frictional testing of the lubricants l and weight loss testing of a bonded solid film lubricant. Tne report concludes that there is a significant amount of variance in the mechanical properties of common bolting materials, that MoS2 can i nydrolyze to form H23 at 100 degrees C and cause stress corrosion cracking (SCC) of bolting materials. One of the most significant findings of.this report is the observation that both A193 B7 and AS40 824 bolting materials are susceptible to transgranular stress j corrosion cracking in domineralized H2O at 280 degrees C in notched tensile tests. NUREG/CR=3776: TESTING OF SAFETY =RELATED NUCLEAR P0hER PLANT Euu1PMENT AT THE CLNTRAL RECEIVER TEST FACILITY, DANDINI,V.J.; ARAGON,J.J. Sandia Laboratories. July 1984 86pp. 8409110074 SAN 083-1960. 26437:198 The use of a solar energy facility for simulating the enormal environment (heat flux) produced as a result of hydrogen burns in a full-scale reactor containment building is described. Using a heat 43
. t. flux profilo davolcpod from colculatienc parforced by the HECTR ~ computer code, the Central Receiver Test Facility simulated the multiple burn thermal environment whien HECTR predicted woule result from the deliberate ignition of hydrogen generated by an 82D accident. 1 Functioning specimens of reactor monitoring and safety system equipment were exposed to this environment. Results of the equipment performance and temperature response are presented. .+. 1 NUNEG/CR-3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETHON= RASTER SYSTEM. BUCKALEn,n.H.; LOCKM000,G.J.; LUKER,S.M.; et al. Sandia Laboratories. July 1984 61pp. 8408080367 j SAN 084-0912. 259763090 The radiation capabilities of the PELLETRON Electron Beam s Accelerator have been expanded to include a controllable, variable i dimension, beam diffusion option. This rastered beam option has been studied in detail. deam cnaracteristics have been determined as a function of incident electron beam energy, current, and deflection 3 system parameters. The Deam diagnostics required to define any given diffuse beam pattern are accurate and predictable. Recently, utility 1 1 of this added PELLETROH capability was demonstrated by simulating the effects of complex nuclear reactor accident electron environments on electrical insulation materials similar to those used in nuclear power j plants. NUREG/CR-3786: A REVIEd UF RLGULATORY REQUIREMENTS GOVERNING CONTROL ROOM HABITABILITY SYSTLMS. JACOBUS,M.J. Sandia Laboratories. August 1984 63pp. 8410120011. SAND 84-0978. 26978 201 Tnis report reviews applicable guides, standards, and codes which i govern the design, manufacture, selection, installation, and surveillance practices for components and systems important to control room habitability. It covers the fundamental guidance contained in General Design Criteria, Regulatory Guides, and applicable sections of the Standard Review Plan, as well as numerous documents referenced by l this guioance. l Instances are cited wnere the present guidance is misleading, contradictory, or vague. In some cases, the problems in the guidance i result from inadequate technical bases; in other cases, the problems } result from several uocuments which are not completely consistent. To independently assess the suitability of the regulatory guide i which covers accidental chlorine releases, a computer program was developed to calculate enlorine concentrations in the control room l following chlorine release. Although problems with the assumptions used to oevelop the guide were found, the conservative nature of the chlorine calculations appears to adequately compensate for these i problems. i NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTEMS ~ IN RETAINING FISSION PRODUCTS. Background Information. MISHIMA,J.; i BLAHNIK,D.E.; HALVERSON,M.A,7 et al. Battelle Memorial Institute, Pacific Northwest Laboratories. August 1984 115pp. 8409170417. j PNL-5101. 26499:206. j' Tne Pacific Northwest Laboratory has compiled and reviewed base line data on the effectiveness of Engineered Safety Feature (ESF) systems in the retention of fission products and part%culate material resulting from a nuclear reactor accident. This work is part of an ~ NRC project to provide the best estimates of the consequences of severe reactor accidents. I 44 i l
Tho rocultino rcpset coceriboo tho ESF oyotono (centoinnont spray, secondary containment filter, containment recirculating filter, pressure suppression pool, ice condenser, and. main steam line-isolation valve leakage control systems). Also described are the anticipated atmospheres in which the ESFs must operate, the experimental studles.of ESF system effectiveness, and the models currently available for assessing the performance of the various EsF systems. The.information gaps identified as a result of.this review have resulted in recommendations for additional work in the areas oft
- 1) performance data and models of containment chiller /cooters; 2)
I continued development ano experimental verification of the ice condenser model; 3) continued development of the pressure suppression pool model; and 4) continued investigations of the behavior of filtration devices. NUREG/CR=3788 V01: STRUCTURAL INTEGRITY OF LIGHT WATER REACTOR PRESSURE GU9NDARY COMPONENTS.Four-Year Plan 1984-1988.
- Materials Er.gineering Associates, Inc.
September 1984 111pp. 8410180235. MEA-2047. 27045:088 Tnis document is the first in a series intended to provide an r up-to-date statement of the four-year plan for the program, " Structural Integrity of Light Water Reactor Pressure Boundary Components," which is being conducted by Materials Engineering Associates, Inc. (MEA). This program consists of engineering and research in fracture, fatigue, and radiation sensitivity of nuclear structural steels and weldments and addresses many of the key uncertainties in the mergin of safety in operating nuclear olents. i All tasks are integrated to focus on structural integrity of LnR pressure boundary components. The approach centers on an experimental + I characterization of nuclear grade steels and an assessment of fracture and fatigue behavior under conditions of a nuclear environment, so l investigation of irradiated materials is a key element of each task. t Experimental studies are supported by analytical models and investigation of the mechanisms responsible for the observed behavior. Data developea in the program will provide the basis for recommendations for the ASME Boiler and Pressure Vessel Code and ASTM test methods, and revisions to NRC Guices. i NUREG/CR-3792: CLOSEQUT OF IL BULLETIN 79-11 FAULTY OVERCURRENT TRIP i DEVICE IN CIRCUIT OREAKCHS FOR ENGINEERED SAFETY SYSTEMS. FOLEY,n.J.; i OEAN,R.S.; HENNICK,A. Parameter, Inc. August 1984 34pp. 8408310084 IEB-79-11. 26348:057 j IE Dulletin 79-11 was issued May 22, 1979 as a result of information received in April 1979 from hostinghouse and an NRC licensee relating to tne potential failure of a circuit breemer in an engieered safety system of a nuclear power plant. The defect of l concern was a smn11 hairline crack in the dashpot end cap of one of the three overcurrent trip devices of a Type 08-75 breaker. Tne Bulletin was also applicable to Type Ob-50 breakers, because tney use the same type of dasnpot end cap. The defective end cap had been installed in 1973 as a replacement, in compliance with IE Bulletin 73-1. Westinghouse Technical Bulletin NSD-TB-79-02 was issued April 4 17,'1979 to alert util1~ ties to the potential proolem, to provide i l background information, to recommend review of calibration test data and retesting of erratic breakers, to advise visual examination of end i caps for cracks and to call for replacement of cracked end caps. Evaluation of utility responses and NRC/IE inspection reports shows that 114 of the 129 current facilities do not use the affected 45 l
breakers in safety-related systems. Followup items for tho fivo facilities with open status are proposed. The Bulletin has been closed out for the remaining ten facilities with safety-related i Westinghouse DB-50 and Du=75 breakers having dashpots, on the oasis of acceptable utility responses and NRC/IE regional inspection reports. Erratic performance.of tnroe DB-50 breakers with worn seals at one facility is identified as a Remaining Area of Concern because the worn seals hao essentially the same effect on performance as cracked eno caps. Tne recommendation is made that preventive maintert-e programs i of Ifconsees be reviewed to make sure that breakers are k6pt clean to avoid plugging dashpot orifices. The Bulletin has served its purpose by resulting in identification of the potential problem at a limited number (15) of facilities and of the need for corrective actions at only five facilities. l NUREG/CR-3795: CLOSEQUT OF It BULLETIN 82-04 DEFICIENCIES IN PRIMARY CONTAINMENT ELECTRICAL PENETRATION ASSEMBLIES. FOLEY,W.J.8 HENNICK,A. Parameter, Inc. July 1984 53pp. 8408090272. IEB-82-04 25983:001 l IE Information Notice 82-40 was issued September 22, 1982 as an l early notification of a potentially significant problem pertaining to electrical penetration assemblies (EPAs) supplied by the Bunker Remo Corporation (BRC) of Cnetsworth, California..All oeficiencies described in the Notice were identified as existing in BHC EPAs with a hard epoxy module design. Utility personnel were asked to review the Notice and take appropriate actions, but were not required to respond or take any specific action. After further study, NRC concluded that there were potential generic safety implications at a limited number of plants. Accordingly, IE Bulletin 82-04 was issueo December 3, 1982 to require responses and specific actions by all licensees and holders of construction permits. Evaluation of utility responses, deficiency reports and NRC/IE inspection reports nas resulted in Bulletin closeout for 124 of 129 current facilities. Deficiencies describeo in the Bulletin were identified at all facilities, of which two are operating and nine.under construction. Followup of corrective actions and verification of inspection procedures are proposed in Appendix C for the five facilities with affecteo assemblies are summarized in Table B.6 Completion by NRC/IE of all the followup items identified in Appendix C is expecteo to resolve fully the specific problem of Bunker Ramo electrical penetrations that utilized a hard epoxy design. NUREG/CR-3796: EMERGENCY PREPAREDNESS SOURCE TERM DEVELOPMENT FOR THE i 0FFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES. SUTTER,S.L.; MISHIMA,J.; BALLINGER,M.Y.f et al. Battelle Memorial Institute, Pacific Northwest Laboratories. August 1984 352pp. 8409040447 PNL-5081. 26363:049 To establish requirements for emergency preparedness plans at i facilities licensed oy the Office of Nuclear Materials Safety and i Safeguards, the Nuclear Regulatory Cormission (NRCO needs to develop source terms (the amount of material made airborne) for accidents. They are used to estimate potential public doses from the events, which will be used to guide whether emergency. preparedness plans are needed. Pacific Northwest Laboratory is providing the NRC with source j terms by developing accident scenarios for fuel cycle and by-product operations. Several scenarios are developed for each operation, leading to the identification of the maximum release considered for emergency preparedness planning (MREPP) scenario. Fire was the MREPP at oxide fuel fabrication, UF(6) production, radiopharmaceutical 46
ccnufotturing, rcdiaphoroccy, cooled scurco acnufccturing, westo warehousing, and university research and development facilities. Tornadoes were MREPP events for_ uranium mills and plutonium contaminated facilities, and criticalities were significant at nonoxide fuel fabrication and nuclear research and development-i facilities. Techniques for adjusting the MREPP release to different facilities are also descrined. NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW ~ LEVEL WASTE INCINERATOR ASH. WESTSIK,J.H.; SUSCHBOM,R.L.; DIVINE,J.H.; et al. Battelle Memorial Institute, Pacific Northwest Lacoratories. August 1984 100pp. 8408310075. PNL-5153. 26347:319 Incinerator ash from the combustion of general trash and ion exchange resins were immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with tne Technical Position on Weste Form. Bitumen had a measured compressive strength of 120 psi and a leashability index of 13 as measured with the ANS 16.1 leach test srocedure. Coment demonstrated a compressive strength of 140C psi and a leachaoility index of 7 Both waste forms easily exceed the minimum compressive strength of 50 ps) and leachability index of 6-specified in the . Technical Position. Irradiation of 10(8) RAD and exposure to thirty +60 degrees to =30 degrees centigrade thermal cycles did not significantly impact these properties. Neither waste form supported 2 bacterial or fungal growth as measured with ASTM G21 and G22 procedures. Neither bitumen nor cement containing incinerator ash j cauted any corrosion or degradation of potential container materials including steel, polyethlyene and fiberglass. However, moist ash did 4 cause corrosion of the steel, i NUREG/CR=3804 V01: PHY3ICS OF REACTOR SAFETY. Quarterly Report January - March 1984. * 'Argonne National Laboratory. July 1984, 1p. 8408080374 ANL-84-35, 25979:356 This quarterly progress report summarizes work done during the l months of January = March 1964 in Argonne. National Laboratory's Applied Physics and Components Technology Divisions for the Division 'of l Reactor Safaty Research in the U. S. Nuclear Regulatory Commission. i The work in the Applied Physics Division includes reports on reactor i safety modeling and assessment by memoors of the Reactor Safety i Appraisals Section. Work on reactor core thermal =hydrauli;a is l performed in ANL's Components Technology Division, emphasizing j 3-dimensional code development for LMFBR accidents under natural convection conditions, An executive summary is provided including a l statement of the findings and recommendations of the report. l NUREG/CR 3606: ENVIRONMENTALLY ASSISTED CRACnING IN LIGHT riATER l REACTOP3: Annual Report,0ctober 1982 - September 1983. SMACK,W.J.; l KASSNAR,T.F.; KUPPERMAN,0 7 et al. Argonne National Laboratory. August 1984 143pp. 6410030340 ANL-84-36 26820:086 This progress report summarizes work on environmentally assisted j cracking in lightwater reactors during the twelve months from October 1982 - September 1983 The objective of this program is to develop an independent capability for prediction, detection, and control of intergranular stress corrosion cracking (IGSCC) in lightwater' reactor i (LnR) systems. The program is primarily directed at IGSCC proolems in 47
existing plants, but also includes the development of recommendations for plants under construction and future plants. The scope includes the followings (1) development of the m.eans to evaluate acoustic leak detection systems objectively and quantitatively; (2) evaluation of the-influence of metallurgical variables, stress, ano the environment on IGSCC susceptibility, including the influence of plant operations on these varisoles; and (3) examination of practical limits for these variables to effectively control IGSCC in LWR systems. The initial experimental work concentrates primarily on problems related to pipe cracking in BWR systems. However, ongoing researen work on otner environmentally assisted cracking problems involving pressure vessels, nozzles, and turbines will be monitored and assessed, and where unanswered technical questions are identified, experimental programs to obtain the necessary information will be developed to the extent that available resources permit. NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADWASTE.CONTAINING ORGANIC 10N-EXCHANGE MEDIA. SWYLER,K.J.; D0DGE,C.J.; DAYAL,R. Brookhaven National Laooratory. September 1984 82pp. 8410030353. BNL=NUREG-51774. 26820:001. Recently, regulatory consideration has been devoted to.the effects of self-irradiation on redweste containing organic ion exchange media. This consideration was prompted by decontamination operations at TMI-II, and by the development of technical positions in support of NRC regulation 10 CFR 60. This report addresses the effects of high radiation dose on the storage and disposal of radwaste ion-exchange media, and the validity of laboratory test procedures for predicting field performance. Our work shows that accelerated testing of ion-exchange media using high-dose-rate external gamma irradiation appears to be a valid procedure for assessing certain aspects of field behavior -i.e., radiolytic scission of the resin functional group, radiolytic gas generation of free liquids and resin agglomeration, provided both the test data and the field conditions refer to storage in a closeo environment. Certain resin decomposition processes appear to depend largely on resin moisture content, and may not be particularly sensitive to resin loading. One practical consequence of radiolytic acidity is to promote the corrosion of mild steel in irradiated resin. However, the corrosion process is very complex. Case-specific, long-term (i.e., low radiation dose) evaluations might be necessary if rigorous guidelines to protect radweste containers against corrosion are required. 1 NUREG/CR-3813 MINET VALIDATION STUDY USING STEAM GENERATOR TRANSIENT DATA. VAN TUYLE,G.J. Uronkhaven National Laboratory. September 1984 39pp. 8410120057. BNL-NUREG-51775 26986:100 Three steam generator transient test cases, that were simulated using the MINET computer code, are described, with computed results compared against experimental data. The MINET calculations closely agreed with the experiment for both the once-through and the U-tube steam generator test cases. The effort is part of an ongoing effort to validate the MINET computer code for thermal-hydraulic plant systems transient analysis, and strongly supports the valtaity of the MINET models. NUREG/CR-3815: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORR-PSF-PVS IRRADIATION EXPERIMENT. STALLMAN,F.W. Oak Ridge National Laboratory. August 1984 34pp. 8409170414 ORNL/TM-9207. 48
_ - _ ~ i 26498:164 A statistical analysis of Charpy test resdits of the two-year Pressure Vessel Simulation metallurgical irradiation experiment was performed. determination of transition temperature and upper shelf energy derived from computer fits compare well with eyeball fits. Uncertainties for all results can be obtained with computer fits. The results were compared with predictions in Regulatory Guide 1.99 and other irradiation damage models. i NUREG/CR-3818 REPORT OF RLSULTS OF NUCLEAR P0nER PLANT AGING WORKSHOPS CLARK,N.H.; BLRMY,D.L. Sandia Laboratories. August 1984. 64pp. 8408240320 SAN 0-64-0374. 26237:284 Two workshops were conducted to identify whether there is any evidence of component or structural aging problems in nuclear power plants, and, if so, what problems are of greatest importance. Fifteen representatives from national Ir.boratories, architect / engineers, nuclear steam supply system vendors, research firms, and a university i participated in the workshops. eased on completed questionnaires and group discussions which screened over 112 components believed to be susceptiole to excessive aging, pressure / temperature sensors, valve operators, and snubbers emerged by consensus as the most important i aging issues. Potential aging problems related to off= normal common ^ mode effects or aging problems which are just now developing were found to be outside the scope of the workshops, because little or no first hand experience is available for these off= normal or yet to 4 develop circumstances. Hecommendations are made for a systematic approach to rate components in terms of overall safety and for a cooperative effort between industry research groups a,nd regulatory i research groups to resolve known aging problems and to identify off= normal or yet to develop aging issues. i l NUREG/CR-3820 V01: THERMAL /HYORAULIC ANALYSIS RESEARCH i PROGRAM.Wuerterly Report, January-March 1984 THOMPSON,S.L. Sandia Laboratories. July 1964 62pp. 8408100147 SANU64-1025/1. 25996:280 The TRAC-PF1/M001 independent assessment program is part of a multi-faceted effort sponsored by the Nuclear Regulatory Commission (NRC) to determine tne spility of various systems codes to predict the detailed thermal / hydraulic response of LhRs during accident and j off= normal conditions. This program is a successor to the RELAPS/M001 independent assessment proaect underway at Sandia for the last two years. I The first quarter of FY84 marked the beginning of the TRAC =PF1/ MOD 1 independent assessment project at SNLA. The code was obtained from Los Alamos National Laboratory (LANL) in October, and brought up on both our Cyber-76 and Cray-1S computers. The assessment [ matrix was formalized, several TRAC nodalizations for the various facilities required were developed, and limited calculations were i begun, all described in the last quarterly. During this quarter, more [ nodalizations were developed and calculations begun, and tne first PF1/ MOD 1 assessment analysis was completed. NUREG/CR-3821: EVALUATION UF CRACK PLANE EQUILIBRIUM MODEL FOR PREDICTING PLASTIC FHACTURE. BUTLER,T.A.; SMITH,F.W. Los Alamos Scientific Laboratory. July 1984 23pp. 6409110107 LA-10129-MS. 26437:328 A simple model for predicting the initiation of crack nrowth ( 49 l
during plootic f rccturo 10 ovoluotod. Tho ccdal to based on rocuirino equilibrium between applied loads and an assumed stress distribution in the uncracked ligament near the crack. The fracture parameters required are the material's ultimate tensile strength and a process =Ione size at the crack tip that is determined from simple fracture tests. The Crack Plane Equilibrium model predicts crack-growth initiation for the crack geometries studied with sufficient accuracy to warrant extending it for investigating other geometries and for predicting stable crack growth and the onset of unstable crack growtn. NUPEG/CR-3822: SOLA-PT3 A Transient,Three-Dimensional Algorithm For Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometrics. DALY,B.J.; TORREY,M.D. Los Alamos Scientific Laboratory. July 1984 103pp. 8409110081. LA-10132-MS. 26445:203 The SOLA-PTS computer code has been developed to snelyze fluid-thermal mix 4ng in the cold legs and downcomer of pressurized water reactors in support of the pressurized thermal shock study. SOLA-PTS is a transient, tnree-demonsional code with the capability of resolving complex geometries using variable cell noding in the three coordinate directions. The computational procedure is second-order accurate and utilizes a state-of-the-art iteration method that allows rapid convergence to an accurate solution for the pressure field. Two different turbulence models are used in the code, a two-equation k-e model that is used in the cold leg pipe away from the Hpl inlet and a three-equation k-e-T' model for use near the HPI inlet and in the downcomer. The physical modeling and the numerical procedure used in SOLA-PTS are described in this report. Applications of the method to two creare 1/Sth-scale experiments are also presented. Two appendices are included. Appendix A provides a comparison of the two-and three= equation turDulence models, while Appendix B provides instructions for setting up and running a problem with SOLA-PTS. NUNEG/CR-3824: CONTING PHOGRAM GUIDE. HENRY,E.b.; GENTILLON,C.D.; STEVERSON,J.A. EG&G, Inc. September 1984 140pp. 8410030079. EGG-2315. 26821:001. CONTING is an interactive computer program for automated trends and pattern analysis of data. The data are License Events Reports, which are coded into a computer-readable, searchable Sequence Coding and Search System (SCSS) format developed by the United States Nuclear Regulatory Commission. In the SCSS, the data are oroken down into occurrences or steps, which are descrioed by categorical variaDies such as system, component, and cause. CONTING searches the steps to obtain counts for contingency tables (hence, its name). The rows and columns of these tables correspond to user specified conditions for the variables. The pattern of counts appearing in such a teole can provide insights concerning the operational experience at nuclear power plants. In addition, CONTING supports trend analysis since the counts can be grouped uy the associated event dates. A statistical package formats the contingency tables; this facilitates the use of log-linear statistical. program may include exposure times for use in normalizing the counts to obtain occurrence rates. CONTING has many features to aid the user in performing this analysis. CONTING was developed at the Idano National Engineering Laboratory (INELJ on tne CYBER 176, using FORlRAN 77. It operates on the SCSS data base located at the INEL. 50 1 1 a
dVREG/CR=3826 RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE SY 8RITTLE FRACTURE IN FERRITIC STEEL SHIPPING CONTAINERS GREATER THAN l FOUR INCHES THICK. SCHWANTZ,M.N. Lawrence Livermore National Laboratory. July 1984 131pp. 8408010155 UCRL-53538 258723067. i Various criteria for protecting against brittle fracture in spent-fuel shipping containers made from ferritic steel forgings greater than four inches thick are evaluated. A fracture initiation criterion based uoon yield stress levels and allowable flew sizes specified in Section XI of the ASME Code is recommended. This recommendation is based upon a value evaluation taking into account i its effect upon industry and the risk of brittle fracture. I I NUREG/CR-3830 V01: AERUSUL RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR QCTOBER 1983 - MARCH 1984 ADAMS,R.E.7 TOBIAS,M.L. Oak Ridge National Laboratory. July 1984. 79pp. i 8409170442. ORNL/TM-9217/v1. 26498:199 This report summariaes progress for the Aerosol Release and 3 Transport Program sponsored by the Nuclear Regulatory Commissions Office of Nuclear Regulatory Research, Division of Accident i Evaluation, for the period October 1983-March 1984 Topics discussed include (1) the experimental program in the Fuel Aerosol Simulant Test (FAST) facility, (2) NSPP experiments involving mixtures of aerosols of iron oxide and uranium in steam and dry atmospheres, (3) support i; work for the OEMONS (nest Germany) and Marviken (Sweden) projects, (4) analysis of core melt experiments involving boric oxide volatility, (5) initial operation of tne new 250-kW induction generator, (6) ) comparisons of NAUA results with experiments, and (7) tests and j improvements in the UVA8uBL-II code. i NUREG/CR-3832: UNCERTAINTIES IN LONG-TERM REPOSITORY PERFORMANCE OuE TO THE EFFECTS OF FUTURE GEULUGIC PROCESSES. SJOREEN,A.L.; KOCHER,0.C. ] Oak Ridge National Laboratory. August 1984. 41pp. 8409110084. ORNL-6049 26447:099 i This report discusses uncertainties in predicting the long-term j performance of geologic repositories for high-level waste that result i from the effects of future geologic processes. This type of uncertainty arises from uncertainties in determining current rates of geologic processes, predicting process rates over long time periods in j the future, and predicting the effects of future geologic processes on performance. This report emphasizes the qualitative and judgmental nature of predictions of future geologic processes and their effects i on repository performance. However, significant changes generally occur over time periods of 100,000 years or more. Thus, at sites chosen for their staoility, geologic processes should not have significant effects on repository performance over a period of 10,000 years. l NUREG/CR-3833: BEHAVION OF SUBCRITICAL AND SL0n-STABLE CRACK GROWTH FOLLOWING A POST-IRRADIATION THERMAL ANNEAL CYCLE, CULLEN,n.H.; i 1 HISER,A.L. Materials Engineering Associates, Inc. August 1984 l 41pp. 8409200295 MEA-4048 26609:061. j This report presents the experimental results of Phase I of a j Small Business Innovation Hesearch Program which investigated the l response of environmentally-assisted monotonic and cyclic crack growth following a simulated unneal of a reactor-pressure vessel weld. Unieradiated steels were used in this (initial) Phase I of the l program. Fatigue cracks were grown in several specimens of a I 51
i submerged arc weld deposit in pressurized, high-temperature reactor = grade water. The specimens were removed from the environment, j and annealed for one week at either 399 cegree C or 454 degree C. Fatigue crack growth in niwh-temperature water was resumed on several i anneated specimens and unannealed controls. No effect of the anneal ] was noted on the fatigue crack growth rates, which continued with about the same degree of environmental assistance as exhibited before s I the anneal. An elastic = plastic fracture specimen, tested in 93 degree C air at a very' slow loading rate, showed that neither annealing nor j the slow rate had a significant effect on the J-R curve i characteristics. However, conducting the tests at a slow loading rate j in 93 degree C PnR water resulted in a 25% to 30% decrease in JIc and ] i a small decrease in T ave. Examination of the oxides on the fatigue i fracture surfaces showed that magnetite (formed during the crack growth in pressurized, high-temperature water) was the predominant i oxide specie. i, I NUREG/CR-3834: ON THE THRESHOLD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRACKING OF SENSITIZED ALLOY 600 IN BORATED THIOSULFATE SOLUTION. SANDY,R.; KELLY,K. Brookhaven National Laboratory. July 1984 3bpp. 8408070014 8NL=NUREG=51785. 25954: 29u. l-The stress corrosion cracking (SCC) of sensitized Alloy 600 was i investigated in serateo solutions of sodium thiosulfate containing l 1.3% boric acid, using U= bends, constant load, and slow strain rate 1 tests. The aim of the investigation, among others, was to determine j the existence, if any, of a threshold level of sulfur, and Li to 3 i ratio governing the SCC. For U= bends, 5 ppm Li as LiOH in the presence of 7 ppm 3 as thiosulfate prevented occurrence of SCC. However, in slow strain rate tests, significant. SCC occurred at a S l 1evel of 30 ppo in the presence of 0.7 ppm of Lt. For a specimen held j under constant load, a propagating crack continued to grow until i fracture during controlled progressive dilution of the bulk solution, 1 leading to final Li concentration of 1.5 ppm and 3 concentration of j 9.6 ppb respectively. Tne implications of the results to initiation j and propagation of SCC in aerated thiosulfate solutions, and their 1 relevance to future operation of the steam generators at Tnroe Mile Island Unit 1 (TMI-1) are oiscussed. NUREG/CR=3835: SIMULATION OF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE DISCHETE VORTEX METHOD. BARR,P.K. Sandia i Laboratories. September 1984 19pp. 8410170076 SAN 084-8715. 27041:045 The interaction of a freely propagating premixed flame with regions of high vorticity in the flow is investigated using a computer model. These vorticity regions are formed due to the flame = generated 1 volume expansion that pushes gas past obstacles ahead of the flame. In the computer model the discrete vortex dynamics method is used to j simulate the time development of the vorticity regions downstream of each obstacle. The flame front is modeled as a wrinkled laminar flame interface that propagates normal to itself at the laminar burning velocity, separet'ing the two different density fluidst burned and unburned. Two different oostacle configurations are discussed in this paper. In the first case, a flame causes unburned gas to exhaust out of a planar duct, and when the flame reaches the duct exit it interacts with the vorticity which was formed at the exit. Two versions of this configuration are considered sharp and square edge exit. The second case involves a series of obstacles in a channel. 52
L Here, the repeated obstacles in the channel leads to acceleration of the flame as indicated by the dramatic increase in fuel consumption. I NUREG/CR-3840: COST ANALYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE ABILITY OF A NUCLEAR PLANT TO ENDURE STATION SLACK 0UT. CLARK,R.A.; THOMAS,W.R.; et al. Science & Engineering Associates, Inc. R10RDON,0.J. MATHTECH, Inc. July 1984 167pp. 8408080472. 25980316C. Cost estimates were required to serve as partial bases for decisions on four potential nuclear reactor facility modifications l being considered In the resolution of USI A-44, Station plackout. The modification constituting the four subtasks in this report are (1) j increasing battery capacity, (2) adding an AC independent charging 7 j pump for reactor coolant seal injection, (3) increasing condensate i storage tank capacity, and (4) increasing compressed air supply for instrument air. ] The cost estimates contained in this report include t5ose for the j followings (1) engineering and design, (2) equipment, meterials, and structures, (3) installation, and (4) present worth of the annual operation and maintenance over the remaining useful life of the reactor. 1 In addition to providing engineering requirements for the four modifications, the report evaluates the potential for synergistic solutions. It was found that some modifications to provide for reactor coolant seal injection would effectively satisfy the DC system l augmentation requirements, with the costs for solving both proolems being competitive with that of additional batteries alone. The report also identifies an innovative potential solution to the DC system capacity problem through the use of high energy density primary batteries which would oe far more cost effective than the add % tion of traditional lead acta catteries for mitigating extended station blackout effects. NUREG/CR-3842: STEAM GtNLRATOR GROUP PROJECT TASK 8 - SELECTIVE TusE i UNPLUGGING. WHEELER,K.H.) _00CTOR,P.G.; FETRON,L.K.; et al. dattelle i Memorial Institute, Pacific Northwest Laboratories. July 1984 200pp. 8410170217 PNL-4876 27029:034 The Steam Generator Group Project utilizes a retired f rom service pressurized water reactor steam generator as a test bed and source of specimens for research. Program oojectIves emphasize validation of the ability to nondestructively characterize the condition of tne steam generator tubing in service. During operation 748 of the 3388 tubes in the research generator were removed from service oy explosive plugging on both ends. Tubes were plugged due to oefect indications, inspectability limits caused by denting, and as a preventative measure. The plugged tunes containoa a substantial portion of the defects necessary for the research program. This report summarizes activities conducted during a campaign removal of 969 explosive tube plugs. The report provides detailed descriptions of the planning, training, supplies, equipment, and operations that led to tne successful completion of the, unplugging in 20 days of operation. Also presented is information on prootems encountered and observations that could aid future unplugning operations. NUNEG/CR-3843: STEAM GLNERATOR GROUP PH0 JECT TASK 10 - SECONDARY SIDE EXAMINATION. SCHMENK,E.8.; WHEELER,K.R. Battelle Memorial Institute, 1 53 l
Pccific NorthuoDt LcDoroterion. July 1984 69pp. 8410120034 PNL-5033. 26985 269 The Steam Generator Group Project utilizes a retired from service pressurized water reactor steam generator as a test bed and source of specimens for research. Program objectives emphasize. validation of the ability to nondestructively characterize the condition of steam generator tubing In service. Remaining integrity of tubing with service induced defects is studied through burst and leak rate tests. Other program objectives seek to characterize overall generator condition, including secondary side structure, and provide realistic samples for development of primary side decontamination, secondary side cleaning, and nondestructive examination technology. This report provioes information on secondary side characterization efforts. The methods and equipment used are discussed, along with comparisons of benefits offered by various j techniques. Details of secondary side steam generator conditions are then presented, emphasizing support plate and U-bend regions. NUREG/CR-3844: CHARACTER 1ZATION OF THE RADIOACTIVE WASTE PACKAGES OF THE MINNESOTA MINING AND MANUFACTURING COMPANY. KEMPF,C.R.7 SISKIND,8.7 BARLETTA,R.E.7 et al. Brookhaven National Laboratory. July 1984 93pp. 8408010151. UNL=NUNEG-51787 25872 001 An evaluation of the low-level waste packages generated by Minnesota Mining and Manufacturing Co. (3M) was made on the oasis of 10 CFR Part 61 criteria and on the Technical Position on Waste Form and neste Classification (TP). This evaluation was the result of a study initiated by the U. S. Nuclear Regulatory Commission (NRC), in which 3M participatea. 3M produces a variety of radioactive products and wastes. The dominant radioisotopes are Po-210 and Cs=137 The Po-210 packages are generally Class A ana meet the requirements in 10 CFR Part 61. The Cs=137 and Sr-90 packages fall into all three waste classifications (A, B, and C). These wastes are packaged by 3M in 30-gallon or 55-gallon carbon steel drums (Class A) or 30-ge11on llned drums (Class B and C). The Class B and greater leso-and concrete-lined packages nave been evaluated with respect to meeting the stability requirements for waste disposed of in a high integrity container. When so evaluated, eleven areas of concern were identified with respect to the regulations and recommendations in the TP. NUREG/CR-3845: PREDICTION OF NONLINEAR STRUCTURAL RESPONSE IN LMF8H ELEVATED-TEMPERATURE PIPING. FARRAR,C. Los Alamos Scientific Laboratory. July 1964 28pp. 8409170447 LA-10090-MS. 2o498:295. The development of structural analysis capabilities to investigate possible accident initiations caused by structural degradation o liquid metal fast breeder reactor (LMFBR) piping is summarized. The ABAWUS finite element code is used to perform a non-linear analysis of a bench mark problem proposed by the Pressure Vessel Research Committee. The problem is representative both in geometry and loading of an LMFBR elevated = temperature piping system, and published analytical results are available for comparison. Results show the system to be most sensitive to large, radial, thermal gradients that occur wnen the system experiences certain thermal transients. Repeated cycles of these transients will lead to thermal ratcheting, causing progressive deformation and strain accumulation in the system. Future work will verify the accuracy of the finite element model and quantify damage accumulated during the lifetime of an LMFBR elevated-temperature piping system. 54
NUREG/CR=3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL i INFORMATION DEVELOPED eY DOE HIGH= LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report for October-December 1983. KELMERS,A.D.; j KESSLER,J.H.; ARNOLD,W.D.; et al. Oak Ridge National Laboratory. August 1984 48pp. 840d300282. ORNL/TM-9191/V1 26331:224 t Oak Ridge National Laboratory (ORNL) is conducting.an experimental investigation of geochemical information for the Nuclear Regulatory Commission (NHC). During this quarter, the project evaluated both radionuclide solubility data and retardation parameters reported by the Basalt Weste Isolation Project (BWIP)f and the i methodologies used to develop those values. Under oxic conditions, neptunium had a sorption ratio of 1.7 L/kg for McCoy Canyon basalt and synthetic groundwater GR-2, which is lower than the " conservative best estimate" value recommended by SWIP. Under anoxic conditions, the basalt showed.little or no soility to remove technetium (VII) from i i GR=2 by sorption or precipitation. Several.important concerns may make it impossible to assert that the addition of hydrazine to groundwater is modeling the repository redox condition. These are (1) its reaction with any reducible solute is undefined, (2) its dissociation to release hydroxide ions probably dominates the groundwater pH, (3) it could react with bicarbonate to form the carbamate ions (4) it is corrosive to polycarbonate or polypropylene test tubes, (5) it may alter or disaggregate clay mineral structure, and (6) uncertainty exists as to the solid phase or solution species formed by reaction with pertechnetate ion. Thus, BWIP data obtained in the presence of hydrazine may be nonconservative for use in assessment studies. I NUREG/CR=3852: INSIGHT INTO PRA METHODOLOGIES, GALLAGHER,0 Science Applications, Inc. Auuust 1984, 121pp. 8409200290 26607:001 This report describes the results of a survey of sie r l probabilistic risk assessments to determine the impact of different aspects of the methodology on dominant sequence ordering and core-melt 4 I likelihood. The results indicate that effort should be given to human error analysis, system dependency analysis, and modeling of AC power systems. l } huREG/CR=3856: AN ULTRASONIC LEVEL AND TEMPERATURE SENSOR FOR POWER REACTOR APPLICATIONS. URESS,d.8.; MILLER,G.N. Oak Ridge National l Laboratory. August 1984 30pp. 8409180280. ORNL/TM-9236 I 26589 001, An ultrasonic waveguide employing torsional and extensional 7 acoustic waves has been developed for use as a level and temperature sensor in pressurized ano coiling water nuclear power reactors. Features of the device include continuous measurement of level, ( density, and temperature producing a realtime profile of these l parameters along a chosen path through the reactor vessel, f NUREG/CR=3867: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTERS AT [ U.S. COMMERCIAL NUCLEAN P0nER PLANTS, JANUARY 1,1976 TO DECEMoER 31,1982. BROWN,S.R.; TH0J0VSKY,M. EGSG, Inc. August 1984 131po. i 8409280077 EGG-2324 26762 206 Tnis report describes a computer-based data file developeu from i License Event Reports (LLRs) of i nverters in U.S. commercial nuclear i power plants for the period January 1, 1976 to December 31, 1982 In addition to the creation of the file, summaries of data contained in the file were made to obtain data for risk assessment and statistical j 56 f -,____nn--- ____n_nnnm-,,---n_w- -.m
s
- purpD00s, G Os0 cGnstcnt fGiluro POtOD ucro COticotGd fCr invcetoro found in selected systems.
Explanations, figures, and summary tables j ( ~ of the results are provided, j ^ ~ NUREG/CR-3868: CONTAINMENT Lu!LOING ATMOSPHERE RESPONSES DUE TO REACTOR + ~ GAS SURNING UNDER SEVERE ACCIDENT CONDITIONS. KROEGEH,P.G. j d Brookhsven National Leooratory. July 1984 42pp. 8410030303. BNL-NUREG-51793. 26811:315 The formation of-combustibje atmospheres during unrestricted core neatup accidents in High Tempvrature Gas-Coo. led Reactors is ceing investigated, consi(oring the e,ffects of only partially mixeo I is found that the previously used assumption of atmospheres. It complete mixingspresents the were severe limit in mast cases. In the few cases where_higpwr loads were obtained, these were still below the invocation of even mare remote failure scenarios. 'A oualitative dt.scussion epplying'the above results to comparable accident at Fort St. Vrain is included'. s NUREG/CR-3869: ANALYSIS OF THE IMPACT OF INSERVICE INSPECTION USING A PIPING RELIABILITY MUDEL., SIMONEN,F.A.; N00,H.H. da t te,11 e Memor'f al Institute, Pacific North, eat Laboratories. August 1984. 55pp. 6408220310.\\ pNL-5149. 26199:249 This repoet presents the results of a study of-the impact o.f inservice (ISI). programs on the rel)Qii,11ty of specific nuc1 war piping systems that have actually failed in dervice. Two mejor factors are considered in the ISI programs: one is the capacility of detecting flaws; 'tne other is the frequency of performing I S I'. A'probabilistic ,g ~ fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PnR.feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident. The results show that an effective inservice inspection requires a suitable combination of tiaw detection. capability and inspection schedule. 3n augmented inspection schedule'is required for piping t ith fast-growing flaws to ensure that the inspection is done.,before the flaws reach critical sizes.
- Also, the epimination of " poor" inspection teams through training and qualification testing can pr9 duce significant bsnefits to ISI effectiveness.
~ = NUREG/CR-3670: RADIATION DOSE ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Rept July 1982 -June 19A3. MEWHINNEY,J.A. Innalation Toxicology Reseorch. Institute. July 1984.* 38pp. 8408160125 LMF-109 26122:313. The objective of this project in to. conduct conffrmatory research on eerosol characteristics and the. resul ting radiation dose disteibution in a11mals 'after Inhalat' ion and to provide prediction of heelth consequences in humans from airborne radioactivity that might be c'eleased in normal operations or-under accident. conditions during production of nuclear fuel composed of mixed oxides of uranium and plutonium. Two research reports summarize the progress of current research. The first paper details results from the completeo radiation dose distribution studies in which dous, monkeys, and rats were exposed to either UU(d) + pug (2) treated ~at 750 degrees centigrade, O!,, Pu)0(2) treated at 175 degrses centigrade, or Pu0(2) '\\ treated at 850 degrees centigrade. This pacer focuses on analysis of the' data f rom' _the last animals sacrifited in the study and updates s m 4 4 s I 1 r w ._.-.c,- ,---.,_,,_..-,,.h, _.-.,---p
ccelice cnolyoco of ~ luno rotantien, ticcuo dictributien, cnd excretion. The second paper details preliminary analyses of the lung retention in Fischer-344 rats exposed to either (U, Pu)o(2) or to pug (2) at one of three levels of projected dose to lung for each aerosol. This paper presents the methods and the application of a rigorous statistical procedure. allowing detection of similarities and differences in the lung retention of rats at different dose levels and for different aerosols. NUREG/CR-3871: AN OVERVIEW OF THE UNIFIED TRANSPORT APPROACH. ERASLAN,A.H.; nITTEN,A.J. Oak Ridge National Laboratory. August 1984 123pp. 8409200399 ORNL-TM-9249 26607:128 The Unified Transport Approach (UTA) consists of a set of nine complementary models developed for assessing the environmental impacts associated with nuclear power plant discharges to receiving water bodies. This set' of models has the capability to simulate natural and plant-induced flow, temperature, salinity, sediment transport, radionuclide transport, and chemical species concentrations. nhile these UTA models were developeo for predicting impacts associated with the operation of nuclear power plants, they are quite general ano can be applied to a variety of situations. The UTA models have oeen used to simuiste the impacts associated with the operation of many industrial and energy production techno) ogles, as well as to simulate laboratory and naturally occurring conditions. In aII cases where data have been available for validation, the UTA model resu*ts have compared favorably. The purpose of this report is to provide an overview of the UTA as a whole, highlighting the important features and unique capabilities of this approach. NUREG/CR-3874: NEAR-GROUND TORNADO WIND FIELDS. MCDONALD,J.R. Texas Tech Univ., Lubbock, TX. July 1984 164pp. 8408220327 26198:197, This report is written as a general treatise on near-ground tornado wind fields. In Section II an engineering perspective on tornadoes is stateo. Section III describes the data available for the study of near-ground tornado wind fields. Section IV discusses tornado wind speeds anu Driefly describes a new method for making more rations) estimates of tornado wind speeds from damaged structures, j Section V describes the damage indicators that are present in the wake of a tornaoo event and discusses other factors that affect tne appearance of damage. A perspective on tornado-generated missiles is presented in Section VI. Conclusions and recommendations for furtner study are contained in the last section of the report. NUNEG/CR-3678: MODELING CONSIDERATIONS FOR THE PRIMANY SYSTEM OF THE EXPERIMENTAL BREEDER REAGTUR-II. MADNI,I.K. Brookhaven National Laboratory. September 1984 45pp. 8410120016 ONL-NUREG-51797. 26978:271 This report describes the additional heat transfer anc coolant dynamic models for components and processes, that are needed for simulation of the primary system of the Experimental Breeder i Reactor-II (EBR-II). This work forms part of the Super System Code (SSC) application efforts to provide preoictions of EBR-II overall plant behavior. NUREG/CR-3d84: EVALUaT10d uF NUCLEAR FACILITY DECOMMISSIONING PROJLCTS PROGRAM - THREL MILE ISLAND UNIT 2 POLAR CHANE RECOVERY. 00ERGE,0.H.; 57 l !l - - - ~ ~ ~
s MILLER,R.L. Unitod Nuc10cr Corp. August 1984 63pp. 8408290185 26309:267 This document summarizes information concerning restoration of the Three Mile Islano-unit 2 Polar Crane to.a fully operstional condition following the loss of coolant accident experienced on March 28, 1979. The data collected from activity reports, reactor containment entry records and other soueces were placed in a computerized information retrieval / manipulation system which permits extraction / manipulation of specific data which could be utilized in plannin9 for recovery activities shoul'd a similar accident occur in a nuclear generating plant. The information is presented in botn computer output form and a manually assembled summarization. This report contains only manpower requirements and radiation exposures actually incurred during recovery operations within the I reactor containment and uoes not include support activities or costs. j NUREG/CR-3888: ANALYSIS UF THE VENUS PhR ENGINEERING MOCKUP EXPEHIMENT -PHASE Is SOURCE DISTRIBUTION. MORAKINYO,P.O.; WILLIAMS,M.L.; KAM,6.K. Oak Ridge National Laboratory. August 1984 81pp. 8410030356 ORNL/TM-9238 26833:162. The neutron fission source distribution in the core of the VENUS PWR Mockup Experiment is computed and compared to experimental l, measurements. Of particular concorr. is the accuracy of the source calculation near the core =Deffle interface, which is the important region for contributing to RPV fluence. Results indicate that the calculated neutron source distribution l within'the VENUS core agrees with the experimentally measured values with an average error of less.than 3%. At the important core; baffle interface, the agreement is within 3% error, except at the baffle corner, where the error is about 6%. better accuracy in the calculations can be obtained by applying a detailed space dependent cross-section weignting procedure to the core-baffle interface region. Using this cross-section weighting, the maximum error introduced into t'he predicted RPV fluence due to source errors should be on the order of 5%. However, in power reactor analysis, additional complexities (such as the time-dependent core composition and the use of few group diffusion tneory) could effect this uncertainty value. NUREG/LR=3892: A RESEARCH PROGRAM FOR SEISMIC WUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Summary Report. KANA,D.D. Southwest Research Institute. August 1984 43pp. 8409070235. 26418:289 This document constitutes the. Summary for the indicated research contract on equipment seismic qualification methodology. Although the program was conducted by Southwest Research Institute, the results were periodically reviewed by a Peer Review Panel of. ten members from various segments of the nuclear industry, and by vari,ous memoers of the NRC staff. In addition, a continuing communication with the IEEE 344 (Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations) revision committee was maintained throughout the program to ensure that the results were disseminated to the industry. Thus, although the results are f r' m principally the findings of SWRI, acknowledgement of ' input o various other sources is recognized. The program has spanned a period of three years and resulted in seven technical summary reports, each of which covered in detail the findings of different tasks and subtasks, and have been combined into 58
five NUREG/CR volumes. This volume is to summarize the entire program from an overeII philosophical point of view. Volume 1 includes Task 1 Summary Reports parts 1, 2, and 3, which describe evaluations of various aspects of equipment qualifications methodology. Volumes 4, 3, and 4 include the summary reports for 4 Tasks 2, 3, and 4, which are concerned with correlation of methodologies, recommendations for improvements, and evaluation of fragility methodology. 4 NUREG/CR-3892 V01: A RESEANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey of Methods For Equipment And Components Evaluation of Methodology; Qualification And Methodology.... KANA,0.0.7 POLCH,E.2.3 POMERENING,0.J.7 et al. Southwest Research Institute. August 1964 393pp. 8409070233, 26413:001. The Research Program for Seismic Qualification of Nuclear Plant Electrical and Mechanical Lquipment has spanned period of three years and resulted in seven technical summary reports, each of which covered in detail the findings of different tasks and subtasks, and have been combined into five NUREG/CR volumes. Volume 1 comprises three parts. Part I reviews the methods i currently utilized for seismic qualification of nuclear plant equipment with emphasis on qualification by testing. In this review various anomalies that are associated with qualification are identified. Part II provides an in-cepth evaluation of the technical issues / anomalies previously identified. Part III provides an evaluation of the method applicable to line mounted items; e.p., valves. NUREG/CR-3892 V02: A RLSLANCH PROGHAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 2-Correlation Of Methodologies For Seismic Qualification Tests Of Nuclear Plant Equipment. KANA,D.D.; P0HENENING,0.J. Southwest Research Institute. August 1984 10Spp. d409070269 26409: 160. The Research Program for Seismic Qualification of Nuclear Plant Electrical and Mechanical Equipment has spanneo a period of three years and resulted in seven technical summary reports, each of which covered in detail the findings of different tasks and suDtasus, and have been combined into five NUREG/CR volumes. Volume 2 presents a general method for correlating the severity of one seismic qualification motion of given dynamic characteristics to another motion, possibly of different dynamic characteristics. The method provides a method of measuring relative damage severity of two different motions in terms of a relative damage severity ratio. NUNEG/CR-3692 V03: A RtSLANCH PROGRAM FOR SEISMIC WUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement Of Equipment Qualification Methodology And Criteria, nANA,0.D.; POMERENING,D.J. Southwest Research Institute. August 1984 74pp. 8409070272. 26409:047 l The Research Program for Seismic Qualification of Nuclear Plant Electrical and Mechanical Louipment has spanned a pariod of three l years ano resulted in seven technical summary reports, each of which l covered in detail the findings of different tasks and suotasks, and have been combined into tive NUREG/CR volumes. Volume 3 prosents recommendations for improvement of equipment j qualification methodology and criteria. These recommendations are 59 l i
grouped into categories standardization of procedures, demonstration of adequate methodology, a new methodology and procedural clarification / modification. The fifth category identifies issues where adequate information does not exist to allow a recommendation to ) be made. NUREG/CR-3892 V04: A RLSLANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 - The use Of Fragility In Design Of Nuclear Plant Equipment. KANA,0,0.; i POMERENING,0.J. Southwest Research Institute. August 1984 44PP. 6409070302, 26409:119 The Research Program for Seismic Gualification of Nuclear Plant Electrical and Mechanical Equipment has spanned a period of tnree years and resulted in seven technical summary reports, each of wnich have covered in detail the findings of different tasks and subtasks, and have been combined into five NUREG/CR volumes. Volume 4 presents a study of the use of fragility concepts in the design of nuclear olant equipment ano compares the results of state-of-the-art proof testing with fragility testing. NUREG/CR-3893: LABORATORY STUDIES DYNAMIC RESPONSE OF PROTOTYPICAL PIPING SYSTEMS. HOMARD,G.E.; WALTON,w.u.; JOHNSON,B.A. ANCO Engineers, Inc. August 1984 101pp. 8409070292 26409:258 This report presents details of the test methods, specimens and a preliminary assessment of results. Two test configurations will be used to achieve the project objectives. Both were three dimensional configurations; the second configuration had branch pipes. The piping systems sustained no apparent damage after being subjected to an earthquake approximately four times greater than tne SSE. Additionally, one of the piping systems resisted five OBE's, nine SSE's and nearly thirty shocks. NUREG/CR-3894: ULTRASONIC AND METALLURGICAL EXAMINATION OF A CRACKt0 TYPE 304 STAINLESS STEEL BnR PIPE nELDMENT. PARK,J.Y.; KUPPERMAN,0 Argonne National Laboratory. July 1984 22pp. 8408240322 ANL-84=1. 26255:169. An ultrasonic in= service inspection (ISI) indicated that a crack had developed in a 22-inch-diameter Type 304 stainless steel pipe manifold endcap weldment of the Hatch-2 Doiling water reactor. A section of the weldment was sent to Argonne National Laboratory (ANL) for further examination. The ANL effort included ultrasonic examinations, destructive crack-depth measurements, metallography, degree of sensitization (DUS) measurements, and chemical analyses of j material. The results showed that the extent of the cracking was j significantly less than indicated by the ISI. j l NUREG/CR-3595: INVESTIGATION OF COLD LEG WATER HAMMER IN A PnR DUE Tu THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. JACKUBEK,A.B.; GRIFFITH,P. Massachusetts Institute of Technology, Cambridge, MA. September 1984 60pp. 8410120001, 26986 141. Experimental studies using a prototypical flow model of a pressurized water reactor (PNR) demonstrate water hammer in the cold legs due to the admission of emergency core. cooling (ECC). Such water hammer can occur in an actual PWR during reflood provided there exists a stratified flow of steam and water in the cold legs. The hydraulic are postulated in this report. Calculations, based on a published 60
i criterion for water hammer initiation, show that the amount of ECC administered by the nigh pressure safety injection (HPSI) system, is 5 not great enough to produce liquid depths in the cold leg which can lead to slug formation and subsequent steam bubble collapse water hammer. However, a few water hammers can occur during ECC as the cold leg is being refilled. A simple analysis developed in this report calculates the water hammer pressures possible under these postulated flow conditions. Potentially dangerous water hammer pressures are predicted during reflood at high system operating pressures characteristic of a small break loss-of-coolant accident (SB LOCA). Similar ca,1culations done for the geometry of the experimental apparatus were compared to measurements taken during water hammer. NUREG/CR-3896: SIMULATION LXPERIMENTS COMPARING ALTERNATIVL PROCESS j FORMULATIONS USING A FACTORIAL DESIGN. MALUZNY,$.P.; SnARTZMAN,G.L. nashington, Univ. of, Seettle, WA. July 1984. 29pp. 8400060386. 2S937:282 This paper reviews methods for exploring the differences oetween alternative equations in complex ecosystem models. A factorial design is proposed as a metnod for exposing possible interactions between equation forms in their effect on model output as well as to clarify differences between the main candidate equations. A number of display methods arising from statistical analysis are used incluaing normal Q-Q plots, linear rank plots, and interaction diagrams. The metnods were illustrated using a complex ecosystem model of Lake Ontaric. We found the methods effective at illustrating major dif ferences oetween equations although several difficulties arose due to the complexity of the models and the diffuse nature of tne data supporting model validation. Qtestions of the method for standardization of equation forms so that the compared (quations are in some way analogous are important. These methods are probably most useful in cases where the data are of sufficient quality to indicate not only h'o w different equations effect model output but also which forms are to be preferred. NUREG/CR-3897: EVALUATION UF ECOSYSTEM SIMULATION MODELS AS TOOLS FOR l ASSESSMENT OF P0nER PLANT IMFACTS ON FISH POPULATIONS. Final Rept. SWARTZMAN,G.L. nashington, Univ, of, Seattle, WA. July 1984, lopp. I 8408010158 25867:294. l This two-volume report presents the procedures and analyses in developing an approach for structuring expert judgments to estimate human error probabilities. Volume I presents an overview of work performeo in developing the approach SLIM-MAUD (Success Limelinood Index Methodology, implemented through the use of an interactive computer program calleo MAUD--Multi-Attribute utility Decomposition). Volume II provides a more uetailed analysis of the technical issues underlying the approacn. t NUREG/CR-3899: UTILITY FINANCIAL STABILITY AND THE AVAILABILITY OF l l FUNDS F0x OECOMMISSIONING. SIEGEL,J.J Engineering & Economics l Research, Inc. Septemoer 1984 28pp. 8410030368 26817:289 l The NRC is currently developing rulemaking in the area of decommissioning nuclear facilities. A part of tnat rulemakina effort is assuring that funds will be available at the time of decommissioning of power reactors. Previous NRC reports have examined this issue by studying various funding methods. Tnis report provides l 61 l t
an update by analyzing tne relative level of assurance of funding methods, considering the present utility financial situation. In its analysis the report menes use of specific case situations. The report concludes that the various funoing methods studied in the earlier reports including the internal reserve method provide assurance of the -availability of funds for decommissioning. NUREG/CR-3900 V01: LONG-TEHM PERF0HMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACAAGING.First Quarterly Report, Year Three,Apri1= June 1984. SIAHL,0.7 MILLER,N.E. Battelle Memorial Institute, Columbus Laboratories. September 1984, 111pp. 8410120024, 26984346 Devitrification severity of glass waste forms is being studied in 1 terme of volume fraction of crystallization and crystal grain size. Glass-water contact during the heating and cooling periods of glass-leaching experiments is oeing evaluated for its effect on the overall results of the isothermal period. Modeling efforts included tne study of possible colloid formation and the change of water chemistry during glass dissolution. The electrochemical properties of container steels were found to be only slightly affected by the groundwater-species concentration, the presence of basalt rock, or the steels' cleanliness or microstructure, Hyorogen-embrittlement susceptibility may increase at expected repository temperatures. Results of the corrosion-modeling effort suggest that radiolysis may significantly affect general-corresion kinetics. The wateraradiolysis model was extended to account for more groundwater species and was used to predict the concentrations of two species in aqueous iron sulfates results were comparea with experimental data. A method was selected for perform 1ag uncertatnty analyses of weste-package models, Integral experiments nave been designed to address the combined effects of i repository conditions on tne waste package. i NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVEr4T REPORTS. Users Guice. GRLENE,N.M.; MAYS,G.T. Oak Ridge National Laboratory. JOHNSON,M.P. JBF Associates. August 1984. 160pp. 1 8409270117 ORNL/NSIC=2d3 26716:177 The Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data has developed, through the Nuclear Operations Analysis Center (NOAC) &t Oak Ridge National Laboratory (ORNL), a system to aio in the evaluation of the Licensee Event Reports (LERs) submitted by the nuclear power plant utilities. The primary objective of tne Sequence Cooing and Search System (SCSS) is to reduce the descriptive text of the incident reports to a cooed sequence that i s botn computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects, as well as personnel errors. The dataoase i contains all current LERs submitted by the nuclear power plant utilities after January 1, 1981, and is updated on a continual basis with new LERs, as they are cubmitted. The database is maintained by NOAC on the IBM-3033 computer system at ORNL. Following a description 1 of SCSS and structure of the database, a tutorial section is provided to acquaint the first-time user with logon procedures and the necessary commands to retriever display, and analyze LERs. Each command is subsequently discussed in detail in the fundamental and advanced command sections. 62 0 - - - =- -. - - -*ww.-e--. -.r--,-.-,-i.----+ ww-.-m-,--+-,v,-..--m-.----.==n =----~,--av- ...---..---ww.e---s-- -. ---m -a -e =*r- "--'F-7
~. _ _ NUREG/CR-3907: GT2R2 AN UPDATED VERSION OF GAPCON-THERMAL-2 CUNNINGHAM,M.E.; BEYER,C.E. Battelle Memorial Institute, Pacific i Northwest Laboratories. September 1984 70pp. 8410100121 PNL-5178 26902:223 l The GAPCON-THERMAL-2 code is used by the U.S. Nuclear Regulatory l Commission for audit calculations of nuclear fuel thermal performance computer codes. Since the code was originally written, errors and needed updates have oeen identified. Revision 2 of GAPCON-THERMAL-2 contains a number of coding corrections and updates, and now conforms with the American National Standards Institute FORTRAN-77 standard. The changes to the code are presentea in detail. Benchmarking calculations, concentrating on fuel temperatures and fission gas release, were performed to qualify the effect of model changes on the performance of GAPCON-THLRMAL-2, Revision 2 It was concluded that use of tne old fuel relocation model combined with the modified ANS 5.4 fission gas release model provides the best overall comparison with the thermal performance and fission gas release data.used for the penchmarking exercise. The use of the new fuel relocation model combined with the Beyer-Hann fission gas release model provided the cost comparisent of thermal behavior but significantly underpredicted ) fission gas release. NUREG/CR-3921: ORY SPENT FUEL STORAGE TEST PLAN FOR FINAL NONDESTRUCTIVE FUEL N00 LXAMINATION. OLSEN,C.S. EG&G. Inc. July 1984 14pp. 8409180283. EGG-2328. 26589:120 A test plan for tne third and final nondestructive examination of eight fuel rods used in a low-temperature, long-term, dry fuel storage 1 program is presented. This examination is part of a long-range project to evaluate the Dehavior of spent fuel during ary storage conditions. The objective of this project is to provide the Nuclear Regulatory Commission with the information to confirm or establish spent fuel ary storage licensing positions for long-term, low-temperature (<523 K), spent fuel rod behavior during dry storage 4 and for radioactive contamination arising from spallation of cladding crud. This examination consists of visual and photographic examinations, dimensional measurements, and gamma scanning of eight fuel rods. NUREG/CR-3929: LOSS =0F-BENEFITS ANALYSIS FOR NUCLEAR POWER PLANT l SHUTD0hNS, Methodology And Illustrative Case Study. PEEREN8004,J.P.; I BUEHRING,W.A.; GUZIEL,K.A. Argonne hetional Labor atory. Septemper 1984 73pp. 8409270130. ANL/AA=29 26719:287. A framework for loss-of-benefits analysis and a taxonomy for identifying and categori41ng the effects of nuclear power plant shutdowns or accidents are presented. The framework consists,of three fundamental steps: (1) cnaracterizing the shutdown; (2) identifying benefits cost as a result of tne shutdown; and (3) quantifying effects. A decision analysis approach to regulatory decision making is presented that expitcitly considers the loss of benefits. A case study of a hypothetical reactor shutdown illustrates one key loss of benefits net replacement energy costs (i.e., change in production costs). Sensitivity studies investigate the responsiveness of case study results to changes in nuclear capacity factor, load growth, fuel price escalation, and discount rate. The effects of multiple reactor shutdowns on production costs are also described. 63
NUREG/CR-3932 BENCHMARK OESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND-RELIABILITY ASSURANCE. l HALVERSON,S.L.; BEZELLA,W.A.; CHARAK,I.; et al. Argonne. National Lacoratory.- August 1964 115pp. 8410030347. ANL-84-34 26819:097 The objectives of this work are to evaluate and benchmark the current safety and reliability assurance-related practices employed Oy the NRC. This effort represents an initial phase of a program wnose overall purpose is to develop a reliability program (RP). A review of NRC regulations relevant to reliability assurance was made for a boiling water reactor using two representative safety systems; the reactor protection system, and the residual neat removal system. The primary sources of information were the standard Review Plan and Title 10 of the Code of Federal Regulations, especially Part 50 In aduition, relevant regulatory guides, NRC branch technical positions and industry consensus standard were identified and catalogued for the two reference safety systems over the plant's life cycle. The identified standards and criteria were then organized into a RP element matrix of current regulatory requirements organized oy life cycle phase, top level assurance function, and items directly auditable by tne NRC. A brief review of the licensing process was l also undertaken to indicate the effectiveness of NRC implementation of l a RP. The results of this work show=d that within the NRC regulations l a framework already exists in which to Integrate, not add, a l reliability assurance program. l NUREG/CR-3933: RISK RELATEU HELIABILITY REQUIREMENTS FOR BhR SAFETY -IMPORTANT SYSTEMS WITH EMPHASIS Ori THE RESIDUAL HEAT. REMOVAL SYSTEM. TZANOS,C.P.; BEZELLA,W.A. Argonne National Laboratory. August 1984 140pp. 8410030385 ANL-84 52. 26819:218 The objective of this study was to identify and evaluate the major safety risk parameters of. typical reactor safety systems for use in developing a reliability program. This effort was part of a larger i research project aiming to evaluate the feasibility and effectiveness of introducing elements of proven reliability programs from otner nigh technology industries into the nuclear industry. As a reference safety system, the Residual Heat Removal (RHR) system of a Bolling Water Reactor (BWR) was selected. A scoping evaluation was also made l for BWR reactor protection system (RPS). Plant information, existing PRA and other relevant analyses, as well as Licensee Event Reports were used as base material for this study. The results of tnis evaluation indicate thats (1) recovery of faults can have a very significant impact on the reliability requirements, (2) there exists an obvious need for an adequate reliability data base, (3) reliability analyses must ce supported by detailed analyses of the plant's response to acciuent sequences, and (4) the development of effective emergency operating instructions and proper operator training must be one of the major elements of 6 Reliability Program. i NUREG/CR-3939: WATER HAMMER, FLOW INDUCED VIBRATION At4D SAFETY /NELIEF VALVE LOADS. UFFER,R.A.; VALANDANI,P.; SEXTON,D. Quadrew Corp. j September 1984 86pp. 8410120003 EGG-2340, 26985 096. This report presents the results of an evaluation performed to determine current and recommended practices regarding the consideration of water hammer flow-induced vibration and safety = relief I valve loads in the design of nuclear power plant piping systems. Current practices were determined by a survey of industry experts. Recommended pract'scos were determined by evaluating factors such as 64
l l Iccd mcenitudo end froquancy content, synton cuocoptibility to loedo, frequency of load occurrence and safety effects of postulated piping damage. l This report was prepared for use by the NRC staff in developing positions regarding consideration of dynamic piping loads for use of the NRC's Piping Review Committee. NUREG/CR-3940 FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN ALKALINE LAKE ENVIRONMENTSe SIMPSON,H.J.; TRIER,H.M.) LI,Y.H.7 et al. Columbia Univ., New York, NY. August 1984 124pp. 8409260650, 26702:037. Measurements of tne radioisotope concentrations of a number of elements (Am, Pu, U, Pa, Th, Ace Ra, Pb, Cs, and Sr) in the water and 4 sediments of a group of alkeline (pH s 9-10), saline lakes demonstrate greatly enhanced soluble-pnase concentrations of elements with oxidation states of (III)-(VI) as the result of complexing by carbonate ion. Ratios of soluble radionuclide concentrations in Mono Lake to those in seawater (ICO3 (2-)) in Mono Lake a 200 times that of seawater) were Pu(a10), (238)U(s150), (231)Pa, (238)The (230)Th(s10(3), and (232)Th(s10(5). Effective distribution coefficients of these radionuclides in high C0(3)(2-) environments are several orders of magnitude lower (i.e., less particle reactive) than in most other natural waters. The importance of C0(3)(2-) ion on effective K(d) values was also strongly suggested by laboratory experiments in which most of the dissolved actinide elements became adsorbed to particles after a water sample normally at a pH of 10 was acidified, stripped of all C0(2) and then returned to pH 10,oy moding NH(4)0H. Furthermore, the effect complexation by organic ligands is of secondary importance in the presence of appreciable carb'nate ion o concentration. Neither pure phase soluDility calculations nor laboratory scale K(d) determinations accurately predicted the measured natural system concentrations. Therefore, measurements of the distribution of radionuclides in natural systems are essential for assessment of the likely fate of potential releases from high level waste repositories to groundwater. NUREG/CR-3951: INTH000CT10w TO BIBELOT:A BIBLIOGRAPHIC FINDING AHD RETRIEVAL SYSTEM.~ COCHHAH,M.I. Battelle Memoria) Institute, Pacific Northwest Laboratories. September 1984 46pp. 8410100776 PNL-5202, 26903:260 Tne BIBELOT System of COBOL and Datatrieve programs for bioliographic storage and retrieval is described. The storage scheme is also briefly descrioeo. The use of unique citation numbers and user defined keywords is illustrated by many retrieval examples. Finally, typical questions about the use of BIdELOT are answered. I NUREG/CR-3988: MARCH 2 (MELTOOAN ACCIDENT RESPONSE CHARACTERISIICS) CODE DESCRIPTION AND USENS MANUAL. WOOTEN,R.O.; CY6ULSKIS,P.; QUAYLE,S.F. Battelle Memorial Institute, Columbus Laboratories. September 1984 400pp. 8410170214. BMI-2115 27030:001. MARCH 2 describes the response of water-cooled reactor systems to severe accidents, particularly those leading to core meltdown. The code performs the calculations from the time of accident initiation through the stages of coolant blowdown and boileff, core heat up and meltdown, pressure vessel bottom nead melting and failure. and debris-water and debris-concrete interactions in the reactor cavity. 65 4 1 -..,...,,,m
i l Both tho primary systoa and the building cro medolod. Mets and onorgy additions to the containment building are evaluated and the l pressure-temperature response of the containment with or without i engineered safety fsatures is calculated. A maximum of eight containment sub-volumes may be modeled. Engineered safety features modeled include emergency core cooling systems, containment sprays, builoing coolers and fans, suppression pool and ice condenser. containments, and emoruency core cooling and spray heat exchangers. Effects of metal-water reactions, combustion of hydrogen and carbon monoxice, heat losses to containment structures, and redistribution of i the decay heat due to loss of volatile fission prooucts from tne core l are considered. MARCH 2 is intended to replace the earlier MARCH 1 l code. It is written in FORTRAN 77 to improve transportability. l NUREG/CR-4001: CONTEMPT 4/MODSIAN IMPROVEMENT TO CONTEMPT 4/ MOD 4 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAIhMENT ANALYSIS. LIN,C.C. Brookhaven National Laboratory. l Septemoer 1964 40pp. 6410180123 Bhl=NUREG-51824, 27045:269 CONTEMPT 4 is a digital computer program for multicompartnent l containment system analysis. Previous version of the CONTEMPT 4 code, MOD 4, consists of an implicit algorithm to computer Junction flow when l numerically induced flow oscillations are encountered. This document presents analytical model and UPDATE statements that are requireo to extend the capability of the MOD 4 implicit routine for ice containment analysis. A sample problem is analyzed both with.and without the use of the implicit routine to demonstrate the effectiveness and the need of an implicit algoritnm for such problems. NUREG/CR-4007: LOhER LIMIT OF DETECTION DEFINITION AND ELABORATION OF A PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL MEASUREMENTS. CURRIE,L.A. Commerce, Dept. of, National dureau of Standards. September 1964 153pp. 8410170308 27031 060 A manual is provideo to define and illustrate a proposed use of the Lower Limit of Detection (LLD) for Radiological Effluent and Environmental Measurements. The manual contains a review of information regarding LLU practices gained from site visitsi a review of the literature and a summary of basic principles underlying tne concept of detection in Nuclear and Analytical Chemistry; a detailed l presentation of the application of LLD principles to a range of l problem categories (simple counting to multinuclide spectroscopy), i including derivations, equations, and numerical examples; and a brief ( examination of related issues such as reference samples, numurical quality control, and instrumental limitations. An appendix contains a summary of notation and terminology, a bibliography, and worked-out l examples. I 1 NUREG/CR-4011: TME 21/55 DATA BASE USER'S MANUAL. SILVER,E.G. Oak 1 Ridge National Laboratory. September 1984 317pp. 8410120007. ORNL/NSIC-221, 269823001 The Nuclear Regulatory Commission's Office for the Analysis and Evaluation of Operational. Data has developed, through the Nucleer Operations Analysis Center (NOAC) at Oak Ridge Nationel, Laboratory (ORNL), a data base for storing and organizing information obtained from the reports on construction deficiencies (CDRs) submitted to NRC. under the requirements of 10 CFR 21 and 10 CFR 50.55(e) by holders of construction permits for nuclear _ facilities. The computerized data base stores coded and textual information about the reports issued and 66
tho ovcnto to which they rofor, including such doto.co dctos of ovonto and reports, affected systems and components, source of information, manufacturers and vendors of affected components and the like. There is also provision for direct access to the data base by NRC Headquarters and Field Office staff both for accessing the information in the data base, and for entry of specific data concerning assignments of NRC follow-up staff and resolution actjons tenen. The document includes a tutorial guide for novice users of the data base. A system of access control to assure the integrity of the NRC= input data was developed and is oescribed. l l l l 67 i
Contractor Report Number index This index lists, in alphabetical order, the contractor-issued report codes for the NRC contractor reports in this compilation. Each contractor code is cross-referenced to the NUREGICR for the report and to the 10-digit NRC Document Control System accession number. SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NdM8EH NUMBER NUM8EH 82 NUREG/CP-0051 BNL=NUREG-51824 NUREG/CR-4001 ANL-83-85 NVREG/CR-3689 V01 EGG-2037 NUREG/CR-0169 V17 ANL-83-85 NUREG/CR=3669 V03 EGG-2307 NUREG/CR=1740 R01 Aht=83-85 NUREG/CR-3669 V02 EGG-2310 NUREG/CR-3761 ANL-84-1 NUREG/CN=3894 EGG =2315 NUREG/CR-3824 ANL-84=34 NUREG/CH-3932 EGG-2324 NUREG/CR-3867 ANL-84-35 NUREG/CR-3804 V01 EGG-2328 NUREG/CR-3921 ANL-84-36 NUREG/CH-3806 EGG =2340 NUREG/CH-3939 ANL=84=52 NuREG/CH-3933 ENIC0=1141 NUREG/CR-3513 ANL/AA=29 NUREG/CN-3929 EPHI NP-2314 NUREG/CR-2576 8HARC=400/84/01 NUREG/CR-3739 EPRI NP-3497 NUREG/CR=36S4 8MI-2115 NUREG/CH-3988 EPRI NP-3602 NUREG/CR-3711 BNL-NUREG-51454 NUREG/CR-2331 V03 N3 GEAP-22054 NUREG/CR-2576 8NL-huREG-51454 NUREG/Cd-2331 V03 N4 GEAP=30496 NUREG/CR 3711 8NL-huREG-51494 NUREG/CR-2482 V05 HEDL-TME 84=1 NUREG/CR-3318 8NL-NUREG-51650 NUREG/CR-3169 IE8-79-11 NUREG/CR=3792 l 8NL-NUREG=51708 NUREG/CR-34o9 V01 IEB-80-08 NUREG/CR-3053 8NL-huREG-51711 NUREG/CN=3493 IES-82-04 NUREG/CR-3795 8NL-hUREG=51716 NUREG/CN-3518 V01 LA-100079-MS NUREG/CH-3735 BNL=huREG-51718 NUREG/Cd=3520 V02 LA-10038 NUREG/CR-3678 l chL=NUREG=51718 NUREG/CR-35do V01 LA-10041-MS NUREG/CR-3679 l 8NL-NUREG=51766 NUREG/CR-3705 LA-10087-MS NUREG/CR-3742 l 8NL-NUREG-51767 NUREG/CR-3706 LA=10090-MS NUREG/CR-3845 8hl=huREG-51769 NUREG/CR-37bo LA-10127-C NUREG/CP-0053 1 BNL-huREG-51774 huREG/CR-3812 LA-10129-MS NUREG/CR-3821 1 8NL-huREG-51775 NUREG/CH-3813 LA-10132=MS NUREG/CR-3822 8NL-NUREG-51785 NUREG/CN-3834 LMF-109 NUREG/CR-3870 l BNL-NUREG-51787 NUREG/CH-3844 MEA-2047 NUREG/CH-3788 V01 8NL-hUREG-51793 NUREG/CH-3868 MEA-2048 NUREG/CR=3833 ShL*NUREG-51797 NUREG/CR-3878 MEP-2051 NUREG/CR-3228 V02 69
SECONDARY SECONDARY REPORT REPURT REPORT REPURT NUMBER NUMdER NUMBER NUMoER
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=====-------- ---- ORNL-5991 NUREG/CH-3618 PNL-5178 NUREG/CR-3907 ORNL-6006 NUREG/CR-3714 PNL-5202 NUREG/CN-3951 ORNL-6010 NUREG/CH-3524 PNL-5214 NUREG/CR-3739 ORNL-6047 NUREG/CR-3671 QUAD-1-84-109 NUREG/CR-3939 ORhL-6049 NUREG/CR-3832 SAI-83/1125 NUREG/CR-3605 V01 ORNL-TM-9249 NUREG/CM-3871 SAI-84/1010 V02 huREG/CR-3665 V02 I ORNL/ENG/TM-25 NUREG/CR-3139 SAI-84/3037 V03 NOREG/CR-3605 V03 i DRNL/NSIC-200 NUREG/CR-2000 V03 N6 SAIC-84/1317 NUREG/CR-36c5 ORNL/NSIC-200 NUREG/CR-2000 V03 N7 SAND-0402 NUREG/CR-3690 ORNL/NSIC-221 NUREG/CR-4011 SAND-84-0374 NUREG/CR-3818 ORhL/NSIC-223 NUREG/CN-3905 SAND 81-0413 NUREG/CR-3662 ORNL/TM-8549 NUREG/CH-2996 SAND 82-1580 NUREG/CR-3190 ORNL/TM-8795 NUREG/CR-3346 SAND 83-1022 NUREG/CH-3273 ORNL/TM-8889 NUREG/CR-3459 SAND 83-1438 NUREG/CR-3369 ORNL/TM-8890 NUREG/CH-3400 SAND 83-1617 NUREG/CR-3418 ORNL/TM-8902 NUREG/CR-3470 SAND 83-1960 NUREG/CR-3776 ORNL/TM-8921/V4 NUREG/CR-3492 V04 SAND 83-2425 NUREG/CR-3589 V01 ORNL/TM-9006 NUREG/CR-3590 SAND 83-2425 NUREG/CR-3589 V02 ORNL/TM-9028 NUREG/CRe3617 SAND 83-2493 NUREG/CR-3643 ORNL/TM=9061 NUREG/CR-3692 SAND 84-0660 NUREG/CR-3724 ORNL/TM=9068 NUREG/CR-3655 SAND 84-0688 NUREG/CR-3734 ORhL/TM-9154/V1 NUREG/CR-3744 V01 SAND 84-0912 NUREG/CR-3777 ORNL/TM-9157 NUREG/CR-37b3 SAND 84-0978 NUREG/CN-3786 ORNL/TM-9191/V1 NUREG/CR*3851 V01 SANDd4-1025/1 huREG/CR-3820 V01 ORNL/TM-9207 NUREG/CR-3815 SAND 84-8715 NUREG/CR-3835 ORNL/TM-9217/V1 NUREG/CR-3830 V01 UCID-19988 huREG/CH-36eo V02 ORNL/TM-9236 NUREG/CR-3856 UCID-20060 NUREG/CR-3758 ORNL/TM-9238 NUREG/CR-3888 UCRL-53021 V08 NUREG/CR-2015 V08 PNL-4163 NUREG/CR-2499 UCRL-53467 NUREG/CR-3593 V01 Pht-4824 NUREG/CN-3474 UCRL-53467 NUREG/CR-3593 V02 PNL-4876 NUREG/CH-3842 UCRL-53489 huREG/CR-3480 PNL-4886 NUREG/CR-3544 UCRL-53500 NUREG/CR-3663 V02 PNL-4915 NUREG/CR-3509 UCRL-53538 NUREG/CR-3826 PNL-4943 NUREG/CR-3610 URNL/NSIC-200 huREG/CP-2000 V03 N8 j PNL-5033 NUREG/CR-3843 nCAP-10415 NUREG/CR-3654 i PNL-5068 NUREG/CR-3751 WINCU-1006 huREG/CR-3513 PNL-5081 NUREG/CR-3796 PNL-5101 NUREG/CR-3767 PNL-5149 NUREG/CR-3869 PNL-5153 NUREC/CR-3798 l 70
Personal Author index l This index lists the personal authors of NRC staff and contractor reports In alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by that author. If further information is needed, refer to the main citation by the NUREG number. 3 ABEL,K.H. NUREG/CR-3474: LONG-LIVEu ACTIVATION PRODUCTS IN REACTOR HATERIALS. ADAMS,R.E. hbREG/CR-3830 V01: AEROSOL RELEASE AND TRANSPORT PROGRAM,SEMIANNOAL PROGRESS REPORT FOR OCT00ER 1983 - MARCH 1984 ADLER,M.V. NUREG/CR-3524: ORGANIZATIONAL INTERFACE IN REACTOR EMERGENCY PLANNING AND RESPONSE. ALAMGIR NUREG/CR-3711 BhR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST HESULTS. ALESSO,H.P. NUNEG/CR-3593 V01: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH HATRIX ANALYSIS OF A NUCLEAR PONER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. NUHEG/CR-3593 v02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR PonER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2. ANDERSON,F.D. NUNEG-1072: TECHNICAL SPECIFICATIONS FOR CATAWBA NUCLEAR STATIUN, Unit
- 1. Uocket No. 50-413.
ANDERSON,R.F. NUREG/CR-3940: FIELD EAPERIMENT DETERMINATI0h3 0F DISTRIBUTION COEFFICIENTS OF ACTINIOE ELEMENTS IN ALKALINE LAKE ENVIRONMENT 3. ARAGGN,J.J. NUREG/CR-3776: TESTING OF SAFETY-RELATED NUCLEAR P0hER PLANT EQUIPMENT AT THE CENTRAL RECEIVEN TEST FACILITY. ARNOLD,W.D. NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GF0 CHEMICAL INFORMATION DEVELOPE 0 dY DOE HIGH-LEVEL NUCLEAR HASTE REPOSITORY SITE PROJECTS. Report for UctoDer-December 1983 i ASHMORE,d.C. NUREG/CR-3593 V01: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH HATRIX j ANALYSIS OF A NUCLEAH P0nEH PLANT'S HIGH PRESSURE SAFETY INJECTION i SYSTEM. NUREG/CR-3593 v02: SYSTEMS InTERACTIUN RESULTS FROM THE DIGRAPH MATP!X l ANALYSIS OF A NUCLEAH P0nER PLANT'S HIGH PRESSURE SAFETY INJECTIUN l SYSTEM. Volume 2. AUSTIN,P.N. AUREG/CR-3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE l l 71 l
ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECUNSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. AZARM,M.A. NUREG/CR-3493: A REVIEW UF THE LIMERICh GtNERATING STATION SEVERE ACCIDENT RISK ASSESSMENT, Review of Core Melt Frequency. BAARS,R.E. NUREG/CR-3679: CALIBRATIuN AND QUALIFICATION OF THE LOS ALAMOS FAILURE MODEL (LAFM). BABCOCK,C.D. NUREG/CR-3742: BUCKLING UF STEEL CONTAINMENT SHELLS UNDER TIME-DEPENDENT LOADING. BAKERan.E. NUREG/CR-37422 BUCKLING UF STEEL CONTAINMENT SHELLS UNDER TIME-DLPENDENT LOADING. BALL,0.G. NUREG/CR-3616: OCA-P,A DETERMINISTIC AND PROBABILISTIC FRACTURE-MECHANICS CUDE FOR APPLICATION TO PRESSURE VESSELS. B BALL,S.J. NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION QUARTERLY PROGRESS REPORT, October-December 1983. BALLINGER,M.Y. NUREG/CR-3796 EMERGENCY PHEPAREDNESS SOURCE TERM DEVELOPMENT FOR THE OFFICE OF NUCLEAR MAIEHIALS SAFETY AND SAFEGUARDS LICENSED F A C li ' t '".S. NUREGiCR-3834: ON THE THRESHULD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRACKING OF SENSITIZED ALLOY 600,IN BORATED THIOSULFATE SOLUTION. BARI,R.A.. NUREG/CR-3493: A REVIEn UF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. BARLETTA,R.E. NUREG/CR-3844: CHARACTLRIZATION OF THE RADI0 ACTIVE WASTE PACKAGES UF THE MINNESOTA MINING Af4D MANUFACTURING COMPANY. BARR,P.K. NUREG/CR-3835: SIMULATION OF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE DISCnETE VORTEX METHOD. BAUM,J.W. NUREG/CR-3469 V01: OCCUPATIONAL DOSE REDUCTION AT NUCLEAR POWER PLANTS ANNOTATED BIBLIOGRAPnY OF SELECTED READINGS IN RADIATION PHOTECTI0re AND ALARA. BEAHM,E.C. NUREG/CR-3617: NOBLE GAS, IODINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT BROWNS FERRY. BENDA,0.J. NUREG/CR-3660 V02: PR0pAu!LITY OF PIPE FAILURE IN THE REACTOR COULANT LOOPS OF WESTINGHOUSE PHR PLANTS. Volume 2 Pipe Failure Induced By Crack Growth. l BENEDICK,M.B. NUREG/CR-3273: COMBUSTION UF HYDROGEN: AIR MIXTURES IN THE VGES CYLINDRICAL TANK. BENNETT,J.G. NUREG/CR-3742: BUCKLING UF STEEL CONTAINMENT SHELLS UNDER TIME-DEPENDENT LOADING. BERMAN,M. NUREG/CR-3369: AN UNCENTnINTY STUDY OF PnR STEAM EXPLOSIONS. NUREG/CR-3734: LIGHT WATLR REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982 - maren 1983. 72
l BERRY,0 L. NUREG/CR-3818: REPORT OF RESULTS OF NUCLEAR POWER PLANT AGING WORKSHOPS, BEYER,C.E'. i NUREG/CR-3907: GT2R2 AN UPDATED VERSION OF GAPCON-THERMAL-2 BEZELLA,W.A. NUREG/CR-3932 BENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. NUREG/CR-3933 RISK RELATEU RELIABILITY REQUIREMENTS FOR BWR SAFETY -IMPORTANT SYSTEMS WITH LMPHASIS ON THE RESIDUAL HEAT REMOVAL SYSTEM. SIRO,S.K. NUREG/CR-35133 MECHANICAL RELIABILITY EVALUATION OF ALTERNATE MOIORS FOR USE IN A RADICIODIHE AIR SAMPLER. BLANNIK,D.E. NUREG/CR-3787: EFFECTIVEhESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTEMS IN RETAINING FISSION PRODUCTS. Background Information. BOCCIO,J.L. NUREG/CR-3493: A REVIEN uF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. 80EGEL,A.J. NUREG/CR-3739: THE OPERATOR FEEDSACK WORKSHOP A TECHNIQUE FOR USTAINING FEEDBACK FROM OPERATIONS PERSONNEL. SOLANDER,M.A. NUREG/CR-3761: RELAPS THERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR THE UCONEE-1 PRESSURIZED WATER REACTOR. BOLSTAD,J.W. NUREG/CR-3735 ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATURE REVIEW. BRACKENBUSH,L. NUREG/CR-3610 NEUTRON 00SIMETRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of Subtask C 3ne Neutron Spectrometer. BROADWATER,R. NUREG/CR-3692: POSSIBLE MODES OF STEAM GENERATOR OVERFILL RESULTING FROM CONTROL SYSTEM MALFUNCTIONS AT OCONEE-1 NUCLEAR PLANT. BROWN,S.R. NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED INSTRUMENTATION AND CONTROL COMPONENTS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. NUREG/CR-3867: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTERS AT U.S. COMMERCIAL NUCLEAR P0nER PLANTS, JANUARY 1,1976 TO DECEMdER 31,1982 BRYSON,M.C. I NUREG/CP-0053: PROCEEDINGS OF THE NINTH ANNUAL STATISTICS SYMPOSIUM ON NATIONAL ENERGY ISSUES,0ctober 19-21,1983. BUCKALEW,W.H. NUREG/CR-3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA i PELLETRON-RASTER SYSTEM. ( BUEHRING,W.A. NUREG/CR-3929 LOSS-OF-BENEFITS ANALYSIS FOR NUCLEAR POWER PLANT SHUTD0nNS. Methodology And Illustrative Case Study. BUSCHBOM,R.L. NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS l CONTAINING SIMULATED LUW-LEVEL WASTE INCINERATOR ASH. BUTLER,T.A. NUREG/CR-3821: EVALUATION OF CRACK PLANE EQUILIBRIUM MODEL FOR l PREDICTING PLASTIC FRACTURE. CANO,G.L. l NUREG/CR-3662: FUEL-DISRUPTION EXPERIMENTS UNDER HIGH-RAMP-RATE HEATING CONDITIONS. CATON,G.M. l 73
1 NUREG/CR-3590: EVALUATION OF ISOTOPE DILUTION MASS SPECTROMETRY FOR BIDASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. CHAPMAN,R.H. NUREG/CR-3459: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRST) BUNDLE B-5 NUREG/CR-3460: EXPERIMENT 0ATA REPORT FOR MULTIROD BURST TEST (MRBT) BUNDLE B-6 CHARAK,I. NUREG/CR-3932 BENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. j CHELLIAH,E. 'NUREG-1068: REVIEW INSIGHTS ON THE PROBABILISTIC RISK ASSESSMENTS FOR l THE LIMERICK GENERATING STATION,0 NIT 1 AND 2 CHEVERTON,R.D. 1 NUREG/CR-3618: OCA-P,4 DETtRMINISTIC AND PROBABILISTIC l FRACTURE-MECHANICS CODE FOR APPLICATION TO PRESSURE VESSELS. CLAPP,N.E. NUREG/CR-3692: POSSIBLE MODES OF STEAM GENERATOR OVERFILL HESULTING FROM CONTROL SYSTEM MALFUNCTIONS AT OCONEE-1 NUCLEAR PLANT. CLARn,F.H. NUREG/CR-3692 POSSIBLL MODES OF STEAM GENERATOR OVERFILL RESULTING FROM CUNTROL SYSTEM MALFUNCTIONS AT OCONEE-1 NUCLEAR PLANT. CLARK,N.H. NUREG/CR-3818: REPORT OF RLSULTS OF NUCLEAR POWER PLANT AGING WORKSHOPS. CLARn,P.V. NUREG/CR-3750: JOB ANALYSIS OF NUCLEAR POWER REACTOR HEALTH PHYSICS TECHNICIANS. CLARK,R.A. NUREG/CR-3840: COST ANALYSIS FOR POTENTIAL MODIFICATIONS To ENHANCE THE ABILITY OF A NUCLEAR PLAHT TO ENDURE STATION BLACKOUT. CLAYTOR,T.N. NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT nATER REACTORS: Annual Report,Uctober 1982 - September 1983 CLEVELAND,J.C. NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-COOLED REACTUR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION GUARTERLY PH0GRESS REPORT, October-December 1983. CLOUGH,R.L. NUREG/CR-3643: HETEROGENLOUS OXIDATIVE DEGRADATION IN IRRADIATED POLYMERS. COATS,0.n. NUREG/CR-3480: VALUE/ IMPACT ASSESSMENT FOR SEISMIC DESIGN CRITERIA USI A-40 COCHRAN,M.I. NUREG/CR-3951: INTRODUCTION TO BIBELOT:A BIBLIOGHAPHIC FINDING AND RETRIEVAL SYSTEM. CODELL,R.B. NUREG-1054: SIMPLIFIED ANALYSIS FOR LIQUID PATHWAY STUDIES. ConEN,J.J. NUREG/CR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PONER PLANTS. Executive Summary. NOREG/CR-3065 V02: OPTIM1ZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonER PLANTS. Considerations In Factorina Occupational Dose Into Value= Impact And Cost-Benefit Analyses. COLMAR,R. NUREG-0933 S01: A PR10HITI4ATION OF GENERIC SAFETY ISSUES. COOK,J.A. NUREG/CR-3708: LhR SPENT FUEL DRY STORAGE GEHAVIOR Al 229 C. COPENHAVER,E.D. 74
NUREG/CR-3524: ORGANIZATIONAL INTERFACE IN REACTOR EMERGENCY PLANNING AND RESPONSE. CORWIN,W.R. NUREG/CR-3071: ASSESSMENT OF RADIATION EFFECTS RELATING TO REACTOR PRESSURE VESSEL CLAD 0!NG. COTTRELL,W.B. NUREG/CR-3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECUNSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTd: 1980-1981 4 Status Report. CRAFT,C.M. NUREG/CR-3418: SCREENING TESTS OF TERMINAL BLOCK PERFORMANCE IN A SIMULATED LOCA ENVIRONMENT. CRONLEY,J.L. NUREG/CR-3459: EXPERIMLNT DATA REPORT FOR MULTIROD BURST TEST (MRBT) SUNDLE B-5 NUREG/CR-3460: EXPERIMENT UATA REPORT FOR MULTIROD BURST TEST (MHST) BUNDLE B-6 CULLEN,W.H. NUREG/CR-3833: BEHAVIOR OF SUBCRITICAL AND SL0n-STABLE CRACK GR0 NTH FOLLonING A POST-IRRADIATION THERMAL ANNEAL CYCLE. CUMMINGS,G.E. NUREG/CR-2015 V08: PHAbE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. 'CUMMINGS,J.C. NUREG/CR-3273: COMBUSTION OF HYDROGEN AIR MIXTURES IN THE VGES CYLINDRICAL TANK. CUNNINGHAM,M.E. NUREG/CR-3907: GT2R2 AN UPDATED VERSION OF GAPCON-THERMAL-2 CURRIE,L.A. NUREG/CR-4007: L0nER LIMIT OF, DETECTION: DEFINITION AND ELABORATION OF A PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL MEASUREMENTS. CUTSHALL,N.H. NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPEu oY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report for Octouer-December 1983 CYBULSKIS,P. NUREG/CR=3988: MARCH 2 (MELTUOWN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USERS MANUAL. CZAJK0nSKI,C.J. NUREG/CR-3766: TESTING OF NUCLEAR GRADE LUBRICANTS AND THEIR EFFECT ON AS40 AND A193 B7 BOLIING MATERIALS. DALY,8.J. NUREG/CR-3822: SOLA-PTS: A Transient,Threc= Dimensional Algorithm For Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometrics. DANDINI,V.J. NUREG/CR-3776: TESTING OF SAFETY-RELATED NUCLEAR POWER PLANT EQUIPMENT AT THE CENTRAL RECEIVER TEST FACILITY, DAVIS,L.T. NUREG/CR-3750: JOB ANALYSIS OF NUCLEAR POWER REACTOR HEALTH PHYSICS TECHNICIANS. t l DAYAL,R. NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA. DEAN,R.S. j NUREG/CR-3053: CLOSE0UT OF IE BULLETIN 80-08 EXAMINATION OF CONTAINMENT LINER PENETRATION WELDS. l NUREG/CR-3792: CLOSEQUI UF IE BULLETIN 79-11 FAULTY UVERCURRENI TRIP l DEVICE IN CIRCUIT 8REAnERS FOR ENGINEERED SAFETY SYSTEMS. 1 i l 75 i - = - - - -. - - -
DEDERER,J.T. NUREG/CR-3654: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14 DIVINE,J.R. NUREG/CR=37983' CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. DOCTOR,P.G. NUREG/CR-3842: STEAM GENERATOR GROUP PROJECT TASK 8 = SELECTIVE TU8E UNPLUGGING. DODGE,C.J. 1 NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADWASTE CONTAINING j ORGANIC ION-EXCHANGE MEDIA. DOERGE,D.H. NUREG/CR-3884: EVALUATION UF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM - THREE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. DOUGHERTY,D. NUREG/CR-3844: CHARACTERAZATION OF THE RADIDACTIVE WASTE PACKAGES OF THE MINNESOTA MINING AND MANUFACTURING COMPANY. DRESS,W.B. NUREG/CR-3856: AN ULTRASUNIC LEVEL AND TEMPERATURE SENSOR FOR POWER REACTOR APPLICATIONS. DYER,F.F. NUREG/CR-3590: EVALUATION UF ISOTOPE DILUTION MASS SPECTROMETRY FOR BI0 ASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. DYnSTRA,J. NUREG/CR-3139: SCENARIUS AND ANALYTICAL METHUDS FOR UF6 RELEASES AT -NRC-LICENSED FUEL CYCLE FACILITIES. ECKER,R.M. NUREG/CR-3751 EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FLOOD PROTECTION COSTS FOR URAdIUM TAILINGS IMPOUNDMENTS. EINZIGER,R.E. NUREG/CR-3708: LnR SPENT FUEL DRV STORAGE SEHAv!OR AT 229 C. EMBREY,0.E. NUREG/CR-3518 V01: SLIM-MAUDIAN APPROACH TO ASSESSING HUMAN ERROR PROBABILITIES USING STRUCTURED EXPERT JUDGEMENT. Volume I Overview of SLIM-MAUD, EMRIT,R. NUREG-0933 S01: A PRIORITI4ATION OF GENERIC SAFETY ISSUES. ENDRES,G.W. NUREG/CR-3544: BETA PARTICLE MEASUREMENT AND 00SIMETRY AT NRC-LICENSED FACILITIES. NUREG/CR-3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITLS. ERASLAN,A.H. NUREG/CR-3871: AN OVERVILW OF THE UNIFIED TRANSPORT APPROACH. EVANS,J.C. NUREG/CR-3474; LONG-LIVF0 ACTIVATION PRODUCTS IN REACTUR MATERIALS. EVANS,R.P. 1 NUREG/CR-0169 V17 LOFT tXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recorded On DAVDS. FAINBERG,A. NUREG/CR-3520 V01: LONG-IERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR P0aER PLANTS. Volume I Summary And Users Guide. NUREG/CR-3520 V02: LONG=IERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR P0aER PLANTS. Volume II: Development Of uetailed Analyses. FARRAR,C. NUREG/CR-3845: PREDICTION UF NONLINEAR STRUCTURAL RESPONSE IN LMFRR ELEVATED-TEMPERATURE PIPING. l 76 1
FETROW,L.K. NUREG/CR-3842: STEAM GENERATOR GROUP PROJECT TASK 8 - SELECTIVE TU8E UNPLUGGING. FIELDS,M.B. NUREG-0978 MARK III LUCA-RELATED HYDRODYNAMIC LOAD DEFINITION. Generic Technical Activity 8-10. final Report. FLETCHER,C.D. NUREG/CR-3761: RELAPS THERMAL-HYDRAULIC ANALYSES OF PRESSURIZEU THERMAL SHOCK SEQUENCES FOR THE.UCONEE-1 PRESSURIZED WATER REACTOR. FLY,J. NUREG/CR-3742 BUCKLING OF STEEL CONTAINMENT SHELLS UNDER TIME-DEPENDENT LOADING. FOLEY,n.J. NUREG/CR-3053: CLOSE0Vi UF IE BULLETIN 80-08 EXAMINATION OF CONTAINMENT LINER PENETRATION WELDS. NUREG/CR-3792: CLOSE0UT UF IE BULLETIN 79-11 FAULTY OVERCURRENT TRIP UEVICE IN CIRCUIT BREAKERS FOR ENGINEERED SAFETY SYSTEMS. NUREG/CR-3795: CLOSEQUI 0F IE BULLETIN 82-04: DEFICIENCIES IN PRIMANY CONTAINMENT ELECTRICAL PENETRATION ASSEMBLIES. FOX,R.A. NUREG/CR-3544 SETA PARTICLE MEASUREMENT AND 00SIMETRY AT NRC-LICENSED FACILITIES. NUREG/CR-3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT 4 4 COMMERCIAL NUCLEAR SITES. FREY,P.R. NUREG/CR-3655: A METHOU FOR ANALYTICAL EVALUATION OF COMPUTER-BASED DECISION AIDS. GALLAGHER,D. NUREG/CR-3852: INSIGHT INTO PRA METHODOLOGIES, GENTILLON,C.0, NUREG/CR-3824: CONTING PROGRAM GUIDE. GEORGE,L.L. huREG/CR-2015 V08: PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. GILLER,K.T. NUREG/CR-3643: HETEROGENEOUS OXIDATIVE DEGRADATION IN IRRADIATED POLYMERS. GINEVAN,M. NUREG-1029: A COMPUTER CODE FOR GENERAL ANALYSIS OF RADON RISKS (GARR). GREENE,N.M. [ NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT i REPORTS. Users Gu'ide. l GREGORY,W.S. NUREG/CR-3735: ACCIDENT-INDUCED FLOW AND MATERI AL TRANSPORT IN NUCLEAR FACILITIES--A LITERATURE REVIEW. GRIFFITH,P. NUREG/CR-3895: INVESTIGATION OF COLD LEG MATER HAMMER IN A PNR DUE TO THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. GUPPY,J.G. J NUREG/CR-3169: SUPER SYSTEM CODE (SSC,REV 0).AN ADVANCED THERM 0 HYDRAULIC SIMULATIUN FOR TRANSIENTS IN LMFBRS. GUZIEL,K.A. NUREG/CR-3929: LOSS =0F-8ENEFITS ANALYSIS FOR NUCLEAR POWER PLANT i SHUTDOWNS Methodology And Illustrative Case Study. HAGEN,E.W. NUREG/CR-3591 V01: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECURSUR3 TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. HAGGARD,D.L. 77
NUREG/CR-3569: SPECIAL AND 00SIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. HALVERSON,M.A. NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTEMS IN RETAINING FISSICN PHODUCTS. Background Information. l HALVERSON,S.L. NUREG/CR-3932: SENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS i AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. ] HANAN,N. NUREG/CR-3493: A REVIEW OF TnE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. HANNON,J.L. NUREG-1075: DECENTRALIZATION OF OPERATING REACTOR LICENSING NEVIEWS.NRR Pilot Program. HARBEN,P. i NUREG/CR-3758: CROSSHOLE GEOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL HASTE REPOSITORIES. HARRINGTON,R.M. NUREG/CR=3470 ATWS AT BH0nNS FERRY UNIT ONE - ACCIDENT SEQUENCE ANALYSIS. NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION QUARTERLY PROGRESS REPORT, October-December 1983. HARRIS,J.D. NUREG/CR-3591 V01: PREGURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECUNSORS TO POTENTIAL SEVERE 60RE DAMAGE ACCIDENTS: 1980-1981 A Status Report. HARVEY,C.O. NUREG/CR-3798: CHAR ;TERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. HENNICK,A. NUREG/CR-3053: CLOSE0UT OF IE BULLETIN 80-08:EXAMINATIUN OF CONTAINMENT LINER PENETRATION nELDS. NUREG/CR=3792: CLOSEOUT OF IE BULLETIN 79-11 FAULTY OVERCURRENT TRIP DEVICE IN CIRCUIT BREAKERS FOR ENGINEERED SAFETY SYSTEMS. NUREG/CR-3795: CLOSE0VT OF IE BULLETIN 82-04 DEFICIENCIES IN PRIMARY CONTAINMENT ELECTRICAL PENETRATION ASSEMBLIES. HENRY,E.B. NUREG/CR-3824: CONTTNG PROGRAM GUIDE. HERCZEG,A.L. NUREG/CR-3940: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN ALKALINE LAKE ENVIRONMENTS. HERRINGTON,W.N. j NUREG/CR-2499: REVIEn UF EMENGENCY RADIOLOGICAL INSTRUMENTATION ANO J ANALYTICAL METHODS AT NMSS-LICENSEE SITES. HISER,A.L. NUREG/CR-3833 BEHAVIOR OF SUSCRITICAL AND SLOW-STABLE CRACK GROWTH FOLLOWING A POST-IRRADIATION THERMAL ANNEAL CYCLE. HOCHREITER,L.E. NUREG/CR-3654: PnR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION j AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14 HODGE,5.A. NUREG/CR-3470 ATHS AT BHonNS FERRY UNIT ONE - ACCIDENT SEQUENCE ANALYSIS. NUREG/CR-3617: NOBLE GAS,IUDINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT BRO *NS FERRY. HOFFMAN,0.R. NUREG-0926 R01: TECHNICAL bPECIFICATIONS FOR GRAND GULF NUCLEAR 78
STATION, UNIT 1. Docket No. 50-416, (Mississippi Power And Light Company) HOLBROOK,K.L. NUREG/CR-3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. HOLMAN,G.S. NUREG/CR-3663 V02: PROBAdILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF CUMBUSTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced oy Crack Growth. HOLT,0.0. NUREG/CR-3139: SCENARIUS AND ANALYTICAL > METHODS FOR UF6 RELEASLS Al NRC-LICENSED. FUEL CYCLE FACILITIES. HORTON,W.H. NUREG/CR-3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR POWER PLANTS.A Calculation Method. HOWARD,G.E. NUREG/CR-3893: LABORATORY STUDIES DYNAMIC RESPONSE OF PROTOTYPICAL PIPING SYSTEMS. HUCHTON,R.L. NUREG/CR-3513 MECHANICAL RELIABILITY EVALUATION OF ALTERNATE MOTORS FOR USE IN A RADIDIOJINE AIR SAMPLER. HUMPHREYS,P. NUREG/CR-3518 V01: SLIM-MAUDIAN APPROACH TO ASSESSING HUMAN ERROR PROBABILITIES USING STRUCTURED EXPERT JUDGEMENT. Volume I Overview of SLIM-MAUD. HUXTABLE,W.P. NUREG/CR-3139: SCENARIUS AND ANALYTICAL METHODS FOR UF6 RELEASES AT NRC-LICENSED FUEL CYCLE FACILITIES. HWANG,W.S. NUREG/CR-3711: BhR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST RESULTS. JACK 00EK,A.B. NUREG/CR-3895: INVESTIGATION OF COLD LEG WATER HAMMER IN A PWR DUE TO THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. JACOBS,G.K. NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPE 0 dY DOE HIGH= LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report for October = December 1983. JACOBUS,M.J. NUREG/CR-3786: A REVIEW OF REGULATORY REQUIREMENTS GOVERNING CONTROL ROOM HABITABILITY SYSTEMS. JAMISON,J.D. NUREG/CR-2499: REVIEW OF EMERGENCY RADIOLOGICAL INSTRUMENTATION AND I ANALYTICAL METHODS AT NMSS-LICENSEE SITES. JOHNSON,B.A. NUREG/CR-3893: LABORATORY STUDIES DYNAMIC RESPONSE OF " ROT 0 TYPICAL j PIPING SYSTEMS. JOHNSON,M.P. t NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Gulde. JUNG,J. l NUREG/CR-3724: ULTIMATE STRENGTH ANALYSES OF THE WATTS BAR, MAINE l YANKEEfAND BELLEFONTE. CONTAINMENTS. JUST,R.A. I NGREG/CR-3139: SCENARIOS AND ANALYTICAL METHODS FOR UF6 RELEASES AT NRC-LICENSED FUEL CYCLE FACILITIES. KAFKA,A. NUREG/CR-3493: A REVIEW 0F THE LIMERICr. GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. KALUZNY,8.P. 79 -~ - -. - - - - -. -.. -. - - -..-.-
NUREG/CR-3896 SIMULATION EXPERIMENTS COMPARING ALTERNATIVE PROCESS FORMULATIONS USING A FACTORIAL DESIGN. KAM,8.K. NUREG/CR-3888: ANALYSIS OF THE VENUS PhR ENGINEERING MOCKUP EXPERIMENT -PHASE I SOURCE DISTRIBUTION. KANA,0.D. NUREG/CR-3892: A RESEARCH PRUGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Summary Report. NUREG/CR-3892 V01: A RESEANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL E4UIPMENT. Task 1 - Survey of Methods For Equipment Ano components Evaluation of 4 Methodology;Wualification And Methodology.... NUREG/CR-3892 V02: A RESEANCH PRUGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 2-Correlation i Of Methoaologies For Seismic Qualification Tests Of Nuclear Plant Equipment. NUREG/CR-3892 V03: # RLSEARCH PROGRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement Of Equipment Qualification Methodology Ano Criteria. NUREG/CR-3892 V04: A RLSEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 - Tne use Of Fragility In Design Uf Nuclear Plant Equipment. KASSNAR,T.F. NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT MATER REACT 0HS: Annual Report,Uctober 1982 - September 1983 KATHREN,R.L. NUREG/CR-2499: REVIEW 0F EMERGENCY RADIOLOGICAL INSTRUMENTATION AND ANALYTICAL METHODS AT NMSS= LICENSEE SITES. KELLY,K. NUHEG/CR-3634: ON THE THNEdHOLD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRACKING OF SENSITIZED ALLOY 600 IN 80 RATED THIOSULFATE SOLUTION. i KELMERS,A.D. NUREG/CR-3763: REVIEn AND ASSESSMENT OF RADIONUCLIDE SORPTION INFORMATION FOR THE BASALT WASTE ISOLATION PROJECT SITE (1979 Through May,1983). NUREG/CR-3851 V01: pHOGRLSS IN EVALUATION OF RADIONUCLIDE GE0 CHEMICAL INFONMATION DEVELOPE 0 dY DUE HIGH= LEVEL NUCLEAR WASTE HEPOSITORY SITE PROJECTS. Report for UctoDer-December 1983 KEMpF,C.R. NUREG/CR-3844: CHARACTERIZATION OF THE RADI0 ACTIVE WASTE PACKAGES OF THE MINNESOTA MINING AHD MANUFACTURING COMPANY. KEN 0YER,J.L. NUREG/CR-2499: REVIEn OF EMENGENCY RADIOLOGICAL INSTRUMENTATION AND ANALYTICAL METHODS AI NMSS-LICENSEE SITES. KESSLER,J.H. l NUREG/CR-3851 V01: PROGRLSS IN EVALUATION OF RADIONUCLIOE GEOCHEMICAL l INFORMATION DEVELOPEp dY DOE HIGH-LEVEL NUCLEAR WASTE REPOSIIORY SITE PROJECTS. Report for Uctooer-December 1983. KIRMAN,B. NUREG/CR-3518 V01: SLIM-MAUD:AN APPROACH TO ASSESSING HUMAN ERHON PROBASILITIES USING STHUCTURED EXPERT JUDGEMLNT. Volume I Overview of SL?_M-MAUD. KISNER,R.A. NUREG/CR-3655: A METH00 FOR ANALYTICAL EVALUATION OF COMPUTER-BASED l OECISION AIDS. KMETY,K.L. NUREG/CR-3690: RELAPS ASSESSMENT:SEMISCALE NATURAL CIRCULATION TESTS S-NC-3,S-NC-4,AND S-NC-8 80 i
K0CHER,D.C. NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONMENTAL RADIATION STANDARDS FOR GEOLOGIC DISPOSAL uF HIGH-LEVEL RADI0 ACTIVE WASTES. NUREG/CR-3832: UNCERTAINTIES IN LONG-TERM REPOSITORY PERFORMANCE DuE TO THE EFFECTS OF FUTURE GE0LUGIC PROCESSES. KOVDALSKJ,F.J. NUREG/CR-3654: PhR FLECHI dEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report .NRC/EPHI/ Westinghouse Report No. 14. KROEGER,P.G. NUREG/CR-3868: CONTAINMENT BUILDING ATMOSPHERE RESPONSES OUE TU REACTOR GAS BURNING UNDER SEVEdE ACCIDENT CONDITIONS. KUDRICK,J.A. NUREG-0978 MARK III LUCA-RELATED HYDRODYNAMIC LOAD DEFINITION. Generic Technical Activity B-10. Final Report. KUPPERMAN,0 NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS: Annual Report,0ctober 1982 - September 1983 NUREG/CR-3894: ULTRAS 0nIC AND METALLURGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS STEEL BnR PIPE nELDMENT. LAPPA,0.A. NUREG/CR-3480: VALUE/ IMPACT ASSESSMENT FOR SEISMIC DESIGN CRITERIA USI A-40 LEE,S.Y. NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORNATION DEVEl ' PED dY DUE HIGH-LEVEL NUCLEAR HASTE REPOSITORY SITE PROJECTS. Report for uctouer-December 1983. LEGGETT,R.W. NUREG/CR-3346: BI0 ASSAY DATA AND A RETENTION-EXCRETION MODEL FOR SYSTEMIC PLUTONIUM. LEPEL,E.L. NUREG/CR-3474: LONG-LIVEU ACTIVATION PRODUCTS IN REACTOR MATERIALS. LEw!S,M. NUREG/CR-3842: STEAM GENERATOR GROUP PROJECT TASK 8 - SELECTIVE TU6E UNPLUGGING. LI,Y.H. NUREG/CR-3940 FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIOE ELEMENTS IN ALKALINE LAKE ENVIRONMENT 3. LIN,C.C, NUREG/CR-4001: CONTEMPT 4/ MUDS AN IMPROVEMENT TO CONTEMPT 4/M004 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. LINDSEY,C.G. NUREG/CR-3796: EMERGENCY PREPAREDNESS SOURCE TERM DEVELOPMENT FOR THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES. LO,T.Y. NUREG/CR-3663 V02: PROBApILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced by Crack Growth. 'LOBNER,P. NUREG/CR-3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR POWER PLANTS.A Review Of Occupational Dose Assessment Considerations In Current Probabilistic Risk Assessment And Cost-Benefit Analyses. LOCKWOOD,G.J. NUREG/CR-3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON=RASTEP SYSTEM. LOKKEN,R.O. NUREG/CR-3798 CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS 81
CONTAINING SIMULATED LUN-LEVEL WASTE INCINERATOR ASH. LONGEST,A.N. NUREG/CR-3459: ExPERIMEN1 DATA REPORT FOR HULTIR0D BURST TEST (MR9T) BUNDLE B-5 NUREG/CR-3460: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MH8T) BUNDLE B-6 LUKER,5.M. NUREG/CR=3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON-RASTER SYSTEM. LYTLE,R.J. NUREG/CR-3758: CROSSHOLE GEOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL WASTE REPOSITORIES. 'MADNI,1,K. NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYSTEM OF THE j EXPERIMENTAL BREEDER REACTOR-II. MAIYA,P.S. t NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATEN REACTORS: Annual Report,Uctober 1982 - September 1983 MARGULIES,T.S. NUREG=0935: ACQUSTIC WAVE PROPAGATION IN FLUIDS WITH COUPLED CHEMICAL REACTIONS. MAROTTA,F.J. NUREG/CR-3750: JOB ANALYSIS OF NUCLEAR POWER REACTOR HEALTH PHYSICS TECHNICIANS. MARSHALL,R.S. NUREG/CR-3678: ES f!MATION METHODS FOR PROCESS HOLDUP OF SPECI AL NUCLE AR MATERIALS. MARTIN,R.A. NUREG/CR-3735: ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR i FACILITIES--A LITERATURE REVIEn. MAST,P.K. NUREG/CR-3190: PLUGM A COUPLED THERMAL-HYDRAULIC COMPUTER MODEL FOR-4 FREEZING MELT FLOW IN A CHANNEL. MATTHEWS,P. i NUREG=0933 S01: A PRIORITIZAIION OF GENERIC SAFETY ISSUES. MAY,M.P. NUREG/CR-3590: EVALUATION UF ISOTOPE DILUTION MASS SPECTROMETRY FOR 6IDASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. MAYS,G.T. NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Guide. MAZOUR,T.J. NUREG/CR-3750: JOB ANALYSIS UF NUCLEAR POWER HEACTOR HEALTH PHYSICS TECHNICIANS. MCCANN,M. NUREG/CR-3493: A REVIEn OF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. MCDONALD,J.R. NUREG/CR-3874: NEAR-GROUND TURNADO WIND FIELDS. MCELROY,n.N. NUREG/CR-3318: LWR PRESSURL YESSEL SURVEILLANCE 00SIMETRY IMPROVEMLNT PROGRAM PCA Experiments,olina Test,And Physics-Dosimetry Support For The PSF Experiments. MCGUIRE,M.V. NUREG/CR-3739: THE OPERATON FEEDBACK WORKSHOP:A TECHNIuuE FOR ORTAINING FEEDBACK FROM OPERATIONS PLRSONNEL. MCKNIGHT,K.D. NUREG/CR-0169 V17: LOFT LXPEHIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume xyII Process Instruments Recorded On DAVDS. MENSTNG,R.M. 82
NUREG/CR-3660 V02: PR0 dab!LITY OF PIPE FAILURE IN THE REACTOR COULANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 2 Pipe Failure Induced By Crack Growth. NUREG/CR-3663 V02: PRodAdILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced by Crack Growth. MERRYMAN,R.G. NUREG/CR-3735: ACCIDENT-INDUCED FLON AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATURE REVIEn. MEnHINNEY,J.A. NUREG/CR-3870: RADIATIUN DUSL ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIOES. Annual Progress Rept July 1982 -June 1983. MEYER,R.E. NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED dY DUE HIGH-LEVEL NUCLEAR nASTE REPOSITORY SITE PRUJECTS. Report for Octouer-December 1983. MILETI,0.5 NUREG/CR-3524: ORGANIZATIONAL INTERFACE IN REACTOR EMERGENCY PLANNING AND RESPONSE. MILLER,G.N. NUREG/CR-3856: AN ULTRASONIC LEVEL AND TEMPERATURE SENSOR FOR PonER REACTOR APPLICATIONS. MILLER,N.E. NUREG/CR-3900 V01: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.First Quarterly Report, Year Three, April-June 1984 MILLER,R.L. NUREG/CR-3884: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM - THREE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. MILSTEAD,W. NUREG-0933 301: A PRIORITIZATION OF GENERIC SAFETY ISSUES. MINARICK,J.W. NUREG/CR-3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE l ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. MINNERS,W. NUREG-0933 Sois A PRIORITIZATION OF GENERIC SAFETY ISSUES. MISHIMA,J. NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS l IN RETAINING FISSION PRODUCTS. Background Information. NUREG/CR-3796: EMERGENCY PREPAREDNESS SOURCE TERM DEVELOPMENT FOR THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES. MORAnINYO,P.O. NUREG/CR-3888: ANALYSIS OF THE VENUS PWR ENGINEERING MOCKUP EXPERIMENT -PHASE Is SOURCE DISTRIBUTION. MOTES,8.G. NUREG/CR-3513: MECHANICAL RELIABILITY EVALUATION OF ALTERNATE MOTORS FOR USE IN A RADIO 100INE AIR SAMPLER. MUELLER,C.J. NUREG/CR-3932: SENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS l AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. MURPHY,E.S. NUREG/CR-0130 ADD 03: TECHNOLuGY, SAFETY AND COSTS OF DECOMMISSIGNING A REFERENCE PRESSURIZED nATER REACTOR PonER STATION. NUREG/CR-Oo72 ADD 02: TECHNOLOGY, SAFETY AND COSTS OF UECOMMISSIONING A l REFERENCE BOILING WATER REACTOR POWER STATION. Classification Of Decommissioning Wastes. 83 I
=- NICHOLS,F.A. NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORSI Annual Report,Uctober 1982 - September 1983 NOVAT,J. NUREG/CR-37358 ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATURE REVIEW. O'BRIEN,J.N. NUREG/CR-3520 V01: LONG-TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR PONER PLANTS. Volume ItSummary And Users Guide. NUREG/CR-3520 V02: LONG-TEHM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR PonER FLANTS. Volume IItDevelopment Of Detailed Analyses. OLSEN,C.S. NUREG/CR-3921 DRY SPENT FUEL STORAGE TEST PLAN FOR FINAL NONDESTRUCTIVE FUEL ROD EXAMINATION. PAPAZOGLOU,I.A. NUREG/CR-3493: A REVIEn UF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. PARK,J.Y. NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS: Annual Report,0ctober 1982 - September 1983 NUREG/CR-3894: ULTRASONIC AND METALLURGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS SIEEL BnR PIPE nELOMENT. PEERENB00M,J.P. = NUREG/CR-39298 LOSS-OF-BENEFITS ANALYSIS FOR NUCLEAR POWER PLANT SHUTD0nNS. Methodology And 111ustrative Case Study. PELROY,R.A. NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. PICARD,R.R. NUREG/CR-3678: ESTIMATION METHODS FOR PROCESS HOLDUP OF SPECIAL NUCLEAR MATERIALS. PILCH,M. NUkEG/CR-3190: PLUGMt A COUPLED THERMAL-HYDRAULIC COMPUTER MODEL FUR FREEZING MELT FLOW IN A CHANNEL. PILLAY,K.S. NUREG/CR-3678: ESTIMATION METH003 FOR PROCESS HOLOUP OF SPECIAL NUCLEAR MATERIALS. PITTMAN,J. NUREG-0933 S01: A PRIORITIZATION OF GENERIC SAFETY ISSUFS. POLCH,E.Z. NUREG/CR-3892 V01: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey of Methods For Equipment And Components: Evaluation Of Methodology 7 Qualification And Methodology.... POMERENING,0.J. NUREG/CR-3892 V01: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey of Methods For Equipment And Components Evaluation Of Methodology 7 Qualification And Methodology.... NUREG/CR-3892 V02: A RESLAdCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRACAL AND MECHANICAL EQUIPMENT. Task 2-Correlation Of Methodologies For Seismic Qualification Tests Of Nuclear Plant Equipment. NUREG/CR-3692 V03: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement Of Equipment Qualification Methodology And Criteria. NUREG/CR-3892 V04: A RESEARCH PROGRAM FOR SEISMIC UUALIFICATION OF The Use Of NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPHENT. Task 4 84
( -d s Fragility In'Desjgn Uf Nuclear Plant Equipment. PO3TMA,A.K. NUREG/CR-3787: EFFECTIVENESS OF EtlGINEERED SAFETY FEATURE (ESP) SYSTEMS IN RETAINING FIOSION PRODUCTS.Dackground Information. PRASSINOS,P.G. NUREG/CR-3273: COMBUSTION OF HYDROGEN: AIR MIXTURES IN THE VGES CYLINDRICAL TANK. ( PUGH,C.E. NUREG/CR-3744 V01: HE AVY-SECTION ' STEEL TECHNOLOGY PRUGR AM SEMI A?JNUAL PROGRESS REPORT FOR UC10dEH 1983 ' MARCH 1984 QUAGLIA,R. NUREG/CR-3654: PWR FLECHT SEASET SiSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION.0ata Evaluation ano Analysis Report NRC/EPRI/Hestinghouse Report No,'14. QUAYLE,S.F. NUREG/CR-3988: MARCH 2 (MELTD0nN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USERS MANUAL. QUINTANA,C.A. NUREG/CR-3643: HETEROGENt00S 0xIDATIVE DEGRADATION IN IRRADIATED POLYMERS.
- s RAMIREZ,A.L.
INVESTIGATE THE NUREG/CR-3758: CROSSHOLE GLOPHYSICAL METHODS USED TU NEAR VICINITY OF HIGH LEVEL HASTE REPOSITORIES. RATHBUN,L.A. NUREG/CR-3544: BETA PARTICLE HEASUREMENT AND 00SIMETRY AT NRC-LICENSED FACILITIES. NUREG/CR-3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITLS. REA,K. NUREG/CR-3518 V01: SLIM-MAUD:AN APPROACH TO ASSESSING HUMAN ERROR PROBABILITIES USING STdVCTURED EXPERT JUDGEMENT. Volume I Overview of SLIM-MAUD. 1.D. /CR-3610: NEUTRON DUSIMLTRY AT COMMERCIAL NUCLEAR PLANTS F'inal . ort Of Subtask C 3He Neutron Spectrometer. 'C R-3493 : A REVIEW UF THE LIMERICK GENERATING STATION SEVERE ' DENT RISK ASSESSMENT. Review of Core Melt Frequency. .P. 2996: SENSITIVITY OF DETECTING IN-CORE VIBRATIONS AND BOILING iSURIZED WATER REACTORS USING EX-CORE NEUTRON NOISE. c-3689 V03: MATERIALS SCIENCE AND TECHNOLOGY DIVISION .nATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress rt, July-September 1983 RIANa,.. NUREG-0933 S01: A PRIOHITIZATION OF GENERIC SAFETY ISSUES. RIGGS,R. NUREG-0933 S01: A PRIORITIZATION OF GENERIC SAFETY ISSUES. Qsp.1g RIORDON,B.J. j%4, NUREG/CR-3840: COST ANALYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE 'yp .c ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKOUT. d.R
- [#f ROBERSON,P.L.
-3$ NUREG/CR=3544: BETA PARTICLE MEASUREMENT AND 00SIMETRY AT NRC-LICENSED FACILITIES. ( W^ NUREG/CR-3569: SPECIAL AND 00SIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. %Y. f ROBERTSON,0.R. MS NUREG/CR-3474: LONG-LIVED ACTIVATION PRODUCTS IN REACTOR MATERIALS. 'fj.
- RODABAUGH,E.C.
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NUREG/CR-35993 SOURCES OF UNCERTAINTY IN THE CALCULATIONS OF LOADS ON SUPPORTS OF PIPING SYSTEMS, ROSA,E.A. NUREG/CR-3518 V01: SLIM-MAUD AN APPROACH TO ASSESSING HUMAN ERROR PROBABILITIES USING STHUCTURED EXPERT JUDGEMENT. Volume I Overview of SLIM-MAUD. ROSAL,E.R. NUREG/CR-3654: PHR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND HEFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Heport No. 14 ROUSE,S.H. NUREG/CR-3655: A METH00 F0H ANALYTICAL EVALUATIOP 9F COMPUTER-BASED DECISION AIDS. ROUSE,W.B. NUREG/CR-3655: A METHOD FOR ANALYTICAL EVALUATION OF COMPUTER-BASEo DECISION AIDS. RUGER,C. NUREG/CR-3493: A REVIEN UF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. RUGGLES,L.E. NUREG/CR-3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON-RASTER SYSTEM. RUPPHECHT,S.D. NUREG/CR-3654: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse HePort No. 14 RUTHER,W.E. NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS Annual Report,0ctober 1982 - September 1983 SACKS,1.J. NUREG/CR-3593 V05; SYSIEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF # NUCLEAR P0nEH PLANT'S HIGH PRESSURE SAFETY INJLCTION SYSTEM. NUHEG/CR-35c3 V02: SYSTEMS INTERACTION RESULTS FHOM 1Hc JIGRAPH MATRIX ANALYS1S 3F A NUCLEAR PonER PLANT'S HIGH PRESSURE SAFFTY INJECTION SYSTEM. Volume 2. SANDERS,R.W. NUREG/CR-3874: LONG-LIVEu ACTIVATION PRODUCTS IN REACTOR MATERIALS. SCHERPELZ,H.I. NUREG/CR-3544: BETA PANTICLE MEASUREMENT AND 00SIMETHY AT NRC-LICENSED FACILITIES. SCHULT,D.A. NUREw/CR-3469 V01: OCCUPATIONAL DOSE REOUCTION AT NUCLEAR P0nER PLANTS ANNOTATED BIBLIOGRAPHY OF SELECTED READINGS IN RADIATION PROIELTION AND ALARA. SCHULTEN,C.S. NUREG-1092 ENVIRONMENTAL ASSESSMENT FOR 10 CFH PART 72, " LICENSING REQUIREMLNTS FOR THE I.NDEPLNDENT STORAGE OF SPENT FUEL AND HIGH-LEVEL RADI0 ACTIVE WASTE
- SCHWARTZ.M.H.
NUREG/CR-3826: RECOMMENDATIONS FOR PHOTECTING AGAINST FAILURE oY BRITTLE FRACTURE IN FEHRATIC STEEL SHIPPING CONTAINENS GREATER inAN FOUR INCHES THICK. SCHWARZ,n.H. NUREG-0935: ACQUSTIC HAVL PROPAGATION IN FLUIDS WITH COUPLED CHEMICAL REACTIONS. SCHWENK,E.B. NUREG/CR-3843: STEAM GLNERATOR GROUP PROJECT TASK 10 - SECONDARY SIDE EXAMINATION. SCOTT,T.G. 86
NUREG/CR-3590: EVALUATION OF ISOTOPE DILUTION MASS JPECTROMETRY FOR BI0 ASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. SEGE,G.. NUREG-0933 801: A PRIORITIZATION OF GENERIC SAFETY ISSUES. SEXTON,D. NUREG/CR-393T: WATER HAMMEM, FLOW INDUCED VIBRATION AND SAFETY / RELIEF VALVE LOADS. SHACK,n.J. NUREG/CR-3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT-nAIER-REACTOR SAFETY RESEARCH PROGRAM: Quarterly Progress Report, January-March 1983 NUREG/CR-3689 V02: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT-WATER REACTOR SAFETY RESEARCH PROGRAM.Guarterly Progress Report, April-June 1983. NUREG/CR-3606: ENVIRONMENTALLY ASSISTED CRACnING IN LIGHT HATER REACTORS: Annual Report,0ctooer 1982 - September 1983, 3HIV,K. NOREG/CR-3493: A REVIEn UF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. SIEGEL,J.J NUREG/CR-3899: UTILITY FINANCIAL STABILITY AND THE AVAILABILITY OF FUNDS FOR DECOMMISSIONING. SILKER,W. NUREG/CR-3474: LONG-LIVED ACTIVATION PRODUCTS IN REACTUR MATERIALS. SILVER,E.G. NUNEG/CR-4011: THE 21/55 DATA BASE USER'S MANUAL. SIMAN-TOV,I. NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION GUARTERLY PROGRESS REPORT, October-December 1983. SIMAN-TOV,M. NUREG/CR-3139: SCENARIOS AND ANALYTICAL METHODS FOR UF6 RELEASES AT NRC-LICENSED FUEL CYCLE FACILITIES. SIMONEN,F.A. NUREG/CR-3869: ANALYSIS OF THE IMPACT OF INSERVICE INSPECTION USING A PIPING RELIABILITY MODEL. SIMONIS,J.C. NUREG/CR-3802 V01: A RESEARCH PROGRAM FOR SEISHIC QUALIFICATION uF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey Of Methods For Equipment Ana Components Evaluation Of Methodology; Qualification Ano Metnodology.... SIMPSON,H.J. NUREG/CR-3940: FIELD EAPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN ALKALINE LAKE ENVIRONMENTS. SINWELLeB.R. NUREG/CR-3654: PWR FLECHT SEASET SfSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14 SISKIND,B. NUREG/CR-3844: CHARACTERIZATION OF THE RADIDACTIVE WASTE PACKAGES OF THE MINNESOTA MINING AND MANUFACTURING COMPANY. SJOREEN,A.L. NUREG/CR-3832: UNCERTAINTIES IN LONG-TERM REPOSITORY PERFORMANCE DUE TO l THE EFFECTS OF FUTURE 4E0 LOGIC PROCESSES. l SMITH,F.W. NUREG/CR-3821: EVALUATION UF CRACK PLANE EQUILIBRIUM MUDEL FOR i PREDICTING PLASTIC FRACTURE. 300,P. NUREG/CR-2482 V05: REVIEn OF 00E WASTE PACKAGE PROGRAM. Subtask 1.1 - National Weste 'ackage Program, April 1983 - September 1983. 87 ....~ . ~
SORENSEN,J.H. NUREG/CR-3524: ORGANIZATIONAL INTERFACE IN REACTOR EMERGENCY PLANNING AND RESPONSE. STAHL,0 NUREG/CR-3900 V01: LONG-lERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACAAGING.First Quarterly Report, Year Three, April-June 1984 STALLMAN,F.W. NUREG/CR-3815: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORR-PSF-PVS IRRADIATION EXPERIMENT. STEPHENS,A.G. NUREG/CR-2576: BhR FULL INTEGRAL SIMULATION TEST (FIST)== Facility Description Report. STEVERSON,J.A. NUREG/CR-3824: CONTING PROGRAM GUIDE. STEnART,0 L. NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. STITT,0.D. NUREG/CR-3761: RELAPS THtRMAL-HYDRAULIC ANALYSES OF PRESSURIZE 0 IHERMAL SHOCK SEQUENCES FOR THE UCONEE-1 PRESSURIZED WATER REACTOR. STOKELY,J.R. NUREG/CR-3590: EVALUATION UF ISOTOPE DILUTION MASS SPECTROMETRY Fog BIDASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. ~ SUTHERLAND,W.A. NUREG/CR-3711: BnR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST RESULTS. SUITER,S.L. NUREG/CR-3796 EMERGENCY PREPAREDNESS SOURCE TERM DEVELOPMENT FOR THE OFFICE OF NUCLEAR MATENIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES. ShARTZMAN,G.L. NUREG/CR-3896: SIMULATION LXPERIMENTS COMPARING ALTERNATIVE PROCESS FORMULATIONS USING A FACT 0 DIAL DESIGN. NUREG/CR-3897: EVALUATION OF ECOSYSTEM SIMULATION MODELS AS TOOLS FOR ASSESSMENT OF POWER PLANT IMPACTS ON FISH POPULATIONS. Final Rept. SWEENEY,F.J. NUREG/CR-2996: SENSITIVITY OF DETECTING IN-CORE VIBRATIONS AND BOILING IN PRESSURIZED WATER REACTURS USING EX-CURE NEUTRON NOISE. SWENSON,0.V. NUREG/CR-3369: AN UNCERTAINTY STUDY OF PnR STEAM EXPLOSIONS. SWYLER,K.J. NUREG/CR-3812: ASSESSMENI UF IRRADIATION EFFECTS IN RAUWASTE CUNTAINING ORGANIC ION-EXCHANGE MEDIA. TANG,P.K. AUREG/CR-3735: ACCIDENT-INDUCED FLOW AND WATERIAL TRANSPORT IN NUCLEnR FACILITIES--A LITERATUNE REVIEW. TANNER,J.E. AUREG/CR-3610: NEUTRON DUSIMLTRY AT COMMERCIAL NUCLEAR PLANTS: Final Report Of Subtask C: 3de Neutron Spectrometer. THATCHER,D. NUREG-0933 S01: A PRIORITIZATION OF GENERIC SAFETY ISSUES. THOMAS,C.W. NUREG/CR-3474: LONG-LIVED ACTIVATION PRODUCTS IN REACTOR MATERIALS. THOMAS,W.R. NUREG/CR-3840: COST ANALYSIS FOR POTENTIAL MUDIFICATIONS TO ENHadCE THE ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKOUT. THOMPSON,S.L. NUREG/CR-3820 V01: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Uuarterly Report, January-March 1984 88 r
TOBIAS,M.L. NUREG/CR-3830 V01: AEROSOL RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR UCIOSEN 1983 - MARCH 1984 L TODD,R.C. l NUREG/CR-3750: JOB ANALYSIS OF NUCLEAR power REACTOR HEALTH PHYSICS TECHNICIANS. i TORREY,M.D. l NUREG/CR-3822: SOLA-PT6: A Transient,Three-Dimensional Algorithm For Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometries.- [ TOSTE,A.P. -NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LUW-LEVEL WASTE INCINERATOR ASH. TREAT,R.L.. NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN nASTE FORMS CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. l TRIER,R.M. NUREG/CR-3940: FIELD EAPLRIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN ALKALINE LAKE ENVIRONMENTS. T*0J0VSKY,M. NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED INSTRUMENTATION AND CONTROL COMPONENTS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. NUREG/CR-3867: DATA SUMMARIES OF LICENSEE EVENT REP 0GTS OF INVERTENS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS, JANUARY 1,1976 TO DECEMdER 31,1982. TZANOS,C.P. 4 NUREG/CR-3933: RISK RELATED RELIABILITY REQUIREMENTS FOR BnR SAFETY -IMPORTANT SYSTEMS WITH EMPHASIS ON THE RESIDUAL HEAT REMOVAL SYSTEM. UFFER,R.A. NUREG/CR-3939: WATER HAMMEH, FLOW INDUCED VIBRATION AND SAFETY / RELIEF VALVE LOADS. UNRUH,J.E. NUREG/CR-3892 VOI: A RLSLANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAN PLANT ELECTRICAL AND MECHANCIAL EQUIFMENT.lask 1 - Survey of Methods For Equipment And Components: Evaluation of Methodology 3 Qualification And Methcdology.... VALANDANI,P. NUREG/CR-3939: WATER HAMMER, PLUM INDUCED VIBRATION AND SAFETY / RELIEF VALVE LOADS. VAN TUYLE,G.J. NUREG/CR-3765: MINET SIMULATION OF A HELICAL COIL SODIUM / WATER STEAM + GENERATOR, INCLUDING STHUCTURAL EFFECTS. NUREG/CR-3813: MINET VALIDATION STUDY USING STEAM GENERATOR TRANS!LNT DATA. VANDER MOLEN,H. NUREG-0933 S01: A PRIORIIIZATION OF GENERIC SAFETY ISSUES. WALKER,R.L. l NUREG/CR-3590: EVALUATION UF ISOTOPE DILUTION MASS SPLCTROMETRY FOR BIDASSAY MEASUREMENT OF URANIUM,PLUT0NIUM,AND THORIUM IN URINE. WALSH,M.E. NUREG/CR-3739: TME OPERATOM FEEDBACK WORKSHOP:A TECHNIQUE FOR OBTAINING l FEEDBACK FROM GPERATIONS PERSONNEL. j WALTON,M.B. NUREG/CR-3893: LABORATORY STUDIES DYNAMIC RESPONSE OF PROTOTYPICAL PIPING SYSTEMS. WANG,S. NUREG/CR-3654: PNR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse deport No. 14. dATERMAN,M.E. 89
l l NUREG/CR=3761: RELAPS IHERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK-SEQUENCES FOR THE UCONEE-1 PRESSURIZE 0 WATER REACTOR. WATLINGTON,B.E. NUREG/CR-3840: COST ANALYSIS FOR POTENTIAL H031FICATIONS TU ENHANCE THE ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKOUT. nE8ER,C.F. NUREG/CR-3617: NOBLE GAS, IODINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT BR0nNS FERRY. hEISS,A.J. NUREG/CR-2331 V03 N3: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RLSEARCH. Quarterly Progress Report, July-Swptember 1983 NUREG/CR-2331 V03 N4: SAFEIY RESEARCH PROGRAMS SPONSURED BY THE OFFICE OF NUCLEAR REGULATORY MESEARCH. Quarterly Progress Report,0ctober 1 -Decemoer 31,1983 WELLS,J.E. NUREG/CR-2015 V08 PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety MarWins Research Program. WESTSIK,J.H. NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL MASTE INCINERATOR ASH. WHEELER,K.R. NUREG/CR-3842 STEAM GENERATUR GROUP PROJECT TASK 8 - SELECTIVE TUBE UNPLUGGING. NUREG/CR-3843: STEAM GENERATOR GROUP PROJECT TASK 10 - SECUNDANY SIDE EXAMINATION. nHITMORE,H.L. NUREG/CR-3735: ACCIDENT-INDUCED l.On AND MATERIAL TRANSPORT IN NUCLEAR FACILITILS--A LITERATURE REVIEn. NICHNER,R.P. NUREG/CR-3617: NOBLE GAS, IODINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT BR0nNS FERRY. WICKETT,A.J. NUREG/CR-33693.AN UNCENTAINTY STUDY OF PnR SIEAM EXPLUSIONS. nILKERSON,C.L. NUREG/CR-3474: LONG-LIVE 0 ACTIVATION PRODUCTS IN REACTOR MATERIALS. nILLIAMS,M.L. NUREG/CR-3886: ANALYSIb 0F THE VENUS PhR ENGINEERING M0CKur EXPERIMENT -PHASE I SOURCE DISTRIBUTION. WILLIAMS,W.R. NUMEG/CR-3139 SCENARIUS AND ANALYTICAL METHODS FOR UF6 RELEASES A[ hRC-LICENSED FUEL CYCLE FACILITIES. WILSON,J.H. NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-C00 LED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIuENT EVALUATION GUARTERLY PROGRESS REPORT, October-December 1983 WINEGARDENER,n. h0 REG /CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS IN RETAINING FISSION Pn00UCTS. Background Information. WING,J. huREG-1081: POST-ACCIDEN[ GAS GENERATION FROM RADIALYSIS OF ORGANIC MATERIALS. nITTEN,A.J. NUREG/CR-3871: AN OVERVIEh 0F THE UNIFIED TRANSPORT APP 90ACH, h0hG,C.C. AUREG/CR-3690: RELAP5 ASSESSMENT SEMISCALE NATURAL CIRCULAi!UN TESIS S-NC-3,S-NC-4,AND S-NC-8. n00,H.H. hUREG/CR-3660 V02: PR0cAoILilY OF PIPE FAILURE IN THE REACTOR C00LANI ' LOOPS UF *ESTINGh0USE PHR PLANTS. Volume 2: Pipe Failure Induced By 90
l Crcck Gronthe NUREG/CR-3663 V02: PR0dA6ILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGIdEERING PHR PLANTS.Vol 2: Pipe Failure Induced oy Crack Growth. NUREG/CR-3869: ANALYSIS OF THE INPACT OF INSERVICE IhSPECTION USING A PIPING RELIABILITY MODEL. dOOTEN,R.O. NUREG/CR-3988: MARCH 2 (MELTUOWN ACCIDENT RESPONSE CHARACTERIS[ICS) CODE DESCRIPTION AND USERS MANUAL. t0RLEDGE,0.H. NUREG/CR-3662: FUEL-DISRUPTION EXPERIMENTS UNDER HIGH-RAMP =RATL nEATING CONDIT106S. cRIGHT,A.L. NUREG/CR-3617: NOBLE GAS, IODINE, AND CESIU'4 TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT BR0nNS FERRY. CRIGHT,S.A. NUREG/CR-3662: FUEL-DISRuPIIUN EXPERIMENTS UNDER HIGH-RAMP = RATE HEATING CONDITIONS. RYANT,F.J. NUREG/CR-3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON-RASTER SYSTEM. ZALOUDEK,F.R. NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS IN RETAINING FISSION Pn00UCTS.dackground Information. l l 91
l Subject index This index was developed from keywords and word strings in titles and ab-stracts. During this development period, there will be some redundancy, which will be removed later when a reasonable thesaurus has been developed through experience. Suggestions for improvements are welcome. ABAOUS NUREG/CR-3845: PREDICTION OF NONLINEAR STRUCTURAL RESPONSE IN LMFBR ELEVATEO-TEMPERATURE PIPING. ALARA NUREG/CR-3469 V01: OCCUPATIONAL DOSE REDUCTION AT NUCLEAR P0nER PLANTS ANNOTATED 6IBLIOGRAPHY OF SELECTED READINGS IN RADIATION PHOTECTION AND ALARA, 4 NUREG/CR-3065 V02: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION ~ PROTECTIuN AT NUCLEAR P0nER PLANTS. Considerations In Factorinn Occupational Dose Into Value-Impact And Cost-Genefit Analyses. AThS NUREG/CR-3470 Aih3 AT BN0nNS FERRY UNIT ONE = ACCIDENT SEuuENCE ANALYSIS. Abnormal Occurrence NUREG-0090 V07 N01: REP 0HT TO CONGRESS ON ABNORMAL OCCURRENCES. January-Maren 1984 Accicent-Induced Flow NUREG/LR-3735: ACCIDENT-INuUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATUNE RLVIEd. Accident NUREG-1080 V01: LONG-RANGE RLSEARCH PLAN FY 1985-1989. NUREG/CR-3139: SCENARIUS AND ANALYTICAL METHODS FOR OF6 RELEASES AT NRC-LICENSED FdEL CYCLE FACILITIES. I NUREG/CR-3591 V01: PRECUdSORS TO POTENTIAL SEVERE CORE DAM. AGE l ACCIDENTS: 1980-1481 A Status Report. i NUREG/CR-3591 v02: PRECURSuRs TO POTENTIAL SEVERE CORE DAMAGE i ACCIDENTb 1980-1981 A Status Report. l NUREG/CR-3617: NOBLE GAS,IuDINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF OECAY NEAT REMUVAL ACCIDENT AT BROWNS FERRY. NUREG/CR-3776: TESl!NG OF SAFETY-RELATED NUCLEAR POWER PLANT EQUIPMENT AT THE CENTRAL RECEIVER TEST FACILITY. l NUREG/CR-3868: CONTAINHE.vT Bu!LDING ATMOSPHERE RESPONSES DuE TO REACTOR GAS 8URNING UNDER SEVEnE ACCIDENT CONDITIONS. NUREG/CR-3988: MARCH 2 (MELTOOHN ACCIDENT RESPONSE CHARACTERISTICS) CODE DLSCRIPTION AND UdERS MANUAL. Actinice NUREG/CR-3763: REVIEW AND ASSESSMENT OF RADIONUCLIDE SURPTION INFORMATION FOR THE uASALT WASTE ISOLATION PROJECT SITE (1979 Through May,1983). Aerosol t l 93 l
t I NUREG/CR-3830 V01: AERUSUL RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR UC100EH 1983 - MARCH 1984 Agenda NUREG=0936 V03 NO2: NRC REGULATORY AGENDA. Quarterly Report, April-June 1984 Aging NUREG/CR-3818: REPORT OF RtSuLTS OF NUCLEAR POWER PLANT AGING WORKSHOPS. Air Sampler NUREG/CR-3513: MECHANICAL HELIABILITY EVALUATION OF ALTERNATE MOTORS FOR USE IN A RADIO 100!NE AIR SAMPLER. Air Storage NUREG/CR-3708: LnR SPEaT FUEL DRV STORAGE BEHAVIOR AT 229 C. Airborne Radionuclides NUREG/CR-3870: RADIATION DUSL ESTIMATES AND NAZARD EVALUATIONS FOR INHALED AINB0RNE RADIONUCLIDES. Annual Progress Rept July 1982 -June 1983 Atmospnere NUREG/CR-3868: CONTAINMENT BUILDING ATMOSPHERE RESPONSES DUE TU REACTOR GAS BURNING UNDER SEVERE ACCIDENT CONDITIONS. BIBELOT NUREG/CR-3951: INTRODUCTION TO BIBELOT A BIBLIOGRAPHIC FINDING AND RETRIEVAL SYSTEM. BW1P NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPE 0 dY DOE HIGH-LEVEL NUCLEAR AASTE REPOSITONY SITE PROJECTS. Report for uctot]r-December 1983. Basalt Haste Isolation Project NUREG/CR-3851 V01: PR0bRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPE 0 dY DOE HIGH-LEVEL NUCLEAR nASTE REPOSITOHY SITE PROJECTS. Report for UctoDer-December 1983 Basalt NUREG/CR-3763: REVIEn ANU ASSESSMENT OF RADIONUCLIDE SURPTION INFORMATION FOR THE oASALT WASTE ISOLATION PROJECT SITE (1979 Through May,1983). Bibliographic Storage NUREG/CR-3951: INTRODUCTION TO BIBELOT;A BIBLIOGRAPHIC FINDING AND RETRIEVAL SYSTEM. Bioassay NUREG/CR-3346: BIDASSAY UATA AND A RETENTION-EXCRETIUN MUDEL FOR SYSTEMIC PLUTONIUM. NUREG/CR-3590: EVALUATION UF ISOTOPE DILUTION MASS SPECTROMETRY FOR BI0 ASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. Biodegradation NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND BITUMEk nASTE F0HMS CONTAINING SIMULATED Lun-LLVEL WASTE INCINERATOR ASH. Bitumen Waste NUREG/CR-3798 CHARACTLRIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LUW-LLVEL WASTE INCINERATOR ASH. Bolting Failure NUREG/CR-2331 V03 N3: SAFETY RESEARCH PROGRANS SFONSURED BY OFFICE OF NUCLEAR REGULATORY DESEARCH. Quarterly Progress Report, July-September 1983. Boric Acid NUREG/CR-3834: ON THE THHESHOLD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRACKING OF SENSITIZED ALLOY 600 IN RORATED THIOSULFATE SOLUTION. drittle NUREG/CR-3826: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE bY BRITTLE FRACTURE IN FEHRITIC STEEL SHIPPING CONTAINERS GREATER THAn 94
FOUR INCHES THICK. Buckling NUREG/CR-3742: BUCKLING OF STEEL CONTAINMENT SHELLS UNDER TIME = DEPENDENT LOADING. Bundle B-5 NUREG/CR-3459: EXPERIMENT DATA REPORT FOR MULTIR00 BURST TEST (MRBT) BUNDLE B-5 COBOL NUREG/CR-3951: INTRODUCTION TO BIBELOT A BIBLIOGRAPHIC FINDING AND RETRIEVAL SYSTEM. CONTEMPT 4 NUREG/CR-4901: CONTEMPT 4/MODb AN IMPROVEMENT TO CONTEMPT 4/ MOD 4 NULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. CONTING NUREG/CR-3824: CONTING PROGRAM GUIDE. Cement NUREG/CR=3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. Cladding s-NUREG/CR-3459: EXPERIMENT DATA REFORT FOR MULTIROD BURST TEST (MRBT) BUNDLE B-5 huREG/CR-3460: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT) SUNDLE B-6. NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANuuAL ~ PROGRESS REPORT FOR OCT0 DER 1983 - MARCH 1984 Code NUREG/CR-2331 V03 N3 SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report, July-September 1983. NUREG/CR-3618: OCA-P,A DLTERMINISTIC AND PROBABILISTIC FRACTURE-MECHANICS CODE FOR APPLICATION TO PRESSURE VESSELS. NUREG/CR-3690 RELAPS ASSESSMENT SEMISCALE NATURAL CIRCULATION TESTS S=NC-3,S=NC-4,AND S-WC-8 NUREG/CR-3845: PREDICTION OF NONLINEAR STRUCTURAL RESPONSE IN LMFBR ELEVATED-TEMPERATURE PIPING. NUREG/CR-3907: GT2R2 AN UPDATED VERSION OF GAPCON-THERMAL-2 NUREG/CR-4001: CONTEMPT 4/MODbs/N IMPROVEMENT TO CONTEMPT 4/ MOD 4 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. Combustion NUREG/CR-3868: CONTAINMENT BUILDING ATMOSPHERE RESPONSES DUE TO REACTOR i GAS BURNING UNDER SEVERE ACCIDENT CONDITIONS. l Components j NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED I INSTRUMENTATION AND CONTHOL COMPONENTS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. Computer Code NUREG-1029: A COMPUTER CODE FOR GENERAL ANALYSIS OF RADON RISKS (GARR). NUREG/CR-3190: PLUGM A COUPLED THERMAL = HYDRAULIC COMPUTER MODEL FOR FREEZING MELT FLOW IN A CHANNEL. NUREG/CR-3079: CALIBRATIUN AND QUALIFICATION OF THE LOS ALAMOS FAILURE MODEL (LAFM). NUREG/CR-3134: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982 - March 1983. NUREG/CR-3761: RELAPS IHERMAL-HYDRAULIC ANALYSES OF PRESSURIZED ThtRMAL SHOCK SEWUENCES FOR THL UCONEE-1 PRESSURIZED WATER REACTOR. NUREG/CR-3765: MINET SIMULATION OF A HELICAL COIL S00IUM/ WATER STEAM GENERATOR, INCLUDING STRUCTURAL EFFECTS. NUREG/CR-3776: TESTING OF SAFETY-RELATED NUCLEAR POWER PLANT EQUIPMENT t N
AT THE CENTRAL RECEIVER TEST FACILITY. NUREG/CR-3813 MINET VALIDATION STUD-Y USING STEAM GENERATOR TRANSIENT DATA. NUREG/CR-3820 vois THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM. Quarterly Report, January-March 1984 NUREG/CR-3822: SOLA-PTS: A Transient,Three= Dimensional Algorithm For 1 Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometries. Computer Model NUREG/CR-3835: SIMULATION UF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE DISCHETE VORTEX METHOD. Computer NUREG/CR-3824: CONTING PH0 GRAM GUIDE. Con 9ress NUREG-0090 V07 N01: REFONT Tu CONGRESS ON ABNORMAL 1 OCCURRENCES. January-March 1984 Construction Deficiency NUNEG/CR-4011: THE 21/55 DATA BASE USER'S MANUAL. Containment NUREG-0800 06.2.1 R6: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition Revision 6 to Section 6.2.1.1.C, " Pressure-Suppression Type SWR Containments." NUREG/CR-3053: CLOSEQUT OF IE BULLETIN 80-08 EXAMINATION OF C04TAINMENT LINER PENETRATION nELDS. NUREG/CR-3610 NEUTRON DUSIMtTRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of JuDtask C 3ne Neutron Spectrometer. NUREG/CR-3742: BUCKLING uF STEEL CONTAINMENT SPELLS UNDER TIME-DEPENDENT LOADING. NUREG/CR=3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS IN RETAINING FISSION PHODUCTS. Background Information. NUREG/CR-3795: CLOSEOUT UF IE BULLETIN 82-04: DEFICIENCIES IN PRIMANY CONTAINMENT ELECTRICAL PENETRATION ASSEMBLIES. NUREG/CR-3868: CONTAINMENT BUILDING ATMOSPHERE RESPONSES DUE TO REACTOR GAS BURNING UNDER SEVEHE ACCIDENT CONDITIONS, NUREG/CR-3988: MARCH 2 (MELT 00WN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USERS MANUAL. NUREG/CR-4001: CONTEMPT 4/ MUDS AN IMPROVEMENT TO CONTEMPT 4/M004 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. Control Room Habitability NUREG/CR-3786: A REVIEW UF REGULATORY REQUIREMENTS GOVERNING CONTROL ROOM HABITABILITY SYSTEMS. Coolant Dynamic Model NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYSTEM OF THE EXPERIMENTAL BREEDER REACTUR-II. Core Catcher i i NUNEG/CR-3190: PLUGM A COUPLED THERMAL-HYDRAULIC COMPUTER MODEL FOR FREEZING MELT FLon IN A CHANNEL. Core Damage NUREG-1068: REVIEW INSIGHTS ON THE PROBA8ILISTIC RISK ASSESSMENTS FOR THE LIMERICK GENERATING STATION, UNIT 1 AND 2 NUREG/CR-3591 V01: PRECUdSQRS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECUNSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. Core Di'sruptive Accident Analysis NUREG/CR-3804 V01: PHYSICS OF REACTOR SAFETY. Quarterly Report January - March 1984 Core Melt NUREG/CR-3734: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982 - March 1983. 96
NUREG/CR-3988: MARCH 2 (MELT 00nN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USERS MANUAL. Core-Melt NUREG/CR-3493: A RE4IEW OF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS IN RETAINING FISSION PH00UCTS.6ackground Information. Corrosion-NUREG/CR-3798: CHARACTERIZATION OF CEMENT AND SITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL-WASTE INCINERATOR ASH. Cost-Benefit NUREG/GR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonEH PLANTS. Executive Summary. NUREG/CR-3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS.A Review Of Occupational Dose Assessment C6nsiderations In Current Probacilistic Risk Assessment And. Cost-Benefit Analyses. NUREG/CR-3665 V02: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION i PROTECTION AT NUCLEAR PonER PLANTS. Considerations In Factoring Occupational Dose Into Value-Impact And Cost-Benefit Analyses. NUREG/CR-3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nEH PLANTS.A Calculation Nethod. Cost NUREG/CR-3751: EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FL0uD PROTECTION COSTS FOR UHANIUM TAILINGS IMPOUNDMENTS. l-NUREG/CR-3840 COST ANALYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKOUT. Crack Growth NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACAING IN LIGHT WATER REACTORS: Annual Reportauctober 1982 - September 1983 Crack Plane Equilibrium i NUREG/CR-3821: EVALUATION OF CRACK PLANE EQUILIBRIUM MODEL FOR PREDICTING PLASTIC FRACTURE. Crack 4 NUREG/CR-3d28 V02: STRUCTUHAL INTEGRITY OF WATER REACTOR PRESSURE 80UNDARY COMPONENTS. Annual Report For 1983 NUREG/CR-3663 V02: PR0dAdILIIY OF PIPE FAILURE IN THE REACT 0H CoulANT LOOPS OF COMBUSTION tNGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced by Crack Growth. DAVDs NUREG/CR=0169 V17: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recordec On DAVDS. i DECON NUREG/CR-0130 ADD 03: TECHNOLuGY, SAFETY AND COSTS OF OECOMMISSIONING A REFERENCE PRESSURIZEU nAIEH REACTOR POAER STATION. i, NUREG/CR-0672 A0002: TtCnNULUGY, SAFETY AND COSTS OF DECOMMISSIONING A l REFERENCE BOILING WArEN HEACTOR POWER STATION. Classification Of I Decommissioning Wastes. Data Acquisition and Visual Oisplay System l NUREG/CR-0169 V17: LOFI LXPENIMENTAL MLASUMEMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recordec On DAVDS. Data Acquisition NUREG/CR-3684: EVALUATION OF NUCLEAR FACILITY OECOMMISSIONING PROJECTS l PROGRAM - THREE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. l _ Decay Heat Removal i NUREG/CR-3017: NOBLE GAS,IuDINL,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF UECAY HEAT REMUVAL ACCIDENT AT BROWNS FERRY. 4 i Decentralization l NUHEG-1075: DECENTRALIZATIUN OF OPLRATING REACTOR LICENSING REVIEWS.NRR f Pilot Program. 1 97
I Decision Making NUREG/CR-3655: A METHOD FOR ANALYTICAL EVALUATION OF COMPUTER-BASED DECISION AIDS. 1 Decommissioning NUREG/CR-0130 ADD 03: TECHNULUGY, SAFETY AND COSTS OF DECOMMISSIUNING A REFERENCE PRESSURIZE 0 nATER~ REACTOR POWER STATION. NUREG/CR-0672 ADD 02: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR PonER STATION. Classification Of Decommissioning Wastes. NUREG/CR-3884: EVALUATIDH UF NUCLEAR FACILITY DECOMMISSIONING PROJECIS PROGRAM - THREE MILE ISLANO UNIT 2, POLAR CRANE RECOVERY. NUREG/CR-3899 UTILITY FINANCIAL STABILITY AND THE AVAILABILITY OF FUNDS FOR DECOMMISSIONING. Decontamination NUREG/CR-3884: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM - THREE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. Defective Fuel NUREG/CR-3708: LWR SPENT FUEL DRY STORAGF.'8EHAVIOR AT 229 C. Deformation NUREG/CR-3459: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT) BUNDLE B-5 NUREG/CR-3460: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT) BUNDLE B-6 Degradation NUREG/CR-3643: HETEROGENEOUS OXIDATIVE DEGRADATION IN IRRADIATED POLYMERS. Design Basis Event NUREG/CR=3418 SCREENING TESTS OF 1ERMINAL BLOCK PERFORMANCE IN A SIMULATED LOCA ENVIRONMENT. Design Flood Event NUREG/CR-3751: EFFECTS OF ROCK RIPRAP DESIGN PARAMETERS ON FLOOD PROTECTION COSTS FOR URANIUM TAILINGS IMPOUNDMENTS. Diffusion Theory NUREG/CR-2996: SENSITIVITY OF DETECTING IN= CORE VIBRATIONS AND BOILING IN PRESSURIZED WATER REACTORS USING EX= CORE NEUTRON NOISE. Digraph-Matrix Analysis NUNEG/CR-3593 V01: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR P0nEM PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. NUREG/CR-3593 V02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR P0nER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2. Discrete' Vortex Method NUREG/CR-3835: SIMULATION OF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE DISCHETE VORTEX METHOD. Disposal NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONMENTAL RADIATION STAHDARDS FOR GEOLOGIC DISPOSAL OF HIGH-LEVEL RADI0 ACTIVE WASTES. Dose l NUREG/CR-3469 V01: OCCUPATIONAL DOSE REDUCTION AT NUCLEAR P0nER PLANTS ANNOTATED BIBLIOGRAPHY OF SELECTED READINGS IN RADIATION PROTECTION AND ALARA. NUREG/CR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonER PLANTS. Executive Summary. NUREG/CR-3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION i PROTEC(ION AT NUCLEAR PonER PLANTS.A Review Of Occupational pose-Assessment Considerations In Current Probabilistic Risk Assessment And Cost-Benefit Analyses. NUREG/CR-3665 V02 OPTIM1ZATION OF PUBLIC AND UCCUPATIONAL RADIATION PROTECTION AT NUCLEAN P0nEH PLANTS. Considerations In Factoring 98
l Occupational Dose Into Value-Impact And Cost-Benefit Analyses. NUREG/CR-3665 V03 OPTIMIZATION OF PUBLIC AND UCCUPATIONAL RADIATION PROTECTION AT NUCLEAR POWEN PLANTS.A Calculation Method. NUREG/CR-3870 RADIATION DOSE ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Rept July 1982 -June l 1983. Dosimetry NUREG/CR-3318 LnR PRESSURL VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM:PCA Experiments,dlind Test,And Physics-Dosimetry Support For The PSF Experiments. NUREG/CR-3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. NUREG/CR-3610: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of Suotask C 3de Neutron Spectrometer. Double-Ended Guillotine dreak NUREG/CR-3663 V02: PRooAoILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced by Crack Growth. Draft Environmental Statement NUREG-1064 DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF MILLSTONE NUCLEAR PonER STATION, UNIT 3.00CKET NO. 50-423 (NORTHEAST NUCLEAR ENERGY COMPANY,et al) NUREG-1073: DRAFT ENVIHONMLNTAL STATEMENT RELATED TO THE OPERATION OF RIVER BEND STATION. Docket No. 50-458.(Gulf States Utilities Company & Cajun Electric Power Cooperative) NUREG-1085: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF NINE MILE POINT NUCLEAH STATION, UNIT NO.2. Docket No. 50-410 (niagara Mohawk Power Corporation, Rochester Gas & Electric Corporation Ana Central Hudson Gas & Electric Corporation) Dry Spent Fuel NUREG/CR-3921: DRY SPENT FUEL STORAGE TEST PLAN FOR FINAL NONDESTRUCTIVE FUEL R00 EXAMINATION. Dry Storage NUREG/CR-3708: LnR SPENT FUEL DRY STORAGE BEHAVIOR AT 229 C. ESF NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS IN RETAINING FISSION PN0 DUCTS. Background Information. Ecosystem NUREG/CR-3897: EVALUATION UF ECOSYSTEM SIMULATION MODELS AS TOOLS FOR ASSESSMENT OF POWER PLANT IMPACTS ON FISH POPULATIONS. Final Rept. Effluent Measurement NUREG/CR-4007: LOWER LIMIT OF DETECTION DEFINITION AND FLABORATION OF A PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL l MEASUREMENTS. Electrical Penetration NUREG/CR-3795: CLUSE0VT OF IL BULLETIN 82-04 DEFICIENCIES IN PRIMARY l CONTAINMENT ELECTRICAL PLNLTRATION ASSEMBLIES. Electron Beam Accelerator j NUREG/CR-3777 CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON-RASTER SYSIEM. Emorittlement NUREG/CR-3228 V02: STRUCIUHAL INTEGRITY OF WATER REACTOR PRESSURE BOUNDARY COMPONENTS. Annual Report For 1983 l NUREG/CR-3689 V02: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT-WATER REACTOR SAFETY RESEARCH PROGRAM.Guarterly Progress Report, April-June 1983. Emergency Core Cooling NUREG/CR-3895: INVESTIGATIUN OF COLD LEG WATER HAMMER IN A PnR DUE TO THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. Emergency Planning i l 99
i NUREG/CR-3524: ORGANIZATIONAL INTERFACE IN BEACTOR EMERGENCY PLANNING AND RESPONSE. Emergency Preparedness NUREG/CR-3196: EMERGENCY PNEPAREDNESS SOURCE TERM DEVELOPMENT FOR THE OFFICE OF NJCLEAR MATERIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES, Emergency Response NUREG/CR.3513: MECHANICAL RELIABILITY EVALUATION OF ALTERNATE MOIORS FOR USE IN A RADIO 100!NE AIR SAMPLER, Enforcement Action NUREG-0940 V03 N02: ENFORCEMLNT ACTIONS:SIGNIFICANT ACTIONS RESOLVED. Quarterly Progress Report, April-June 1984 Engineered Safety Feature NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTLMS IN RETAINING FISSION PRODUCTS.8ackground Information. Engineered Safety System NUREG/CR=3792: CLOSEQUT OF IE SULLETIN 79-11: FAULTY OVERCURRENT TRIP DEVICE IN CIRCUIT BREAnERS FOR ENGINEERED SAFETY SYSTEMS. Environmental Measurement NUREG/CR-4007: LOWER LIMIT OF DETECTION: DEFINITION AND ELABORATION OF A PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENv!RONMENTAL MEASUREMENTS. Environmental Radiation Standards NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONMENTAL RADIATION STANDAROS FOR GEOLUGIC DISPOSAL OF HIGH-LEVEL RADI0 ACTIVE WASTES. Ergonnmics NUREG-0985 R01: U.S. NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN. Evaluation NUREG/CR-3655: A METHOD FOR ANALYTICAL EVALUATION OF COMPUTER =dASE0 DECISION AIDS. Event Tree NUREG/CR=3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. NUREG/CR-3591 V02: PRECURSUR3 TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. Ex-Core NUREG/CR=2996: SENSITIVITY OF DETECTING IN= CORE VIBRATIONS AND BOILIN3 IN PRESSURIZED WATER REACTURS USING EX= CURE NEUTRON NOISE. Exposure NUREG/CR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS, Executive Summary. NUREG/CR=3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS.A Review Of Occupational Dose Assessment Considerations In Current Probabilistic Risk Assessment And Cost-Benefit Analyses. NUREG/CR-3665 v02: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PohER PLANTS. Considerations In Factoring Occupational Dose Into Value= Impact And Cost-Benefit Analyses. NUREG/CR-3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RAOIATION PROTECTION AT NUCLEAR POWER PLANTS.A Calculation Methoa. NUREG/CR-3870: RADIATION DOSE ESTIMATES AND HAZARD EVALUATIONS FL: INHALED AIRBORNE RADIONUCLIDES. Annual Progress Rept July 1982 -June 1983 FAST NUREG/CR-3830 V01: AEROSOL RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR QCTOWER 1983 = MARCH 1984 FIST NUREG/CR=2576: BMR FULL INTEGRAL SIMULATION TEST (FIST)--Facility Description Report. 100
NUREG/CR=3711: 8WR FULL INTEGRAL SIMULATION' TEST (FIST) PHASE I TEST RESULTS. -FLECHT l' NUREG/CR-3654: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14 FORTRAN-77 NUREG/CR=39073,GT2R2 AN UPDATED VERSION OF GAPCON-THERMAL-2. Fcctorial Experiments NUREG/CR-3896 SIMULATION EXPERIMENTS COMPARING ALTERNATIVE PROCESS FORMULATIONS USING A FACTORIAL DESIGN. Fo11ure NUREG/CR-3369: AN UNCERTAINTY STUDY OF PnR STEAM EXPLOSIONS. NUREG/CR-3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. NUREG/CR-3591 V02: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980=1981 A Status Report. NUREG/CR-3593 V01: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX 4 ANALYSIS OF A NUCLEAR P0nER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. NUREG/CR-3593 V02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR POWER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2. NUREG/CR-3663 V02: PR00AulLITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS UF COM8USTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced by Crack Growth. NUREG/CR-3826: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE BY BRITTLE FRACTURE IN FERRITIC STEEL SHIPPING CONTAINERS GREATER THAN FOUR INCHES THICK. NUREG/CR-3867: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTEMS AT U.S. COMMERCIAL NUCLEAR PohER PLANTS, JANUARY 1,1976 TO DECEHdER 31,1982. Fotigue NUREG/CRa3788 V01: STRUCTURAL INTEGRITY OF LIGHT WATER REACTOR PRESSURE BOUNDARY COMPONENTS.Four= Year Plan 1984-1988 Fcult Rates NUREG/CR=1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED INSTRUMENTATION AND COHTHOL COMPONENTS AT U.S. COMMEdCIAL NUCLEAR POWER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. Forritic Steel NUREG/CR-3826: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE dY BRITTLE FRACTURE IN FERRITIC STEEL SHIPPING CONTAINERS GREATER THAN l FOUR INCHES THICK. Filter System l NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS l IN RETAINING FISSION PRouuCTS.8ackground Information. Financial Assurance NUREG/CR-3899: UTILITY FINANCIAL STABILITY AND THE AVAILABILITY OF FUNDS FON DECOMMISSIONING. I Fish Populations NUREG/CR-3897 EVALUATION UF ECOSYSTEM SIMULATION MODELS AS TOULS FOR ASSESSMENT OF PunER PLANT IMPACTS ON FISH POPULATIONS. Final Rept. Fission Product Release NUREG/CR=3734: LIGHT WATtR REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982 - Maren 1983. Fission Product i NUREG/CR-3617: NOBLE GAS,IUDINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT RLMOVAL ACCIDENT AT 8H0nNS FERRY. I NUREG/CR-3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT-nATER-REACTOR SAFETY RESEARCH PROGRAM Quarterly Progress l 101
Reporta. January-March 1983 NUREG/CR-3689 V02: MATtRIALS SCIENCE-AND TECHNOLOGY DIVISION LIGHT-WATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress i Report, April-June 1983. NUREG/CR-3o89 V03: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT-HATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress ~ Report, July-September 1983 NUREG/CR-3763: REVIEn ANU ASSESSMENT OF RADIONUCLIDE SURPTION INFORMATION FOR THE dASALT WASTE ISOLATION PROJECT SITE (1979 Througn May,1983). NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTEMS IN RETAINING FISSION PROUUCTS. Background Information. Five-Year Research j NUREG-1080 V01: LONG-RANGE RESEARCH PLAN FY 1985-1989, i l Flame Propagation NUREG/CR=3835: SIMULATION OF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE DISCRETE VORTEX METHOD. Flaw NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR UClodER 1983 - MARCH 1984 NUREG/CR-3826: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE oY SRITTLE FRACTURE IN FENRITIC STEEL SHIPPING CONTAINERS GREATER THAN FOUR INCHES THICK. Flood NUREG/CR-3751: EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FLOOD PROTECTION COSTS FOR URANIUM TAILINGS IMPOUNDMENTS. l Fluia-Thermal Mixing [ NUREG/CR-3822: SOLA-PTS: A Transient,Three-Dimensional Algorithm For Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometrics. ! ~ Fracture Mecnanic j NUREG/CR-3618: OCA-P,A DETERMINISTIC AND PROBABILISTIC FRACTUNE-MECHANICS CODE FON APPLICATION TO PHESSURE VESSELS. i NUREG/CP-0051: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON ( LEAK-BEFORE-BREAK IN NUCLEAR REACTOR PIPING. i NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLUGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR UCI0dEH 1983 - MARCH 1984 Fracture Toughness NUREG/CR-3671: ASSESSMENT OF RADIATION EFFECTS RELATING TO REACTOR PRESSURE VESSEL CLAD 0!NG. i Fracture NUREG/CR-3788 V01: STRUCTUNAL INTEGRITY OF LIGHT WATER REACTOR PNESSURE BOUNDARY CUMPONENTS.Four-Year Plan 1984-1988 NUREG/CR-3826: RECOMMENDATIOWS FOR PROTECTING AGAINST FAILURE SY 6RITTLE FRACTURE IN FENRITIC STEEL SHIPPING CONTAINENS GREATER THAN FOUR INCHES THICK. Freezing Melt Flow NUREG/CR-3190: PLUGM A COUPLED THERMAL-HYDRAULIC COMPUTER MODEL FOR FREEZING MELT FLOW IN A CHANNEL. Fuel Aerosol Simulant' Test i NUREG/CR-3830 V01: AEROSQL RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR UCTOBER 1983 - MARCH 1984 Fuel Expulsion huREG/CR-3679: CALIBRATIUN AND QUALIFICATION OF THE LOS ALAMOS FAILUHE MODEL (LAFM). Fuel Faerication NUREG/CR-3139: SCENARIOS AND ANALYTICAL METHODS FOR UF6 RELEASES AT NRC-LICENSED FUEL CYCLE FACILITIES. Fuel Oxidation NUREG/CR-3708: LnR SPENT FUEL DRY STORAGE BEHAVIOR AT 229 C. Fuel Rods 102
l l NUREG/CR=3921: ORY SPENT FUEL STORAGE TEST PLAN FOR FINAL [ NONDESTRUCTIVE FUEL R0D EXAMINATION.
- i. '
. Fuel Storage NUREG/CR=3921: ORY SPENT FUEL STORAGE TEST PLAN FOR FINAL NONDESTRUCTIVE FUEL ROD LXAMINATION. Puol Thermal Performance 'NUREG/CR=3907: GT2R2 AN UPDATED VERSION OF GAPCON-THERMAL-2. Fuel = Disruption l NUREG/CR=3662: FUEL =0ISRUPTION EXPERIMENTS UNDER HIGh= RAMP = RATE HEATING-CONDITIONS. Full Integral Simulation Test NUREG/CR=2576: BNR FULL INTEGRAL SIMULATION TEST (FIST)== Facility Description Report. NUREG/CR-3711: BAR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST RESULTS. - GAPCON=THENMAL-2 NUREG/CR=3907: GT2R2 AN UPDATED VERSION OF GAPCON= THERMAL-2 GARR NUREG=1029: A COMPUTER CUDE FOR GENERAL ANALYSIS OF RADON RISKS (GARR). GPU y 8&N NUREG=0680 S05: THI-1 RESTART.An Evaluation of The Licensee's Management Integr}ty As It Affects Restart Of Three Mile Island Unit 1 Docket 50-289. GT2R2 NUREG/CR=3907: GT2R2 AN UPUATED VERSION OF GAPCON-THERMAL-2. Gamma Ray NUREG/CR=3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. Gas-Cooled NUREG/CR=3492 V04: HIGH= TEMPERATURE GAS = COOLED REACTOR SAFETY STUDIES .FOR THE DIVISION OF ACCIDENT EVALUATION GUARTERLY PROGRESS REPORT, October = December 1983. 4 Generic Issue B-10 NUREG=0800 06.2.1 R6 STANUARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition Revision 6 to Section 6.2.1.1.C, " Pressure = Suppression Type 8dR Containments.' Geologic Repositories NUREG/CR=3758: CROSSHOLE GLOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL HASTE REPOSITORIES. NUREG/CR=3832: UNCERTAINTIES IN LONG= TERM REPOSITORY PERFORMANCE DuE TO THE EFFECTS OF FUTURE GEULOGIC PROCESSES. Geotomography NUREG/CR=3758: CROSSHOLE GEOPHYSICAL METHODS USED TO INVESTIGATE THE l NEAR VICINITY OF HIGH LEVEL WASTE REPOSITORIES, Glass Waste NUREG/CR=3900 V01: LONG= TERM PERFORMANCE OF MATERIAL.S USED FOR l HIGH= LEVEL WASTE PACKAGING.First Quarterly Report, Year l Three, April-June 1984. l-Groundwater NUREG-1054: SIMPLIFIED ANALYSIS FOR LIQUID PATHWAY STUDIES. Guicance NUREG/CR=3786: A REVIEW UF REGULATORY REQUIREMENTS GOVERNING CONTROL ROOM HABITABILITY SYSTEMS. HECTR NUREG/CR=3734: LIGHT WATLR REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982 = deren 1983. l' NUREG/CR=3776: TESi!NG OF SAFETY =RELATED NUCLEAR P0nER PLANT EQUIPMENT AT THE CENTRAL RECE!VER IEST FACILITY. i HEU NUREG/CR=3678: ESTIMATION METHODS FOR PROCESS HOLOUP OF SPECIAL NUCLEAR 133
MATERIALS, Hazard Evaluation NUREG/CR-3870 RADIATION DOSE ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Rept July 1982 -June 1983. Health Physics Technician NUREG/CR=3750: JOB ANALYSIS OF NUCLEAR POWER REACTOR HEALTH PHYSICS TECHNICIANS. Heat Flux NUREG/CR-3776: TESTING OF SAFETY-RELATED NUCLEAR POWER PLANT EuulPMENT AT THE CENTRAL RECEIVEN IEST FACILITY. Heat Transfer NUREG/CR-3820 V01s THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM. Quarterly Report, January-March 1984. NUREG/CR-3822: SOLA-PTS: A Transient,Three= Dimensional Algorithm For Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometrics. NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYSTEM UF THE EXPERIMENTAL BREEDER REACTOR-II. Heated Feedwater System NUREG/CR-2576: BHR FULL INTEGRAL SIMULATION TEST (FIST)== Facility Description Report. Heavy-Section Steel NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAH SEMIANNUAL PROGRESS REPORT FOR UCT0 DER 1983 - MARCH 1984 Helical Coil NUREG/CR-3765: MINET SIMULATION OF A HELICAL COIL SODIUM / WATER STEAM GENERATOR, INCLUDING STRUCTURAL EFFECTS. High Pressure Safety Injection NUREG/CR-3593 V01: SYSIEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR POWER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. NUREG/CR-3593 V02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH HATRIX ANALYSIS OF A NUCLEAR POWER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2. NUREG/CR-3895: INVESTIGATION OF COLD LEG WATER HAMMER IN A PnR DUE TO THE ADMISSION OF ECC DURING A SMALL SREAK LOCA. High Temperature NUREG/CR-2331 V03 N3: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report, July-September 1983 High-Level Weste NUREG/CR-2482 V05: REVIEn 0F DOE WASTE PACKAGE PROGRAM. Subtask 1.1 - National Weste Package Program, April 1983 - September 1983. NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONHENTAL RADIATION STANDARDS FOR GEOLOGIC DISPOSAL OF HIGH-LEVEL RADI0 ACTIVE WASTES. NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONMENTAL RADIATION STANDAR08 FOR GEOLOGIC DISPOSAL OF HIGH-LEVEL LADI0 ACTIVE WASTES. NUREG/CR-3758: CROSSHOLE GEOPHYSICAL NETHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL M ASTE REPOSITORIES.
- NUREG/CR-3763: REVIEW AND ASSESSHENT OF RADIONUCLIDE SORPTION INFORMATION FOR THE dASALT WASTE ISOLATION PROJECT SITE (1979 Through May,1983).
NUREG/CR-3832: UNCERTAINTIES IN LONG-TERM REPOSITORY PERFORMANCE DUE TO THE EFFECTS OF FUTURE GEOLOGIC PROCESSES. NUREG/CR-3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GE0 CHEMICAL INFORMATION DEVELOPED WY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITONY SITE PROJECTS. Report for Octocer-December 1983 NUREG/CR=3900 V01: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.First Quarterly Report, Year Three, April-June 1984. 104
High=Rocp-Roto H00 tins NUREG/CR-3662: FUEL-DISRUPTION EXPERIMENTS UNDER HIGH RAMP-RATE HEATING CONDITIONS. High= Temperature l NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS = COOLED REACTOR SAFETY STUDIES FOR-THE DIVISION OF ACCIUENT EVALUATION QUARTERLY PROGRESS REPORT, October-December 1983. Highly Enriched Uranium NUREG/CR-3678: ESTIMATION METHODS FOR PROCESS HOLDUP OF SPECIAL NUCLEAR MATERIALS. Human Error NUREG/CR-2331 V03 N3: SAFETY RESEARCH PRUGRAMS SPONSORED BY OFFICE OF NUCLEAR NEGULATORY RESEARCH.Guarterly Progress Report, July-Septem*,e 1983. NUREG/CR-3518 V01: SLIM =MAUDIAN APPROACH TO ASSESSING HUMAN ERHr i. PR08A81LITIES USING STHUCTURED EXPERT JUDGEMENT. Volume Isove view of SLIM-MAUD. Human Factors i NUREG=0985 R01: U.S. NUCLEAR REGULATORY COMMISSION HUMAC cACTORS PROGRAM PLAN. NUREG/CR-2331 V03 N4: SAFETY RESEARCH PROGRAMS '>"'NSr.ED BY THE OFFICE l OF NUCLEAR REGULATORY HESEARCH. Quarterly Prot s ceport,0ctober 1 -December 31,1983 NUREG/CR-3520 V01: LONG=TEMM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR POWER PLANTS. Volume I Summary And Users Guide. NUREG/CR-3520 V02: LONG-TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR POWER PLANTS. Volume II Development Of Getailed i Analyses. i NUREG/CR-3524: ORGANIZATIONAL INTERFACE IN REACTOR ENERGENCY PLANNING AND RESPUNSE. I Hyorogen Comoustion NUREG/CR-3273: COMBUSTION OF HYDROGEN AIR MIXTURES IN THE VGES i CYLINDRICAL TANK. IE Sulletin 79-11 NUREG/CR-3192: CLOSEQUT OF IE BULLETIN 79-11 FAULTY OVERCURRENT [ RIP DEVICE IN CIRCUIT BREAKENS FOR ENGINEERED SAFETY SYSTEMS. i IE Bulletin 80-08 NUREG/CR-3053: CLOSEQUI 0F IE BULLETIN 80-08: EXAMINATION OF CONTAINMLNT LINER PENETRATION nELDS. IE Bulletin 82-04 NUREG/CR-3795: CLOSEQUT UF IE BULLETIN 82-04: DEFICIENCIES IN PRIMARY CONTAINMENT ELECTRICAL PENcTRATION ASSEMBLIES. IGSCC NUREG/CR=3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT nATER REACTORS: Annual Report,0ctober 1982 - September 1983 Ice Containment NUREG/CR-4001: CONTEMPT 4/MUDSIAN IMPROVEMENT TO CONTEMPT 4/M004 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. In-Core NUREG/CR-2996: SENSITIVITY OF DETECTING IN CORE VIBRATIONS AND ROILING IN PRESSURIZED WATER REACTURS USING EX-CURE NEUTRON NOISE. Inhalation NUREG/CR-3870 RADIATION DUSE ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBURNE RADJ0NUCLIDES. Annual Progress Rept July 1982 -June 1983. Inservice Inspection NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS: Annual Report,0ctober 1982 - Se ptember 1983 NUREG/CR-3869: ANALYSIS UP THE IMPACT OF INSERVICE INSPECTION USING A 105
PIPING RELIABILITY MODEL. NUREG/CR-3894: ULTRASONIC AND METALLURGICAL EXAMINATION OF A CRACKED 2 TYPE 304 STAINLESS STEEL BnR PIPE WELOMENT. Inspection a NUREG/CR-3610: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS: Final _= Report Of Subtask C 3de Neutron Spectrometer. NUREG/CR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION lE PROTECTION AT NUCLEAR PonER PLANTS. Executive Summary. ^3 NUREG/CR-3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR POWER PLANTS.A Calculation Method, 2 Instrumentation and Control i NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEf CVENT REPORTS OF SELECTED INSTRUMENTATION AND CONTROL COMPONENTS AT U.S. COMMERCIAL NUCLEAR _j POWER PLANTS JANUARY 1,1976 To DECEMBER 31,1981, 2 Inventory i NUREG-0430 V04 N02: LICENSED FUEL FACILITY STATUS REPORT. Inventory ] Difference Data, July 1983-December 1983.(Buff Book) g = Inverter j NUREG/CR-3667: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTERS AT U.S. COMMERCIAL NUCLEAR PohER PLANTS, JANUARY 1,1976 TO DECEMdER ] 31,1982 3 Ion-Exchange Media 7 NUREG/CR-3812: ASSESSMENT OF IRRADIATIGH EFFECTS IN RADWASTE CONTAINING ORGANIC ION-EXCHANGE ME01A. Iaradiat6d Polymer A NUREG/CR-3643: HETERUGENLOUS OXIDATIVE DEGRADATION IN IRRADIATED POLYMERS. j Irradiation Effect NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADhASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA. Irradiation 4 NUREG/CR-3615: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN j = THE ORR-PSF-PVS IRRADIATION EXPERIMENT. Isotope Dilution Mass Spectrometry NUREG/CR-3590: EVALUATION UF ISOTOPE DILUTION MASS SPECTROMETRY FOR j 3 BIDASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. Joe Analysis i NUREG/CR-3750: JOB ANALYSIS OF NUCLEAR PONER REACTOR HEALTH PHYSICS a7 TECHNICIANS. d 1,AFM NUREG/CR-3679: CALIBRATION AND QUALIFICATION OF THE LOS ALAMOS FAILURE 4 MODEL (LAFM). 3 LER -i NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED INSTRUMENTATION AND CONTROL COMPONENTS AT U.S. COMMERCIAL NUCLEAR s POWER PLAf4TS JANUARY 1,1976 TO DECEMBER 31,1981, s NUREG/CR-2000 V03 N6 LICENSEE EVENT REPORT (LER) COMPILATION For Month 3 of June 1984 R NUREG/CR-2000 V03 N7 LICENSEE EVENT REPORT (LER) Compilation For Montn 7" of July 1984 NUREG/CR-2000 V03 N8: LICENSEE EVENT REPORT (LER) Compilation For Month j Of August 1984 WREG/CR-3867: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVE"TERS AT 5 U.S. COMMERCIAL NUCLEAR PonER PLANTS, JANUARY 1,1976 TO DECEMelR 4 31,1982. NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT Y REPORTS. Users Guide. h LOCA NUREG=0800 36.2.1 R6 STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY 1 ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition Revision 6 to l 106
l 3:ctien 6.2.1.1.C, "Proccuro-Supprocotcn Typ3 St:R CCntoinnonto." NUREG/CR-3418: SCREENING TESTS OF TERMINAL BLOCK PERFORMANCE IN A SIMULATED LOCA ENVIRONMENT. L NUREG/CR=3711: BWR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST I RESULTS. NUNEG/CR-4001: CONTEMPT 4/ MUD 5 AN IMPROVEMENT TO CONTEMPT 4/ MOD 4 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. LOFT NUREG/CR=0169 V17: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recorded On DAVDS. Leak Detection NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT nATER 4 REACTORS: Annual Report,0ctober 1982 = September 1983 Leak-Sofore-Break NUREG/CP=0051: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON LEAK-3EFORE-8REAK IN NUCLEAR REACTOR PIPING. Legal Issuances NUREG-0750 V19 101: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JANUARY-MARCH 1984 NUREG-0750 V19 NO3 NUCLEAR REGULATORY COMMISSION ISSUANCES F0R MARCH j 1984. Pages 555-936 i NUREG-0750 V19 N04: NUCLEAR NEGULATORY COMMISSION ISSUANCES FOR APNIL 1984.Pages 937-1,149 NUREG-0750 V19 N95: NUCLEAN HEGULATORY COMMISSION ISSUANCES FOR MAY 1984.Pages 1,151-1,321. Licensed Operating Reactors NUREG=0020 V08 N06: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPT. Data As of May 31,1984.(Grey Book) NUREG-0020 V08 N08: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT.Date As of July 31,1984.(Grey Book) Licensee Contractor and Vendor Inspection 4 i NUREG=0040 V08 N02: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Quarterly Report, April-June 1984.(hhite Book). i i Licensee Event Report NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED INSTRUMENTATION AND CONTHOL COMPONENTS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. 4 NUREG/CR-2000 V03 N6 LICENSEE EVENT REPORT (LER) COMPILATION For Month of June 1984 NUREG/CR-2000 V03 N7: LICENSEE EVENT REPORT (LER) Compilation For Month i j of July 1984 NUREG/CR-2000 V03 N8: LICENSEE EVENT REPORT (LER) Compilation For Month Of August 1984 NUREG/CR-3824 CONTING PH0 GRAM GUIDE. NUREG/CR-3867: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTERS AT U.S. COMMERCIAL NUCLEAN P0HER PLANTS, JANUARY 1,1976 TO DECEMdEH 31,1982. NUREG/CR-3905: SEQUENCL COUING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Guide. Licensing Examination NUREG/CR-3739: THE OPERATON FEEDBACK WORKSHOP: A TECHN!QUE FOR OBTAINING FEEDBACK FROM OPERATIONS PLRSONNEL. l Liquid Pathway NUREG-1054: SIMPLIFIED ANALYSIS FOR LIQUID PATHWAY STUDIES. Load Ratio NUREG/CR-3228 V02: STRUCTUHAL INTEGRITY OF r!ATER REACTOR PRESSURE SOUNDARY COMPONENTS. Annual Report For 1983 Long-Range Research 'NUREG-1080 V01: LONG= RANGE RESEARCH PLAN FY 1985-1989. 107
Leng-Toro R0pecitery-NUREG/CR=3832: UNCERTAINTIES IN LONG-TERM REPOSITORY PERFORMANCE DUE TO THE EFFECTS OF FUTURE GEULOGIC PROCESSES. Los Alamos Failure Model NUREG/CR=3679: CALIBRATIUN AND QUALIFICATIUN OF THE LOS ALAMOS FAILURE MODEL (LAFM). Loss =0f=8enefits Analysis NUREG/CR=3929: LOSS =0F-8ENEFITS ANALYSIS FOR NUCLEAR P0hER PLANT SHUTDOWNS. Methodology And Illustrative Case Study. Loss-Of= Coolant Accident NUREG/CR=3418: SCREENING TESTS OF TERMINAL BLOCK PERFORMANCE IN A I SIMULATED LOCA ENVIRONMENT. NUREG/CR=3459: EXPERIMENI DATA REPORT FOR MULTIR0D BURST TEST (MH8T) BUNDLE.8=5 NUREG/CR=3460 EXPERIMLNT DATA REPORT FOR MULTIROD BURST TEST (MHST) 8UNDLE 8-6 4 I NUREG/CR=3711: BWR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST i RESULTS. l Lors-Of= Flow Accident-NUREG/CR=3662: FUEL = DISRUPTION EXPERIMENTS UNDER HIGH= RAMP = RATE HEATING CONDITIONS. Loss-Of= Fluid Test i NUREG/CR=0169 V17: LOFT EXPEHIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recorded On DAVDS. Low = Level Weste NUREG/CR=3798 CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS l CONTAINING SIMULATED LOW-LEVEL WASTE INCINERATOR ASH. ~ NUREG/CR=3844: CNARACTERIZATION OF THE RADIDACTIVE WASTE PACKAGES OF THE MINNESOTA MINING AND MANUFACTURING COMPANY. Lung Cancer NUREG=1029: A COMPUTER COOL t0R GENERAL ANALYSIS OF HADON HISKS (GARR), MARCH 2 NUREG/CR=3988: MARCH 2 (MELTUOWN ACCIDENT NESPONSE CHARACTERISTICS) i CODE-DESCRIPTION AND USENS MANUAL, MINET i NUREG/CR=2331 V03 N3: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report, July-September 1983 NUREG/CR=2331 V03 N4 SAFETY RESEARCH PROGRAMS SPONSORED BY THE OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report,0ctober 1 = December 31,1983 NUREG/CR=3165: MINET SIMULATION OF A HELICAL COIL SODIUM / WATER STEAM 4 i GENERATOR,1NCLUDING STHUCTURAL EFFECTS. i NUREG/CR=3813 MINET VALIDATION STUDY USING STEAM GENERATOR TRANSIENT DATA. M001 l NUREG/CR=3690: RELAPS ASSESSMENT SEMISCALE NATURAL CIRCULATION TESTS i S=NC=3,3=NC=4,AND S=NC=8 M005 NUREG/CR=4001: CONTEMPT 4/ MUD $3AN IMPROVEMENT TO CONTEMPT 4/ MOD 4 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE l CONTAINMENT ANALYSIS. i MRST Sundle S=6 Test Data NUREG/CR=3460: EXPERIMENI DATA REPORT FOR MULTIROD BURST TEST (MRST) SUNDLE 8-6 Maintenance l NUREG/CR=3610: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of Subtask C 3He Neutron Spectrometer. 'Markov Modeling l NUREG/GR=3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION 108 i . ~ ~
PROTECTION AT. NUCLEAR P0:ER PLANTS.A Colculation Mothed. Meterial Transport NUREG/CR-3735: ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATURE REVIEW. Measurement NUREG/CR-3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTUN FIELDS AT COMMERCIAL NUCLEAR SITES. NUREG/CR-3590: EVALUATION OF ISOTOPE DILUTION MASS SPECTROMETRY FOR SIDASSAY MEASUREMENT OF ORAMIUM, PLUTONIUM,AND THORIUM IN URINE. NUREG/CR-3856: AN ULTRASUNIC LEVEL AND TEMPERATURE SENSOR FOR PonER REACTOR APPLICATIONS. Mechanistic Model NUREG/CR-3346: 810 ASSAY DATA AND A RETENTION-EXCRETION MODEL FOR 1 SYSTEMIC PLUTONIUM. Meltdown Accident Response Cnaracteristics NUREG/CR-3988: MARCH 2 (MELT 90nN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USERS MANUAL. Methodology NUREG/CR-3892: A RESEARCH PRUGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Summary Report. i NUREG/CR-3892 V01: A RESEAMCh PROGRAM FOR SEISMIC QUALIFICATION OF i NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey of Methods For Equipment And Components Evaluation of l MethodologyJQualification And Methodology.... l NUREG/CR-3892 V02: A RESLANCH PROGRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 2-Correlation Of Methodologies For Seismic Qualification Tests Of Nuclear Plant i Equipment. NUREG/CR-3892 V03: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement Of Equipment Qualification Methodology And Criteria. NUREG/CR-3892 V04: A RESEARCH PROGRAM FOR SEISMIC WUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 - Tne Use Of Fragility In Design of Nuclear Plant Equipment. Modeling NUREG/CR-3139: SCENARIUS AND ANALYTICAL METHODS FOR UF6 RELEASES AI NRC-LICENSED FUEL CYCLE FACILITIES. Multirod Surst Test NUREG/CR-3459: EXPERIMENT DATA REPJRT FOR MULTIROD BURST TEST (MRST) BUNDLE B-5 National Energy Issues NUREG/CP-0053: PROCEEDINGS OF THE NINTH ANNUAL STATISTICS SYMPUSIUM UN NATIONAL ENERGY ISSUES,0ctober 19-21,1983. Natural Circulation NUREG/CR-3654: PWR FLECHf SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14. NUREG/CR-3090 RELAPS ASSE3SMENTsSEMISCALE NATURAL CIRCULATION TESTS $=NC-3,S=NC-4,AND S-NC-8 Natural Convection Core Cooling NUREG/CR-3804 vols PHYSICS OF REACTOR SAFETY. Quarterly Report January March 1984 Near-Ground Tornado Wind F,ields NUREG/CR-3869: ANALYSIS UF THE IMPACT OF INSERVICE INSPECTION USING A PIPING RELIABILITY MUDEL. Near-Surface Disposal NUREG/CR-0130 AD003: TECHNOLUGY, SAFETY AND CUSTS OF OECOMMISSIONING A REFEREhCE PRESSURIZE 0 AAIER REACTOR PonER STATION. Noutron Transport 109
NUREG/CR=2996: SENSITIVITY OF DETECTING IN= CORE VIBRATIONS AND BOILING IN PRESSURIZED WATER REALTORS USING EX-CORE NEUTRON NOISE. Nondestructive Evaluation NUREG/CR=3689 V02: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT = WATER HEACTOR 3AFETY RESEARCH PROGRAM. Quarterly Progress heport, April-June 1983. NUREG/CR=3594: ULTRAS 0dIC ANU HETALLURGICAL EXAMINATION OF A CHACKED TYPE 304 STAINLESS STEEL BnR PIPE nELDMENT, Nondestructive Examination NUREG/CR=3921: DRY SPENT FUEL STORAGE TEST PLAN FOR FINAL NONDESTRUCTIVE FUEL 80D EXAMINATION. Notch Ductility NUREG/CR=3228 V02: STRUCTURAL INTEGRITY OF WATER REACTOR PRESSURE BOUNDARY COMPONENTS. Annual Report For 1983 OCA=P NUREG/CR=3618: OCA=P,A DETLRMINISTIC AND PROBABILISTIC FRACTURE-MECHANICS COOL FOR APPLICATION TO PRESSURE VESSELS. Operating Reactors Licensing Action NUREG=0748 V04 N05: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As of May 31,1984.(Orange uook) NUREG=0748 V04 N06: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of June 30,1983 (Orange Book) NUREG=0748 V04 N07: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of July 31,1984.(Orange Book) Operator Feedback NUREG/CR=3739: THE OPERATOM FEED 8ACK WORKSHOP A TECHNIQUE FOR OBrAINING FEEDBACK FROM OPERATIONS PERSONNEL. Organic Matter NUREG-1081 POST = ACCIDENT GAS GENERATION FROM RADIALYSIS OF ORGANIC MATERIALS. PCA Experiment NUREG/CR=3318: LWR PRESSURE VESSEL SURVEILLANCE 00SIMETRY IMPROVEMENT PROGRAM PCA Experiments,dlina Test,And Physics-Dosimetry Support For The PSF Experiments. PELLETRON NUREG/CR=3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON= RASTER SYSTEM. PLUGH NUREG/CR=3190: PLUGM A COUPLED THERMAL-H~DRAULIC COMPUTER MODEL FOR FREEZING MELT FLOW IN A CHANNEL. PRA NUREG-1050 PROBABILISTIC HISK ASSESSMENT (PRA) REFERENCE DOCUMENT. Final Rept. NUREG=1066: REVIEW INSIGHTS UN THE PROBABILISTIC RISK ASSESSMENTS FOR THE LIMERICK GENERATING STATION, UNIT 1 AND 2. NUREG/CR-2015 V08 PHASE I FINAL REPORT = SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. NUREG/CR=3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS. Executive Summary. NUREG/CR=3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS.A Review Of Occupational Dose Assessment Considerations In Current Probabilistic Risk Assessment And Cost = Benefit Analyses. PTS NUREG/CR=3888: ANALYSIS OF THE VENUS PNR ENGINEERING MOCKUP EXPERIMENT = PHASE Is SOURCE DISTRIBUTION. PVS NUREG/CR=3515: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORR= PSF =PVS IRRAp!ATION EXPERIMENT. Penetration neld 110 ?
NUREG/CR-3053: CLOSEOUT OF IE SULLETIN 80-08: EXAMINATION OF CONTAINMENT LINER PENETRATION MELDS. Phenomenological Research NUREG/CR-3589 V01: REACTOR SAFETY RESEARCH QUARTERLY l REPORT. January-March 1983 NUREG/CR-3589 V02: REACTOR SAFETY RESEARCH GUARTERLY REPORT. April = June 1983 Photon Field NUREG/CR-3569: SPECIAL AND 00SIMETRIC MEASUREMENTS OF PHOTON FIELOS AT i COMMERCIAL NUCLEAR SITES. Pilot Program NUREG=1075: DECENTRALIZATION OF OPERATING REACTOR LICENSING REVIEWS.NRR Pilot Program. Pipe Cracking NUREG/CR=3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT = HATER-REACTOR SAFETY RESEARCH PROGRAM Quarterly Progress Report, January-March 1983 NUREG/CR-3689 V02: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT-MATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress Report, April = June 1983. NUREG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT nATER REACT 0HS: Annual Report,Uctober 1982 - September 1983 NUREG/CR-3894 ULTRASONIC ANU METALLURGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS STEEL BnR PIPE nELDMENT. Piping NUREG/CP=0051: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON LEAK-SEFORE-SREAK.IN NUCLEAR REACTOR PIPING. NUREG/CR-3228 V02: STRUCTURAL INTEGRITY OF WATER REACTOR PRES $URL SOUNDARY COMPONENTS. Annum) Report For 1983 l NUREG/CR-35998 SOURCES OF UNCERTAINTY IN THE CALCULATIONS OF LOADS ON SUPPORTS OF PIPING SYSTEMS. NUREG/CR-3845: PREDICTION OF NONLINEAR STRUCTURAL RESPONSE IN LMFBR ELEVATED-TEMPERATURE PIPING. NUREG/CR-3869: ANALYSIS UF THE IMPACT OF INSERVICE INSPECTION USING A 4 PIPING RELIABILITY MODEL. i NUREG/CR-3893: LABORATORY STUDIES DYNAMIC RESPONSE OF PROT 0 TYPICAL PIPING SYSTEMS. NUREG/CR-3939: WATER HAMMER, FLOW INDUCED VIBRATION AND SAFETY / RELIEF VALVE LOADS. Planning NUREG-0985 R01: U.S. NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN. Plastic Fracture NUREG/CR-3821: EVALUATION OF CRACK PLANE EQUILIBRIUM MODEL FOR PREDICTING PLASTIC FHACTURE. Plugging NUREG/CR-3190: PLUGM A COUPLED THERMAL-HYDRAULIC COMPUTER MODEL FOR l FREEZING MELT FLOW Id A CHANNEL. Polar Crane Recovery NUREG/CR-3884: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM - THREE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. l Post-Accident Gas Generation NUREG-1081: POST-ACCIDLNT GAS GENERATION FROM RADIALYSIS OF ORGANIC MATERIALS. Power Plant Impact NUREG/CR-3897 EVALUATION OF ECOSYSTEM SIMULATION MODELS AS TOOLS FOR ASSESSMENT OF POWER PLANT IMPACTS ON FISH POPULATICNO. Final Rept. Practice and Procedures U1 gest hUREG-0386 003: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. 111
Pressure Soundary Components
- NUREG/CR=3788 V01: STRUCTURAL INTEGR.ITY OF LIGHT WATER REACTOR PRESSURE 800NDARY COMPONENTS.Four-Year Plan 1984-1988.
F Pressure Soundary NUREG/CR=3228 V02: STRUCTURAL INTEGRITY OF WATER REACTOR PRESSURE 80UNDANY COMPONENTS.Annuel Report For 1983 Pressure Vessel Cladding j~ NUREG/CR=3671: ASSESSMENI OF RADIATION EFFECTS RELATING TO REACTOR j. PRESSURE VESSEL CLADUING. Pressure Vessel Research Committee NUREG/CR=38458 PREDICTION UF NONLINEAR STRUCTURAL RESPONSE IN LMF8R ELEVATED = TEMPERATURE PIPING. Pressure Vessel Simulation NUREG/CR=3815: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORR= PSF =PVS IRRAUIATION EXPERIMENT. Pressura Vessel i NUREG/CR=3318: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPRUVEMENT PROGRAM PCA Experiments,dlind Test,And Physics = Dosimetry Support For The PSF Experiments. 1 NUREG/CR=3616s OCA=P,A DETERMINISTIC AND PR06A81LISTIC I . FRACTURE = MECHANICS CODE FOR APPLICATION TO PRESSURE VESSELS. NUREG/CR=3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR UCT0 DER 1983 = MARCH 1984 NUREG/CR=3988: MARCH 2 (MELTOOWN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTIUN AND USERS MANUAL. Pressurized Thermal Shock } NUREG/CR=2331 V03 N43 SAFETY RESEARCH PROGRAMS SPONSORED BY THE OFFICE l OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report,0ctober i {- =0ecember 31,1983 1 NUMEG/CR=3618 OCA-P,A DETERMINISTIC AND PROBASILISTIC FRACTURE = MECHANICS CODE FOR APPLICATION TO PRESSURE VESSELS. i NUREG/CR=3761t RELAPS THERNAL= HYDRAULIC ANALYSES OF PRESSURIZED THERMAL i SHOCK SEQUENCES FOR THE OCONEE=1 PRESSURIZED WATER REACTOR. I NUREG/CR=38883 ANALYSIS OF THE VENUS PnR ENGINEERING MOCKUP EXPERIMENT -PHASE 18 SOURCE DISTRIBUTION. Primary Coolant System NUREG/CR=3663 V02: PROBABILITY OF PIPE FAILURE'IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced cy Crack Growth, i Probabilistic Fracture Meenanics NUREG/CR=3618: OC A=P, A DETERMINIS11C AND PROBASILISTIC FRACTURE = MECHANICS CUDE FOR APPLICATION TO PRESSURE VESSELS. Probabilistic Risk Assessment i NUREG-1050 PROSABILISTIC RISK ASSESSMENT (PRA) REFERENCE DOCUMENT. Final Rept. NUREG=10683 REVIEW INSIGHTS ON THE PR08ABILISTIC RISK ASSESSMENTS FOR THE LIMERICK GENERATING STATION, UNIT 1 AND 2 NUREG/CR=2015 V088 PHASE I FINAL REPORT = SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. NUREG/CR=3665: OPTIMIZATION UF PUBLIC AND OCCUPATIONAL RADIAT!UN 1 I PROTECTION AT NUCLEAR PonEN PLANTS. Executive Summary. NUREG/CR=3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION l PROTECTION AT NUCLEAR PonER PLANTS.A Review of occupational Dose Assessment Considerations In Current Probabilistic Risk Assessment And Cost-Senefit Analyses. Probabilistic Risk NUREG/CR=3665 V028 OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION l PROTECTION AT NUCLEAR POWER PLANTS. Considerations In Factoring l Occupational Dose Into Value= Impact And Cost =8enefit Analyses. 1 Process Equation Comparison 112
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hvREG/CR-3896: SIMULATION EXPERIMENTS COMPARING ALTERNATIVE PROCESS FORMULATIONS USING A FACTORIAL DESIGN. Process Holdup NUREG/CR-3678: ESTIMATION METHODS FOR PROCESS HOLDUP OF SPECIAL NUCLEAR MATERIALS. Protection NUREG/CR-3751: EFFECTS OF ROCK RIPRAP DESIGN PARAMETERS ON FLOOD PROTECTION C09TS FOR UNANIUM TAILINGS IMPOUNDMENT 5 RAMONA=38 NUREG/CR-2331 V03 N3 SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEANCH. Quarterly Progress Report, July-September 1983. l NUREG/CR-2331 Vr,3 N4 SAFEIY RESEARCH PROGRAMS SPONSORED BY THE OFFICE OF NUCLEAR REGULATORY HESEARCH.Ouarterly Progress Report,0ctober 1 -December 34,1983 RELAP5 NUREG/CR-3690 RELAPS ASSESSMENT SEMISCALE NATURAL CIRCULATION TESTS S-NC-3,3-NC-4,AND S-NC-8 NUREG/CR-3761: RELAP5 IHERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL I l SHOCK SEGUENCES FOR THE UCONEE-1 PRESSURIZED WATER REACTOR. RETS NUREG/CR-4007: LOWER LIMIT OF DETECTION DEFINITIUN AND ELABORATION OF A i PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL l MEASUREMENTS. l RHR NUREG/CR-3933: RISK RELATED RELIABILITY REQUIREMENTS FOR BWR SAFETY -IMPORTANT SYSTEMS WITN LMPHASIS ON THE RESIDUAL HEAT REMOVAL SYSTEM. Radiation Protection NUREG/CR-3469 V01: OCCUPATIONAL 00SE REDUCTION AT NUCLEAR POWER PLANTS ANNOTATED BIBLIOGRAPHY OF SELECTED READINGS IN RADIATIGN PROTECTION AND ALARA. NUREG/CR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nEd PLANTS. Executive Summary. NUREG/CR-3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS.A Review Of Occupational Vose Assessment Considerations In Current Probabilistic Risk Assessment And Cost-Benefft Analyses. NUREG/CR-3665 v02: OPTIMIZATION OF PUBLIC AND UCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nEH PLANTS. Considerations In Factoring Occupational Dose Into Value-Impact And Cost-Benefit Analyses. NUREG/CR-3665 V03 OPTIMIZATION OF PUBLIC AND UCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonEH PLANTS.A Calculation Method. Radiation i NUREG/CR-3569: SPECIAL AND DUSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. NUREG/CR-3665 V02: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nEH PLANTS. Consider ations In Factoring Occupational Dose Into Value-Impact And Cost-Benefit Analyses. NUREG/CR-3671: ASSESSMcNT UF RADIATION EFFECTS RELATING TO REACTOR PRESSURE VESSEL CLADu!NG. NUREG/CR-3788 V01: STRUCTUHAL INTEGRITY OF LIGHT WATER REACTOR PRESSURE BOUNDARY COMPONENTS.Fouc-Year Plan 1984-1988 NUREG/CR-3798: CHARACTERIZATION OF CENLNT AND DITUMEN WASTE F0HMS CONTAINING SIMULATED,LOh-LEVEL WASTE INCINERATOR ASH. NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADWASTE CUNTAINING ORGANIC ION-EXCHANGE MEDIA. NUREG/CR-3870: RADIATION OUSE ESTIMATES AND HAZARD EVALUATIONS FOR INH ALED AIRBORNE R ADIOr.UCLIDES. Annual Progress Rept July 1982 -June 1983. Radioactive naste 113
~ NUREG/CR-0130 ADD 03: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIGNING A REFERENCE PRESSURIZE 0 nATER REACTOR P0nER STATION. NUREG/CR-0672 AD002: TECHNULUGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE SOILING WATER HEACTOR POWER STATION. Classification of Decommissioning destes. NUREG/CR-3844: CHARACTLRIZATION OF THE RADIDACTIVE WASTE PACKAGES UF THE MINNESOTA MINING AND MANUFACTURING COMPANY. i Radiciodine NUREG/CR-3513: MECHANICAL RELIABILITY EVALUATION OF ALTERNATE MOTORS l FOR USE IN A RADICIOUIdE AIR SAMPLER. Radiological Effluent NUREG/CR-4007: LOWER LIMIT OF DETECTION DEFINITION AND ELABORATIUN OF A PROPOSED POSITION FOR HADI0 LOGICAL EFFLUENT AND ENVIRONMENTAL MEASUREMENTS. Radiological Monitoring NUREG/CR-4007: LohER LIMIT OF DETECTION: DEFINITION AND ELABORATION OF A PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL MEASUREMENTS. Radiolysis NUREG-1081: POST-ACCIDENT GAS GENERATION FROM RADIALYSIS OF ORGANIC MATERIALS. Nadionuclide NUREG/CR-3763: REVIEW AND' ASSESSMENT OF RADIONUCLIDE SORPTION INFORMATION FOR THE uASALT WASTE ISOLATION PROJECT SITE (1979 Through May,1983). NUREG/CR-3651 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPEu dY DUE HIGH-LEVEL NUCLEAR nASTE REPOSITORY SITE PROJECTS. Report for Octooer-December 1983 Redon 'JUREG-1029: A COMPUTER CUDE FOR GENERAL ANALYSIS OF RADON RISKS (GARR). i Radweste NUREG/CR-3812: ASSESSMENT UF IRRADIATION EFFECTS IN RADWASTE CONTAINING URGANIC ION-EXCHANGE MEDIA. Rastered NUACG/CR=3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETRON-RASTER SYSTEM. Reactor Coolant Loop NUREG/CR-3663 V02: PROBABILITY OF PIPE FAILUME IN THE REACTOR COOLANT LOCPS OF COMSUSTION ENGINEERING PWR PLANTS vol 2: Pipe Failure Induced by Crack Growth. Reactor Core Thermal Hydraulic NUREG/CR-3804 V01: PHYSICS OF REACTOR SAFETY. Quarterly Report January 1 March 1984 Reactor Pressure Vesse) NUREG/CR-3888: ANALYSIS OF THE VENUS PhR ENGINEERING MOCKUP EXPERIMENT -PHASE I SOURCE DISTRIBUTION, Reactor Protection System i NUREG/CR-3776: TESTING OF.3AFETY-RELATED NUCLEAR POWER PLANT EQUIPMENT AT THE CENTRAL RECEIVER TEST FACILITY. Reactor Safety NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION QUARTERLY PROGRESS REPORT, October-December 1983. NUREG/CR-3680 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT = WATER-REACTOR SAFETY RESEARCH PROGRAeroyeeterly Progrege Report, January-March 1983 NUREG/CR-3734: LIGHT WATER RLACTOR SAFETY RESEARCH PROGRAM. Semiannual f Report,0ctober 1982 - March 1983. NUREG/CR-3804 V01: PHYSICS OF REACTOR SAFETY. Quarterly Report January - March 1984 114 . _ -, _ _ - ~ _
R0 circulation Air Cooler NUREG/CR=3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTEMS IN RETAINING FISSION PRODUCTS. Background Information. Reflux Condensation NUREG/CR-3654: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14. Regionalization j NUREG=10753 DECENTRALIZATION OF OPERATING REACTOR LICENSING REVIEWS.NRR Pilot Program. Regulatory Requirements NUREG/CR-3932: BENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS i AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. l Regulatory and Technical Report NUREG-0304 V09 N02s REGULATONY AND TECHNICAL REPORTS. Compilation For ) Second Quarter 1984 Release NUREG/CR-3139: SCENARIOS AND ANALYTICAL METHODS FOR UF6 RELEASES AT l NRC-LICENSED FUEL CYCLE FACILITIES. Reliability NUREG/CR=3932: BENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. NUREG/:R-3933: RISK RELATED RELIABILITY REQUIREMENTS FOR BWR SAFETY -IMPORTANT SYSTEMS WITH EMPHASIS ON TME RESIDUAL HEAT REMOVAL SYSTEM. Research Frogram NUREG/CR-3892: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EGUIPMENT. Summary Report. NUREG/CR-3892 V01: A RESEARCN PA0 GRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey Of Methods For Equipment And Components Evaluation Of MethodologyJQualification And Methodology.... HUREG/CR-3892 V02: A RESEANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL, EQUIPMENT. Task 2-Correlation Of Methodologies For Seismic Qualification Tests Of Nuclear Plant Equipment. NUREG/CR-3892 V03: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement of Equipment Qualification Methodology Ano Criteria. NUREG/CR-3892 V04: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 - The Use of I Fragility In Design of Nuclear Plant Equipment. Residual Heat Removal i NUREG/CR-3933 RISK RELATED HELIABILITY REQUIREMENTS FOR BWR SAFETY -IMPORTANT SYSTEMS WITH EMPHASIS ON THE RESIDUAL HEAT REMOVAL SYSTEM. Rosin Degradation NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADWASTE CONTAINING ORGANIC ION-EXCHANGE MLDIA. Retention NUREG/CR-3346: 810 ASSAY UATA AND A RETENTION-EXCRETION MODEL FUR SYSTEMIC PLUTONIUM. Retrieval System NUREG/CRe3951: INTRODUCTION TO BIBEL0T A BIBLIOGRAPHIC FINDING AND RETRIEVAL SYSTEM. Risk Analysis NUREG=1054: SIMPLIFIED ANALYSIS FOR LIQUID PATHWAY STUDIES. NUREG/CR-3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A' Status Report. NUREG/CR-3591 V02: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. 115
Rick AcconcGont NUREG/CR-34933-A REVIEn 0F THE LIMERICK GENERATING STATION SEVERE i ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. NUREG/CR-3929: LOSS-OF-BENEFITS ANALYSIS FOR NUCLEAR POWER PLANT f SHUTD0hNS Methodology And Illustrative Case Study. Risk NUREG-1029: A COMPUTER CUDL FOR GENERAL ANALYSIS OF RADON RISKS (GARR). NUREG-1050: PROBABILISTIC HISK ASSESSMENT (PRA) REFERENCE DOCUMENT. Final Rept. i NUREG/CR-2015 V08: PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. NUREG/CR-3665: OPTIMIZATION UF PUBLIC AND OCCUPATIONAL RADIATIUN PROTECTION AT NUCLEAR P0nEk PLANTS. Executive Summary. NUREG/CR-3870 RADIATION DUSL ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Rept July 1982 -June 1983 NUREG/CR-3933: RISK RELATEU HELIABILITY REQUIREMENTS FOR BWR SAFETY -IMPORTANT SYSTEMS WITH EMPHASI* ON THE RESIDUAL HEAT REMOVAL SYSTEM. Rock R1 prep + NUREG/CR-3751: EFFECTS OF HOCK RIPRAP DESIGN DARAMETERS ON FLOOD PROTECTION COSTS FOR UHANIUM TAILINGS IMPOUNDMENTS. Rules NUREG-0386 003: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. Rupture NUREG/CR-3663 V02: PR08A61LITY OF PIPE FAILURE IN THE REACTOR COOLANT ~ l LOOPS OF COMBUSTION ENGINEERING PWR PLANTS.Vol 2 Pipe Failure Induced i by Crack Growth. SCSS NUREG/CR-3824: CONTING PROGRAM GUIDE. NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Guide. SEASET NUREG/CR-365a PnR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report i NRC/EPRI/ Westinghouse Heport No. 14 l SLIM-MAUD NUREG/CR-3518 V01: SLIM-MAUDIAN APPROACH TO ASSESSING HUMAN ERROR PROBASILITIES USING STHUCTURED EXPERT JUDGEMENT. Volume Isoverview of SLIM-MAUD. SOLA-PTS NUREG/CR-3822: SOLA-PTS: A Transient,Three-Dimensional Algorithm For l Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometries. SSC NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYS'EM OF THE EXPERIMENTAL BREEDER REACTUR-II. Safeguards j NUREG/CR-3520 V01: LONG-TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING l SAFEGUARDS AT NUCLEAR POWER PLANTS.Vclume I Summary And Users Guide. NUREG/CR-3520 V02: LONG-TEHM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR 90nER PLANTS. Volume II Development Of Detailed Analyses. l Safety Evaluation Report NUREG-0420 306: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SHQRENAM NUCLEAR POWER STATIONrUNIT N0, 1. Docket No. 50-322 (Long Island Lighting Company) NUREG-0420 807: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SHOREHAM NUCLEAR POWER STATION UNIT NO.1. Docket No. 50-322 (Long Island Lighting Company) NUREG-0675 324: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF 116
DIABLO CANYON NUCLEAR PonER PLANT,0 NITS 1 & 2. Docket Nos. 50=275 & 50-323.(Pacific Gas & Electric Company) NUREG=0675 525: SAFETY EVALUATION REFORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR PonER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50=323.(Pacific Gas Ano Electric Company) NUREG-0675 S26: SAFETY EVALUtiTION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR P0nER PLANT, UNITS I AND 2 NUREG-0675 327: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF I DIABLO CANYON NUCLEAR P0nEN PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas Ana Electric Company) NUREG-0680.S05 TMI-1 RESTART.An Evaluation Of Tne Licensee's Management Integrity As It Affects Restart of Three Mile Island Unit 1 Docket 50-289 NUREG-0787 S07: SAFETY EVALUATION REPORT RELATED TO-THE OPERATION OF j WATERFORD NUCLEAR P0nEN PLANT, UNIT 3. Docket No. 50-382. (Louisiana Power & Light Company) NUREG=0798 SO43 SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF ENRICO FERMI ATOMIC POWER PLANT, UNIT NO. 2. Docket No. 50-341. (Detroit Edison Company) NUREG-0831 805: SAFETY EVALUATION REPORT RELATED TO THE UPERATION OF GRAND GULF NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-416 Ano 50-417.(Mississippi Power And Light Company, Middle South Energy,Inc And Soutn Mississippi Electric Power Association) NUREG-0954 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF CATAhBA NUCLEAR STATIOH,0 NITS 1 AND 2. DOCKET Nos. 50-413 And 50-414.(Duke Power Company,et al) NUREG-10313 SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MILLSTONE-NUCLEAR POWER STATION, UNIT NO. 3. Docket No. 3 50-423.(Northeast Nuclear Energy Company) NUREG-1069: SAFETY EVALUATION REPORT RELATED TO THE RENEHAL OF THE OPERATING LICENSE FOR THE GENERAL ELECTRIC-NUCLEAR TEST REACTOR (GE-NTR).00CKET NO. 50-73.(General Electric Company) NUREG-10838 SAFETY EVALUATION REPORT RELATED TO THE HENEnAL OF THE OPERATING LICENSE FOR THE nESTINGHOUSE RESEARCH REACT 0k AT ZION,ILLINDIS. DOCKET No. 50-87.(nestinghouse Electric Company) NUREG/CR-0130 A0003: TECHNOLUGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZE 0 nATEH REACTOR P0nER STATION, Safety Goals 5 NUREG-10503 PROBABILISIIC RISK ASSESSMENT (PRA) REFERENCE DOCUMENT. Final Rept. Safety Researen NUREG/CR-2015 V08 PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Margins Research Program. NUREG/CR-2331 V03 N3: SAFETY RESEARCH PROGRAMS SPONSURED BY OFFICE OF NUCLEAR REGULATORY REStANCH. Quarterly Progress Report, July-September i 1983 NUREG/CR-2331 V03 N4 SAFELY RESEARCH PROGRAMS SPONSURED BY THE UFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report,0ctober 1 -December 31,1983 NUREG/CR-3589 V01: REACTOR SAFETY RESEARCH QUARTERLY HEPORT. January-March 1983 NUREG/CR-3589 V02: REACTuR SAFETY RESEARCH QUARTERLY REPORT. April-June 1983. NUREG/CR-3689 V023. MATERIALS SCIENCE AND TECHNOLOGY DIVISION 4 LIGHT-nATER REACTOR SAFEIY RESEARCH PROGRAM. Quarterly Progress Report, April-June 1983. NUREG/CR-3689 V03: MATERIALS SCIENCE AND TECHNOLOGY DIVTSION i LIGHT-WATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress Report, July-September 1983. Safety-117
NUREG/CR-39323 BENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. NUREG/CR-3933: RISK RELATEU HELIABILITY REWUIREMENTS FOR BnR SAFETY, -IMPORT ANT SYSTENS WITri tMPHASIS ON THE RESIDUAL HEAT REMOV AL SYSTEM. Safety / Relief Valve NUREG/CR-3939 WATER HAMMER,FLOh INDUCED VIBRATION AND SAFETY /RELILF VALVE LOADS. Search System NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Guide. Security NUREG/CR-3520 V01: LONG-TEHM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAd P0nEn PLANTS. Volume I: Summary And users Guide. NUREG/CR-3520 V02: LONG-TENM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR P0nER PLANTS. Volume II Development Of Detailed Analyses. Seismic Capacity NUREG/CR-3893: LABORATURY STUDIES: DYNAMIC RESPUNSE OF PROTOTYPICAL PIPING SYSTEMS. Seismic Loading NUREG/CR-3742: BUCKLING OF STEEL CONTAINMENT SHELLS UNDER TIME-DEPENDENT LOADING. Seismic Qualification NUREG/CR-3692: A RESEANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Summary Report. NUREG/CR-3892 V01: A RESEARCH PROGRAM FOR SEISMIC UUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey Uf Methods For Equipment And Components Evaluation Of MethodologyJQu61ification And Methodology.... NUREG/CR-3892 V02: A RESLANCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 2-Correlation Of Methodologies For Seismic Qualification Tests Of Nuclear Plant Equipment. NUREG/CR-3892 v03: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement Of Equipment Qualification Methodology And Criteria. NUREG/CR-3892 V04: A RESEARCH PROGRAM FOR SEISMIC WUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 - Tne Use Of Fragility In Design Of Nuclear Plant Equipment. Seismic Risk NUREG/CR-2015 V08: PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT = VII). Seismic Safety Margins Research Program. Semiscale NUREG/CR-3690: RELAPS ASSESSMENT SEMISCALE NATURAL CIRCULATION TESTS S-NC-3,S-NC-4,AND S-NC-8. Sequence Coding and Search System NUREG/CR-3624: CONTING PROGRAM GUIDE. Sequence coding NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Guide. Severe Accident Sequence Analysis NUREG/CR-3470s ATWS AT BH0HNS FERRY UNIT ONE - ACCIDENT SEGUENCE ANALYSIS. Severe Accident NUREG/CR-3493: A REVIEn UF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. NUREG/CR-3988: MARCH 2 (MELT 00HN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USENS MANUAL. Severe Reactor Accident 118
NUREG/CR-3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTtMS IN RETAINING FISSION PH000 CTS. Backs?ound Information. Shipping Containers NUREG/CR-3826: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE SY BRITTLE FRACTURE IN FEHRITIC STEEL SHIPPING CONTAINERS GREATER THAN FOUR INCHES THICK. Shutdown NUREG/CR-3929: LOSS =0F-BENEFITS ANALYSIS FOR NUCLEAR POWER PLANT SHUTD0hNS. Methodology And Illustrative Case Study. Simulation Models NUREG/CR-3896: SIMULATION EXPERIMENTS COMPARING ALTERNATIVE PROCESS FORMULATIONS USING A FACTORIAL DESIGN. Small Broek Loss-Of-Coolant Accident NUREG/CR-3895: INVESTIGATIUN OF COLD l.EG WATER HAMMER IN A PnR DUE TU THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. Sodium Thiosulfate NUREG/CR-3834: ON THE THHESHOLD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CURR0SION CRACKING OF SENSITIZED ALLOY 600 IN 80 RATED THIOSULFATE SOLUTION. Source Term NUREG/CR-3735: ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATURE REVIEW. NUREG/CR-3796: EMERGENCY PHEPAREDNESS SOURCE TERM DEVELOPMErT F0H THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES. Spent Fuel NUREG/CR-3708: LnR SPENT FUEL, DRY STORAGE dEHAVIOR AT 229 C. NUREG/CR-3826: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE sy l BRITTLE FRACTURE IN FERRITIC STEEL SHIPPING CONTAINENS GREATER THAN l FOUR INCHES THICK. l Stainless Steel Welds NUREG/CR-3671: ASSESSMENT OF RADIATION EFFECTS RELATING TO REACTOR PRESSURE VESSEL CLAD 0ING. Standard Review Plan NUREG-0800 06.2.1 R6: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition Revision 6 to Section 6.2.1.1.C, ' Pressure-Suppression Type ShR Containments." Station Blackout NUNEG/CR-3640 COST ANALYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE ABILITY OF A NUCLEAR PLAUT TO ENDURE STATION BLACK 0UT. Steam Explosion NUMEG/CR-3369: AN UNCENTAINTY STUDY OF PMR STEAM EXPLOSIONS. Steam Generator NUREG/CR-3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nEH PLANTS.A Calculation Method. NUREG/CR-3765: MINET SIMULATION OF A HELICAL COIL SODIUM / WATER STEAM GENERATON, INCLUDING STduCTURAL EFFECTS. NUREG/CR-3813 MINET VALIDATION STUDY USING STEAM GENERATOR TRANSIENT DATA. NUREG/CR-3842: STEAM GENERATOR GROUP PROJECT TASK 8 - SELECTIVL TUdE UNPLUGGING. NUREG/CR=3843: STEAM GENERATOR GROUP PROJECT TASK 10 - SECONDARY SIDE EXAMINATION. Steel Containment NUREG/CR-3742: BUCKLING uF STEEL CONTAINMENT SHELLS UNDER TIME-DEPENDENT LOADING. Stress Corrosion Cracking NUREG/CR-2331 v03 N3 SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR NEGULATORY RLSLAHCH. Quarterly Progress Report, July-September
- 1983, 119
NUREG/CR-2331 V03 N4 SAFETY RESEARCH PROGRAMS SPONSORED BY THE OFFICE OF NUCLEAR REGULATORY HESEARCH. Quarterly Progress Report,0ctober 1 -December 31,1983 NUNEG/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS: Annual Report,0ctober 1982 - September 1983 NUREG/CR-3834: ON THE THNESHOLC SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRECNING OF SENSITIZED ALLOY 600 IN HORATED THIOSULFATE SOLUTION. Stress NUREG/CR-3821: EVALUATION OF CRACK PLANE EQUILIBRIUM MODEL FOR PREDICTING PLASTIC FRACTURL. Structural Analysis NUREG/CR-3645: PREDICTION OF NONLINEAR STRUCTURAL RESPONSE IN LMFBH ELEVATED-TEMPERATURE PIPING. Structural Integrity NUREG/CR-3T88 V01: STRUCTURAL INTEGRITY OF LIGHT WATER REACTOR PRESSURE BOUNDARY COMPONENTS.Four= Year Plan 1984-1988 Structural Steel NUREG/CR-3788 V01: STRUCTURAL INTEGRITY OF LIGHT WATER REACTOR PRESSURE BOUNOARY COMPONENTS.Four-Year Plan 1984-1988. Super System Code NUREG/CR-2331 V03 N4: SarETY RESEARCH PROGRAMS SPONSORED BY THE OFFICE OF NUCLEAR REGULATORY HESEARCH. Quarterly Progress Report,0ctober 1 -December 31,1983 NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYSTEM OF THE EXPERIMENTAL BREEDER REACTOR =II. Surveillance NUREG/CR-3318: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM PCA Experiments,dlind Test',And Physics-Dosimetry Support For The PSF Experiments. Systems Interaction NUREG/CR-3593 V01: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR P0nEN PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. NUNEG/CR-3593 V02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR P0nER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2. TMI=1 Restart NUREG-0680 305: TMI-1 RESTART.An Evaluation Of The Licensee's Management Integrity As It Affects Restart of Three Mile Island Unit 1 Occket 50-289. TRAC-PF1/ MOD 1 NUNEG/CR-3820 V01: THENMAL/HYORAULIC ANALYSIS RESEARCH PROGRAM. Quarterly Report, January-March 1984 Task Action Plans NUREG=0649 R01: TASK ACTION PLANS FOR UNRESOLVED SAFETY ISSUES RELATED TO NUCLEAR POWER PLANTS. Technical Specification NUREG-10T2 TECHNICAL SPECIFICATIONS FOR CATAWBA NUCLEAR STATION, Unit
- 1. Docket No. 50-413.
Temperature NOREG/CR-3815: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORN-PSF-PVS IRRADIATION EXPERIMENT, NUREG/CR-38568 AN ULTRASUNIC LEVEL AND TEMPERATURE SENSOR FOR POWEH REACTOR APPLICATIONS. NUMEG/CR-3856: AN ULTRASONIC LEVEL AND TEMPERATURE SENSOR FOR P0nER REACTOR APPLICATIONS. Terminal Block NUREG/CRa3418: SCREENING TESTS OF TERMINAL BLOCK PERFORMANCE IN A SIMULATE 0 LOCA ENVIRONMENT. 120
j Thcrmol Shcak NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR UCTOWEN 1983 = MARCH 1984 Thermel-Hydraulic l NUREG/CR=2331 V03 N3: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAN REGULATORY RESEARCH. Quarterly Progress Report, July-September 1983 NUREG/CR-2331 V03 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY THE OFFICE OF NUCLEAR REGULATORY HESEARCH. Quarterly Progress Report,0ctober 1 -December 31,1983 NUREG/CR-3761: RELAPS THERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEGUENCES FOR THE OCONEE-1 PRESSURIZED WATER REACTOR. NUREG/CR-3765: MINET SIMULATION OF A HELICAL COIL SODIUM / WATER STEAM CENERATOR, INCLUDING STHUCTURAL EFFECTS. NUREG/CR-3820 V01: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Uuarterly Report, January-March 1984 NUREG/CR-3678: MODELIN4 CONSIDERATIONS FOR THE PRIMARY SYSTEM OF THE EXPERIMENTAL BREEDER REACTOR-II. Title List NUREG=0540 V06 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.May !=31, 1984 NUREG-0540 V06 N07: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILAULE July 1-31, 1984 l Training l NUREG/CR-3520 V01: LONG-TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARuS AT NUCLEAR PonER PLANTS. Volume I: Summary And Users Guide. NUREG/CR-3520 V02: LONG-TEMM RESEARCH PLAN FOR HUMAN FACTORS AFFLCTING i SAFEGUARDS AT NUCLEAR P0nER PLANTS. Volume II: Development Of uetailed Analyses. NUREG/CR=3739: THE OPERATOR FEEDBACK WORKSHOP:A TECHNIQUE FOR UBIAINING FEEDSACK FROM OPERATIONS PERSONNEL. Transient NUREG/CR-3888: ANALYSIS UF THE VENUS PWR ENGINEERING MOCKUP EXPEHIMENT -PHASE I SOURCE DISTRIBUTION. Tube Plugging NUREG/CR-3642: STEAM GENERATOR GROUP PROJECT TASK 8 - SELECTIVE TUuE UNPLUGGING. TuDe Rupture NUREG/CR-3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAN P0nEH PLANTS.A Calculation Method. Tubes NUREG/CR-3459: ExPERIMLNi UATA REPORT FOR HULTIROD BURST TEST (MRBT) BUNDLE 8-5. NUWEG/CR-3460: EXPERIMtNT DATA REPORT FOR MULTIR00 BURST TEST (MHBT) BUNDLE 8-6 Type 304 Stainless Steel Pipe NUREG/CR-3894: ULTRAS 0dIC AND METALLURGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS STELL BnR PIPE HELDMENT. U-Bends NUREG/CR-3634: ON THE IHHESHuLD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CURROSION CRACKING OF SENSITIZE 0 ALLOY 600 IN BORATED i THIOSULFATE SOLUTION. l USI A-44 l NUREG/CR-3840: COST A,NnLYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE l ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKUUT. Uncertainty Analysis NUREG/CR-3369: AN UNCERTAINTY STUDY OF PHR STEAM EXPLOSIONS. Unresolved Safety Issues NUREG-0649 Rois TASK ACTION PLANS FOR UNHESOLVED SAFETY ISSUES RELATED TO NUCLEAR POWER PLANTS. 121
Urcniua Toilingo NUREG/CR-3751: EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FLOUD PROTECTION COSTS FOR URANIUM TAILINGS IMPOUNDMENTS. drinalyses NUREG/CR-3346: BIDASSAY 0ATA AND A RETENTION-EXCRETION MODEL FOR SYSTEMIC PLUTONIUM. User's Manual NUREG/CR-4011: THE 21/55 DATA BASE USER'S MANUAL. VENUS NUREG/CR-3888: ANALYSIS UF TnE VENUS PMR ENGINEERING MOCKUP EXPERIMENT -PHASE Is SOURCE DISTRIBUTION. VGES NUREG/CR-3273: COMBUSTION OF HYDROGEN AIR MIXTURES IN THE VGES CYLINDRICAL TANK. Value/ Impact NUREG/CR-3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0aER PLANTS. Executive Summary. NUREG/CR-3665 V01: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR POWER PLANTS.A Review Of Occupational pose Assessment Considerationu In Current Procabilistic Risk Assessment And Cost-Benefit Analyses. NUREG/CR-3065 V02: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonER PLANTS. Considerations In Factoring Occupational Dose Into Value= Impact And Cost-Benefit Analyses. Variable Geometry Experimental System NUREG/CR-3273: COMBUSTION UF HYDROGEN: AIR MIXTURES IN THE VGES CYLINORICAL TANK. Vibration NUREG/CR-3939: MATER HAMMER, FLOW INDUCED VIBRATION AND SAFETY / RELIEF VALVE LOADS. Maste Package NUREG/CR-2482 V05: REVIEn OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 - National Waste Package Program, April 1983 - September 1983. NUREG/CR-3900 V01: LONG-IERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PAC 6 AGING.First Quarterly Report, Year Three, April-June 1984 Weste Repositories NUREG/CR-3758: CROSSHOLE GLOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGd LEVEL HASTE REPOSITORIES. Water Chemistry NUREG/CR-3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT-WATER-REACTOR SAFETY RESEARCH PROGRAM Quarterly Progress Report, January =Maren 1983 nater Hammer NUREG/CR-3895: INVESTIGATION OF COLD LEG WATER HAMMER IN A PWR DUE TO THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. NUREG/CR-3939: WATER HAMMEH, FLOW INDUCED VIRRATION AND SAFETY / RELIEF VALVE LOADS. neldment NUREG/CR-3788 V01: STRUCTURAL INTEGRITY OF LIGHT WATER REACTOR PHESSURE BOUNDARY COMPONENTS.Four-Year Plan 1984-1988 NUREG/CR-3694: ULTRASONIC AN0 METALLURGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS STEEL BAR PIPE nELOMENT. Mhole Rod Testing NUREG/CR-3708: LWR SPENT FUEL DRY STORAGE BEHAVIOR AT 229 C. Workshop NUREG/CR-3739: THE OPERATOR FEEDBACK WORKSHOP A TECHN!QUE FOR OBTAINING FEED 8ACK FROM OPERATIONS PERSONNEL. Zircatoy NUREG/CR-3459: EXPERIMENT DATA REPORT FOR HULT! ROD BURST TEST (MNBT) 122 {
SUNDLE 8*5 NUREG/CR-3460s EXPERIMENT DATA REPORT FOR MULTIR00 BURST TEST (MRST) SUNDLE. 8=6.-. l I l { l 2 I i I, i 4 1 123 l _ _,_______._ _ _..,.~ _ _
NRC Originating Organization Index (Staff Reports) This index lists those NRC organizations that have published staff reports. The Index Is arranged alphabetically by major NRC organizations (e.g., program offices) and then by subsections of these (e.g., divisions, branches) where ap-propriato. 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. EDO - 0FFICE OF ADMINISTRATION DIVISION OF TECHNICAL INFOMMATION E DOCUMENT CONTROL NUHEG-0304 V09 N02: HEkULATORY AND TECHNICAL REPORTS. Compilation For Second Quarter 1984 NUREG-0540 V06 N05: IITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.May 1-31, 1964 NUREG-0540 V06 N06: TITLL LIST OF DOCUMENTS MADE PUBLICLY AVAILAoLE. JUNE 1-30, 1984 NUREG-0540 V06 N07: TITLL LIST OF DOCUMENTS MADE PUBLICLY AVAILAbLL July 1-31, 1984 NUREG-0750 V19 101: INDEAES TO NUCLEAR REGULATORY COMMISSIUN ISSUANCES FOR JANUARY-MANCH 1984 NukEG-0750 V19 NO3: nUCLLAN REGULATORY COMMISSION ISSUANCES FOR MARCH 1984 Pages 555-936 NUREG-0750 V19 N04: NUCLLAd HEGULATORY COMMISSION ISSUANCES F0H APHIL 1984.Pages 937-1,149 NUNEG-0750 V19 N05: NUCLLAN REGULATORY COMMISSION ISSUANCES FOR MAY 1984.Peges 1,151-1,341. DIv!SIUN OF RULES AND HECONDb HUREG-0936 V03 N02: 4RC NE(,ULATORY AGENDA.Guarterly Report, April-June 1984 E00 - 0FFICE OF EXECUTIVL LEGAL DIRECTOR OFFICE OF THL EXECUTIVL LEGAL DIRECTOR NUREG-0306 003: UNITLD SLATES NUCLEAR REGULATORY CCMMISSION STAFF PRACTICE AND PROCE00HE DIGEST. EDO - 0FFICE FOR ANALYSIb 6 LVaLUATION OF OPERATIONAL DATA DIHECTOR'S OFFICE NUREG-0090 V07 N01: HEPONT TO CONGRESS ON ABNORMAL OCCURRLNCES. January = March 1984 OFFICE OF INSPECTION & EhFORLEMENT (POST 12/11/80) 125
DIRECTOR'S OFFICE, OFFICL OF INSPECTION AND ENF0HCEMENT NUREG-0430 V04 N02: LICENSLO FUEL FACILITY STATUS REPORT. Inventory Difference Dato, July 1983-December 1983.(Buff book) ENFORCEMENT STAFF NUNEG-0940 V03 N02: ENF0HCLMENT ACTIONS SIGNIFICANT ACTIONS RESOLVED. Quarterly Progress Report, April-June 1984 DIVISION OF QA, SAFEGUARDS & INSPECTION PROGRAMS (POST 830103) NUREG-0040 V08 N02: LICENSLE CONTRACTOR AND WENDUR INSPECTION STATUS REPORT. Quarterly Report, April-June 1984.(nhite book). OFFICE OF NUCLEAR REGULATONY RESEARCH (POST 4/05/81) 0FFICE OF NUCLEAR REGULAf0HY RESEARCH, DIRECTOR NUREG-1080 V01: LONG-RANGE RESEARCH PLAN FY 1985-1989 DIVISION OF RISK ANALY3Is 6 UPERATIONS (POST 840429) NUREG-0935: ACQUSTIC WAVL PROPAGATION IN FLUIDS WITH COUPLLD CHEMICAL REACTIONS. NUREG-1050: PROBABILISTIC HISK ASSESSMENT (PRA) REFERENCE DOCUMENT. Final Rept. DIv!SION OF RADIATION PROGHAMS & EARTH SCIENCES (POST 840429) NUREG-1029: A COMPUTER COOL FOR GENERAL ANALYSIS OF NA00N RISKS (GARH). DIVISION OF ENGINEERING TELHNOLOGY NUREG-1092 ENVIRONMLNTAL ASSESSMENT FOR 10 CFR PANT 72, " LICENSING HEOUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT FULL AND HIGH-LEVEL RADIDACTIVE WASTE." NUkEG/CP-0051: PROCELDINuS OF THE CSN! SPECIALIST MEETING ON I. LEAK-BLFORE-bREAn IN NUCLEAR REACTOR PIPING. INTRA AGENCY COMMITTEES, REVIEn GROUPS, ETC. PIPING REVIEn COMMITTEL NUREG-1061 V01: HEP 0d7 0F THL U.S. NUCLEAR REGULATOPY COMMISSION PIPING REVIEn COMMIT 1EL. Volume 1 Investigation And Evaluation Of Stress Corrosion Cracking In Piping Of Boiling Water Reactor Plants. EDO-RESOURCE MANAGEMENT OFFICE OF HESOURCE MANAGtMLNT, DIRECTOR NUREG-0748 V04 N05: UPcRATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As of May 31,1984.loranne Book) NUREG-0748 V04 406: UPERATANG HEACTORS LICLNSING ACTIONS SUMMANY. Data As of June 30,1984 (Orange book) NUNEG-0748 V04 N07: OPLRATING HEACTOHS LICENSING ACTIONS SUMMANY. Data As Of July 31,1954.(Urange Book) DIVISION OF BUDGET & AHALYa!S NUREG-0020 V06 N06: LICLHSLO OPERAT!kG REACTURS STATUS
SUMMARY
REPT. Data As of May 31,1984.(Grey book) NUREG-00d0 V08 N07: LICENSLD OPERATING RLACTORS STATUS SUMMANY REPORT. Data As Of June 30,1984.(Grey Book) 126
NUREG-0020 VOS N08: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT.Date As Of July 31,1984.(Grey Book) 0FFICE OF NUCLEAR REACT 0H HEGULATION (POST 4/28/80) 0FFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR NUREG-0767 307: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WATERFORD NUCLEAR P0nEN PLANT,0 NIT 3. Docket No. 50-382 (Louisiana Power & Light Company) ) NUREG-0800 06.2.1 R6 STANDARD REVIEh PLAN FOR THE REVIEW OF SAFETY ] ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition Revision 6 to [ Section 6.2.1.1.C, " Pressure-Suppression Type BhR Containments." DIVISION OF ENGINEERING NUREG-1054: SIMPLIFIED ANALYSIS FOR LIQUID PATHWAY STUDIES. NUREG-1061: POST-ACCIDENT GAS GENERATION FROM HADIALYSIS OF ORGANIC MATERIALS. DIVISION OF HUMAN FACTURS SAFETY NUREG-0985 R01: U.S. NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN. DIVISION OF SYSTEMS INTEGRATION (POST 811005) NUREG-0978: MARK III LOCA-RELATED HYDRODYNAMIC LOAD DEFINITION. Generic Technical Activity B-10. Final Report. DIVISION OF LICENSING NUREG-0420 S06: SAFETY EVALUATION REPORT RELATED TO THE UPERATION OF SHOREHAM NUCLEAR POWLR STATION, UNIT NO. 1. Docket No. 50-322 (Long Island Lighting Company) NUREG-0420 307: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SHOREHAM NUCLEAR POWER STATION UNIT NO.1. Docket No. 50-322 (Long Island Lighting Company) NUREG-0675 324 SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR P0nER PLANT, UNITS 1 & 2. Docket Nos. 50-275 & 50-323.(Pacific Gas e Electric Company) NUREG-0675 325: SAFEIY EVALUATION REPORT RELATED To THE OPERATION OF DIABLO CANYON NUCLLAH PonER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company) NUREG-0675 326: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR P0nER PLANT,0 NITS 1 AND 2 NUREG-0675 S2T SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company) NUREG-0680 805: TMI-1 RESTART.An Evaluation of The Licensee's Management Integrity As It Affects Restart Of Three Mile Island Unit 1 Docket 50-269 NUREG=0798 304: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF ENRICO FERMI ATOMIC P0pFR PLANT, UNIT NO. 2. Docket No. 50-341. (Detroit Edison Company) NUREG-0831 305: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF GRAND GULF NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-416 Ans 50-417.(Mississippi Power And Light Company, Middle South Energy,Inc And South Mississippi Electric Power Association) NUREG=0831 S06: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF GRAND GULF NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-416 And 50-417 (Mississippi Power And Light Company) NUREG-0926 R01: TECHNICAL OPECIFICATIONS FOR GRAND GULF NUCLEAR STATION, UN!1 1. Docket No. 50-416. (Mississippi Power And Light Company) NUREG-0954 303: SAFEIY EVALUATION REPORT RELATED TO THE OPERATION OF 127
CATAWSA NUCLEAR STATION, UNITS 1 AND 2.00CKET Nos. 50=413 And 50-414.(Duke Power Company,et al) 1 NUREG-1031: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF j MILLSTUNE NUCLEAR PodER STATION, UNIT NO. 3. Docket No. 50-423.(Nortneast Nuclear Energy Company) j NUREG=10648 DRAFT ENVIRONMkNTAL STATEMENT RELATED To THE OPERATION OF 1 MILLSTONE NUCLEAR POWER STATION, UNIT 3. DOCKET NO. 50-423. (NORTHEAST NUCLEAR ENERGY COMPANY,et al) NUREG=10693 SAFETY EVALUATION REPORT RELATED TO THE RENEdAL UF THE OPERATING LICENSE FOR THE GENERAL ELECTRIC = NUCLEAR TEST 4 REACTOR (GE=NTR). DOCKET NU. 50-73.(General Electric Company) NUREG-10T2 TECHNICAL SPECIFICATIONS FOR CATAWDA NUCLEAR STATION, Unit 4 i
- 1. Docket No. 50=413 l
NUREG=10733 DRAFT ENVINONMENTAL STATEMENT RELATED TO THE OPERATION OF RIVER dENO STATION. Docket No. 50-458.(Gulf States Utilities Company & Cajun Electric Power Cooperative) NUREG-1075: DECENTRALIZAI!UN OF UPERA11NG REACTOR LICENSING i REVIEWS.NRR Pilot Program. NUREG-1033: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE 4 OPERATING LICENSE FOR THE nESTINGHOUSE RESEARCH REACTOR AT ZION,ILLINDIS. DOCKET NU. 50=87.(nestinghouse Electric Company) NUREG-1084: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REACTOR AT MICHIGAN STATE UNIVERSITY.0ccket do. 50-294 NUREG-1085: ORAFT ENVIHONMENTAL STATEMENT RELATED TO THE OPERATION OF i NINE MILE POINT NUCLEAR STATION, UNIT NO.2. Docket No. 50-410 (Niagara Mohawk Power Corporation, Rochester Gas & Electric Corporation And Central Hudson Gas & Electric Corporation) DIVISION OF SAFETY TECHNULUGY j NUREG=0606 V06 NO3: uNHESOLVED SAFETY ISSUES
SUMMARY
. Data As Of August 17,1984 (Aque dook) NUREG=0649 R01: TASK ACTION PLANS FOR UNRESOLVED SAFETY ISSUES RELATED TO NUCLEAR POWER PLANTS. NUREG=0933 S01: A PR10NITIZATION OF GENERIC SAFETY ISSUES. NUREG-1068: REVIEW INSIGHTS ON THE PROBASILISTIC RISK ASSESSMENTS FOR THE LIMERICK GENERATING STATION, UNIT 1 AND 2. j i l 128
NRC Contract Sponsor index (Contractor Reports) This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC l organization (e.g., program office) and then by subsections of these (e.g., l divisions) where appropriate. The sponsor organization is followed by the NUREG/CR number and title of the report (s) prepared by that organization. If i further information is needed, refer to the main citaticn by the NUREG/CR number. E00 = OFFICE FOR ANALYSIS a EVALUATION OF OPERATIONAL DATA 0! RECTOR'S OFFICE PCJREG/CR=2000 V03 N6 LICENSEE EVEr4T REPORT (LER) COMPILATION For Month of June 1984 N'JREG/CR=2000 V03 N7 LICENSEE EVENT REPORT (LER) Comptistion For Month of July 1984 NUREG/CR=2000 V03 N8 LICENSEE EVENT REPORT (LER) Compilation For Month Of August 1904 NUREG/CR=3824: CONTING PHOGRAM GUIDE. huREG/CR=3905: SEQUENCE Cou!NG AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS Users Guice. NUREG/GR=4011: THE 21/h5 DATA BASE USER'S MANUAL. OFFICE OF INSPECTION & El4FORCEMENT (POST 12/11/80) DIVISION OF EMERGENCY PREPARLDhESS & ENGINEERING RESPONSE (POST 830103) NUREG/CR=3053: CLOSEQUT uF IE BULLETIN 80-08 EXAMINATION OF CONTAlhMENT LINER PENETRATION nELDS. huREG/CR=3792: CLOSEuur UF IE BULLETIN 79-II FAULTY OVERCURRENT TRIP DEVICE IN CIRCU!i dREAnENS FOR ENGINEERED SAFETY SYSTEMS. NUREG/CR=3195: CLOSEUUf UF IE BULLETIN 82-04 DEFICIENCIES IN PRIMARY CONTAINMENT ELECTRICAL PENETRATION ASSEMBLIES. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS DIVISION OF FULL CYCLE & MATERIAL SAFETY NUREG/CR=2499 REVIEd uF EMERGENCY RADIOLOGICAL INSTRUMENTATION AND ANALYTICAL METHODS AT NMSS= LICENSEE SITES. URAN!UM FUEL LICENSING BRANCn NUREG/CR=3796: EMERGENLY PHEPAREDNESS SOURCE TERM DEVELOPMENT FOR THE OFFICE :P NUCLEAR HAIEHIALS SAFETY AhD SAFEGUARDS LICENSED FACILITIES. DIVISION OF WASTE MANAGEMENT NUREG/CR=2482 V05: REVIEn UF 00E WASTE PACKAGE PROGRAM. Subtask 1.1 = National Weste Pacxage Program, April 1983 - September 1983 129
NUREG/CR=3763: REVIEn AND ASSESSMENT OF HADIONUCLIOE SORPTION INFORMATION FOR THE uASAsT WASTE ISOLATION PROJECT SITE (1979 Through May,1983). NUREG/CR=3844: CHARACTtRIZATION OF THE RADI0 ACTIVE WASTE PACnAGES UF THE MINNESOTA MINING AHD MANUFACTURING COMPANY, NUREG/CR=3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEUCHEMICAL INFORMATION DEVELOPED dY DOE HIGH= LEVEL NUCLEAR nASTE HEPOSITONY SITE PROJECTS. Report for October =Decemoer 1983. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81) 0FFICE OF NUCLEAR REGULATONY RESEARCH, DIRECTOR NUREG/CR=3469 V01: OCCUPATIONAL 00SE REDUCTION AT NUCLEAR P0nEN PLANTS ANNOTATLD BIBLIOGHAPHY OF SELECTED READINGS IN RADIAT10N PROTECTION AND ALAHA. NUREG/CR=3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LIGHT =MATERaREACTON SAFETY RFSEARCH PROGRAM Quarterly Progress Report, January-Maren 1983. ACCIDENT SOURCE TERM PH0wRAM OFFICE NUREG/CR=3929: LOSS =UF-BLNLFITS ANALYSIS FOR NUCLEAR PUnER PLANT SHUTD0nNS. Methodology And Illustrative Case Stuoy. DIVISION OF ACCIDENT EVALUATION (PRE 840101) NUREG/CR=3821: EVALUATION OF CRACK PLAkE Euu1LIBR!uM H0 DEL FOR PREDICTING PLASTIC FNACTURE. DIVISION OF ACCIDENT EVALUATION NUREG/CR=0169 V17 LUFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recoroeo On DAVDS. NUREG/CR=2331 V03 N3 SAFETY RESEARCH PROGRAMS SPONSORED BY UFFICE OF NUCLEAR REGULATORY RLSLANCH. Quarterly Progress ReportrJuly-Septemper 1963. NUREG/CR-2576: OnR FULL INTEGRAL SIMULATION TEST (FIST)== Facility Description Report. NUREG/CR=3169: SUPER SISTEM CODE (SSC,HEV. 0).AN ADVANCED THERM 0MYDRAULIC SIMULATIUN FOR TRANSIENTS IN LMFbRS. NUREG/CR=3190: PLUGM A C0uPLED THERMAL-HYDRAULIC COMPUTER MUDEL FUR FREEZING MELT FL0n IN A CHANNEL. NUREG/CR=3273: COM8UST10d UF HYDROGEN: AIR MIXTURLS IN THE VGES CYLINDRICAL TANK. NUREG/CR=3369: AN UNCENTAINTY STUDY OF PnR STEAM EXPLOSIUNS. I NUREG/CR=3459: EXPERIMcNT DATA REPORT FOR HULTIROD BURST TEST (MHOI) ouh0LE B-5. NUREG/CR=3460 EXPERIMLNT UAT A REF URT FOR MULTIROD RuRST TEST (MHBT) BUNDLE 8=6. NUREG/CR=3470 ATWS AT BH0nNS FERRY UNIT ONE = ACCIDENT SEGUENCE ANALYSIS. NUREG/CR=3492 V04: HIGH-TEMPERATURE GAS =COULED RLACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIUENT EVALUATION QUARTERLY PROGHESS RtPORT, October = December 1983. NUREG/CR=3589 V01: REACTUR SAFETY RESEARCH QUARTERLY REP 0HT. January = March 1983. NUREG/CR=3589 V02: RLACTUR SAFETY RESEARCH GUARTLRLY REPORT. April-June 1963. NUREG/CR=3617: NOBLE GASr!UDINE,AND CES!UM TRANSPORT IN A POSTULATED LOSS OF DECAY NEAT RLMUVAL ACCIDENT AT BR0aNS FERRY. NUREG/CR=3654: PnR FLECHI SEASET SYSTEMS EFFECTS NATURAL CIRCULATIUN AND REFLUX CONDENSATION.bata Eva)ustion ano Analysis Report NRC/LPHI/ Westinghouse Neport No. 14. NUREG/CR=3662: FvEL-0ISRUPTIUN EAPERIMENTS UNDLR HIGH= RAMP = RATE 130
HEATING CONDITIONS. NUREG/CR=3679: CALIBRAlION AND QUALIFICATION OF THE LOS ALAMOS FAILURE MODEL (LAFM). NUREG/CR=3690 RELAPS ASSESSMENT SEMISCALE NATURAL CIPCULATION TESTS S=NC=3,S=NC=4,AND S=NC=8.. NUREG/CR=3711: SnR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST RESULTS. NUREG/CR-3724: ULTIMATE STRENGTH ANALYSES OF THE WATTS SAR, MAINE o YANKEE,AND BELLEFONTE CONTAINMENTS. NUREG/CR=3734: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report,0ctober 1982 = Herch 1983 NUREG/CR=3761: RELAP5 lHERMAL= HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR THE OCONEE-1 PRESSURIZED WATER REACTUR. NUREG/CR=3765: MINET SIMOLATION OF A HELICAL COIL 30010M/ WATER STEAM GENERATOR, INCLUDING STRUCTURAL EFFECTS. NUREG/CR=3776: TESTING OF SAFETY =RELATED NUCLEAR POWER PLANT EQUIPMENT AT THE CENTRAL RECEIVER TEST FACILITY. NUREG/CR=3804 V01: PHYSICS OF REACTOR SAFETY. Quarterly Report January - Marcn 1984 NUREG/CR=38133 MINET VALIDATION STUDY USING STEAM GENERATOR TRAN61LNT DATA. NUREG/CR=3820 V01: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM. Quarterly Report, January = March 1984 NUREG/CR=3822 SOLA-PTS: A Transient,Three= Dimensional Algorithm For Fluid = Thermal Mixing And Wall Heat Transfer In Complex Geometrics. NUREG/CR=3830 V01: AEROSuL RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR UCT0bER 1983 - MARCH 1984 NUREG/CR=3835: SIMULATION OF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE WISCRETE VORTEX METHOD. NUREG/CR=3845: PREDICTION OF NONLINEAR STRUCTURAL RESPONSE IH LMERN ELEVATLD-TEMPERATURE PIPING. NUREG/CR-3868: CONTAANMENT BUILDING ATMOSPHERE RESPONSES DUE TU REACTOR GAS $URNING UNDER SEVERE ACCIDENT CONDITIONS, NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYSTEM OF THE EXPERIMENTAL 8REEDER REACTOR =II. NUREG/CR=3895: INVESTIGATION OF COLD LEG WATER HAMMER IN A PAR DUE TO THE ADMISSION OF EGC OURING A SMALL BREAM LOCA. O! VISION OF FACILITY OPERAi!UNS NUREG/CR=3346: 810 ASSAY DAIA AND A RETENTION = EXCRETION MODEL FOR SYSTEMIC PLUTONIUM. NUMEG/CR=3418: SCREENING TESTS OF TERMINAL BLOCK PERFORMANCE IN A SIMULATED LOCA ENVIRONMENT. NUREG/CR=3513: MECHANICAL RELIABILITY EVALUATION OF ALTERNATE m0 TORS FOR USE IN A RADIO 100!NE AIR SAMPLER. NUREG/CR=3518 V01: SLIM-MAUD:AN APPROACH TO ASSESSING HUMAN ERROR PROSASILITIES USING STRUCTURED EXPERT JUDGEMENT. Volume I 0verview of SLIM.MAUO. NUREG/CR=3520 V01: LUNG-TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUAR05 AT NUCLEAR POWER PLANTS. Volume I Summary And Users Guide. NUREG/CR=3520 V02: LUNG = TERM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARUS AT NUCLEAR POWER PLANTS. Volume I! Development of Detailed Analyses. NUREG/CR=3524: ORGANIZATIONAL INTERFACE IN REACTOR EMERGENCY PLANNING AND RESPONSE. NUREG/CR=35448 8 ETA PARTICLE MEASUREMENT AND DOSIMETRY AT NRC= LICENSED FACILITIES. NUREG/CR=3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELOS AT COMMERCIAL NUCLEAR SITES. NUREG/CR=3590: EVALUATION UP ISOTOPE DILUTION MASS SPECTROMETRY FOR 131
810A88AY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URINE. NUREG/CR=3678: ESTIMATION METH003 FOR PROCESS HOLDUP OF SPECIAL NGCLEAM MATERIALS. NUREG/CR*3750: JOS ANALYSIS OF NUCLEAR P0MER REACTOR HEALTH PHYSICS TECHNICIANS. NUREG/CR=3786: A REVIEn 0F REGULATORY REQUIREMENTS GOVERNING CONTROL ROOM HABITASILITY SYSTEMS. NUREG/CR=3856: AN ULTRASONIC LEVEL AND TEMPERATURE SENSOR FOR P0nEN REACTOM APPLICATIONS. DIVISION OF HEALTH, SITING & WASTE MANAGEMENT NUREG/CR=3714: ON THE uEVELOPMENT OF ENVIRONMENTAL RADIATION STANDAR03 FOR GEOLOGIC DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES. NUREG/CR=3896: SIMULATION LXPERIMENTS COMPARING ALTERNATIVE PRUCESS FORMULATIONS USING A FACTORIAL DESIGN. NUREG/CR=3897: EVALUATION OF ECOSYSTEM SIMULATION MODELS AS TOOLS FOR ASSESSMENT OF POAER PLANT IMPACTS ON FISH POPULATIONS. Final Rept. DIVISION OF RISK ANALYSIS 6 UPERATIONS (POST 840429) NUREG/CR-1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED INSTRUMENTATION AND CONTROL COMPONENTS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS JANUARY 1,1976 TO DECEM8ER 31,1981. NUREG/CR=3139: SCENARIOS AND ANALYTICAL METHODS FOR UF6 RELEASES AT NRC= LICENSED FUEL CYCLE FACILITIES. NUREG/CR-3591 V01: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PREGUNSuRS TO POTENTIAL SEVERE CONE DAMAGE ACCIDENTS: 1980=1981 A Status Report. NUREG/CR=3655: A METHOD FON ANALYTICAL EVALUATION OF COMPUTER =uASED DECISION AIDS. NUREG/CR=3867: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTENS AT U.S. COMMERCIAL NUCLEAR POWER PLANTS, JANUARY 1,1976 TO DECEMBER 31,1982 " NUREG/CR=3932: SENCHMARK DESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. NUREG/CR=3933: RISK RELATED RELIABILITY REQUIREMENTS FOR BnR SAFETY =IMPORTANT SYSTEMS MITH EMPHASIS ON THE RESIDUAL HEAT REMOVAL SYSTEM. NUREG/CR=3988: MARCH 2 (MELTD0nN ACCIDENT RESPONSE CHARACTERISTICS) CODE DESCRIPTION AND USERS MANUAL. DIVISION OF RADIATION PRUGHAMS & EARTH SCIENCES (POST 840429) NUREG/CR=3610 NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of Subtask C 3He Neutron Spectrometer. NUREG/CR=3665: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR P0nER PLANTS. Executive Summary. NUREG/CR=3665 V01: OPTIM1ZATION OF PU8LIC AND OCCUPATIONAL RA01AT10N PROTECTION AT NUCLEAN P0nER PLANTS.A Review of Occupational pose Assessment Considerations In Current Probabilistic Risk Assessment And Cost =Senefit Analyses. NUREG/CR=3665 V02: OPTIM1ZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonER PLANTS. Considerations In Factoring Occupational Dose Into Value= Impact And Cost =Senefit Analyses. NUREG/CR=3665 V03: OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAN P0nER PLANTS.A Calculation Method. NUREG/CR=3751: EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FL000 PROTECTION COSTS FOR UNANIUM TAILINGS IMPOUNDMENTS. NUREG/CR=3758: CROSSNOLE GEOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL HASTE REPOSITORIES. NUREG/CR=3798: CHARACTLRIZATION OF CEMENT AND SITUMEN nASTE FONMS 132
. ~. _ _ CONTAINING SIMULATED LUW LEVEL WASTE INCINERATOR ASM. NUREG/CR=3812: ASSESSMENT UF IRRADIATION EFFECTS IN RADWASTE CONTAINING ORGANIC ION = EXCHANGE MEDIA. NUREG/CR=3832: UNCERIAINTIES IN LONG= TERM REPOSITORY PLRFORMANCE DUE TO THE EFFECTS OF FUTUHE GEOLOGIC PROCESSES. NUREG/CR=3871 AN OVERVILW OF THE UNIFIED TRANSPORT APPROACH. NUREG/CR=3874: NEAR-GRUUND TORNADO NIND FIELDS. NUREG/CR=3900 V01: LONG=IEHM PERFORMANCE OF MATEHIALS USED FOR HIGH-LEVEL WASTE PACKAGING.First Quarterly Report, Year Three, April = June 1984 NUREG/CR-3940: FIELD EAPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN ALKALINE LANE ENVIRONMENTS. NUkEG/CR=3951: INTHODUCTION TO BIBELOT A BIBLIOGRAPHIC FINDING AND RETRIEVAL SYSTEM. DIVISION OF ENGINEERING TECHNOLOGY NUREG/CR=0130 A0003: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENQE PRESSUNIZE0 nATER REACTOR PonER STATION. NUREG/CR=0672 ADD 02: TECHNULOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING HATEN HEACTOR P0hER STATION. Classification Of Decommissioning Aastes. NUREG/CR=2015 V08: PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT VII). Seismic Safety Mergins Research Program. NUREG/CR=2996: SENSIT!v!TY OF DETECTING IN-CORE VIBRATIONS AND BOILINW IN PRESSURIZED WATER REACTORS USING EX-CORE NEUTRON NOISE. NUREG/CR=3228 V02: STRUCTURAL INTEGRITY OF NATER REACTOR PRESSURE BOUNDARY COMPONENTS. Annual Report For 1983 NUREG/CR=3318: LnR PNESSURL VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM PCA Experiments,ulind Test,And Pnysics= Dosimetry Support For The PSF Experiments. NUREG/CR=3599: SOURCES OF UNCERTAINTY IN THE CALCULATIONS OF LUAUS ON SUPPORIS OF PIPING SYSTEMS. NUREG/CR=3618: OCA=P,A DLTERMINISTIC AND PROBABILISTIC FRACTURE = MECHANICS CODE FOR APPLICATION TO PRESSURE VESSELS. NUREG/CR-3643: HETEROGENc003 OXIDATIVE DEGRADATION IN IRHADIATED POLYNENS. NUREG/CR-3660 V02: PH0dA6!LITY OF PIPE FAILURE IN THE HEACTON COOLANT LOOPS OF WESTINGHOUSE PHH PLANTS. Volume 2 Pipe Failura Induced By Crack Growth. NUREG/CR-3663 V02: PROSAd!LITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS UF COMBUST!01 ENGINEERING PHR PLANTS.Vol 2 Pipe Failure Induced by Crack Growtn. NUREG/CR=3671: ASSESdMENT OF RADIATION EFFECTS RELATING TO REACTOR PRESSUME VESSEL CLADOING. NUREG/CR=3689 V02: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHTanATER REACT 0H SAFETY RESEARCH PROGRAM. Quarterly Prouress Report, April-June 1983 NUREG/CR=3689 V03: MATERIALS SCIENCE AND TECHNOLOGY DIVISION LIGHTanATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress Report, July-September 1983. NUREG/CR=3692: POSSluLE N00CS OF STEAM GENERATOR OVERFILL RESULTING FROM CONTROL SYSTEM MALFUNCTIONS AT OCONEE=1 NUCLEAR PLANT. NUREG/CR-3708: LNR SPENT FUEL DRY STORAGE BEHAVIOR AT 229 C. NUREG/CR=3742: BUCKLING OF STEEL CONTAINMENT SHELLS UNDEH TIME-DEPENDENT LOADING. NUREG/CR=3744 V01: HEAVY =SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT F0H OCf0 DER 1983 = MARCH 1984 NUREG/CR=3766: TESTING OF NUCLEAR GRADE LUBRICANTS AND THEIR EFFECT ON A540 AND A193 81 uCLTING MATERIALS. NUREG/CR=3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA PELLETHON=HASTER SYSTEM. 133
l l -NUREG/CR=3787: EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESF) SYSTEMS IN RETAINING FISSION PRODUCTS.Beckground Information. NUREG/CR=3784 V01: STRUCTURAL INTEGRITY OF LIGHT NATER REACTOR PRESSURE-SOUNDARY COMPUNtNTS.Four= Year Plan 1984=1988 NUREG/CR=3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS: Annual Report,0ctober 1982 - September 1983. NUREG/CR=3815: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORR-PSF =PVS IRRADIATION EXPERIMENT. NUREG/CR=3818: REPORT OF RESULTS OF NUCLEAR POWER PLANT AGING WORMSHOPS. NUREG/CR=3626: RECOMMENDATIONS FOR PROTECTING AGAINST FAILURE BY BRITTLE FRACTURE IN FERRITIC STEEL SHIPPING CONTAINERS GREATER THAN FOUR INCHES THICK. NUREG/CR=3833: BEHAVIOR OF SUSCRITICAL AND SLOM-STABLE CRACK GRONTn FOLLOWING A POST = IRRADIATION THERMAL ANNEAL CYCLE. NUREG/CR=3642: STEAM GENERATOR GROUP PROJECT TASK 8 - SELECTIVE TU8E UNPLUGGING. NUREG/CR=3843: STEAM GENERATOR GROUP PROJECT TASK 10 - SECONDARY SIDE EXAMINATION. NUREG/CR=3869: ANALYSIS OF THE IMPACT OF INSERVICE INSPECTION USING A PIPING RELIABILITY MODEL. NUREG/CR=3884: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM = THRtE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. NUREG/CR=3888: ANALYSIb QF THE VENUS PnR ENGINEERING MOCnUP EXPERIMENT = PHASE I: SOURCE DISTRIBUTION. NUREG/CR=3892: A RESEARCH PROGRAM FOR SEISMIC GUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Summary Report. NUREG/CR=3892 V01: A RESEARCH PROGRAM FOR SEISMIC WUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey Of Methods For Equipment And Components Evaluation of MethodologypGualification And Methodology.... NUREG/CR=3892 v02: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 2= Correlation Of Metnodologies For Seismic Qualification Tests of Nuclear Plant Equipment. NUREG/CR=3892 V03: A RtSEARCH PROGRAM Fok SEISMIC uuALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3= Recommendations For Improvement Of Equipment Qualification Methodology And Criteria. NUREG/CR=3892 V04: A RtSEARCH PROGRAM FOR SEISMIC GUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 = ine Use Of Fragility In Design of Nuclear Plant Equipment. NUREG/CR=3893: LABORATORY STUDIES DYNAMIC HESPONSE OF PROTOTYPICAL PIPING SYSTEMS. NUREG/CR=3894: ULTRA 30n!C AND METALLURGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS STEEL BWR PIPE nELDMENT. NUREG/CR=3899 UTILITY FINANCIAL STABILITY AND THE AVAILABILITY OF FUNDS FOR DECOMMIss!UNING. NUREG/CR=3921: DRY SPENT FUEL STORAGE TEST PLAN FOR FINAL NONDESTRUCTIVE FUEL HOW CXAMINATION. EDO= RESOURCE MANAGEMENT DIVISION OF BUUGET & ANALYSIS NUREG/CR=3840: COST ANALYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE ADILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKUUT. OFFICE OF NUCLEAR REACTON REkULATION (POST 4/28/80) 0FFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR 134
NUREG/CR=34$9 V01: OCCUPATIONAL DOSE REDUCTION AT NUCLEAR P0cER PLANTS ANNOTATED SISLIOGRAPHY OF SELECTED READINGS IN RADIATION PROTECTION AND ALARA. DIVISION OF ENGINEERING NUREG/CR=3834: ON THE THRESHOLD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRACKING OF SENSITIZED ALLOY 600 IN BORATED THIOSULFATE SOLUTION. DIVISION OF HUMAN FACTORS 3AFETY NUREG/CR-3739: THE OPERATOR FEEDBACK WORKSHOP:A TECHNIQUE FOR OBTAINING FEEDBACK FROM QPERATIONS PERSONNEL. DIVISION OF SYSTEMS INTEGRATION (POST 811005) NUREG/CR=3907: GT2R28AN UPDATED VERSION OF GAPCON-THERMAL-2 NUREG/CR=4001: CONTEMPT 4/ MODS AN IMPROVEMENT TO CONTEMPT 4/ MOD 4 MULTICUMPARTMENT CUNTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE CONTAINMENT ANALYSIS. NUREG/CR=4007: LOWER LIMIT OF DETECTION: DEFINITION AND ELABORATION OF A PROPOSED POSITION FOR RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL MEASUREMENTS. DIVISION OF SAFETY TECHNOLOGY NUREG/CR=3474: LONG= LIVE 0 ACTIVATION PRODUCTS IN REACTOR MATERIALS. NUREG/CR=3480: VALUE/ IMPACT ASSESSMENT FOR SEISMIC DESIGN CRITERIA USI A=40 NUREG/CR=3493: A REVIEn UF THE LIMERICK GENERATING STATION SEVERE ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequency. NUREG/CR=3593 V01: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR POWER PLANT'S HIGH PRESSURE SAFETY i INJECTION SYSTEM. NUREG/CR=3593 V02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX ANALYSIS OF A NUCLEAR POWER PLANT'S HIGH PRESSURE SAFETY INJECTION SYSTEM. Volume 2 NUREG/CR-3852: INSIGHT INTO PRA METHODOLOGIES. NUREG/CR=3939 WATER HAMMER, FLOW INDUCED VIBRATION AND SAFETY / RELIEF VALVE LOADS. 135
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. ANALYSIS & TECHNOLOGY, INC. NUREG/CR-3750: JOB ANALYSIS OF NUCLEAR POWER REACTOR HEALTH PHYSICS TECHNICIANS. ANCO ENGINEERS, INC. NUREG/CR-3893: LABORATURY STUDIES DYNAMIC RESPONSE OF PROTOTYPICAL PIPING SYSTEMS. ARGONNE NATIONAL LABORATURY NUREG/CR-3689 V01: MATERIAL SCIENCE AND TECHNOLOGY DIVISION LluHT-HATER-REACTOR SAFETY RESEARCH PROGRAM Quarterly Progress Report, January-Maren 1983 NUREG/CR-3689 V02: MATLRIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT HATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress Report, April-June 1983. NUREG/CR-3689 V03: MATLRIALS SCIENCE AND TECHNOLOGY DIVISION LIGHT = HATER REACTOR SAFETY RESEARCH PROGRAM. Quarterly Progress Report, July-September 1983 NUREG/CR-3804 V01: PnYdICS OF REACTOR SAFETY. Quarterly Report January = March 1984 NURE4/CR-3806: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS: Annual Report,0ctober 1982 - September 1983 NUREG/CR-3894: ULTRASONIC ANu METALLUdGICAL EXAMINATION OF A CRACKED TYPE 304 STAINLESS STEEL BWR PIPE nELOMENT. NUREG/CR-3929: LOSS-OF=BENtFITS ANALYSIS FOR NUCLEAR POWER PLANT SHUT 00nNS. Metnodology And Illustrative Case Study. NUREG/CR-3932: SENCHMARK OESCRIPTION OF CURRENT REGULATORY REQUIREMENTS AND PRACTICES IN NUCLEAR SAFETY AND RELIABILITY ASSURANCE. NUREG/CR-3933: RISK RELATE 0 HELIABILITY REQUIREMENTS FOR BnR SAFETY -!MPORTANT SYSTEMS WITH EMPHASIS ON THE RESIDUAL HEAT REMOVAL SYSTEM. ASPEN SYSTEMS, INC. 7 huREG-0386 003: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDUNE DIGEST. BATTELLE HUMAN AFFAIRS RESEARCH CENTERS NUHEG/CR-3739: THE OPERATOR FEED 8ACK WORKSHOP:A TECHNIQUE FOR OBTAINING FEEDSACK FROM GPERATIONS PLRSONNEL. BATTELLE MEMORIAL INSTITUTE, CULUM8US LABORATORIES NUREG/CR-3900 V01: LONG=iEHM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACnAGING.First Quarterly Report, Year Three, April = June 1984 NUREG/CR-3988: MARCH 2 (MELTDONN ACCIDENT RESPONSE CHAHACTERISTICS) CODE DESCRIPTION AND USEHS MANUAL. 137
BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORIES NUREG/CR=0130 AD003: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE-PRESSURIZED aATER REACTOR P0nER STATION. NUREG/CR=0672 ADD 02: TECHNULOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE. BOILING MATER HEACTOR P0nER STATION. Classification Of Decommissioning hastes. NUREG/CR-2499 REVIEW 0F EMERGENCY RADIOLOGICAL INSTRUMENTATION ANU ANALYTICAL METHODS AT NHSS-LICENSEE SITES. NUREG/CR=3474: LONG=LIVEu ACTIVATION PRODUCTS IN REACTOR MATERIALS. NUREG/CR=3544: BETA PARTICLE MEASUREMENT AND 00SIMETRY AT NRC= LICENSED FACILITIES. NUREG/CR=3569: SPECIAL AND DOSIMETRIC MEASUREMENTS OF PHOTON FIELDS AT COMMERCIAL NUCLEAR SITES. NUREG/CR=3610 NEUTRON DUSIMETRY AT COMMERCIAL NUCLEAR PLANTS Final Report Of Subtask C: 3He Neutron Spectrometer. NUREG/CR=3751: EFFECTS OF HOCK RIPRAP DESIGN PARAMETERS ON FLOOD PROTECTION COSTS FOR UR4NIUM TAILINGS IMPOUNDMENTS. NUREG/CR=378T EFFECTIVENESS OF ENGINEERED SAFETY FEATURE (ESP) SYSTEMS IN RETAINING FISSION PHODUCTS.dackground Information. NUREG/CR=3796: EMERGENCY PHEPAREONESS SOURCE TERM DEVELOPMENT F0h THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS LICENSED FACILITIES. NUREG/CR=3798: CHARACTERIZATION OF CEMENT AND BITUMEN WASTE FORMS CONTAINING SIMULATED LOW-LEVEL HASTE INCINERATOR ASH. NUREG/CR=3842 STEAM GENERATOR GROUP PROJECT TASK 8 = SELECTIVE TU8E UNPLUGGING. NUREG/CR=3843: STEAM GENERATUR GROUP PROJECT TASK 10 = SECONDARY SIDE EXAMINATION. NUREG/CR=3869: ANALYSIS OF THE IMPACT OF INSERVICE INSPECTION USING A PIPING RELIABILITY MUDEL. NUREG/CR=3907: GT2R2 AN UPDATED VERSION OF GAPCON= THERMAL-2 NUREG/CR=3951: INTRODUCTION TO BIBELOT A BIBLIOGRAPHIC FINDING AND RETRIEVAL SYSTEM. BROOKHAVEN NATIONAL LAB 0 HAT 0HY NUREG/CR-2331 V03 N3 SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEANCH. Quarterly Progress Report, July-September 1983. NUREG/CR=2331 V03 N4: SAFETY RESEARCH PROGHAMS SPONSORED BY THt UFFICE OF NUCLEAR REGULATORY HESEARCH. Quarterly Progress Report,0ctober 1 = December 31,1983 NUREG/CR=2482 V05: REVIEn UF DOE HASTE PACKAGE PROGRAM. Subtask 1.1 - National Waste Package Program, April 1983 - September 1983 NUREG/CR=3169: SUPER SYSTEM CODE (SSC,REv. 0).AN ADVANCED THERM 0HYORAULIC SIMULATION FOR THANSIENTS IN LMF8RS. NUREG/CR=3469 V01: OCCUPATIONAL DOSE REDUCTION AT NUCLEAR P0nEH PLANTS ANNOTATED BIBLIOGRAPnY OF SELECTED READINGS IN RADIATION PROTECTION AND ALARA. NUREG/CR=3493: A REVIEn 0F THE LIMERICK GENERATING STATION SEVERL l ACCIDENT RISK ASSESSMENT. Review of Core Melt Frequercy. NUREG/CR=3518 V01: SLIM-MAUD:AN APPROACH TO ASSESSING HUMAN ERROR PROBABILITIES USING STHUCTURED EXPERT JUDGEMENT. Volume I Overview of = SLIM-MAUD. NUREG/CR=3520 V01: LONG=iEMM RESEARCH PLAN FOR HUMAN FACTORS AFFLCTING SAFEGUARDS AT NUCLEAN P0nER PLANTS. Volume I Summary And users Guide. NUREG/CR=3520 V02: LONG=iEHM RESEARCH PLAN FOR HUMAN FACTORS AFFECTING SAFEGUARDS AT NUCLEAR P0nER PLANTS. Volume II Development of Oetailed Analyses. NUREG/CR=3750 JOB ANALYSIS OF NUCLEAR POMER REACTOR HEALTH PHYSICS TECHNICIANS. NUREG/CR=3765: MINET SIMULATION OF A HELICAL COIL SODIUM / WATER STEAM 138
GENERATOR, INCLUDING STRUCTURAL EFFECTS. NUREG/CR-3766: TESTING OF NULLEAR GRADE ~LUSRICANTS AND THEIR EFFECT ON AS40 AND A193 87 80LTING MATERIALS. NUREG/CR-3812: ASSESSMENT OF IRRADIATION EFFECTS IN RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA. NUREG/CR-3813 MINET VALIDATION STUDY USING STEAM GENERATOR TRANSIENT DATA. NUREG/CR-3834: ON THE THRESHULD SULFUR AND LITHIUM TO SULFUR RATIO IN STRESS CORROSION CRACKING OF SENSITIZED ALLOY 600 IN BORATED l THIOSULFATE SOLUTION. NUREG/CR-3844: CHARACTERIZATION OF THE RADICACTIVE WASTE PACKAGES OF l THE MINNESOTA MINING AND MANUFACTURING COMPANY. NUREG/CR-3868: CONTAINMENT BUILDING ATMOSPHERE RESPONSES DUE TO REACTOR l GAS SURNING UNDER-SEVERE ACCIDENT CONDITIONS. i NUREG/CR-3878: MODELING CONSIDERATIONS FOR THE PRIMARY SYSTEM UF THE EXPERIMENTAL 8REEDER REAGTOR=II. l NUREG/CR-4001: CONTEMPT 4/M005 AN IMPROVEMENT TO CONTEMPT 4/ MOD 4 MULTICOMPARTMENT CONTAINMENT SYSTEM ANALYSIS PROGRAM FOR ICE l CONTAINMENT ANALYSIS. COLUM81A UNIV., NEW YORK, NY NUREG/CR-3940 FIELO EXPERIMENT DETERMINATIONS OF DISTRIBUTION i COEFFICIENTS OF ACTINIDE ELEMENTS IN ALKALINE LAKE ENVIRONMENTS. I COMMERCE, DEPT. OF, NATIONAL SUREAU OF STANDARDS NUREG/CR-4007 L0hER LIMIT OF DETECTION 0EFINITION AND ELAdORATION OF A PROPOSED POSITION FOR HADIULOGICAL EFFLUENT AND ENVIRONMENTAL i MEASUREMENTS. j E.C. RODABAUGH ASSOCIATES, INC. i NUREG/CR-3599: SOURCES OF UNCERTAINTY IN THE CALCULATIONS OF LUADS ON SUPPORTS OF PIP!NG SYSTEMS. i EG&G, INC. i NUREG/CR=0169 V17: LOFT EXPEN! MENTAL MEASUREMENTS UNCERTAINTY ANALYSIS. Volume XVII Process Instruments Recorded On DAVDS. NUREG/CR=1740 R01: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF SELECTED I INSTRUMENTATION AND CONTHOL COMPONENTS AT U.S. COMMERCIAL NUCLEAR P0nER PLANTS JANUARY 1,1976 TO DECEMBER 31,1981. i NUREG/CR-3513: MECHANICAL HELI ABILITY EVALUATION OF ALTERNATE MOTONS i I FOR USE IN A RA010100INE AIR SAMPLER. NUREG/CR-3708 LhR SPENT FUEL DRY STORAGE WEHAVIOR AT 229 C. j NUREG/CR-3761: RELAP5 THERMAL HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR THE UCUNEE-1 PRESSURIZED WATER REACTOR. NUREG/CR-3824: CONTING PdOGRAM GUIDE. I NUREG/CR-3667: DATA SUMMARIES OF LICENSEE EVENT REPORTS OF INVERTERS AT O.S. COMMERCIAL NUCLEAN P0nER PLANTS, JANUARY 1,1976 TO DECEMdER i 31,1982. NUREG/CR-3921: DRY SPENT FUEL STORAGE TEST PLAN FOR FINAL j NONDEsiRUCTIVE FUEL N00 EXAMINATION. ELECTRIC POWER RESEARCH INSTITUTE NUREG/CR-3111 BWR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST i RESULTS. ENGINEERING & ECONUMICS HESEARCH, INC. l l NUREG/CR-38993 UTILITY FINANCIAL STABILITY AND THE AVAILABILITY UF FUNDS FON DECOMMISS!UNINb. GENERAL ELECTRIC CU. l j NUREG/CR-2576: BnR FULL INTEGRAL SIMULATION TEST (FIST)--Facility Description Report, NUREG/CR-3711: SnR FULL INTEGRAL SIMULATION TEST (FIST) PHASE I TEST RESULTS. HANFORD ENGINEERING DEVELOPMENT LABONATORY NUREG/CR-3318: LnR PHES$URE VESSEL SURVEILLANCE DOS! METRY IMPROVEMENT PROGRAM PCA Emperiments,dlind Test,And Physics-Dosimetry Support For s l
Tho PSF Expericcnto. IDAHO NATIONAL ENGINEERING LABURATORY NUREG/CR-35133 MECHANICAL HELIABILITY EVALUATION OF ALTERNATE MOTORS l FOR USE IN A PADIOI0 DINE AIR SANPLER. l INNALATION T0XICOLOGY RESEARCH INSTITUTE l NUREG/CR-3870 RADIATION DOSE ESTIMATES AND HAZARD EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Rept July 1982 -June l 1983 J8F ASSOCIATLS i NUREG/CR-3905: SEQUENCE COUING AND SEARCH SYSTEM FOR LICENSE EVENT l REPORTS. Users Guide. l LAhRENCE LIVERMORE NATIONAL L4dORATORY NUREG/CR-2015 V08: PHASE I FINAL REPORT - SYSTEMS ANALYSIS (PROJECT l VII). Seismic Safety Margins Research Program. NUREG/CR-3480 VALUE/ IMPACT ASSESSMENT FOR SEISMIC DESIGN CRITtRIA USI A-40 l NUREG/CR=3593 V01: SYSTEMS INTERACTION AESULTS FROM THE DIGRA.PH MATRIX ANALYSIS OF A NUCLEAR P0nEN PLANT'S HIGH PRESSURE SsFETY INJECTION SYSTEM. NUREG/CR-3593 V02: SYSTEMS INTERACTION RESULTS FROM THE DIGRAPH MATRIX i ANALYSIS OF A NUCLEAR POWER PLANT'S HIGH PRESSURE UAFETY INJECTION SYSTEM. Volume 2. NUREG/CR-3660 v02: PROdApILITY OF PIPE FAILURE IN THE REACTOR COULANT LOOPS OF wESTINGHOUSL PNd PLANTS. Volume 2 Pipe Failure Induced By Crack Growth. NUREG/CR-3663 V02: PROOAdILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINELRING PWR PLANTS.Vol 2 Pipe Failure Induced j by Crack Growth. NUREG/CR-3758: CROSSHOLE GEOPHYSICAL METHODS USED TO INVESTIGATE THE NEAR VICINITY OF HIGH LEVEL HASTE REPOSITORIES. j NUREG/CR-3826: RECOMMENDATIONS FOR FROTECTING AGAINST FAILURE SY BRITTLE FRACTURE IN FEHRITIC STEEL SHIPPING CONTAINERS GREATER THAN FOUR INCHES THICK. LOS ALAMOS SCIENTIFIC LAdONATONY NUREG/CP-0053: PROCEEDINGS OF THE N! NTH ANNUAL STATISTICS SYMPOSIUM ON NATIONAL ENERGY ISSUES,0ctober 19-21,1983, NUREG/CR-3678: ESTIMATION METHODS FOR PROCESS HOLDUP OF SPECIAL NUCLEAR i l MATERIALS. NUREG/CR-3679: CALIBRATION AND QUALIFICATION OF THE LOS ALAMOS FAILudE 3 MODEL (LAFM). NUREG/CR-3735: ACCIDENT-INDUCED FLOW AND MATERIAL TRANSPORT IN NUCLEAR FACILITIES--A LITERATUNE REVIEh. NUREG/CR-3742: BUCKLING OF STEEL CONTAINMENT SHELLS UNDER l TIME-DEPENDENT LOADING. { NUREG/CR-3821: EVALUATION OF CRACK PLANE EGUILIBRIUM MODEL FOR PREDICTING PLASTIC FRACTURE. NUREG/CR-3822: SOLA-PTS: A Transient,Three= Dimensional Algorithm For l Fluid-Thermal Mixing And Wall Heat Transfer In Complex Geometries. I NUREG/CR-3845: PREUICTION OF NONLINEAR STRUCTURAL RESPUNSE IN LMF03 l ELEVATED = TEMPERATURE PIPING. l MASSACHUSETTS INSTITUTE UF TECHNOLOGY, CAMURIDGE, MA NUREG/CR-3895: INVESTIGATION OF COLD LEG WATER HAMMER IN A PnR DUE TU l THE ADMISSION OF ECC DURING A SMALL BREAK LOCA. MATERIALS ENGINEERING ASSOCIATES, INC. NUREG/CR-3228 V02: STRUCTURAL INTEGPITY OF WATER REACTOR PRESSURE BOUNDARY COMPONENTS. Annual Report For 1983 NUREG/CR-3788 V01: STRUCTUHAL INTEGRITY OF LIGHT WATER REACTOR PRESSURE 80VNDARY COMPONENTS.Four-Year Plan 1984=1988 NUREG/CR-3833: 8EHAV10R QF SU8 CRITICAL AND SL0n-STABLE CRACK GR0 NTH FOLLOWING A POST =IRRADIAT!UN THERMAL ANNEAL CYCLE. 140
MATHTECH, INC, NUREG/CR-3840: COST ANALYSIS FOR POTENTIAL MODIFICATIONS TO ENHANCE THE -ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKOUT. 0AK RIDGE NATIONAL LABORATORY NUREG/CR-2000 V03 N6 LICENSEE EVENT REPORT (LER) COMPILATION For Month of June 1984 NUREG/CR-2000 V03 N7: LICENSLE EVENT REPORT (LER) Compilation For Month of July 1984 NUREG/CR-2000 V03 N8: LICENSEE EVENT REPORT (LER) Compilation For Montn Of August 1984 NUREG/CR-2996: SENSITIVIIY OF DETECTING IN-CORE VIBRATIONS AND ROILING IN PRESSURIZED HATER RLACTURS USING EX-CORE NEUTRON NOISE. NUREG/CR-3139: SCENARIOS AND ANALYTICAL METHODS FOR UF6 RELEASES Al NRC. LICENSED FUEL CYCLE FACILITILS. NUREG/CR-3346: 810 ASSAY DATA AND A RETENTION-EXCRETION MODEL FOR SYSTEMIC PLUTONIUM. NUREG/CR=3459: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MdBT) SUNDLE B-5 NUREG/CR-3460: EXPERIMLNT DATA REPORT FOR MULTIROD BURST TEST (MNBT) BUNDLE 8-6, NUMEG/CR-3470: ATWS AT BR0aNS FERRY UNIT ONE = ACCIDENT SEQUENCE ANALYSIS. NUREG/CR-3492 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDEa7 EVALUATION QUARTERLY PROGRESS REPORT, October-December 1983. NUREG/CR-3524: ORGANIZATIONAL INTERFACL IN REACTOR EMERGENCY PLANNING AND RESPONSE. NUREC/CR-3590: EVALUATION OF ISOTOPE DILUTION MASS SPECTROMETRf FOR 810 ASSAY MEASUREMENT OF URANIUM, PLUTONIUM,AND THORIUM IN URTNE. NUREG/CR-3591 V01: PRECURSORS TO POTENTIAL SLVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3591 V02: PRECURSURS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1980-1981 A Status Report. NUREG/CR-3599: SOURCES OF UNCERTAINTY IN THE CAR.CULAlIONS OF LUADS ON SUPPORTS OF PIPING SYSTEMS. NUREG/CR-3617: NOBLE GAS,IuCINE,AND CESIUM TRANSPORT IN A POSTULATED LOSS OF DECAY HEAT REMOVAL ACCIDENT AT BR0nNS FERRY. NUREG/CR-3618: OCA-P,A DETERMINISTIC AND PROBA8ILISTIC FRACTURE-MECHANICS CUDL FON APPLICATION TO PRESSURL VESSELS. i NUREG/CR-3655: A METH00 F0d ANALYTICAL EVALUATION'0F COMPUTER-BASED DECISION AIDS. NUREG/CR-3671: ASSESSMLNT OF RADIATION EFFECTS RELATING TO REACTOR PRESSURE VESSEL CLADDING. NUREG/CR-3692: POSS! ALE MODES OF STEAM GENERATOR OVERFILL RESULTING FROM CONTROL SYSTEM MALFUNCTIONS AT OCONEE-1 NUCLEAR PLANT. NUREG/CR-3714: ON THE DEVELOPMENT OF ENVIRONMENTAL RADIATION STANDARUS FOR GEOLOGIC DISPOSAL OF HIGH-LEVEL RADICACTIVE WASTES. NUREG/CR-3744 V01: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR QC10 DER 1983 - MARCH 1984 NUREG/CR-3763: REVIEn AND ASSESSMENT OF RADIONUCLIOE SORPTION INFORMATION FOR THE dASALT WASTE ISOLATION PROJECT SITE (1979 Through May,1983). NUREG/CR-3815: STATISTICAL EVALUATION OF THE METALLURGICAL TEST DATA IN THE ORR-PSF =PVS IRRADIATION EXPERIMENT. NUREG/CR-3830 V01: AEROSul RELEASE AND TRANSPORT PROGRAM, SEMIANNUAL PROGRESS REPORT FOR OCTosER 1983 - MARCH 1984 NUREG/CR-3832: UNCERTAINTIES IN LONG-TERM REPOSITORY PERFORMANCE DUE TO THE EFFECTS OF PUTURE GEOLOGIC PROCESSES. NUREG/CR.3851 V01: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPE 0 dY DOE HIGH-LEVEL NUCLEAR WASTE REPOSIIORY SITE 141
? ~ PROJECTS. Report for October = December 1983. NUREG/CR-3856: AN ULTRASUNIC LEVEL AND TEMPERATURE SENSOR FOR P0nER REACTOR APPLICATIONS. NUREG/CR=3871: AN QVERVIEW OF THE UNIFIED TRANSPORT APPROACH. NUREG/CR=3888: ANALYSIS OF THE VENUS PhR ENGINEERING MOCKUP EXPERIMENT -PHASE 1: SOURCE DISTRIBUTION. NUREG/CR-3905: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSE EVENT REPORTS. Users Guide. NUREG/CR-4011: THE 21/55 DATA BASE USER'S MANUAL. PARAMETER, INC. NUREG/CR-3053: CLOSEQUT OF IE BULLETIN 80-08: EXAMINATION OF CONTAIN ENT M i LINER PENETRATION nELDS. r NUREG/CR=3792: CLOSE0UT uF It BULLETIN 79=11 FAULTY OVERCURRENT TRIP DEVICE IN CIRCUIT BREAnERS FOR ENGINEERED SAFETY SYSTEMS. 4 NUREG/CR-3795: CLOSEQUT UF IE BULLETIN 82-04 DEFICIENCIES IN PRIMARY CONTAINMENT ELECTRICAL PLNETRATION ASSEMBLIES. = QUADREX CORP. NUREG/CR=3939: MATER HAMMEH, FLOW INDUCED VIBRATION -AND SAFETY / RELIEF VALVE LOADS, a SANDIA LABORATORIES ^ P NUREG/CR-3190: PLUGM A COUPLED THERMAL-HYORAULIC COMPUTER MODEL FOR FREEZING MELT Flow Id A CHANNEL. NUREG/CR-3273: COM8USTION UF HYDROGEN AIR MIXTURES IN THE VGES b CYLINORICAL TANK. NUREG/CR-3369: AN UNCENTAINTY STUDY OF PnR STEAM EXPLOS' NS. a; NUREG/CR-3418: SCREENING TESTS OF TERMINAL BLOCK PERFORMANCE IN A e SIMULATED LOCA ENVIRONMENT. ..NUREG/CR=3589 V01: REACTOR SAFETY RESEARCH QUARTERLY REPORT. January-March 1983 = g NUREG/CR-3589 V02: REACTOR SA5ETY RESEARCH QUARTERLY REPORT. April-June r-1983 NUREG/CR-3643: HETEROGENLOUS OXIDATIVE DEGRADATION IN IRRADIATED POLYMERS. . )4 5 NUNEG/CR-3662: FUEL-DISRUPTION EXPERIMENTS UNDER HIGH-RAMP = RATE HEATING 9'A CONDITIONS. NUREG/CR-3690: RELAPS ASSESSMENT SEM! SCALE NATURAL CIRCULATION TESTS S=NC=3,S=NC=4,AND S=NC-8 P NUREG/CR-3724: ULTIMATL STHENGTH ANALYSES OF THE WATTS BAR, MAINE YANKEE,AND BELLEFONTE CONTAINMENTS. = NUREG/CR-3734: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual t Report,0ctober 1982 - Maren 1983 NUREG/CR-3776: TESTING OF 3AFETY=RELATED NUCLEAR POWER PLANT EQUIPMENT E AT THE CENTRAL RECEIVER TEST FACILITY. NUREG/CR-3777: CAPABILITIES AND DIAGNOSTICS OF THE SANDIA r PELLETHON= RASTER SYSTEM. NUREG/CR=3786: A RLVIEn UF REGULATORY REQUIREMENTS GUVERNING CONTROL L ROOM HABITABILITY SYSTEMS. NUREG/CR-3618: REPORT OF RESULTS OF NUCLEAR POWER PLANT AGING E WORKSHOPS. NUREG/CR-3820 V01: THENHAL/ HYDRAULIC ANALYSIS RESEARCH PROGRAM.uuarterly ReportsJanuary-March 1984 NUREG/CR-3835: SIMULATION UF FLAME PROPAGATION THROUGH VORTICITY REGIONS USING THE DISCRETE VORTEX METHOD. SCIENCE & ENGINEERING ASSOCIATES, INC. p-NUREG/CR=3840: COST ANALYSIS FOR POTENTIAL MODIFICATIONS 70 ENHANCE THE ABILITY OF A NUCLEAR PLANT TO ENDURE STATION BLACKOUT. SCIENCE. APPLICATIONS, INC. I NUREG/CR=3190: PLUGM A COUPLED THERMAL = HYDRAULIC COMPUTER MODEL FOR e FREEZING MELT FLOW IN A CHANNEL. NUREG/CR-3665: OPTIMIZATION UF PUBLIC AND OCCUPATIONAL RADIATION z 142
PROTECTION AT NUCLEAR PogER PLANTS.Exocutive Su==ary. NUREG/CR=3665 Von OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR POWER PLANTS.A Review Of 0;.upational D'se o Assessment Considerations In Current Probabilistic Risk Assessment And Cost = Benefit Analyses. NUREG/CR=3665 V02 OPTIMIZATION OF PUBLIC AND OCCUPATIONAL RADIATION PROTECTION AT NUCLEAR PonER PLANTS. Considerations In Factoring Occupational Dese Into Value= Impact And Cost-Benefit Analyses. NUREG/CR=3665 V03: OPTIMIZATION OF PO6LIC AND UCCUPATIONAL RADIATION l PROTECTION AT NUCLEAR P0nER PLANTS.A Calculation Method. NUREG/CR=3852: INSIGHT INTU PRA METHODOLOGIES. SEARCH TECHNOLOGY, INC. NUREG/CR=3655: A METHOD FOR ANALYTICAL EVALUATION OF COMPUTER =dASEC DECISION AIDS. SOUTHWEST RESEARCH INSTITUTE 1 NUREG/CR=3892 A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT.Sunimary Report. NUREG/CR=3892 V01: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANCIAL EQUIPMENT. Task 1 - Survey of Methods For Equipment And Components Evaluation of Methodology 3 Qualification And Methodology.... 4 NUREG/CR=3892 V02: A RESEAMCH PROGRAM FOR SEISMIC QUALIFICATION OF NUCLEAR PLANT ELECTRICAL AND MECHANICAL. EQUIPMENT. Task 2= Correlation Of Methodologies For Seismic Qualification Tests Of Nuclear Plant Equipment. -NUREG/CR=3892 V03: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 3-Recommendations For Improvement Of Equipment Qualification Methodology And Criteria. i NUREG/CR=3892 V04: A RESEARCH PROGRAM FOR SEISMIC QUALIFICATION UF NUCLEAR PLANT ELECTRICAL AND MECHANICAL EQUIPMENT. Task 4 - Tne Use of Fragility In Design Of Nuclear Plant Equipment. TENNESSEE TECH. UNIV., COOKEVILLE, TN i NUREG/CR=3692: POSSIBLE MODES OF STEAM GENERATOR OVERFILL RESULTING FROM CONTROL-SYSTEM MALFUNCTIONS AT OCONEE-1 NUCLEAR PLANT. TEXAS TECH UNIV., LUBBOCK, TX NUREG/CR=3874: NEAR= GROUND TURNA00 MIND FIELDS. UNITED NUCLEAR CORP. NUREG/CR=3884: EVALUATION UF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM = THREE MILE ISLAND UNIT 2 POLAR CRANE RECOVERY. WASHINGTON, UNIV. OF, SEATTLL, WA NUREG/CR=3896: SIMULATION EXPERIMENTS COMPARING ALTERNATIVE PROCESS FORMULATIONS USING A FACTORIAL DESIGN. NUREG/CR=3897: EVALUATION UF ECOSYSTEM SIMULATION MODELS AS TOULS FOR ASSESSMENT OF PonER PLANT IMPACTS ON FISH POPULATIONS. Final Rept. MESTINGHOUSE ELECTRIC CORP. 1 NUREG/CR=35133 MECHANICAL RELIABILITY EVALUATION OF ALTERNATE MOTONS FOR USE IN A RAD 10100INE AIR SAMPLER. i l NUREG/CR=3654: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Data Evaluation and Analysis Report NRC/EPRI/ Westinghouse Report No. 14 NUREG/CR=3708: LhR SPENT FUEL DRY STORAGE BEHAVIOR AT 229 C. i 143 - - - - - - -. ~. -. -
Licensed Facility Index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If fur-ther information is needed, refer to the main citation by the NUREG number. 30-23,9 Seewns Ferry Nuclear Power Statten. Unit 1. Tennessoa Valley Authern NUREC,/CR-3470 j 50-23 S..uns....g N.cio..,ew.e St. tion. Unit i. 7enn.sse. v. n ey A.th. 1 NuREO CR-3 i, S0-260 Steens Fetty Nuclear Power Station. Unit 2. Tennessee Valley Autheet NURE0/CR-3617 30-296 Srowns Ferry Nuclear Power Statten. Unit 3. Tennessee Valley Authert NUREO/CR-3617 50-413 Catawba Nvelear Stetten. Unit 1, Duke Power Co. NUREG-0954 SO3 30-413 Catawba Nucleae Statten. Unit 1. Duke Power Co. NUREG-1072 50-414 Catawba Nucleae Station. Unit 2. Duke Power Co. NURE0-09S4 803 S0-275 Diante Cangen Nuclear Power Plant. Unit 1. Pacific Gas & Electric Ce NUREG-0675 524 30-27S Diante Cangen Nuclear Power Plant. Unit 1. Pacific Gas & Electric Ce NUREO-0675 S25 50-275 Dian te Congen Nuclear Powee Plant. Unit 1. Pacific Gas & Electric Ce NUREG-067S S26 S0-275 Diable Cangen Nuclear Power Plant. Unit 1. Pacific Gas & Electric Ce NUREG-067S 827 S0-275 Diable Cangen Nuclear Power Plant. Unit le Pacific Gas & Electric Ce NUREO-0675 S27 ERR S0-323 Diable Cangen Nuclear Power Plant. Unit 2. Pacific Gas & Electric Ce NURE0-0675 S26 S0-323 Diatte Cangen Nuclear Power Plant. Unit 2. Pacific Gas & Electric Ce NUREG-0675 827 50-323 Diatte Canyon Nuclear Power Plant. Unit 2. Pacific Gas & Electric Ce NUREO-0675 S27 ERR S0-323 Diante Cangen Nwcaear Power Plant. Unit 2. Pacific Gas & Electric Ce It) REG-067S 824 30-323 Diat te Cangen Nuclear Power Plant. Unit 2. Pacific Gas & Electric Ce NUREG-067S 825 S0-341 Enrice Fermi Atomic Power Plant. Unit P. Detroit Edison Co. NUREG-0799 504 30-73 General Electric Nucleae Test Reactor. General Electric Ceepang NUREG-1069 50-414 Orend Swif Nuclear Station. Unit 1. Mississippi Power & Light Co. NUREG-0831 SOS 30-416 Grand Gulf Nuclose Station. Unit 1. Mississippi Power & Light Co. NUREO-0831 806 30-416 Grand Gulf NwClear Station. Unit 1. Mississippi Power & Light Co. NUREG-0926 R01 30-417 Orand 0w17 Nuclear Statten. Unit 2. Mississippi Power & Light Co. NUREG-0831 SOS S0-417 Orand Swif Nuclear Station. Unit 2. Mississippi Powee & Light Co. NURE0-0831 806 S0-3S2 Limerick Generating Station. Unit 1. Philadelphia Electric Co. NUREO-1069 S0-352 Limerick Generating Station. Unit 1. Philadelphia Electric Co. NURE0/CR-3493 S0-333 Lieerict Generating Station. Unit 2. Philadelphia Electric Co. NUREG-1069 50-333 Limeetch Generating Station. Unit f. Philadelphia Electric Co. NUREO/CR-3493 S0-309 Maine Yankee Atomic Power Plant. Maine Vantee Atenic Powee Co. NURE0/CR-3724 S0-294 Michigan State Univ. Research Reacter NUREG-10G4 50-423 Millstone Nuclear Power Station. Unit 3, Northeast Nuclear Energy Ce NUREO-1031 50-423 Millstone Nucleae Power Statten. Unit 3. Northeast Nwctear Energy Ce NUREO-1064 50-410 Nine Mile Point Nvelear Statten. Unit 2. Niage*a Mohawk Power Corp. NUREG-1085 S0-269 Oconee Nuclear Statten. Unit 1. Duke Power Co. NURE0/CR-3692 30-269 Oconee Nuclear Statten. Unit 1. Duke Power Co. NUREO/CR-3761 50-458 River Send Station. Unit 1. Gute States Utilities Co. NUREG-1073 S0-322 Shefeham Nuclear Power Station. Long Island Lighting Co. NUREG-0420 S06 50-322 She*eham Nuclear Power Station. Long Island Lighting Cs NUREG-0420 S07 30-289 Three Mile Island Nuttee Statten. Unit 1. Metropolitan Edison Co. NUREG-0680 SOS 30-382 Waterford Generating Station. Unit 3, Lewistana Power & Light Co. NUREG-0787 S07 S0-390 Watts Sat Nuclear Plant. Unit 1. Tennessee Va11eg Authority NURE0/CR-3724 50-87 Westinghouse Crit 14al Esperiment. Settis (Westinghouse) Ateatc Power NURiG-1083 8 145
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- 2. TITLE AND Sutil 3 LEAVE SLANK R:gulatory d Technical Reports (Abstract Index Journal)
Compilation f Third Quarter 1984 . qre RuORT CO-tenO July - Septembe =O~Ts l veAa / u.uT ORisi / 8 OATE REPORT issufD Np(ember l .O~Ta v AR 1984 F. s EAFORMiNG ORGANi2 ATION N AME AND iLiNG ADDRESSisarwal. Camp 8 MOJECTITASK/VWORit umai NUM9tR Division of Technical 1 formation and Document Control / Office of Administratio i *i= Oa caA~T au-na U.S. Nuclear Regulatory ission Washington, DC 20555 ii..rvreO RuORT in,0~ sori ~a ORoA~a2 ATiO..A.E A~0. ail, o AooR ,,,,,..,.c, j / Quarterly Same as 7, above. /
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July - September 1984 u sv,uoinTARv ~Oni i3 ABSTR ACT (20D words er tesst This compilation lists all NRC regulatory /an technical reports published under the NUREG series during the third quar 'r o f 84. 14 DOCUMENT AN ALYSi$ -e KtvWORQ$/DE iPTORS IS AVAILABILITY STATEWiNT abstract index (nlimited 18 hCuRITY CLAS$1FICATION 6EcTa'ssi fied oa~TiriiRs,0,i= =~Ono nRus IThe reportl Unclassi fied 17.NUMSER OF PAGE5 18 PRICE
UNITED STATES POUMM CLASS MAR NUCLEAR REGULATCRY COMMISSION P "^y,* "c '^'8 WASHINGTON, D.C. 20666 waSa o.c. l PtRMIT No. G 47 OFFICIAL BUS! NESS PENALTY FOR PRNATE USE. 6300 Main Citations cnd Abstracts t Contractor Report Number index l Personal Author index l !u-i i; Subject Index m L. NRC Originating Organization Index NRC Contractor Sponsorindex r L Contractor index i~ l M Licensed Facility Index l -}}