ML20071M846
| ML20071M846 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1983 |
| From: | NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V08-N01, NUREG-304, NUREG-304-V8-N1, NUDOCS 8306030488 | |
| Download: ML20071M846 (64) | |
Text
--.
NUREG-0304 Vol. 8, No.1 Regulatory and Technical Reports Compilation for First Quarter 1983 January - March U.S. Nuclear Regulatory Commission office of Administration pe"%,
4
- s
(.
i
'...../
PDR UR Q 0304 R pop
e i
i 4
a P
Available from NRC/GPO Sales Program f
Superintendent of Documents Government Printing Office Wahinf.on, D. C. 20402 i
A year's subscription cons;sts of 4 issues for this publication.
4 Single copies of this publication are available from National Technical information Service, Springfiald, VA 22161 f
Microfiche of single copies are available from NRC/GPO Sales Program Washington, D. C. 20555 4
l t
l l
.f I
l r
,,r- - - -,
r
,m._
.r-,
-.-..,c-+-,--e
.. -c--..-
--m..
NUREG-0304 Vol. 8, No.1 Regulatory and Technical Reports Compilation for First Quarter 1983 January - March r
3 Date Publishad: May 1983 Division of TechnicalInformation and Document Control Off;ce of Admin!stration U.S. Nuclear Flagulatory Commission Washington, D.C. 20655
?
l l
f
4 4
F t
4 i
1 e
t s
CONTENTS Preface..........
......................... v 4
Ir.dex Tab W:n Citation and Abstracts.......
..... 1 S<xtf Repo:tt.
Conferone.e Procssiings....
Oc ntractor R spor ts.................................................
Contractor Heport Nurnber index.........,
.2 Pe sanal Author index........................................................... 3 S e t$M:t Inde x........................................................
... 4 NHC 0iiginat;ng Organization index (Staff heports)...
... 5 NRC Contract Sponsor Index (Contractor P.aports)...
6 Ccatrscter index................................................
...... 7 Licensud Focility nndex.
...... 8 l
l r
I I
4 1
i i
l
}
l-l PREFACE j
This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Comniission (NRC) Staff and its contractors. It is NRC's
.l intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-i preciated. Please send them to:
1 Division of Technical Information -
and Document Control Attn: Ann W. Savolainen Landow 212 l
U.S. Nuclear Regulatory Commission j
Washington, D.C. 20665 1
The main citations aad aLstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, and NUREG/CR-XXXX. These preceds the following indexos:
j Contractor Report Number index Persnnal Auth9r Index Stbiect inoex NRC C,.-igineting Organization :nder (Staff Reports) flP C Cct.trat i Sponsor index (Contractor Reports)
Cartractor ladex i
Licer. sed Facility Index A deta.iled explar.ation of the entries precedes each indax.
Tne bibliag aph;c elements of the main citations are the following:
Stoff He. port i
NUREG-05N. F1 ARK ll CONTAINMENT PROGHAM EVALUATION AND ACCEPTANCE CRITERIA.
[
ANDERSON, C.J.. Divis5r. of Safety Technology. August 1981. 90 pp. 8109140048 09 5 0:200.
j i
Whera the enWs 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) tt'e NRC Document Control System accession number, (8) the microfeche address (for intemal NRC use).
l Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND i
RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 81nn9an9em. ANL-813. 08632:070.
Where the entries are (1) report number, (2) report title, (3) repo.-t author, (4) organization that compiled the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-4 ment Control System accession number, (8) the report number of the originating organization, (9) the i
microfiche address (for NRC intomal use).
Contractor Report NUREG/CR-1666: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R.
l Sandia Laboratories. May 1981.100 pp. 8107010440. SAND 8(M)829. 08912:242.
Where the entries are (1) report nu nber, (2) report title, (3) report authors, (4) organizational unit of i
authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC i
Doc:.* ment Control System accession number, (8) the report number of the originating organization (if j
given), and (9) the microfiche address (for NRC intemal use).
t l
f v
The following abbreviations are used to identify the document status of a report:
ADD - addendum APP - appendix DRFT - dratt ERR
- errata N - number revision R
S - supplement V - volume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the NRC-GPO Sales Office or from the Etional Techn. cal Information Service, Springfield, Virginia 22161. To purchase documar.ts from the NRC-GPO Sales Office e,end a check or money order, payable to the Superintendent of Documente. to the foJo.ving ar' dress:
U S. Nucker Regulatory Commission ATTN: S$es Muapet Wa:Nngton, D.C. 20555 t
You reay chargo any purchsa to your GPO Dcptit Accotet, Master Chege =rd, or VISA charge card b/ colling the NRC-GPO Salea Ohire on Gel) 492 9530. Non-U.S. customeis rnest rnake payrnent in efv. ace e,: thor by laterrat:orM Postsi Myey Order, paabts tc tt,3 Supsrintendent of Docuraents, or by draft on a United S%tes er Canadi:n bank, payable to the Superintevatit o' Documerits.
NFC Rergrt Code.s The NUREG desgnat!on, NURcG-XXXX, and; cates that the document is a fctmal NRC statf-generated report. Contractor-prepared tcmil f JRC toputs carry the report code NUREG/CR-XXXX. TMs type of idertification replects contrec:or esta%shed codes such as ORNL/!JUREG/TM,vyX ond TREE-NUllEG-XXXX, es we:1 as various whe. numNrs t1at could not be corrriated wth NRC spnnsorship of thq work bc!ng reported, in addition to tim? NUMEG aad NUREG/CR cocea, ' 0 PlG/CD :s used for ';RC-sponsored conferecca procee. dings.
All these report codes are contro: led and essigned by the NFiC Dinsion cf Technical Infortnation and Document Control.
vi
1 Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is an NRC staff originated report. NUREG/CP-XXXX is an NRC sponsored cor.forence report, and NUREG/CR XXXX is an NRC centractor-prepared report. The bibliographic information (see Preface for details) is followed by a brief abstract of the report, NUREC-0020 VC6 f tO9: LICENFED OPERATING REACTORS STATUS
SUMMARY
REPORT.Dats As Of A.*Lat 31,199k (Crey Book)
- Management Information Branch.
January 1983.
366pp.
8302170107, 17194:200.
The O?ERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS prcvides desta un the creration of nuclear units as timely and accurately as poss1513.
This information is collected by the Office of Management and Program Analysis from the Headquarters staff of NRC's Office of Inspection cad Enfurcement, from N9C's Regicnal Offices, and from utilities.
The three sect: ions cf the v aport are:
monthly h2ghlights and statistics for comercial op erati r,g units, and errata from previously reported dataJ a compilation of detailed information on sach unit, provided by HRC's Region.nl 08Pices, IE Headqur.rters and the utilitiess and an appendix for miscellaneous information such ar spent fuel stcrape capability, reactor-years of eaperience and non-power reacters in the U.S.
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.
NUREC-0020 V06 N10: LICENSED GPERATINC REACTORS STATUS
SUMMARY
REPORT. Data As Of Septemb er 30,19B2. (Greg Book )
- Management Information Branch.
February 1983.
410pp.
8303110145.
17525:006.
The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS l
provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office of Management and Program Analysis from the Headquarters staff of l
NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are:
monthly highlights and statistics for commercial operating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters and the utilitiess and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of esperience and non-power reactors in the U.S.
It is hoped the report is help ful to all agencies and individuals interested in maintaining an awareness of the U.S.
energy situation as a whole.
NUREQ-0020 V06 N11: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of October 31,1982.(Greg Book)
- Management
(
1
Information Branch.
March 1983.
393pp.
8303300426.
17768:356.
Th e OPERATINC UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office of Management and Program Analysis from the Headquarters staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are:
monthly highlights and statistics for commercial operating units, and errata from previously reported dates a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters and the utilitiess and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the U. S.
It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S.
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.
NUREC-0040 VO6 NO4-LICENSEE CONTRACTOR AND VENDOR INSPECTICN STATUS R EPOR T.
Quarterly Reporw Octoler 1982 - Dec*mber 1982.
- Region 4, 1
Office of Dftector.
January 1983.
Uf 9pp.
83C2100265.
17133:187.
This peilodical covers the results af inspections performad by the NRC 's Vendor Program Branch that have been distributed to the l
inspected cryanizations during the period from October 1962 through Decen,ber 1982.
Also included in this issue are the results of cartain inspections performed prior to 0:tober 1982 that were not included in previous issues of NUNEG-0040.
HUREC-0090 VCS NO3: REPGRT TO CONGRESG ON ASNG9M AL OCCU9RENCES. duly -
September 1982.
- Director's Office January 1983.
40pp.
830217002d.
17192 052.
Section 2CB of the Energy Raurganizat4on /.ct of 1974 identifies an abnurmal occurrence er an enscheduled incident or event which the Nuclear Rrgulatery Corrnission deteroines to be signif2 cant from the s tandpoint of public health or safety and requires a quarterly report of such events to be made to Congress.
This report covers the period July 1 to September 30,1902.
During the report period, there were two abnormal occurrencess one at the nuclear power plants licensed to operate and one at other NRC licensees.
The first involved loss of auxiliary electrical power and the second involved rupture of at least one americium 241 well logging source.
The Agreement States reported no abnormal occurrences to the NRC.
The report also contains information updating some previously reported abnormal occurrences.
NUREC-0304 VO7 N04: REGULATORY AND TECHNICAL REPORTS. Compilation For 1982.
- Division of Technical Information & Document Control.
February 1983.
630pp.
8303160606.
17600:341.
This compilation lists all NRC regulatory and technical reports published under the NUREC series during 1982.
NUREC-0383 VO1 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
- Division of Fuel Cycle & Material Safety.
January 1983.
404pp.
8303140704.
17578:140.
This directory contains a Summary Report of NRC Approved Packages (Volume 1 ),
Certificates of Compliance (Volume 2), and a Summary 2
i i
j.
Report of NRC Approved Quality Assurance Programs for Radioactive f
Material Packages (Volume 3).
The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the U. S.
Noclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of 4
Compliance number is included at the back of each volume of the directory. The Summary Report includes a listing of all users of each package design prior to the publication date of the directory.
Shipments of radioactive material utilizing these packagings must be in accordance with the provisions of 49 CFR 173.393a and 10 CFR Part 71, as applicable.
In satisfying the requirements of Section i
71.12, it is the responsibility of the licensees to insure themselves that they have a copy of the current approval and conduct their transportation activities in accordance with an NRC approved quality l
assurance program.
Copies of the current approval may be obtained from the U.S.
Nuclear Regulatory Commission Public Document Room files (see docket number listed on each certificate) at 1717 H Street, Washington, DC 20555.
Note that the general license of 10 CFR 71.12 does not authorize the receipt, possession, use or transfer cf nyproduct, sour:e, or special nuclear materials sach authorization mort be obtainei pursuant to 30 CFR Parts 30 to 26, 40, 50, or 70.
i NUREC-0383 VO2 h05: DIftECfCRY OF CERTIFICATES OF COMPLIANCE FOR R ADIGSCTIVE MATERIALS PACM AGES. Certificates of Compliance.
- Division of Fuel Cycla & Material Safety.
January 1983.
611pp.
C3031407C1.
17579.184.
i This directory contaires a Scesary Report of NRC Approved Pcckenes (Volume 1 ),
Certificates of Compliance (Volume 2), and a Sun. mary t
Report of NRC Approved Ouality Assurance Programs for Radioictive l
Haterial Packages (Volume 3).
The purpose of this etrectory is to make available a convenient i
source of information on packagings which have been approved by the l
U. S.
Nuclear Regulatory Commission. To assist in identifying l
packeging,.in index by Model Number and corresponding Ca'rtificate of Compliance number is included at the back of each volume of the directory. The Summaty Report includes a listing of all users of each package design prior to ths publication date of the directory.
Shipments of radioactive material utilizing these packagings must be in accordance with the provisions of 49 CFR 173.393a and 10 CFR Part 71, as applicable.
In satisfying the requirements of Section 71.12, it is the responsibility of the licensees to insure themselves that they have a copy of the current approval and conduct their transportation activities in accordance with an NRC approved quality assurance program.
Copies of the current approval may be obtained from the U.S.
Nuclear Regulatory Commission Public Document Room files l
(see docket number listed on each certificate) at 1717 H Street, l
Washington, DC 20555.
Note that.the general license of 10 CFR 71.12 does not authorize the receipt, possession, use or transfer of byproduct, source, or special nuclear materials such authorization must be obtained pursuant to 10 CFR Parts 30 to 36, 40, 50, or 70.
NUREG-0303 VO3 RO2: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Quality Assurance Programs For Radioactive Material Packages.
- Division of Fuel Cycle & Material Safety.
January 1983.
74pp.
8303140690.
17581:075.
This directory contains a Summary Report of NRC Approved Packages 3
l
(Volume 1 ),
Certificates of Compliance (Volume 2), and a Summarg Report of NRC Approved Quality Assurance Programs for Radioactive i
Material Packages (Volume 3).
The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the U. S.
Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance number is included at the back of each volume of the directory. The Summary Report includes a listing of all users of each package design prior to the publication date of the directory.
l Shipments of radioactive material utilizing these packagings must be in accordance with the provisions of 49 CFR 173.393a and 10 CFR Part 71, as applicable.
In satisfying the requirements of Section r
71.12, it is the responsibility of the licensees to insure themselves that they have a copy of the current approval and conduct their i
transportation activities in accordance with an NRC approved quality assurance program.
Copies of the current approval may be obtained from the U.S.
Nuclear Regulatory Commission Public Document Room files (see docket number lasted on each certificate) at 1717 H Street, Washington, DC 20555.
Note that the general license of 10 CFR 71.12 does not authurixe the receipt, possessian, use or transfer nf j
byproduct, source, or special nuclear materials such notharization I
l must be obtained pursuant to 10 CFR Parts 30 to 3% 40, 50, or 70.
NUNEG-0390 VO6 NO2: TOPICAL REPORT REVIFW STATUS, Data As Of January l
20,1983. (Blue Book)
- Mar;agement Inf orma tion Branch.
ifarc h 1983.
211pp.
8304200522.
16077:001.
The primarg purpose of this report is to provide periodic j
progress reports of on-going topical report revious, to identify those topical reports for which the Nuclear Regulatory Commission (NRC) i j
staff review has been completed and, to the extent practicable.
provide the NRC management with sufficient Anformation regarding the t
conduct of the topical report program to permit tating whatever actions deemed necesary or appropriate.
r l
i f
NUREQ-0420 SO3: BAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SHOREHAM NUCLEAR POWER STATION UNIT NO.
- 1. Docket No. 50-322.
(Long Island Lighting Company)
- Office of Nuclear Reactor Regulation, Director.
February 1983.
70pp.
8302170477.'
17197:309.
Supplement No. 3 to the Safety Evaluation Report of Long Island Lighting Compeng's application for a license to operate the Shoreham Nuclear Power Staticn, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the U.
S.
Nuclear Regulatory Commission.
This supplement addresses several items that have come to light since the previous supplement was issued.
NUREG-0422 SO6: SAFETY EVALUATION REPORT RELATED TO OPERATION OF MCOUIRE NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-369 And 50-370.(Duke Power Company)
- Office of Nuclear Reactor Regulation, Director.
February 1983.
58pp.
8303230239.
17685:030.
This report supplements the " Safety Evaluation Report Related to the Operation of McGuire Nuclear Station, Units 1 and 2" (BER (NUREG-0422) issued in March 1979 by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, as applicant and owner, for licenses to operate the McGuire Nuclear Station, Units 1 and 2 4
i (Docket Nos. 50-369 and 50-37).
The facility is located in Mecklenburg County, North Carolina, about 17 mi north-northwest of Charlotte, North Carolina.
This supplement provides information related to issuance of a full power license for Unit 2 and the staff's evaluation of the licensee's compliance with requirements and conditions contained in the Unit 1 operating license.
Subject to favorable resolution of the items discussed in this report, the staff concludes that the station can be operated by the licensee without 4
endangering the health and safety of the public.
NUREG-0430 VO3 NO1: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. January 1982 - June 1982.
- Director's Offices Office of Inspection and Enforcement.
February 1983.
14pp.
[
8303100427.
17503:332.
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the infcrmation and completion of any related investigations.
Information in this report includes inventory difference data for active fuel fabrication Pacilities possessing more than cne effective kilogram of
+
i high enriched uranium, low enriched uranium, plutonium or uranium-233.
NUREG-0495 VO4 N12: SY3TEMATIC EVALUATION PRDORAM STATUS
SUMMARY
REFORT. Data As Of December 312 1982.(Suff Book)
- Office of Resource Nanagement, Director-January 1983.
59pp.
8332170028.
17192:069.
l The Systematic Evaluation Program is intendted to examine mang safety-related aspects of 11 of tha older light water reactors.
This document provides the eristing status of the review process including individual topic and overall completion status.
frJREO-0485 VOS NO1: SYSTEMATIC EVALUATION PROGRAM STATUS S')MMARY REPORT. Data As OF January 31, J DC2. (Buf f Book )
- Office of Rescu. ace Managenent. Director.
February 1983.
44PP.
8303100432.
17503:252.
l The Systematic Evaluetion Program is intended to examine many sataty-related aspects of 11 of the older light water reactors.
This document provides the existing status of the review process including 2ndividual topic and overall completion status.
NUREG-0485 VO5 NO2: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Buff Book)
- Division of Data Automation & Management Information.
- Office of Nuclear Reactor Regulation, Director.
March 1983.
65pp.
8304060014.
17864:092.
The Systematic Evaluation Program is intended to examine mang l
safety-related aspects of 11 of the older light water reactors.
This i
document provides the existing status of the review process including individual topic and overall completion status, i
NUREG-0525 R06: SAFEQUARDS
SUMMARY
EVENT LIST (SSEL).
- Office of Nuclear Material Safety & Safeguards, Director.
February 1983.
42pp.
8303160626.
17611:233.
The Safeguards Summary Event List (SSEL) provides brief summaries l
of several hundred safeguards-related events involving nuclear i
material or facilities regulated by the U.S.
Nuclear Regulatory Commission (NRC).
Events are described under the categories of bomb-related, intrusion, missing /ellegedig stolen, transportation, i
vandalism, arson, firearms-related, radiological sabotage and i
5
l I
miscellaneous.
The information contained in the event descriptions is derived primarily from official NRC reporting channels.
NUREG-0540 VO4 NO9: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. September 1-30,1982.
- Division of Technical Information &
Document Control, January 1983.
520pp.
8302100279.
17136:015.
This document is a monthly publication containing 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) nondocketed -
material received and generated by NRC pertinent to its role as a regulatory agency.
The following indexes are included: Personal Author Index, Corporate Source Index, Report Number Index, and Cross j
Reference to Principal Documents Index.
NUREG-0540 VO4 N10: TITLE LIST DF DGCUMENTS MADE PUBLICLY AVAILABLE.0ctober 1 -31,1982.
- Diviston of Technical Information &
1 Document Control.
January 1983.
506pp.
8302170062.
17193: 034.
This document is a monthly public.Stton containing descriptions of information received and generated cy the U.S.
NR C.
This information includes (1) docketed eaterial associated with civilian nuclear power i
p) ants and other uses of radicactive materials, ano (2) nondocketed Paterfil received and generaced by NRC pertinent to Ate role us a regulatory agency.
The fellouing ir.denes are includei Parsonal Author Index, Corporate Ecurce Index. Repcst Nomber Index, and Cross Reference to Principal Decouents Indum.
l NUREG-0540 VO4 N11: TITLE LIST OF 00CUMENTS MADE PUBLICLY l
AVAILABLE. November 1-30,1952.
- Division of Technical Information &
Document Control.
February 1983.
671pp.
8303100446.
17497: 0P8.
This document is a nonthly publication coutrir.ing coscriptions of i
infarmation received and generated by the U.S.
NR O.
This infernation I
includes (1) docketed material associated with civilian tuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its
- ole as e regulatory agency.
The following indexes are included: Personal Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.
NUREG-0540 VO4 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1982.
- Division of Technical Information &
Document Control.
March 1983 537pp.
8303310369.
17797:135.
This document is a monthly publication-containing descriptions of 1
information received and generated by the U.S.
NRC.
This information in:1udes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency.
The following indexes are included: Personal Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Princirs1 Documents Index.
NUREG-0566 VO2 NO4: S(ANDARDS DEVELOPMENT STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Green Book)
- Management Information Branch.
February 1983.
188p p.
8303110137.
17524:178.
The Standards Development Status Summary Report is designed for 6
1 i
scheduling, monitoring, and controlling the process by which Regulatory Standards, Guides, Reports, Petitions, and Environmental Statements are written.
It is a summary of the current schedule plans for development of the above products.
NUREG-0500 Vil N12: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Blue Book)
- Management Information Branch.
January 1983.
65pp.
8302150723.
17160:174.
Provides a review of the status of the progress of the licensing reviews for all construction permits, operating licenses, special projects and non power reactor renewals under review, as reported to 1
Congress.
NUREG-05BO V12 NO1: REQULATORY LICENSING STATUS
SUMMARY
REPORT Data As Of January 31,1983. (Blue Book)
- Management Information Branch.
February 19G3.
68pp.
G302250140.
17299:175.
Provides a review of the status of the progress of the licensing reviews for all construction permits, uperating licenses, special projects and non power reactor renew.als under review, as reported to Congress.
t4UREG-0580 V12 NO2: REQULATORY LICENSING STATUS SUMilARY PEPORT. Data As Of February 28,1933.(Blua Book)
- Management Information Branch.
March 1983.
40pp.
83031G0016.
17586:172.
Provides a review of the status of the progress of the licensing reviews for all construction permits, operating license, special project and non power reactor renewels under review, as reportes to Congress.
NUREG-0564 RO3 DRFT: ASSURING THE AVAILABILITY CF FUNDS.cDR DECOMMISSIONING NUCLEAR CACILITIES. WOOD,H.S.
08fice of State Programs, Director.
March 1983.
38pp.
B303220417.
17667:193.
This report discusses methods that can be used to assure the availability of funds for decommissioning nuclear facilities.
Although the report focuses on rommercial reactors, other NRC-licensed facilities and activities are ceasidered.
Six basic funding alternatives are analyzed by using five evaluation criteria.
The author has drawn conclusions that will form the basis of the NRC staff's recommendations on decommissioning funding to the Commission.
NUREG-0606 VO5 NO1: UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As Of February 18, 1983.(Aqua Book)
- Management Information Branch.
l February 1983.
49pp.
83031406Y8.
17569:257.
l Provides an overview of the status of the progress and plans for resolution of the generic tasks addressing " Unresolved Safety Issues" as reported to Congress.
L NUREG-0737 S01: CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS: REGUIREME NTS FOR EMERGENCY RESPONSE CAPABILITY.
- Division of Licensing.
January 1983.
28pp.
8302250043.
17301:324.
This document, Supplement 1 to NUREG-0737, is a letter from D.
Q.
Eisenhut, Director of the Division of Licensing, NRR, to licensees of operating power reactors, applicants for operating licenses, and holders of construction permits forwarding post-TMI. requirements for 7
emergency response capability which have been approved for implementation.
On October 30, 1980, the NRC staff issued NUREG-0737, which incorporated into one document all TMI-related items approved for implementation by the Commission at that time.
In this NRC report, additional clarification is provided regarding Safety Parameter Display Systems, Detailea Control Room Design Reviews, Regulatory Guide 1.97 (Revision 2) - Application to Emergency Response Facilities, Upgrade of Emergency Operating Procedures, Emergency Response Facilities, and Meteorological Data.
NUEEG-0748 VO2 N12: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of December 31,1982.(Orange Book)
- Management Information Branch.
January 1983.
292pp.
8302170022.
17191:120.
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 V03 NO1: OPERATING REACTOR 3 LICENSING ACTIONS EUMMARY. Data As Of January 31,1983.(Orange Book!
- Management Information Branch.
February 1983.
293pp.
8303140dE6.
17581:217.
The Oparating Reactors Licensing Actions Summerg is designeo' to provide the mar.agement of the Nuclear R9gulatory Commission (NRC) with an overview of licensing actions dealing with uperating power and nonpower reactors.
NUREG-0748 V01 NO2: DPERATINO LICENSING ACTIONS SUM?tARY_ Data As Dt February 28, 1963,(Orange Book) <
Management Information Branch.
March 1983.
301 p p.
8304010675.
17924:120.
The Operating Reactors Licenving Ac tions Summary is designed te provide the. management of the Nuclear Regulatory Commission (NHC) with an overview of licensing actions dealing with operating power and 1
nonpower reactors.
NUREG-0776 S05: SAFETY EVALUATIDN REPORT RELATED TO THE OPERATION DF SUSGUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-387 And 50-388. (Pennsylvania Power & Ligh*t Compangs Allegheng Electric Cooperative Incorporated)
- Division of Licensing.
March 1983.
18pp.
8304130566.
17961:141.
In April 1981, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0776) regarding the application of the Pennsylvania Power and Light Company (the applicant and/or licensee) and the Allegheng Electric Cooperative, Inc.
(co-applicant) for licenses to operate the Susquehanna Steam Electric Station, Units 1 and 2, located on a site in Luzerne County, Pennsylvania.
Supplement 1 to NUREG-0776 was issued in June 1981 and addressed several outstanding issues.
Supplement 2 was issued in September 1981 and addressed additional outstanding issues.
Supplement 2 also contains NRS staff responses to the comments made by the Advisory Committee on Reactor Safeguards in its report dated August 11, 1981.
Supplement 3 was issued in July 1982 and addressed five items that remained open and closed them out.
On July 17, 1982, Operating License NPF-14 was issued to allow Unit 1 operation at power levels not to exceed 5% of rated power.
Supplement 4 was issued November 1982 and discusses the resolution of several license conditions.
On 8
November 12, 1982, Operatg License NPF-14 was amended to remove the 5% power restriction, thereby permitting full-power operation of Unit 1.
This supplement, No.
5, addresses s2veral issues that require resolution before licensing operation of Unit 2.
NUREC-0798 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF ENRICO FERMI ATDMIC POWER PLANT, UNIT NO.
- 2. Docket No. 50-341.
(Detroit Edison Company)
- Office of Nuclear Reactor Regulation, Director.
January 1983.
190pp.
8302170216.
17189:319.
Supplement No. 3 to the Safety Evaluation Report related to the operation of the Enrico Fermi Atomic Power Plant, Unit 2 provides the staff's evaluation of additional information submitted by the applicant regarding outstanding review issues identified in Supplement No. 2 to the Safety Evaluation Report, dated January 1982.
NUP.EQ-OO22: INTECRATED PLANT CAFETY ASSESSMENT. SYSTEMATIC EVALUATION PROGRAN,0YSTER CREEK NUCLEAR GENERATINO STATION. Doc ket No.
50-219. (;F U Nuclear Corporation And Jersey Central Fower & Light Company)
- Division oF Licensing.
January 1993.
Sa3pp.
G302200005 17323:299 The Nuclear Regulatory Commis51on (NRC) has published its Final Integrated Plant, Safety Assessment Report (IPSAR) (NUREG-0822), under the scope of tne Systematic Evaluation Program (SEP), for the Dyster Creek Nuclear Generating Station located in Ocean County, New Jersey, and cperaten' by QPU Neclear Corporation and Jersey Centra 7 Power &
Light Company (colicenseer).
The SEP was initiated by the NRC to review the design cf claer operating nuclear plants to recenfirm and documen t their saf ety.
This report documents the review completed under tne SEP for the Oyster Creek Nuc19ar Generatann Stetion.
The review has provfoed for (1) en essessment of the significance of differences between current technical positions on set cted safety issues and those that existed when the Oyster Creek plant was licensed, (2) a batis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when all supplements to the Final IPSAR and the Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued.
The report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in September 1982.
The Final IPSAR and its supplements will form part of the bases for considering the conversion of the existing provisional operating license to a full-term operating license.
NUR EG-0023: INTEGRATED SAFETY ASSESSMENT SYSTEMATIC EVALUATION PRDORAM, DRESDEN NUOLEAR POWER STATION UNIT 2. Docket No. 50-237. (Commonwealth Edison Company)
- Division of Licensing.
February 1983.
590pp.
8303090709.
17478:003.
The Nuclear Regulatory Commission (NRC) has published its Final Integrated Plant Safety Assessment Report (IPSAR) (NOREG-0823), under the scope of the Systematic Evaluation Program (SEP), for Commonwealth Edison Company's Dresden Nuclear Power Station, Unit 2 located in Grundy County, Illinois.
The SEP was initiated by the NRC to review the design of older operating nuclear reactor plar.ts to reconfirm and document their safety.
This report documents the review completed under the SEP for Dresden Unit 2.
The review has provided for (1) an 9
~
assessment of the significance of differences between current technical positions on selected safety issues and those that existed when Dresden Unit 2 was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when all supplements l
to the Final IPSAR and Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued.
The report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in October 1982.
The Final IPSAR and its supplements will form part of the bases for considering the conversion of the existing provisional operating license to a full-term operating license.
NUREG-0824: IN1EGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATIDN PROGRAM-MILLSTONE NUCLEAR POWER STATION, UNIT 1. Docket No.
SO-245.(Northeast Nuclear Energy Company) e Divisien of Licensing.
February 1983.
540pp.
8303230417.
17675:281.
The Systematic Evaluation Program was initiated in February 1977 by the U.S.
Nuclear Regulatory Commission to review the designs of cider operating nuclear reactor plants to reconfirm and dccument their 7
safety.
The review provides (1) an assessment of isow these plants l
compare with current licensing safety requirements relating tc i
selected issues. (2) a basis for deciding on how these differences r
should be resolved in an integrated plant review, and (3) a documenteo l
evt.luation of plant safety when all supplenests to thw Final l
Integrated Plant Safety Assessment Report and the Ssfety Evaluation I
Caport for convertang the Ifcensee from a provisional to a full-tern j
licarse have been issued.
l This report documents the review of the Hillstor.e Nuclear Power i
Station, Unit 1, operated by Northeast NucIcar Energy Company lec2ted in Waterford, Connecticut.
Millstcne Nuclear Power Station. Unit 1 is l
one of ten plants reviewed under Phase II of this program.
This l
report indicates how 137 topics selected for review under Phase I of the program were addressed.
Equipment and procedural changes have I
been identified as a result of the review.
It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license.
This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in November 1982.
NUREG-0825: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM-YANKEE NUCLEAR POWER STATION. Doc k et No.
j 50-029.(Yankee Atomic Electric Company)
- Division of Licensing.
February 1983.
548pp.
8303090583.
17474:001.
The Systematic Evaluation Program was initiated in February 1977 by the U.S.
Nuclear Regulatory Commission to review the designs of older operating plants to reconfirm and document their safety.
The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety.
This report documents the review of the Yankee Nuclear Power Station, operated by the Yankee Atomic Electric Company, located in Framingham, Massachusetts.
Yankee Nuclear Power Station is one of ten 10
~
l i
i plants reviewed under Phase II of this program.
This report indicates how 137 topics selected for review under Phase I of the program were addressed.
Equipment and procedural changes have been identified as a result of the review.
i NUREG-0837 VO2 NO3: TLD DIRECT RADIATION MONITORINC NETkORK. Progress Report July-September 1982. CDSTELLO F.s THOMPSON,T.s CDEEN,L.
1 Region 1.
Office of Director.
March 1983.
199pp.
8303D90694.
17737:059.
This report provides the status and results of'che NRC Thermoluminescent Dosimeter (TLD) Director Radiation Monitoring Network.
It presents the radiation levels measured in the vicinity of NRC licensed facility sites throughout the country for the-third
(
quarter of $982.
NUREG-0845: AGENCY PROCEDURES FDR THE NRC INCIDENT RESPONEE PLAN.
I WEIN3TEINcE.
Incident Response & Development Branch.
Fobcuary 1983.
296pp.
8303090505.
i 747S: 189.
The NRC Incident Response Plan, NUREG-0728/HC 05C2, describes the functions of the NRC during an incident and the kindt of actions that comprise an NRC responss.
Tha NHC response plan uill be activated in i
arcordance with thresho?d criteria descrined in the plan fcr incidents occuring at nuclear reactora. Fuel facilities and materials licensees, during tear.sportation of licensed meterial. and for threats against fec111tien oc licensed material.
In contrast'to the gen 2ral overview providuJ by the Plun, the purpose of these agency procedures is to delineate:
1.
The manner in which each planned response fcnction is performeds i'.
The criteria for waking those response decisions which
~
can be precplanneds 3.
The hdormation and other resources needed during i
a responsa.
An inexperienced but qualified person should be able to perform functi.)ns assigned by the Plan and make necessary decasions, given the specified information, by becoming familiar with these precedures.
This rule of thumb has been used to determine the amount of detail in which the agency procedures are described.
These procedures form a foundation for the training of response personnel both in their r.: ara,a1 working environment and during planned emergency exercises.
These procedures also form a ready reference or reminder checklist for technical team members and managers during a response.
NUREG-0851: NOMOORAMS FOR EVALUATION OF DOSES FROM FINIYE NODLE GAS CLOUDS. PASCIAK,W.J.s FAIROBENT,L.A.: WAN91.ER, M. E. 4 at al.
Division of Systems Integration (post 811005).
January 1983.
249pp.
8302250091.
17300:001.
This document describes a simple mothef using a finite-cloud approach for calculating doses due to elevated atmospheric releases of noble gases.
This method uses nomogram, which allow doses at selected points downwind to be read directly, and regression equations, which permit simple calculation of doses.
Nomograms and regression equations for 9 radionuclides, 2 release heights (ground and 100 meter elevated), and 7 Pasquill stability classes are presented for typical wind speeds. Dose calculations may be made for gsmes air dose, whole body dose, or skin dose.
The results of this method are compared with the results of three other finite plume model dose calculational
~
11
2 e
e l
~
techniques.
These nomograms and regressiort equatio'ns can be used tn situations where finite modeling 'is' required,and computer routine or a computer is not available.
Applicants for-Itcenses For nuclear power treactors'a.ng find;them useful in developinD.their offsite-dose t' calculation manual.
Licensees may fAno' them,pseful in evaluating compliance with regulations, either for ahnormal or routine releases.
end in developing ne implementing emergen,dy' response plans.
They can ha applied nob.onig to commercial light water reactors _Lut also to f
reactors used to. generate radjoisotopes or to This be used byfthe DSI Staff in,research reactors.
the evaluation of final j
d ocument will also emergency plans for the Divisicn of Emergency Preparedness.
i l
,/
r NUREG-0852 SO1: SAFETY EVALUATION ftLPORT RELATED TO THE FINAL DESIGN OF
(
THE STANDARD NUCLEAR STEAM SUPPLY'8EFE9ENCE SYSTEM CESSAR SYSTEM
- 80. Docket No. STN 50-a70.(Combustion Enginsering, Incorporated)
- Office of Nuclear Reactor Regulation, Director.
March 17 B3.
eor,.
8304130530.
27960:001.
,c Supplement No.
1 to tha<3afety hvaluation Report for the application filed by Combustion Engineeringi inc. For a Final Oesign Approval for the Combustion Engineering Standard Safetg Analgs s Report (STN 50-470) hus been prepared b9 the Office et Nucleat ficac tor 9egulation of the Nuclear Regulatory Comsissien.
The purpose of this
'4upplement is to update the Gafety Evalcatio'n hg providsng (i) the evaluation of additional information submita7J by the applicanc since the safety Evaluation Report was issued. (2) the evaluatir,a of the mattets t h s-staff Sad ur:dar review when the Safety Evaluation Kaport l
was issued, and (3) the response to come.ents made by the r.dviaary l
Ccemittee.on Reactor Safeguards.
j NOREG-0057 SO4: SAFETY ENLUATION REPORT RELAYED TO T!E OPERATICE OF l
P ALO VERDE HUCLEAR OENERATINO STATION, L* NITS 1,2 AND 3. Doc ket Not GTN le 50-52P, STN 50-SCV And STN 50-530. ( Arizona Public Service Compang, et al.)
- Office cf Nuclear Reactor Regulation, Director.
March 1983.
45pp.
8303300463.
17763:304.
j SJpplement No. 4 to the Safety Evaluation Report for the a pp 1'J cati on filed og Arizona Public Service Company, et al, for liceeses to operate the Palo Verde Nuclear Generating Station, Units
~
1, 2 and 3 (Docket Nos. STN 50-5DS/529/530), located in Maricopa County, Arizona"hes been prepared by the,0ffice of Nuclear Reactor
- +
Regulation of the Nuclear Regulatoro commissian.
The purpose of this supplement is to update the Safeiv Evaluation Report by providing (1) i an evaluation =of' additional inF(enation submitted by the applicants l
since Supplement No. 3 was issued, and (2),an evaluation of the l
matters that*the staff had under revism when Supplement No. 3 was issued.
i NUREQ-0071 VO2 NO1: ~
SUMMARY
INFORMATIDN REPORT.0ctober 1 - December 31, 1982.- (Brown Book)
- Management Information Branch.
February 1983.
52pp.
8302280024.
17303:270.
Provides summary data concerning NRC and its licensees for general use_by the Chairman, other Commissioners and Commission staff o f f i c' e s, the Executive Director for Operations, and the Office Directors.
~
6 12
.n
~
L sy s
T
\\
NUREO 0876 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF BYRON STATION, UNITS 1 AND 2. Coc k etiNos. STN S0-454 And STN 50-455.
(Commonwealth Edison Cespang)'* OPfice of Nuclear Reactor Regulation, Director.
January 1983.
30pp.
8302170481.
17197:264.
Supplement No, 2 to,the Safety Evaluation Report related to Commonwealth Edison Company 's application for licenses to operate the Byron Station, Uni t s 1 and 2, located in Rockvale Township, Ogle County, Illinois,'has been preparedsby the Office of Nuclear Reactor Regulation of the U.S' Nuclear Regulatory Commission.
This supplement l
' reportikthe, status of certain items that had not been resolved at the l
3(1-time of2 publication. of the Safety Evaluation Report.
.r h
q
.g.
._g-l NUR EG-OD&5 102: US NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING s
' GUIDANCE 'f 983.
- NHC f No Detailed Affiliation Given.
January 1983.
l 23pp. s8301190179.
16861:092.
r The purpose of the Policy and Planning 3 0uidance is to provide a
. ' common basis for establishing priorities throughout the NRC.
The guidance should also'be used for developing. budget requirements.
The goal of the document is to make the'whole regulatory process more e effective and more e f fi c i ent.-
The document is organized in terms of seven ' major themes: Safe Operation of Licensed Plantss Near-Term Licensing ^ Problems and Responsess Coordinating Regulatorg Requirementsa Improving the Licensing Processs Supporting New Initiatives in Waste Management and the Cleanup of Three Mile Islands Improving Related~ Regulatory Tools and Safeguards.
The policy
'section in each theme is intended to establish a general framework for siaping NRC plans and programs.
Planning guidance is furnished in thost ~ areas where the Commission believes more detail is warranted to 9 7
, Meetospecific priorities and schedules or where major assumptions are
,b
.'needed.for program development.
Guidance with respect to each and
.every activity within NRC is not furnished, since it is not intended
~
that th e documen~h b'e all-inclusive.
However this should not be perceived as.a Commission belief that other areas are not important to
. protecting the public health and safety.
NUREG-OB87 SO2: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF l
PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2. Doc ket Nos.,50-440 And
'^
50-441.(Cleveland Electric ~ Illuminating Company)
- Office of Nuclear Reactor Regulation, Director?
January 1983.
60pp.
8302100276.
17135:286.
3 Supplement No. 2 to the' Safety Evaluation Report on the application filed by the Cleveland Electric Illuminating Company on b ehalf of itself and as ' agent for the Duquesne Light Company, theichio s
' Edison Company, the Pennsgivania Power Company and the Toledo Edison Company-(the Central Area Power Coordination Group, CAPCO), as applicants and owners, for a license to operate the Perry Nuclear 1
Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441).
The
~
. facility is located near Leke Erie in Lake County, Ohio.
This I
supplement has been prepared by the Office of Nuclear Reactor
-Regulation of the U.
S.
Nuclear Regulatory Commission and reports the i
status of certain items that had not been resolved at the time of publication of the Safeta Evaluation Report.
NUREG-OB9E: SAFETY EVALUATION REPORT REL^ATED TO THE OPERATION OF SEAbH03M. STATION, UNITS'l AND 2.Dochat Nos. 50-443 And 50-444.(Public Service Compang Of-New Hampshire,et al.)
- Division of Licensing.
l March'1983. '660pp.
8303220444.
17460:112, 13 '
i
i The Safety Evaluation Report for the application filed by Public Service Company of New Hampshire, et al for licenses to operate the Seabrook Station, Units 1 and 2 (Docket Nos. 50-443 and 50-444) i located in Rockingham County, New Hampshire, has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory i
Commission.
Subject to favorable resolution of the items discussed in the Safety Evaluation Report, the staff concludes that the plant can I
be operated by the Public Service Company of New Hampshire without endangering the health and safety of the public.
I NUR EG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN. LARKINS,J.T.s C UNNINGHAM, M.
Office of Nuclear Regulatory Research, Director.
January 1983.
84pp.
8302170029.
17192:149.
The Severe Accident Research Plan (SARP) covers research for the time period of January 1982 through January 1986, aimed at providing technical information necessary to support regulatory decisions in the severe accident area for existing or planned nuclear power plants.
j SARP has been formulated to develop generic bases to determine how safe the plants are and generic guidance on where and how their level of safety ought to be improved.
There are thirteen program elements in the plan and the work is phased in two parts, with the first phase l
being completed in earig 1984, at which time an assessment will be made whether or not any major changes will be recommended to the Commission for operating plants to handle severe accidents.
i Additionally at this time, all of the thirteen program elements in Chapter 5 will be reviewed and assessed in terms of how much additional work is necessary and where major impacts in probabilistic risk assessements might be achieved.
The plan covers work sponsored by the NRC's Office of Nuclear Regulatory Research, however, work being sponsored by the nuclear industry and foreign countries will also be utilized as much as practical.
i NUREG-0907: ACCEPTANCE CRITERIA FOR DETERMINING ARMED RESPONGE FORCE SIZE AT NUCLEAR POWER PLANTS.
- Power Reactor Safeguards Licensing Branch.
February 1983.
14pp.
8303100429.
17503:317.
This guidance document contains acceptance criteria to be used in the NRC license review process.
It consists of a scored worksheet and guidelines for interpreting worksheet score that can be used in determining the adequacy of the armed response force size at nuclear power reactor facility.
i l
NUREG-0921: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF l
C ATAWBA NUCLEAR STATION, UNITS 1 AND 2. Doc k et Nos. 50-413 And 414.(Duke Power Company,et al.) a Office of Nuclear Reactor
)
i l
Regulation, Director.
January 1983.
288p p.
8301190461.
16851:001, t
This Final Environmental Statement contains the second assessment of the impact associated with the operation of the Catawba Nuclear J
l Station, Units 1 and 2, pursuant to the National Environmental Policy act of 1969 (NEPA) and 10 CFR 51, as amended, of the NRC regulations.
This statement examines: the affected environment, environmental 1
conseguences and mitigating actions, and environmental and economic benefits and costs.
Land use and terrestrial and aquatic-ecological impacts will be small.
Operational impacts to hisioric and archeological sites will be negligible.
The effects of routine operations, energy transmissions, and periodic maintenance of rights-of-way and transmission facilities should not jeopardize any populations of threatened or endangered species.
No significant 14
4 l
impacts are anticipated from normal operational releases of radioactivity.
The risk associated with accidental radiation exposure is very low.
The not socioeconomic effects of the project will be a
beneficial.
The action called for is the issuance of operating licenses for Catawba Nuclear Stations, Units 1 and 2.
NUREG-0936 Vol NO4: NRC REQULATORY AGENDA.Guarterly Report. September
-December 1982.
- Division of Rules and Records.
January 1983 210pp.
8302280004, 17313:113.
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 been received by the Commission and are pending disposition by the Commission.
The Regulatory Agenda is updated and l
issued each guarter.
The Agendas for April and October are published in their entirety in the Federal Register while a notice of availability is published in the Federal Register for the January and July Agendas.
NUREG-0940 VO1 NO4: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS 1
RESOLVED.Guarterly Progress Report.0ctober - December 1982.
- Director's Office, Office of Inspection and Enforcement.
January 1983.
152pp.
8302220438.
17237:025.
This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (October - December 1982) and includes copies of letters, notices, and orders sent by the Nuclear Regulatory Commission to the licensee with respect to the enforcement action.
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.
This publication is issued on a quarterly basis to include significant enforcement actions resolveo during the preceding guarter.
l l
NUREG-0943: THREADED FASTENER EXPERIENCE IN NUCLEAR POWER PLANTS.
KOO, W. H.
Division of Licensing.
January 1983.
30pp.
8302250112.
17299:284.
This report identifies 44 incidents of threaded-fastener degradation and failure in nuclear power plants from October 1964 to March 1982.
It provides an overview of some of the threaded-fastener problems that have occurred since 1964.
Safety implications of these incidents are discussed, and short-term regulatory actions and ongoing long-term regulatory actions are described.
Information included in this report represents the current NRC staff understanding of each I
issue.
l
[
NUREG-0947: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE TEXAS A&M UNIVERSITY TRIGA REACTOR. Docket No. 50-128. License R-83.
- Office of Nuclear Reactor Regulation, Director.
March 1983.
102pp.
8304120683.
17942:104.
This Befety Evaluation Report for the application filed by the fosas A&M University for a renewal of operating license R-83 to continue to operate a 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 Texas A&M University and is located 3 miles from the Texas A&M campus in College Station. Texas, Brazos County, Texas.
The staff concludes that the 15 i
i TRIGA reactor facility can continue to be operated by the Texas A&M without endangering the health and safety of the public.
NUREG-0948: SPECIAL INSPECTION REPDRT OF QUADREX CORPORATION REPORT ON DESIGN REVIEW OF BROWN & ROOT ENGINEERING WORK FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And Power Company) OBERG,C.R.
Division of Resident, Reactor Project & Engineering Programs.
January 1993.
466pp.
8302170221.
17188:213.
This report is the NRC staff's review of the South Texas Project's "Guadres Report."
The Guadrez Report contained a limited assessment of the Brown & Root engineering design ef forts for STP.
This design review was conducted during January-May 1981 by the l
Guadron Corporation at the request of Houston Lighting & Power Company.
Subsequently, the Guadrex Report will become the subject of a public hearing to be held by the ASLB at a later date.
In November 1981. HL&P announced that Bechtel Power Corporation would replace B&R as the AE.
As a result, Bechtel was assigned the task of resolving the Guadron findings.
Their report, EN-619 (Guadrex Work Package) was completed in September 1982.
In February 1982, Region IV requested i
HL&P to provide information on their transition program pursuant to 10 CFR 50.54(f).
Information on the Guadrex Report resolution was specifically requested.
A NRC team composed of personnel from DIE, l
NRR, and Region IV conducted the staff review, principally in the l
Bechtel offices in Houston, Texas.
This report details the results of the review of approximate 1g 351 separate Guadrex findings.
Each finding was reviewed for reportability under 10 CFR 50.55(e), safety significance and generic implications, and adequacy of resolution.
Assessments and conclusions are given in the report along with the individual inspection findings.
NUR EG-0953: FY 1994/85 BUDGET ESTIMATES.
- Office of Resource Management, Director.
January 1983.
92pp.
8302250113.
17302:099.
This report contains the fiscal year budget justifications to Congress.
The budget estimates for salaries and expenses for fiscal year 1984-1985 provide for obligations of $466,800.00 to be funded to total by a new appropriation.
NUR EG-0954: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF CATAWBA NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-414.
(Duke Power Compangeet al.)
- Office of Nuclear Reactor Regulation.
Director.
February 1983.
566pp.
8302280007.
17318:001.
The Safety Evaluation Report for the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Electric Membership Corporation, and Saluda River Electric I
Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414) has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is located in York County, South Carolina, approsimately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie.
Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public.
t l
16
E N
NUREG-0959: USER'S QUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
WIDMAYER,D.A.
Division of Waste Management.
January 1983.
72pp.
8302170483.
17198:073.
This document explains how to use the Impact Analysis Codes used in the Draft Environmental Impact Statement (DEIS) (NUREG-0782, Vol.
1-4) supporting 10 CFR 61, " Licensing Requirements for Land Disposal of Radioactive Waste."
The mathematical development of the Impact Analysis Codes and other information necessary to understand the results of using the Codes is contained in the DEIS, and in a supporting document, " Data Base for Radioactive Waste Management" (NUREG-1759, Vol.
1-3).
NUREG-0960 V01: DRAFT SITE CHARACTERIZATION ANALYSIS OF THE SITE l
CHARACTERIZATION REPORT FOR THE BASALTa.MASTE ISOLATION PROJECT. Main Report And Appendices A Through D.
- Division of Waste Management.
March 1983.
414pp.
8304200500.
18115:028.
Dn November 12,1982 the U.S.
Department of Energy submitted to the U.S.
Nuclear Regulatory Commission the " Site Characterization Report for the Basalt Weste Isolation Project" (DOE /RL 82-3).
The Basalt Weste Isolation Project is located on DOE's Hanford Reservation in the State of Washington.
NUREG-0960 contains the detailed analysis, by the NRC staff, of the site characterization report.
Supporting technical material is contained in Appendices A through W.
NUREG-0960 V02: DRAFT SITE CHARACTERIZATION ANALYSIS OF THE SITE CHARACTERIZATION REPORT FOR THE BABALT WASTE ISOLATION PROJECT. Appendices E-W.
- Division of Waste Management.
March 1983.
430pp.
8304200532.
18075:289.
Dn November 12, 1982, the U.S.
Department of Energy submitted to the U.S.
Nuclear Regulatory Commission the " Site Characterization Report for the Basalt Waste Isolation Project" (DOE /RL 82-3).
The Basalt Waste Isolation Project is located on DOE's Hanford Reservation l
in the State of Washington.
NUREG-0960 contains the detailed analysis, by the NRC staff, of the site characterization report.
Supporting technical material is contained in Appendices A through W.
I NUREG-0963: REVIEW AND EVALUATION OF THE NUCLEAR REQULATORY COMMISSION SAFETY REGEARCH PR00 RAM FOR FISCAL YEARS 1994 AND 1985.
- ACRS -
Advisory Committee on Reactor Safeguards. " February 1983.
86PP.
8303090715.
17477:282.
Public Law 95-209 includes a requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on l
the safety research program of the Nuclear Regulatory Commission.
This taport presents the results of the ACRS review and evaluation of l
the NRC safety research program for Fiscal Years 1984 and 1985.
The report contains a number of comments and recommendations.
NUREG-0964: TECHNICAL SPECIFICATIONS FOR MCOUIRE NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-369 And 50-370.(Duke Power Company)
ANDERSON,F.
Division of Licensing.
March 1983.
592pp.
8303220439.
17670:343.
The McGuire Nuclear Station, Units 1 and 2.
Technical Specifications were prepared by the U.S.
Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements 17
i applicable to a nuclear reactor facility as set forth in Section 50.36 of 10CFR Part 50 for the protection of the health and safety of the public.
4 NUR EG-0965: NRC INVENTORY OF DAMG. LEAR,C.E.s THOMPSON,D.D.
hWIC - No Detailed Affiliation Given.
January 1983.
98pp.
8302250115.
17302:001.
The NRC Inventory of Dams has been prepared as required by the charter of the NRC Dam Safety Officer.
The inventory lists 51 dans associated with nuclear power plant sites and 14 uranium tailings dans (licensed by NRC) in the U.S.
as of February 1, 1982.
Of the 85 listed nuclear power plants (145 units), 26 plants obtain cooling water from impoundments formed by dams.
The 51 dans associated with the plants are:
(a) located on a plant site (29 dans at 15 plant i
sites), (b) located off-site but provide plant cooling water (18 dans at 11 additional plant sites), (c) located upstream from a plant (4 dams) -- they have been identified as dams whose failure, and ensuing plant flooding, could result in a radiological risk to the public health and safety.
The dans that might be considered NRC's responsibility in terms of the Federal dam safety program are identifieds this group of dans (20 on nuclear power plant sites and 14 uranium mill tailings dams) was obtained by eliminating dans that do not pose a flooding hazard (e.g.
submerged dams) and dams that are regulated by another Federal agency.
The report includes the principal design features of all dans and related useful information.
NUR EG-0966: SAFETY EVALUATION REPORT RELATED TO THE D2/D3 STEAM QENERATOR DESION MODIFICATION.
- Division of Licensing.
March 1983.
147pp.
8303310361.
17798:312.
This Safety Evaluation Report (SER) related to the D2/D3 steam generator design modification has been prepared by the Office of Nuclear Regulatory Commission.
The purpose of this SER is to issue the staf f's evaluation of the acceptability of the design modification for both installation and full-power operation in the D2/D3 steam generators based on the Design Review Panel Report of January 1983.
NUR EG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATION PROGRAM
- A USE OF PROBABILITY IN DECISION MAMING. REITER,L.s JACM90N,R.E.
Division of Engineering.
March 1983.
65pp.
8303300435.
17763:181.
This document presents the U.S.
Nuclear Regulatory Commission (NRC) Geosciences Branch review and recommendations with rsspect to earthquake ground motion considerations in the Systematic Evaluation l
Program (SEP) Phase I and II.
It evaluates the probabilistic estimates presented in the 5 volume report entitled " Seismic Hazard Analysis" (NUREO CR-1582) and compares and modifies them to take into account deterministic estimates.
It presents the NRC's Geosciences i
Branch first approach to utilizing complex state-of-the art probabilistic studies in an area where probabilistic criteria have not get been set and where decisions for specific plants have been previously made in a non-probabilistic way.
NUREG-0968 VO1: SAFETY EVALUATION REPORT RELATED TO THE CONSTRUCTION OF
[
THE CLINCH RIVER BREEDER REACTOR PLANT. Main Report. Docket No.
l 50-537.(U.S. Department of Energy, Tennessee Valley Authority And l
Project Management Corporation)
- Clinch River Breeder Reactor Program 0ffice.
March 1983.
930pp.
8303300448.
17772:134.
l 18
The Safety Evaluation Report for the application by the United States Department of Energy, Tennessee Valley Authority, and the Project Management Corporation, as applicants and owners, for a license to construct the Clinch River Breeder Reactor Plant (Docket No. 50-537) has been prepared by the Office of Nuclear Reactor Regulatfor of the United States Nuclear Regulatory Commission.
The facility will be located on the Clinch River approximately 12 miles southwest of downtown Oak Ridge and 25 miles west of Knoxville.
Subject to resolution of the items discussed in this report, the staff concludes that the construction permit requested by the applicants should be issued.
NUREG-0968 VO2: SAFETY EVALUATION REPORT RELATED TO THE CONSTRUCTION OF THE CLINCH RIVER BREEDER REACTOR PLANT. Appendices A-H. Docket No.
50-537.(U.S. Department of Energy, Tennessee Valley Authority And Project Management Corporation)
- Clinch River Breeder Reactor Program Office.
March 1983.
373pp.
8303300092.
17752:001.
The Safety Evaluation Report for the application by the United States Department of Energy. Tennessee Valley Authority, and the Project Management Corporation, as applicants and owners, for a license to construct the Clinch River Breeder Reactor Plant (Docket No. 50-537) has been prepared by the Office of Nuclear Reactor Regulation of the United States Nuclear ReDulatory Commission. The facility will be located on the Clinch River approximately 12 miles t
southwest of downtown Oak Ridge and 25 miles west of Knoxville, Tennessee.
Subject to resolution of the items discussed in this report, the staff concludes + ggt the construction permit requested by the applicants should be issuef.
NUREG-0975 VO1: COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEERING TECHNOLOGY. Annual Report For FY 1982.
- Materials Engineering Branch.
March 1983.
275pp.
8304050661.
17849:028.
/
This report presents summaries of the research work performed during Fiscal Year 1982 by laboratories and organizations under contracts administered by the NRC's Materials Engineering Branch, Office of Nuclear Regulatory Research.
Each contractor has written a more complete and detailed annual report of their work which can be obtained by writing to NRCs however, we believe it is useful to have a summary of each contractor's efforts for the year combined into one volume.
l l
NUREG-0977: NRC FACT-FINDING TASK FORCE REPORT ON THE ATWS EVENT AT I
SALEM NUCLEAR GENERATING STATION, UNIT 1.ON FEBRUARY 22 AND 25,1983.
- Region 1.
Office of Director.
March 1983.
731pp.
8303310358.
17799:099.
An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to an anticipated transient without scram (ATWS) event at the Public Service Electric and Gas Company's Balen Generating Station, Unit 1 on February 25, 1993.
The charter of the l
Task Force was to determine the factual information pertinent to management and administrative controls which should have ensured proper operation of the Reactor Trip Circuit Breakers in the Solid State Protection System.
This report documents the findings of the l
Task Force along with its conclusions and recommendations.
19
t NUREQ/CP-OO27 VO1: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -Geptember 2,1982.
- American Nuclear Society.
February 1983.
749pp.
8303140707.
17573:001.
The Proceedings of the International Meeting on Thermal Nuclear Reactor Safety, held at Chicago, Illinois, August 29-September 2, 1982, contain the entire collection of papers submitted for presentation at the meeting, as well as two special addresses and four summarizing review articles.
The papers deal with a wide spectrum of subjects pertaining to the area of thermal nuclear reactor safety, including: licensing criteria, safety goals, probabilistic risk assessment, reliability analysis, safety-related operational experience, man / machine interface, human factors, transient analysis, loss-of-coolant analysis, structural analysis, fuel performance evaluation, severe accident analysis, radiological source term i
evaluation, pressurized thermal shock.
In addition to papers on the above technical subjects, the Proceedings contain a number of papers describing safety-related programs in a number of countries, including Argentina, Brazil, Canada, Fed. Rep. of Germany, Finland, France, Greece, Italy, Japan, Mexico. Spain. Sweden, and United Kingdom.
The Meeting was Jointly sponsored by the American Nuclear Society, the European Nuclear Society, the Canadian Nuclear Society, and the Japan Atomic Energy Society.
It was, furthermore, organized and conducted in cooperation with the U.
S.
Nuclear Regulatory Commission and the International Atomic Energy Ageneg.
NUREQ/CP-OO27 VO2: PROCEEDINGS OF THE INTERNATIDNAL MEETING ON THERMAL l
NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
- American Nuclear Society.
February 1983.
750p p.
8303140709.
17576:110.
The Proceedings of the International Meeting on Thermal Nuclear i
Reactor Safety, held at Chicago, Illinois, August 29-September 2, 1982, contain the entire collection of papers submitted for presentation at the meeting, as well as two special addresses, and four summarizing review articles.
The papers deal with a wide spectrum of subjects pertaining to the area of thermal nuclear reactor l
safety, including: licensing criteria, safety goals, probabilistic l
risk assessment, reliability analysis, safety-related operational experience, man / machine interface, human factors, transient analysis, loss-of-coolant analysis, structural analysis, fuel performance evaluation, severe accident analysis, radiological source term evaluation, pressurized thermal shock.
In addition to papers on the above technical subjects, the Proceedings contain a number of papers describing safety-related programs in a number of countries, including Argentina, Brazil, Canada, Fed. Rep. of Germany, Finland, France, Greece. Italy, Japan, Mexico. Spain. Sweden, and United Kingdom.
The Meeting was jointig sponsored by the American Nuclear Society, the European Nuclear Society, the Canadian Nuclear Society, and the Japan Atomic Energy Gociety.
It was, furthermore, organized and conducted in cooperation with the U.
S.
Nuclear Regulatory Commission and the International Atomic Energy Agency.
NUREQ/CP-OO27 VO3: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
- American Nuclear Society.
February 1983.
709pp.
8303140710.
17570:019.
The Proceedings of the International Meeting on Thermal Nuclear Reactor Safety, held at Chicago, Illinois. August 29-September 2, 20
1 1982, contain the entire collection of papers submitted for presentation at the meeting, as well as two special addresses, and four summarizing review articles.
The papers deal with a wide spectrum of subjects pertaining to the area of thermal nuclear reactor safety, including: licensing criteria, safety goals, probabilistic risk assessment, reliability analysis, safety-related operational experience, man / machine interface, human factors, transient analysis, loss-of-coolant analysis, structural analysis, fuel performance evaluation, severe accident analysis, radiological source term evaluation, pressurized thermal shock.
In addition to papers on the above technical subjects, the Proceedings contain a number of papers describing safety-related programs in a number of countries, including Argentina, Brazil, Canada, Fed. Rep. of Germany, Finland, France, Greece. Italy, Japan, Mexico, Spain, Sweden, and United Kingdom.
The l
Meeting was jointig sponsored by the American Nuclear Society, the European Nuclear Society, the Canadian Nuclear Society, and the Japan i
Atomic Energy Society.
It was, furthermore, organized and conducted in cooperation with the U.
S.
Nuclear Regulatory Commission and the International Atomic Energy Agency.
NUREC/CP-OO28 VO3: PROCEEDINGS OF THE SYMPOSIUM ON LOW-LEVEL WASTE DISPOSAL: Facility Design, Construction And Operating Practices.
YALCINTAS.M.G.
Oak Ridge National Laboratory.
March 1983.
47pp.
8303300472.
CONF-820911.
17767:011.
This document is a compilation of the papers presented at the Low-Level Waste Symposium on Facility Design, Construction, and Operating Practices.
Both proven practices, as well as innovative Practices, were discussed as a means to supplement site Characteristics.
j NUREQ/CP-OO41 VO1: PROCEEDINGS OF THE TENTH WATER REACTOR BAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A.
Szewlewiez, S. A.
January 1983.
443pp.
8302170487.
17198:209.
This report is a compilation of papers which were presented at the Tenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, October 12-15, 1982.
It consists of six volumes.
The papers describe recent results and planning of safety research work sponsored by the Office of Nuclear Regulatory Research, NRC.
It also includes a number of invited papers on water reactor safety research prepared by the Electric Power Research Institute and various government and industry organizations from Europe and Japan.
r NUREQ/CP-OO41 VO2: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A.
Szawlewicz, S. A.
January 1983.
504pp.
8302220441, 17238:001.
This report is a compilation of papers which were presented at the Tenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, October 12-15, 1982.
It consists of six volumes.
The papers describe recent results and planning of safety research work-sponsored by the Office of Nuclear Regulatory Research, NRC.
It also includes a number of invited papers on water reactor safety research preparea by the Electric Power Research Institute and various government and industry organizations from Europe and Japan.
21
_=
l 1
MUREQ/CP-0041 VO3: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A.
Srawlewic:,
S. A.
January 1983.
312pp.
8302220442.
17236:001.
This report is a compilation of papers which were presented at r
the Tenth Water Reactor Sa'ety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, October 12-15, 1982.
It consists of six volumes.
The papers describe recent results and planning of safety research work sponsored by the Office i
of Nuclear Regulatory Research, NRC.
It also includes a number of l
invited papers on water reactor safety research prepared by the Electric Power Research Institute and various government and industry organizations from Europe and Japan.
i NUREQ/CP-DO41 VO4: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A.
Srawlewicz, S. A.
January 1983.
326pp.
8302220443.
17235:001.
This report is a compilation of papers which were presented at the Tenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, October 12-15, 1982.
It consists of six volumes.
The papers describe recent results and planning of safety research work sponsored by the Office of Nuclear Regulatory Research, NRC.
It also includes a number of invited papers on water reacter safety research prepared by the Electric Power Research Institute and various government and industry organizations from Europe and Japan.
i l
NUREQ/CP-0041 VO5: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A.
Szawlewicz, S. A.
January 1983.
399pp.
8302170491.
17199:292.
This report is a compilation of papers which were presented at the Tenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Osithersburg, Maryland, October 12-15, 1982.
It consists of six volumes.
The papers describe recent results and planning of safety research work sponsored by the Office of Nuclear Regulatory Research, NRC.
It also includes a number of invited papers on water reactor safety research prepared by the Electric Power Research Institute and various government and industry organizations from Europe and Japan.
NUREQ/CP-OO41 VO6: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETINC. SZAWLEWICZ,S.A.
Srawlewicz, S. A.
I January 1983.
2 bop p.
6302170496.
17200:331.
This report is a compilation of papers which were presented at the Tenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, October 12-15, 1982.
It consists of six volumes.
The papers describe recent results and planning of safety research work sponsored by the Office of Nuclear Regulatory Research, NRC.
It also includes a number of invited papers on water reactor safety research prepared by the Electric Power Research Institute and various government and industry organizations from Europe and Japan.
I NUREQ/CR-0169 V21: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
C HEEVER, 0. C.
EG&O, Inc.
March 1983.
46pp.
8304060011.
EGG-2037.
17862:169.
22 i
A performance analysis of the Loss-of-Fluid Test (LDFT) modular drag-disc turbine transducer (MDTT) is presented.
Specific sources of measurement uncertainty are identified, quantified, and combined to provide an a-sessment of the ability of the MDTT to satisfy the requirement for measurement of single-and two-phase flow.
NUREQ/CR-1120 V10: SEISMIC SAFETY MARCING RESEARCH PRDORAM. Progress Report No. 14. BDHN, M. P. s BERNREUTER D.L.s CHUANG, T. V. s et al.
Lawrence Livermore Laboratory.
Januerg 1983.
42pp.
8302170469.
17197:185.
The Seismic Safety Margins Research Program (SSMRP) is a NRC-funded, multigear program conducted by Lawrence Livermore National Laboratory (LLNL).
Its goal is to develop a complete, fully coupled i
analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from a commercial nuclear power plant.
The analysis precedure is based on a state-of-the-art evaluation of the current seismic analysis and design process and explicitly includes the uncertainties inherent in such a process.
The results will be used to improve seismic licensing requirements for nuclear power plants.
This document is a progress. report on the Seismic Safety Margins Research Program covering the period April 1,
198L through June 30, 1982.
The report gives a general description of the program, together with financial summaries and individual project details.
Each project is summarized to show accomplishments, schedules, milestones and completion dates, budget and expenditures, and any concerns that may a f f ect the project.
NUREC/CR-1391: MAEROS USER MANUAL. CELBARD,F.
Sandia Laboratories.
February 1983.
56pp.
8303300527.
SAND 80-OS22.
17764:013.
This manual discusses the capabilities and implementation procedures for MAEROS, the stand-alone multicomponent aerosol module of CONTAIN.
The module calculates aerosol composition and mass concentration as a function of particle size and time.
The processes that may be incorporated are 1) coagulation due to Brownian motion, gravity and turbulences 2) particle deposition due to gravitational settling, diffusion and thermophoresiss 3) particle growth due to condensation of a gas, typically water vapors and 4) time varying sources of particles of different sizes and chemical compositions.
The capabilities of the code are illustrated through simulating an aerosol released in a hypothetical situation.
NUR EQ/CR-1894: MECHANICAL RELIABILITY EVALUATION OF A PROPOSED EMEROENCY RESPONSE RADIDIODINE AIR SAMPLER. KRAUPA,J.F.s BIRD,S.K.:
MOTES.B.C.
Idaho National Engineering Laboratory.
February 1983.
68pp.
8304050620.
ENICD-1075.
17846:296.
The purpose of this study was to evaluate the mechanical reliability of the air sampler component of the prototype system.
l Three air samplers previously used at Three Mile Island subsequent to the accident of March 1979 were tested.
During the tests, the repeatability and uniformity of the air sampler flowrates were monitored to determine the effects of the temperature, relative humidity, rainfall, dusty air, vibration, and mechanical shock.
Although all three air samplers eventually failed due to scoring and seizure of the motor shafts and/or bearings, prior to failure the three samplers exhibited uniform and reproducible flowrates at all l
test conditions except one.
The three air samplers would not operate j
23
(
l reliably on direct cu rent voltage at or below 0 F.
Based on the number of hours of operation in this study only, the minimum average lifetime of the three air sampler motors was determined as 44.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
In conjunction with the air sampler study, three additional type motors were investigated briefly: a dual voltage motor similar to the original air sampler motors and two single voltage motors.
l NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III I
CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE. SHARMA,S.s REICH,M.s CHANO, T. V. s et al.
Brookhaven National Laboratory.
March 1993.
69PP.
8304060017.
BNL-NUREG-51543.
17864:246.
An analysis of a Mark III reactor containment vessel sub Jected to a uniformly increasing internal pressure and gravity loads is carried out in order to ascertain the load carrying capacity of the structure l
under hydrogen burn.
The analysis is conducted by using a nonlinear i
(
finite element model that includes nonlinearities in the i
strain-displacement relations as well as in the material constitutive equations.
In this analysis, the nonlinear behavior of the liner and reinforcement steels is described by a von Mises elastic plastic model with isotropic hardening.
A recently developed elastic plastic-Fracture model that includes both the cracking and crushing limit states is used for the plain concrete.
Consistent i
smearing and de-smearing procedures are then used to represent the composite material properties of the reinforced concrete by an anisotropic and locally homogeneous continuum.
i l
NUREC/CR-2OOO VO1N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month j
Of December 1982.
- Dak Ridge National Laboratory.
January 1983.
112pp.
8302170033.
ORNL/NSIC-2OO.
17192:282.
This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file of l
the Nuclear Safety Information Center (NSIC) during the one month i
period identified on the cover of this document.
The LERs. from which this information is derived, are submitted to the Nuclear 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 l
Preparation of Data Entry Sheets for Licensee Event Reports.
The LER l
summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility.
Component, system, and keyword indexes follow the summaries.
The components and systems are those identified by the utility when the LER form is initiateds the keywords are assigned by the NSIC staff when the summaries are prepared for computer entry.
NUREO/CR-2OOO VO2 N1: LICENSEE EVENT REPORT (LER) CDMPILATION: For Month Of January 1983.
- Dak Ridge National Laboratory.
February 1983.
118pp.
8303100433.
ORNL/NSIC-2OO.
17496:305.
This 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 Nuclear Regulatory j
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 24 1
d summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility.
Component, system, and keyword indexes follow the summaries.
The components and systems are those identified by the utility when the LER form is initiateds the keywords are assigned by the NSIC staff when the summaries are prepared for computer entry.
1 NUREQ/CR-2OOO VO2 N2: LICENSEE EVENT REPORT (LER) COMPILATIDN: For Month j
Of February 1983.
- Dak Ridge National Laboratory.
March 1983.
t 70pp.
8303230408.
DRNL/NSIC-2OO.
17675:084.
This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file of f
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 Nuclear 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 NUREC-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 facility.
Component, system, i
and keywords indexes follow the summaries.
The components and systems are those identified by the utility when LER form is initiated: the keywords are assigned by the NSIC staff when the summaries are prepared for computer entry.
NUREC/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH PROGRAM, PHASE I FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
BERNREUTEP,D.L.s CHUNG,D.H.s MORTCAT.C.P.
Lawrence Livermore Laboratory.
January 1983.
142pp.
8302170213.
UCRL-53021.
17190:149.
Project II was charged with developing "a probabilistic statement of the seismic hazard" at the Zion site.
The definition of the seismic hazard included both the time histories upon which the SMACS computation was based and the hazard curve necessary for the calculation of unconditional release probabilities by SEISM.
In this volume, we discuss the application of the probabilistic approach using expert opinion to obtain estimates of the seismic hazard at the Zion site.
We also discuss and evaluate the ground motion models used to develop the seismic hazard at the the Zion site and the extensive sensitivity studies which were performed to determine the important parameters and the significance of uncertainty in them.
We also discuss the event specific approach which we developed to provide the correlated spectral information needed to generate the simulated time histories for the Zion site needed as input for the soil-structure interaction analysis.
From the many such spectra generated, we randomly selected 30 in each of six ranges of peak ground acceleration (0.15-0.3Og, 0.30-0.45g, 0.45.065g,.06.075g, 0.75.098g). Finally, time histories were generated for each response spectrum.
NUREQ/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S. Commercial Nuclear Power Plants, January 1,1972 Through September 30,1980. ATWOOD,C.L.
- EGEC, Inc.
February 1983.
206pp.
8303220394.
EOC-EA-5289.
17666:005.
25
l This report presents estimates of common cause fault rates and 1
related quantities, based on Licensee Event Reports for pumps in nuclear reactors.
The Licensee Event Report data base is described.
For estimating rates. the binomial failure rate model is used, extended to allow for the substantial observed plant-to plant variability, and for shocks that by their nature make all the pumps in 1
a system inoperable.
Every quantity is estimated by both a point estimate and a 90 percent interval.
All rates are expressed per hour.
NUREG/CR-23OO VO1: PRA PROCEDURES OUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants. HICMMAN,J.W.
American Nuclear Society.
Institute of Electrical & Electronic Engineers.
January 1983.
487pp.
B302100284.
17137:197.
This procedures guide describes methods for performing probabilistic risk assessments (PRAs) for nuclear power plants at three levels of scope:
(1) systems analysiss (2) systems and containment analysiss and (3) systems, containment, and consequence analysis.
After reviewing its objectives and limitations, this
~
document describes the organization and management of a PRA project and then presents procedures for accident-sequence definition and j
systems modeling, human-reliability analysis, the development of a I
data base, and the quantification of accident sequences.
Precedures for evaluating the physical processes of core meltdown are presented next, followed by guidance on the evaluation of radionuclide releases from the containment as well es the analysis of environmental transport and offsite consequences.
The analysis of external hazards is discussed next, including procedures for seismic, fire, and flood analyses.
The guide concludes with suggestions for the development and interpretation of results and the performance of uncertainty analyses.
NUREQ/CR-23OO VO2: PRA PROCEDURES OUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants. HICKMAN,J.W.
American Nuclear Society.
Institute of Electrical & Electronic Engineers.
January 1983.
447pp.
8302100281.
17138:324.
This procedures guide describes methods for performing probabilistic risk assessments (PRAs) for nuclear power plants at three levels of scope:
(1) systems analysiss (2) systems and containment analgsj ss and (3) systems, containment, and consequence analysis.
After reviewing its objectives and limitations, this l
document describes the organization and management of a PRA project l
and then presents procedures for accident-sequence definition and systems modeling, human-reliability analysis, the development of a data base, and the quantification of accident sequences.
Procedures for evaluating the physical processes of core meltdown are presented next, followed by guidance on the evaluation of radionuclide releases from the containment as well as the analysis of environmental transport and offsite consequences.
The analysis of external hazards is discussed next, including procedures for seismic, fire, and flood analyses.
The guide concludes with suggestions f or the development and interpretation of results and the performance of uncertainty analyses.
l l
NUREQ/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Guarterly Progress Report, April 1 -June 30,1982. BARI,R.A.s CERBONE,R.J.s GINSBERC. T. s et al.
Brookhaven National Laboratory.
March 1983.
155pp.
B304060057.
l 26
BNL-NUREC-51454.
17866:083.
This progress report describes 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.
2 The projects reported are the following: HTOR Safety Evaluation, SBC Development, Validation and Application, Generic Balance of Plant Modeling, Thermal-Hydraulic LWR nd LMFBR Safety Experiments, RDMONA-3B Code Modification and Evaluation, LWR Plant Analyzer Development, LWR Code Assessment and Applicatior.s Stress Corrosion Cracking of PWR Steam Generator Tubing, Standards for Materials Integrity in
- LWRs, Probability Based Load Combinations for Structural Design, Mechanical Piping Benchwork Problems, Soil Structure Interactions and Human Error l
Rate Data Analysis.
The previous reports have covered the period October 1, 1976 through March 31, 1982.
NUREQ/CR-2391: DNET SELF-TEACHING CURRICULUM. CRANWELL,R.M.s C AMPBELL, J. E. s STUCKWISCH, S. E. s et al.
Sandia Laboratories.
March 1983.
130pp.
8304200508.
This report contains a series of sample problems nd solutions for the Dynamic Network (DNET) model developed at Sandia National Laboratories for the Risk Methodology for Geologic Disposal of Radioactive Waste Project.
With this document and the DNET User's Manual (NUREC/CR-2343), the user may familiarize himself with the computer program, its capabilities and limitations.
When the user has completed this curriculum, he or she should be able to prepare data input for DNET and have some insights into interpretation of model output.
This report is one of a series of self-teaching curricula prepared under a technology transfer contract for the U.S.
Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards.
I NUR EQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No. 12. ROSAL, E. R. s HOCHREITER,L.E.s ANDREYCHEM,T.S.s et al.
Westinghouse Electric Corp.
February 1983.
413pp.
8303090659.
17481:252.
This report presents a descriptive plan of tests for the Systems Effects Task of the Full Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects (FLECHT SEASET) program.
This task is designed to produce experimental data which can be used to address issues related to natural circulation cooling modes in a pressurized water reactor (PWR).
The natural circulation tests were planned in direct response to the accident at Three Mile Island.
The tests consist of natural circulation and reflux condensation cooling esperiments using electric heating rods to simulate current nuclear core arrays of PWR and PWR fuel vendors.
The FLECHT BEASET systems effects test facility with a scale factor of 1/307 with respect to a l
four-loop 3411 MWt PWR was used.
The facility was designed with all elevations identical to those of a PWR.
Two full-height steam g enerators with active secondary side heat removal are also part of the system design.
All tests were conducted with a consine axial l
power profile.
The data obtained from these tests will be used to l
evaluate the effects of components and systems parameters during natural circulation cooling modes.
l 27 l
1 NUREQ/CR-2422: DOSIMETRY AND HEALTH EFFECTS SELF-TEACHING CURRICULUM.
Illustrative Problems to Supplement The User's Manual For The Dosimetry And Health Effects Computer Code. RUNKLE,0.E.s FINLEY,N.C.
Sandia Laboratories.
March 1983.
115pp.
8304220289.
SAND 81-2488.
18138:001.
This report contains a series of sample problems and solutions for the Dosimetry and Health Effects (DHEJLS1) model developed at Sandia National Laboratories for the Risk Methodology for Geologic Disposal of Radioactive Waste Project.
With this document and the DHEJL81 User's Manual (NUREO/CR-2346), the user mag familiarize himself with the computer program, its capabilities and limitations.
When the user has completed this curriculum, he or she should be able to prepare data input for DHEJL81 and have some insights into interpretation of model output.
This report is one of a series of self-teaching curricula prepared under a technology transfer contract for the U.S.
Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards.
NURE9/CR-2443: A MANUAL FOR UGING ENERGY ANALYSIS FOR PLANT SITING.
ODUM,H.T.s LAVINE. M. J. s WANG.R.C.s et al.
Florida, Univ. of, Gainesville.
February 1983.
242pp.
8303090520.
17476:161.
This is an instructional manual for choosing among possible sites for power plants by selecting the one with the least diversion of 1
resources of the environment and of the economy.
For each alternative site, changes of embodied energy in flows and storages are estimated in solar equivalent Calories.
Then a dollar equivalent is estimated from the ratio of total solar equivalent flows to gross national product.
Sample calculations are provided for LaSalle power plant west of Chicago considering alternatives of cooling reservoirs, cooling from a natural water body, and cooling towers.
In order to facilitate calculation, an appendix provides procedures and data for evaluating embodied energy of the environment.
NUREQ/CR-2482 VO3: REVIEW OF DDE WASTE PACKAGE PRDORAM. Subtask
- 1. 1 -
National Waste Package Program. April 1982 - September 1982. SDO, P.
Brookhaven National Laboratorg.
March 1983.
289pp.
8304130108.
BNL-NUREG-81494.
17958:001.
This is part of an ongoing task to evaluate the national high-level waste package program.
The contributions of reference waste package components, i. e., waste form, container and packaging material, to containment and controlled release of radionuclides in basalt and salt repositories are evaluated.
Chemical and mechanical failure / degradation modes for the waste package are reviewed.
Data required to demonstrate that the waste package will meet the performance objectives of 10 CFR 60 are identified.
NUREQ/CR-2524: EVALUATION OF PERSONNEL NEUTRON DOSIMETRY AT (PERATING NUCLEAR POWER PLANTS. RYAN,R.M.
Rensselaer Polytechnic Inst.
March i
1983.
83pp.
8304200606 18080:265.
l The basic objective of this research program titled, " Evaluation of Personnel Neutron Dosimetry at Operating Nuclear Power Plants,"
sponsored by the United States Nuclear Regulatory Commission was to i
evaluate neutron personnel monitoring devices and/or methods used at l
nuclear power plants.
To accomplish this research, measurements were made in areas where personnel are likely to have access during reactor i
i operation.
Measurements were made of spectral and flux distribution of neutrons.
Dose equivalent surveys were made of neutrons and gammas 2s
in operating areas with portable survey meters.
Gamma to neutron ratios were obtained.
A new computer code, BONABS was developed to incorporate the response to portable neutron survey meters with an j
existing code BONNER used for neutron spectrum unfolding techniques.
NUREQ/CR-2527: LOCA SIMULATION IN NRU PRDORAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3). MOHR,C.L.s HESSON,0.E.s MING.L.L.s et a l.
Battelle Memorial Institute, Pacific Northwest Laboratory.
l Mtrch 1983.
120pp.
8304120660.
PNL-4165.
17942:210.
A series of in-reactor experiments are being conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory (PNL) Loss-of-Coolant Accident (LOCA) Simulation Program.
The fifth experiment in the series of thermal-hydraulic nd materials deformation experiments (TH-3) is described in this report.
The experiments are being conducted in the National Research Universal (N9U) reactor, Chalk River, Ontario, Canada.
The objective of TH-3 was to further refine the feedback control parameters developed in the TH-2 experiment and to re-establish the operability of the loop prior to the subsequent materials deformation and repture test (MT-3).
The TH-3 and MT-3 experiments were planned for the same reactor window and were run within two days of each other.
The TH-3 test results insured the success of MT-3 nd provided the opportunity to demonstrate the reactor control improvements and to evaluate a new desuperheater concept that would allow the test to run for extended times at high temperatures.
The control system improvements and the addition of the new desuperheater resulted in fuel cladding tersperatures above 1033K (1400 degrees farenheit) for 340 s.
Experimental data and initial results are presented in this report.
NUREQ/CR-2530: REVIEW OF THE GRAND QULF HYDROGEN IONITER SYSTEM.
CUMMINGS, J. C. s CAMP,A.L.s SHERMAN,M.P.s et al.
Sandia Laboratories.
March 1983.
225pp.
8304220626.
18136:071.
The Mississippi Power and Light Company has proposed installation of a Hydrogen Igniter Systems (HIS) at the Grand Gulf Nuclear Station l
(BWR Mark III) to burn hydrogen generated during accidents more severe than the design-basis accidents.
Sandia National Laboratories, under a contract with the U.S.
Nuclear Regulatory Commission, has performed a technical evaluation of the adequacy of the proposed HIS to meet the threat posed by hydrogen combustion.
Areas considered in this review include HIS design and testind location and distribution of igniters, containment pressure and temperature response calculations, detonations, containment atmosphere mixing mechanisms, actuation criteria for the HIS, and the spectrum of hydrogen generating accidents.
NUREQ/CR-2531 RO1: INTRODUCTORY UtsERS MANUAL FOR THE US NUCLEAR REQULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
SCDFIELD,N.R.s HARAY,H.A.s LAATS E.T.
EQ&Q, Inc.
March 1983.
94pp.
8304060050.
EQQ-2164.
17863:153.
The United States Nuclear Regulatory Commission (NRC) has established the NRC/ Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs.
The NRC/DAE Data Bank Program provides a central computer stora5e mechanism and access software for data that is to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry.
The administrative portion of the 29
program provides data entry, documentation, training, and advisory services to users and the NRC.
The NRC/DAE Data Bank an the capabilities of the data access software are described in this document.
NUREQ/CR-2659: IDDINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM GENERATOR TUBES. POSTMA A.K.3 HESSON, 0. M. s FALETTI,D.H.
Battelle Memorial Institute, Pacific Northwest Laboratory.
February 1983.
36pp.
8303140696.
PNL-3794.
17582:197.
Iodine transport in the primary and secondary cooling systems of 4
a pressurized water reactor following a postulated steam line break with concurrent rupture of steam generator tubes was analgred.
The l
goal of the study was a conservative estimate of the quantity of i
iodine which could escape from the braak to the atmosphere.
Ionic iodine, elemental icoine, and methyl iodine were studied.
Ionic l
iodine was the dominant species and its release was governed by the carrgover of atomized primary water.
The number of ruptured tubes was i
treated as a parameter.
While both an instantaneous and a linear release of iodine from fuel pins to the primary coolant were included in the models, it was found that the instantaneous release dominated l
because the blowdown transient (IRT Code calculations) lasted Just a l
few minutes.
The fraction of an instantaneous iodine source term l
(total of the three forms) which could escape from the steam line l
break was predicted to be 3.7% for one rupture tube and 12.5% for five ruptured tubes.
Thus, most of the iodine released from the fuel as a result of the postulated accident will be retained within the primary and secondary coolant systems.
NUREC/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
AHMAD A. s CATTON.I.s CUESTA-00NZALEZs et al.
California, Univ. o f, Los Angeles.
January 1983.
276pp.
8302170198.
UCLA-ENG-8284.
17187:137.
This is the final report for a study of various accident mitigation schemes for Light Water Reactor (LWR) containments.
This work is part of the NRC's program assessing degraded core and core-melt accidents beyond the design basis.
Included are studies aimed at estimating the risk reduction potential for filtered-vented containment systems, passive containment heat removal, and features to mitigate against hydrogen burns and basemat penetration.
In addition, specific aspects of mitigation for the Zion. Indian Point and Limerick plants were considered.
NUREQ/CR-2668: JOB ANALYSIS OF THE MAINTENANCE SUPERVISOR AND INSTRUMENT AND CONTROL SUPERVISOR POSITIONS FOR THE NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL. BARTTER,W.D.s SIEGEL,A.I.: FEDERMAN,P.J.
Oak Ridge National Laboratory.
February 1983.
127pp.
8303300651.
ORNL/TM-8299.
17768:153.
This report is one of a series that is planned to describe the results of a program undertaken by the Oak Ridge National Laboratory (ORNL) for the U.S.
Nuclear Regulatory Research, to define, develop, validate, and disseminate a methodology for the quantitative prediction of human reliability in the conduct of maintenance tasks in nuclear power plants (NPPs).
DRNL has subcontracted portions of this effort to Applied Psychological Services, Inc.
A program scoping / feasibility study has been completed which consisted of four tasks:
1.
Completion of a structured user survey to assess the likelg l
30
l utility of the proposed methodology, to identify user output requirements, and to identify critical input data needs.
2.
Assessment of available methodologies via completion of a literature survey.
3.
Completion of the initial job analysis of key maintenance positions.
4.
Definition of a comprehensive program plan for development, validation, and dissemination of the proposed i
methodology.
NUREQ/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REQULATION OF l
NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites. MCKENZIE,D.H.s CADWELL,L.L.; EBERHARDT.L.E.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
January 1983.
34pp.
8302250053.
PNL-4241.
17300:250.
The purpose of the work reported here was to estimate the l
potential dose to man resulting from biotic transport mechanisms at a r ef erenc e. humid low-level waste site.
The site description includes waste inventories, site characteristics, and biological communities.
Parameter values for biotic transport are based on data reported in current literature.
Calculations for radionuclide decay and waste container decomposition are made to estimate the amount of radioactive material available for biotic transport and exposure scenarios during 500 years following site closure.
Five-hundred year dose to man estimates based on biotic transport are estimated to be of the same order of magnitude as dose estimates resulting from the more common 1g evaluated intrusion-agricultural scenario reported in NRC's DEIS for 10 CFR 61.
These results indicate that biotic transport has the potential to influence low-level waste site performance.
The eported lack of potential importance of biotic transport at low-level waste sites in earlier assessments studies is not confirmed by the findings presented in this report.
Through biotic transport, radionuclides may be moved to locations where they can enter exposure pathways to man.
NUREQ/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE i
EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
GANAPATHY,S.s SCHMULT,B.4 WU,W.S.s et.al.
Michigan, Univ. o f.
January 1983.
87pp.
8303020266.
17355:150.
Progress in the development of a special purpose system for use in a real-time in-service inspection system for reactor vessels and piping components is described in this report.
An analysis of the synthetic aperture processing algorithm is presented and new methods of speedup are described.
A number of special purpose processor architectures are presented and two of the more promising ones are l
described in detail and are compared and evaluated.
Proposed l
specifications for an initial field inspection system are presented.
A brief description of the capabilities of a laboratory prototype processor (to be fabricated) is given.
i l
NUREQ/CR-2716 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report. July -September 1982. EDLER,S.K.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
84pp.
8304010668.
P NL-427 5-3, 17824:005.
This document summarizes work performed by Pacific Northwest Laboratory (PNL) from July 1 through September 30, 1982, for the l
Division of Accident Evaluation and the Division of Engineering l
Technology, U. S.
Nuclear Ragulatory Commission (NRC).
Evaluations of I
31
.. _ = _
nondestructive examination (NDE) techniques and instrumentation are reporteds areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analysing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated.
Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to pipe impacts following postulated breaks in high-energy fluid system piping.
Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions.
Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canadas fuel rod deformation, severe fuel damage, and i
postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italys the instrumented fuel assembly irradiation program at Halden, Norways and experimental programs at the Power Burst Facility. Idaho National Engineering Laboratory, Idaho Falls, Idaho.
These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating l
conditions.
l NUREQ/CR-2728: INTERIM RELIABILITY EVALUATION PRDORAM PROCEDURES OUIDE.
CARLSON,D.D.s CALLUP,D.R.s KOLACZK0WSKI,A.
et al.
Sandia Laboratories.
March 1983.
151pp.
8304060072.
SANDB2-1100.
17866:238.
l This document presents procedures for conducting analyses of a scope similar to those performed in Phase II of the Interim Reliability Evaluation Program (IREP).
It documents the current state of the art in performing the plant systems analysis portion of a probabilistic risk assessment.
Insights gained into managing such an analysis are discussed.
Step-by-step procedures and methodological guidance constitute the major portion of the document.
While not to be viewed as a " cookbook," the procedures set forth the principal steps in performing an IREP analysis.
Guidance for resolving the problems encountered in previous analyses is offered.
Numerous esamples and representative products from previous analyses clarify the discussion.
l NUREQ/CR-2729: USER'S GUIDE TO BFR.A Computer Code Based Dn The l
Binomial Failure Rate Common cause Model. ATWOOD,C.L.s SUITT,W.J.
EG&G, Inc.
February 1983.
30pp.
8303230281.
EOC-EA-5502.
17686:015.
The computer code BFR finds estimates of common cause failure rates and related quantities.
The estimates are based on the binomial failure rate model.
This user's guide describes the model and tells how to prepare input for BFR.
The output of BFR is described.
Twc esamples are given.
i NUREG/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
l TOPMILLER. D. A. s ECKEL.J.S.s K0ZINSKY,E.J.; et al.
General Physics Corp.
March 1983.
221pp.
8304060027.
SAND 82-7057/1.
17867:155.
This report describes a survey and comparative analysis of previous and current attempts to quantify and predict human operator 32
=
and maintainer performance as a function of design, training, precedural, or situational factors.
An assessment was made of these methods and techniques as to their potential applicability to PRA and as a supplement to the data and procedures in NUREQ/CR-1278.
Five previously established human reliability data banks were reviewed along with five current systems, all of which include estimates of j
l human error related events.
The data banks were evaluated against a set of criteria intended to serve as guidelines far an idealized human reliability data reporting, storage, and retrieval system.
It was concluded that insufficient data currently exist in these systems to adequately support nuclear PRA activities, and it was therefore recommended that a human reliability data bank specific to nuclear power plant PRA applications be developed.
Volume 2 of this report contains the results of the date bank concept development task.
NUREC/CR-2751 VO2: HEAVY SECTION STEEL TECHNOLDQY PROGRAM GUARTERLY PROGRESS REPDRT FOR APRIL-JUNE 1982. WHITMAN,0.D.s BRYAN,R.H.
Oak Ridge National Laboratory.
January 1983.
150pp.
8303300638.
ORNL/TM-8369/V2.
17766:167.
l The Heavy-Section Steel Technology Program comprises studies l
related to all areas of the technology of materials fabricated into thick-section primary-coolant containment systems of I
l light-water-cooled nuclear power reactors.
The investigation focuses on the behavior and structural integrity of steel pressure vessels containing crack-like flaws.
Current work is organized into six t
tasks:
(1) program administration and procurement, (2) fracture-mechanics analyses and investigations, (5) pressure vessel investigations, and (6) stainless steel cladding investigations.
The three-dimensional finite-element program for elasticplastic fracture analysis was adapted to the analysis of stainless-steel-clad structures under combined pressure and thermal shock loads.
Subcontractors continued their support of thermal-shock experiments and investigation of fracture mode transitions and crack-arrest testing.
Statistical analyses of irradiated Charpy impact specimens I
were completed, and plans were inade for two new irradiation studies, high-copper welds and stainless steel cladding.
Further overcooling accident analyses were made, and preliminary studies of the next thermal-shock experiments were conducted.
l NUREQ/CR-2751 VO3: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JULY-SEPTEMBER 1982. WHITMAN,G.D.s BRYAN,R.H.
Oak Ridge National Laboratory.
February 1983.
119pp.
8303290673.
ORNL/TM-8369/V3.
17738:246.
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 comprises studies related to all areas of the technology of materials I
fabricated into thick-section primary-coolant containment systems of light-water-cooled nuclear power reactors.
The investigation focuses on the behavior and structural integrity of steel pressure vessels i
containing crack-like flaws.
Current work is organized into six tasks:
(1) program administration and procurement, (2) fracture-mechanics analyses and investigations. (3) investigations of irradiated materials, (4) thermal-shock investigations. (5) pressure vessel investigations, and (6) stainless steel cladding investigations.
33
i NUREQ/CR-2755: PACKING MATERIAL TESTING REQUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests. BIDA.C.s EASTWOOD.D.
Brookhaven National Laboratory.
January 1983.
38pp.
8302250129.
B NL-NUR EG-51544.
17299:246.
As part of NRC's development of an understanding of phenomena related to' waste package containment, this report addresses whether discreet backfill alone can contain radionuclides for 1000 years.
To do so, the backfill would have to control the flow of groundwater to the container and, after failure of the container, retard migration of the radionuclides.
The report identifies properties of backfill i
associated with control of groundwater flow and retardation of radionuclides that could be measured to demonstrate 1000 year containment. -Methods of performing the measurements are also discussed, as are two candidate backfill materials, bentonite and
~
roolite.
Two general conclusions are that (1) because backfill materials have not been tested under the range of conditions expected in a repository. the data base is inadequate to demonstrate 1000 containment and (2) standardization of test methodology (e.g.,
a Materials Characterization Center - type approach) seems essential.
More specific conclusions are also discussed.
NUREQ/CR-2770: CDMMON CAUSE FAULT RATES FOR VALVES: Estimates Based On Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, 1976-1980. ATWOOD,C.L.s STEVERSON J.A.
EQ&G, Inc.
February 1983.
163pp.
8303220400.
ECO-EA-5485.
17665:282.
This report presents estimates of common cause fault rates and related quantities, based on Licensee Event Reports for valves in nuclear reactors.
The Licensee Event Report data base is described.
For estimating rates, the binomial failure rate model is used, extended to allow for the substantial observed plant-to-plant variability, and for shocks that by their nature make all the valves in a system inoperable.
Every quantity is estimated by both a point l
estimate and a 90 percent interval.
l NUREQ/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants, 1976-1978. ATWOOD,C.L.
EC&G, Inc.
February 1983.
107pp.
8303220409.
ECG-EA-5623.
17660:225.
This report presents estimates of common cause fault rates and related quantities, based on Licensee Event Reports of instrumentation and control assemblies in nuclear reactors.
The Licensee Event Report data base is briefly described, and imperfections in the data are discussed.
The components are grouped into assemblies, for which rates are estimated.
For estim4 ting rates, the binomial failure rate I
model is used, extended to allow for the substantial observed p lant-to plant variability, and for shocks that by their nature cause all the assemblies in a system to fail.
Every quantity is estimated i
by both a point estimate and a 90 percent interval.
All rates are expressed per calendar hour.
NUREQ/CR-2774 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July -
September 1982.
- Argonne National Laboratory.
February 1983.
23pp.
8304060010.
ANL-82-24 VOL.3.
17863:250.
This Guarterly progress report summarizes work done during the months of July-September 1982 in Argonr,e National Laboratory's Applied Physics and Components Technology Divisions for the Division of 34 i
Reactor Safety Research of the U.S.
Nuclear Regulatory Commission.
The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section.
Work on reactor core thermal-hydraulic is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.
An executive summary is provided including a statement of the findings and recommendations of the report.
NUREQ/CR-2780: EVALUATION OF SYSTEM REGUIREMENTS AND STANDARDS DEVELOPMENT FOR THERMAL ANNEALING OF REACTOR PRESSURE VESSELS.
i SERVER,W.L.
EC&O, Inc.
March 1983.
50pp.
8304200581.
EQQ-FM-6174.
18081:254.
The material property data on thermal annealing of reactor pressure vessels have been reviewed.
The most critical materials are high copper welds the data indicate that close to full recovery of Charpy-V notch properties can be realized by annealing at 850 F for 1 week (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />).
However, the variability and sparcity of annealing recovery data dicatate appropriate surveillance and experimental test programs.
Of particular concern are the actual fracture toughness changes and the differences between test and power reactor conditions.
NUREG/CR-2800: GUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT. ANDREWS,W.B.s GALLUCCI,R.H.V.s HEABERLIN,S.W.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
February 1983.
274pp.
8303150600.
P NL-429.
17583:062.
Pacific Northwest Laboratory has developed a methodology, with examples, to calculate the risk, dose and cost impacts of implementing i
resolutions to reactor safety issues.
This report is an applications guide to issue-specific calculations.
A description of the approach, i
mathematical models, work sheets and step-by-step examples are j
provided.
Analyses using this method are intended to provide comparable results for many issues at a cost of two staff-weeks per issue.
Results will be used by the NRC to support decisions related to issue priorities in allocation of resources to complete safety issue resolutions.
NUR EG/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REQULATION DEVELOPMENT. WOO,H.H.s CHOU,C.K.
Lawrence Livermore Laboratory.
January 1983.
63pp.
8302170208.
UCRL-53040.
17188:096.
This report has two purposes: (1) to present the validation l
results for the piping reliability model previously developed at the Lawrence Livermore National Laboratory (LLNL) and (2) to evaluate the potential use of the reliability approach for licensing regulation development.
It includes five major parts.
The first part reviews the piping reliability model developed during fiscal years 1980 and 1981.
Two failure modes-fatigue failure and stress corrosion cracking failure j
resulting from vibratory, seismic, assembig, and operating stress are l
considered in the piping model.
This piping reliability model is developed on the basis of probabilistic fracture mechanien.
The second part presents the validation results for the piping l
reliability model.
The failure case chosen for comparison with the l
analytical result is pressurized water reactor (PWR) feedwater line 36
cracking incidents.
The estimated leak probabilities correlate reasonably well with the failures observed at certain nuclear power plants.
NUR EC/CR-2804: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSING IMPACTS. Final Report. MCKENZIE D.H.s SIMMONS. M. A. s SKALSKI,J.R.
Battelle Memorial Institute, Pacific Northwest Laboratory.
January 1983.
58pp.
8302250146.
17299:073.
Monitoring methods used in fisheries management assessments were examined and their potential applicability in confirmatory impact monitoring were evaluated using case studies from selected nuclear p ower plants.
A report on Task I of the project examined the application of Catch-Per-Unit-Effort (CPUE) techniques in monitoring programs at riverine, large lake and ocean sites.
Included in this final report is an examination of CPUE data for the Oconee Nuclear Plant on Lake Keowee, a reservoir site.
This report also presents a summary of results obtained over the life of the project and guidelines for designing and implementing data collection programs and for data analysis and interpretation. Analysis of monitoring data from Lake Keowee confirmed findin2s from previous analyses of surveys at nuclear power plants on large lakes, rivers and coastal sites.
CPUE techniques as applied to these monitoring programs do not provide data necessary to separate changes induced by plant operation from naturally occurring changes, i
NUREC/CR-2805 VO1: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report. January 1982 - March 1982. GUTHRIE,0.L.s MCELROY,W.N.
Hanford Engineering Development Laboratory.
January 1983.
130pp.
8302250135.
HEDL-TME 82-18.
17301:186.
The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) has been established by NRC to improve, test, verify, and standardize the physics-dosimetry-metallurgy, damage correlation, and the associated reactor analysis methods, procedures and data that are used to predict the integrated effect of neutron exposure to LWR pressure vessels and their support structures.
A vigorous research effort attacking the same measurement and analysis problems exists worldwide, and strong cooperative links between the US NRCJsupported activities at HEDL, ORNL NBS, and MEA-ENSA and those supported by CEN/SCK (Mol, Belgium),
EPRI (Palo Alto. USA), KFA (Julich, Germany), and several UK laboratories have been extended to a number of otner countries and laboratories.
These cooperative links are strengthened by the active membership of the scientific staff from many participating countries and laboratories in the ASTM E10 Committee on Nuclear Technology and j
Applications.
Several subcommittees of ASTM E10 are responsible for l
the preparation of LWR surveillance standards.
NUREQ/CR-2805 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PRDORAM. Guarterly Progress Report. April 1982 - June 1982.
QUTHRIE,0.L.s MCELROY,W.N.
Hanford Engineering De'velopment Laboratory.
January 1983.
50pp.
8302220435.
HEDL-TME B2-19.
17237:177.
The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvements Program (LWR-PV-SDIP) has been established by NRC to improve, test, verify, and standardize the physics-dosimetry-metallurgy, damage correlation, and the associated 36 l
reactor analysis methods, procedures and data that are used to predict the integrated effect of neutron exposure to LWR pressure vessels and their support structures.
A vigorous research effort attacking the same measurement and analysis problems exists worldwide, and strong cooperative links between the US NRC-supported activities at HEDL, ORNL, NBS, and MEA-ENSA and those supported by CEN/SCK (Mol, Belgium),
EPRI (Palo Alto, USA), KFA (Julich, Germany), and several UK laboratories have been extended to a number of other countries and laboratories.
These cooperative links are strengthened by the active membership of the scientific staff from many participating countries and laboratories in the ASTM E10 Committee on Nuclear Technology and Applications.
Several subcommittees of ASTM E10 are responsible for l
the preparation of LWR surveillance standards.
I NUREC/CR-2005 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982). MCELROY,W.N.
Hanford Engineering Development Laboratory.
KAM F.B.K.
Oak Ridge National Laboratory.
GRUNDL.J.A.: et al.
Commerce, Dept. o f, National Bureau of Standards.
January 1983.
173pp.
8302280020.
HEDL-TME 82-20.
17313:323.
The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) has been established by NRC to improve, test, verify, and standardize the physics-dosimetry-metallurgy, dam' age correlation, and the associated reactor analysis methods, procedures and data that are used to predict the integrated effect of neutron exposure to LWR pressure vessels and their support structures.
A vigorous research effort attacking the same measurement and analysis problems exists worldwide, and strong cooperative links between the US NRC-supported activities at HEDL, ORNL, NBS, and. MEA-ENSA and those supported by CEN/SCK (Mol, Belgium),
i EPRI (Palo Alto, USA), KFA (Julich, Germany), and several UK laboratories have been extended to a number of other countries and laboratories.
These cooperative links are strengthened by the active membership of the scientific staff from many participating countries and laboratories in the ASTM E10 Committee on Nuclear Technology and Applications.
Several subcommittees of ASTM E10 are responsible for the preparation of LWR surveillance standards.
NUR EC/CR-2006: A KINETIC NODEL FOR THE CHLORINATION OF POWER PLANT COOLING WATERS. JOHNSON, J. D. s GUALLS,R.C.
North Carolina, Univ. of January 1983.
ESpp.
8302100273.
17135:194.
In this study, we developed kinetic expressions for the short term reactions of chlorine consumption by organic substances in natural freshwater.
These expressions were developed to use in a kinetic model to predict the free and total available chlorine l
discharged in cooling water.
This model uses common 1g available water quality data.
It assumes that most of the chlorine consuming substances are:
(1) NH (3), (2) chloramine-forming organic-N, and (3) humic substances.
It uses the Morris-Wei model of chlorine-ammonia reactions.
Chloramine formation from organic-N was represented by a model compound, gigcylglycine.
We estimated that about 10% of the dissolved organic-N formed chloramines in our river water samples.
Isolated fulvic acid was used to model the kinetics of chlorine c onsump tion by humic substances.
The humic reactions were adequately described by the sum of two second order reactions.
Concentrations of fulvic acid reaction sites were related to fulvic acid carbon
(
concentration.
Rates of chlorine consumption by fulvic acids from 3 l
diverse water sources were guite similars thus, our kinetic model mag 37
l I
be generally applicable.
Simulations of chlorine consumption in several river water samples matched the measured data reasonably well.
l Using conditions typical of once-through cnlorination, the most i
important reactions consuming free residual chlorine were NH (2) C1 i
formation, and consumption by humic substances in those samples.
Most total residual chlorine reduction was caused by humic substances.
NOREQ/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982. ADAMS,R.E.s TOBIAS,M.L.
i Dak Ridge National Laboratory.
March 1993.
41pp.
8304060005.
ORNL/TM-B397/V3.
17862:069.
This report summarizes progress for the Aerosol Relsase and Transport Program for the period July-September 1992.
Topics discussed include (1) the source-term experimental program in the Fuel l
Aerosol Simulant Facilitys (2) NSPP experiment 601 involving the generation of mixed concrete-iron oxide aerosols in steams (3) small vessel tests of plasma torch techniques for concrete aerosolsa (4)
{
technical support work for the Marviken and DEMONA test programs: (5) core-melt experiments involving cesium vaporization and manganese I
r elease s (6) summaries of calculational results obtained in the ABCOVE e m ercises (7) hot-film anemometer experiments on fan-miner effects in the NSPP.
i NUR EQ/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
BAUMANN,W.L.s DOMANUS,H.M.s MOHR,D.s et al.
Argonne National i
Laboratory.
January 1983.
87pp.
8303140617.
ANL-B2-66.
17582:257.
The in-vessel thermal-hydraulic analysis of the EBR-II Pool Reactor for Transient Test No. 10. Phase 2, has been performed using the COMMIX-1A computer code.
The analysis includes all reactor components inside the reactor vessel.
COMMIX-1A employs the i
porous-media formulation in which the concepts of volume pccosity, surface permeability, and distributed resistance and heat source are used to model the internal structures.
The governing equations of conservation of mass, momentum, and energy are solved as a boundary problem in space and as an initial-value problem in time.
This report presents the steady-state and transient in-vessel thermal-hydraulic results of the EBR-II natural-circulation simulation.
Comparisons show close agreement between computational and experimental data.
The phenomenon of reversed flow in the low-pressure plenum, which was observed during the EBR-II transient test, is confirmed by the l
simulation.
[
l NURE0/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PRDORAM. Guarterly Progress Report For Period Ending June 30,1982.
DODD,C.V.4 DEEDS,W.E.s MCCLUNG,R.W.
Dak Ridge National Laboratory.
February 1983.
10pp.
8303140622.
DRNL/TM-8418/V2.
17548:313.
Computer-based multifrequency, multiproperty eddy-current I
techniques and equipment are being developed to reduce ambiguities during in-service inspection of steam generator tubing.
Recent I
calculations show that an array of small pancake coils pressed against the inner wall of the tubing can e tect and locate small flaws on the outer wall of the tubing with much greater accuracy and reliability than can the usual large circumferential coils.
We are continuing to construct, test, and develop such pancake coils and arrays, as well as the instrumentation to go with them.
38 i
~-=-
l NUPEQ/CR-2824 VO3: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending September 30, 1982. DODD. C. V. s DEEDS W.E.s MCCLUNG.R.W.
Oak Ridge National Laboratory.
February 1983.
9pp.
8303290710.
DRNL/TM-8418/V3.
17742:234.
This program was established to develop improved eddy-current techniques and equipment for the in-service inspectiJn of steam generator tubing.
The purpose is to separate the effects of variables such as denting, probe wobble, tubesheets, tube supports, and conductivity variations from defect sire, depth.'and wall thickness variations.
The program consists of design calculations based on theoretical models, construction of optimum equipment, laboratory tests of the best design, and field tests of the eg'o i p men t.
Previous 1g reported computer calculations have shown that verd small, flat " pancake" coils pressed against the inside wall of the tubing are an orde? of magnitude more sensitive to small flaws in the tubing and much less sensitive to outside influences, such as tube supports, than are the conventional large circumferential coils, which must~be made somewhat smaller than the inside diameter of the tubing to pass dents or deposits.
NUREQ/CR-2843 V01: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report, January - March 1992. THOMPSON.S.L. ;Sandia Laboratories.
March 1983.
63pp.
8303300514.
SANDB2-1788.
17749:037.
This guarterly report (1) includes further results.of several RALOC analysen which were performed both for assessment purposet and for use in the Crand Gulf hydrogen igniter study.
The second RALOC calculation which was performed this guarter using the Grand Gulf nadalization investigated the effect of having a pure hydro,nen source instead of the 2 parts hydrogen, 1 part air mixture that was assumed for the source in all the previous Grand Gulf' analyses.
For the same quantity of hydrogen injected, the pure hydrogen injection source produced higher end-of-calculation hydrogen concentrations and slight 1g more asymmetric concentration gradients during the injection period.
The results folloued the general trends seen,in previous RALOC calculations.
This quarter, brief current status reports on the LDFT, PKL, LOBI and Semiscale Mod-3 analyses are given.
Most steady state calculations have been completed and n number of I
transients begun.
Only a few transients were actually completed this guarter.
Results are discussed For several LOFT transient calculations completed for the L6-7/L9-2 test, and steady calculations completed for the L9-1/L3-3, L3-6/L8-1, and L5-1/LG-2 tests.
NUREC/CR-2847: COGAP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL i
SYSTEM EVALUATION CODE. GIDO,R.J. 'Los Alamos Scientific Laboratory.
l January 1983.
31pp.
8302100286.
LA-9459-MS.
17135:346.
i The accounting of containment gas concentration following s l
loss-of-coolant accident is important in the safety evaluation of hydrogen combustible gas control systems for nuclear power plants.
The COGAP code provides such accounting including the offects of (1) l the reaction of airconium and waters (2) radiolysis of core and sump waters (3) corrosion of rine, aluminum and coppers (4) recirculation between compartmentss (5) hydrogen recombinerss (6) purgings (7) nitrogen additions and (8) atmospheric steam.
Controls are available to determine when options are initia teds for' example, the hydrogen recombiner can be started when the hydrogen concentration reaches a user-specified value.
39
_r_
l
.~
NUREG/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
l G I DO, R. J. s LANKIN,D.'i KDESTEL,A.
Los Alamos Scientific Laboratory, i
January'1993.
.53pp.
.3302150718.
LA-9460-MS.
17160:117.
i Procedures for performing mechanistic dry-pressure-containment LOCA analyses are presented, evaluated.' applied and coepared with other approaches.
The procedures are based on (1) the blowdown-introduced small drops (10 to-100 um) being' homogeneously mixed into the atmosphere, (2) drop (particla) turbulent deposition on i
vertical surfaces, and (3) terminal velec'ity gravity deposition on the floor.
Variation of drop sire'and mass transfer deposition velocity was found to have a small,_effect on calculated results, except for the i
atmosphere water mass retention.
The primary effect of the mechanistic approach was a saturated containment atmosphere, with significant atmosphere water retention.
The calculated containment pressure of'the mschanistic approach was lower, before the spray initiation, than that calculated by otteer current procedures.
l l
[,
NUREG/CR-2953: NON-CONDENSIBLE GAS FRACY1DN PREDICTIONS AT ELEVATED I
TEMPERATURES AND PRESSURE USING WET AND DRY BULB TEMPERATURE l
, MEASUREMENTS. BOWMAN.J.K s GRIFFITH,P.
Massachusetts Institute of Technology.
March 1993.
162pp.
830427C221.
18240:108.
A technique is presented whereby non-condensible gas mass fractions in a closed system can be determined using wet bulb and dry
(
b ulb temperature and ' system pressure measurements.
This technique l
would have application in situations where sampling techniques could not be used..
Using an energy balance about the wet bulb wick, an e xpression 'is obtained which relates the vapor concentration difference between the wet bulb ard a heat to mass transfer
~
c oef fic ient ratio.
This coefficient ratio was examined for Forced and l
natural convection flows.
This analysis was verified with forced and I
natural convection tests over-the range of pressure and temperature from 50-557 psig and 415-576 F.
All the data could best be fit by the natural convection analysis.
This is useful when no information about the field flow is known.
1 NUREQ/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP OEOLOGIC REPOSITORIES. 00 NANO,L.4 FINDLEY,D.s WILDANGER,W.s et al.
Golder Associates.
March 1993.
185pp.
'9303250113.
813-1164D.
1771,0:239.
The purpose of the complete project is to provide the NRC with J
technical assistance to enable the focused, adeguate review by NRC of specific aspects related to design and construction of an in situ test facility and final geologic repository, as presented by the DOE in l
Site Characterization Reports.and License Applications (LA).
I This report provides a comparative evaluation of various shaft sinking techniques for production shafts for a repository.
The primary comparative evaluation has been conducted fo'r 14 ft internal diameter shafts developed in two composite media using five different i
methods of sinking / lining.
The technical, cost and schedule comparisons distinguish between shafts sunk blind and those which utilize bottom access.
Based on the system ranking, it is concluded
'that no one particular method of sinking exhibits a clear overall superiority.
NUR EG/CR-2856: A REVIEW OF FUGITIVE DUST CONTROL FOR URANIUM MILL TAILINGS. LI. C. T. s ELMGRE,M.R s HARTLEY,J.N.
Battelle Memorial Institute, Pacific Northwest Laboratory.
January 1983.
50pp.
40
'l
~_
l x yt Qc t
s 1
\\ \\
8302250094.
PNL-4360.
17300:309.
,An immediate concern associated with the disposal of uranium mill t'al l i ng s is that wind erosion of the tailings from an impoundment area will subsequently. deposit tailings on surrounding areas.
Pacific Northwest Laboratory (PNL), under contract to the U.S.
Nuclear Regulatory Commission, is investigating the current technology for fugitive dust control.
Different methods of fugitive dust control, including chemical, l
physical, and vegetative, have been used or tested on mill tailings piles.
This report presents the results of a literature review and i
discussion with manufacturers and users of available stabilization materials and techniques.
NUR EQ/CR-2863: VALENCE EFFECTS DN ADSORPTION: A PRELIMINARY ASSESSMENT OF<THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
MEYERiR.E.s, ARNOLD,W.D.s CASE,F.s et al.
Dak Ridge National La!s ora t ory.
February 1983.
31pp.
8303170624. 'ORNL-5905.
- ~17615: 160, Estimation of the rates of migration of nuclides from nuclear il.
waste repositories requires knowledge of,the. interaction of these nuclides with the compon'ents of the geolcgical formations in the path of the mi ration.
Determination of these interactions requires that b
the valence state of the,nuclide be known.
If the valence state is not known,\\ then there _ can be nos confidence in use of the data for safety analysis.
The vslence state of some nuclides can be determined by means of solvent extraction techniques.
Spectrophotometry can also-be used but only at concentration high enough to give enough absorption of light.
The,une of Eh pH' diagrams along with an indicator electrode in the solution to predict valence states assumes lp#
that 'the entire system it at aquilibrium andJthat the electrode is at
,1 i-equilibrium with the redor.' systems in solution.
Attainment of equilibrium is not often attained howevern,and unless equilibrium can be-demonstrated, indicator'clectrodes along with Eh-pH diagrams most be used with considerable, caution. ' Electrochemical arguments are t
advanced to illustrate that'what is usually measured in practice is a l
mired potential determined by tha'tinatics of than electrode processes l
occurring at the indicator electrode.
Valence states can be altered i
electrochemical 1g or by use of added chemical reagents,' including l
redex couples which can hold the potential to~relatively specific l
p otenti al s.
The disadvantage of added chemical reagents is that they may alter the characteristics of the sorption reactions by interaction
.with the sorbent.
l cNUREQ/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FDR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report.
l April 1 - June 30,1982. BALL, S. J. s CLAPP,N.E.s' CLEVELAND,J.C.s et al.
N Dak Ridge Naticnal Laboratory.
February 1983.
30pp.
8303160009.
ORKL'/ 7M-84 43/V2.
17615:076.
Continuing work su High-Temperature Cas-Cooled Reactor (HTOR) severe accident analyses included a study of a hypothetical large-scale release following a permanent loss-of-coolant accident at the Fort St. Vrain reactor and further development of the ORECA code l
for siting ~ntudies-of the 2240 MW(t) cogeneration plant HTOR.
Work on i
fission product release and transport includee investigations of i
alternative iodine chemistry scenarios and an analysis of'the major areas of uncertainties. in release predictions during. severe accidents.
Code development work showed further progress inisteam generator 41
<t I
e
l 4'
modeling, development of a multiloop HTOR simulation, and testit.g of an alernative simplified core model.
NUR EQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS. CRAWFDRD,S.L.s DOCTDR,S.R.s TAYLOR. T. T. s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
February 1983.
25pp.
8303220422.
PNL-4373.
17667:110.
To provide confidence in the integrity of a reactor during an over cooling transient, it is necessary for nondestructive evaluation to demonstrate high probabilities of detecting cracks located 6.0 mm deep and deeper at the pressure vessel clad surface.
The cracks of interest may be parallel or perpendicular to the clad lag.
Ultrasonic techniques developed and used in Europe are evaluated in this paper for their use on U. S.
Flaw detectability experiments were carried out by testing the inspection technique's ability to detect artificial flaws under several types of clad, including some Manual Metal Arc (MMA) clad.
Both ground and unground clad surfaces were evaluated.
Crack sizing tests of the inspection technique were made using a crack tip diffraction technique.
The data reported here indicate that for sufficient 1g smooth clad surfaces, the 70 degree compressional wave technique is extremely effective for detecting under-clad cracks.
In addition, results show that dramatic signal-to-noise improvements can be made by grinding the clad surface.
Specifically, a reduction in noise level of 10 to 12 dB was achieved by improving the surface condition by a factor of two from 0.012 in.
R MS t o 0. 006 in. RMS.
This reduction in noise moves the crack detectability confidence level from low to very high.
NUR EQ/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT OF AVAILABLE MARGINS INHERENT IN FLOOD PROTECTION OF NUCLEAR POWER PLANTS.
BOROMAN L.E.
Army, Dept. of, Army Engineer Waterways Experiment Station.
January 1983.
163pp.
8302280006.
17320:136.
{
The design of structures to withstand long-term environmental hazards is based on uncertain estimates of the severity of the i
conditions which the structure will encounter.
The uncertainty of the l
estimates depends on (1) inaccuracies in the available data, (2) sample variability in the data, (3) inadequate understanding of the basic mathematical and probability structure of the natural phenomena, and (4) inability to catch all possible extremal conditions in the finite data base.
These sources of uncertainty are present whether one seeks to make classical extremal extrapolations or tries to define a " design hazard set."
In the reported research, the particular problems related to the i
determination of measures of reliability for long-term maximum flooding at nuclear power plants sites were studied by evalua ting thoroughly current procedures in the field.
l NUREC/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT. DRAGO,J.P.s BORKDWSKI, R. J. s FRAGOLA J.R.s et al.
Dak Ridge National Laboratory.
January 1983.
158pp.
8303290717.
DRNL/TM-8465.
17745:206.
The objective of the In-Plant Reliability Data (IPRD) pilot program is to develop a comprehensive, component-specific data base for probabilistic risk assessment and for other statistical analyses relevant to component reliability evaluations.
This objective was attained through a cooperative effort with several utilities, wherein 42 l
each utility provided access to the maintenance files and pertinent population information, and in return, received a computerized listing of both the component populations and the component maintenance records.
This data base includes (1) a comprehensive component population list for each plant including electromechanical and mechanical equipment, i. e., pumps, valves, diesel generators, inverters, and batteries, and (2) a comprehensive cciponent failure and repair history including all corrective maintenance action on each component.
This document details the data collection and analysis related to pumps in nuclear power generating stations.
The data base is developed primarily from a comprehensive record of corrective mintenance actions obtained directly from nuclear plant maintenance files.
A comprehensive pump population is also included in the data base.
This report represents data and reliability statistics on PWR and BWR power plants.
NUREQ/CR-2887: RELAP5 ASSESSMENT: FLECHT SEASET STEAM QENERATOR TEST 23402. MMETYK,L.N.
Sandia Laboratories.
March 1983.
63pp.
8303290677.
SANDB2-2894.
17738:032.
The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal / hydraulic response of LWR's during accident and off-normal conditions.
The RELAP5 code is being assessed at SNLA against test data from various integral and separate effects test facilities.
As i
part of this assessment matrix, a steam generator transient performed at the FLECHT SEASET test facility has been analgred.
NUR EQ/CR-2891: PERFORMANCE TESTING DF PERSONNEL DDSIMETRY SERVICES: Final Report Of Test 3.
PLATO,P.s MIKLDS,J.
Michigan, Univ.
o f.
February 1983.
175pp.
8303100443.
17497:063.
In September, 1977, the University of Michigan began a pilot study of the Health Physics Society Standards Committee (HPSSC)
Standard entitled, " Criteria for Testing Personnel Dosimetry Performance."
Approximately 70 dosimetry processors volunteered to participate in one or more of three tests of the HPSBC Standard.
The results from Tests #1 and #2 were used to evaluate and revise the Standard which was then adopted by the HPSSC in June, 1981.
The Standard was also adopted by the American National Standards Institute as ANSI N13.11-1982 in June, 1982.
Test #3 of the revised HPSSC Standard was condected from November, 1981 to April, 1902.
The objectives of Test #3 were to determine if the Standard is acceptable for future testing programs, and to provide experience with the final version of the Standard.
The passing rate among all the processors for Test #3 was 75%
l compared to passing rates of 48% and 62%.for Tests #1 and #2, respectively, with adjustments made for changes in the Standard following Test #2.
Among all the individual dosimeters irradiated during Test #3, 89% had a reported dose within plus or minus 50% of the delivered dose compared to 79% and 85% of the dosimeters irradiated for Tests #1 and #2.
The HPSSC Standard was found to be an acceptable measure of minimum performance and an appropriate basis for a regulatory program to accredit dosimetry processors.
NUREQ/CR-2892: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY BERVICES: A Revised Procedures Manual. MIKLDS,J.s PLATO,P.
Michigan, Univ. of, 43
Medical School.
February 1983.
216pp.
8303100453.
17497:238.
The U.S.
Nuclear Regulatory Commission's pilot study of the Health Physics Society Standards Committee Standerd, " Criteria for Testing Personnel Dosimetry Performance," was begun in 1977.
A third test of this Standard was conducted from November, 1981 through April, 1982.
The objective of this Procedures Manual is to describe the procedures used for Test #3 which reflect the changes in the Standard from Tests #1 and #2.
This Manual describes each of the radiation sources us 2d for Test #3, as well as administrative procedures used during the testing program.
Methods of irradiation, quality control, data analysis, record keeping, and handling large numbers of dosimeters are presented.
This Manual discusses the role of the National Bureau of Standards in verifying the validity of the calibration of each radiation source.
Suggestions for improving irradiation procedures are included as
~
well as recommendations that will facilitate the operation of the permanent testing facility.
NUR EQ/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERC00 LING ACCIDENTS: A PARAMETRIC ANALYSIS. CHEVERTON, R. D. s ISKANDER,S.K.s BALL,D.C.
Oak Ridge National Laboratory.
February 1983.
59pp.
8304060012.
ORNL/TM-7931.
17862:108.
There are certain hypothetical accidents associated with pressurized-water reactors that can result in severe thermal shock to the reactor pressure vessel at a time when the primary-system pressure l
is substantial.
These overcooling accidents, coupled with a reduction
~
in fracture toughness due to the exposure of the vessel to fast neutrons, introduce the possibility of propagation of preexistent flaws on the inner surface of the vessel.
In order to evaluate the magnitude of the problem and to provide a " handbook" assessment capability, a fracture-mechanics parametric-type study was conducted for a number of postulated transients, assuming an initial flaw in the form of a long axial crack.
In addition to the large-break i
loss-of-coolant accident, the postulated transients consisted of a constant pressure and an exponential decay of the temperature of the coolant in the downcomer.
Parameters that were varied in the study included the thermal decay constant, coolant asymptotic temperature, primary system pressure, initial toughness of the material, copper i
concentration in the material, and crack depth.
NUR EC/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL STATEMENTS. JOHNSON J.D.s RITCHIE,L.T.
Sandia Laboratories.
March 1983.
35pp.
8303220429.
SANDB2-1693.
17667:046.
The CRAC2 computer code has been adapted to the calculation requirements of Draft / Final Environmental Impact Statement (DES /FES) l casework analysis for the Nuclear Regulatory Commission.
CRAC2 is a revised version of the CRAC (Calculation of Reactor Accident Consequences) computer code developed in support of the Reactor Safety Study, WASH-1400.
A graphical output package has been developed for displaying CRAC2 computed results.
All phases of the casework i
analysis calculations from initial data formatting to plotting of calculated results are executed through the use of procedure files on the Idaho National Engineering Laboratory (INEL) computing system at Idaho Falls, Idaho.
The INEL computing system operates under the j
Control Data Corporation (CDC) NOS/BE Operating System (Level 518) and Intercom Version 5.
44
NUREQ/CR-2902:
SAFETY CODE: CLAD RELOCATION. FLOW RECIME MODELING STUD CANAPOL.B.D.
Los Alamos Scientific Laboratory.
February 1983.
30pp.
8303230362.
LA-9499-NS.
17673:330.
A clad relocation model has been developed for the SIMMER-II LMFBR safety analysis code.
The new model introduces the appropriate momentum and energy exchange function the current SIMMER-II calculational framework.for clad motion without changing Results using this new model were compared to the TREAT R5 experiment and to several calculational models with reasonably good agreement.
other NUREO/CR-2904: SIMULATION OF OROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL. MCKAY,E.D.s JOHNSON.T.M.s READE.R.T.
COS, Inc.
January 1983.
32pp.
8303150130.
SAND 82-7077.
17600:103.
The purpose oP this report is to describe through examples the operation of the deterministic probabilistic containment transport (DPCT) model.
Results from ten example simulations are presented to illustrate the use and capabilities of the model and to provide users with the basic knowledge necessary to become proficient with its use.
The development of the DPTC model, its verification, and a description input and output parameters are discussed in detail in Schwartz and Crowe, 1980.
Much of the material presented in the User Meaual is not repeated here.
Users of this problem set should obtain copies of the Schwartz and Crowe (1980) report.
Section 2 of the present report describes the containment transport model DPCT.
Section 3 presents ten example simulations.
The hydrogeologic systems and the computer simulations were chosen to illustrate the effects of model parameters on predicted groundwater flow patterns and containment examined include hydraulic conductivity, dispersivity, radioactive decay, cation exchange and speci fied boundary conditions.
NUREC/CR-2907 VO1: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980. TICHLER,J.s BENKOVITZ,C.
Brookhaven National Laboratory.
January 8983.
215pp.
8302220426.
BNL-NUREC-51581.
l 17240:114.
Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1980 have been i
compiled and reported.
Data on solid waste shipments as well as t
selected operating information have been included.
This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission.
The 1980 release data are compared with previous years' releases in tabular form. Data covering specific radionuclides are summarized.
NUREQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SUR. TACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE. WANQ,J.S.V.s MANCOLD,D.C.s SPENCER,R.K.s et al.
Lawrence Berkeleg Laboratory.
March 1983.
239pp.
8304050637.
LBL-13341.
17848:149.
This report studies the thermal effects associated with the emplacement of aged radioactive waste in a geologic repository, in particulars the waste characteristics, repository structure, and rock properties controlling the thermally induced effects, the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous 45
studies that assume 10-year-old wastes.
The thermal criteria used to determine the repository waste loading densities, and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts.
The waste loading densities determined by repository designs for 10 year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations.
The eutension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have onig modest long-term effects for spent fuel based on far-field thermomechanical and thermohydrologic considerations.
More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.
NUR EC/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS. BURKE R.P.s HEISING,C.D.
Massachusetts Institute of Technology.
ALDRICH,D.C.
Sandia Laboratories.
March 1983.
150pp.
8303290689.
SANDB2-2OO4.
17738:096.
Offsite response decision-making methods based on in plant conditions are developed for use during severe reactor accident situations.
Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent fuel storage pool from consideration for immediate offsite emergency response during accident situations.
A simple plant information management scheme is developed for use in offsite response decision-making.
Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite response implementation for a range of PWR accidents involving the reactor core.
In plant decision criteria for offsite response implementation are defined.
The definition of decision criterie is based on consideration of core accident physical processes, in plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public protective measures.
It is recommended that a small area near the plant be evacuated when large fractions of the core gap inventory are released to the containment structure.
Extension of the protective action area is appropriate if fuel melting within the core region is indicated.
NUR EC/CR-2928: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-1. OSBORNE,M.F.s LORENZ,R.A.s TRAVIS.J.R.s et al.
Oak Ridge National Laboratory.
February 1983.
50pp.
8303290698.
ORNL/TM-8500.
17742:270.
This first in a series of high-temperature fission product j
release tests was conducted for 30 min at 1400 degrees centigrade, with the release taking place into flowing steam.
The fuel specimen was a 20-cm-long section of H.B.
Robinson fuel rod, irradiated to 28,000 mwd per metric ton (t).
After tha test, the Zircalog cladding of the specimen was almost completely oxidized and was quite fragile.
The fission product collection system included a thermal gradient tube (700-150 degrees centigrade), filters, heated charcoal, and cooled charcoal.
Gamma ray analysis of apparatus components and collectors showed that about 2.83% of the (85) Kr nd 1.75% of the (137) Cs were released from the fuel.
Activation analysis of leach solutions from these components indicated that 2.04% of the (129) I was released.
Other analyses revealed small but significant release of the radionuclides (125) Sb and (106) Ru, and of the elements, Br, Rb, Sr, 46
Z r, Ag, Sn, Te, Ba, and La.
In addition, several other elements which originate as furnace materials or as impurities were collected.
These preliminary results are in general agreement with release values predicted from previous studiess more complete data evaluation will be reported later.
NUREC/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTOR CORE SUPPORT ORAPHITE--PART III. MORGAN,W.C.s DAVIS,T.J.s THOMAS,M.T.
Battelle Memorial Institute, Pacific Northwest Laboratory.
February 1983.
49pp.
8303100421.
PNL-4449.
17503:347.
Methods are being developed to monitor, in-situ, the strength changes of graphite core-support components in a High-Temperature Gas-Cooled Reactor (HTGR).
The results reported herein pertain to the development of techniques for monitoring the core-support blocks; the PQX graphite used in these studies is the grade used for the core-support blocks of the Fort St. Vrain HTOR, and is coarser-grained than the grades used in our previous investigations.
- The, thrcugh-transmission ultrasonic velocity technique, developed for monitoring strength of the core-support posts, is not suitable for use on the core-support blocks.
Eddy-current and ultrasonic backscattering techniques have been shown to be capable of measuring the density-depth profile in oxidized PCX and, combined with a correlation of strength versus density, could vield an estimate of the strength-depth profile of in-service HTCR core support blocks.
Correlations of. strength versus density and other properties, and progress on the development of the eddy-current and ultrasonic backscattering techniques are reported herein.
1 NUR EC/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS. EGGERS,R.F.s BROUNS,R.J.s BRYANT.J.L.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
January 1983.
294pp.
8303100416.
PNL-4336, 17498:094.
Proposed regulations for NRC licensees authorized to possess and process formula quantities of strategic special nuclear material (SSNM) would require each licensee to implement a material control and accounting (MC&A) system capable of prompt loss detection and alarm resolution.
In support of the loss detection and alarm response activities an overcheck program would also be implemented.
This program would include personnel qualification and training, quality control, inventory verification and shipper-receiver transaction verification.
However, the frequency of physical inventory verification would be about once per year rather than one every two months.
In addition MC&A activities would include procedures for the prevention and detection of data falsification and other forms of l
deceit that might undermine the performance of the loss detection and I
response systems.
This report provides examples of prompt accountability systems for four plants as follows:
l Mixed Oxide Fuel Fabrication Uranium Hexafluoride Conversion High Enriched Uranium Fuel Fabrication High Enriched Uranium Scrap Recovery.
NURES/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CONVECTION HEAT TRANSFER OF INTERNALLY HEATED LIGUIDS. GREENE,0.A.
Brookhaven National Laboratory.
February 1983.
56pp.
8303230512.
BNL-NUREG-51585.
17677:119.
Boundary heat transfer from a liquid pool with a uniform internal l
l 47 i
l I
heat source to a vertical or inclined boundary was investigated.
The experiments were performed in an open rectangular liquid pool in which the internal heat source was generated by electrical heating.
The local heat flux was measured to a boron nitride test wall which was able to be continuously inclined from vertical.
Gold plated microthermocouples of 0.01 inch outside diameter were developed to measure the local surface temperature, both front and back, of the boron nitride.
The local heat flux and, thus, the local heat transfer coefficient was measured at nineteen locations along with vertical axis of the test plate.
Under the assumptions of laminar natural convection, it was found that the local heat transfer data could be l
empirically correlated by the relation Nu (x) = 0.153 Ra (x) (0.25) with a standard deviation in the coefficient of plus/minus 14%, and the average heat transfer data could be empirically correlated by Nu =
f 0.127 Ra(L) (0.25) with a standard deviation in the coefficient of plus/minus 8%.
The component of the gravitational acceleration along l
the wall, g (eff) = g cos was found to help account for the l
inclination of the boundary although some systematic effect was still
{
evident in the data.
l NUREC/CR-2944: TORNADO DAMAGE RISK ASSESSMENT. REINHOLD,T.A.;
ELLINOWOOD,B.
Commerce, Dept. o f, National Bureau of Standards.
February 1983.
62pp.
8303290707.
BNL-NUREG-51586.
17737:329.
A study has been conducted to evaluate several proposed models for predicting tornado wind speed probabilities at nuclear plant sites as part of a program to develop statistical data on tornadoes needed for probability-based load combination analysis.
A unified model was i
developed which synthesized the desired aspects of tornado occurrence and damage potential.
The sensitivity of wind speed probability estimates to various tornado modeling assumptions are examined, and the probability distributions of tornado wind speed that are needed for load combination studies are presented.
NUREC/CR-2945: CHARACTERIZATION OF EARTHGUAME FORCES FOR PROBABILITY-BASED DESIGN OF NUCLEAR STRUCTURES. ELLINOWOOD,B.4 MATTS M.
Commerce, Dept. o f, National Bureau of Standards.
Brookhaven National Laboratory.
February 1993.
70pp.
8303290704.
B NL/NUREG-51587.
17737:257.
Research is underway to develop probability-based loading criteria for the design of safety-related steel and reinforced concrete nuclear plant structures.
The random nature of earthquake loads and related structural response must be established before load criteria for design that include earthquake effects can be developed.
This report provides a general statistical description of earthquake actions.
Sources of uncertainty that might affect structural response are considered, including uncertainties in the peak. ground motion at the site and in the structural response to prescribed ground motions.
Power spectral density representations of earthquake ground motion J
l needed for predicting response using random vibration analysis are examined and related to site conditions.
The emphasis is on probabilistic models that are consistent with available data on seismicity and structural response.
I NUR EQ/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER l
PLANTS: THE FOREIGN EXPERIENCE. MELBER.B.D.s SCHREIBER,R.E.
Battelle Memorial Institute, Pacific Northwest Laboratory.
February 1983.
21pp.
8303090510.
17476:125.
l 48
t This report describes the practices of selected foreign countries in providing engineering expertise on shift in nuclear power plants.
The extent to which engineering expertise is made available and the alternative models of providing such expertise are presented.
The implications of foreign practices for U.S.
consideration of alternatives are discussed, with reference to the shift technical advisor (STA) position and to a proposed shift engineer position.
The procedure used to obtain information on foreign practices was l
primarily a review of the literature, including publications.
l presentations, and government and utility reports.
There are two l
approaches that are in use to make engineering expertise available on j
shift: (1) employing a graduate engineer in a line management operations position, and (2) creating a specific engineering position for the purpose of providing expertise to the operations staff.
The comparisons of these two models did not indicate that one system inherent 1g functions more effectively than the other.
However, the alternative models are likely to affect crew relationships and performances labor suppig, recruitment, and retentions and system implementation problems.
l NUREC/CR-2956: NEUTRON DOSIMETRY AT CDMMERCIAL NUCLEAR PLANTS. Final l
Report Of Substask B: Dosimeter Response. CUMMINGS.F.M.: ENDRES.C.W.s BRACKENBUSH,L.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
115pp.
8304130448.
PNL-4471.
17959:222.
As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterire neutron dosimeter responses inside the containment structure of commercial nuclear plants.
In order to characterize those responses, dosimeters were irradiated inside containment at 2 pressurized water reactors and at pipe penetrations outside the biological shield at two boiling water reactors while the reactors were operating at full power.
Additionally, the dosimeters were t
irradiated (1) using monoenergetic neutrons produced by an accelerator l
and (2) using the filtered reactor irradiations simultaneous measurements were taken using a tissue equivalent proportional counter and portable remmeters, SNOOPY, RASCAL and PNR-4.
The results of the analyses of dosimeter responses indicate that (1) the dosimeters irradiated inside containment' of PWR's respond as if the dosimeters were irradiated using moncenergetic neutrons below 100 kev, (2) that the use of bare neutron sources for dosimeter calibrations is inappropriate for the in-containment irradiations, (3)
I the TLD-albedo dosimeter is the only type of dosimeter available that demonstrated both adequate precision and sensitivity, (4) that the polycarbonate track etch dosimeter which uses radiators was sensitive enough, but demonstrated inadequate precision for the reactor irradiations and that (5) CR-39 and polycarbonate track etch dosimeters which do not use radiators were inadequate for use inside containment of nuclear power plants.
NUREC/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP l
l GEDLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE. CATES.R.s i
BAUHOF,F.s CONANO,L.
Golder Associates.
March 1983.
52pp.
8303300446.
813-1165.
17763:245.
The purpose of the complete project is to provide the NRC with technical assistance to enable the focused, adequate review of specific aspects related to design and construction of an in situ test facility and final geologic repository as presented by DOE in site characterization reports and license applications.
49
i This report evaluates and presents recommendations on the relationship between an in situ test facility (exploratory shaft and underground test facility) and the repository.
Technical and construction issues are addressed in generic terms for an in situ test facility as an integral part of the repository.
Media / site specific considerations are also presented.
These issues include location and size of the test facility, the design / construction / operational features of the in situ test facility, and the nature and duration of in situ testing (as presented in the Task 2 report - NUREQ/CR-3065).
It is recommended that the exploratory shaft and test facility allow for adeguate characterization of the strata surrounding the repository horizon and for explicit full-scale sampling of the engineering behavior of the materials of the repository site.
1 Construction of the exploratory shaft and test facility should utilize the same methods as planned for the proposed repository structures.
WREC/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE B-4.
LONGEST,A.W.s CHAPMAN,R.H.s CROWLEY,J.L.
Oak Ridge National Laboratory.
January 1983.
66pp.
8303170616.
ORNL/TM-8509.
17622:309.
A compilation of bundle B-4 test data is presented.
These data were obtained during the test and from pretest and posttest l
examination of the test array.
This research is designed to investigate Zircalog cladding deformation behavior under simulated light-water reactor loss-of-coolant accident conditions.
The specific objectives of the B-4 test were to investigate axial propagation of cladding ballooning as the result of rod-to-rod contact and to determine the effect of a relatively cold fuel pin simulator on the l
deformation behavior of its hotter neighbors under test conditions known to produce large deformation.
These objectives were not realized, however, because the temperature transient did not proceed l
to the planned failure conditions.
Electrical power to the bundle was lost when the bundle average cladding temperature was 675 degrees centigrade.
The bundle temperature slowly decreased (*0.2 K/s for
'380 s),
and the tubes deformed (by creep) until pressure was vented from the tubes to terminate the test.
Significant deformation occurred (up to 18% average strain over the heated length), but none of the tubes burst.
The bundle was disassembled, and deformation profiles of the individual tubes were measured.
Although objectives of the test were not met, the data appear useful for model development and verification.
NUREC/CR-2972: AN ANALYSIS OF DENSITY-WAVE OSCILLATIONS IN VENTILATED CHANNELS. TALEYARKHAN,R.s PODOWSKI,M.s LAHEY,R.T.
Rensselaer Polytechnic Inst.
March 1983.
545pp.
8304200504.
18113:203.
A mathematical model has been developed for the linear stability analyses of a system of ventilated parallel boiling channels.
The model can accomodate phasic slip, arbitrary non-uniform axial power distributions, distributed local losses, heater wall dynamics, channel-to-channel radial power skews, discrete or continuous ventilation between the channel, turbulent mixing between the channels, various donor-cell options for the lateral transport of energy nd momentum, and a transverse momentum equation, including storage and cross-flow inertial.
A special matrix reduction scheme was developed to efficient 1g solve the system of linearized, Laplace transformed, nodal equations.
A parametric study revealed that phasic slip, axial power distribution, heat wall dynamics, local losses, lateral ventilation 50
and radial power skew can have a significant effect on the stability characteristics of the system.
Comparison were made with the limited experimental data that exists, and good agreement was achieved.
NUR EC/CR-2974: USER'S MANUAL FOR LPQS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere. WHITE,J.E.s ECKERMAN,K.F.
Oak Ridge Nationa1 Laboratory.
March 1983.
167pp.
8303280066.
ORNL/TDMC-2.
17711:214.
The LPCS computer program was developed to calculate the radiological impacts resulting from radioactive releases to the hydrosphere.
The hydrosphere is represented by the following types of l
water bodies:
estuary, small river, well, lake, and one-dimensional l
(1-D) river.
The program is principally designed to calculate radiation dose (individual and population) to body organs as a function of time for the various exposure pathways.
The radiological consequences to the aquatic biota is estimated.
Several simplified i
radionuclide transport models are employed with built-in formulations to describe the release rate of the radionuclides.
Optionally, a tabulated user-supplied release model can be input.
Printer plots of dose versus time for the various exposure pathways are provided.
NUR EC/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation. ARNAHAN C.L.s DELANY, J. M. s LONG.J.C.S.s et al.
Lawrence Berkeley Laboratory.
January 1983.
1BOpp.
8302250122.
i LBL-15029.
17301:006.
Four issues are considered that must be addressed by a site characterization program designed to evaluate the suitability of the flood basalts of the Pasco Basin in central Washington as a site for the construction of a repository for the disposal of high-level nuclear waste.
The four issues are (1) identification of hydrostratigraphie units within a sequence of flood basalts, (2) mechanisms and points of groundwater recharge and discharge. (3) s solubility of radionuclides, and (4) phas> transformation of fracture filling materials.
Each issue is discussed in terms of its significance to waste isolation.
Available approaches for resolving the issues are presented and their limitations identified.
When appropriate, research programs for overcoming these limitations are indicated.
NUREC/CR-2993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM GENERATOR AT THE MAINE YANKEE NUCLEAR POWER STATION. CZAJK0WSKI,C.
Brookhaven National Laboratory.
February '1983.
32pp.
8303160620.
l B NL-NUR EG-51594.
17602:251.
l Three studs removed from service on the primary manway cover from steam generator #2 of the Maine Yankee station were sent to Brookhaven National Laboratory (BNL) for examination.
The examination consisted of visual / dye penetrant examination, optical metallography and r
l Scanning Electron Microscopg/ Energy Dispersive Spectrocopy (SEM/EDS) evaluation.
One bolt was "through cracked" and its fracture face was generally transgranular in nature with numerous secondary intergranular cracks.
The report concludes that the environmentally l
assisted cracking of the stud was due to the interaction of the l
various lubricants used with steam leaks associated with this manway cover.
l 51 l
NUREC/CR-2997 VO2: APPLICATIDNS OF ENERGY RELFASE RATE TECHNIQUES TO P ART-THROUGH CRACMS IN PLATES AND CYLINDERS. Volume 2: DRVIRT: A Finite Element Program For Energy Release Rate Calculations For 2-Dimensional And 3-Dimensional Crack Models. BASS,B.R.s BRYSON J.W.
Oak Ridge National Laboratory.
February 1983.
52pp.
8303300511.
ONNL/TM-8527/V2.
17748:213.
In nonlinear applications of computational fracture mechanics, energy release rate techniques are used increasing 1g for computing stress intensity parameters of crack configurations.
- Recent1g, deLorenzi used the virtual-crack-extension method to derive an analytical expression for the energy release rate that is better suited for 3-dimensional calculations than the well-known J-integral.
Certain studies of fracture phenomena, such as pressurized-thermal-shock of cracked structures, require that crack tip parameters be determined for combined thermal and mechanical loads.
A method is proposed here that modifies the isothermal formulation of deLorenzi to account for thermal strains in cracked bodies.
This combined thermo-mechanical formulation of the energy release rate is valid for general fracture, including nonplanar fracture, and applies to thermo-elastic as well as deformation plasticity material models.
The formulation has been implemented in the virtual-crack-extension program ORVIRT (Dak Ridge virtual-Crack-Extension).
Program DRVIRT performs energy release rate l
calculations for both 2-and 3-dimensional nonlinear models of crack c onfig urations in engineering structures.
Two applications of the ORVIRT program are described here.
In the firstesemielliptical surface cracks in an experimental test vessel are analyzed under l
elasticplastic conditions using the finite element method.
l NUREC/CR-2999: A CDMPARISDN OF BWR STABILITY MEASUREMENTS WITH C ALCULATIONS USING THE CODE LAPUR-IV. MARCH-LEUBA, J. s OTADUY,P.J.
Oak Ridge National Laboratory.
February 1983.
42pp.
8303140430.
ORNL/TM-8546.
17569:308.
A parametric study of stability characteristics in boiling water reactors (BWRs) was performed using the frequency domain code L APUR-I V.
Two different reactors, Peach Bottom Unit 2 and Vermont Yankee, were considered in a total of 17 different operating conditions that corresponded to three series of low-flow stability tests performed in these two reactors.
Stability margins, in terms of decay ratio and natural frequency of the closed loop reactivity-to power transfer function (T.F.),
were calculated and then l
i compared with the experimental results.
In addition, a sensitivity l
analysis was performed to determine the changes in calculated results to be expected in response to alterations in the density reactivity coefficient (DRC), the recirculation loop pressure-to-flow T.F.
parameters, and the fuel-to-cladding-gap heat conductance.
This i
allows assessment of the effect of input data uncertainty on the calculated results.
Satisfactory agreement was found between the stability margins calculated using LAPUR-IV and the experimental results.
The sensitivity analysis shows that the DRC is the most critical of the parameters investigated for accurate stability calculations.
l NUREQ/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE i
CONCRETE INTERACTIONS. RANDICH,E.s SMAARDYK,J.E.s ACTON,R.U.
Sandia Laboratories.
March 1983.
115pp.
8304060020.
SANDB2-2315.
17868:016.
Eleven large scale tests examining the interaction of molten l
N
- i sodium and limestone-(calcite) concrete were performed.
The tests typically used between 100 and 200 kg of sodium at temperatures between 723 K and 973 K and a total sodia/ concrete contact area of
'1. Om (2).
The results show that energetic reactions can occur between sodium and limestone concrete.
Delay times of less than 30 minutes were observed before the onset of the energetic phase.
Not all tests exhibited energetic reactions and the results indicate that there is a sodium temperature threshold of 723 K to 773 K which is necessary to initiate the energetic phase.
Maximum heat fluxes during the energetic phase were measured at 3.6x10 (5) J/m (2)-s.
Maximum p enetration rates were 4 mm/ min.
Total concrete erosion varied from 1 tc 15 cm.
NUREC/CR-3OO8: AUDITORY PERCEPTIDN IN LOOSE-PARTS MONITORING.
HOWARD,J.H.
Catholic Univ.
January 1983.
50pp.
8302170205.
17188:161.
i This survey assessed the safety information potentially available to operators in the audio output of loose parts monitors for nuclear power reactors.
Three tasks were completeds (1). Literature reviewed to identify acoustic signal parameters of primary interest and relate these physical parameters to known human auditory detection and recognition capabilitiess (2). Current use of auditory information examined product descriptions and visits to a limited number of manufacturers and nuclear power plants, and.(3). Optimal use of human auditory capabilities in loose-parts monitoring were recommended.
Conclusions were:
(1).
Audio surveillance can be enhanced by automatic monitoring supplemented by periodic operator listening for alarmss (2).
Training procedures could improve human auditory detections (3).
Audio information may be used to diagnose location, number and size of detected loose-parts.
Since both practical and theoretical limitations exist for the full automation of this j
function, human ability to perceive subtle spectral differences may be useful when combined with current automatic diagnosiss and (4).
Audio l
data can provide feedback on the response of sonically active remote equipment in limited access areas.
i NUR EC/CR-3OO9: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
KNOROVSKY,0.A.; KRIEC, R. D. s ALLEN,0.C.
Sandia Laboratories.
February 1983.
225pp.
8303160614.
SAND 78-2347.
17602:318.
The objective of the Fracture Toughness of Component Supports Program was to perform a generic fracture toughness evaluation of materials used in operating Pressurized Water Reactor component supports.
Historically, the program was initiated as a result of experiences that occurred during the licensing of the Virginia Electri c Power Company 's North Anna Station.
The materials used in the component supports are classified according to three categories:
(1) structural materials, (2) weld consumables, and (3) bolting materials.
Material property data from numerous literature sources for these steels were assessed.
1 1
NUREQ/CR-3010: POST EVENT HUMAN DECISION ERRORS: DPERATOR ACTIDN l
TREE / TIME RELIABILITY CORRELATION. HALL. R. E. s FRAGOLA,J.s WREATHALL,J.
Brookhaven National Laboratory.
March 1983.
49pp.
8304060036.
BNL-NUREC-51601.
17B64: 043.
This report documents an interim framework for the quantification l
of the probability of errors of decision on the part of nuclear power plant operators after the initiation of an accident.
The framework 53
l l-l l
can easily be incorporated into an event tree / fault tree analysis.
The method presented consists of a structure called the operator action tree and a time reliability correlation which assumes the time available for making a decision to be the dominating factor in l
situations requiring cognitive human response.
This limited approach i
decreases the magnitude and complexity of the decision modeling task.
I Specifically, in the past, some human performance models have attempted prediction by trying to emulate sequences of human actions, or by identifying and modeling the information processing approach applicable to the task.
Although such modeling approaches can provide considerably greater insight into individual human behavior and the reasons for that behavior, this type of modeling for the full spectrum i
of relevant nuclear power plant tasks is extremely ambitious and goes I
beyond the requirements.
The model developed here is directed at describing the statistical performance of a representative group of hypothetical individuals responding to generalized situations.
NUR EQ/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
GREENE, 0. A. s QINSBERG,T.s KAZIMI,M.S.
Brookhaven National Laboratory.
March 1983.
126pp.
8304060024.
BNL-NUREG-51621.
17867:029.
The technology of thermal hydraulic aspects of the transition phase accident sequence in liquid metal fast breeder reactors has been t
reviewed.
Previous analyses of the transition phase accident sequence have been reviewed and the current understanding of major thermal hydraulic phenomenology has been assessed.
As a result of the foregoing, together with a scoping analysis of the transition phase accident sequence, major transition phase issues have been defined and research needs have been identified.
The major conclusion of transition phase scoping analysis is that fuel dispersal cannot be relied upon to rule out the possibility of recriticalities during this stage of the accident.
The potential for fuel blowdown to a subcritical configuration immediately following fuel disruption has been greatly overestimated in much of the previous l
work.
Material freezing in the subassembig blanket structure is the i
major factor expected to prevent blowdown.
While our understanding of fuel motion and freezing phenomena is incomplete, the available evidence strong 1g suggests the likelihood of only limited fuel relocation prior to freezing.
NUR EQ/CR-3027: OVERLAND EROSION OF URANIUM MILL TAILINGS IMPOUNDMENTS:
PHYSICAL PROCESSES AND COMPUTATIONAL METHODS. WALTERS,W.H.
Battelle i
Memorial Institute, Pacific Northwest Laboratory.
March 1983.
63pp.
8304200590.
PNL-4523.
18081:056.
This report reviews the surface runoff and erosional processes of watersheds caused by rainfall-runoff.
Soil properties, topography, l
{
and rainstorm distribution are discussed with respect to their effects on soil erosion.
The effects of climate and vegetation are briefly presented. Regression models and physical process simulation models I
are reviewed.
NUREQ/CR-3028: A REVIEW OF THE LIMERICK QENERATING STATION PROBABILISTIC RISK ASSESSMENT. PAPAZOOLOU,I.A.s KAROL,R.J SHIU,K.s et i
a l.
Brookhaven National Laboratory.
February 1983.
410pp.
8303140664.
BNL-NUREC-51600.
17575:030.
A review of the Probabilistic Risk Assessment of the Limerick 54
Generating Station was conducted with the broad objective of evaluating its risk in relation to those identified in the Reactor Safety Study (WASH-1400).
The review included a technical assessment of the assumptions and methods used in the Limerick study.
It also included a re-evaluation of the main results within the scope and I
general methodological framework of the study.
This included both qualitative and quantitative analyses of accident initiators, data bases, accident sequences which result in core damage, core melt phenomena, fission product behavior, and offsite consequences.
Specific comparisons were made between the Limerick study, the Brookhaven review, and the WASH-1400 reactor for the core damage frequency and the average frequencies of acute and latent fatalities.
l The effect of uncertainties was considered throughout the review process and the uncertainty bands for the risk indices were i
quantified.
NUREQ/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT DF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
SHERWOOD,D.R.s SERNE,R.J.
Battelle Memorial Institute, Pacific Northwest Laboratory.
February 1983.
33pp.
8303160611.
PNL-4524, 17602:283.
Laboratory experiments were conducted to evaluate the performance of selected neutralizing agents for the treatment of uranium tailings solutions.
Highly acidic tailings solutions (pH<2) from the Lucky Mc 4
Mill in Cas Hills, Wyoming and the Exxon Highlands Mill near Casper, Wyoming were neutralized to pH of 7 or greater using seven neutralizing agents.
Reagents used included: 1) Fig Ash from Boardman Coal Plant, Boardman, Oregons 2) Fly Ash from Wyodak Coal Plant, l
Gillette, Wyomings 3) Calcium carbonate (CaCD(3)) reagent grades 4)
Calcium hydroxide CCa(OH)(2)3 reagent grades 5) Magnesium oxide (MgD) reagent grades 6) Sodium carbonate (Na(2)CD(3)) reagent grades and 7)
Sodium hydroxide (NaOH) reagent grade.
i Evaluation of the effectiveness for the treatment of uranium tailings soluions for the selected neutralizing agents under controlled laboratory conditions was based on three criteria.
The criteria are:
- 1) treated effluent water quality. 2) neutralized sludge handling and hydraulic properties, and 3) reagent costs and acid neutralizing efficiency.
On the basis of these limited laboratory results calcium hydroxide or its dehydrated form Ca0 (lime) appears to be most effective option for treatment of uranium tailings solutions.
NUREQ/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS DF AN L
UNPROTECTED TRANSIENT UNDERCDOLING ACCIDENT IN A l
L ARGE, HETEROCENEDUS-CDRE, LIGUID-METAL-COOLED FAST BREEDER REACTOR.
LUCK, L. B. s DEVAULT,0.P.s ASPREY,M.W.
Los Alamos Scientific l
Laboratory.
February 1983.
211pp.
B303230517.
LA-9553-MS.
j 17684:078.
Disrupted-core (transition phase) behavior has been evaluated fo~
a hypothetical, unprotected transient undercooling accident in an early version of the heterogeneous-core liquid-metal-cooled fast breeder reactor (LMFBR) developed for the Conceptual Design Study.
SAS3D was used to predict initiating-phase phenomena.
Two complete calculations were performed, one at the beginning-of-life cycle and j
one at the end-of-equilibrium cycle.
As a result of low sodium-void reactivity and high extensive blockage formation in the axial blankets j
restricted fuel escape in the transition phase.
However, because several critical masses of fuel were available, only after the removal se i
of large quantities of fissile fuel from the core would permanent suberiticality be assured.
Although dispersive phenomena were predicted, ths effects were temporary and several recriticalities occurred.
Permanent subcriticality was predicted onig after i
large-scale disruption of core structure and intermixing of blanket materials.
Separate-effects calculations were performed to determine the adequacy of SIMMER-II modeling of important predicted phenomena.
The predicted behaviors agreed well with available experimental data and thus supported results of the two transition-phase calculations.
I NUR EQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIDNUCLIDES THROUGH SOIL AT THE MAXE',' FLATS, KENTUCKY, WASTE-BURI AL SITE. FDWLER E.B.s I
POLZER, W. L. s ESSINGTON,E.H.
Los Alamos Scientific Laboratory.
March 1983.
75pp.
8304060030.
LA-9565-MS.
17863:328.
Laboratory and field studies conducted in FY 1990 at the Maxey Flats Waste-Burial Site in northcentral Kentucky relate to the movement of water into waste trenches, the movement of waste liquid and radionuclides from trenches, and the effect of soil on this movement.
Two areas at the waste-burial site are being used to study the interaction of soil with liquid waste-one near Trench 19S and the l
other between an experimental trench and Trench 27-using porous cups to obtain samples of the soil solution.
Analyses of soil solutions near Trench 198 indicate that radionuclides have migrated from the waste-burial trench.
The observed distribution of radionuclides in the area suggests that (3) H, as tritiated water, has moved the greatest distance.
Movement of (137) Cs is essentially nonexistent.
The migration of (238) Pu and (60) Co lies between those two extremes.
The distance that (3) H has moved, at an approximated depth of 4 m, is about 9 m.
Additional porous cup samplers were installed at depths to 8 m to better evaluate the distribution of radionuclides near Trench 199.
I NUREQ/CR-3033: MODELING AOKI ET AL. EXPERIMENTS ON CONDENSATION OF FLOWING STEAM DNTO INJECTED WATER VIA K-FIX. MILDISNESS,R.C.s DALY,B.J.
Los Alamos Scientific Laboratory.
February 1983.
12pp.
8303300518.
LA-9574-MS.
17748:286.
The experiments and the 1-D theoretical calculations of Aoki et a l.
are reviewed.
K-FIX calculations of these experiments are discussed and the modifications of the condensation model and the momentum-exchange model needed to secure agreement with experiments are presented.
Calculations are presented for the cases of 30 degrees centigrade and 5 degrees centigrade subcooled water for a particular set of parameters for these models.
Agreement with experiment is very good for the case of 30 degrees centigrade subcooled water and good for the case of 5 degrees centigrade subcooled water.
1 NUREO/CR-3045 VO1: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT OUTAGES. Vol.-1: Approach And Analgsis. BUEHRING,W.A.s PEERENBOOM,J.P.
Argonne National Laboratory.
February 1983.
161pp.
8303300628.
ANL/AA-28.
17766:006.
This report discusses and analgres some of the important consequences of nuclear power plant unavailability, and quantifies a number of technical measures of loss of benefits that may help the Nuclear Regulatory Commission make decisions involving nuclear power plant licensing and operation.
The consequences include increased costs of system generation, increased demand for nonnuclear and often scarce fuels, and reduced system reliability.
Argonne National 56 1
l Laboratory (ANL) developed case studies to investigate the effects of hypothetical nuclear plant shutdowns.
The studies developed quantitative measures of both short-and long-term economic, fuel use, and reliability effects that could result from the unavailability of nuclear generating units.
Results showed that production costs (fuel costs plus operation and maintenance costs) increase significant1g whenever an operating reactor is shut down.
Production-cost increases ranged from less than 10% to over 60%s the normalized increases for the first year of reactor outage ranged from $0.125 million per MWe year to $0.33 million per MWe year.
The discounted production-cost increases for the first 10 years of shutdown in the four ANL cases ranged from $1.8 billion t $4.2 billion.
l l
NUREQ/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT OUTAGES. Vol 2: Appendines. BUEHRING,W.A.s PEERENBOOM,J.P.
Argonne National Laboratory.
February 1983.
149pp.
8304050633.
ANL/AA-28 VOL.2.
17847:221.
This report discusses and analgres some of the important consequences of nuclear power plant unavailability, and quantifies a number of technical measures of loss of benefits that may help the Nuclear Regulatory Commission make decisions involving nuclear power plant licensing and operation.
The consequences include increased costs of system generation, increased demand for nonnuclear and often scarce Fuels, and reduced system reliability.
Argonne National Laboratory (ANL) developed case studies to investigate the effects of hypothetical nuclear plant shutdowns.
The studies developed quantitative measures of both short-and long-term economic, fuel use, and reliability effects that could result from the unavailability of nuclear generating units.
Results showed that production costs (fuel costs plus operation and maintenance costs) increase significant1g whenever an operating reactor is shut down.
Production-cost increases ranged from less than 10% to over 60%s the normalized increases for the first year of reactor outage ranged from $0.125 million per i
MWe-year to $0.33 million per MWe year.
NUREC/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models. THURCDOD M.J.s KELLY,J.M.s GUIDOTTI.T.E.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
186pp.
8304060044.
PNL-4385.
17865:257.
The COBRA / TRAC computer program has been developed to predict the thermal-hydraulic raspons; of r.u:Isar reactor primary cool:nt syctems to small and large break loss-of-coolant accidents and other l
anticipated transients.
The code solves the compressible three-dimensional, two-fluid, three-field equations for two phase flow in the reactor vessel.
The three fields are the vapor field, the continuous liquid field, and the liquid drop field.
a five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, and accumulators.
The heat generation rate of the core is specified by input and no reactor kinetics calculations are included in the solution.
This volume describes the conservation equations and physical models used in the vessel model.
NUREQ/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT 57
SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
THURODOD,M.J.s GEDROE.T.L.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
64pp.
8304050627.
PNL-4385.
17847:111.
The COBRA / TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients.
The code solves the compressible three-dimensional, two-fluid, three-field equations for two phase flow in the reactor vessel.
The three fields are the vapor field, the continuous liquid field, and the liquid drop field.
A five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, nd accumulators.
The heat generation rate
+
l of the core is specified by input and no reactor kinetics calculations are included in the solution.
This volume describes the finite-difference equations and the numerical solution methods used to solve these equations.
It is directed toward the methods used to obtain a solution to the hydrodynamic equations.
NUREQ/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Users ' Manual. THURODOD,M.J.s CUTA J. M. s KDONTZ,A.S.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
319pp.
8304120245.
PNL-4305.
17943:001.
The COBRA / TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients.
The code solves the compressible three-dimensional, two-fluid, three-field equations for two phase flow in the reactor vessel.
The three fields are the vapor field, the continuous liquid field, and the liquid drop field.
A five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, and accumulators.
The heat generation rate of the core is specified by input and no reactor kinetics calculations are included in the solution.
This volume is the User 's Manual.
It contains the input instructions for COBRA / TRAC and its auxiliary programs, SPECSET and ORAFIX.
It also includes a users' guide to the code and is intended to aid the new user in becoming familiar with the capabilities and modeling conventions of the code.
NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMO. Volume 4: Developmental Assessment And Data. THURGOOD M.J.s GUIDOTTI,T.E.s SLY,G.A.s et al.
Battelle Memorial Institute, Pacific Northwest Lchoratory.
March 1983.
221pp.
8304130310.
PNL-4385.
17959:001.
The COBRA / TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients.
The code solves the compressible three-dimensional, two-fluid, three-field equations for two-phase flow in the reactor vessel.
The three fields are the vapor field, the continuous liquid field, and the liquid drop field.
A five-equation drift flux model is used to modelfluid flow in the primary system piping, pressurizer, pumps, and accumulators.
The heat generation rate of the core is specified by input and no reactor kinetics l
l calculations are included in the solution.
This volume documents the major data comparisons made with COBRA / TRAC during the process of code j
58 j
I I
development.
These data comparisons were extremely useful in detecting programming errors and defining deficiencies in the code's physical models.
The data comparisons presented in this volume document the results obtained on developmental ve'rsions of the code.
A separate document will be released at a later date containing data comparisons run on the final released version of the code.
l j
NUREC/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Programmers Manual. KOONTZ, A. S. s CUTA,J.M.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
302pp.
8304060033.
PNL-4385.
17864:315.
The COBRA / TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients.
The code solves the compressible t
three-dimensional, two-fluid, three-field equations for two-phase flow in the reactor vessel.
The three fields are the vapor field, the continuous liquid field, and the liquid drop field.
A five-equation drift flux model is used to model fluid flow in the primary system piping, pressuirizer, pumps, and accumulators.
The heat generation rate of the core is specified by input and no reactor kinetics calculations are included in the solution.
This volume explains the
(
details of COBRA / TRAC's working parts from the programmer's viewpoint.
The code's overlag structure is discussed.
The memory management and COMMON block manipulation are explained, as are the =estart/ dump logic and the graphics logic.
Suggestions for code conversion to "non-LANL" CDC computers and non-CDC computers are given.
i NUREQ/CR-3059: PARAMETRIC CALCULATIONS OF FATIQUE CRACK QRDWTH IN PIPING. SIMONEN,F.A.s GOODRICH,C.W.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
42pp.
8304050624.
P NL-4537.
17B47: 175.
This study presents calculations of the growth of piping flaws produced by fatigue.
Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors.
The results indicate that the present flaw acceptance standards of ASME Section XI provide a relatively consistent set of allowable flaw sizes because the predicted life of flawed piping is relatively insensitive to pipe wall l
thickness, flaw aspect ratio, and piping material (ferritic versus l
austenitic).
On the other hand, the results show that flaws that are I
acceptable under ASME Section XI can grow at unacceptable rates if the cyclic stresses are at the maximum level permitted by the design rules of ASME Section III.
However, a review of the conservatisms inherent to the ASME code rules is presented to explain the low occurrence of piping. fatigue failures in service.
It is concluded that decreases in the allowable flaw sizes are not justified.
NUREG/CR-3062: STATUS OF QE0 CHEMICAL PROBLEMS RELATING TO THE BURIAL OF HIGH-LEVEL RADIDACTIVE WASTE,1982. APPS, J. A. s CARNAHAN,C.L.4 LICHTNER,P.C.s et al.
California, Univ. o f, Berkeley.
March 1983.
3 bop p.
8304200515.
LBL-15103.
19074:025.
Geochemistry research supporting high-level radioactive waste burial in underground repositories is evaluated.
General research covering the whole repository includes the bounding of physical nd i
chemical conditions in the repository environment, estimation of toxic 50
radionuclides present in spent fuel and high-level waste at various times after 1,000 years, the forms in which radionuclides e.ight be transported in groundwater, the mechanisms by which radionuclides are retarded, and transport models incorporating chemical reactions.
Specific issues relating to the backfil, near field, and far field are also examined.
Each examination results in a series of conclusions and recommendations.
The recommendations are prioritized, and further research needed to resolve current uncertainties is j
suggested.
NUREQ/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
WYANT,F.J.; CROPP L.O.4 ARONSON,E.A.s et al.
Sandia Laboratories.
l March 1983.
88p.
8304060021.
SANDB2-2730.
17864:157.
This report summarizes Sandia National Laboratories' efforts in
}
assessing nuclear power plant electrical system performance.
Initial j
scoping study results, including an LER search and a survey of PES analysis techniques, are discussed.
Descriptions of the limits of the proposed analysis and tasks, broken down by activities, are given.
A number of large network codes were evaluated, and a summary of the characteristics and suitability of each code is presented.
Results of test runs made on two of the codes (EMTP and SCEPTRE) are discussed.
Recommendations for additional study of nuclear power plant electrical systems are presented.
l NUREQ/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HICH LEVEL NUCLEAR WASTE (HLW) DEEP CEOLOCIC REPOSITORIES. Main Rep ort. ROBERDS,W.s BAUHOF,F.; CONANO L.s et al.
Golder Associates.
March 1983.
219pp.
8303280114.
813-1163.
17712:021.
The purpose of the complete project is to provide the NRC with technical assistance to enable the focused, adequate review of NRC of specific aspects related to design and construction of an in situ test facility and final geologic repository, as presented by DOE in Site Characterization Reports and License Applications (LA).
This report includes the general recommendation of available tests which should be considered in designing media / site specific in situ test programs.
Tests will be conducted in an exploratory shaft and an underground test facility at the prospective repository horizon.
Testing plans will be presented by DOE in an SCR and the complete results presented in the LA.
A licensing perspective is outlined and a defensible rationale developed and utilized for the test selection process.
This rationale consists of matching the capabilities of tests to the perceived information needs for LA.
The information needs are a function of significant engineered components, site characteristics, available information and the acceptable level of confidence in satisfactory performance for each licensing step.
j Tests which are available and respond to the perceived media / site j
specific information needs, either by simulation or assessment of the characteristics, are identified and their capabilities assessed.
Specific in situ are investigated and described in detail.
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HICH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC R EPOSITORIES. Ap p en d i c e s.
ROBERDS,W.s BAUHOF,F.s GONAND,L.s et al.
Golder Associates.
March 1983.
694pp.
8303280097.
813-1163.
17714:001.
The purpose of the complete project is to provide the NRC with l
80
technical assistance to enable the focused, adequate review by NRC of specific aspects related to design and construction of an in situ test facility and final geologic respository, as presented by DOE in Site i
Characterization Reports and License Applications (LA).
This report includes the general recommendation of available tests which should be considered in designing media / site specific in situ test programs.
Tests will conducted in an exploratory shaft and an underground tast facility at the prospective repository horizon.
Testing plans will be presented by DOE in an SCR and the complete results presented in the LA.
A licensing perspective is outlined and a defensible rationale developed and utilized for the test selection process.
This rationale consists of matching the capabilities of tests to the perceived information needs for LA.
The information r
needs are a function of significant engineered components, site characteristics, available information and the acceptable level of confidence in satisfactory performance for each licensing step.
Tests which are available and respond to the perceived media / site specific information needs, either by simulation or assessment of the characteristics, are identified and their capabilities assessed.
Specific in situ tests are investigated and described in detail.
NUREC/CR-3069 VO1: INTERACTION OF ELECTRDMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary. ERICSDN,D.M.
Sandia Laboratories.
STRAWE,D.F.s SANDBTRG,S.J.7 et al.
Boeing Co.
February 1983.
26pp.
8303100558.
SANDB2-2738/1.
17500:339.
This study examines the interaction of the electromagnetic pulse from a high altitude nuclear burst with commercial nuclear power plant systems.
The potential vulnerability of systems required for safe shutdown of a specific nuclear power plant are explored.
EMP signal coupling, induced plant response and component damage thresholds are established using techniques developed over several decades under Defense Nuclear Agency sponsorship.
A limited test program was j
conducted to verify the coupling analysis technique as applied to a i
nuclear power plant.
Based upon the analysis, it was concluded that:
(1) Diffuse fields inside a Seismic Class I buildings are negligibles (2) EMP signal entry points are identifiables (3) Interior signal attenuation can be reasonably modeleds (4) Damage thresholds, even for equipment containing solid state components are highs (5) EMP induced signals at the critical equipment in the example plant are much less than nominal operating levels, but plant topology and cabling practice have a strong influence on responsess (6) The likelihood that individual components examined will fail is smalls therefore, it is unlikely that an EMP event would fail sufficient equipment so as to prevent safe shutdown.
i l
NUREG/CR-3069 VO2: INTERACTION DF ELECTRDMAGNETIC PULSE WITH COMMERCI AL l
NUCLEAR POWER PLANT SYSTEMS. Main Report. ERICSON,D.M.
Sandia l
Laboratories.
STRAWE. D. F. s SANDBERG, S. J. s et al.
Boeing Co.
February 1983.
421pp.
8303100460.
SAND 82-2738/2.
17501:149.
This study examines the interaction of the electromagnetic pulse from a high altitude nuclear burst with commercial nuclear power plant systems.
The potential vulnerability o8 systems required for safe shutdown of a specific nuclear power plant are explored.
EMP signal coupling, induced plant response and component damage thresholds are established using techniques developed over several decades under Defense Nuclear Agency sponsorship.
A limited test program was conducted to verify the coupling analysis technique as applied to a nuclear power plant.
Based upon the analysis, it was concluded that:
61 J
.._ -.-. _ __-. - _ ~,
l l
l 1
(1) Diffuse fields inside a Seismic Class I buildings are negligibles (2) EMP signal entry points are identifiables (3) Interior signal attenuation can be reasonably modeleds (4) Damage thresholds, even for equipment containing solid state components are highs (5) EMP induced signals at the critical equipment in the example plant are much less than nominal operating levels, but plant topology and cabling practice have a strong influence on responsess (6) The likelihood that individual components examined will fail is smalls therefore, it is unlikely that an EMP event would fail sufficient equipment so as to prevent safe shutdown.
NURE0/CR-3076: COMPUTER PREDICTION OF SUBSURFACE RADIONUCLIDE TRANSPORT 1
-AN ADAPTIVE NUMERICAL METHOD. NEUMAN,S.P.
Arizona, Univ. o f.
{
l January 1983.
43pp.
8302250151.
17299:130.
i Radionuclide transport in the subsurface is often modeled with the aid of the advection-dispersion equation.
A review of existing computer methods for the solution of this. equation shows that there is need for improvement.
To answer this need, a new adaptive numerical method is proposed based on on Eulerian-Lagrangian formulation.
The method is based on a decomposition of the concentration field into two parts, one advective and one dispersive, in a rigorous manner that does not leave room for ambiguity.
The advective component of steep concentration fronts is tracked forward with the aid of moving particles clustered around each front.
Away from such fronts the advection problem is handled by an efficient modified method of characteristics called single-step reverse particle tracking.
When a front dissipates with time, its forward tracking stops automa tically and the corresponding cloud of particles is eliminated.
The dispersion problem is solved by an unconventional Lagrangian finite element formulation on a fixed grid which involves only symmetric and diagonal matrices.
Preliminary tests against analytical solutions of one-and two-dimensional dispersion in a uniform steady state velocity field suggest that the proposed adaptive method can handle the entire range of Peclet numbers from 0 to infinity, with Courant numbers well in excess of 1.
NUR EQ/CR-3078: MODEL EVALUATION OF SEEPACE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE. NELSON,R.W.s MEYER,P.R.s OBERLANDER,P.L.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
140pp.
8304130008.
17961:001.
l Model simulations identify the rate and amount of leachate released to the environment if disposed uranium mill tailings come 1
into contact with ground water or if seepage from tailings reaches ground water.
In this study, simulations o* disposal above and below 4
the water table, with various methods of leachate control, were compared.
Three leachate control methods were used in the I
comparisons: clay bottom linersa stub-sidewall clay linerss and tailing drains with sumps, with the below the water table is combination of the three methods.
The combined methods intercept up to 80% of the leachate volume in pits above the water table and intercept essentially all of the leachate in pits below the water table.
Effluent pumping, however, requires continuous energy costs and an alternative method of disposal for the leachate that cannot be reused as makeup water in the mill process.
Without the drains or offluent pumping, the clay bottom liners have little advantage in terms of the total volume of leachate lost.
The clag liners reduce the rate of leachate flow to the ground water, but the flow continues for a longer time.
The buffering, sorption, and chemical reactions of 62
1 the leachate passing direct 1g through the liner are also advantages of the liner.
i NUREQ/CR-3079: EARTHGUAKE HAZARD STUDIES IN NEW YORK STATE AND ADJACENT
]
AREAS. Final Report. April 1976 - June 1982. KUFKA.A.L.
Columbia Univ.
January 1983.
70pp.
8302230025.
17241:225.
Lamont-Doherty Geological Observatory (LDOO) current 1g operates a network of 38 short period seismic stations in the states of New York, i
New Jersey and Vermont.
It is part of the larger Northeastern United States Seismic Network (NEUSSN) operated by several university groups in New York, New Jersey, Pennsylvania and New England.
These networks provide a wealth of data to study seismicity, earthquake hazards, earthquake source properties, tectonic processes, and crustal and upper mantle structure in the northeastern United States and adjacent parts of Canada.
The LDGO network provides data for more spe:ific studies of earthquake processes in New York State and adjacent areas.
The operation and maintenance of the LDOO network has been supported primarily by funds from the United States Geological Survey (USGS), the United States Nuclear Regulatory Commission (NRC), and the New York State Energy Research and Development Authority (NYSERDA).
This report discusses results of research related to the operation of the network during Phase I through Phase VII of our contract with NYSERDA, and diso introduces current directions of research for future studies.
4 NUR EQ/CR-3000: NETWORK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA. SIBOL,M.S.s BOLLINGER.C.A.
Virginia Polytechnic Institute.
January 1983.
80pp.
8302170251.
17196:011.
Twenty-four blasts from three guarries operating in the central Virginia area were used, first to test the locational capabilities of the Central Virginia - North Anna Network and then to generate i
relative station delag suites for network stations.
i Using two different methods of approximating blast origin times, i
the Closest Station Method (CSM) and the Single Iteration Method (SIM), station delays were derived for different areas within central Virginia.
Application of these station delay suites reduced locational errors in the general area from an average of 3.0 plus or minus 1.2 to 1.7 plus or minus 0.6 km (95% confidence level).
In both cases, the average equivalent radii, a linear measure of error ellipse size, were 1.3 km.
However, this result depends primarily on the improvement at one of the three quarries, where the locational error was reduced from 6.5 km to 2.6 km.
Utilizing one of these methods (the SIM), lateral variational
[
patterns in velocity were inferred and determined to be velocity banding similar to that observed in the Piedmont province in Georgia, j
1 North and South Carolina.
l NUREQ/CR-3082: PROBABILISTIC APPRDACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS. LOFOREN. E. V. s VARCOLIK,F.
Science Applications. Inc.
- Brookhaven National Laboratory.
February 1983.
195pp.
8303230282.
BNL-NUREG-51628.
17685:180.
l This final report presents results of a comprehensive analysis of risk-based methods for establishing Limiting Conditions for Operation (LCO) and surveillance requirements for on-line test and repair of nuclear power plant safety system components.
Limiting Conditions for Operation refers to the legal constraint on safety system component outage times that are imposed by the NRC as part of the reactor l
N
a operating license.
Generally, when a safety system component is removed for repair or test for a period of time there is a period of increased vulnerability concerning the probability that the affected safety system will be available to mitigate an accident.
This period of increased vulnerability exists until the component is restored to service.
The constraint on the duration of this period, the allowed outage time (AOT), is the aspect of LCOs that is of interest here.
In particular, methods are reviewed and developed that relate measures of risk to the AOT.
Chapter 2.0 describes the review and analysis of j
risk related methods for establishing LCOs.
Chapter 3.0 describes the analysis of the relationship between periodic testing requirements and risk, as measured by system everage unavailability.
1 NUREQ/CR-3084: LOW-LEVEL NUCLEAR WASTE SHALLOW LAND BURIAL TRENCH I
I SOLATION. MCCRAY,J.G.s NOWATZKI,E.A.s 1 HOMPSON, 0. M. s et al.
Arizona, Univ. o f.
March 1983.
136pp.
8304200611.
18061:118.
The scope of this work covers an evaluation of trench cap designs, trench construction and trench Icading artifically speeding up the creation of void space in the trenches and monitoring structural and hydraulic effects under these conoitions.
Eight trenches were constructed and instrumented, four in a semi-arid region and four in a more humid mountainous region.
The two main concepts being tested are the " soil arch" and the " soil beam."
i Additionally, it is expected that methods will be developed for optimization of monitoring of sites as well as improved predictions of the relative long-term stability of the tested designs.
I The results to date indicate that the " soil arch" are viable l
concepts that can be utilized in the design of shallow-land burial l
trenches.
The presence of geotentiles reinforcement does diminish the deleterious effects of moisture infiltration into the waste area.
However, there are some constraints that may limit these applications to full scale, commercial operations.
These constraints are:
a.
Need for moisture and soil match between trench sidewalls and backfill.
b.
The maximum depth to width ratio of the backfill cap is dependent upon soil properties along the trench wall-cap interface.
c.
The soil beam should be sufficiently rigid not to slide off its ends supports.
NURE3/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING RESPONSES. WANG,Y.K.s SUBUDHIIM.s BEZLER P.
Brookhaven National Laboratory.
January 1983.
84pp.
8302220431.
BNL-NUREG-51629.
17237:231.
This report includes the findings of an investigation of the l
conservatism associated with different combinations between the l
primary and secor.dary stress components for piping systems under l
dynamic loading, such as in an earthquake event.
The primary stresses are induced by piping response to its mass inertia effects.
The secondary stresres are induced by relative displacements of piping supports.
It is found that the SRSS combination of the primary and secondary stress components yield acceptable results provided the secondary stress components are calculeted in the most unfavorable phasing relationship among displacements of piping supports.
The absolute sua combination as recommended in the current Standard Review Plan is found to yield very conservative results when compared to the time history solutions.
l l
64 l
NUREC/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR i
ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION. BEARE,A.N.s DORRIS,R.E.s KOZINSKY,E.J.s et al.
General Physics Corp.
February 1983.
61pp.
8303230142.
ORNL/TM-8599.
17689:154.
This report presents preliminary comparisons of field and simulator operator performance data collected in an NRC-funded research program directed by Oak Ridge National Laboratory.
The i
primary objective of the program is to develop an empirical data base on operator performance to support development of criteria and standards for safety-related operator actions.
The comparison of simulator and field data is intended to provide a " calibration" of simulator results so that they can be more confidently extrapolated to field conditions.
Collection of PWR/BWR field data was performed by the Memphis State University Center for Nuclear Studies.
The collection of PWR/BWR simulator data was performed by General Corporation, using the Electric Power Research Institute's Performance Measurement System.
The performance measure used in this study was the time required for operators to initiate the first correct manual action in response to an abnormal or emergency event.
Techniques for i
data collection as well as the problems and limitations of field data are reported, along with the initial results.
NUR EC/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED GROUND MOTIONS. APSEL,R.J.s HADLEY,M.s H ART, R. S.
Lawrence Livermore Laboratory.
March 1983.
139pp.
8304050635.
UCRL-15495.
17848:010.
This report summarizes an extensive simulation study of 5460 components of ground motion representing a model parameter study for magnitude, distance, source orientation, source depth and near-surface site conditions for a generic EUS crustal model.
The simulation methodology represents a hybrid approach to modeling strong ground motion.
Wave propagation is modeled with an efficient frequency-wavenumber integration algorithm.
The source time function used for each grid element of a modeled fault is empirical, scaled from near-field accelerograms.
This study finds that each model i
parameters ha, a significant influence on both the shape and amplitude of the simulated response spectra.
The combined effect of all l
parameters predicts a dispersion of response spectral values that is j
consistent with strong ground motion observations.
This study provides guidelines for scaling WUS data from shallow earthquakes to the source depth conditions more typical in the EUS.
The modeled site j
L conditions range from very soft soil to hard rock.
To the extent that these general site conditions model a specific site, the simulated response spectral information can be used to either correct spectra to site-specific environment or used to compare expected ground motions a
at different sites.
NUREC/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL. VENHUIZEN,J.R.s GRIFFITH,J.M.
EG&G, Inc.
February 1983.
19pp.
8303140628.
EGG-2235.
17576: 0B0.
We introduce the concept of predictor displays for use in a nuclear power plant control environment.
[A predictor display is defined as a display on a cathode ray tube (CRT) givin 6 the status of a portion of the plant at some future time as predicted by computer simulation.3 Properties of the predictor are given, and direct relationships between supervisory control, optimal estimation theory, and prediction are presented.
Two examples of predictor displays are included:
steam generator water level and primary system coolant l
E
inventory.
We recommended additional research to determine the ultimate value of predictor displays in a nuclear power plant control environment.
NUREG/CR-3105: SECURORS APPLICATION TO A GENERIC NUCLEAR POWER PLANT.
ROUNDTREE, S. L.
Sandia Laboratories.
February 1983.
52pp.
8303230413.
SANDB2-2821.
17674:186.
The Security Officer Response Strategies (SECURORS) technique is i
applied to a nine-level generic nuclear power plant to determine security officer deployment locations within the facility subsequent to detection of adversary intrusion.
Extensive use has been made of the facility layout drawings, facility region information, and pathfinder critical path results readily available from the Safeguards Automated Facility Evaluation (SAFE) analysis of the nine-level facility.
An additional pathfinder analysis, using both the deterministic and the stochastic pathfinder vsutines from SAFE, has been conducted and the results compared to the SECURORS results.
NUREC/CR-3106: COMPARISON OF-FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS. YOUNG.J.A.s l
JACKSON,P.O.s THOMAS,V.W.
Battelle Memorial Institute, Pacific Northwest Laboratory.
January 1983.
23pp.
8302250046.
PNL-4591.
17299:316.
i Pacific Northwest Laboratory (PNL) has made side-by-side five-minute air filter radon daughter measurements, 100-hour RPISU radon daughter measurements, and Terradex Track Etch radon i
measurements in buildings in Edgemont, South Dakota.
The standard deviation of the difference between the (natural) logarithms of the RPISU average annual radon daughter concentrations and the logarithms of single air filter measurements (S.D.-In) was found to be 0.52.
Usiveg the S.D.-In of 0.52 it can be calculated that there would onig be a 5% probability in Edgemont that the RPISU annual average would be greater than 0.015 WL if the five-minute measurement were less than O.010 WL was reasonable.
Comparison of the S.D.
-in 's between the RPISU annual averages and indivsdual RPISU measurements in Edgemont indicates that a single RPISU measurement taken in the months of September threugh April would provide a somewhat better estimate of l
the annual average of two RPISU measurements taken six months apart and would provide a considerably better estimate of the annual average.
It also appears that a two-month Track Etch measurement l
would provids e considerably poorer estimate of the annual average than would a five-minute measurement, but that a six-month Tract Etch measurement or a one year Track Etch measurement would provide a better estimate of the annual average than would an air filter i
measurement.
NUR EC/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS. WILSON,W.B.s ENGLAND,T.R.; LABAUVE.R.J.
Los Alamos Scientific Laboratory.
February 1983.
96pp.
8302280013.
I LA-9563-MS.
17314:136.
A set of radionuclide inventories that bounds the equilibrium cycle inventories expected (without anticipated irregularities in fuel management) for various enrichments, reactor types, and burnup have been calculated.
Limiting inventory values applicable to a range of operating conditions do not occur under a single set of operating conditions.
Eight actinide and 48 fission product nuclides, which 66
have been identified as important to redirlogical consequences of reactor accidents, have been considered.
Fasic nuclear properties from ENDF/B-V have been processed and-combined with specific reactor engineering parameters in EPRI-CELL code calculations.
The' temporal flux and cross section data thus generated were combined with nuclear property libraries in CINDER-2 nuclide inventory code calculacions.
Comparisons with DRIGEN have been~made for a limited subset of the reactor conditions.
NUREQ/CR-3113: A TORSIONAL ULTRASONIC TECHNIQUE FOR LWR LIGUID LEVEL MEA'SUREMENT. DRESS.W.B.
Oak Ridge National Laboratory.
February 1983.
13pp.
8303220453.
ORNL/TM-8505.
17670:096.
In the late 1970's, a technique for determining the mean density of fluid surrounding a waveguide of non-circular cross section was developed using slow torsional ultrasonic pulses.
With the waveguide j
partially inserted in a liquid of known density, the fraction of the probe immersed in the liquid can be derived, and hence the level.
Since 1980, the Instrumentation and Controls Division of DRNL has been involved in the evaluation of sensors using torsional ultrasonic pulses to make level and density measurements, and extensional ultrasonic pulses for temperature measurements and correction of the torsional measurments.
Work at DRNL through 1981 demonstrated the feasibility of transmitting torsional pulses in a ribbon of stainless steel and the sensor was used to measure the level of a steam / water interface at 290 degrees centigrade and 10 MPa.
Subsequent work under the Advanced Two-Phase Instrumentation Program has been directed towards an operational prototype instrument that can be, installed in functional power reactor.
NUR EQ/CR-3114: PROCEEDINGS OF WORKSHOP DN CDONITIVE MODELING OF NUCLEAR PLANT CONTROL ROOM DPERATORS. August 15-18,1982,Dedham, Massachusetts.
ABBOTT,L.S.
Oak Ridge National Laboratory.
February 1983.,
213pp.
8303230467.
DRNL/TM-8614.
17683:225.
I This document presents 11 invited papers and the deliberations of four working groups at a Workshop on Cognitive Modeling of Nuclear Plant Control Room Operators that was held in Dedham, Massachusetts, under the sponsorship of the U.S.
Nuclear Regulatory Commission.
The purpose of the workshop was to review the status of " cognitive modeling and to recommend to the NRC whether it should support research directed toward the development of a' cognitive model of a reactor oper3 tor that could be useful by itself or as a part of a larger model of the human-machine system.
It was the consensus of the invited papers ano the working groups that some cognitive models developed for other types of systems can be adapted to the reactor l
operator under limited and precisely defined, conditions (and, indeed, some already are being used).
However, the development of a l
comprehensive model for the reacter operator should be preceded by an i
improved understanding of the task.
In the meantime, the need for further applied research in operator cognition is apparent, as isi the need for supporting data collection.
Work in these areas could begin immediately.
~
NUREQ/CR-3115: EXPERT DPINION AND RANKING METHODS. UPPULURI.V.R.R. 'Dak Ridge National Laboratory.
March 1983.
23pp.
8303290683.
ORNL/CSD/TM-201.
17742:244.
Suppose we have k ob Jects and wish to rank them according to a characteristic.
A Judge compares these objects two at a time and l
I I
67 l
I
indicates whether one object is better or worse or equal to another object.
We analyze this data by log least squares procedure and the
()embda man) procedure suggested by Saaty.
We show by examples, how onte can incorporate the data from several Judges to obtain a ranking of the objects.
Finally, we consider the case when a Judge compares j
the objects two at a time and indicates the probability that one object is better than another object, and show how one can get a ranking of the objects.
NUREQ/CR-3116: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Project Summary Report,Enrico Fermi-1 Reactor. MILLER R.L.s LINK,B.W.
United
' Nuclear Corp.
February 1983.
28pp.
8303220413, 17667:142.
This report summarizes information concerning the decommissioning l
of the Enrico Fermi-1 reactor.
Decommissioning data from available dccuments and other decommissioning records were input into a computerized data-handling system in a manner that permits specific information toibe readily retrieved.
The information is presented both in detail:in its computer output form, and as an assembled summarization, generated for the purpose of highlighting the more important aspects of the escommissioning program.
NUREG/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. Annual, Report,For FY 1992. EISENHOWER,E.H.s EHRLICH,M.s SOARES. C. s et al.
Commerce,' Dept. of National Bureau of Standards.
February 1983.
73pp.
8303100424.
17503:178.
This report describes results of the second year of a program that will enable the Nuclear Regulatory Commission to improve, p
[
demonstrate, and document traceability of its measurements to the l
national physical measurement standards for ionizing radiation.
The l
principal actioni being taken are:
(1) characterization of the response of thermoluminescence dosimetry systems used for routine surveillance of nuclear facilitiess (2) type testing and characterization.of, portable survey instrumentss and (3) implementation of routine quality assurance services that will demonstrate that regulatory measurements are sufficient 1g consistent (in agreement) with national measurement standards.
During the year, some tests of the TLD system were performed as specified in American Netional StandardcN545-1975, specifically self-irradiation, directional dependence, humidity dependence at extreme temperatures, and salt spray.
Measurement assurance tests were conducted for the NRC Region-1 laboratory.
The response of six portable surveg l
instruments to. beta radiation from menon-133 was determined using a specially-build chamber for exposure to gaseous sources.
The basic principles under-which the long-range interactive MGA program will operate have been developed'and documented, along with procedures for obtaining essential information from participating laboratories.
l NUREC/CR-3126: SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTS COMPARISDN. BOUCHER,T.J.I DIMENNA,R.A.
EGarc Inc.
February
~ 1983.
53pp.
8303150614.
EQG-2238.
17584:176.
Results are presented from an analysis of Semiscale Mod-2A i
Intermediate Break, Tests S-IB-1,
-2, and
-3.
The tests were 100%
l
~
(percentage of cold leg ripe flow area), 50%, and 21.7%, respectively, communicative cold leg break loss-of-coolant experiments.
They were intended to provide reference data for evaluation and assessment of treactor safety code capabilities to predict integral blowdown, refill /reflood experiments for intermediate break sizes, and.
OB
-r
---+-o
l s
particular1g, to~ provide data to extend the code into the reflood regime.
Comparisons of Eweiscale intermediate break test results with those from large and small break. tents provided characterization of the phenomena observed duringuthe intermediate break tests.
An 1 additional objective of: Test S-IB-3 was to provide reference data for comparison of Semiscale test results with results from LCFT Test L5-1 and LDBI Test B-RIMi NUREQ/CR-3130: INFLUENCE DF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MIGRATION. OZTUNALI,0.I.s PJTT,C.J.s FURFARD.J.P.
Dames
& Moore.
February 1983.
300pp.
8303150031.
17597:181.
A sensitivity analysis is performed in which the, leaching characteristics of selrcted. low-level radioactive waste streams are j
varied, and the result'ing potential groundwater impacts calculated, as a function of alternative disposal site environments and alternative facility designs and operating practices.
The analysis also assesses whether a leaching standard for solidified waste streams is needed to ensure safe low-level waste disposal.. In the analysis, the leach rate of particular waste streams suitable for solidification (LWR concentrated liquids. Class B and C LWR ion-exchange resins and filter sludge) is varied, while the leach rate of other waste streams unsuitable for solidification is held constant.
The results indicate that measures which reduce percolation into disposal cells and reduce contact of waste by water are less expensive and more effective in reducing groundwater impacts than large reductions in the leach rates of the waste streams suitable for solidification.
Thus, a;1each criteria which can be met by existing binders such as cement, bitumen, vinyl ester styrene, or synthetic polymers would probably be acceptable provided that the solidified product is structurally stable.
The analysis also indicates that the presence of small guantities of free standing liquids within some of the waste containers would probably not result in significant additional groundwater impacts.
l NUREC/CR-3135: BUCKLING INVESTICATION DF RING-STIFFENED CYLINDRICAL.
SHELLS WITH REINFORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
BAKER W.s BENNETT,J.
Los Alamos Scientific Laboratory.
March 1983.
A9pp.
8304060028.
LA-9646-MS. -17863:275.
Four steel shells having features representative of steel containment vessels for nuclear power plants were fabricated and
, tested to failure under unsymmetrical axial loading.
All of the ring-stiffened shells were 698 mm.(27.5 in.) in diam by 0.762 mm I
(0.030.in.) in wall thickness.
Each one had a penetration that was reinsforced in accordance with the area-replacement rule of the applicable American, Society of Mechanical Engineers (ASME) code and of a design to simulate actual practice for steel containments.
The penetrations were of four-different; diameters, cutting no ring stiffeners, and cutting one, two, and three ring stiffeners.
Before
. testing, imperfections were measured, and. strain gages were applied to i
characterize the strain field.at an end and around the penetration.
~
Buckling loads were determined with application parallel to the exis at an eccentricity of R/ ware compared with the result's from a numarical solution.
l NUREG/CR-3141: CHARACTERIZATION OF TWO-PHASE FLDW USING NEUTRONIC FLUCTUATIONS. ALBRECHT,R.E.s CROWE,R.D.s DAILEY, D. J. s et al.
Washington, Univ. o f.
Ma ch 1983.
328pp.
8304200567.
19077:212.
89
_,__m y
Results of an analytical and esperimental investigation are presented which demonstrats that neutron fluctuations measurements can be used to yield quantitative information on: void fractions, phase velocities, and flow regimes which characterine the state of a two-phase mixture.
These results have opened the possibility of using existing instrumentation in BWR plants to obtain, in situ, data on these two-phase flow parameters, that is, on void fraction, 11guld and vapor velocities and flow regimes.
N'JREO/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIGUES FOR LOW-LEVEL RADIDACTIVE WASTE DISPOSAL. TUCKER.P.O.
Army, Dept. o f, Army Engineer Waterways Esperiment Station.
February 1983.
221pp.
8303220432.
17667:250.
With federal regulations requiring that all states (whether on an individual basis or through a compact of states ) have a designated low-level radioactive waste (LLW) disposal site by 1986 and in the light of the U.S.
Nuclear Regulatory Commission's rule 10 CFR Part 61,
" Licensing Requirements for Land Disposal of Radioactive Waste " there exists a need for providing technical guidance to the federal, state, and private sectors involved with this type of disposal operation.
This document provides information on trench design and construction techniques which can be used in the disposal of LLW by shallow land burial.
" Trench Design and Construction Techniques for Low-Level 4
Radioactive Waste Disposal" covers practices currently in use not only in the LLW disposal, field, but also methods and materials being used in areas of hazards and municipal waste disposal which are compatible with the performance objectives of 10 CFR Part 61.
Information is provided on the properties, selection, and installation of various materials such as bentonite, soil-coment, polymeric materials, asphaltic materials, and geotechnical fabrics.
This is not intended to outline step-by-step procedures but rather to serve a guide to acquaint the reader with some of the methods and materials available.
NUREQ/CR-3145 VO1: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHID-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
CHRISTENSEN, D. s WIEDENBECK.M.G.s JACKSON,P.L.
Michigan, Univ. o f.
February 1983.
64pp.
8303220407.
17666:291.
Earthquake activity in the Western Ohio-Indiana region is monitored with a precision seismograph network:
Nine stations clustered about Anna, Ohio, and four stations in Indiana centered about Indianapolis.
During this period no earthquakes were detected in the region of the network.
This seismic inactivity contrasts with the previous year, during which nine small (unfelt) earthquakes occurred within and adjacent to the Western Ohio portion of the l
network.
Digital triggering and recording capability was added to the analog recording capability.
Using the new digital capability, we were able to more precisely locate the focus of some of these small earthquakes.
As part of the inception of a modeling investigation, a preliminary location of the Grenville Front was populated as eastward l
of the network.
Azimuthal changes in travel time residuals of teleseises were LO*d for this estimate.
l NUREO/CR-3149: DISPERSION COEFFICIENTS FOR COASTAL REGIONS.
MACRAE, B. L. s KALEEL.R.J.s SHEARER,D.L.s et al.
TRC Environmental l
Consultants, Inc.
March 1993.
52pp.
8304200604.
PNL-4627.
19081:001.
l 70
i The work reported here was directed toward obtaining and presenting atmospheric dispersion coefficients representative of coastal regions from previous 1g conducted field measurement programs.
The results will assist in minimizing the need for future measurement efforts and should improve those efforts still required.
The data i
base is comprised of 495 tracer experiments from 9 field programs.
A compendium of dispersion coefficients is presented and conditions for daylight hours and continuous emissions for a varfety of terrain conditions are well represented.
Recommendations for needed research relative to poorig represented conditions are given.
These include measurements of vertical dispersion. instantaneous point source emissions dispersion for a variety of terrain, dispersion in stable atmospheres and in nonsteady-state conditions, and dispersion at distances less than one kilometer and greater than five kilometers.
It is further recommended that standardization of measurement and classification techniques be employed in future measurement programs for the important reason of intercomparability.
The report complements previous 1g-issued NUREQ/CR-2754, PNL-4292, Critical Review of Studies on Atmospheric Dispersion in Coastal Regions, published in June 1982.
NURE0/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION. CALVD, J. s BUYUK0ZTURK,0.s C ONNOR. J.
Massachusetts Institute of Technology.
March 1983.
228p p.
8304060077.
17868:131.
Behavioral models are developed for reinforced concrete containment wall panels having orthogonal reinforcement and for those having both orthogonal and diagonal reinforcement.
An initial axial tension which causes cracking in the concrete, is applied to the orthogonal reinforcement to simulate the normal membrane stresses induced by an accidental internal pressure load.
This load is followed by the application of cyclic membrane shear histories representing the tangential shear stresses generated in the event of l
an earthquake.
Using these two models, the experimentally measured strengths of various specimens are evaluated in view of important i
behavioral parameters.
Then, an equation in terms of the governing parameters is proposed to predict the ultimate strength of these panels.
Finally, a practical design procedure is developed and recommended as a replacement for the design criteria specified by the current code.
Applications of the proposed design procedure are shown and the results obtained are evaluateC NUREO/CR-3166: RECOMMENDED PROCEDURES FOR MEASURING RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS. YOUNG,J.A.s THOMAS,V.W.s JACKSON,P.O.
Battelle Memorial Institute, Pacific Northwest Laboratory.
March 1983.
31pp.
8304010671.
PNL-4597.
17824:006.
This report recommends instrumentation and methods suitable for measuring radon fluxes emanating from covered disposal sites of residual radioactive materials such as uranium tailings.
Problems of spatial and temporal variations in radon flux are discussed and the advantages and disadvantages of several instruments are examined.
A year-long measurement program and a two month measurement methodology are then presented based on the inherent difficulties of measuring average redon flux over a cover using the recommended. instrumentation.
l 71
_ ~.
i i
a i
NURE0/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS. GRUSZCZYNSKI,M.s VISKANTA,R.
Argonne National Laboratory.
March 1983.
57pp.
8304060417.
ANL-83-7.
17879:300.
The natural circulation heat transfer data for longitudinal flow of water outside a vertical rod bundle are needed for developing correlations -hich can be used in best estimate computer codes to w
model thermal-hydraulic behavior of nuclear reactor cures under l
accident or shutdown conditions.
The heat transfer coefficient between the fuel rod surface and the coolant is the key parameter required to predict the fuel temperature.
Because of the absence of the required heat transfer coef*1cient data base under natural circulatien conditions, experiments have been performed in a natural circulation loop.
A seven-tube bundle having a pitch-to-diameter ratio of 1.25 was used as a test heat exchanger.
A circulation flow was established in the loop, because of bougancy differences between its two vertical legs.
Steady-state and transient heat transfer measurements have been made over as wide a range of thermal conditions as possible with the system.
Steady state heat transfer data were correlated in terms of relevant dimensionless parameters.
Empirical correlations for the average Nusselt number, in terms of Reynolds number, Rayleigh number and the ratio of Grashof to Reynolds number are given.
i 1
l NUREQ/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS. DURGIN,W.W.s NORELKA,J.
Alden R6 search Laboratory.
- Sandia Laboratories.
March 1983.
60pp.
8304130465.
SANDB3-7008.
17958:293.
In the event of a loss of coolant accident (LOCA) in a nuclear power plant, it is possible that insulation for pipes or other items inside the containemnt building could be dislodged by the high energy break Jet.
This insulation debris could affect the recirculation of water from the sump of the emergency core cooling system (ECCS) by collecting on the screen currounding this sump.
To help in assessing the susceptibility of fibrous insulation pillows to debris formation under impingement by break-flow Jets, three types of insulation pillows were tested using licuid water jets.
The Jet stagnation pressures required for cover fabric damage and for pillow failure through insulation release were determined for each of the three types at two impingement angles.
In all cases it was found that these pressures were substantially higher than that suggested in NUREQ/CR-2791 (Methodology for Evaluation of Insulation Debris Effects) for evaluation of debris generation.
Based on the experiments conducted here, a value of 20 psi for incipient insulation material release could be used for evaluation purposes for insulation pillows of construction similar to those tested.
NURE0/CR-3177 VO1: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 1: Methodologios. VONHERRMANN,J.
Wood-Leaver & Associates, Inc.
- EchC, Inc.
March 1983.
95pp.
8304060015.
EOG-2243.
17862:227.
Systematic methods for reviewing and evaluating improved emergency procedure guidelines are presented.
The deficiencies of existing " event-oriented" emergency procedures are discussed and the industry efforts to produce improved guidelines in the aftermath of TMI are summarized.
It is concluded that the function-or symptom-oriented approaches which have evolved since TMI have, in theory, the potential to produce effective guidelines.
However, when 72 r
-__--_g-=-
e i
attention is focused on a limited number of critical safety functions (or symptoms indicative of the performance of these functions), the concern arises that diverse accident conditions which exhibit common or similar symptoms can result in ambiguous operator diagnosis and ineffective response.
Methods for systematically examining potential accident sequences using " operator action event trees" are developed in this first volume which can help ensure that functional or symptomatic guidance can, in reality, lead to unambiguous and effective diagnosis and response regardless of the specific failure events.
Subsequent volumes of this report will appig these methods to Westinghouse, General Electric, Babcock & Wilcox, and Combustion Engineering plant designs.
NUREO/CR-3177 VO2: METHDDS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
VDNHERRMANN, J. s BRINSFIELD,W.A.s BRDWN R.O.s et al.
Wood-Leaver &
Associates, Inc.
March 1983.
120pp.
0304060064.
E00-2243.
17863:033.
Systematic methods for finalizing, reviewing, or developing improved emergency procedure guidelines are applied to a Westinghouse a
PWR plant.
The methods are based on the use of operator action event trees (DAETS) which document the key operator actions and plant symptoms associated with the various stages of risk significant multiple failure accident sequences.
The application of the methodology utilizes DAETS developed for the Zion 1 PWR and the Westinghouse Owners Group 's Emergency Rssponse Guidelines.
Since the details of the Westinghouse Function Restoration Guidelines (FRCs) were not get complete, this examination did not take the form of an evaluation of the correctness or completeness of the final guidelines.
Rather, the product of this examination was a delineation of the necessary characteristics which each FRG must possess when it is developed in complete detail.
In addition to methodology l
demonstration, goals of the project included the identification of i
those aspects of Westinghouse plant design, operation, or response to multiple failure accident sequences which could result in incomplete, cabiguous, or incorrect guidance to the operator if not carefully oddressed in the guideline development or utilization process.
NUREQ/CR-3206: UNSATURATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK-RELATED TO HIGH LEVEL WASTE REPOSITORIES. EVANS,D.D.
- Arizona, Univ. o f.
March 1983.
244pp.
8304200527.
18075:045.
The principal objective of this study was to assess the state-of-the-art of characterizing unsaturated fractured rock gelogic cettings for possible high-level radioactive waste isolation.
The primary approach has been to evaluate existing methods and computer models developed for unsaturated granular material nd naturated fractured rock for application to unsaturated rock.
The experimental asethod s explored relate to:
(1) rock matrix characterization, (2) fracture characterization, (3) rock moisture status, and (4) water movement on a mesoscale.
A model was developed to demonstrate the decrease in water conductance of fractures with decrease in water potential.
Another l
codel was developed to show vapor transport in the nonisothermal zone currounding a repository, thus causing a drying and a wetting r en e.
For macroscale modeling of a repository setting, three fields are considered: (1) near-field, the zone surrounding the repository which is influenced by the heat generated by the radioactive waste, (2) meso-field, the unsaturated zone other than the r. ear-field and to the 73
- ~, - - - - - -
regional water table, and (3) far-field, the zone beneath the regional water table.
Clearig the nonisothermal unsaturated near-field is most complex and no existing model is available to realistically model this zone.
Models are available for the meso-Field and far-field but parameter estimation is a serious problem.
NUREC/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOQ1 CAL ASSESSMENT.
MILLS,M.s VOOT D.
Teknekron Research. Inc.
March 1983.
198pp.
8303310366.
17796:297.
This report is the second in a series of five reports that will provide critical reviews and summaries of computer programs that can
(
be used to analyze the potential performance of a high-level radioactive waste repository.
The computer programs identified address the following areas:
radionuclide source term, radiation shielding, airborne transport, surface water transport, dose to man, and health effects.
The report identifies 242 computer codes for use l
in analyzing radiological assessment issues and provides a summary description of 17 computer programs.
These 17 computer programs are being used by the U.S.
Department of Energy, the U.S.
Nuclear Regulatory Commission, or the U.S.
Environmental Protection Agency to ana19:e high-level waste management or reactor effluent releases.
NUREC/CR-3212: SELECTED REVIEW AND EVALUATION OF U.S.
SAFETY RESEARCH VIS-A-VIS FOREIGN SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
STEVENSON.J.D.s THOMAS,F.A.
Stevenson & Associates.
March 1983.
150pp.
8303280061.
17711:064.
A review of current 1g available nuclear research resources in a selected group of United States government national laboratories is presented.
The current nuclear safety research interests of industry organizations, particular1g the Electric Power Research Institute, are also faentified.
In addition, suggestions for potential Joint or cooperative funding of light water reactor safety research in the U.S.
between the NRC and other organizations, both foreign and domestic, are presented.
The topics of research considered are associated with the areas of Siting. Structural Engineering, Metallurgy. Materials, and Mechanical Engineering.
l l
l l
74
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 NUREG/CR for the report and to the 10-digit NRC Document Control System accession number.
SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER 813-1163 NUREG/CR-3065 V02 ENICD-1075 NUREG/CR-1894 813-1163 NUREG/CR-3065 V01 EPRI NP-2015 NUREQ/CR-2401 813-1164D NUREG/CR-2854 HEDL-TME B2-18 NUREG/CR-2805 V01 813-1165 NUREG/CR-2959 HEDL-TME 82-19 NUREO/CR-2805 V02 ANL-82-24 VOL.3 NUREC/CR-2774 V03 HEDL-TME 82-20 NUREG/CR-2805 V03 ANL-82-66 NUREQ/CR-2821 LA-9459-MS NUREQ/CR-2847 ANL-83-7 NUREC/CR-3167 LA-9460-MS NUREC/CR-2848 ANL/AA-28 NUREC/CR-3045 VO1 LA-9499-MS NUREC/CR-2902 ANL/AA-20 VOL.2 NUREC/CR-3045 V02 LA-9553-MS NUREO/CR-3031 BNL-NUREG-51454 NUREC/CR-2331 V02 N2 LA-9563-MS NUREQ/CR-3108 BNL-NUREC-51543 NUREG/CR-1967 LA-9565-MS NUREG/CR-3032 BNL-NUREC-51544 NUREC/CR-2755 LA-9574-MS NUREG/CR-3033 BNL-NUREC-51581 NUREG/CR-2907 UO1 LA-9646-MS NURE0/CR-3135 BNL-NUREC-51585 NUREC/CR-2939 LDL-13341 NUREQ/CR-2910 BNL-NUREG-51586 NUREC/CR-2944 LDL-15029 NUREG/CR-2983 l
BNL-NUREC-51594 NUREC/CR-2993 LDL-15103 NUREC/CR-3062 BNL-NUREG-51600 NUREC/CR-3028 ORNL-5905 NUREC/CR-2863 BNL-NUREG-51601 NUREC/CR-3010 ORNL/CSD/TM-201 NUREQ/CR-3115 BNL-NUREC-51621 NUREC/CR-3014 ORNL/NSIC-200 NUREC/CR-2OOO V01N12 BNL-NUREG-51628 NUREC/CR-3082 ORNL/NSIC-200 NUREG/CR-2000 V02 N1 BNL-NUREC-51629 NUREQ/CR-3086 ORNL/NSIC-200 NUREC/CR-2000 VO2 N2 BNL-NUREG-81494 NURE0/CR-2482 V03 ORNL/TDMC-2 NUREC/CR-2974 BNL/NUREG-51587 NUREG/CR-2945 ORNL/TM-7931 NUREG/CR-2895 CONF-820911 NUREG/CP-0028 V03 ORNL/TM-8299 NUREG/CR-2668 EGC-2037 NUREC/CR-0169 V21 ORNL/TM-8369/V2 NUREG/CR-2751 V02 ECQ-2164 NUREG/CR-2531 RO1 ORNL/TM-8369/V3 NUREQ/CR-2751 V03 I
ECG-2235 NUREG/CR-3103 ORNL/TM-8397/V3 NUREC/CR-2809 V03 ECO-2238 NUREC/CR-3126 ORNL/TM-8418/V2 NUREG/CR-2824 VO2 ECC-2243 NUREC/CR-3177 Vol ORNL/TM-8418/V3 NUREC/CR-2824 V03 EGO-2243 NUREC/CR-3177 VO2 ORNL/TM-8443/V2 NUREC/CR-2874 V02 75
l l
SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER ECO-EA-5289 NUREC/CR-2098 ORNL/TM-8465 NUREQ/CR-2886 ECO-EA-5485 NUREG/CR-2770 ORNL/TM-8500 NUREC/CR-2920 EOG-EA-5502 NUREC/CR-2729 ORNL/TM-8509 NUREG/CR-2968 ECG-EA-5623 NUREC/CR-2771 ORNL/TM-8527/V2 NUREQ/CR-2997 V02 EGG-FN-6174 NUREC/CR-2780 ORNL/TM-8546 NUREG/CR-2998 L
E ENICO-1075 NUREC/CR-1894 ORNL/TM-8585 NUREQ/CR-3113 EPRI NP-2015 NUREG/CR-2401 ORNL/TM-8599 NUREQ/CR-3092 ORNL/TM-8614 NUREG/CR-3114 SAND 81-2488 NUREC/CR-2422 PNL-3794 NUREC/CR-2659 SANDB2-1100 NUREC/CR-2728 FNL-4165 NUREG/CR-2527 SANDS 2-1693 NUREQ/CR-2901 PNL-4241 NURE0/CR-2675 VO3 SAND 82-1788 NUREC/CR-2843 V01 PNL-4275-3 NUREC/CR-2716 V03 SANDB2-2004 NUREC/CR-2925 PNL-429 NUREC/CR-2800 SANDB2-2315 NUREG/CR-3000 PNL-4336 NUREC/CR-2935 SAND 82-2730 NUREQ/CR-3063 PNL-4360 NUREC/CR-2856 SAND 82-2738/1 NUREQ/CR-3069 V01 PNL-4373 NUREG/CR-2878 SANDB2-2738/2 NUREC/CR-3069 V02 PNL-4385 NURE0/CR-3046 V02 SAND 82-2821 NUREQ/CR-3105 PNL-4385 NUREO/CR-3046 VOS SANDB2-2894 NUREG/CR-2887 PNL-4385 NUREO/CR-3046 V01 SANDB2-7057/1 NUREC/CR-2744 V01 PNL-4385 NUREC/CR-3046 V03 SAND 82-7077 NURE0/CR-2904 PNL-4385 NUREG/CR-3046 V04 OAND83-7008 NUREG/CR-3170 PNL-4449 NUREC/CR-2929 UCLA-ENG-8284 NUREC/CR-2666 PNL-4471 NUREQ/CR-2956 UCRL-15495 NUREQ/CR-3102 PNL-4523 NUREC/CR-3027 UCRL-53021 NUREC/CR-2015 V03 PNL-4524 NUREQ/CR-3030 UCRL-53040 NUREC/CR-2801 PNL-4537 NUREC/CR-3059 WCAP-9973 NUREC/CR-2401 PNL-4591 NUREC/CR-3106 PNL-4597 NUREG/CR-3166 PNL-4627 NUREO/CR-3149 SAND 78-2347 NUREG/CR-3009 SAND 80-0822 NUREG/CR-1391 SAND 81-2256 NUREC/CR-2391 76
1 Personal Author Index This iridex 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 furtheiinformation is needed, refer to the main citation by the NUREG number.
C-I ABBOTT,L.S.
NUR EQ/CR-3114: PROCEEDINGS OF WORKSHOP DN COGNITIVE MODELING OF NUCLEAR PLANT CONTROL ROOM OPERATORS. August 15-18,1982,Dedham, Massachusetts.
ACTON,R.U.
NUR EC/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE CONCRETE INTERACTIONS.
ADAMS,R.E.
NUREC/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
AHMAD,A.
NUR EQ/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
ALBRECHT,R.E.
NUR EQ/CR-3141: CHARACTERIZATION OF TWO-PHASE FLOW USING NEUTRONIC FLUCTUATIONS.
ALDRICH,D.C.
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
ALEXANDER,J.F.
NUREC/CR-2443: A MANUAL FOR USINC ENERQY ANALYSIS FOR PLANT SITINC.
ALLEN,O.C.
NUR EC/CR-3OO9: FRACTURE TOUCHNESS OF PWR COMPONENT SUPPORTS.
ANAOLIA,R.
NUR EC/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No. 12.
ANDERSON,F.
NUR EG-0964: TECHNICAL SPECIFICATIONS FOR MCGUIRE NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-369 And 50-370.(Duke Power Company)
ANDREWS,W.B.
NUR EC/CR-2800: CUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
ANDREYCHEK,T.S.
NUR EC/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION Task Plan Rsport.NRC/EPRI/ Westinghouse Report No.
12.
APPS,J.A.
NUREQ/CR-3062: STATUS OF GEOCHEMICAL PROBLEMS RELATING TO THE BURIAL OF HIGH-LEVEL RADIOACTIVE WASTE,1982.
APSEL,R.J.
NUREQ/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL 77
~
CONDITIONS ON SIMULATED GROUND MOTIONS.
ARNAHAN,C.L.
NUREQ/CR-2983: SELECTED HYDROLOGIC AND QEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
ARNOLD,W.D.
NUREG/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
ARONSON,E.A.
NUREQ/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
ASPREY,M.W.
NUREC/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERCOOLING ACCIDENT IN A LARCE, HETEROGENEOUS-CORE, LIQUID-METAL-COOLED FAST BREEDER REACTOR.
ATWOOD,C.L.
NUR EC/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S. Commercial Nuclear Power Plants, January 1,1972 Through September 30,1980.
NUREQ/CR-2729: USER'S CUIDE TO BFR.A Computer Code Based On The Binomial Failure Rate Common Cause Model.
NUREQ/CR-2770: COMMON CAUSE FAULT RATES FOR VALVES: Estimates Based Gn Licensee Event Reports At U.S. Commercial Nuclear Power Plants, 1976-1980.
NUREC/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants, 1976-1978.
BAKER,W.
NUREC/CR-3135: BUCKLING INVESTICATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
BALL,D.G.
NUREG/CR-2895: PWR PRESSURE VESSEL INTEORITY DURING OVERCOOLING ACCIDENTS: A PARAMETRIC ANALYSIS.
BALL,S.J.
NURE0/CR-2874 V02: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION GuarterIg Progress Report, April 1 - June 30,1982.
BARI,R.A.
NUREC/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
NUREG/CR-3020: A REVIEW OF THE LIMERICK CENERATING STATION PROBABILISTIC RISK ASSESSMENT.
BARTTER,W.D.
NUR EC/CR-2668: JOB ANALYSIS OF THE MAINTENANCE SUPERVISOR AND INSTRUMENT AND CONTROL SUPERVISOR POSITIONS FOR THE NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
BASS,B.R.
NUREQ/CR-2997 V02: APPLICATIONS OF ENERGY RELEASE RATE TECHNIQUES TO P ART-THROUGH CRACKS IN PLATES AND CYLINDERS. Volume 2: ORVIRT: A Finite Element Program For Energy. Release Rate Calculations For i
2-Dimensional And 3-Dimensional Crack Models.
BATTS M.
NUR EQ/CR-2945: CHARACTERIZATION OF EARTHOUAKE FORCES FOR PROBABILITY-BASED DESIGN OF NUCLEAR STRUCTURES.
BAUHOF,F.
NUREQ/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP CEOLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
NUREG/CR-3065 V01: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP CEOLOGIC REPOSITORIES. Main Report.
F 78
l 1
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC I
REPOSITORIES. Appendices.
BAUMANN,W.L.
NUREQ/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
BEARE,A.N.
NUREO/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
BELL,C.R.
NUREQ/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN I
UNPROTECTED TRANSIENT UNDERC00 LING ACCIDENT IN A LARGE, HETER 00ENEDUS-CORE,LIGUID-METAL-COOLED FAST BREEDER REACTOR.
BENKOVITZ,C.
NUREQ/CR-2907 VO1: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
BENNETT,J.
l NURE9/CR-3135: BUCKLING INVESTIGATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
BERNREUTER,D.L.
NUREQ/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
NUREC/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH PROGRAM,PHABE I FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
BEZLER,P.
NUREG/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING RESPONSES.
BIAN,S.H.
NUREC/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
BICKFORD,W.E.
NUREQ/CR-2800: QUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
BIDA,Q.
NUREQ/CR-2755: PACKING MATERIAL TESTING REQUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On. Waste Package Verification Tests.
BIRD,S.K.
NUREQ/CR-1894: MECHANICAL RELIABILITY EVALUATION OF A PROPOSED EMERGENCY RESPONSE RADIDIODINE AIR SAMPLER.
B0HN,M.P.
l NUREQ/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
BOLLINGER,G.A.
l NURES/CR-3000: NETWORK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA.
BOROMAN,L.E.
NUREQ/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT OF AVAILABLE MARGINS INHERENT IN FLOOD PROTECTION (F NUCLEAR POWER PLANTS.
BORK0WSKI,R.J.
NUREO/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT.
BOUCHER,T.J.
NUREO/CR-3126: SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTS COMPARISON.
BOWMAN,J.K.
NUREO/CR-2853: NON-CONDENSIBLE GAS FRACTION PREDICTIONS AT ELEVATED TEMPERATURES AND PRESSURE USING WET AND DRY BULB TEMPERATURE MEASUREPENTS.
79
BRACKENBUSH,L.
NUREQ/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substesk B: Dosimeter Response.
BRINSFIELD,W.A.
NURE0/CR-3177 V02: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY j
PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
BRISTOR,D.
NURE0/CR-2703: Investigation of special purpose processors for real-time synthetic aperture focusing techniques for nondestructive evaluation of nuclear reactor vessels and piping components.
BRITE,D.W.
NURE0/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECTFICATIONS.
BROUNS,R.J.
NURE0/CR-2930:. EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATICNS.
BROWN,R.G.
l NURE0/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
BRYAN,R.H.
NUREO/CR-2751 VO2: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR APRIL-JUPE 1982.
NURE0/CR-2751 VO3: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JULY-SEPTEMBER 1982.
BRYANT,J.L.
NURE0/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
BRYSON,J.W.
NURE0/CR-2997 VO2: APPLICATIONS OF ENERGY RELEASE RATE TECHNIGUES TO PART-THROUGH CRACKS IN PLATES AND CYLINDERS. Volume 2: 0RVIRT: A Finite Element Program For Energy Release Rate Calculations For 2-Dimensional And 3-Dimensional Crack Models.
SUEHRING,W.A.
MJREQ/CR-3045 V01: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT GUTAGES. Vol. 1: Approach And Analysis.
NURE0/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT GUTAGES. Vol 2: Appendixes.
BURKE,R.P.
NUREC/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
BURNHAM,B.W.
NUPEQ/CR-2530: REVIEW OF THE GRAND QULF HYDROGEN IGNITER SYSTEM.
BUTLER,T.
NURE0/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
BUYUK0ZTURK40.
NURE0/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
BYERS,R.K.
NURE0/CR-2030: REVIEW OF THE GRAND QULF HYDROGEN IGNITER SYSTEM.
BYRNE,J.
NURE0/CR-3065 V01: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Main Report.
NURE0/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP OEOLDOIC REPOSITORIES. Appendices.
CADWELL,L.L.
NURE0/CR-2675 V03: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF
(
NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Hamid Low-Level Sites.
80
8 CALVO,J.
NURE0/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS 4
UNDER COMBINED ACTION OF SHEAR AND TENSION.
CAMP,A.L.
NUREG/CR-2530: REVIEW OF THE GRAND QULF HYDROGEN IGNITER SYSTEM.
CAMPBELL,J.E.
NUREG/CR-2391: DNET SELF-TEACHING CURRICULUM.
CARLSON,D.D.
NUREG/CR-2729: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
CARNAHAN,C.L.
i NUREG/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE
(
CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The i
Hanford Reservation.
i NURE0/CR-3062: STATUS OF QE0 CHEMICAL PROBLEMS RELATING TO THE BURIAL OF i
HIGH-LEVEL RADIDACTIVE WASTE,1982.
CASE,F.
NUREO/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
CATTON,I.
NUR ES/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
CERBONE,R.J.
NUREG/CR-2331 VO2 N2: SAFETY RGSEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
CHANO,T.Y.
NUR EC/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
CHAPMAN,R.H.
NUREQ/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST HWtBT)
BUNDLE B-4.
CEEVER.G.C.
NUREQ/CR-0169 V21: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY l
ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
CEVERTON,R.D.
NUREG/CR-2895: PWR PRESSURE VESSEL INTEORITY DURING OVERC00 LING ACCIDENTS: A PARAMETRIC ANALYSIS.
CHOU, C. K.
NUREG/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REQULATION DEVELOPMENT.
CHRISTENSEN,D.
NUREG/CR-3145 VO1: GEOPHYSICAL INVESTICATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
CHUANO,T.V.
NUREG/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
l CHUNO,D.H.
NUREO/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH PROORAM, PHASE I FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
CLAPP,N.E.
NURE9/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE T,IVISION OF ACCIDENT EVALUATION. Guarterly Progress Report, April 1 - June 30,1982.
CLEVELAND,J.C.
NUREQ/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES i
FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report.
April 1 - June 30,1982 COHEN,L.
NUREG-0837 VO2 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1992.
CONKLIN,J.C.
81
NUREO/CR-2874 V02: HIGH-TEMPERATURE QAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report, April 1 - June 30,1982.
CONNOR,J.
NUREG/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
COSTELLO,F.
NUREG-OS37 VO2 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Repcet July-September 1982.
COVER,L.E, NUREG/CR-1120 V10: BEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress t
Report No. 14.
]
CRANWELL,R.M.
NURE0/CR-2391: DNET SELF-TEACHING CURRICULUM.
CRAWFORD.S.L.
NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
CROPP,L.O.
NURE0/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
j CROWE,R.D.
NUREQ/CR-3141: CHARACTERIZATION OF TWO-PHASE FLOW USING NEUTRONIC FLUCTUATIONS.
CROWELL,K.R.
NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT l
SYSTEMS. Volume 1: Equations And Constitutive Models.
NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAN REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
CROWLEY,J.L.
NUREQ/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE B-4.
CUESTA-OONZALEZ NUREQ/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
CUMMINOS,F.M.
NUR EC/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
CUMMINGS,J.C.
NUR EO/CR-2530: REVIEW OF THE GRAND QULF HYDROGEN IGNITER SYSTEM.
CUNNINGHAM,M.
NUREG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN.
CUTA,J.M.
NUREQ/CR-3046 V03: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY C00LANT SYSTEMS. Volume 3: Users ' Manual.
NURE0/CR-3046 V04: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS Volume 4: Developmental Assessment And Data.
NURE0/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Programmers Manual.
CIAJK0WSKI,C.
NURE0/CR-2993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM OENERATOR AT THE MAINE YANKEE NUCLEAR POWER STATION.
DAILEY,D.J.
NURE0/CR-3141: CHARACTERIZATION OF TWO-PHASE FLOW USING NEUTRONIC FLUCTUATIONS.
DALY,B.J.
NUREO/CR-3033: MODELING AOKI ET AL. EXPERIMENTS ON CONDENSATION OF FLOWING STEAM ONTO INJECTED WATER VIA K-FIX.
E i
I DAVENPORTeL.C.
NURE0/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
DAVIS,T.J.
NURE0/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTCR CORE SUPPORT ORAPHITE-PART III.
DEEDS,W.E.
NUREQ/CR-2824 V02: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Guarterly Progress Repcet For Period Ending June 30,1982.
NUREQ/CR-2824 VO3: ED0Y-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending September 30, 1982.
DELANY,J.M.
NUREQ/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
DERSHOWITZ,W.
NUREG/CR-3065 V01: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Main Report.
NURES/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Appendices.
DEVAULT,G.P.
NUREQ/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERC00 LING ACCIDENT IN A LARGE,HETEROGENE0VS-CORE,LIGUID-METAL-COOLED FAST BREEDER REACTOR.
DIMENNA,R.A.
NUR EQ/CR-3126: SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTS COMPARISON.
DOCTOR,S.R.
NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
DODD,C.V.
NUREQ/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM CENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending June 30,1982.
NURE0/CR-2824 V03: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending September 30, 1982.
DOMANUS,H.M.
NUREQ/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
DORRIS,R.E.
NUR EO/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
DRAGO, J. P.
NUREQ/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT.
DRESS,W.B.
7 NURE0/CR-3113: A TORSIONAL ULTRASONIC TECHNIGUE FOR LWR LIGUID LEVEL MEASUREMENT.
DURGIN,W.W.
NURE0/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
EASTWOOD,D.
NUREO/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
EBERHARDT,L.E.
NURE0/CR-2675 V03: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid 83
r Low-Level Sites.
ECKEL,J.S.
NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review Of Esisting Human Reliability Data Banks.
ECKERMAN, K. F.
NUREG/CR-2974: USER'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive l
Releases To The Hydrosphere.
EILER, S. K.
NUREG/CR-2716 VO3: REACTOR SAFETY RESEARCH PR00 RAMS.Guarterly Report. July -September 1982.
E00ERS,R.F.
NURE0/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
EHRLICH,M.
NUREG/CR-3120: GUALITY' ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. Annual Report For FY 1992.
EISENHOWER,E.H.
NUREQ/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. Annual Report For FY 1982.
ELLINOWOOD,B.
NURES/CR-2944: TORNADO DAMAGE RISK ASSESSMENT.
NUR E0/CR-2945: CHARACTERIZATION OF EARTHOUAKE FORCES FOR PROBABILITY-BASED DESIGN OF NUCLEAR STRUCTURES.
ELHORE, N. R.
NUREQ/CR-2856: A REVIEW OF FUGITIVE DUST CONTROL FOR URANIUM NILL TAILINGS.
ENDRES,G.W.
NUREQ/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
ENGLAND,T.R.
NUREQ/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
ERICSON,D.M.
l NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREG/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL l
NUCLEAR POWER PLANT SYSTEMS. Main Report.
ESSINGTON,E.H.
NUREQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
EVANS,D.D.
l NUREG/CR-3206: UNSATURATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK-RELATED TO HIGH LEVEL WASTE REPOSITORIES.
FABRY,A.
NUREO/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
i FAGER,J.E.
NUREQ/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
l FAIROBENT,L.A.
NUREG-0851: NOM 00 RAMS FOR EVALUATION N DOSES FROM FINITE NOBLE CAS CLOUDS.
FALETTI,D.W.
NURE9/CR-2659: 10 DINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM GENERATOR TUBES.
I FEDERMAN, P. J.
NURE0/CR-2668: JOB ANALYSIS OF THE MAINTENANCE SUPERVISOR AND INSTRUMENT AND CONTROL SUPERVISOR POSITIONS FOR THE NUCLEAR POWER 84 I
PLANT MAINTENANCE PERSONNEL RELIADILITY MODEL.
FIARMAN,S.
NUREO/CR-3028: A REVIEW OF THE LIMERICK GENERATING STATION PROBABILISTIC RISK ASSESSMENT.
FINDLEY,D.
NUREG/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIOUES FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES.
FINLEY,N.C.
NUREQ/CR-2391: DNET SELF-TEACHING CURRICULUM.
NUR EQ/CR-2422: DOSIMETRY AND HEALTH EFFECTS SELF-TEACHING CURRICULUM.
Illustrative Problems to Supplement The User's Manual For The Desimetry And Health Effects Computer Code.
F0WLER,E.B.
NUREQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SGIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
FRAGOLA,J.
NUREQ/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
FRAQuLA,J.R.
NUREQ/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT.
FURFAPD,J.P.
NUREQ/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MIGRATION.
GALLUCCI,R.H.V.
NUREQ/CR-2000: GUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIDRITIZATION INFORMATION DEVELOPMENT.
GALLUP,D.R.
NUR EC/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
GANAPATHY,S.
3 NUREQ/CR-2703: INVESTICATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
GANAPOL,B.D.
NUR EC/CR-2902: FLOW RECIME MODELING STUDY FOR THE SIMMER-II LMFBR SAFETY CODE: CLAD RELOCATION.
GATES,R.
NUREC/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP CEOLOGIC REPOSITORIES.
NUREQ/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP l
GEDLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
NUREC/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE ('HLW) DEEP OEOLOGIC I
REPOSITORIES. Main Report.
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND l
CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Appendices.
GAZILLO,F.
NUREC/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
l GELBARD,F.
NUR EO/CR-1391: MAEROS USER MANUAL.
GEORGE T.L.
NUREG/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
GIDO,R.J.
NUREG/CR-2847: COGAP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL SYSTEM EVALUATION CODE.
NURE0/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
GINSBERO,T.
IMi
l NURE0/CR-2331 V02 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF
]
NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
NURE9/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
00 NANO,L.
NUREQ/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOCIC REPOSITORIES.
i NUREQ/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP OEOLOGIC REPOSITORY FOR HICH LEVEL NUCLEAR WASTE.
NU"tEC/CR-3065 V01: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Main Report.
NURE0/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP CEOLOGIC i
REPOSITORIES. Appendices.
000DR ICH, C. W.
NUREG/CR-3059: PARAMETRIC CALCULATIONS OF FATIQUE CRACK OROWTH IN PIPING.
I GREENE,G.A.
l NUREC/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REOULATORY RESEARCH.GuarterIV Progress Report, April 1 -June 30,1982.
I NURE0/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CONVECTION HEAT TRANSFER OF INTERNALLY HEATED LIGUIDS.
NURE0/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
ORIFFITH,J.M.
l NURE0/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL.
ORIFFITH,P.
j NURE0/CR-2853: NON-CONDENSIBLE CAS FRACTION PREDICTIONS AT ELEVATED TEMPERATURES AND PRESSURE USING WET AND DRY BULB TEMPERATURE MEASUREM8.?NTS.
GRUNDL,J.A.
NUREG/CR-2005 V03: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PRDORAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
GRUSZCZYNSKI,M.
NUREO/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
QUIDOTTI,T.E.
NURE0/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1:Estuations And Constitutive Models.
NURE0/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT i
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT l
SYSTEMS, Volume 4: Developmental Assessment And Data.
GUPPY,J.G.
NURES/CR-2331 V02 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR RESULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
00THRIE,G.L.
NURE0/CR-2805 V01: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IPPROVEMENT PROGRAM. Guarter1y Progress Report. Janvary 1982 - March 1982.
[
NURE0/CR-2805 V02: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY l
IMPROVEMENT PROGRAM.Guarter1y Progress Report.Apri1 1982 - June 1982.
l HAAS,P.M.
NURE0/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT DPERATOR 86
-ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
HADLEY,M.
NUREQ/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED QROUND MOTIONS.
HALL,R.E.
NUREQ/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSOPED BY OFFICE OF NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
NUREQ/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
HAMAND,N.
NUREQ/CR-2703: Investigation oP special purpose processors for real-time synthetic aperture focusing techniques for nondestructive evaluation of nuclear reactor vessels and piping components.
HANSON,R.J.
t NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR PCWER PLANT SYSTEMS. Executive Summary.
NUREQ/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
HARAY,H.A.
NUREQ/CR-2531 RO1: INTRODUCTORY USERS MANUAL FOR THE US NUCLEAR REQULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
HARRINGTON,R.M.
NUREQ/CR-2874 VO2:. HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report, April 1 - June 30,1982.
HART,R.S.
NUREQ/CR-3102: EFFECTS OF EARTHGUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED GROUND MOTIONS.
HARTLEY,J.N.
NUREQ/CR-2856: A REVIEW OF FUCITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
HEABERLIN,S.W.
NUREQ/CR-2800: QUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
HEISING,C.D.
NUR EQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
HESSON,G.E.
NUREQ/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
HESSON,G.M.
I NUREQ/CR-2659: IODINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM QENERATOR TUBES.
HICKMAN,J.W.
NUREQ/CR-23OO VO1: PRA PROCEDURES QUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREQ/CR-23OO VO2: PRA PROCEDURES QUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
HOCHREITER, L. E.
NUREQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION t
AND REFLUX CONDENSATION. Task Plan Report.NRC/EPHI/ Westinghouse Report No. 12.
HORTON,W.H.
I NUREQ/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
HOWARD,J.H.
NUREQ/CR-3008-AUDITORY PERCEPTION IN LOOSE-PARTS MONITORING.
ISKANDER,S.K.
NUREQ/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERC00 LING ACCIDENTS: A PARAMETRIC ANALYSIS.
87
JACKSON,P.L.
NURE0/CR-3145 VO1: CEOPhYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
JACKSON,P.O.
NUREG/CR-3106: COMPARISDN OF FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
NUREQ/CR-3166: RECOMMENDED PRDCEDURES FOR MEASURING RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS.
JACKSON,R.E.
l NUREG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATION PROGRAM i
- A USE OF PROBABILITY IN DECISION MAKING.
I JOHNSON,J.D.
NURE0/CR-2806: A KINETIC MODEL FOR THE CHLORINATION OF PDWER PLANT l
COOLING WATERS.
i NUR EQ/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL
{
STATEMENTS.
l JOHNSON,J.J.
NUREO/CR-1120 V10: SEISMIC SAFETY MARGING RESEARCH PROGRAM. Progress Report No. 14.
JOHNSON,J.W.
NUREO/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT.
i JOHNSON,T.M.
NUR EG/CR-2904: SIMULATION OF GROUNDWATER FLOW AND CONTAMINANT TRANSPORT i
USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
JONES,K.
NUREG/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP CEDLOGIC REPOSITORIES. Main Report.
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP OEOLOGIC REPOSITORIES. Appendices.
JONES,V.K.
NUREC/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREQ/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
KALEEL,R.J.
NUREG/CR-3149: DISPERSION COEFFICIENTS FOR COASTAL REGIONS.
KAM,F.B.K.
NUREC/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PRDORAM.1982 ANNUAL REPORT (Octaber 1,1981 - September 30, 1982).
KAROL,R.
NUREQ/CR-3028: A REVIEW OF THE LIMERICK OENERATING STATION PROBABILISTIC RISK ASSESSMENT.
j KASTENBERG,W.
l NURE0/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
l KAVALKOVICH,W.
NUREQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report. NRC/EPRI/ Westinghouse Report No. 12.
KAZIMI,M.S.
NURE9/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
KELLY,J.M.
NURE9/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models.
88 L
l NURE0/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Us ers ' Manual.
NUREC/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
KELMERS,A.D.
NUREQ/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Gu>arterly Progress Report, April 1 - June 30,1982.
KENNEDY,W.E.
NUREQ/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REQULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
KING,L.L.
NURES/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
KINNISON,R.R.
NUREQ/CR-2935: EXAMPLES OF MC8 A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
KMETYK,L.N.
NUR EQ/CR-2887: RELAP5 ASSESSMENT: FLECHT SEASET STEAM QENERATOR TEST 23402.
KNOROVSKY,C.A.
NUREC/CR-3OO9: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
KDESTEL,A.
NUR EQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
KOHRT,R.J.
NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models.
NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
KOLACZK0WSKI,A.
NUREC/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
KOLB, 9. J.
NUREC/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
MONZEK,C.J.
NURE0/CR-2800: QUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
KOO, W. H.
NUR EG-0943: THREADED FASTENER EXPERIENCE IN NUCLEAR POWER PLANTS.
i KOONTZ,A.C.
l NUREG/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Us ers ' Manual.
NUREO/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Programmers Manual.
KORDALSKI,F.J.
NUREQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report. NRC/EPRI/ Westinghouse Report No. 12.
KORNEGAY,F.C.
I NURE0/CR-2874 VO2: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report, April 1 - June 30,1982.
KOSALY,9.
I NURE0/CR-3141: CHARACTERIZATION OF TWD-PHASE FLOW USING NEUTRONIC 89 l
l
FLUCTUATIONS.
K0ZINSKY,E.J.
NUREC/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
NUREQ /CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
KRAUPA,J.F.
NUREG/CR-1894: MECHANICAL PELIABILITY EVALUATION OF A PROPOSED EMERGENCY RESPONSE RADIDIODINE AIR SAMPLER.
KRIEC,R.D.
NUREG/CR-3OO9: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
KUFKA,A.L.
NUREQ/CR-3079: EARTHOUAKE HAZARD STUDIES IN NEW YORK STATE AND ADJACENT AREAS. Final Report. April 1976 - June 1982.
LAATS,E.T.
NUREG/CR-2531 Rol: INTRODUCTORY USERS MANUAL FOR THE US NUCLEAR REQULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
LABAUVE,R.J.
NUR EO/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATINC REACTOR RELOAD REVIFWS.
LAHEY,R.T.
NUREG/CR-2972: AN ANALYSIS OF DENSITY-WAVE OSCILLATIONS IN VENTILATED CHANNELS.
LAMKIN,D.
NUREQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
LARKINS,J.T.
NUR EG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN.
LAVINE,M.J.
NUREO/CR-2443: A MANUAL FOR USING ENERQY ANALYSIS FOR PLANT SITINC.
LAWSON,J.E.
NURE0/CR-3109: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA-Final Report.
LEAR,G.E.
NUREQ-0965: NRC INVENTORY OF DAMS.
LEDERMAN,L.
NUREO/CR-3020: A REVIEW OF THE LIMERICK OENERATING STATION I
PROBABILISTIC RISK ASSESSMENT.
LEE,N.
NUR EQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report. NRC/EPRI/ Westinghouse Report
(
No. 12.
LI, C. T.
I NUREQ/CR-2856: A REVIEW OF FUCITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
LICHTPER, P. C.
NURES/CR-3062: STATUS OF OEOCHEMICAL PROBLEMS RELATING TO THE BURIAL OF HIGH-LEVEL RADIDACTIVE WASTE 1982.
l LINK,B.W.
NURE0/CR-3116: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Project Summary Report,Enrico Fernt-1 Reactor.
l LOBNER,P.R.
l NUREG/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES OUIDE.
l LOFOREN, E. V.
I NURE0/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES OUIDE.
NUREO/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS.
LONO,J.C.S.
NUREO/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
90
I l
LONGEST,A.W.
NURE0/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE P-4.
LONGSINE,D.E.
NUR E0/CR-2391: DNET SELF-TEACHING CURRICULUM.
LORENZ,R.A.
NUR E0/CR-2928: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-1.
I LUCK,L.B.
NURE0/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERC00 LING ACCIDENT IN A LARGE, HETER 00ENEOUS-CORE,LIGUID-METAL-COOLED FAST BREEDER REACTOR.
(
LUCKAS,W.J.
NUREO/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
i LUDEW10.H.
NUREC/CR-3029: A REVIEW OF THE LIMERICK OENERATING STATION PROBABILISTIC RISK ASSESSMENT.
LUZA,K.V.
NUREC/CR-3109: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA i
UPLIFT IN OKLAHOMA-Final Report.
MACRAE,B.L.
NUR EC/CR-3149: DISPERSION COEFFICIENTS FOR COASTAL REGIONS.
MANGOLD,D.C.
NURE0/CR-2910: THERMAL IMACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH OEOLDOIC DISPOSAL OF NUCLEAR WASTE.
MANNING,J.J.
NURE0/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
MARCH-LEUBA,J.
NUREQ/CR-2998: A COMPARISON OF BWR STABILITY MEASUREE NTS WITH C ALCULATIONS USING THE CODE LAPUR-IV.
l MARINELLI,F.
I NURE0/CR-3065 VOI: IN SITU TEST PROGRAMS RELATED TO DESIGN AND l
CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (M.W) DEEP OE0 LOGIC REPOSITORIES. Main Report.
NUREC/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLDOIC REPOSITORIES. Appendices.
MARSHALL,R.K.
NURE0/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
MAXWELL,R.L.
STUDY.
NURE0/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES)
MAYER,D.W.
NURE0/CR-3078: MODEL EVALUATION OF SEEPAGE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW TE WATER TABLE.
MCCLUNO,R.W.
NURE0/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM OENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending June 30,1982.
NURE0/CR-2824 VO3: EDDY-CURRENT INSPECTION FOR STEAM OENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending September 30, 1982.
MCELROY,W.N.
NURE0/CR-2805 VO1: LWR PRESSURE' VESSEL SURVEILLANCE DOSIETRY I MROVE E NT PROGRAM.Guarter1y Progress Report.Janvarg 1982 - March 1982.
NURE0/CR-2805 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY I MROVEMENT PROGRAM.Guarterly Progress Report. April 1982 - June 1982.
91
NUREQ/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
MCOARRY,E.D.
NUREQ/CR-2005 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY l
l IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
McKAY,E.D.
NUREO/CR-2904: SIMULATION OF CROUNDWATER FLOW AND CONTAMINANT 1RANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
MCKENZIE,D.H.
NURE9/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
NURES/CR-2804: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSING IMPACTS. Final Report.
MELBER,B.D.
NUREQ/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER PLANTS: THE FOREIGN EXPERIENCE.
MEYER,P.R.
NUR EG/CR-3078: MODEL EVALUATION OF SEEPACE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
MEYER,R.E.
NUREC/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
MICHEL.M.C.
NUREC/CR-3062: STATUS OF GEOCHEMICAL PROBLEMS REL4 TING TO THE BURIAL OF HIGH-LEVEL RADI0 ACTIVE WASTE.1982.
l MIKLOS,J.
NURES/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
NUR E9/CR-2892: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: A Revised Procedures Manual.
MILLER,M.A.
NUREO/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
MILLER,R.L.
NUREG/CR-3116: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Project Summary Report,Enrico Fermi-1 Reactor.
MILLS,M.
NUR E0/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSESSMENT.
MILDISNESS,R.C.
NURE0/CR-3033: MODELING AOKI ET AL. EXPERIMENTS ON CONDENSATIDN OF FLOWING STEAM ONTO INJECTED WATER VIA K-FIX.
MO, T.
NUREG-0851: NOMOORAMS FOR EVALUATION OF DOSES FROM FINITE NOBLE GAS CLOUDS.
MOHR,C.L.
NURE0/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIPEENT (TH-3).
MOHR,D.
NUREQ/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
MORGAN,W.C.
NURE0/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTCR CORE SUPPORT ORAPHITE--PART III.
l MOR TG AT, C. P.
j NUREG/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH PROGRAM,PHABE I FINAL REPORT-DEVELOPMENT OF SEIGNIC INPUT (PROJECT II).
MOTES,B.G.
NUREO/CR-1994: MECHANICAL DELIABILITY EVALUATION OF A PROPOSED 92
EMERGENCY RESPONSE RADIDIODINE AIR SAMPLER.
NEBBITT,B.
NUREQ/CR-3065 VO1: IN SITU TEST PRDORAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEDLDOIC REPOSITORIES. Main Report.
NUREQ/CR-3065 VO2: IN SITU TEST PRDORAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Appendices.
NELSON, R. W.
NUREQ/CR-3078: MODEL EVALUATION OF SEEPAGE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
NEUMAN,S.P.
NUR EQ/CR-3076: COMPUTER PREDICTION DF SUBSURFACE RADIONUCLIDE TRANSPORT l
-AN ADAPTIVE NUMERICAL METHOD.
NOREIKA,J.
NUR EC/CR-3170: THE SUSCEPTIBILITY DF FIBROUS INSULATION PILLDWS TO DEBRIS FORMATIDN UNDER EXPOSURE TO ENERGETIC JET FLOWS.
OBERO.C.R.
NUREG-0948: SPECIAL INSPECTION REPORT OF GUADREX CORPORATION REPDRT ON DESIGN REVIEW DF BROWN & RDOT ENGINEERING WORK FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And Power Company)
OBERLANDER,P.L.
NUR EQ/CR-3075: MODEL EVALUATION OF SEEPAGE FRDM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
ODUM,H.T.
NUR EQ/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FDR PLANT SITING.
OSBOR NE, M. F.
NUR EQ/CR-2929: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-1.
OTADUY,P.J.
NUREQ/CR-2999: A COMPARISON OF BWR STABILITY MEASUREMENTS WITH CALCULATIONS USING THE CODE LAPUR-IV.
OZTUNALI,0.I.
NUREQ/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS DN GROUNDWATER MIGRATION.
PALMER,D.A.
NUREQ/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORrTION MEASUREMENTS.
PAPAZOGLOU,I.A.
NUREQ/CR-3028: A REVIEW DF THE LIMERICK QENERATING STATIDN PROBABILISTIC RISK ASSESSMENT.
PARCHEN,L.J.
NUREQ/CR-2527: LOCA SIMULATIDN IN NRU PRDORAM - DATA REPDRT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
PASCIAK,W.J.
NUREG-OB51: NOMOCRAMS FOR EVALUATION OF DOSES FROM FINITE NOBLE GAS CLOUDS.
PEERENBOOM,J.P.
NUREQ/CR-3045 VO1: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT GUTAGES. Vol. 1: Approach And Analysis.
NUREQ/CR-3045 VO2: LOSS OF BENEFITS RESULTING FRDM NUCLEAR POWER PLANT GUTAGES. Vol 2: Appendises.
PELOGUIN,R.A.
NUREQ/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REQULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
PENTZ,D.
NUREQ/CR-3065 VOI: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEDLOGIC 93
~
REPOSITORIES. Main Report.
NUREG/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP OEDLOGIC REPOSITORIES. Appendices.
PERRY,D.
NUR EQ/CR-3062: STATUS OF OE0 CHEMICAL PROBLEMS RELATING TO THE BURIAL OF l
HIGH-LEVEL RADIDACTIVE WASTE,1992.
PHILIPS,S.
j NUREQ/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIQUES FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES.
PILGER,J.P.
NUREG/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
i PITT,C.J.
NURE0/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MIORATION.
PLATO,P.
NUREQ/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
NUR EQ/CR-2992: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: A Revised Procedures Manual.
I PODOWSKI,M.
NUREO/CR-2972: AN ANALYSIS OF DENSITY-WAVE OSCILLATIONS IN VENTILATED CHANNELS.
j POLZER,W.L.
NUREQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
POSAKONY,9.J.
NUREG/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
POSTMA,A.K.
NUREO/CR-2659: IDDINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM QENERATOR TUBES.
PRATT,W.T.
NUREO/CR-3028: A REVIEW OF THE LIMERICK OENERATING STATION PROBABILISTIC RISK ASSESSMENT.
GUALLS,R.C.
NURE0/CR-2OO6: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT COOLING WATERS.
RANDICH,E.
NURES/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE CONCRETE INTERACTIONS.
l READE,R.T.
l NURE0/CR-2904: SIMULATION OF GROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT) l MODEL.
REICH,M.
NUREG/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAIN^iENT STRUCTURE UNDER UNIFGRH FRE55URE.
r NURE0/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REQULATORY RESEARCH.GuarterIg Progress Report, April 1 -June 30,1982.
REINHOLD,T.A.
1 NUREG/CR-2944: TORNADO DAMAGE RISK ASSESSMENT.
REISENAUER,A.E.
NUREO/CR-3078: MODEL EVALUATION OF SEEPACE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
REITER,L.
NUREG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATION PROGRAM I
- A USE OF PROBABILITY IN DECISION MAKING.
94
t l
RENSNER,G.D.
NUREO/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NURE9/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
RICSON,D.M.
NUREG/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Fieport.
RITCHIE,L.T.
NUREQ/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL STATEMENTS.
ROBERDS W.
NUREG/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR. WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Main Report.
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR' WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Appendices.
l ROSAL,E.R.
l NUR EQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION l
AND REFLUX CONDENSATION.Tash Plan Report.NRC/EPRI/ Westinghouse Report No. 12.
ROUNDTREE,S.L.
NUREQ/CR-3105: SECURORS APPLICATION TO A GENERIC NUCLEAR POWER PLANT.
RUNKLE,G.E.
NURE9/CR-2422: DOSIMETRY AND HEALTH EFFECTS SELF-TEACHING CURRICULUM.
Illustrative Problems to Supplement The User's Manual For The Dosimetry And Health Effects Computer Code.
l RUPPRECHT, S. D.
~
NUREO/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No. 12.
RUSSCHER,G.E.
NUREQ/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
RYAN,R.M.
NUREG/CR-2524: EVALUATION OF PERSONNEL NEUTRON DOSIMETRY AT CPERATING NUCLEAR POWER PLANTS.
SAHA,P.
NUREQ/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1 -June 30,1982.
SANCHEZ,R.
NURES/CR-3141: CHARACTERIZATION OF TWO-PHABE FLOW USING NEUTRONIC FLUCTUATIONS.
l SANDBERO,S.J.
NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREO/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
SASTRE, C.
NUREG/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY REGEARCH.Guarterly Progress Report, April 1 -June 30,1982.
SCHIMA,F.J.
NUREO/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. Annual Report For FY 1982.
SCHMITT,R.C.
NURE0/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
SCWIULT,B.
96
i i
NUR EG/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIQUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
SCHREIBER,R.E.
NUREO/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER PLANTS: THE FOHEIGN EXPERIENCE.
SCHWARTZ,F.W.
. NUREG/CR-2904: SIMULATION OF QROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINIf?'IC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
SCOFIELD,N.R.
NURE0/CR-2531 RO1: INTRODUCTORY USERS MANUAL FOR THE US NUCLEAR REOULATORY COMMIGSION REACTOR SAFETY RESEARCH DATA BANK.
SELTZER,S.
NUREG/CR-3120: QUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING j,'
R ADIATION. A'nnual Report For FY 1982.
SERNE,R.J.
NUREQ/CR-303C: EVALUATICN OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT OF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
SERVER,W.L.
NUREQ/CR-2780: EVALUATION OF SYSTEM REGUIREMENTS AND STANDARDS DEVELOPMENT FOR THERMAL ANNEALING OF REACTOR PRESSURE VESSELS.
SHA,W.T.
NUREG/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
SHARMA,S.
NURE0/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III i
l CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
c SHEARER,D.L.
NUREC/CR-3149: DISPERSION COEFFICIENTS FOR COASTAL REGIONS.
SHERMAN,M.P.
NUREQ/CR-2530: REVIEW OF THE GRAND GULF HYDROGEN IGNITER SYSTEM.
SHERWOOD,D.R.
NURE0/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT OF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
SHIAO,S.Y.
NUREG/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT I
OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
SHIEH,L.C.
NUREQ/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
SHIU, K.
NUR EQ/CR-3028: A REVIEW OF THE LIMERICK CENERATING STATION PROBABILISTIC RISK ASSESSMENT.
SHOUP,R.W.
NUREG/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COPTIERCI AL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
l NUREQ/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PUL9E WITH COMMERCIAL l
NUCLEAR POWER PLANT SYSTEMS, Main Report.
SHTEYNOART,S.
NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.,
SHUKLA.S.N.
NURE9/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PRDORAM. Progress Report No. 14.
SIBOL,M.S.
NUREQ/CR-3080: NETWORK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA.
SIEGEL.A.I.
NURE0/CR-266S: JOB ANALYSIS OF THE MAINTENANCE SUPERVISOR AND INSTRUMENT AND CONTROL SUPERVISOR POSITIONS FOR THE NUCLEAR POWER 96
~
l 1
s n
PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
SILVA,R.J.
L NUREQ/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN 82ITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
NUR EQ/CR-3062: STATUS OF QEOCHEMICAL PROBLEMS RELATING TO THE BURIAL OF HIGH-LEVEL RADIOACTIVE WASTE,1982.
SIMMONS,M.A.
l I
NUREQ/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REQULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
NUREQ/CR-2004: THE APPL 4 CATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSING IMPACTS. Final Report.
SIMONEN,F.A.
NUREQ/CR-3059: PARAMETRIC CALCULATIONS OF FATIQUE CRACK QROWTH IN PIPING.
SINWELL,B.R.
. PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION NUP50/CR-2401:
AND REFLUX CONDENSATION. Task ~ Plan Report.NRC/EPRI/ Westinghouse Report No. 12.
SKALSKI,J.R.
l NUREQ/CR-2804: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIQUES TO l
ASSESSING IMPACTS. Final Report.
SLY,G.A.
NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
SMAARDYK,J.E.
NUREQ/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE l
CONCRETE INTERACTICNS.
SMITH,R.I.
NUREQ/CR-2EOO: QUIDELINES-FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
I iSNEIDER,S C NUREQ/CR-3078: MODEL EVALUATION OF SEEPAGE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
SOARES,C.
NUREQ/CR-3120: GUALITY' ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. ' Annual Report For FY 1982.
SOO, P.
NUREQ/CR-2482 VG3: REVIEW OF DOE WASTE PACMAGE PROGRAM. Subtask 1.1 -
National Waste Package F' Neam. April 1982 - Sep terabar 1982.
SPENCER, R. K.
NUR EQ/CR-2910: THERP'U. *," A T OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCI ATED WITH (";G.'.W
',ISPOSAL OF NUCLEARI WASTE.
STACK, D. W.'
NUREQ/CR-2728: INTEkiM RELIA 61LITY EVALUATION PROGRAM PROCEDURES QUIDE.
STEVENSON,J.D.
SAFETY etESEARCH NUREQ/CR-3212: SELECTED REVIEW AND EVALUATION OF U.S.
VIS-A-VIS FOREZON SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
STEVERSON,J.A.
NUREQ/CR-2770: COMMON CAUSE FAULT RATES FOR VALVES: Estin,etes Based On Licensee Ev?nt Reports At U.S.
Commerci&] Nuclear Power Plants, 1976-1980.
STEWART,M.
NUREQ/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Main, Report.
NUREQ/CR-3065 VO2: IN, SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLDOIC 97 l
l t
REPOSITORIES. Appendices.
STRAWE,D.F.
NURE0/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL I
NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NURE0/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
STRENGE,D.L.
NUREO/CR-2800: OUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
STUCKWISCH,S.E.
NUREQ/CR-2391: DNET SELF-TEACHING CURRICULUM.
I SUBUDHI,M.
NUREG/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING RESPONSES.
SUITT,W.J.
NUR EQ/CR-2729: USER'S GUIDE TO BFR.A Computer Code Based On The Binomial Failure Rate Common Cause Model.
SULLIVAN,J.E.
NUREO/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
SZAWLEWICZ,S.A.
NUREQ/CP-OO41 VO1: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETINC.
NUREQ/CP-OO41 VO2: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
(
NUREG/CP-OO41 VO3: PROCEEDINCS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETINC.
NUREQ/CP-OO41 VO4: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NUREQ/CP-OO41 VO5: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NURE0/CP-OO41 VO6: PROCEEDINGS OF THE TENTH MATER REACTOR SAFETY RESEARCH IlJFORMATION MEETING.
TALEYARKHAN,R.
NUREQ/CR-2972: AN ANALYSIS OF DENSITY-WAVE OSCILLATIONS IN VENTILATED CHANNELS.
TAYLOR,T.T.
NUREG/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
THOMAS,F.A.
NUREG/CR-3212: SELECTED REVIEW AND EVALUATION OF U.S.
SAFETY RESEARCH VIS-A-VIS FOREIGN SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
THOMAS,M.T.
NUREG/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTOR CORE SUPPORT ORAPHITE--PART III.
THOMAS,V.W.
NUREQ/CR-3106: COMPARISON OF FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONC-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
NUREG/CR-3166: RECOMMENDED PROCEDURES FOR MEASURING RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS.
THOMPSON,D.D.
NUREG-0965: NRC INVENTORY OF DAMS.
THOMP SON, S. L.
NUREG/CR-2843 VO1: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.GuarterIg Report, January - March 1982.
THOMP SON, T.
NUREG-0837 VO2 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1982.
THURGOOD,M.J.
NURE0/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT 98
i ANALYSIS OF NUCLEAtt REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models.
NUREG/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
NUREG/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Us ers
- Manual.
NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
TICHLER,J.
NUREG/CR-2907 VO1: RADIDACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
TOBIAS,M.L.
NUREG/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PRDORAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
TCHASKO,D.
NUR EQ/CR-2530: REVIEW OF THE GRAND CULF HYDR 00EN IGNITER SYSTEM.
TOPMILLER,D.A.
NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review 0F Existing Human Reliability Data Banks.
TRAVIS,J.R.
NUREQ/CR-2928: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-1.
TSANG,C.F.
NUREC/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH OEOLOGIC DISPOSAL OF NUCLEAR WASTE.
TUCKER,P.C.
NUREO/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIGUES FOR LOW-LEVEL RADIOACTIVE WASTE DISPOSAL.
UPPULURI V.R.R.
NUREG/CR-3115: EXPERT OPINION AND RANKING METHODS.
VARCOLIK,F.
NUREG/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS.
VENHUIZEN,J.R.
NUREG/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL.
VISKANTA,R.
NUREG/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
VDOT, D.
NUREG/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSESSMENT.
VONHERRMANN,J.
NUREQ/CR-3177 VO1: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 1: Methodologies.
NUREQ/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
WALTERS,W.H.
NUREG/CR-3027: OVERLAND EROSION OF URANIUM MILL TAILINGS IMPOUNDMENTS:
PHYSICAL PROCESSES AND COMPUTATIONAL METHODS.
WANG,J.S.V.
NUREG/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH OEOLOGIC DISPOSAL OF NUCLEAR WASTE.
WANG,R.C.
NUREQ/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
WANO,Y.M.
NUR EG/CR-3086-INVESTICATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING 99
e l
RESPONSES.
WANGLER.M.E.
NUR EG-0951: NOMOGRAMS FOR EVALUATION OF DOSES FROM FINITE NOBLE CAS CLDUDS.
WATKINS D.J.
NURE0/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATIDN FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
WEAKLEY,S.A.
NUR EG/CR-2800: QUIDELINES rDR NUCLEAR PDWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
WEBB.B.J.
NUREG/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FDR THERMAL HYDRAULIC EXPERIMENT (TH-3).
WEBSTER,C.S.
NUR EQ/CR-2928: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-1.
WEEKS,J.R.
NUREG/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Guarterly Progress Report, April 1 -June 30,1982.
WEINSTEIN,E.
NUR EQ-OB45: AGENCY PROCEDURES FOR THE NRC INCIDENT RESPONSE PLAN.
WELDON,C.
NUREQ/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Main Report.
NUREC/CR-3065 VO2: IN SITU TEST PRDCRAMS RELATED TO DESIGN AND CONSTRUCTION DF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOCIC REPOSITORIES. Appendices.
WELLS,J.E.
NUREQ/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
WENZEL A.C.
NUREQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No.
12.
WERES,0.
NUR EQ/CR-3062: STATUS OF GEDCHEMICAL PROBLEMS RELATING TO THE BURIAL OF HIGH-LEVEL RADIOACTIVE WASTE.1982.
WESTER,M.J.
NUREQ/CR-2530: REVIEW OF THE CRAND GULF HYDROGEN ICNITER SYSTEM.
WHITE,A.F.
NUREG/CR-2903: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE l
CHARACTERIZATIDN FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The l
Hanford Reservation.
NUREQ/CR-3062: STATUS OF GEOCHEMICAL PROBLEMS RELATING TO THE BURIAL OF HIGH-LEVEL RADIOACTIVE WASTE,1982.
WHITE,J.E.
NUREC/CR-2974: USER'S MANUAL FOR LPCS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive l
Releases To The Hydrosphere.
WHITMAN,0.D.
NUREQ/CR-2751 VO2: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PRDORESS REPORT FOR APRIL-JUNE 1982.
NUREQ/CR-2751 VO3: HEAVY-SECTION STEEL TECHNDLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JULY-SEPTEMBER 1982.
WIDMAYER,D.A.
NUREG-0959: USER'S CUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
WIEDENBECK M.G.
100
NUREC/CR-3145 VO1: GEDPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
WILDANGER,E.
NUREC/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED.TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEDLOGIC REPOSITORIES. Main Report.
NUREG/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Ap p end i c e s.
WILDANGER,W.
NUREG/CR-2834: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FDR i
HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEDLOGIC REPOSITORIES.
i' WILDUNG,N.J.
NUR EC/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
WILKINS,C.A.
NUREC/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
WILLIAMS,C.B.
NUREG/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREG/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
WILLIAMS,R.C.
NUREC/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
WILSON,C.L.
NUR EG/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
WILSON,C.R.
NUREG/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
WILSON,R.L.
NUREC/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
WILSON, W. B.
NUREG/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
WISMER,M.C.
NUREG/CR-2527: LOCA SIMULATIDN IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
WONG,R.
I NUREC/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
(
WONG,S.
i NUR EC/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No. 12.
I WOO,H.H.
NUR EG/CR-2001: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REGULATION DEVELOPMENT.
l WOOD,R.S.
NUREG-0584 RO3 DRFT: ASSURING THE AVAILABILITY OF FUNDS FOR DECOMMISSIDNING NUCLEAR FACILITIES.
WREATHALL,J.
NUREC/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
WU,W.S.
NUR EG/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FOR 101
REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE l
EVALUATION DF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
WULFF,W.
NURE0/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Guarterly Progress Report April 1 -June 30,1982.
WYANT,F.J.
NUR EG/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
YALCINTAS,M.C.
NUREC/CP-OO28 VO3: PROCEEDINGS OF THE SYMPOSIUM ON LOW-LEVEL WASTE DISPOSAL: Facility Design, Construction,And Operating Practices.
YOUNC,J.A.
NUR EG/CR-3106: COMOARISON OF FIVE-MINUTE RADON DAUCHTER MEASUREMENTS j
WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
NUREQ/CR-3166: RECOMMENDED PROCEDURES FOR MEASURINC RADON FLUXES FRDM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS.
l l
?
l l
1 102
1 l
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.
ATWS NUR EC-0977: NRC FACT-FINDING TASK FORCE REPORT ON THE ATWS EVENT AT SALEM NUCLEAR GENERATING STATION, UNIT 1, ON FEBRUARY 22 AND 25,1983.
Abnormal Occurrence NUREG-OO90 VOS NO3: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. July -
September 1982.
Accident Mitigation NUREQ/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
Accident-Sequence Definition NUREQ/CR-23OO VO1: PRA PROCEDURES QUIDE.A Quide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREQ/CR-23OO VO2: PRA PROCEDURES QUIDE.A Quide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
I Accident NUR EG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN.
(
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO l
REACTOR ACCIDENTS.
I NUREQ/CR-2716 VO3: REACTOR PAFETY RESEARCH PROGRAMS.Guarterly Report. July -September 1982.
NUREQ/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
NUR EQ/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
NUREQ/CR-2887: RELAPS ASSESSMENT: FLECHT SEASET STEAM QENERATOR TEST 23402.
NUREQ/CR-3177 VO1: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCED'JRE QUIDELINES. Volume 1: Methodologies.
NUR EQ/CR-3(,? O: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION I
TREE / TIME RiELIABILITY CORRELATION.
I NUREQ/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
NUREQ/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERCOOLING ACCIDENTS: A PARAMETRIC ANALYSIS.
Acoustic NUREQ/CR-3OOB: AUDITORY PERCEPTION IN LOOSE-PARTS MONITORING.
Adsorption NUREQ/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
Advisory Committee on Reactor Safeguards NUR EG-0963: REVIEW AND EVALUATION OF THE NUCLEAR REQULATORY COMMISSION 103
SAFETY RESEARCH PROGRAM FOR FISCAL YEARS 1984 AND 1985.
Aeroso!
NUREG/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
Air Sampler NUR EQ/CR-1894: MECHANICAL RELIABILITY EVALUATION OF A PROPOSED EMERGENCY RESPONSE RADIDIODINE AIR SAMPLER.
Airborne NUREC/CR-2907 VO1: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
Anticipated Transient Without Scram NUR EC-0977: NRC FACT-FINDING TASK FORCE REPORT ON THE ATWS EVENT AT SALEM NUCLEAR GENERATING STATION, UNIT 1,0N FEBRUARY 22 AND 25,1983.
Aqua Book NUREG-0606 VOS NO1: UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As Of February 18, 1983.(Aqua Book)
Armed Response Force NUR EG-0907: ACCEPTANCE CRITERIA FOR DETERMININO ARMED RESPONSE FORCE SIZE AT NUCLEAR POWER PLANTS.
NUR EC-0907: ACCEPTANCE CRITERIA FOR DETERMININC ARMED RESPONSE FORCE SIZE AT NUCLFAR POWER PLANTS.
Atmospheric Releases NUR EQ-0851: NOMOGRAMS FOR EVALUATION DF DOSES FROM FINITE NOBLE CAS CLOUDS.
Axial Tension E
BFR NUREQ/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER CDMBINED ACTION OF SHEAR AND TENSION.
NUR E0/CR-2729: USER'S CUIDE TO BFR.A Computer Code Based On The Binomial Failure Rate Common Cause Model.
Backfill NUR EQ/CR-2755: PACKING MATERIAL TESTINC REGUIRED TO DEMONSTRATE COMPLI ANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Peport On Waste Package Verification Tests.
Basalt NUREC/CR-2482 VO3: REVIEW OF DOE WASTE PACKAGE PROCRAM. Subtask 1.1 -
National Waste Package Program. April 1982 - September 1982.
Biological Shield NUREQ/CR-295(: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
Biotic Transport NUREC/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
Blue Book NUREC-0580 Vil N12: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Blue Book)
NUREG-0580 V12 NO2: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Blue Book)
Boundary Heat Transfer NUR EQ/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CONVECTION
=
HEAT TRANSFER OF INTERNALLY HEATED LIQUIDS.
Boundary NUR EQ/CR-2904: SIMULATIDN OF GROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
Break Tests NUR EC/CR-3126: SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTS COMPARISDN.
Budget Estimates 104
NUREG-0953: FY 1984/85 BUDGET ESTIMATES.
Budget NUREG-0885 102: US NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING l
GUIDANCE 1983.
l Burnup I
NUR EQ/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
COBRA / TRAC l
l NUREQ/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
NUREQ/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT j
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT
[
SYSTEMS. Volume 5: Programmers Manual.
NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT
. ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models.
NUREQ/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Users ' Manual.
COGAP NUREQ/CR-2847: COGAP: A NUCLEAR POWER PLANT CONTAINMENT HYDR 00EN CONTROL SYSTEM EVALUATION CODE.
COMMIX-1A NUREC/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
CPUE NUREQ/CR-2804: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSING IMPACTS. Final Report.
CRAC2 NUREG/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL STATEMENTS.
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
CRT NUREC/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL.
Certificates of Compliance NUREG-0383 VO2 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates of Compliance.
NUREG-0383 VO3 RO2: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Summary Report Of NHC Approved Guality
)'
Assurance Programs For Radioactive Material Packages.
l 0
NUREG-0383 VO1 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR R ADI0 ACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
Chloramine NUR EQ/CR-2806: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT COOLING WATERS.
Chlorine NUREQ/CR-2806: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT COOLING WATERS.
Clad Motion NUREQ/CR-2902: FLOW REGIME MODELING STUDY FOR THE SIMMER-II LMFBR SAFETY CODE: CLAD RELOCATION.
Cladding NUREO/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (P5tBT)
BUNDLE B-4.
106
=
Codes NUR EQ/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSESSMENT.
Cognitive Modeling NUREQ/CR-3114: PRDCEEDINGS OF WORKSHOP DN COGNITIVE MODELING OF NUCLEAR PLANT CDNTROL ROOM OPERATORS. August 15-18,1982, Dedham, Massachusetts.
Common Cause Failure Rates NUREQ/CR-2729: USER'S QUIDE TO BFR.A Computer Code Based On The Binomial Failure Rate Common Cause Model.
Common Cause Fault Rates NUR EQ/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CDNTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants. 1976-1978.
NUREQ/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S. Commercial Nuclear Power Plants, January 1,1972 Through September 30,19G0.
NUREQ/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S. Commercial Nuclear Power Plants, January 1.1972 Through September 30,1980.
Component Supports NUREQ/CR-3009: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
Computer Programs NUREQ/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSESSMENT.
Congress NUREG-0090 V05 NO3: REPORT TO CONORESS ON ABNDRMAL OCCURRENCES. July -
September 1982.
Container NUR EQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
NUREQ/CR-2482 VO3: REVIEW OF DDE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program. April 1982 - September 1982.
Containment Atmosphere NUR EQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
Containment Qas NUREQ/CR-2847: CDGAP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL SYSTEM EVALUATIDN CODE.
it.
Containment Pressure NUREQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
Containment Vessel NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
NUREQ/CR-3135: BUCKLING INVESTICATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINCORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
Containment NUR EQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
Contaminant Transport NUREQ/CR-2904: SIMULATION OF QRDUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
Contract Research NUREG-0975 VO1: COMPILATION OF CONTRACT RESEARCH FDR THE MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEERING TECHNOLOGY. Annual Report For FY 1982.
Control Assemblies NUREQ/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants. 1976-1970.
Cooling Systems 106
NUREQ/CR-2659: IDDINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE i
BREAK WITH CONCURRENT RUPTURES OF STEAM GENERATOR TUBES.
Cooling Water NUR EG/CR-2806: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT COOLING WATERS.
Core-Melt NUREQ/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSENT.
Core-Support NUREG/CR-2929: FEASIBILITY OF MONITORING TE STRENGTH OF HTGR CORE SUPPORT GRAPHITE--PART III.
Costs NUREQ/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT t
OUTAGES. Vol 2: Appendixes.
l Crack NUR EQ/CR-3059: PARAMETRIC CALCULATIONS OF FATIGUE CRACK GROWTH IN P IP ING.
Cyclic Membrane Shear NUREQ/CR-3I 57: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
DPCT NU? EG/CR-2904: SIMULATION Or: GROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
Dams NUR EG-0965: NRC INVENTORY OF DAMS.
Daughter Measurements NUREG/CR-3106: COMPARISDN OF FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
Dearis NUREG/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO i
DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
I Decommissioning NUREG-0584 RO3 DRFT: ASSURING THE AVAILABILITY OF FUNDS FOR DECOMMISSIONING NUCLEAR FACILITIES.
NUREQ/CR-3116: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Project Summary Report,Enrico Fermi-I Reactor.
Degraded Core NUR EQ/CR-2666: PWfi SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
Design Review l
NUR EG-0948: SPECIAL INSPECTION REPORT OF GUADREX CORPORATION REPORT DN DESIGN REVIEW OF BROWN & ROOT ENGINEERING WORK FOR SOUTH TEXAS PROJECT, UNITS I AND 2. Docket Nos.50-49S And 50-499. (Houston Lighting And Power Company)
Disposal Site NUREQ/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MICRATION.
NUREQ/CR-3166: RECOMMENDED PROCEDURES FOR MEASURING RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS.
Disposal NUREG/CR-3078: MODEL EVALUATION OF SEEPAGE FROM URANIUM TAILINGS
[
l DISPOSAL ABOVE AND BELOW THE WATER TABLE.
NUREQ/CR-2983: SELECTED HYDROLOGIC AND GE0 CHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
Dose Projections NUREG/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
Dose NUREC/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid 107
1 Low-Level Sites.
NUREG-0851: NOMOCRAMS FOR EVALUATION OF DOSES FROM FINITE NOBLE CAS CLOUDS.
Dosimeters NUREO/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
Dosimetry NUREO/CR-2992: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: A Revised Procedures Manual.
NUREQ/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
Draft Environmental Impact Statement NUREG-0959: USER'S QUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
Dust 4
NUREG/CR-2856: A REVIEW OF FUCITIVE DUST CONTROL FOR URANIUM MILL
{-
TAILINGS.
l Dyna:aic Loading NUR EC/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS DETWEEN PRIMARY AND SECONDARY PIPING RESPONSES.
EBR-II Pool Reactor NUREQ/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
i ECCS NUREC/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERCETIC JET FLOWS.
EMP i
NURE0/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL l
NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREC/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
EMTP 3
{
NUREQ/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
Ear th qua k e NUR EG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATION PROGRAM l
- A USE OF PROBABILITY IN DECISION MAKING.
I NUR EQ/CR-3086: INVESTICATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING j
RESPONSES.
NUREC/CR-3079: EARTHOUAKE HAZARD STUDIES IN NEW YORK STATE AND ADJACENT AREAS. Final Report. April 1976 - June 1982.
NUREC/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
NURE0/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
NUREQ/CR-3145 VO1: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
)
NUREG/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED GROUND MOTIONS.
Eddy-Current NUREQ/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING l
PROGRAM. GuarterIg Progress Report For Period Ending June 30,1982.
Electrical Systems NUREG/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
Electromagnetic Pulse l
NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NURE9/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
Emergency Core Cooling System NUREQ/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO 108
l DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
Emergency Procedure NUREQ/CR-3177 VO1: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 1: Methodologies.
NUREQ/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE GUIDELINES. Volume 2: Applications To Westinghouse Plants.
1 Emergency Response NUREG-0737 SO1: CLARIFICATION OF TMI ACTION PLAN REGUIREMENTS: REGUIREME NTS FOR EMERGENCY RESPONSE CAPABILITY.
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
NUREQ/CR-1894: MECHANICAL RELIABILITY EVALUATION OF A PROPOSED EMERGENCY RESPONSE RADIDIODINE AIR SAMPLER.
Energetic Reactions NUREG/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE CONCRETE INTERACTIONS.
Energy Analysis NUR EQ/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
Enforcement Actions NUREG-0940 VOi NO4: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED.Guarterly Progress Report. October - December 1982.
Engineering NUREG/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER PLANTS: THE FOREIGN EXPERIENCE.
Enrichments NUREQ/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
Environmental Harerds NUR EQ/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT OF AVAILABLE MARQINS INHERENT IN FLOOD PROTECTION OF NUCLEAR POWER PLANTS.
Environmeatal Impact Statement NUR EQ/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL STATEMENTS.
Environmentally Assisted Cracking i
NUREQ/CR-2993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM QENERATOR l
AT TE MAINE YANKEE NUCLEAR POWER STATION.
i Errors NUREG/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
Eulerian-Lagrangian NUREQ/CR-3076: COMPUTER PREDICTION OF SUBSURFACE RADIONUCLIDE TRANSPORT
-AN ADAPTIVE NUMERICAL METHOD.
Expert Opinion NUREQ/CR-3115: EXPERT OPINION AND RANKING METHODS.
Exposure Pathqgs NUREQ/CR-2673 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
Exposure NUREQ/CR-2974: USER'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
Extensional Ultrasonic Pulses NUREO/CR-3113: A TORSIONAL ULTRASONIC TECHNIGUE FOR LWR LIGUID LEVEL MEASUREENT.
External Hazards NUREO/CR-23OO VO1: PRA PROCEDURES OUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREO/CR-23OO VO2: PRA PROCEDURES CUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
109 l
FLECHT SEASET NUREQ/CR-2687: RELAP5 ASSESSMENT:FLECHT SEASET STEAM QENERATOR TEST 23402.
i NUREG/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No.
12.
Fast Aerosol Simulant Facility NUREC/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
i Fatigue Failure NUR EG/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REGULATION DEVELOPMENT.
Fatigue NUREQ/CR-3059: PARAMETRIC CALCULATIONS OF FATIGUE CRACK GROWTH IN PIPING.
i Feedwater Line Cracking NUREC/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR l
LICENSING REQULATION DEVELOPMENT.
Final Environmental Statement NUREG-0921: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF CATAdBA NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And a'4..(Duke Power Company,et al.)
Fint-integrated Plant Safety Assessment NV ' J-OS22: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION Ph0 GRAM,0YSTER CREEK NUCLEAR GENERATING STATION. Docket No.
50-219.(GPU Nuclear Corporation And Jersey Central Power & Light Company)
Fiscal Year Budget NUREG-0953: FY 1984/85 BUDGET ESTIMATES.
Fisheries Management NUREQ/CR-2004: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSING IMPACTS. Final Report.
Flaws NUREC/CR-3059: PARAMETRIC CALCULATIONS OF FATIGUE CRACK GROWTH.IN 4
PIPING.
NUREQ/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERC00 LING ACCIDENTS: A PARAMETRIC ANALYSIS.
Flood Basalts NUREC/CR-2983: SELECTED HYDROLOGIC AND GE0 CHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
Flooding NUREG-0965: NRC INVENTORY OF DAMS.
i NUREC/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT OF AVAILABLE MARGINS INHERENT IN FLOOD PROTECTION CF NUCLEAR PCBJER PLANTS.
Foreign NUR EG/CR-3212: SELECTED REVIEW AND EVALUATION OF U.S.
SAFETY RESEARCH VIS-A-VIS FOREIGN SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
NUREG/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER i
PLANT 9: THE FOREIGN EXPERIENCE.
Fracture Toughness NUREC/CR-3OO9: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
NUREC/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERCOOLING ACCIDENTS: A PARAMETRIC ANALYSIS.
Fuel Blowdown l
NUREC/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LPFBR.
Fuel Bundles NUREG/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
110 m-
Fuel Dispersal NUREO/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF TE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
Fuel Pin NUREQ/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE B-4.
Fuel NUREG/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
Fugitive Dust NUREQ/CR-2856: A REVIEW OF FUCITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
Funds NUREG-0584 RO3 DRFT: ASSURING THE AVAILABILITY OF FUNDS FOR DECOMMISSIONING NUCLEAR FACILITIES.
Geologic Repository NUREC/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH QEOLOGIC DISPOSAL OF NUCLEAR WASTE.
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP OEOLOGIC REPOSITORIES. Appendices.
NUREQ/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP QEOLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
NUREQ/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Main Report.
NUREC/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR HICH LEVEL NUCLEAR WASTE (HLW) DEEP OEDLOGIC REPOSITORIES.
Oraphite NUREQ/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTGR CORE SUPPORT ORAPHITE--PART III.
Orenville Front NUREG/CR-3145 VO1: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHID-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
Ground Mstion NUR EG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATION PROGRAM
- A USE OF PROBABILITY IN DECISION MAKING.
NUREQ/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED GROUND MOTIONS.
Ground Water NUREO/CR-3078: MODEL EVALUATION OF SEEPACE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
Oroundwater Flow NUREQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE I
COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
l NUREQ/CR-2904: SIMULATION OF QROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
Groundwater NUREQ/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MIGRATION.
Ouidelines NUREQ/CR-2800. QUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
HPSSC Standard NUREG/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
Harard NUREG/CR-3079: EARTHGUAKE HAZARD STUDIES IN NEW YORK STATE AND ADJACENT 111
i AREAS. Final Report. April 1976 - June 1982.
Heat Flux NUREQ/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CONVECTION HEAT TR ANSFER OF INTERNALLY HEATED LIGUIDS.
Heat Transfer NUREG/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
High-Level Nuclear Waste NURE0/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
Human Error NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT i
OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
I Human Operator l
NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
l Huron Performance NUREC/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
Human Reliability Data Banks NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
Hydrogen Burn NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
Hydrogen Combustible-Gas Control Systems l
NUREQ/CR-2847: C00AP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL SYSTEM EVALUATION CODE.
Hydrogen Igniter NUREC/CR-2843 VO1: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report January - March 1982.
Hydrogeologic Systems NUREC/CR-2904: SIMULATION OF QROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
Hydrosphere NUREQ/CR-2974: USER 'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
IPSAR NUREG-0822: INTEORATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION PROGRAM, OYSTER CREEK NUCLEAR GENERATING STATION. Docket No.
50-219.(OPU Nuclear Corporation And Jersey Central Power & Light Company)
NUREG-0823: INTEORATED SAFETY ASSESSMENT SYSTEMATIC EVALUATION PROGRAM, DRESDEN NUCLEAR POWER STATION UNIT 2. Docket No. 50-237. (Commonwealth Edison Company)
IREP NUR EC/CR-2728: INTERIM RELI ABILITY EVALUATION PROGRAM PROCEDURES GUIDE.
Impact Analysis Codes NUREG-095": USER'S GUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
Impact Monitoring NUR EG/CR-2004: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSING IMPACTS. Final Report.
Impoundment NUR EG/CR-2856: A REVIEW OF FUGITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
NUREG-0965: NRC INVENTORY OF DAMS.
In Situ Test Facility 112
NUREC/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Main Rep ort.
NUR EG/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FDR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEDLOGIC REPOSITORIES.
NUREC/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Ap p end ic es.
NUREC/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP GEOLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
In-Service Inspection NUR EC/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE l
EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
NUREG/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending June 30,1982.
Incident Response Plan NUR EG-084 5: AGENCY PROCEDURES FOR THE NRC INCIDENT RESPONSE PLAN.
Inspection Report NUR EG-0948: SPECIAL INSPECTION REPORT OF GUADREX CORPORATION REPDRT ON DESIGN REVIEW OF BROWN & ROOT ENGINEERING WORK FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And Power Company)
Inspections NUREG-OO40 VO6 NO4: LICENSEE CONTRACTOR AND VENDOR INSPECTIDN STATUS REPORT. Guarterly Report,0ctober 1982 - December 1982.
Instrumentation NUR EQ/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants, 1976-1978.
Insulation NUREC/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLDWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
Integrated Plant Safety Assessment NUREG-0824: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATION PROGRAM-MILLSTONE NUCLEAR POWER STATION, UNIT 1. Docket No.
50-245.(Northeast Nuclear Energy Company)
Interim Reliability Evaluation Program NUR EC/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES GUIDE.
Intrusion NUR EC/CR-3105: SECURORS APPLICATION TO A GENERIC NUCLEAR POWER PLANT.
Inventorg NUREG-0430 VO3 NO1: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. January 1982 - June 1982.
Iodine Transport NUR EC/CR-2659: IODINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM GENERATOR TUBES.
Ionizing Radiation NUREC/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. Annual Report For FY 1982.
Irradiated NUREG/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
L APUR-IV NUREG/CR-2998: A COMPARISON OF BWR STABILITY MEASUREMENTS WITH C ALCULATIONS USING THE CODE LAPUR-IV.
LER l
NUREC/CR-2OOO VO2 N1: LICENGEE EVENT REPDRT (LER) CDNPILATION: For Month Of January 1983 NUREC/CR-2OOO VO2 N2: LICENBEE EVENT REPORT (LER) COMPILATION: For Month 113
Of February 1983.
NUREG/CR-2OOO V01N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of December 1982.
LMFBR NUREQ/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERC00 LING ACCIDENT IN A LARGE, HETEROGENEOUS-CORE,LIGUID-METAL-COOLED FAST BREEDER REACTOR.
LOCA i
NUREG/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL i
HYDRAULIC EXPERIMENT (TH-3).
NUREQ/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
LOFT NUREG/CR-0169 V21: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY i
ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
LPGS NUREC/CR-2974: USER'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
LWR-PV-BDIP j
NUREQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report
)
On Waste Package Verification Tests.
NUREQ/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (Octaber 1,1981 - September 30, 1982).
Land Disposal of Radioactive Weste NUREG-0959: USER'S GUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
Leach Rate NUR EQ/CR-3130: INFLUENCE T.,F LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MIGRATION.
Leachate NUREG/CR-3078: MODEL EVALUATION OF SEEPACE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
Level Measurement NUREC/CR-3113: A TORSIONAL ULTRASONIC TECHNIGUE FOR LWR LIGUID LEVEL MEASUREMENT.
Licensed Fuel Facility NUREG-0430 VO3 NO1: 8.ICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. January 1982 - June 1982.
Licensed Operating Reactors NUREG-OO2O VO6 N11: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of October 31,1982.(Greg Book)
NUREG-OO2O VO6 N10: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of September 30,1982.(Greg Book) l NUREG-OO20 VO6 N09: LICENSED DPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of August 31,1982.(Greg Book)
Licensee Contractor And Vendor Inspection NUREG-OO40 VO6 NO4: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report,0ctober 1982 - December 1982.
Licensee Event Report NUREG/CR-2OOO V02 N1: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of January 1983.
NUREG/CR-2OOO V01N12: LICEldSEE EVENT REPORT (LER) COMPILATION: For Month Of December 1982.
NUREC/CR-2000 V02 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month 0f Febrvarg 1983.
Licensing Actions NUREG-0748 VO3 NO1: OPERATING REACTOR 8 LICENSING ACTIONS
SUMMARY
. Data As Of January 31,1983.(Grange Book) 114 I
(
NUREG-0748 VO3 NO2: OPERATING LICENSING ACTIONS
SUMMARY
. Data As Of February 28, 1983.(Orange Book)
Licensing Regulation NUREQ/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REGULATION DEVELOPMENT.
Licensing Requirements NUR EG-0959: USER'S GUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
Licensing Reviews NURE3-0580 V12 NOI: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of January 31,1983. (Blue Book)
NUREG-0580 V12 N12: REQULATONY LICENSING STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Blue Book)
Licensing Safety NUREG-0826 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION PROGRAM-HADDAM NECK PLANT. Docket No. 50-213.(Connecticut Yankee Atomic Power Company)
Licensing NUREG-0580 V12 NO2: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Blue Book)
Limestone NUREG/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE CONCRETE INTERACTIONS.
Limiting Conditions For Operation NUR EQ/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS.
Liquid Effluents NUREQ/CR-2907 VO1: RADIDACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
Liquid Metal Fast Breeder NUR EQ/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF TE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
Liquid Pool NUREG/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CCMVECTION HEAT TRANSFER OF INTERNALLY MATED LIQUIDS.
Liquid Water Jets NUREQ/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
Loads NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
Loose-Parts Monitors NUREQ/CR-3OOS: AUDITORY PERCEPTION IN LODSE-PARTS MONITORING.
Loss Of Benefits NUREQ/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT OUTAGES. Vol 2: Appendixes.
Loss-Of-Coolant Accident NUR EQ/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE B-4.
NUREQ/CR-2874 VO2: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report.
April 1 - June 30,1982.
NUREC/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Users ' Manual.
NUREQ/CR-2847: C00AP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL SYSTEM EVALUATION CODE.
NUREC/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Programmers Manual.
NUREC/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT 115
e i
ANALYSIS OF NUCLEAR REACTOR VEBSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
NUREC/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT 1
SYSTEMS. Volume 4: Developmental Assessment And Data.
NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT I
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT i
SYSTEMS. Volume 1: Equations And Constitutive Models, NUREO/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATIDN PILLOWS TO i
DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
NUREG/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL.
l HYDRAULIC EXPERIMENT (TH-3).
Loss-Of-Coolant NUREQ/CR-3126: SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTS COMPARISON.
Loss-Of-Fluid Test
{=
NURE0/CR-0169 V21: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
Low Level Waste NUREQ/CP-OO28 VO3: PROCEEDINGS OF THE SYMPOSIUM ON LOW-LEVEL WASTE DISPOSAL: Facility Design, Construction,And Operating Practices.
Low-Flow Stability NUREG/CR-2998: A COMPARISON OF BWR STABILITY MEASUREMENTS WITH CALCULATIONS USING THE CODE LAPUR-IV.
Low-Level Waste Site NUREC/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REGULATION OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid i
Low-Level Sites, j
MDTT i
NUREC/CR-0169 V21: LOFT EXPERIMENTAL MEASUREMENTS UACERTAINTY ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
Material Control and Accounting NUREQ/CR-2935: EXAMPLES OF MC8A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
Mater ials Deformation NUR EO/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
l Materials Engineering NUREG-0975 VO1: COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS l
ENGINEERING BRANCH DIVISION OF ENGINEERING TECHNOLOGY. Annual Report For FY 1992.
Maxey Flats NUREQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
Mechanistic Drg-Pressure-Containment l
NUREQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
q Migration NUREO/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL 1
AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
NUREO/CR-2863: VALENCE EFFECTS ON AD90RPTION: A PRELIMINARY ASSESSMENT l'
NUREQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
l Modular Drag-Disc Turbine Transducer l
NUREQ/CR-0169 V21: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY I
ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
Molten Sodium l
NUREG/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE l
CONCRETE INTERACTIONS.
116
Mult1 rod Burst Test NUREQ/CR-2968: EXPERIMENT DATA REPORT FOR MULTIRGD BURST TEST (MRBT)
BUNDLE B-4.
NDE NUREC/CR-2716 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report. July -September 1982.
NRC/DAE Data Bank Program NUREQ/CR-2531 RO1: INTRODUCTDRY USERS MANUAL FOR THE US NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
Natural Circulation Cooling NUR EG/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATIDN AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No. 12.
Natural Circulation NUR EG/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
NUR EG/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
Nemaha Ridge NUREC/CR-3109: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA-Final Report.
Network Stations NUR EC/CR-3080: NETWORK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA.
Neutralizing Agents NUREC/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT OF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
Neutron Exposure NUREG/CR-2005 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
NUREG/CR-2005 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Guarterly Progress Report. April 1982 - June 1982.
NUREG/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
Nomograms NUR EG-0851: NOMOGRAMS FOR EVALUATION OF DOSES FROM FINITE NOBLE CAS CLOUDS.
Nondestructive Evaluation NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
Nondestructive Examination NUREG/CR-2716 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report. July -September 1982.
Nuclear Durst i
NUREO/CR-3069 VO1: INTERACTION OF. ELECTROM/4GNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREG/CR-3069 VO2: INTERACTION DF ELECTRDMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
Nuclear Power Plant Sites NUf!EG-0965: NRC INVENTORY OF DAMS.
Nuclear Research Resources NUR EG/CR-3212: SELECTED REVIEW AND EVALUATION OF U.S.
SAFETY RESEARCH VIS-A-VIS F0F.EIGN SAFETY RESEARCH FOR NUCLEAR PDWER PLANTS.
Nuclear Waste Repositories NUREC/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
DAETS NUREQ/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
117
Offsite Response NUREC/CR-2925: IN-PLANT CONSIDERATIONS FOR DPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
Oklahoma City Uplift NUREO/CR-3109: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA-Final Report.
Operating Reactors Licensing Actions NUREG-0748 VO2 N12: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of December 31,1982.(Orange Book)
NUREG-0748 VO3 NO2: OPERATING LICENSING ACTIONS
SUMMARY
. Data As 0F February 28, 1983.(Orange Book) 1 NUREG-0748 VO3 NO1: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data i
As Of January 31,1983.(Orange Book)
Operating Units Status NUREG-OO2O VO6 NO9: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of August 31,1982.(Greg Book)
NUREG-OO2O VO6 N11: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of October 31,1982.(Greg Book)
NUREG-OO2O VO6 N10: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of September 30,1982.(Greg Book)
Operator Action Event Trees NUREO/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE GUIDELINES. Volume 2: Applications To Westinghouse Plants.
NUREQ/CR-3177 VO1: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 1: Methodologies.
Operator Action Tree NURES/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
Operator Performance NUREO/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT DPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
Operators NUREO/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TIME RELIABILITY CORRELATION.
NUREO/CR-3114: PROCEEDINGS OF WORKSHOP ON COGNITIVE MODELING OF NUCLEAR PLANT CONTROL ROOM OPERATORS. August 15-18,1982,Dedham, Massachusetts.
Outages NUREG/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT GUTAGES. Vol 2: Appendixes.
Overcooling NUREO/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERCOOLING ACCIDENTS: A PARAMETRIC ANALYSIS.
PES NUREG/CR-3063: A SUMMAdY OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
PRA I
NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT i
OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
Packing Material 1.1 -
NUREO/CR-2482 VO3: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask National Waste Package Program. April 1982 - September 1982.
Penetrations NUREO/CR-3135: BUCKLING INVESTIGATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINGS UNDER UNSYMPEETRICAL AXIAL LOADS.
Personnel Dosimetry NUREO/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
NUREC/CR-2892: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY BERVICES: A Revised Procedures Manual.
Personnel Neutron Dosimetry NUREO/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final 118
Report Of Substask B: Dosimeter Response.
Petitions For Rulemaking NUREG-0936 VO1 NO4: NRC REGULATORY AGENDA.Guarterly Report. September
-December 1982.
Physical Inventory NUREG/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
Piping Components NUREG/CR-2703: INVESTIGATIDN OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIQUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
Piping Flaws NUREG/CR-3059: PARAMETRIC CALCULATIONS OF FATIGUE CRACK GROWTH IN I
PIPING.
Piping Reliability NUREG/CR-2001: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REGULATION DEVELOPMENT.
Piping NUREG/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING RESPONSES.
Plant Systems Analysis NUREG/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES GUIDE.
Plasma Torch NUREG/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPDRT FOR JULY - SEPTEMBER 1982.
Policy And Planning Guidance NUREG-0885 102: US NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING GUIDANCE 1983.
Post-TMI Requirements NUREG-0737 SO1: CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS: REGUIREME NTS FOR EMERGENCY RESPONSE CAPABILITY.
Predictor Displays NUREQ/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL.
Pressure Vessel Clad Surface NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
Pressure Vessel Surveillance Dosimetry l
NUREG/CR-2805 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report. April 1982 - June 1982.
Pressure Vessel l
NUREC/CR-2895: PWR PRE 5 CURE VESSEL INTEGRITY DURING DVERC00 LING l
ACCIDENTS: A PARAMETRIC ANALYSIS.
NUREG/CR-2DO5 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September i
30, 1982).
NUR EG/CR-2755: PACKING MATERIAL TESTING REQUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
Pressure NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
Primary Coolant NUREG/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 4: Developmental Assessment And Data.
NUREG/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Us ers ' Manual.
119
1 NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1:Eguations And Constitutive Models.
NUREG/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT l
SYSTEMS. Volume 5: Programmers Manual.
NURE9/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT i
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
Pricary and Secondary Stress Components NUREQ/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING RESPONSES.
Probabilistic Risk Assessment NUREQ/CR-23OO VO1: PRA PROCEDURES QUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREC/CR-3028: A REVIEW OF THE LIMERICK CENERATING STATION PROBABILISTIC RISK ASSESSMENT.
NUREQ/CR-23OO VO2: PRA PROCEDURES GUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUR EG/CR-2728: INTERIM RELIABILITY EVALUATION PRDORAM PROCEDURES OUIDE.
Procedures Guide NUREQ/CR-23OO VO1: PRA PROCEDURES GUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREC/CR-23OO VO2: PRA PROCEDURES QUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUR EG/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PRDCEDURES CUIDE.
Procedures Manual NUREQ/CR-2892: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: A Revised Procedures Manual.
Pumps NUR EG/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S.
Commercial Nuclear Power i
Plants, January 1,1972 Through September 30,1980.
l Cusdrez Report NUR EG-0948: SPECIAL INSPECTION REPORT OF GUADREX CORPORATION REPORT ON l
DESIGN REVIEW OF BROWN 4r ROOT ENGINEERING WORK FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And Power Company)
Guality Assurance Programs NUREG-0383 VO1 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR R ADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
NUREC-0383 VO2 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Certificates of Compliance.
NUREC-0303 VO3 RO2: DIRECTORY OF CERTIFICATES OF COMPLIANCE FDR R ADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Quality j
Assurance Programs For Radioactive Material Packages.
RALDC NUREQ/CR-2843 VO1: THERMAL / HYDRAULIC ANALYSIS RESEARCH PRDORAM.Guarterly Report. January - March 1982.
RELAPS NUR EQ/CR-2887: RELAP5 ASSESGMENT: FLECHT SEASET STEAM QENERATOR TEST l
23402.
Radiation Monitoring NUREG-0837 VO2 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1982.
Radiation NUR EO/CR-3120: GUALITY ASSURANCE FDR MEASUREMENTS OF IONIZING RADIATION. Annual Report For FY 1982.
120
. ~
Radioactive Material Packages NUREG-0383 VO3 RO2: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Guality Assurance Programs For Radioactive Material Packages.
NUREG-0383 VO2 ROS: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKACES. Certi ficates of Compliance.
NUREG-0383 VO1 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR j
RADI0 ACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
Radioactive Materials NUREO/CR-2907 VO1: RADI0 ACTIVE MATERIALS RELEASED FROM NUCLEAR POWER i
PLANTS - 1980.
l Radioactive Release l
NUREQ/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
NUREQ/CR-2974: USER'S MANUAL FOR LPGS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
Radioactive Waste Repository NUR EC/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSESSMENT.
Radioactive Waste Streams NUREQ/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MICRATION.
Radioactive Waste NUREQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH CEOLDOIC DISPOSAL OF NUCLEAR WASTE.
NUREQ/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIGUES FOR LOW-LEVEL l
RADI0 ACTIVE WASTE DISPOSAL.
Radionuclide Inventories NUREQ/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
Radionuclide Transport NUREQ/CR-3076: COMPUTER PREDICTION OF SUBSURFACE RADIONUCLIDE TRANSPORT
-AN ADAPTIVE NUMERICAL METHOD.
I Radionuclides NUREC/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL L
l AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
Radon Daughter Measurements NUREC/CR-3106: COMPARISON OF FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
Radon Measurements NUREQ/CR-3106: COMPARISON OF FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
Radon NUR EG/CR-3166: RECOMMENDED PROCEDURES FOR MEASURING RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS.
Rank NUREG/CR-3115: EXPERT OPINION AND RANKING METHODS.
Reactor Pressure Vessels NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
Reactor Safety Issuss NUREQ/CR-2800: GUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
Reactor Safety Research NUREQ/CR-2716 VO3: REACTOR SAFETY RESEARCH PROGRAMS Guarterly Report. July -September 1982.
NUREG/CR-2531 RO1: INTRODUCTORY USERS MANUAL FOR THE US NUCLEAR REQULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
Reactor Safety Study 121
NURE0/CR-3028: A REVIEW OF THE LIMERICK QENERATING STATION PROBABILISTIC RISK ASSESSMENT.
Reactor Safety NUREO/CR-2774 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July -
September 1982.
Reactor Trip Circuit Breakers NUREG-0977: NRC FACT-FINDING TASK FORCE REPORT DN THE ATWS EVENT AT SALEM NUCLEAR GENERATING STATION, UNIT 1.ON FEBRUARY 22'AND 25,1983.
Reactor Vessels NUR EG/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FCR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
Recriticalities NUR EO/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
Reflux Condensation Cooling NUREG/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No.
12.
'+
Regulatory Actions NUREG-0943: THREADED FASTENER EXPERIENCE IN NUCLEAR POWER PLANTS.
Regulatory Agenda NUREG-0936 VO1 NO4: NRC REGULATORY AGENDA.Guarterly Report. September
-December 1982.
Regulatory And Technical Reports NUREG-0304 VO7 NO4: REQULATORY AND TECHNICAL REPORTS. Compilation For 1982.
Regulatory Licensing NUREG-0580 V12 NO2: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Blue Book)
NUREG-0580 V12 NO1: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of January 31,1983. (Blue Book)
NUREG-0580 V11 N12: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Blue Book)
Release NUREQ/CR-2BO9 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
NUREQ/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report, April 1 - June 30,1982.
NUREQ/CR-2907 VO1: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
Repository NUREQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report
)
On Waste Package Verification Tests.
Ring-StiPfened Shells NUREQ/CR-3135: BUCKLING INVESTIGATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
Risk NUREG/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No.
14.
Rod Bundle NUR EQ/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
Rules NUREG-0936 VOR NO4: NRC REQULATORY AGENDA.Guarterly Report. September
-December 1982.
Rupture NUREQ/CR-2659: IODINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE 122
t.
BREAK WITH CONCURRENT RUPTUREG OF STEAM QENERATOR TUBES.
SAFE NUREQ/CR-3105: SECURORS APPLICATION TO A GENERIC NUCLEAR POWER PLANT.
SARP NUREG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN.
SAS3D NUREG/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERC00 LING ACCIDENT IN A L ARGE, HETEROGENEDUS-CORE, LIGUID-METAL-COOLED FAST BREEDER REACTOR.
SCEPTRE NUREG/C9-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
SECURORS NUREQ/CR-3105: SECURORS APPLICATION TO A GENERIC NUCLEAR POWER PLANT.
I SEISM NUREC/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH PROGRAM, PHASE I FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
SEP NUR EG-0822: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION PROGRAM. 0YSTER CREEK NUCLEAR GEERATING STATION. Doc k et No.
50-219. (GPU Nuclear Corporation And Jersey Central Power 8: Light Company)
NUREG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATION PROGRAM
- A USE OF PROBABILITY IN DECISION MAKING.
SIMMER-II NUREQ/CR-2902: FLOW REGIME MODELING STUDY FOR THE EIMMER-II LMFBR SAFETY CODE: CLAD RELOCATION.
SMACS NUREQ/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH--PROGRAM, PHASE I l
FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
SRSS NUREQ/CR-3086-INVESTIGATION OF TE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETkEEN PRIMARY AND BECONDARY PIPING RESPONSES.
SSMP NUREG/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROGttAM. Progress Report No. 14.
[
NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMERCI AL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREC/CR-3069 VO2: INTERACTION OF ELECTROMAGNETIC PULSE WITH-COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
Safeguards Summary Event, List NUREG-0525 R06: SAFEQUARDS
SUMMARY
EVENT LIST (SSEL).
l Safeguards-Related Events NUREG-0525 R06: SAFEGUARDS
SUMMARY
EVENT LIST (SSEL).
Safety Analysis NUREC/CR-2902: FLOW REGIME MODELING STUDY FOR THE SIMMER-II LMFBR SAFETY CODE: CLAD RELOCATION.-
Safety Assessment Report NUREG-0826 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC l
EVALUATION PROGRAM-HADDAM NECK PLANT. Docket No. 50-213.(Connecticut Yankee Atomic Power Company) l Safety Evaluation Report NUREG-0422 SO6: SAFETY EVALUATION REPORT REL/sTED TO OPERATION OF MCGUIRE NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-369 And 50-370.(Duke Power Company)
NUREG-0420 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERAT, ION OF SHOREHAM NUCLEAR POWER STATION, UNIT NO.
- 1. Docket No.s 50-322.'(Long Island Lighting Company)
NUREG-0776 SO5: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF 123
4 w+e*
mvin-w---w--rw---
wi.ewi----
s-*-
s
i-----
m--w--e. - - - - -
w v-m-r--
SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-387 And 50-388.(Pennsgivania Power & Light Compangs Allegheng Electric Cooperative, Incorporated) l NUREG-0796 303: SAFETY EVALUATIDN REPORT RELATED TO THE DPERATION OF ENRICO FERMI ATOMIC POWER PLANT, UNIT NO.
- 2. Docket No. 50-341.
(Detroit Edison Company)
NUREG-0857 SO4: ~ SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF P ALD VERDE N8) CLEAR QENERATING STATION, UNITS 1,2 AND 3. Doc ket Nos. STN 50-528,STN 50-529 And STN 50-530.(Arizona Public Service Compang,et al.)
NUREG-0852 501: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN OF THE STANDARD NUCLEAR STEAM SUPPLY REF,ERENCE SYSTEM CESSAR SYSTEM l
- 80. Docket No. STN 50-470.(CombustionEEngineering, Incorporated) t NUREG-0876 S02: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF BYRON STATION UNITS 1 AND 2. Doc ke t Nos. STN 50-454 And STN 50-455.
(Commonwealth Edison Company)
NUREG-0896: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF l
SEABROOK STATION, UNITS 1 AND 2. Docket Nos. 50-443 And 50-444.(Public Service Compang Of New Hampshire,et al.)
i NUREG-0947: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE TEXAS A&M UNIVERSITY TRICA REACTOR. Docket No. 50-128. License R-83.
NUREC-0954: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF C ATAWBA NUCLEAR STATION, UNITS 1 AND 2. Doc ket Nos. 50-413 And 50-414.
(Duke Power Ccapangeet al.)
NUREC-0968 VC2: SAFETY EVALUATION REPORT RELATED TO THE CONSTRUCTION OF
(
THE CLINCH RIVER BREEDER REACTOR PLANT. Appendices A-H. Doc ket No.
50-537.(U.S. Department of Energy, Tennessee Valley Authority And Project Management Corporation)
' NUR EG-0966: SAFETY EVALUATIOff REPORT RELATED TO THE D2/D3 STEAM QENERATOR DESIGN MODIFICATION.
NUREG-0968 VO1: SAFETY EVALUATION REPORT RELATED TO THE CONSTRUCTION OF THE CLINCH RIVER BREEDER REACTOR PLANT. Main Report. Docket No.
i 50-537.(U.S. Department.of Energy, Tennessee Valley Authority And l
Project Management Corporation) l NUR EQ/CR-2806: A KINETIC MODEL FOR YHE CHLORINATION OF POWER PLANT CDOLING WATERS.
l Safetg Functions NUREQ/CR-3177 VO1: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE CUIDELINES. Volume 1: Methodologies.
Safety Issue l
NUREG/CR-2800: QUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
Safety Research L
NWEG-0963: REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION I'
/ SAFETY RESEARCH PROGRAM FOR FISCAL YEARS 1984 AND 1985.
j NUREG/CP-OO41 VO6: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY
'RESEARCH INFORMATION MEETINC.
)
i NUREC/CP-OO41 VO4: PROCEEDINGS OF THE TENTH WATER REACTDR SAFETY 1
RESEARCH INFORMATION MEETINC.
NUREQ/CP-OO41 VG.I; PROCEEDINGS OF THE TENTH WATER REACTOR BAFETY i
RESEARCH INFORMA7 ION MEETINC.
NUREQ/Ct -OO41 VOli ' PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY 8
RESEARCH INFORMATION MEETING.
- NUREQ/CP-OO41 VO2: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
9
.HUREQ/CP-OO41 VO5: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY l-RESEARCH INFORMATION MEETINC.
_NVR.EC/CR-3212: SELECTED REVIEW AND EVALUATION DF U.S.
SAFETY RESEARCH VIS-A-VIS FOREIGN SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
N 124
^
1
')
1 4
~
NUREQ/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REQULATDRY RESEARCH. Guarterly Progrest ' Report. April 1 -June 30,1982.
Safety System Cosponents NUR EO/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND BURVEILLANCE REGUIREMINTS FOR ETAN28Y SAFETY SYSTEMS.
Safety-Relates Coerator Actions l
NURE0/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
Safety-Related Programs NUREQ/CP-OO27 VO3: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
NUREQ/CP-OO27 VO1: PROCEEDINGS OF THE INTERNATIONAL MEETING"ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
NUREO/CP-OO27 VO2: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Hold At Chicago, Illinois, August 29 -September 2,1982.
Safety NUREG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN.
NUR EG-0943: THREADED FASTENER EXPERIENCE IN NUCLEAR POWER PLANTS.
Salt Repositories NUREQ/CR-2482 VO3: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1. 1 -
National Maste Paciagu1 Program. April 1982 - September 1982.
Securi.tg. Officer NUR E0/CR-3105: SECURORS APPL 10ATION TO A GENERIC NUCLEAR POWER. PLANT.
Seismic Analysis, NUREQ/CR-1120 ViO: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Repoet'No.
14.
. Seismic Hazard NUREG/CR-2015 VO3: ' SEISMIC SAFETY MARGINS RESEARCH PROGRAM, PHABE I
- FINAL REPORT-DEVELOPMENT 09 SEISMIC INPUT (PROJECT II).
Seismic Licensing Requirements,i NUREQ/CR-1120 V10: SEISMIC.3AFETY MARGINS RESEARCH PROGRAM. Progress Report No.
14.
Seismic Safety Margins Research Program.
NUREO/CR-2015 VO3: SEISMIC SAFETY MARCINS RESEARCH PROGRAM PHASE I FINAL REPORT-DEVELOPMENT DF SEISMIci!NPUT (PROJECT II).
NUREQ/CR-1120 V10: SEISMIC" SAFETY MARGINS RESEARCH PROGRAM. Progress Report No.
14.
Seismicity NUR EO/CR-3079: EARTHOUAME HAZARD STUDIES IN NEW YORK STATE AND ADJACENT.
AREAS. Final Report. April 1976 - June 19821 NURES/CR-3109: SEISMICITY AND TECTONIC RCLATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA-Final Report.
Seismograph Network NUREQ/CR-3145 VO1: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
Semiscale Mod-2A NUREQ/CR-3126: SEMISCALE MOC-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTE COMPARISON.
Severe Accident Analyses I
NUREQ/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY DTUDIES FOR THE DIVISION.0F ACCIDENT EVALUATION. Guarterly Progress Report, April 1._- June 30,1982.
Severe Accident Research Plan NUREG-0900: NUCLEAR FLANT SEVERE ACCIDENT RESEARCH PLAN.
Shaft NUREG/CR-2654: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR 125 I
!i..
- _ - _ ~ -
l i
HIGH LEVEL NUCLEAR WASTE (HLW) DGEP GEOLOGIC REPOSITORIES.
Shallow Land Burial NURE0/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIOUES FOR LOW-LEVEL RADIDACTIVE WASTE DISPOSAL.
Shear Stresses NURE0/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
Shift NURE0/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER PLANTS: THE FOREIGN EXPERIENCE.
Shipments NUREG-0393 VO2 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERI ALS PACKAGES. Certificates of Compliance.
NUREG-0383 VO1 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
NUREG-0383 VO3 RO2: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR R ADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Guality l
Assurance Programs For Radioactive Material Packages.
Shutdown NUREG/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER 3
NATURAL CIRCULATION CONDITIONS.
NURE0/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT GUTAGES. Vol 2: Appendixes.
Site Characteristics NUREQ/CP-OO28 VO3: PROCEEDINGS OF THE SYMPOSIUM ON LOW-LEVEL WASTE DISPOSAL: Facility Design, Construction.And Operating Practices.
Site Characterization NURE0/CR-2983: SELECTED HYDROLOGIC AND OE0 CHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
Sites NUREC/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
Soi!
NURES/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON 9IMULATED QROUND MOTIONS.
NUREG/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL l
AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
Solid State Protection System NUREG-0977: NRC FACT-FINDING TASK FORCE REPORT ON THE ATWS EVENT AT SALEM NUCLEAR GENERATING STATION, UNIT 1,0N FEBRUARY 22 AND 25,1983.
l Solid Waste Shipments NURE0/CR-2907 Vol: RADI0 ACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
Stability i
(
NUREQ/CR-2998: A COMPARISON OF BWR STABILITY MEASUREMENTS WITH CALCULATIONS USING THE CODE LAPUR-IV.
Standard Nuclear Steam Suppig NUREG-0852 901: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN OF THE STANDARD NUCLEAR STEAM SUPPLY REFERENCE SYSTEM CESSAR SYSTEM
- 80. Docket No. STN 50-470.(Combustion Engineering. Incorporated)
Standards Development NUREO-0566 VO2 NO4: STANDARDS DEVELOPPENT STATUS Gl#9tARY REP (MtT. Data As Of December 31,1982.(Green Book)
Station Delags NUREO/CR-3080: PETWORK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA.
Steam Generator Tubing NURE0/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM QEPERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending June 30,1982.
126
Steam Generator NUREC-0966: SAFETY EVALUATIDN REPORT RELATED TO THE D2/D3 STEAM QENERATOR DESIGN MODIFICATION.
NUREQ/CR-2887: RELAP5 ASSESSMENT: FLECHT SEASET STEAM QENERATOR TEST 23402.
NUREQ/CR-2716 V03: REACTOR SAFETY RF0EARCH PROGRAMS. Guarterly Reoort. July -September 1982.
NUR EQ/CR-2993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM QENERATOR AT THE MAINE YANKEE NUCLEAR POWER STATION.
Steam Line Break NUREQ/CR-2659: IODINE TRANSPORT PREDICTED FOR A POSTULA* ED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM QENERATOR TUBES.
Strategic Special Nuclear Material NUREQ/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
Strength NUR EQ/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTOR CORE SUPPORT ORAPHITE--PART III.
Stress Corrosion Cracking Failure NUR EQ/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENGINC REQULATION DEVELOPMENT.
Structures NUR EQ/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT DF AVAILABLE MARGINS INHERENT IN FLOOD PROTECTION OF NUCLEAR PDWER PLANTS.
Studs NUR EQ/CR-2993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM QENERATOR AT THE MAINE YANKEE NUCLEAR POWER STATION.
Summary Information NUREG-0871 V02 N01:
SUMMARY
INFORMATION REPORT.0ctober 1 - December 31, 1982. (Brown Book)
Surface Cooling NUR EQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE CDOLING ASSOCIATED WITH QEOLOGIC DISPOSAL OF NUCLEAR WASTE.
Surveillance Requirements NUREQ/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS.
Synth etic Aperture Processing NUREQ/CR-2703: INVESTICATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIQUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND P1 PING COMPDNENTS.
Systematic Evaluation Program NUREQ-0485 V04 N12: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Buff Book)
NUREC-0485 V05 NO2. SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPDRT. Data As Of February 28,1983.(Buff Book)
NUREG-0485 V05 N01: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of January 31,1983.(Buff Book)
NUR EC-0824: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATIDN PROGRAM-MILLSTONE NUCLEAR POWER STATION. UNIT 1. Docket No.
50-245.(Northeast Nuclear Energy Company)
NUREG-OR26 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT. SYSTEMATIC EVALUATION PROGRAM-HADDAM NECK PLANT. Doc ket No. 50-213.(Connecticut Yankee Atomic Power Company)
NUR EC-0823: INTEGRATED SAFETY ASSESSMENT SYSTEMATIC EVALUATION PRDORAM, DRESDEN NUCLEAR POWER STATION UNIT 2. Docket No. 50-237. (Commonwealth Edison Company)
NUREC-0822: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION PROGRAM. OYSTER CREEK NUCLEAR QENERATING STATION. Docket No.
50-219.(GPU Nuclear Corporation And Jersey Central Power & Light Company) 127
NUREG-0825: INTEGRATED PLANT SAFETY ASSESSMENT REPDRT, SYSTEMATIC EVALUATION PRDORAM-YANKEE NUCLEAR POWER STATIDN.Dochet No.50-029.(Yankee Atomic Electric Company)
NUR EG-0967: SEISMIC HAZARD REVIEW FDR THE SYSTEVI, EVALUATION PROGRAM
- A USE OF PROBABILITY IN DECISION MAKING.
Systems Modeling NUREQ/CR-23OO VO2: PRA PROCEDURES GUIDE. A Guide To The Perf ormance DF Probabilistic Risk Assessments For Nuclear Power Plants.
1 NUREQ/CR-23OO YO1: PRA PROCEDURES QUIDE.A Guide To The Performance OF Probabilistic Risk Assessments For Nuclear Power Plants.
TH-3 NUREC/CR-2527: LDCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL HYDRAULIC EXPERIMENT (TH-3).
TLD NUREG-0837 VO2 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1982.
NUREG/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IDNIZING RADIATIDN. Annual Report For FY 1982.
Tailings NUREG/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT DF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
NUREC/CR-2856: A REVIEW OF FUCITIVE DUST CONTROL FOR URANIUM MILL TAILINCS.
Technical Specifications NUREG-0964: TECHNICAL SPECIFICATIONS FOR MCGUIRE NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-369 And 50-370.(Duke Power Company)
Tectonic Processes NUR EQ/CR-3079: EARTHGUAKE HAZARD STUDIES IN NEW YDRK STATE AND ADJACENT AREAS. Final Report. April 1976 - Juna 1982.
Tectonic NUREC/CR-3109: SEISMICITY AND TECTDNIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA-Final Report.
l Thermal Effects NUREQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE CDOLING ASSOCIATED WITH OEDLOGIC DISPOSAL OF NUCLEAR WASTE.
Thermal Nuc lear Reac tor Saf ety NUREG/CP-OO27 VO2: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
NUREC/CP-OO27 VO3: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1992.
NUREC/CP-OO27 VO1: PROCEEDINGS OF THE INTERNATIDNAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
Thermal Shock i
NUR EC/CR-2895: PWR PRESSURE VESSEL INTEORITY DURING DVERCOOLINC ACCIDENTS: A PARAMETRIC ANALYSIS.
Thermal / Hydraulic Analysis NUREO/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
I NUREG/CR-2843 VO1: THERMAL / HYDRAULIC ANALYSIS RESEARCH PRDORAM.Guarterly Report. January - March 1982.
Thermal / Hydraulic Response NUREQ/CR-2887: RELAP5 ASSESSMENT:FLECHT SEASET STEAM GENERATOR TEST 23402.
Thermal / Hydraulic NUREG/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FDR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Users' Manual.
NUREQ/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THERMAL 128 I
i HYDRAULIC EXPERIMENT (TH-3).
NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY CDOLANT SYSTEMS. Volume 1:Eguations And Constitutive Models.
I NUREQ/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMG. Volume 4: Developmental Assessment And Data.
i NUREQ/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
NUREQ/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
I NUREQ/CR-3046 VO5: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Programmers Manual.
NUREQ/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLDQY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
NUREQ/CR-2774 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July -
September 1982.
Thermoluminescent Dosimeter NUREG-0837 VO2 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1992.
Threaded-Fastener NUREC-0943: THREADED FASTENER EXPERIENCE IN NUCLEAR POWER PLANTS.
Title List d
NUREG-0540 VO4 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1982.
NUREG-0540 VO4 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY j
AVAILABLE. November 1-30,1982.
NUREC-0540 VO4 N10: TITLE LIST OF DOCUMENTS MADE PUSLICLY AVAILABLE. October 1 -31,1982.
NUREG-0540 VO4 NO9: TITLE LIST (F DOCUMENTS MADE PUBLICLY AVAILABLE.Septomber 1-30,1982.
Torsional Ultrasonic Pulses NUREQ/CR-3113: A TORSIONAL ULTRASONIC TECHNIGUE FOR LWR LIGUID LEVEL MEASUREMENT.
Track Etch NUREC/CR-3106: COMPARISON OF FIVE-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONO-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
Transients NUREO/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY CDOLANT l
SYSTEMS. Volume 3: Users ' Manual.
NURE9/CR-3046 VO2: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
l NURE0/CH-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY CDOLANT RYSTEMS. Volume 4: Developmental Assessment And Data.
NUREQ/CR-3046 VOS: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Pragrammers Manual.
NURE0/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CDDE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEnS. Volume 1: Eguations And Constitutive Models.
NUREO/CR-2895: PWR PRESSURE VESSEL INTEORITY DURING DVERCOOLING ACCIDENYS: A PARAMETRIC ANALYSIS.
Transport NUREO/CR-2909 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PRDORESS REPORT FOR JULY - SEPTEMBER 1982.
129
l l
Trench NUREO/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIGUES FOR LOW-LEVEL RADIDACTIVE WASTE DISPOSAL.
NURE0/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SDIL AT THE MAXEY FLATS,MENTUCKY, WASTE-BURIAL SITE.
Ultrasonic Techniques NURE0/CR-2979: DETECTION OF SMALL-GIZED NEAR-SURFACE UNDER-CLAD CRACKS i
FOR REACTOR PRESSURE VESSELS.
Undercooling Accident NUREG/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERC00 LING ACCIDENT IN A LARGE, HETER 00ENEDUS-CORE,LIGUID-METAL-COOLED FAST BREEDER REACTOR.
Underground Test Facility i
NURE0/CR-3065 V02: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GE0 LOGIC REPOSITORIES. Appendices.
NURES/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP GEOLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
N9RE0/CR-3065 V01: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP OEOLOGIC REPOSITORIES. Main Report.
Unresolved Safety Issues NUREG-0606 VOS N01: UNRESOLVED BAFETY ISSUES
SUMMARY
. Data As 0F February 19, 1983.(Aqua Book)
Unsymeetrical Axial Loading i
NURE0/CR-3135: BUCKLING INVESTIGATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINOB UNDER UNSYMMETRICAL AXIAL LDADS.
Upper Mantle Structure NUR EG/CR-3079: EARTHGUAME HAZARD STUDIES IN NEW YORK STATE AND ADJACENT AREAS. Final Report. April 1976 - June 1982.
Uranium Mill Tailings NUR EG-0965: NRC INVENTORY OF DAMS.
NUR E0/CR-3078: MODEL EVALUATION OF SEEPAGE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
NUREO/CR-3166: RECOMMENDED FROCEDURES FOR MEASURING RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIDACTIVE MATERIALS.
NUR E0/CR-2856: A REVIEW OF FUGITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
Uranium Tailings NURE0/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE j
TREATMENT OF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
User's Guide NURE0/CR-2729: USER 'S GUIDE TO.BFR. A Computer Code Based On The 1
l Binomial Failure Rate Common Cause Model.
User's Manual NURES/CR-3046 V03: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Users ' Manual.
NURE0/CR-2531 RO1: INTRODUCTORY USERS MANUAL FOR 'fHE US NUCLEAR REGULATORY CDPMISSION REACTOR SAFETY RESEARCH DATA BAM.
NUREO/CR-2974: USER'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
Valence State NUREG/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION M ASUREMENTS.
Valves i
NURE0/CR-2099: COMMON CAUSE FAULT RATES FOR PUPPS: Estimates Based On Licensee Event Reports At U.S. Commercial Nuclear Power Plants, January 1,1972 Through September 30,1990.
(
130
Vaporization NUREC/CR-2009 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
WASH-1400 NUREQ/CR-3028: A REVIEW OF THE LI:lERICK OENERATING STATION PROBABILISTIC RISK ASSESSMENT.
Waste Burial NUREG/CR-3032: STUDIES OF TRANSPOR' 0F WASTE RADIONUCLIDES THROUGH SOIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
Waste Disposal NUREO/CP-OO2B VO3: PROCEEDINGS OF THE SYMPOSIUM ON LOW-LEVEL WASTE DISPOSAL: Facility Design,Construrtion,And Operating Practices.
NUREO/CR-3130: IPsFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON GROUNDWATER MIGRATION.
Weste Form NUREQ/CR-2492 VO3: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask
- 1. 1 -
National Waste Packa2e Program. April 1982 - September 1982.
Waste Package NUREQ/CR-2482 VO3: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask
- 1. 1 -
National Waste Package Program. April 1982 - September 1982.
NUREG/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
Waste NUR EQ/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
NUREQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH OEOLOGIC DISPOSAL OF NUCLEAR WASTE.
Water Quality NUREQ/CR-2006: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT COOLING WATERS.
Water Reactor Safetg Research NUREC/CP-OO41 VO3: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NUREC/CP-OO41 VO1: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY l
RESEARCH INFORMATION MEETING.
NURE0/CP-OO41 VO5: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NURE0/CP-OO41 VO2: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NUREQ/CP-OO41 VO6: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY l
RESEARCH INFORMATION MEETINO.
NUREQ/CP-OO41 VO4: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
Water Table NUREQ/CR-3078: MODEL EVALUATION OF BEEPAGE FROM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
Wind Erosion NUREQ/CR-2856: A REVIEW OF FUGITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
l Workshop NUREO/CR-3114: PROCEEDINGS OF WORKSHOP DN COGNITIVE MODELING OF NUCLEAR PLANT CONTROL ROOM OPERATORS. August 15-18,1982,Dedham, Massachusetts.
Zircalog Cladding NURE0/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (PWIBT)
BUNDLE B-4.
131
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-propriate. Each entry is followed by a NUREG number and title of the report (s).
If further information is needed, refer to the main citation by NUREG number.
ADVISORY COMMITTEE (S)
ACRS - ADVISORY COMMITTEE ON REACTOR SAFECUARDS NUREG-0963: REVIEW AND EVALUATION OF THE NUCLEAR RECULATORY COMMISSION SAFETY RESEARCH PROGRAM FOR FISCAL YEARS 1984 AND 1985.
OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
REGION 1, OFFICE OF DIRECTOR NUREC-0837 V02 NO3: TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1982.
NUREG-0977: NRC FACT-FINDING TASK FORCE REPORT ON THE ATWS EVENT AT I
SALEM NUCLEAR CENERATING STATION, UNIT 1,ON FEBRUARY 22 AND 25,1983.
l REGION 4, OFFICE OF DIRECTOR NUREC-OO40 V06 N04: LICENSEE CONTRACIDR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report,0ctober 1982 - December 1982.
DIVISION OF RESIDENT, REACTOR PROJECT & ENGINEERING PROGRAMS NUREQ-0948: SPECIAL INSPECTION REPORT OF QUADREX CORPORATION REPORT ON DESIGN REVIEW OF BROWN & ROOT ENGINEERING WORK FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And Power Company)
EDO - OFFICE OF ADMINISTRATION DIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL NUREC-0304 VO7 N04: REGULATORY AND TECHNICAL REPORTS. Compilation For 1982.
NUREC-0540 V04 N09: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. September 1-30,1982, NUREG-0540 V04 N10: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. October 1 -31,1982.
NUREG-0540 V04 Nil: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. November 1-30,1982.
NUREC-0540 V04 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1982.
133
DIVISIDN OF RULES AND RECORDS NUREC-0936 VO1 NO4: NRC REGULATORY AGENDA.GuarterIg Report. September
-December 1982.
EDO - OFFICE OF STATE PROGRAMS OFFICE OF STATE PROCRAMS, DIRECTOR NUREG-0584 RO3 DRFT: ASSURINO THE AVAILABILITY OF FUNDS FDR DECOMMISSIDNING NUCLEAR FACILITIES.
EDO - OFFICE FOR ANALYSIS & EVALUATION OF DPERATIONAL DATA DIRECTOR'S OFFICE NUREG-OO90 VOS NO3: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. July -
September 1982.
OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)
DIRECTOR'S OFFICE. OFFICE OF INSPECTION AND ENFORCEMENT NUREC-0430 VO3 NO1: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. January 1982 - June 1982.
NUREC-0940 VOI NO4: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIDNS RESOLVED.GuarterIg Progress Report. October - December 1982.
INCIDENT RESPONSE & DEVELOPMENT BRANCH NUREC-0845: ACENCY PROCEDURES FOR THE NRC INCIDENT RESPONSE PLAN.
OFFICE OF NUCLEAR MATERI AL SAFETY & SAFEGUARDS OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEQUARDS. DIRECTOR NUREG-0525 R06: SAFECUARDS
SUMMARY
EVENT LIST (SSEL).
DIVISION OF FUEL CYCLE & MATERIAL SAFETY l
NUREG-0383 VO1 RO5: DIRECTDRY OF CERTIFICATES OF COMPLIANCE FOR l
RADIOACTIVE MATERIALS PACKACES. Summary Report Of NRC Approved Packages.
NUREG-0383 VO2 RO5: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKACES. Certificates of Compliance.
NUREC-0383 VO3 RO2: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS FACKACES. Summary Report Of NRC Approved Guality Assurance Programs For Radioactive Material Packages.
POWER REACTOR SAFEQUARDS LICENSING BRANCH NUREG-0907: ACCEPTANCE CRITERIA FOR DETERMINING ARMED RESPONSE FORCE SIZE AT NUCLEAR POWER PLANTS.
DIVISION OF WASTE MANAGEMENT NUREC-0959: USER 'S GUIDE FOR 10 CFR 61 IMPACT ANALYSIS CODES.
NUREG-0960 VO1: DRAFT SITE CHARACTERIZATION ANALYSIS OF THE SITE CHARACTERIZATION REPORT FOR THE BASALT WASTE ISOLATION PROJECT. Main Report And Appendices A ThrouDh D.
NUREC-0960 VO2: DRAFT SITE CHARACTERIZATION ANALYSIS OF THE SITE CHARACTERIZATION REPORT FOR THE BASALT WASTE ISOLATION PROJECT. Appendices E-W.
U. S.
NUCLEAR RECULATORY COMMIS9 ION 134
NRC - NO DETAILED AFFILIATION CIVEN NUREG-OB85 102: US NUCLEAR REQULATORY COMMISSION POLICY AND PLANNING OUIDANCE 1983.
NUREC-0965: MtC INVENTORY OF DAMS.
OFFICE OF NUCLEAR REQULATORY RESEARCH (POST 4/05/81)
DFFICE OF NUCLEAR REQULATORY RESEARCH, DIRECTOR 1
NUREG-0900: NUCLEAR PLANT SEVERE ACCIDENT RESEARCH PLAN.
MATERIALS ENGINEERING BRANCH NUREG-0975 VO1: COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS 3
ENGINEERING BRANCH DIVISION OF ENGINEERING TECHNOLOGY. Annual Report For FY 1982.
EDO-RESOURCE MANAGEMENT OFFICE OF RESOURCE MANAGEMENT, DIRECTOR NUREG-0485 VO4 N12: SYSTEMATIC EVALUATION PROGRAM STATUS SLNMARY REPORT. Data As Of December 31,1982.(Buff Book)
NUREG-0485 VO5 NO1: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of January 31,1983.(Buff Book)
NUREQ-0953: FY 1984/85 BUDGET ESTIMATES.
DIVISION OF DATA AUTOMATION L MANAGEMENT INFORMATION NUREG-0485 VO5 NO2: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Buff Book)
MANAGEMENT INFORMATION BRANCH NUREG-0020 VO6 NO9: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of August 31,1982.(Greg Book)
NUREG-OO2O VO6 N10: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of September 30,1982.(Greg Book)
NUREG-OO2O VO6 N11: LICENSED OPERATING REACTORS STATUS
SUMMARY
. REPORT. Data As Of October 31,1982.(Greg Book)
NUREC-0390 VO6 NO2: TOPICAL REPORT REVIEW STATUS. Data As Of January 20,1983. (Blue Book)
NUREG-0566 VO2 NO4: STANDARDS DEVELOPMENT STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Green Book) i NUREG-05BO V11 N12: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of December 31,1982.(Blue Book)
I NUREG-0580 V12 NO1: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of January 31,1983. (Blue Book)
NUREG-0580 V12 NO2: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Blue Book)
NUREG-0606 VOS NO1: UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As Of l
l February 18, 1983.(Aqua Book)
NUREG-0748 VO2 N12: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data r
As Of December 31,1982.(Orange Book)
NUREG-0748 VO3 NO1: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of January 31,1983.(Orange Book)
NUREG-0748 VO3 NO2: OPERATING LICENSING ACTIONS SJMMARY. Data As Of February 28, 1983.(Orange Book)
NUREG-0871 VO2 NO1:
SUMMARY
INFORMATION REPORT. October 1 - December 31, 1982. (Brown Book) 0FFICE OF NUCLEAR REACTOR REQULATION (PDST 4/28/80) 0FFICE OF NUCLEAR REACTOR REQULATION, DIRECTOR 136
l NUREG-0420 SO3: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF SHOREHAM NUCLEAR POWER STATION, UNIT NO.
- 1. Docket No. 50-322. (Long Island Lighting Company)
NUREG-0422 SO6: SAFETY EVALUATION REPORT RELATED TO DPERATION DF MCQUIRE NUCLEAR STATION, UNITS 1 AND 2.0cchet Nos. 50-369 And 50-370.(Duke Power Company) l NUREG-0485 V05 NO2: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPD.9T. Data As Of February 28,1983.(Buff Book)
NUREG-0798 SO3: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF ENRICO FERMI ATDMIC POWER PLANT, UNIT NO.
- 2. Docket No. 50-341.
4 (Detroit Edison Company)
NUREG-0852 SO1: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN 0F THE STANDARD NUCLEAR STEAM SUPPLY REFERENCE SYSTEM CESSAR SYSTEM
- 80. Docket No. STN 50-470. (Combustion Engineering. Incorporated)
NUREG-0857 SO4: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF PALD VERDE NUCLE,AR GENERATING STATION, UNITS 1,2 AND 3. Docket Nos.STN 50-528,STN 50-529 And STN 50-530.(Arizona Public Service l
Compang,et al.)
NUREG-0876 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPERATIDN OF BYRON STATION, UNITS 1 AND 2. Docket Nos. STN 50-454 And STN 50-455.
(Commonwealth Edison Company)
NUREG-0B87 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PERRY NUCLEAR PDWER PLANT, UNITS 1 AND 2. Docket Nos. 50-440 And 50-441.(Cleveland Electric Illuminating Company) j NUREG-0921: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE DPERATION OF CATAWBA NUCLEAR STATION. UNITS 1 AND 2. Docket Nos. 50-413 And 414.(Duke Power Compang et al.)
NUREG-0947: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE I
DPERATING LICENSE FDR THE TEXAS A&M UNIVERSITY TRIGA REACTOR. Docket No. 50-128. License R-83.
NUREG-0954: SAFETY EVALUATION REPORT RELATED TO THE OPERATION DF CATAWBA NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-414. (Duke Power Compang,et al.)
6 l
CLINCH RIVER BREEDER REACTOR PRDCRAM DFFICE l
NUREC-0968 V01: SAFETY EVALUATION REPORT RELATED TO THE CONSTRUCTION OF THE CLINCH RIVER BREEDER REACTOR PLANT. Main Report. Docket No.
50-537.(U.S. Department of Energy, Tennessee Valley Authority And Project Management Corporation)
NUREG-0968 V02: SAFETY EVALUATION REPORT RELATED TO THE CONSTRUCTION OF THE CLINCH RIVER BREEDER REACTDR PLANT. Appendices A-H. Docket No.
50-537.(U.S. Department of Energy, Tennessee Valley Authority And Project Management Corporation)
DIVISION OF ENGINEERING NUREG-0967: SEISMIC HAZARD REVIEW FOR THE SYSTEMATIC EVALUATIDN PROGR AM - A USE OF PROBABILITY IN DECISION MAKING.
J DIVISION DF SYSTEMS INTEORATION (PDST 811005)
NUREG-OB51: NOMOORAMS FOR EVALUATION OF DOSES FRDM FINITE NOBLE GAS CLOUDS.
DIVISION OF LICENSING NUREG-0737 S01: CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS: REGUIREMENTS FOR EMERGENCY, RESPONSE CAPABILITY.
NUREO-0776 S05: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SUSOUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-387 And 50-388.(Pennsylvania Power & Light Companys Allegheng Electric Cooperative, Incorporated)
NUREG-0822: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION PRDORAM,0YSTER CREEK NUCLEAR GENERATING STATION. Docket No.
136
50-219.(GPU Nuclear Corporation And Jersey Centra! Power & Light Company)
NUREG-0823: INTEGRATED SAFETY ASSESSMENT SYSTEMATIC EVALUATION PRDORAM, DRESDEN NUCLEAR POWER STATION UNIT 2. Docket No. 50-237.
(Commonwealth Edison Company)
NUREG-OS24: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATION PRDORAM-MILLSTONE NUCLEAR POWER STATION, UNIT 1. Docket No.
50-245.(Northeast Nuclear Energy Company)
NUREG-0825: INTEGRATED PLANT SAFETY ASSESSMENT REPORT.BYSTEMATIC EVALUATION PROGRAM-YANKEE NUCLEAR POWER STATION. Docket No.50-029.(Yankee Atomic Electric Company)
NUREG-0826 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT, SYSTEMATIC EVALUATION PROGRAM-HADDAM NECK PLANT. Docket No. 50-213.(Connecticut Yankee Atomic Power Company)
NUREG-OB96: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SEABROOK STATION, UNITS 1 AND 2. Docket Nos. 50-443 And 50-444.(Public Service Compang Of New Hampshire,et al.)
NUREG-0943: THREADED FASTENER EXPERIENCE IN NUCLEAR POWER PLANTS.
NUREG-0964: TECHNICAL SPECIFICATIDNS FOR MCQUIRE NUCLEAR STATION, UNITS I AND 2. Docket Nos. 50-369 And 50-370.(Duke Power Company)
NUREG-0966: SAFETY EVALUATION REPORT RELATED TO THE D2/D3 STEAM GENERATOR DESIGN MODIFICATION.
l l
177 y
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 organization (e.g., program office) and then by subsections of these (e.g.,
divisions) where appropriate. The sponsor organization is followed by the NUREG/CR number and title of the report (s) prepared by that organization. If further information is needed, refer to the main citation by the NUREG/CR number.
EDO - OFFICE OF ADMINISTRATION DIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL NUREQ/CR-2974: USER 'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA DIRECTOR'S OFFICE NUREQ/CR-2OOO VO1N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month DF December 1982.
NUREQ/CR-2OOO VO2 N1: LICENSEE EVENT REPORT (LER) COMPILATION: For Month OF January 1983.
NUREQ/CR-2OOO VO2 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of February 1983.
OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEQUARDS DIVISION OF WASTE MANAGEMENT NUREC/CR-2391: DNET SELF-TEACHING CURRICULUM.
I NUREQ/CR-2422: DOSINETRY AND HEALTH EFFECTS SELF-TEACHING CURRICULUM.
Illustrative Problems to Supplement The User's Manual For The Dosimetry And Health Effects Computer Code.
NUREQ/CR-2482 VO3: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
f National Waste Package Program. April 1982 - September 1982.
NUREQ/CR-2755: PACKING MATERIAL TESTING REGUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
NUREQ/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES.
NUREQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH QEOLOGIC DISPOSAL OF NUCLEAR WASTE.
NUREQ/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP QEOLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
NUREQ/CR-2983: SELECTED HYDROLOGIC AND QEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
139
NUREQ/CR-3062: STATUS OF CEOCHEMICAL PROBLEMS RELATING TO THE BURIAL OF HICH-LEVEL RADIOACTIVE WASTE 1982.
NUREC/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HICH LEVEL NUCLEAR WASTE (HLW) DEEP GEDLOGIC REPOSITORIES. Main Report.
NUREC/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP CEOLOCIC REPOSITORIES. Appendices.
NUREQ/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON l
GROUNDWATER MIGRATION.
NUREQ/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIQUES FOR LOW-LEVEL RADIOACTIVE WASTE DISPOSAL.
l NUREQ/CR-3166: RECOMMENDED PROCEDURES FOR MEASURING RADON FLUXES FROM DISFOSAL SITES DF RESIDUAL RADIOACTIVE MATERIALS.
NUREC/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSESSMENT.
OFFICE OF NUCLEA1 REGULATORY RESEARCH (POST 4/05/81)
OFFICE OF NUCLEAR RECULATORY RESEARCH, DIRECTOR i
l NUREC/CR-23OO VO1: PRA PROCEDURES CUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREC/CR-23OO VO2: PRA PROCEDURES CUIDE.A Guide To The Performance C?
Probabilistic Risk Assessments For Nuclear Power Plants.
NUREC/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR i
THERMAL HYDRAULIC EXPERIMENT (TH-3).
NUREC/CR-2751 VO3: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM QUARTERLY I
PROGRESS REPORT FOR JULY-SEPTEMBER 1982.
NUREQ/CR-2774 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July -
September 1982.
NUREC/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROCRAM. Guarterly Progress Report For Period Ending June 30,1982.
NUREQ/CR-2853: NON-CONDENSIBLE CAS FRACTION PREDICTIONS AT ELEVATED l
TEMPERATURES AND PRESSURE USING WET AND DRY BULB TEMPERATURE MEASUREMENTS.
NUREQ/CR-2928: DATA SU 1 MARY REPORT FOR FISSION PRODUCT RELEASE TEST HI-1.
NUREG/CR-3045 VO1: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER l
PLANT DUTACES. Vol. 1: Approach And Analysis.
NUREC/CR-3045 VO2: LOSS OF BENEFITS RESULT NQ FROM NUCLEAR POWER PLANT OUTAGES. Vol 2: Appendixes.
NUREC/CR-3114: PROCEEDINGS OF WORKSHOP DN COGNITIVE MODELING DF NUCLEAR PLANT CONTROL ROOM OPERATORS. August 15-18,1982, Dedham, Massachusetts.
6 NUREC/CR-3115: EXPERT OPINION AND RANKING METHODS.
DIVISION OF ACCIDENT EVALUATION NUREQ/CR-0169 V21: LOFT EXPERIMENTAL NEASUREMENTS UNCERTAINTY ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
NUREQ/CR-1391: MAEROS USER MANUAL.
NUREC/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR RECULATORY RESEARCH.Guarterly Progress Report April 1 -June 30,1982.
NUREC/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report No.
12.
NUREG/CR-2531 RO1: INTRODUCTORY USERS MANUAL FOR THE US NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
NUREG/CR-2716 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly i
Report. July -September 1982.
NUREC/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY 140
l
-PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
NURES/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
NUREG/CR-2874 VO2: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report, April 1 - June 30,1982.
NUREQ/CR-2887: RELAP5 ASSESSMENT: FLECHT SEASET STEAM OENERATOR TEST j
23402.
NURES/CR-2902: FLOW REGIME MODELING STUDY FOR THE SIMMER-II LMFBR SAFETY CODE: CLAD RELOCATION.
l NURES/CR-2929: FEASIBILITY OF MONITORING THE STRENGTH OF HTOR CORE SUPPORT GRAPHITE -PART III.
l NUREO/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CONVECTION MAT TRANSFER OF INTERNALLY HEATED LIGUIDS.
NURES/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
NUREG/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE B-4.
NUREQ/CR-2972: AN ANALYSIS OF DENSITY-WAVE OSCILLATIONS IN VENTILATED CHANNELS.
NURE0/CR-3OOO-LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE CONCRETE INTER %CTIONS.
NURE0/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLDOY OF THE TRANSITION PHASE OF A CORE-DISRUPTIVE ACCIDENT IN A LMFBR.
NURES/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UPPROTECTED TRANSIENT UNDERCOOLING ACCIDENT IN A LARGE, HETER 00EPEDUS-CORE, LIGUID-METAL-COOLED FAST BREEDER REACTOR.
NURES/CR-3033: MODELING AOKI ET AL. EXPERIMENTS ON CONDENSATION OF I
FLOWING STEAM ONTO INJECTED WATER VIA K-FIX.
i NUREC/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models.
NUREO/CR-3046 VO2: COBRA / TRAC - A TERMAL-HYDRAULICS CODE FOR l
TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT l
SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
NUREQ/CR-3046 VO3: COBRA /7RAC - A TERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Users ' Manual.
NURES/CR-3046 VO4: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VFSSELS AND PRIMARY COOLANT t
SYSTEMS. Volume 4: Developmental Assessment And Data.
NUREQ/CR-3046 VOS: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT l
SYSTEMS. Volume 5: Programmers Manual.
NURES/CR-3126: DEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST l
RESULTS CDPFARISON.
NUREQ/CR-3141: CHARACTERIZATION OF TWO-PHASE FLOW USING NEUTRONIC FLUCTUATIONS.
NUREQ/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
ANALYTICAL MODELS BRANCH NUREQ/CR-2843 VO1: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report, January - March 1982.
DIVISION OF FACILITY OPERATIONS NURES/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review Of Existing Human Reliability Data Banks.
NURES/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
NURES/CR-2892: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: A 141
. - ~.
Revised Procedures Manual.
NUREQ/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
NUREQ/CR-3OOS: AUDITORY PERCEPTION IN LODSE-PARTS MONITORING.
NUREQ/CR-3010: POST EVENT HUMAN DECISION ERRORS: OPERATOR ACTION TREE / TINE RELIABILITY CORRELATION.
NUREQ/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
NUREQ/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
NUREQ/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL.
j l
NUREQ/CR-3105: SECURORS APPLICATION TO A GENERIC NUCLEAR POWER PLANT.
NUREQ/CR-3113: A TORSIONAL ULTRASONIC TECHNIGUE FOR LWR LIGUID LEVEL MEASUREMENT.
NUREQ/CR-3177 Vol: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY t
PROCEDURE QUIDELINES. Volume 1: Methodologies.
NUREQ/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY l
PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants, i
DIVISION OF HEALTH, SITING & WASTE MANAGEMENT NUREQ/CR-1894: MECHANICAL RELIABILITY EVALUATION OF A PROPOSED EMERGENCY RESPONSE RADI0 IODINE AIR SAMPLER.
NUREQ/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
NUREQ/CR-2524: EVALUATION OF PERSONNEL NEUTRON DOSIMETRY AT OPERATING NUCLEAR POWER PLANTS.
NUREQ/CR-2675 VO3: RELEVANCE OF BIOTIC PATHWAYS TO REQULATION OF NUCLEAR WASTE DISPOSALS Topical Report On Reference Eastern Humid l
Low-Level Sites.
NUREQ/CR-2006: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT l
COOLING WATERS.
NUREQ/CR-2856: A REVIEW OF FUGITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
NUREQ/CR-2863: VALENCE EFFECTG ON ADSORPTION: A PRELIMINARY ASSESSMENT OF THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
NUREQ/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT OF AVAILABLE MARGINS INHERENT IN FLOOD PROTECTION OF NUCLEAR POWER PLANTS.
NUREQ/CR-3027: OVERLAND EROSION OF URANIUM MILL TAILINGS IMPOUNDMENTS: PHYSICAL PROCESSES AND COMPUTATIONAL METHODS.
NUREO/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT OF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
NUREQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
NUREQ/CR-3078: MODEL EVALUATION OF SEFPAGE FCM URANIUM TAILINGS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
NUREQ/CR-3079: EARTHGUAKE HAZARD STUDIES IN NEW YORK STATE AND ADJACENT AREAS. Final Report. April 1976 - June 1982.
NUREQ/CR-3080: NETWORK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA.
I NUREQ/CR-3106: COMPARISON OF FIVti-MINUTE RADON DAUGHTER MEASUREMENTS WITH LONG-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
NUREQ/CR-3109: SEISM 7 CITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN OKLAHOMA-Final Report.
NUREQ/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. Annual Report For FY 1982.
NUREQ/CR-3145 VO1: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
NUREQ/CR-3149: DISPERSION COEFFICIENTS FOR CDASTAL REGIONS.
NUREQ/CR-3206: UNSATURATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK-RELATED TO HIGH LEVEL WASTE REPOSITORIES.
WASTE MANAGEMENT BRANCH 142
NUREQ/CR-3076: COMPUTER PREDICTION OF SUBSURFACE RADIONUCLIDE TRANSPDRT -AN ADAPTIVE NUMERICAL METHOD.
DIVISION OF RISK ANALYSIS NUREQ/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, January 1,1972 Through September 30,1990.
NUREQ/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
NUREQ/CR-2729: USER 'S QUIDE TO BFR. A Computer Code Based On The J
Binomial Failure Rate Common Cause Model.
NUREQ/CR-2770: COMMON CAUSE FAULT RATES FOR VALVES: Estimates Based i
On Licensee Event Reports At U.S.
Commercial Nuclear Power l
Plants, 1976-1980.
NUREQ/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants, 1976-1978.
NUREQ/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT.
NUREQ/CR-2904: SIMULATION OF QROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT) MODEL.
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
NUREG/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS.
DIVISION OF ENGINEERING TECHNOLOGY NUREG/CR-1120 V10: SEISMIC SAFETY MARGINS RESEARCH PROCRAM. Progress Report No. 14.
NUREQ/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
NUREO/CR-2015 VO3: SEISMIC SAFETY MARGINS RESEARCH PROGRAM, PHASE I FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
NUREQ/CR-2751 VO2: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR APRIL-JUNE 1982.
NURE0/CR-2780: EVALUATION OF SYSTEM REGUIREMENTS AND STANDARDS DEVELOPMENT FOR THERMAL ANNEALING OF REACTOR PRESSURE VESSELS.
l NUREQ/CR-2801: PIPING RELIABILITY MODE VALIDATION AND POTENTI AL USE FOR LICENSING REQULATION DEVELOPMENT.
NUREG/CR-2905 VO1: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IPPROVEMENT PROGRAM. Guarter1y Progress Report. January 1982 - March 1982.
NUREC/CR-2005 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report. April 1982 - June 1982.
NURES/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
l NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD l
CRACKS FOR REACTOR PRESSURE VESSELS.
NUREQ/CR-2895: PWR PRESSURE VESSEL INTEORITY DURING DVERC00 LING ACCIDENTS: A PARAMETRIC ANALYSIS.
NUREG/CR-2944: TORNADO DAMAGE RISK ASSESSMENT.
l NUREG/CR-2945: CHARACTERIZATION OF EARTHOUAKE FORCES FOR I
PROBABILITY-BASED DESIGN OF NUCLEAR STRUCTURES.
NURE0/CR-2997 VO2: APPLICATIONS OF ENERGY RELEASE RATE TECHNIGUES TO PART-THROUGH CRACKS IN PLATES AND CYLINDERS. Volume 2: ORVIRT: A Finite Element Program For Energy Release Rate Calculations For 2-Dimensional And 3-Dimensional Crack Models.
NUREG/CR-3059: PARAMETRIC CALCULATIONS OF FATIQUE CRACK OROWTH IN 143
PIPING.
NUREG/CR-3116: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Project Summary Report,Enrico Fermi-1 Reactor.
NUREG/CR-3135: BUCKLING INVESTIGATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
NUREG/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
NUREG/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
NUREG/CR-3212: SELECTEL REVIEW AND EVALUATION OF U.S.
SAFETY RESEARCH VIS-A-VIS FOREIGN SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
MATERIALS ENGINEERING BRANCH NUREG/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
EDD-RESOURCE MANAGEMENT l
DIVISION OF DATA AUTOMATION & MANAGEMENT INFORMATION NURFQ/CR-2907 VO1: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1990.
1 0FFICE OF NUCLEAR REACTOR REQULATION (POST 4/28/80)
DIVISION OF ENGINEERING NUREQ/CR-2804: THE APPLICATION OF FISHERIES MANAGF_ MENT TECHNIGUES TO ASSESSING IMPACTS. Final Report.
NUREQ/CR-2824 VO3: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending September 30, j
1982.
NUREG/CR-2993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM GENERATOR AT THE MAINE YANKEE NUCLEAR POWER STATION.
NURES/CR-3OO9: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
NUREG/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT COMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPINO RESPONSES.
NUREG/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED GROUND MOTIONS.
JIVISION OF HUMAN FACTORS SAFETY NUREQ/CR-2668: JOB ANALYSIS OF TE MAINTENANCE SUPERVISOR AND INSTRUMENT AND CONTROL SUPERVISOR POSITIONS FOR THE NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
i NUREG/CR-2952: ENGINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER PLANTS: THE FOREIGN EXPERIENCE.
DIVISION OF SYSTEMS INTEGRATION (POST B11005)
NUREG/CR-2530: REVIEW OF THE GRAND GULF HYDROGEN IGNITER SYSTEM.
NUREG/CR-2659: IDDINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM GENERATOR TUBES.
NUREG/CR-2666: PWR SEVERE ACCIDENT DELINEATION AND ASSESSMENT.
NUREG/CR-2847: COGAP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL SYSTEM EVALUATION CODE.
NUREQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
NUREG/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL STATEMENTS.
NUREG/C R-2998I. A COMPARISON OF BWR STABILITY MEASUREENTS WITH CALCULATIONS USING THE CODE LAPUR-IV.
144
NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREQ/CR-3069 VO2: INTERACTION OF ELECTRDMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
NUREC/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
. DIVISION OF SAFETY TECHNOLOGY NUREG/CR-2800: CUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.
NUREQ/CR-3C 28: A REVIEW OF THE LIMERICK QENERATING STATION PROBABILISTIC RISK ASSESSMENT.
l l
l P
l 145
Contractor !ndex l
This index lists, in alphabetical order, the contractors that prepared the l
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.
ALDEN RESEARCH LABORATORY NUR EC/CR-3170: THE SUSCEPTIBILITY OF FIBROUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
AMERICAN NUCLEAR SOCIETY NUREQ/CP-OO27 VO1: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
NUREQ/CP-OO27 VO2: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
NUREC/CP-OO27 VO3: PROCEEDINGS OF THE INTERNATIONAL MEETING ON THERMAL NUCLEAR REACTOR SAFETY. Held At Chicago, Illinois, August 29 -September 2,1982.
NUREC/CR-23OO VO1: PRA PROCEDURES CUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREG/CR-23OO VO2: PRA PROCEDURES GUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
ARGONNE NATIONAL LABORATORY NUREQ/CR-2774 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July -
September 1982.
NUR EC/CR-2821: EBR-II IN-VESSEL NATURAL-CIRCULATION ANALYSIS.
l NUREQ/CR-3045 VO1: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT OUTAGES. Vol. 1: Approach And Analysis.
NUREQ/CR-3045 VO2: LOSS OF BENEFITS RESULTING FROM NUCLEAR POWER PLANT OUTAGES. Vol 2: Appendixes.
NUREQ/CR-3167: HEAT TRANSFER TO WATER FROM A VERTICAL TUBE BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.
ARIZONA, UNIV. OF NUREO/CR-3076: COMPUTER PREDICTION OF SUBSURFACE RADIONUCLIDE TRANSPORT
-AN ADAPTIVE NUMERICAL METHOD.
NUR EG/CR-3206: UNSATURATED FLOW AND TRANSPORT THROUGH FRACTURED r
ROCK-RELATED TO HIGH LEVEL WASTE REPOSITORIES.
l ARMY, DEPT. OF, ARMY ENGINEER WATERWAYS EXPERIMENT STATION NUR EC/CR-2879: FEASIBILITY FOR GUANTITATIVE ASSESSMENT OF AVAILABLE MARCINS INHERENT IN FLOOD PROTECTION OF NUCLEAR POWER PLANTS.
NUREG/CR-3144: TRENCH DESIGN AND CONSTRUCTION TECHNIGUES FOR LOW-LEVEL RADIOACTIVE WASTE DISPOSAL.
BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORY NUR EQ/CR-2527: LOCA SIMULATION IN NRU PROGRAM - DATA REPORT FOR THEHMAL HYDRAULIC EXPERIMENT (TH-3).
147
NUREQ/CR-2659: IODINE TRANSPORT PREDICTED FOR A POSTULATED STEAM LINE BREAK WITH CONCURRENT RUPTURES OF STEAM QENERATOR TUBES.
NUREQ/CR-2675 V03: RELEVANCE OF BIOTIC PATHWAYS TO REQULATIDN OF NUCLEAR WASTE DISPOSALS. Topical Report On Reference Eastern Humid Low-Level Sites.
NUREQ/CR-2716 V03: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report. July -September 1982.
NUR EC/CR-2800: QUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIDRITIZATION INFORMATION DEVELOPMENT.
NUR EC/CR-2004: THE APPLICATION OF FISHERIES MANAGEMENT TECHNIGUES TO ASSESSINC IMPACTS. Final Report.
NUREQ/CR-2056: A REVIEW OF FUQITIVE DUST CONTROL FOR URANIUM MILL TAILINGS.
NUREQ/CR-2878: DETECTION OF SMALL-SIZED NEAR-SURFACE UNDER-CLAD CRACKS FOR REACTOR PRESSURE VESSELS.
NUREQ/CR-2929: FEASIBILITY'OF MONITORING THE STRENGTH OF HTOR CORE SUPPORT ORAPHITE--PART III.
NUREQ/CR-2935: EXAMPLES OF MC&A SYSTEMS TO MEET PROMPT ACCOUNTABILITY SPECIFICATIONS.
NUR EQ/CR-2952: ENCINEERING EXPERTISE ON SHIFT IN NUCLEAR POWER PLANTS: THE FOREIGN EXPERIENCE.
NUREQ/CR-2956: NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS. Final Report Of Substask B: Dosimeter Response.
NUREQ/CR-3027: OVERLAND EROSION OF UPANIUM MILL TAILINGS IMPOUNDMENTS:
PHYSICAL PROCESSES AND COMPUTATIONAL METHODS.
NUREQ/CR-3030: EVALUATION OF SELECTED NEUTRALIZING AGENTS FOR THE TREATMENT OF URANIUM TAILINGS LEACHATES. Laboratory Progress Report.
NUREQ/CR-3046 VO1: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 1: Equations And Constitutive Models.
NUREQ/CR-3046 V02: COBRA / TRAC - A THERMAL-HYDRAULICS CODE FOR TRANSIENT l
ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY CODLANT SYSTEMS. Volume 2: COBRA / TRAC Numerical Solution Methods.
NUREQ/CR-3046 VO3: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 3: Us ers ' Manual.
NUREQ/CR-3046 V04: COBRA / TRAC - A THERMAL HYDRAULIC CODE FOR TRANSIENT i
l ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT l
SYSTEMS. Volume 4: Developmental Assessment And Data.
NUREQ/CR-3046 V05: COBRA / TRAC - A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR VESSELS AND PRIMARY COOLANT SYSTEMS. Volume 5: Programmers Manual.
i NUREQ/CR-3059: PARAMETRIC CALCULATIONS OF FATIQUE CRACK QROWTH IN PIPING.
NUREC/CR-3078: MODEL EVALUATION OF SEEPAGE FROM URANIUM TAILINOS DISPOSAL ABOVE AND BELOW THE WATER TABLE.
NUR EQ/CR-3106: COMPARISON OF FIVE-MINUTE RADON DAUQHTER MEASUREMENTS WITH LONO-TERM RADON AND RADON DAUGHTER CONCENTRATIONS.
NUREQ/CR-3149: DISPERSION COEFFICIENTS FOR COASTAL REGIONS.
NUREG/CR-3166: RECOMMENDED PROCEDURES FOR MEASURINO RADON FLUXES FROM DISPOSAL SITES OF RESIDUAL RADIOACTIVE MATERIALS.
7 BELGIUM NUREC/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
BOEING CO.
NUREQ/CR-3069 V01: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCI AL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREQ/CR-3069 VO2: INTERACTION OF ELECTROMAQNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
148
l BOOZ ALLEN & HAMILTON, INC.
NUREC/CR-3069 Vol: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL i
NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREO/CR-3069 VO2: IN1ERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCI AL NUCLEAR POWER PLANT SYSTEMS. Main Report.
i BROOKHAVEN NATIONAL LABORATORY NUREG/CR-1967: FAILURE EVALUATION OF A REINFORCED CONCRETE MARK III CONTAINMENT STRUCTURE UNDER UNIFORM PRESSURE.
NUREO/CR-2331 VO2 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REQULATORY RESEARCH. Quarterly Progress Report, April 1 -June 30,1982.
NUREC/CR-2482 VO3: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program. April 1992 - September 1982.
NUREG/CR-2755: PACKING MATERIAL TESTING REQUIRED TO DEMONSTRATE COMPLIANCE WITH 1000-YEAR RADIONUCLIDE CONTAINMENT: Semiannual Report On Waste Package Verification Tests.
NURE9/CR-2907 VO1: RADIDACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS - 1980.
NUR EQ/CR-2939: EXPERIMENTAL AND ANALYTICAL STUDY OF NATURAL CONVECTION HEAT TRANSFER OF INTERNALLY HEATED LIQUIDS.
NUREC/CR-2945: CHARACTERIZATION OF EARTHOUAKE FORCES FOR PROBABILITY-BASED DESIGN OF NUCLEAR STRUCTURES.
NUREQ/CR '?993: EXAMINATION OF FAILED STUDS FROM NO. 2 STEAM OENERATOR AT THE MAINE YANKEE NUCLEAR POWER STATION.
NUREQ/CR-3010: POST EVENT HUMAN DECISION ERRORS: 0PERATOR ACTION TREE /TI E RELIABILITY CORRELATION.
NUREQ/CR-3014: ASSESSMENT OF THE THERMAL HYDRAULIC TECHNOLOGY OF THE TRANSITION PHASE OF A CORE-DISRUPIIVE ACCIDENT IN A LMFBR.
NUREG/CR-3028: A REVIEW OF THE LIMERICK OENERATING STATION PROBABILISTIC RISK ASSESSMENT.
NUREG/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FOR STANDBY SAFETY SYSTEMS.
NUREG/CR-3086: INVESTIGATION OF THE CONSERVATISM ASSOCIATED WITH DIFFERENT CCMBINATIONS BETWEEN PRIMARY AND SECONDARY PIPING RESPONSES.
CALIFORNIA, UNIV. OF, BERKELEY NUR EO/CR-3062: STATUS OF OE0 CHEMICAL PROBLEMS RELATING TO THE BURIAL OF l
HIGH-LEVEL RADIDACTIVE WASTE 1982.
CALIFORNIA, UNIV. OF, LOS ANGELES NUREO/CR-2666: PWR BEVERE ACCIDENT DELINEATION AND ASSESSMENT.
CATHOLIC UNIV.
NUREO/CR-3OOS. AUDITORY PERCEPTION IN LODSE-PARTS MONITORING.
CSS, INC.
NUREQ/CR-2904: SIMULATION OF GROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
COLUMBIA UNIV.
NUREO/CR-3079: EARTHGUAKE HAZARD STUDIES IN NEW YORK STATE AND ADJACENT AREAS. Final Report. April 1976 - June 1982.
COMERCE. DEPT. OF, NATIONAL BUREAU OF STANDARDS NUREQ/CR-2005 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IPPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September i
30, 1982).
NUREG/CR-2944: TORNADO DAMAGE RISK ASSESSMENT.
NURE0/CR-2945: CHARACTEftIZATION OF EARTH 00AKE FORCES FOR PROBABILITY-BASED DESIGN OF NUCLEAR STRUCTURES.
NUREG/CR-3120: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING
(
R ADI ATION. Annual Report For FY 1982.
DAPES & MOORE NUREC/CR-3130: INFLUENCE OF LEACH RATE AND OTHER PARAMETERS ON 149 l
OROUNDWATER MIGRATION.
EQ4rQ, INC.
NUREQ/CR-0169 V21: LDFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES. Volume XXI. Modular Drag-Disc Turbine Transducer.
NUREQ/CR-2098: COMMON CAUSE FAULT RATES FOR PUMPS: Estimates Based On Licensee Event Reports At U.S.
Commercial Nuclear Power Plants January 1,1972 Through September 30,1980.
NUREQ/CR-2531 RO1: INTRODUCTORY USERS MANUAL FOR THE US NUCLEAR REQULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
NUR EQ/CR-2729: USER'S GUIDE TO BFR.A Computer Code Based On The Binomial Failure Rate Common Cause Model.
NUREQ/CR-2770: COMMON CAUSE FAULT RATES FOR VALVES: Estimates Based On Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, 1976-1980.
NUREO/CR-2771: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES: Estimates Based On Licensee Event Reports At U. S.
Commercial Nuclear Power Plants, 1976-1978.
NUR EQ/CR-2700: EVALUATION OF SYSTEM REGUIREMENTS AND STANDARDS DEVELOPMENT FOR THERMAL ANNEALING OF REACTOR PRESSURE VEBSELS.
NUREQ/CR-3103: PREDICTOR DISPLAY CONCEPTS FOR USE IN NUCLEAR PLANT CONTROL.
j NUREQ/CR-3126: SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES-TEST RESULTS COMPARISON.
NUREQ/CR-3177 VO1: NETHCDS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE QUIDELINES. Volume 1: Methodologies.
NUREQ/CR-3177 VO2:' METHODS FOR REVIEW AND EVALUATION OF' EMERGENCY PROCEDURE QUIDELINES. Volume 2: Applications To Westinghouse Plants.
ELECTRIC POWER RESEARCH INSTITUTE I
NUR EQ/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report.NRC/EPRI/ Westinghouse Report l
No. 12.
FLORIDA, UNIV. OF, GAINESVILLE NUREQ/CR-2443: A MANUAL FOR USING ENERGY ANALYSIS FOR PLANT SITING.
GENER AL PHYSICS CORP.
i NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS.Vol 1: A Review 0F Existing Human Reliability Data Banks.
NUREQ/CR-3092: CRITERIA FOR BAFETY-RELATED NUCLEAR POWER PLANT DPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
00LDER ASSOCIATES t
NUREQ/CR-2854: EVALUATION OF ALTERNATIVE SHAFT SINKING TECHNIGUES FOR l
HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES.
l NUR EQ/CR-2959: RELATIONSHIP OF AN IN SITU TEST FACILITY TO A DEEP
(
GEDLOGIC REPOSITORY FOR HIGH LEVEL NUCLEAR WASTE.
NUREQ/CR-3065 VO1: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QEOLOGIC REPOSITORIES. Main Report.
NUREQ/CR-3065 VO2: IN SITU TEST PROGRAMS RELATED TO DESIGN AND CONSTRUCTION OF HIGH LEVEL NUCLEAR WASTE (HLW) DEEP QE0 LOGIC REPOSITORIES. Appendices.
HANFORD ENGINEERING DEVELOPMENT LABORATORY NUREQ/CR-2805 VO1: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report. January 1982 - March 1982.
NUREQ/CR-2005 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report. April 1982 - June 1982.
NUREQ/CR-2005 VO3: LWR PRESSURE VESSEL SURVEILLANCE DDSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
IDAHO NATIONAL ENGINEERING LABORATOGY NUREG/CR-1894: MECHANICAL RELIABILITY EVALUATION OF A PROPOSED 150
I EMERGENCY RESPONSE RADIOIODINE AIR SAMPLER.
INSTITUTE OF ELECTRICAL & ELECTRONIC ENGINEERS NUREQ/CR-2300 VO1: PRA PROCEDURES QUIDE.A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREQ/CR-2300 V02: PRA PROCEDURES QUIDE.A Quide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
IRT CORP.
NUREC/CR-3069 V01: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREQ/CR-3069 V02: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Main Report.
LAWRENCE BERKELEY LABORATORY NUR EQ/CR-2910: THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH CEOLOGIC DISPOSAL OF NUCLEAR WASTE.
NUREQ/CR-2983: SELECTED HYDROLOGIC AND GEOCHEMICAL ISSUES IN SITE CHARACTERIZATION FOR NUCLEAR WASTE DISPOSAL: Flood Basalts At The Hanford Reservation.
LAWRENCE LIVERMORE LABORATORY NUREQ/CR-1120 VIO: SEISMIC SAFETY MARGINS RESEARCH PROGRAM. Progress Report No. 14.
NUREQ/CR-2015 V03: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT-DEVELOPMENT OF SEISMIC INPUT (PROJECT II).
NUR EQ/CR-2001: PIPING RELIABILITY MODE VALIDATION AND POTENTIAL USE FOR LICENSING REQULATION DEVELOPMENT.
NUREQ/CR-3102: EFFECTS OF EARTHOUAKE RUPTURE SHALLOWNESS AND LOCAL SOIL CONDITIONS ON SIMULATED GROUND MOTIONS.
LOS ALAMOS SCIENTIFIC LABORATORY NUREQ/CR-2847: C00AP: A NUCLEAR POWER PLANT CONTAINMENT HYDROGEN CONTROL SYSTEM EVALUATION CODE.
NUREQ/CR-2848: MECHANISTIC DRY-PRESSURE-CONTAINMENT LOCA ANALYSIS.
NUREC/CR-2902: FLOW REGIME MODELING STUDY FOR THE SIMMER-II LMFBR SAFETY CODE: CLAD RELOCATION.
NUR EQ/CR-3031: AN EVALUATION OF THE CALCULATED RESULTS OF AN UNPROTECTED TRANSIENT UNDERCOOLING ACCIDENT IN A LARGE, HETEROQENEOUS-CORE, LIQUID-METAL-COOLED FAST BREEDER REACTOR.
NUREQ/CR-3032: STUDIES OF TRANSPORT OF WASTE RADIONUCLIDES THROUGH SOIL AT THE MAXEY FLATS, KENTUCKY, WASTE-BURIAL SITE.
NUR EQ/CR-3033: MODELING AOKI ET AL. EXPERIMENTS ON CONDENSATION OF FLOWING STEAM ONTO INJECTED WATER VIA K-FIX.
NUR EQ/CR-3108: EXTENDED BURNUP CALCULATIONS FOR OPERATING REACTOR RELOAD REVIEWS.
NUREQ/CR-3135: BUCKLING INVESTIGATION OF RING-STIFFENED CYLINDRICAL SHELLS WITH REINFORCED OPENINGS UNDER UNSYMMETRICAL AXIAL LOADS.
MASSACHUSETTS INSTITUTE OF TECHNOLOGY NUR EQ/CR-2853: NON-CONDENSIBLE GAS FRACTION PREDICTIDNS AT ELEVATED TEMPERATURES AND PRESSURE USING WET AND DRY BULB TEMPERATURE MEASUREMENTS.
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
NUREQ/CR-3157: DESIGN OF REINFORCED CONCRETE CONTAINMENT WALL ELEMENTS UNDER COMBINED ACTION OF SHEAR AND TENSION.
MICHIQAN, UNIV. OF NUR EQ/CR-2703: INVESTIGATION OF SPECIAL PURPOSE PROCESSORS FOR REAL-TIME SYNTHETIC APERTURE FOCUSING TECHNIGUES FOR NONDESTRUCTIVE EVALUATION OF NUCLEAR REACTOR VESSELS AND PIPING COMPONENTS.
NUR EQ/CR-2891: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: Final Report Of Test 3.
NUREQ/CR-3145 V01: GEOPHYSICAL INVESTICATIONS OF THE WESTERN OHIO-INDIANA REGION. Annual Report,0ctober 1981-September 1982.
MICHICAN, UNIV. OF, MEDICAL SCHOOL 151
?.
Y y
, ~
.g.
y
. :v
_ _ - =
i
{
NUREQ/CR-2892: PERFORMANCE TESTING OF PERSONNEL DOSIMETRY SERVICES: A 1
1 Revised Procedures Manual.
l i
NORTH CAROLINA, UNIV. OF I
NUREQ/CR-2006: A KINETIC MODEL FOR THE CHLORINATION OF POWER PLANT l
COOLING WATERS.
DAK RIDGE NATIONAL LABORATORY l
NUREQ/CP-OO28 VO3: PROCEEDINGS OF THE SYMPOSIUM ON LOW-LEVEL WASTE DISPOSAL: Facility Design, Construction And Operating Practices.
I NUREQ/CR-2OOO VO1N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of December 1982.
NUREQ/CR-2OOO VO2 N1: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of January 1983.
NUREQ/CR-2OOO VO2 N2: LICENSEE EVENT REPORT (LER) COMPILATION:For Month l
Of February 1983.
NUREQ/CR-2668: JOB ANALYSIS OF THE MAINTENANCE SUPERVISOR AND INSTRUMENT AND CONTROL SUPERVISOR POSITIONS FOR THE NUCLEAR POWER l
PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
i NUREQ/CR-2751 VO2: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY f
PROGRESS REPORT FOR APRIL-JUNE 1982.
NUREQ/CR-2751 VO3: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JULY-SEPTEMBER 1982.
NUREQ/CR-2805 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.1982 ANNUAL REPORT (October 1,1981 - September 30, 1982).
t NUREQ/CR-2809 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM GUARTERLY PROGRESS REPORT FOR JULY - SEPTEMBER 1982.
J NUREG/CR-2824 VO2: EDDY-CURRENT INSPECTION FOR STEAM CENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending June 30,198E NUREQ/CR-2824 VO3: EDDY-CURRENT INSPECTION FOR STEAM CENERATOR TUBING PROGRAM. Guarterly Progress Report For Period Ending September 30, i
1982.
NUREC/CR-2863: VALENCE EFFECTS ON ADSORPTION: A PRELIMINARY ASSESSMENT 0F THE EFFECTS OF VALENCE STATE CONTROL ON SORPTION MEASUREMENTS.
NUREC/CR-2874 VO2: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Guarterly Progress Report, i
April 1 - June 30,1982.
NUR EQ/CR-2886: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT l
COMPONENTS: INTERIM DATA REPORT-THE PUMP COMPONENT.
NUREC/CR-2895: PWR PRESSURE VESSEL INTEGRITY DURING DVERC00 LING ACCIDENTS: A PARAMETRIC ANALYSIS.
NUREG/CR-2928: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-1.
NUR EC/CR-2968: EXPERIMENT DATA REPORT FOR MULTIROD BURST TEST (MRBT)
BUNDLE B-4.
NUREQ/CR-2974: USER'S MANUAL FOR LPOS: A Computer Program For Calculating Radiation Exposure Resulting From Accidental Radioactive Releases To The Hydrosphere.
NUREQ/CR-2997 VO2: APPLICATIONS OF ENERQY RELEASE RATE TECHNIGUES TO PART-THROUGH CRACKS IN PLATES AND CYLINDERS. Volume 2: 0RVIRT: A Finite Element Program For Energy Release Rate Calculations For 2-Dimensional And 3-Dimensional Crack Models.
NUREQ/CR-2998: A COMPARISON OF BWR STABILITY MEASUREMENTS WITH CALCULATIONS USING THE CODE LAPUR-IV.
NUREG/CR-3092: CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR ACTIONS: INITIAL SIMULATOR TO FIELD DATA CALIBRATION.
NUREQ/CR-3113: A TORSIONAL ULTRASONIC TECHNIGUE FOR LWR LIGUID LEVEL MEASUREMENT.
NUR EC/CR-3114: PROCEEDINGS OF WORKSHOP DN COGNITIVE MODELING OF NUCLEAR PLANT CONTROL ROOM OPERATORS. August 15-18,1982,Dedham, Massachusetts.
NUREQ/CR-3115: EXPERT OPINION AND RANKING METHODS.
152 i
-_. ~.
OKLAHOMA, UNIV. OF NUREQ/CR-3109: SEISMICITY AND TECTONIC RELATIONSHIPS OF THE NEMAHA UPLIFT IN DKLAHOMA-Final Report.
RENSSELAER POLYTECHNIC INST.
NUREQ/CR-2524: EVALUATION OF PERSONNEL NEUTRON DOSIMETRY AT OPERATING NUCLEAR POWER PLANTS.
NUR EQ/CR-2972: AN ANALYSIS OF DENSITY-WAVE OSCILLATIONS IN VENTILATED CHANNELS.
SANDIA LABORATORIES NUREQ/CR-1391: MAEROS LSER MANUAL.
NUREQ/CR-2391: DNET SELF-TEACHING CURRICULUM.
NUREQ/CR-2422: DOSIMETRY AND HEALTH EFFECTS SELF-TEACHING CURRICULUM.
l Illustrative Problems to Supplement The User's Manual For The Dosimetry And Health Effects Computer Code.
NUREQ/CR-2530: REVIEW OF THE GRAND QULF HYDROGEN IGNITER SYSTEM.
i NUR EQ/CR-2728: INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES QUIDE.
NUREQ/CR-2744 VO1: HUMAN RELIABILITY DATA BANK FOR NUCLEAR POWER PLANT OPERATIONS. Val 1: A Review 0F Existing Human Reliability Data Banks.
NUREQ/CR-2843 VO1: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report, January - March 1982.
NUR EQ/CR-2887: RELAP5 ASSESSMENT: FLECHT SEASET STEAM QENERATOR TEST 23402.
NUREQ/CR-2901: CRAC CALCULATIONS FOR ACCIDENT SECTIONS OF ENVIRONMENTAL STATEMENTS.
NUR EQ/CR-2904: SIMULATION OF QROUNDWATER FLOW AND CONTAMINANT TRANSPORT l
USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
NUREQ/CR-2925: IN-PLANT CONSIDERATIONS FOR OPTIMAL OFFSITE RESPONSE TO REACTOR ACCIDENTS.
NUREQ/CR-3OOO: LARGE-SCALE EXPLORATORY TESTS OF SODIUM / LIMESTONE CONCRETE INTERACTIONS.
NUR EQ/CR-3OO9: FRACTURE TOUGHNESS OF PWR COMPONENT SUPPORTS.
NUREQ/CR-3063: A
SUMMARY
OF THE PLANT ELECTRICAL SYSTEMS (PES) STUDY.
NUREQ/CR-3069 VO1: INTERACTION OF ELECTROMAGNETIC PULSE WITH COMMERCIAL NUCLEAR POWER PLANT SYSTEMS. Executive Summary.
NUREQ/CR-3069 VO2: INTERACTION OF ELEC110 MAGNETIC PULSE WITH COMMERCI AL NUCLEAR POWER PLANT SYSTEMS. Main " " et.
NUREQ/CR-3105: SECURORS APPLICATIC' r0 A GENERIC NUCLEAR POWER PLANT.
NUREQ/CR-3170: THE SUSCEPTIBILITY OF FIBRDUS INSULATION PILLOWS TO DEBRIS FORMATION UNDER EXPOSURE TO ENERGETIC JET FLOWS.
SCIENCE APPLICATIONS, INC.
I NUREQ/CR-3082: PROBABILISTIC APPROACHES TO LCO'S AND SURVEILLANCE REGUIREMENTS FDR STANDBY SAFETY SYSTEMS.
SIMCO GROUNDWATER ASSOCIATION NUR EQ/CR-2904: SIMULATION OF QROUNDWATER FLOW AND CONTAMINANT TRANSPORT USING THE DETERMINISTIC-PROBABILISTIC CONTAMINANT TRANSPORT (DPCT)
MODEL.
STEVENSON & ASSOCIATES NUREQ/CR-3212: SELECTED REVIEW AND EVALUATION OF U.S.
SAFETY RESEARCH VIS-A-VIS FOREIGN SAFETY RESEARCH FOR NUCLEAR POWER PLANTS.
SZAWLEWICZ, S. A.
NUREQ/CP-OO41 VO1: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NUREQ/CP-OO41 VO2: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
I NUREQ/CP-OO41 VO3: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY i
RESEARCH INFORMATION MEETING.
NUREQ/CP-OO41 VO4: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.
NUREQ/CP-OO41 VO5: PROCEEDINGS OF THE TENTH WATER REACTOR SAFETY 153
RESEARCH INFORMATION MEETING.
NUREQ/CP-OO41 VO6: PROCEEDINGS'OF THE TENTH WATER REACTOR S/FETY RESEARCH INFORMATION MEETINCc TEKNEKRON RESEARCH, INC.
NUREQ/CR-3209: A
SUMMARY
OF COMPUTER CODES FOR RADIOLOGICAL ASSES 3 MENT.
TRC ENVIRONMENTAL CONSULTANTS, INC.
NUREQ/CR-3149: DISPERSION COEFFICIENTS FOR CUASTAL REGIONS; x
UNITED NUCLEAR CORP.
NUREG/CR-3116: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Project Summary Report,Enrico Fermi-1 Reactor.
(
VIRGINIA POLYTECHNIC INSTITUTE NUREQ/CR-3080: NETWCRK LOCATIONAL TESTING AND VELOCITY VARIATIONS IN CENTRAL VIRGINIA.
WASHINGTON, UNIV. OF NUR EQ/CR-3141: CHARACTERIZATION OF TWO-PHASE FLOW USING NEUTRONIC FLUCTUATIONS.
~
WESTINGHOUSE ELECTRIC CORP.
NUREC/CR-2401: PWR FLECHT SEASET SYSTEMS EFFECTS NATURAL CIRCULATION AND REFLUX CONDENSATION. Task Plan Report. NRC/EPRJ/ Westinghouse Report No. 12.
WOOD-LEAVER 8: ASSOCIATES, INC.
NUREC/CR-3177 VO1: METHODS F08t REVIEW AND EVALUATION OF EMERGENCY PROCEDURE CUIDELINES. Volume 1: Methodologies.
NUREG/CR-3177 VO2: METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE GUIDELINES. Volume 2: Applications To Westinghouse Plants.
154
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.
STN-50-454 Byron Station. Unit t Commonsealth Edison Co.
NUREG-OO76 SOP E T7.-50-4 55 Byron Station. Unit b Commonwealth Edison Co.
NUREG-0076 SO2 S TN-50-470 CESSAR 6e tem d0, Combustion Engineering NUREG-0052 501 s
50-413 Catawba Nuclear Stat tor:. Unit 1.
Duke Power Co.
NUREG-0921 50-413 Catawba Nuclear Station. Unit 1.
Duke Power Co.
NUREG-0954 50-414 Catawea Eclear Station. Unit 2.
Duke Power Co.
NUREG-0921 50-414 Catawba Nuclear Station. Unit 2.
Duke Power Co.
NUREG-0954 50-537 Clinch Hiver Breeder Reactor. Project Management Corp.
NUNEG-0968 VOP 50-537 Clinch River Breeder Heector. Project Management Corp.
NUNEG-0968 VOJ SC-237 Ivesden Nutteer Power Station. Unit 2.
Commonwealth Edison Cn.
NUREG-0823 50-16 Enrico Fermi Atomic Pet >er Plant. Unit 1,
Power Reactor Development Co NUREQ/CR-3116 50-341 Enrico Fermi Atomic Power Plant. Unit 2.
Detroit Edison Co.
NUREG-0798 S03 50-416 Grand Ov1f Lclear Station. Unit 1. Mississippi Power & Light Co.
NUREQ/CR-2530 50-417 Orand Gulf Nuclear Station. Unit 2.
Mississippi Power & Light Co.
NUREQ/CR-2530 50-213 Haddam Neck Plant. Cor.ne c t ic u t Yankee 4tomic Power Co.
NUREG-0826 DRFT 50-352 Limerick Generating Station. Unit 1.
Philadelphia Electric Co.
NUREQ/CR-3028 50-309 Maine Yankee Atomic Power Plant. Maine Yankee Atomic Power Co.
NUREQ/CR-2993 50-245 Millstone Nuclear Pw.er Station. Unit 1. Northeast Nuclear Energy Co NUREG-0024 50-219 Oyster Creek Nuclear Power Plant. Joretg Central Power & Light Co.
NUREG-0022 STN-DO-528 Pelo Verde Nucleer Statica. Unit 1.
Arisona Public Service Co.
NUREG-0857 SO4 STN-50-529 Palo Verse Nuclear Station. Ur.it 2.
Arizona Public Service Co.
NUREG-0837 804 STN-50-530 Palo Verde Nuclear F.tation. Unit 3.
Arizona Public Service Co.
NUREG-0857 SO4 50-440 Perry Nvetear Power Plant. Unit 1. Cleveland Electric illuminating C NUREC-0087 SO2 50-441 Perry Noclear Power Plant. Un e t 2.
Cleveland Electric illuminating C NUREG-OOO7 802 50-272 Salms itsclear Generating Statisn. Unit 1.Public Service Electric & Q NUREG-0977 l
50-443 Sem6 coos Nuclear Station. Unit 1 Public Service Co. of New Hampshir NUREG-0896 l
50-444 Bestrack Na,cl e me Station, Uni'; 2. Public Service Co. of New Hampshir NUREG-0096 30-322 Short t:aa Nuclear Power Sta tion. Lo,g Island Lighting Co.
NUREG-0420 S03 STN-50-499 Scuth Te sas Projec t.
Unit 1.
Houston Lighting & Power Co.
NUREQ-0948 874-50-499 Sou th ' Ye xas Projec t.
Unit 2.
Houston Lighting & Power Co.
'50 *.87 Susqtevhanna Steam Electric Station. Unit 1. Pennsg1vania Power & Lig NUREO-0776 805
':0-380 Susqueherna Steam Electric Station. Unit 2 Pennsgivania Power & Lig NUREQ-0776 805 50 36n Willinei B.
McGuire Nuclear Station. Unit 1.
Duke Power Co.
NUREG-0964 50-349 William B.
McGuire Nuclear Station. Unit 1.
Duke Power Co.
NUREO-0422 606 50-370 William B.
McQuire Nuclear Station. Unit 2.
Duke Power Co.
NUREG-0422 SO6 50-370 William B.
McGuire Nuclear Station. Unit 2.
Duke Power Co.
NUREG-0964 50-29 Yankee-Rows Nuclear Power Station. Yankee Atomic Electric Co.
l l
1 l
l l
l l
155
'0""
U.S. NUCLEAR REGULATORY COMMIS$10N BIBLIOGRAPHIC DATA SHEET NUREG-0304, Vol. 8, No. 1 t
- 4. TITLE AND SUBTITLG (Add Volume No, of apprmnate)
- 2. (Leave Dianki Regulatory and Technical Reports Compilation for First Quarter 1983
- 3. RECIPIENT'S ACCESSION NO
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLE TED M ON T H l YEAR
- 9. PE RF ORMING ORGANIZATION N AME AND MAILING ADDRSS tincluue lep Codel DATE REPORT ISSUED I u^a Division of Technical Information and Document Control mon'"
May 1983 Office of Administration 6 '"' * "
U.S. Nuclear Regulatory Comission Washington, DC 20555 8 (Leave biwk)
- 12. SPONSORING ORGANIZ ATION N AME AND M AILING ADDRE SS (include I<p Codel p
i
' t FIN NO Same as 9, above.
i j
- 13. TYPE OF REPORT PE RIOD COV E RE D (inclusive dJtest Reference January - March 1983
- 15. SUPPLEMENTARY NOTES 14 (Leave olan&/
- 16. ABSTR ACT (200 words or less)
This compilation lists all NRC tcgulatory and technical reports published under the NUREG series during the first quarter of 1983.
r 17 KE Y WOh0S AND DOCUME NT AN AL YSIS 17a DE SC RIP TORS I
17b IDEN TIFIE RS OPE N EN DE D TE RVS
- 18. AV AIL ABILITY ST AT E ME N T 19 SE CURITY CLASS (T&,s reporr) 21 NfL OF P AGES Unclassified Unlimi ted 22 gn,ce
,,c q
JRc,ORu 33s m.,,
j
UNITE 3 STATE 3 souu n ctsss:x.it NUCLEA3 REIULATORY COMMISSION Postus a stEs Pt.no WASHINGTON, D C. 20586
."$"$ c Ptnuitto M OFFICIAL SUSpeESS PENALTY FOR PRIVATE USE,8300 l
l i
a 9
0 120555078877 1
99999 US NRC ADM DIV CF TIDC PCLICY & PUB MGT BR-POR NUREG W-501 WASHINGTCN CC 20555 i
d b
i i
a J
l t
i 1
4 m e3
_