ML20095L484
| ML20095L484 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1984 |
| From: | NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V09-N02, NUREG-304, NUREG-304-V9-N2, NUDOCS 8408300279 | |
| Download: ML20095L484 (180) | |
Text
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NUREG-0304 Vol. 9, No. 2
! Regulatory and Technical Reports i
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l Compilation for Second Quarter 1984 l
April - June l
U.S. Nuclear Regulatory Commission Office of Administration i
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I Available from NRC/GPO Sales Program -
Superintendent of Documents Government Printing Office Washington, D. C. 20402 A yest's subscription consists of 4 issues for this publication.
Single copies of this publication are available from National Technical information Service, Springfield, VA 22161 Microfiche of single copies are available from N RC/GPO Sales Program Washington, D. C. 20555
NUREG-0304 Vol. 9, No. 2 Regulatory and Technical Reports Compilation for S::cond Quarter 1984 April - June D:te Published: August 1984 Division of Technical information and Document Control Office of Administration U.S. Nuclear Regulatory Commission Washington, D.C. 20665
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CONTENTS Pref ace..............
v index Tub Main Citation and Abstracts......................................
1 Staff Reports.
Conference Proceedings.................................
Contractor R eports.............................
2 Contractor Report Number Index.........
Personal Author I ndex..................................................
3 S u bject i n d ex................................................
4 5
NRC Originating Organization index (Staff Reports)..............
6 NRC Contract Sponsor Index (Contractor Reports)...
Contractor index............
7 Licensed Facility index.................................................
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PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuc! ear Regulatory Commission (NRC) Staff and its contractors. It is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to:
Division of TechnicalInformation and Document Control k-Policy and Publications Management Branch Publishing and Translations Section Woodmont 501 r
U.S. Nuclear Regulatory Commission y
Washington, D.C. 20555
=
The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, and NUREG/CR-XXXX. These precede the following indexes:
h~
Contractor Report Number Index i
Personal Author Index Subject Index
[
NRC Originating Organization Index (Staff Reports)
F NRC Contract Sponsor index (Contractor Reports)
Contractor index Licensed Facility Index A detailed explanation of the entries precedes each index.
k.
The bibliographic elements of the main citations are the following:
Staff Report E
NUREG-0508: MARK ll CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.
ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048 09570:200.
Where the entries are (1) report number, (2) report title, (3) report author, (t) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control
,g System accession number, (8) the microfiche address (for internal NRC use).
Conference Report
=_
E NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTUPE ROLE OF RISK ASSESSMENT AND 1
RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070.
F Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled eE the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-L ment Control System accession number, (8) the report number of the originating organization, (9) the
+--
microfiche address (for NRC internal use).
Contractor Report y
NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R.
p Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242.
Where the entries are (1) report nurnber, (2) report t tle, (3) report authors, (4) organizational unit of i
5 authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use).
E h
v 7
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- The following abbreviations are used to identify the document status of a report:
3 FC ; e
-ADD addsndum
~
APP '-
appendix --
DRFT -- draft
. ERR ~ - errata N- - number R
revision S - supplement V-volume Availability of NRC Publications.
Copies of NRC staff and contractor reports may be purchased either from b NR'C-GPO Salos Office or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the NRC-GPO Sales Office send a check or money order, payable to the Superintendent of Documents, to the following address:
U.S. Nuclear Regulatory Commission ATTN: Sales Manager Washington, D.C. 20555 You may charge any purchase to your GPO Deposit Account, Master Charge card, or VISA charge card by calling the NRC-GPO Sales Office on (301) 492-9530. Non-U.S. customers must make payment in advance either by intomational Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents.
NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor estabiished codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported.
'n addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings.
All these report codes are controlled and assigned by the NRC Division of Technical Information and
' Document Control.
1 i
i b
vi I
1 Main Citations and Abstracts The report listings in this cornpilation are arranged by report number, where NUREG-XXXX is an NRC staff originated report. NUREG/CP-XXXX is an NRC sponsored conference report, and NUREG/CR-XXXX is an NRC contractor-prepared report. The bibliographic information (see Preface for details)is followed by a brief abstract of the report.
NUREO-0020 VO8 NO3: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of February 29,1984.(Grey Book)
- Division of Dudget
& Analysis.
April 1984.
386pp.
8405220049.
24563:007.
The OPERATING 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 Resource Management from the Headquarters staff of NRC's Office of Inspection and Enforcement, f rom NRC 's Reg ional Of fices, and from utilities.
The three sections of the report are: munthly 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 utilitiesa 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 hopea the report is helpful to all agencies and individuals interested in maintaining an awareness of the U. S.
energy situation as a whole.
NUREO-0020 V08 N04: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As of March 31,1984.(Grey Book)
- Division of Dudget &
Analysis.
May 1984.
407pp.
8406120532.
24916:063.
See NUREG-0020,VO8,NO3 abstract.
NUREO-OO20 VO8 NOS: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of April 30,1984.(Grey Dool)
- Division of Budget &
Analysis.
June 1984.
372pp.
840718002"..
25653:001.
See NUREG-OO20.VO8,NO3 abstract.
NUREO-0040 VO8 N01: LICENSEE CONIRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report, January 1984 - March 1984.(White Book)
- Region 4, Office of Director.
April 1984.
309pp.
8405020039.
24297:126.
This periodical covers the results of inspections performed by the NRC 's Vendor Program Branch that have been distributed to the l
inspected organizations during the period from January 1984 through March 1984.
Also included in this issue are the results of certain 1
in,pections performed prior to January 1984 that were not included in previous' issues of,NUREG-OO40.
1 NUREO-OO90 VO6 NO3: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. July-September 1983.
- Director's Office.
April 1984.
57pp.
8405220091.
24601:296.
Section 208 of.the_ Energy Reorganization Ac,t.of 1974 identifies tan abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report
.of such events to be made to Congret This report covers the period July 1 to' September 30, 1983.
During the report period, there were three abnormal occurrences at the nuclear power plants licensed by the NRC to operate.
The first involved large diameter pipe cracking in boiling water. reactors: the second involved an uncontrolled leakage of. reactor coolant outside primary containment; and the third. involved improper control rod manipulations.
There were seven abnormal occurrences for the other NRC licensees.
Three involved overexposuress two involved medical misadministrationsa.one involved widespread radiological c on tami na t i ons and one involved willful violation of license and a material false statement to the NRC.
There were:no abnormal occurrences reporte_d by the Agreement States.
The report also contains information updating some previously reported abnormal occurrences.
NOREO-OO90'VO6 NO4: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. October
-December 1983.
- Director's Office.
May 1984.
29pp.
8406190041.
25025:220.
See NUREG-OO90,VO6,NO3 abstract.
NUREO-0304 VO9 NO1: REGULATORY AND TECHNICAL REPORTS. Comp ila tion For First Guarter 1984.
- Division of Technical Information & Document Control.
May 1984.
146pp.
8407110023.
25544:227.
This compilation lists all NRC regulatory and technical reports published under the NUREC series during the'first quarter of 1984.
l NOREG-0420 S05: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF SHOREHAM NUCLEAR POWER STATION, UNIT NO.
- 1. Docket No. 50-322.(Long l
Island Lighting Company)
- Division of Licensing.
April 1984.
i 36pp.
8405220021.
24556:341.
Supplement No. 5 (SSER 5) to the Safety Evaluation Rep ort on Long Island Lighting Company 's application f or a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County,.New York, has been prepared by the Of fice of Nuclear Reactor Regulation uf the U.S.
Nuclear Regulatory Commission.
This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued.
NUREG-0525 RO9: SAFECUARDS
SUMMARY
EVENT LIST (SSEL).
- Licensing l
Policy & Programs Branch.
June 1984 53pp.
8407180039.
25654.272.
i The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred saftguards-related events involving nuclear material or facilities regulated by'the U.S.
Nuclear Regulatory Commission (NRC).
Events'are described under the categories of i
I' i
l 2
bomb-related, intrusion, missing / allegedly stolen, transportation, tampering / vandalism, arson, firearms-related, radiological sabotage and miscellaneous.
The information contained in the event descriptions is derived primarily from official NRC reporting channels.
NUREG-0540 VO6 NO1: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. January 1-31,1984.
- Division of Technical Information &
Document Control.
April 1984.
599pp.
8404250005.
24210:184.
This document is a monthly publication containing descriptions of information received and generated by the U.S.
NRC.
This information includes (1) docketed material essociated 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 Referenca to Principal Documents Index.
NUREG-0540 VO6 NO2: TITLE LIST OF DOCUMENTS MADE PJBLICLY AVAILABLE. February 1-29,1984.
- Division of Tec nical Information &
Document Control.
April 1984.
669pp.
84052200'/2.
24553:001.
See NUREG-0540,VO6,NO1 abstract.
NUREG-0540 VO6 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. March 1-31, 1984.
- Division of Technical Information &
Document Control.
May 1984.
632pp.
8406190044 25026:001.
See NUREG-0540,VO6,NO1 abstract.
NUREG-0540 VO6 NO4: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. April 1-30,1984.
- Division of Technical Information &
Document Control.
June 1984.
644pp.
8407170553.
25628:001.
See NUREG-0540,VO6,NO1 abstract.
NUREO-0606 VO6 NO2: UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As Of May 18, 1984. (Aqua Dook)
- Division of Safety Technology.
May 1984.
57pp.
8406120260.
24910:236.
Provides an overview of the status of the progress and plans for resolution of the generic tasks addressing " Unresolved Safety Issues" as repor^ed to Congress.
NUREG-0675 S23: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF DIADLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)
- Division of Licensing.
June 1984.
46pp.
8407110014.
25546:280.
Supplement No. 23 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate the Diablo Canyon Nuclear Pouer Plants (Docket Nos. 50-275 and 50-323),
located in San Luis Obispo County, California, has teen prepared by the Office of Nuclear Reactor negulation of the U.S.
Nuclear Regulatory Commission.
This supplement addresses the licensee's requests for deviations from Section III.G in Appendix R (related to fire protection) of Title 10 of the Code of Federal Regulations Part 50, presents the staff's evaluation and conclusion regarding each request, and summarises the staff's review of the licensee 's requests.
3
NUREO-0725 RO4: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
- Division of Safeguards.
June 1984.
51pp.
8407f90487.
25693:177.
This circular has been prepared in response to numerous requests for information regarding routes used for the shipment of irradiated reactor (spent fuel) subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96-295 The NRC staff must approve such routes prior to their first use in accordance with the regulatory. provisions of Section 73.37 of 10 CFR Part 73.
The information included reflects NRC staff knowledge as of June 1, 1984.
Spent fuel shipment routes, primarily for road transportation, but also including one rail route, are indicated on reproductions of DOT maps.
Also included are the amounts of materiel shipped during the approximate three year period that safeguards regulations for spent fuel shipments have been effective.
In addition, the Commission has chosen to provide information in this document regarding the NRC's safety and safeguards regulations for spent fuel shipments as well as safeguards incidents regarding spent fuel shipments (cf which none have been reported te date).
This additional information is furnished by the Commission in order to convey to the public a more complete picture of NRC regulatory s
practices concerning the shipment of spent f u e '. than could be obtained by the publication of the shipment routes and quantities alone.
NUREG-0748 VO4 NO2: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of February 29,1984.(Orange Book)
Management Inf orma tion 4
Branch.
April 1984.
150pp.
8404240178.
24190:001.
The Operating Reactors Licensing Actions Summary is designed to provide the Management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with the operating power and nonpower reactors.
NUREO-0748 VO4 NO3: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of March 31,1984.(Orange Book)
- Management Information Branch.
May 1984.
355pp.
8405210574.
24530:001.
See NUREG-0748,VO4,NO2 abstract.
NUREG-0748 VO4 NO4: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of April 30,1984. (Orange Book)
- Management Information Branch.
June 1984.
304pp.
8406210444, 25099:073.
See NUREG-0748,VO4,NO2 abstract.
NUREO-0750 V17: NUCLEAR RECULATORY COMMISSION ISSUANCES. January-June 1983. Pages 1-1,196.
- Division of Technical Inf ormation & Document Control.
June 1983.
1,268pp.
8406200559.
25059:019.
Legal issuances of the Atomic Safety and Licensing Board and Appeal Panels, the Commission, the Administrative Law Judge, and NRC Program Offices.
NUREO-0750 V18 102: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES.
July-December 1983.
- Division of Technical Information & Document Control.
December 1983.
131pp.
8406200209.
25209:001.
See NUREG-0750,V17 abstract.
4
NUREG-0750 V18 N06: NUCLEAR REGULATORY COMMISSION ISSUANCES. December 1983.Pages 1,303-1,482.
- Division of Technical Information &
Document Control.
December 1983, 179pp.
8405220259, 24603:012.
See NUREG-0750,V17 abstract.
NUREG-07"O V19 NO1: NUCLEAR REGULATORY COMMISSION ISSUANCES. January 1984 Pp 1-485.
- Division of Technical Information & Document Control.
January 1984.
487pp.
8407130479.
25582:311.
See NUREG-0750,V17 abstract.
NUREG-0750 V19 NO2: NUCLEAR REGULATORY COMMISSION ISSUANCES. February 1984. Pp 487-554.
- Division of Technical Information & Document Control.
February 1984.
75pp.
8407130393.
25575:242.
See NUREG-0750,V17 abstract.
NOREG-0776 SO7: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-387 And 50-388.(Pennsylvania Power And Light Company, Allegheny Electric Division of Licensing.
May 198'-
19pp.
Cooperative Incorporated) 8406190069.
25025:253.
In April 1981, the staff of the Nuclear regulatory Comma=='.an issued its Safety Evaluation Report (NUREG-0776) regarding the application of the Pennsylvania Power & Light Company (the applicant and/or licensee) and the A11egheng Electric Cooperative, Inc.
(co-applicant) for licensos to operate the Susquehanna Steam Electric Station, Units 1 and 2, located on a site in Luzerne County, Pennsylvania.
Supplements 1 and 2 were issued in June 1981 and September 1981, respectively.
Supplement No. 2 also contains NRC staff responses to the comments made by the Advisory Committee on Reactor Safeguards in its report, dated August 11, 1981.
Supplement No. 3 was issued in July 1982 and closed out 5 remaining items.
On July 17, 1982, Operating License NPF-14 was issued to Unit i to allow operation at 5%
cf rated power.
Supplement r.o.
4 was issued in November 1982 and discusses the resolution of several license conditions.
On November 12, 1982, Operating License NPF-14 was amended to remove the 5% power restriction, thereby permitting full power operation of Unit 1.
Supplement 5 was issued March 1983, Supplement 6 was issued in March 1984 and both addressed remaining issues that required resolution prior to operating Unit 2.
On March 23, 1984 Operating License NPF-22 uas issued to allou Unit 2 operation not to exceed 5% of rated power.
This Supplement addresses those issues which require resolution prior to allowing Unit 2 operation at power levels exceeding 5% rated power.
NUREG-0787 SO6: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF WATERFORD STEAM ELECTRIC STATION, UNIT 3. Docket No. 50-382.
(Louisiana Power And Light Company)
- Division of Licensing.
June 1984.
168pp.
8407110007.
25545:008.
Supplement 6 to the Safety Evaluation Report for the application filed by Louisiana Power & Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commisrion.
The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of in f orma t i on submitted by 5
the applicant since the Safety Evaluation Report and its five previous supplements were issued.
NUREO-0800 03.9.3 R1: STANDARD REVIEW PLAN FOR THE REVIEW OF GAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 1 To Section 3.9.3, Appendix A.
SERKIZ,A.W.
Division of Safety Technology.
April 1984.
lipp.
8404170399.
24068:311.
Revision No.
1 to Appendix A of Standard Review Plan Section 3.9.3 incorporates changes that have been developed since the original issuance in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREO-0800 03. 9. 4 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 3.9.4,
" Control Rod Drive Systems." SERKIZ,'A.W.
Division of Safety Technology.
April 1984.
9pp.
8404170381.
24068:322.
Revision No. 2 to Standard Review Plan Section 3.9.4 incorporates changes that have been developed since the issuance of Revision 1 in July 19Et This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer".
NUREO-0800 05.4.6 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 5.4.6,
" Reactor Core Isolation Cooling System (BWR)."
SERKIZ,A.W.
Division of Safety Technology.
April 1984.
lipp.
8404170467.
24091:227.
Revision No. 3 to Standard Review Plan Section 5.4.6.
incorporates changes that have been developed since the issuance of Revision 2 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREO-0800 05.4.7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 5.4.7,
" Residual Heat Removal (RHR) System." SERKIZ,A.W.
Division of Safety Technology.
April 1984.
20pp.
8404170350.
24069:253.
Revision No. 3 to Standard Review Plan Section 5.4.7 incorporates changes that have been developed since the issuance of Revision 2 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREG-OOOO 06. 3 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 6.3,
" Emergency Core Cooling System." SERKIZ,A.W.
Division of Safety Technology.
April *984.
16pp.
8404170375.
24068:331.
Revision No. 2 to Standard Review Plan Section 6.3 incorporates changes that have been developed since the issuance of Revision 1 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" DTP RSB 6-1 is also included with revised page numbers-no other changes were made to the BTP.
NUREO-0800 09.2.1 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No. 3 To Section 9.2.1,
" Station Service Water System." SERKI Z, A. W.
6
Division of Sofoty Technology.
April 1984.
10pp.
8404170057.
24091:257.
Revision No. 3 to Standard Review Plan Section 9.2.1 incorporates changes that have been developed since the issuance of Revision 2 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREG-0800 09. 2. 2 R2: STANOARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 9.2.2,
" Reactor Auxiliary Cooling Water Systems." SERKIZ,A.W.
Division of Safety Technology.
April 1984.
12pp.
8404170042.
24091:283.
Revision No. 2 to Standard Review Plan Section 9.2.2 incorporates changes that have been developed since the issuance of Revision 1 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREG-0800 10. 3 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No.3 To Section 10.3,
" Main Steam Supply System." SERKIZ,A.W.
Division of Safety Technology.
April 1984.
12pp.
8404170062.
24069:241.
Revision No. 3 to Standard Review Plan Section 10.3 incorporates changes that have been developed since the issuance of Revision 2 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREO-0800 10.4.7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Ed i t ion. R evi sion 3 To Section 10.4.7,
" Condensate And Feedwater System" And BTP ASD 10-2,
" Design Guidelines For Avoiding Water Hammer.
" SERKI Z, A. W.
Division of Safety Technology.
April 1984.
lipp.
8404170353.
24068:347.
Revision No. 3 to Standard Review Plan Section 10.4.7 and BTP ASD 10-2 incorporates changes that have been developed since the issuance of Revision 2 in July 1981.
This revision incorporates the resolution of Unresolved Safety Issue A-1,
" Water Hammer" NUREG-0828: INTEGRATED PLANT SAFETY ASSESSMENT REPORT. SYSTEMATIC EVALUATION PROGRAM. Big Rock Point Plant. Docket No.
50-155.
(Consumers Power Company)
- Division of Licensing.
May 1984.
800pp.
8406120255.
24917:115.
The Systematic Evaluation Program was initiated in February 1977 by the U.S.
Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm 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 Big Rock Point Plant, operated by Consumers Pouer Company located in Charlevoix, Michigan.
Big Rock Point is one of ten 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.
It also addresses a najority of the pending licensing actions for Dig Rock Point, which include TMI Action Plan requirements and implementation criteria for 7
i
-~.-
. roco' vad f genovic :iosues.
Equipasnt and procedural changos have.been 1
. identified as a result of the review.
+
SAFETYL ' VALUATION REPORT RELATED TO THE OPERATION OF NUREO-0830 SO3:
E C ALLAWAY PLANT, UNIT NO.1. Doc ke t No. 50-483.(Union Electric Company) g Division of Licensing.
May 1984.
-194pp.
8405290428.
24695:074.
Supplement No. 3 to the Safety Evaluation Report related to the operation of-the Callaway-Plant, Unit No.
1' resolves open items and updates information contained-in the~ Safety Evaluation, dated October.
1981.
Supplements 1 and 2 -dated January 1982'and. June'1983, respectively also updates the information-contained in the Safety.
Evaluation Report.
Supplement No.
1 contained the ACRS Report issued on November 17,'1981.
The Safety. Evaluation Report and its supplements pertain to the application for a license to-operate the Callaway Plant. filed by the Union Electric Company-on October 19, 1979.
NUREO-0837 VO3 NO4: NRC..TLD DIRECT RADIATION MONITORING NETWORK. Progress Report September-December 1983. COSTELLO,F.J THOMPSON,T.s COMEN, L. 's et al.
Region 1, Office of Director.
May.
~
1 1984.
247pp.
8406060392.
24741:148.
2 This report provides -the status and results of the NRC 1
Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring
^
Network.
It presents the-radiation levels measured in the vicinity of j
NRC licensed facility sites throughout the country for the fourth quarter of 1983.
i NUREG-0853 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF CLINTON POWER STATION, UNIT NO.1. Docket No. 50-461.(I111nois Power t
Company,et al)
- Division of Licensing.
May 1984.
40pp.
l 8406190045 25025:274 i
Supplement No. 3 to the Safety Evaluation Report on the l
application filed by Illinois-Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1 has been prepared by the Office of Nuclear Reactor l
Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is i
located in Harp Township, DeWitt County, Illinois.
This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplements No.
1 and No.
2.
j NUREO-0871 VO3 NO1:
SUMMARY
INFORMATION REPORT. Data As Of December j
31,1983. (Brown Book)
- Management Information Branch.
June 1984 52pp.
8406250269.
25138:239.
Provides summary data concerning NRC and its licensees for j
general'une by the Chairman, other Commissioners and Commission staf f l
offices, the Executive Director for Operations, and the Office Directors.
.4
[
i l
NUREO-0876 SO4: SAFErY EVALUATION REPORT RELATED TO THE OPER ATION OF THE BYRON ' STATION, UNITS 1 AND 2. Doc k et Nos. STN 50-454 And STN 50-455.(Commonwealth Edison Company)
- Division of Licensing.
May 1984.
32pp.
8406060010.
24847:254.
l Supplement No.-
4 to the Safety Evaluation Report related to I
s
Commanuealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplements 1 through 3.
NUREG-0892 SO5: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF WPPSS NUCLEAR PROJECT NO.
- 2. Docket No. 50-397.(Washington Public
- Division of Licensing.
April 1984.
41pp.
Power Supply System) 8404240005.
24189:087.
Supplement No. 5 to the Safety Evaluation Report on the application filed by Washington, Public Power Supply System for a license to operate the WPPSS Nuclear Project No.
2, located in Richland, Washington, has been prepared by tne Division of Licensing, Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Comnission.
This supplement is to update our evaluations on issues identified in the previous Safety Evaluation Report and Supplements that need resolution prior to issuance of the full power operating license.
NUREG-0936 VO3 N01: NRC REGULATORY AGENDA. Quarterly Report January-March 1984.
- Division of Rules and Record s.
April 1984.
182pp.
8405020032.
24287:128.
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 issued each quarter.
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 VO3 NO1: ENFORCEMENT ACTIONS:SIGNIFICANT ACTIONS RESOLVED. Quarterly Progress Report (January - March 1984).
- Director 's Of fice, Office of Inspection and Enforcement.
April 1984.
347pp.
8405220263.
24595:008.
This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (January - March) 1984 and includes copies of letters, notices, and orders sent by the Nuclear Regulatory Commission to licensees with respect to these enforcement actions and the licensees' responses.
It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, in the interest of promoting public health and saf ety as well as common defense and security.
NUREG-0954 802: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF CATAWBA NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-414.
(Duke Power Company,et al.)
- Division of Licensing.
June 1984.
134pp.
8407130504.
25578:156.
This report supplements the Safety Evaluation deport (NUREG-0954) issued in February 1983 and Supplement 1 issued in April 1983 by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission with respect to the application filed by Duke Power 9
Ce:pcny, N3rth Corolino Municipal Pcwor Agoney Number 1 North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc., as app licants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively).
The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie.
This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory issues and license conditions identified in the Safety Evaluation Report.
4 NUREG-0974: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF LIMERICK GENERATING STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-353.(Philadelphia Electric Company)
Division of Licensing.
April 1984.
320pp.
8404170288.
24089:096.
The information in this Final Environmental Statement is the second assessment of the environmental impact associated with the construction and operation of the Limerick Generating Station, Units 1 and 2.
The first assessment was the Final Environmental Statement related to the construction of the facilities.
The present assessment is the result of the NRC Staff review of the activities associeted with the proposed operation of the station.
NUREG-0980: NUCLEAR REGULATORY LEGISLATION. FOTIAS,A.
Office of the Executive Legal Director.
June 1984.
649pp.
8407130401, 25580:001.
NUREG-0980 is a comp ilation of. nuclear regulatory leg islation and other relevant material through the 97th Congress, 2nd Session.
This compilation has been prepared for use as a resource document, which the NRC intends to update at the end of every Congress.
Contents of NUREG-0980 include:
The Atomic Energy Act of 1954, as amendeds Energy Reorganization Act of 1974, as amendeds Uranium Mill Tailings Radiation Control Act of 1978: Low-Level Radioactive Waste Policy Acts Nuclear Waste Policy Act of 19823 and NRC Authorization and Appropriations Acts.
Other materials included are statutes and treaties on export licensing, nuclear non-proliferation, and environmental protection.
Sections of Title 5, United States Code, on Administrative Procedure are also included.
NUR EG-0989: SAFETY EVALUATION REPORT RELATED TO THE OI7PATION OF RIVER BEND STATION. Docket No. 50-458.(Gulf States Utilities tompany, Cajun Electric Power Cooperative)
- Division of Licensing.
May 1984.
597PP.
8405310124.
24735:001.
The Safety Evaluation Report for the application filed by the Gulf States Utilities Company, as applicant and owner, for a license to operate the River Bend Station (Docket No. 50-458) has been prepared by the office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is located near St.
Francisv111e, Louisiana.
Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public.
NUREG-1020LD VO1: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al.
v.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1,
Docket 10
50-299.
- Office of Nuclear Reactor Regulation, Director.
June
'1994.
152pp..8407130502.
25579:089.
LThis report. documents a review by'the Nuclear Regulatory
-Commission (NRC) staff of the General Public Utilities v.~
Babcock &
Wilcox-lawsuit-record to assess whether any of the staff's previous conclusions or-their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in-light of the information contained in the lawsuit record.
Details of the lawsuit record-are provided in the appendices contained in Volume 2 of this report.
I i
NUREO-1020LD VO2: OPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON i
TMI-1. General Public Utilities Corporation.et al.
v.
The Babcock &
l Wilcox Company.et al.Three Mite Island Nuclear Station, Unit 1.
Docket 50-289..*
Office of Nuclear Reactor Regulation, Director.
June 1984.
875pp.
8407130415.
2"576:001.
See NUREG-1020LD,VO1 abstract.
NUREO-1026: FINAL ENVIRONMENTAL STATEMF.NT RELATED TO.THE OPERATION OF BRAIDWOOD STATION.UNTTS 1 AND 2. Docket Nos. STN 50-456 And STN 50-457.(Commonwealth Edison Company)
- Division of Licensing.
June 1984.
276pp.
D407180017.
25682:126.
The information in this statement is the-second assessment of the environmental impact associated with'the construction and operation of the Braidwood Station, Units 1 and 2, located in northeastern Illinois l
within Reed Township, Will County, Illinois.
The first assessment was the Final Environmental Statement related to construction issued in.
July 1974 prior to issuance of the Braidwood-Construction Permits.
The present assessment is the result of the NRC' staff review of the activities associated with the proposed operation of the plant.
NUREO-1028: RUPTURED CESIUM-137 WELL-LOQGING SOURCE AT SHELWELL SERVICES,INC., HEBRON, OHIO. AXELSON,W.
Division of Radiological &
Materials Safety Programs.
April 1984.
135pp.
8405220266.
i 24601:162.
l This U.S.
Nucler.' Regulatory Commission report documents the-circumstances-surrounding the September 13, 1983, cesium-137 sealed source rupture incident at Shelwell Services, Inc., facility in Hebron, Ohio.
It focuses on the period from approximately 4:00 p.m.
(EOT) on September 134 1983, when the source ruptured, to October 5, 1983, when the radiological emergency response aspects of the' event were concluded.
Information outside these periods is recounted as j
necessary.
The incident resulted in radiation doses to two licensee employees that exceeded the regulatory limits for whole-body and
{
extremity exposures, and contamination of the licensee's employees,
-families, and friends.
The emergency response required th e combined i
efforts of NRC, the U.S.
Department of Energy, and state p ersonnel.
The report describes the factual information and significant findings associated with the event and, thereby, provides a data base for l
l l
subsequent detailed analyses and recommendations by various NRC l
offices.
i a.
NUREO-1038 S01: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF
)
SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1. Docket No. STN 50-400.
l (Carolina Power And Light Company, North Carolina Eastern Municipal j
Power Agency)
- Division of Licensing.
June 1984.
52pp.
!~
l 11 t
3 n.
C407180053.
25665:341.
. Supplement No. I to the Safety Evaluation Report'for the-
-application-filed by Carolina Power and Light Company and North' Carolina Eastern Municipal Power Agency for a license to operate the l-
-Shearon. Harris Nuclear-Power Plant, Unit 1 (Docket-No. 50-400),
located.in Wake and Chatham Counties, North Carolinai has been prepared by'the Office of Nuclear Reactor Regulation of the U.S.
l
' Nuclear Regulatory Commission.
.This supplement provides more recent j
information regarding resolution of some of the open items identified i
in'the Safety Evaluation Report.
This supplement also provides and discusses the: recommendations of the Advisory Committee on Reactor Safeguards in its report on Shearon Harris, dated' January 16, 1984.
3 s
NUREO-1051: l SAFETY EVALUATION REPORT RELATED TO THE' RENEWAL OF THE OPERATING l.ICENSE FOR THE RESEARCH REACTOR AT THE UNIVERSITY OF KANSAS.Dotset.No. 50-148.
- Division of Licensing. :May 1984.
68pp.
t 840606Ct19.
24847:182.
};
This Safety Evaluation Report for the application filed by the University of Kansas (KU). for a renewal of Operating. License R-78 to continue to operate the KU 250-kw open pool -training reactor has been
}
prepared by the Office oft Nuclear Reactor Regulation of the U.S.
i Nuclear Regulatory Commission.
The facility is owned and operated by i
the University of Kansas and is located on the KU campus in Lawrence, Douglas County, Kansas.
the staff concludes that the reactor facility can continue to be operated by KU without endangering the health and safety of the'public.
NUREG-1052: FEDERAL / STATE COOPERATION IN THE LICENSING OF A NUCLEAR POWER PROJECT. A Joint Process Between The U. S.
Nuclear Reg ulatory Commission And The Washington State Energy Facility Site Evaluation Council.
- Office of Nuclear Reactor Regulation, Direc tor.
May 1984.
53pp.
8406230318.
25131:265.
ThisTreport summarires and documents a Joint environmental review and licensing process established between the U.S.
Nuclear Regulatory Commission (NRC) and the Washington State Energy Facility Site 1
l Evaluation Council (EFSEC) in 1980-83 for the Skagit/Hanford Nuclear project (S/HNP).
It documents the agreements made between the e
agencies to prepare a Joint environmental impact ststement responsive l-tc the requirements.of the National Environmental Policy Act~1969 (NEPA) and the Washington State Environmental Policy Act.
These agreements also established protocol to conduct ' Joint public evidentiary hearings on matters of mutual jurisdiction,thereby reducing'the duplication of effort and in c r e'a s in g the efficiency of the resources of Federal and State governments and other entities involved in the process. 'This report may provide guidance and rationale to licensing bodies that may wish to adopt some of the procedures discussed in the report in the event that they become involved in the licensing-of a nuclear power plant project.
The history of the S/HNP and of the agreement processes are discussed.
Discussions are provided on implementing the Joint review process.
A separate section'is included which presents independent evaluations of l-the process by the applicant, NRC, and EFSEC.
Cooperating Federal agencies in the environmental review included the U.S.
Department of Energy, the Bonneville Power Administration, and the Bureau of Reclamation.
)
12
~.
'NUREO-1055i.IMPROVINO QUALITY AND'THE ASSURANCE OF GUALITY IN THE.
DESIGN AND CONSTRUCTION OF COMMERCIAL NUCLEAR POWER PLANTS. A Report To~ Congr ess. : ALTMAN, W. J. ANKRUM, T. 0 BRACH, W.
GA Branch.
May-1984.
524pp.
8406010533. :24763:001.
At the request of. Congress, NRC conductedLa' study of' existing and
' alternative programs for improving quality and1the assurance of quality in the. design and construction of commercial nuclear power plants.
A' primary focus 1off the study was to determine.the underlying causes:of major quality-related problems in the construction of some
- nuclear : power plants: and the untimely detection and correc tion of
-these problems.
.The study concluded that the root cause>for major.
l quality-related problems was the failure orzinability of some utility nanagements to. effectively implement a management system'that ensured-2 adequate control over all aspects of the project.
-Thesa management
- shortcomings arose in part.from inexperience on the part of some Project: teams in the construction of nuclear power-plants.
As a corollary,.NRC's past licensing and inspection peactices did not adequately screen construction permit applicants for overall capability to manage-or provide effective management oversight over-the construction project.'
The study recommends a number of improvements in industry and NRC programs.
' NUR EG-1056:. REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM GENERATORS.
- Office of Nuclear Reactor Regulation. Director.
April 1984.
124pp.
8404240014.
24189:131.
'This is a report on'a trip to Japan'by personnel of the U.S.
Nuclear Regulatory Commission in 1983 to exchange information on steam generators of nuclear pouer plants.
Steam generators of Japanese-pressurized water reactors have experienced nearly all'of the forms of-degradations-that have been experienced in'U.S.
recirculating-type steam generators, except for denting and pitting.
More tubes have been plugged per year of reactor operacion in Japanese than in U.S.
steam generacors, but much of the Japanese tube plugging is preventative rather than the result of leaks experienced.
The number of leaks per reactor year is much smaller f or. Japanese than for U.S.
No steam generators have been replaced in Japan uhile several have replaced-in the U.S.
The Japanese esperience may be related to their very stringent inspection and maintenance programs for steam generators.
NUR EG-1058: TECHNICAL SPECIFICATIONS FOR CALLAWAY PLANT, UNIT NO.
1.
Docket No. STN'50-483.(Union Electric Company) ANDERSON,F.D; Division of Licensing.
June 1984.
490pp.
8407020225.
25230:206.
The Calloway Plant, Unit No.
1, Technical Specifications-were prepared by the U.S.
Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable'to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the' protection of the health and safety of the public.
NUREG-1059: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE
-OPERATING LICENSE FOR THE UNION CARBIDE SUBSIDIARY B,INC. RESEARCH REACTOR. Docket No. 50-54.
- Division of Licensing.
June 1984.
98pp.
8407180046.
25683:041.
-This Safety Evaluation' Report for the application filed by the Union Carbide Subsidiary B, Inc. (UNC) for a renesal of operatin0 license R-81 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.
S.
e 13
._.=
NuclocrTRsgulatory Conmission.
The facility is owned-and operated by thel Union Carbide, Subsidiary B, Inc. and is located in the City of
- Tuxedo, Orange County, New York..The staff concludes that the reactor, facility can continue to be operated by UNC without endangering the v-n>
health and' safety of the public.
'NUREG-1062: DOSE: CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
MARQULIE3.T.S.; MARTIN,J.A.
Division-of Risk' Analysis & Operations (post 840429).
May 1984.
227pp.
8406230205.
25132:045.
i This report presents a set of precalculated doses based on a set j
of. postulated accident releases and intended for use in emergency planning-and~ emergency' response.
Doses were calculated ~for the PWR
. (Pressurized Water Reactor) accident categories of the' Reactor Safety
. Study (WASH-1400) using t h e CR AC (Calculations of Reactor Accident Consequences)' code.
Whole body-and thyroid doses are presented for a selected set of weather cases.
For each weather case these calculations uere performed for various times and-distances including three different dose pathways--cloud (plume) shine, ground. shine and inhalation.
During an emergency this information can'be useful since
'it is immediately available for projecting offsite radiological doses
. based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms).
It can be used for emergency drill-scencrio-development as well.
NUREC-1063: STEAM GENERATOR OPERATING EXPERIENCE UPDATE 1982-1983.
FRANK L.
Division of Engineering.
June 1984.
50pp.
8406270122.
25173:231.
This report is a continuation of earlier reports by the' staff addressing pressurized water reactor. steam generator op era ting experience.
NUREG-0886, " Steam Generator Tube Experience," published in February 1982 summarized experience in domestic'and foreign plants through December 1981.
This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983.
Included are new problems encountered with secondary-side loose parts, sulfur-induced stress assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators.
The status of Unresolved Safety Issues A3, A4, and A5 is also discussed.
NUREG-1065: ACCEPTANCE CRITERI A FOR THE LOW ENRICHED URANIUM REFORM AMENDMENTS. EMEIGH,C.W.s GUNDERSEN,G.E.s WITHEE.C.J.
Division of Safeguards.
May 1984.
49pp.
8406080305.
24877:126.
This report documents a standard format suggested by the NRC for use in. preparing fundamenta! nuclear material control plans as required by the Low Enriched Uranium Reform Amendments (portions of 10 CFR Part 74).
The report also describes the necessary contents of a comprehensive plan and provides example acceptance criteria which are intended'to communicate acceptable means of achieving the performance capabilities of the Reform Amendments.
By using the suggested format, the license applicant uill minimize administrative problems associated with the submittal, review and approval of the FNMC plan Preparation of the plan'in accordance with this format will assist the NRC in evaluating the plan and in standardizing the review and licensing process.
However, conformance with this guidance is not required by the NRC.
A license = applicant who employs a format that provides an equal level of completeness and detail may use their own format.
14 l
l 4
s NUREC-1066:-COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REGUIREMENTS ON OPERATING. PLANTS DESIGNED BY BABCOCK AND WILCOX.-
-THOMA,'J. O. s - HERNAN, R.J KADAMB I, N.~ P. ; et al.
Division of. Licensing.
.May 1984.
.186pp..8406020464.
24800:001.
i This = report provides :the results of a study conducted by:the U.S.
Nuclear' Regulatory Commission staff to compare the degree to which
.eight. Babcock and Wilcox (B&W) designed licensed nuclear p ower plants
.have complied with the requirements in NUREG-0737, " Clarification of
'TMIl Action' Plan Requirements".
The eight-licensed operating plants e xamined ' are as f ollows:
Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3i Davis Besse, Oconee Units 1,
2 and 3, Rancho Seco, and Three Mile Island Unit.1 (TMI-1).
The purpose of this audit was to
}
establish the progress of the'TMI-1 licensee, General Public Utilities (CPU) Nuclear Corporation, in: completing the long-term requirements in NUREG-0737: relative to the other B&W licensees examined.
f NUREC-1071: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL'OF SOURCE MATERIAL LICENSE NO. SUB-526. Docket No. 40-3392. ( A111ed.Ch emical Company UF6 Conversion Plant)
- Division'of Fuel Cycle & Material l
Safety.
May 1984.
110pp.
_8405310034.
24737:072.
l This-Environmental Impact Appraisal-is issued by the U.S.
Nuclear Regulatory Commission in response to an application by Allied Chemical l
Company for renewal of Source Material License No. SUB-526.
NUREG-1074: DRAFT ENVIRONMENTAL. STATEMENT RELATED TO THE OPERATION OF HOPE CREEK GENERATING STATION. Docket No. 50-354.(Public Service-Electric And Oas Co And Atlantic City Electric.Co)
- Office of l
Nuclear Reactor Regulation, Director.
June 1984.
227pp.
8407110001.
25544:001.
The Draft Environmental Statement related to the operation of Hope Creek Generating Station, located in Salem County, New Jersey, has been prepared by the Office of Nuclear Reactor Regulation of the U. S.
Nuclear Regulatory Commission.
The statement reports on staff's i
review of the environmental and socio-economic impacts.of plant i
operation.
Comments received on this document will be
- included and addressed in the Final Environmental Statement.
NUREC-1077: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SPECIAL j
NUCLEAR MATERIAL LICENSE NO. SNM-21. Docket No. 70-25.
(Energy Systems Group Rockwell Intsrnational. Corporation)
- Division of Fuel Cycle &
i Material. Safety.
June 1984.
121pp.
8406280455.
25195:047.
This Environmental Impact Appraisal is issued by the U.S.
Nuclear Regulatory Commission in response to an application by Energy Systems j
Oroup,.Rockwell International Corporation, f or renewal of Special l
Nuclear Material (SNM) License No. SNM-21.
i i
NUREG-1078: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SPECIAL I
NUCLEAR MATERIAL LICENSE NO. SNM-1097. Docket No. 70-1113. (General Division of Electric Company,Wilmington Manufacturing Department) l Fuel Cycle & Material Safety.
June 1984 84pp.
8407020195.
25275:094.
This Environmental Impact Appraisal is issued by the U.S.
Nuclear i
~ Regulatory Commission in response to an application by General Electric Company. Wilmington, NC, for renewal of Special Nuclear l
Material (SNM) License No. SNM-1097.
l 15 1
O
NURED-109(
- l. S.
NUCLEAR RECULATORY COMMISSION 1983 ANNUAL REPOR T.
MAHER,W Jffice of Resource Management, Directo..
June 1984.
206pp.
8406250266.
25138:001.
This-report addresses all NRC activities, policies, and decisions made during the reporting period, complete with illustrations,, charts, and treatment of technical material in lay language for consumption by the lay public.
NUREO/CP-OO52: - NRC NUCLEAR WASTE MANAGEMENT GEOCHEMISTRY
'83.
ALEXANDER,D.H.r BIRCHARD,G.F.
Division of Health, Siting & Waste Management.
May 1984.
541pp.
8406060366.
24846:001.
This document-summarizes papers and panel discussions presented at the Office of Nuclear Regulatory Research sponsored conference on
" Nuclear Waste Management Research on Geochemistry of HLW Disposal" The conference was held at the United States Geological Federal Center in Reston, Virginia on August 30-31, 1983.
The purpose of the meeting uas to present results from NRC sponsored research and to identify regulatory research issues which need to be addressed prior to licensing a high level waste repository.
Important summaries of technical issues and recommendations are included with each paper.
The issues reflect areas of technical uncertainty addressed by the NRC Research program in geochemistry.
The objectives of the NRC Research Program in geochemistry are to provide a technical basis for waste management rulemaking, to provide the NRC Waste Management Licensing Office with information that can be used to support sound licensing decisions, and to identify investigations that need to be conducted by DOE to support a license application.
NUREO/CR-2OOO VO3 N3: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of March 1984.
- Dak Ridge National Laboratory.
April 1984.
175pp.
8405010064.
ORNL/NSIC-2OO.
24257:064.
This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER da ta 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 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-Ol61. Instruction for Preparation of Data Entry Sheets f or Licensee Event Rep orts.
The LER summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility.
Component, system, keywords, and component vendor indexes follow the summaries.
The components, systems, and vendors are those identified by the utility when the LER form is initiated; the keywords are assigned by the computer using correlation tables from the Sequence and Search i
System.
NUREO/CR-2OOO VO3 N4: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of April 1984.
- Dak Ridge National Laboratory.
May 1984.
180pp.
8406040026.
ORNL/NSIC-2OO.
24805:078.
l See NUREG/CR-2OOO,VO3,N3 abstract.
l NUREO/CR-2OOO VO3 N5: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of May 1984.
- Dak Ridge National Laboratory.
June 1984.
129pp.
8407160280.
ORNL/NSIC-2OO.
25625:070.
l 16
~
b Sao NUREG/CR-2000,V03,N3 cbotreet.
NUREC/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE J
TRANSPORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /
Radionuclide Transport Model FETRA.' DNISHI,Y.s THOMPSON,F.L.
j Battelle Memorial Institute, Pacific Northwest Laboratories.
May
-1984.
112pp.
8406230270.
PNL-5088-1.
25132:269.
.The finite element nodel, FETRA, is an unsteady, two-dimensional (longitudinal and lateral) model for simulating the transport of
~
sediment and contaninants (e.g.,
radionuclides, heavy metals, pesticides) in coastal uaters.
FETRA includes major transport and fate mechanisms explicitly, including sediment / contaminant 1
interactions.
The model was tested by applying it to the Irish Sea to simulate wind generated uaves and the migration of sediment and (137)Cs.
The model predicted distributions of suspended sands suspended silts suspended clays (137)Cs sorbed by each of the three sizes of suspended sediments; dissolved (137) Css bed sediment size fractionss and (137)Cs sorbed by bed sand, bed silt, and bed clay over a tuo-nonth period in 1974.
FETRA predicted that approximately 82%.
0.002%, and 18% of the total (137)Cs remaining in this study area were dissolved,. suspended sediment-sorbed, and bed-sediment-sorbed radionuclides, respectively.
NUREC/CR-2424 VO2: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing f or FETRA.
ONISHI,Y.; THOMPSON,F.L.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May 1984.
89pp.
8407110172.
PNL-SOO8.
25542:134.
FETRA is a finite element model for simulating the sediment and containment transport to surface water.
The model was applied to a test site in the Irish Sea and modified to account for wave mechanisms that affect sediment suspension.
Volume 2 of this report presents a very brief users guide for FETRA and a computer program listing of the model.
NUREC/CR-2531 R02: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
SCOFIELD,N.R.s HARDY,H.A.; LAATS.E.T.
EG&G, Inc.
April 1984, 102pp.
8405220080.
EOC-2164.
24556:173.
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 th e many domestic and foreign water reactor safety research programs.
The NRC/DAE Data Bank Program provides a central computer storage 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 administration portion of the program provides data entry, documentation, training, and advisory services to users and the NRC.
The NRC/DAE Data Bank and the capabilities of the data access software are described in this document.
NURE0/CR-2552: CRAC2 MODEL DESCRIPTION. RITCHIE, L. T. J ALPERT,D.J.s BURKE,R.P.s et al.
Sandia Laboratories.
April 1984 95PP.
8405220186.
SAND 82-0342.
24602:188.
The CRAC2 computer code is a revised version of CRAC (Calculation 17
Of Roccter Accidsnt Consequancos) which-wcs dovolopGd for tho Rocctor Safety Study.
This document provides an overview of th e CRAC2 code and a description of each of the models used.
Significant improvements incorporated into CRAC2. include an improved weather sequence sampling technique, a new evacuation model, and new output capabilities.. In addition, refinements have been made to the atmospheric < transport and deposition model.
Details of the modeling differences between CRAC2 and CRAC are emphasized in the model descriptions.
NUREC/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT. RAWLINGS,G.s ANTONNEN,G.1 CHAMNESS,M.s et al.
Golder Associates.
April 1984 171pp.
d 8405220085.
24594:100.
The purpose of the complete project is to provide NRC with technical assistance to enable the focused, adequate review by NRC of the aspects related to design and construction of an undergound test facility and final geologic repository as presented by the Department of Energy (DOE).
The study presented in this report covers the identification of characteristics which influence design and construction of a geologic repository in domal salt.
This report has identified five key issues, i. e., constructibility, thermal response, mechanical response, hydrologic response, and geochemical response.
This report involves both short-term (up to closure) and long-term (post closure) effects.
The characteristics of domal salt and its environment are described under the headings of stragraphic/ structural, tectonic, mechanical, thermal and hydrologic.
Characteristics are separated into parameters (quantified and measured) and factors (qualitative).
The characteristics are then subjectively ranked by their influence on the key issues.
This takes into account the availability and suitability of conservative design / construction techniques, uncertainty in model and model sensitivity to the characteristic.
NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF. RAWLINGS,G.i ANTONNEN,G.J FINDLEY,D.4 et a l.
Golder Associates.
April 1984.
156pp.
8405220065.
813-1162C.
'24564:216.
The purpose of the complete projecf'is to provide NRC with technical assistance to enable the focused, adequate review by NRC of the aspects related to design and construction of an underground test facility and final geologic repository as presented by the Department of Energy (DOE).
The study presented in this report covers the identification of characteristics which influence design and construction of a geologic repository in tuff at the Nevada Test Site (NTS).
This report has identified five key issues, i.e.,
constructibility, thermal response, mechanical response, hydrological response, and geochemical response.
This report involves both short-term (up to closure) and long-term (post closure) effects.
The characteristics of tuff and its environment are described under the headings of stratigraphic /structual tectonic, mechanical, thermal and hydrologic.
Characteristics are separated into parameters (quantified and measured) and factors (qualitative).
The characteristics are then subjectively ranked by their influence on the key issues.
This ranking took into account availability and suitablity of conservative design / construction techniques, uncertainty in model and the model sensitivity to characteristics.
18 i
,, - ~,. - -. +
Y V
- NURED/CR-2675 VOS: RELEVANCE OF BIOTIC PATHWAYS TO THE LOND-TERM REOULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
MCKENZIE, D. H.'s CADWELL,L.L.s.'EBERHARDT L.E.s e c - a 1.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May 1984.
49pp.
8406230239.
PNL-4241, 25130:267.
Licensing'and regulation of commercial low-level. waste-(CLLW)
~ burial facilities require that anticipated. risks associated _with
^
burial sites be evaluated for the life of the facility.
This. work
- reviewed:the existing-capability to evaluate dose.to man resulting
~from the potential redistribution of. buried radionuclides'by plants and animals.' _Through' biotic transport,-radionuclides can be moved to
' locations where-they can enter exposure pathways:to man.
Wo found thatl predictive models currently in use did not address the long-term r
risks resulting from the cumulative transport _of_ radionuclides.
Although reports'in the--literature confirm that biotic transport phenomena are common, assessments routinely ignore;the associated i:
- risks or dismiss them as' insignificant.
To determine.the potential.
impacts of biotic transport, we made order-of-magnitude _ estimates of the dose to man for biotic transport processes at reference arid and
~
4-humid CLLW disposal-sites.
Es timated doses to site residents after_
assumed loss of institutional control were comparable to dose estimates for.the. intruder-agricultural scenario defined in the DEIS for 10 CFR 61 (NRC)._ The reported lack of potential importance of 4
biotic transport at l ou-l eve l wa s t e sites in earlier assessnent studies.is'not confirmed by order of magnitude estimates presented.in 4
j this study.
i NUREG/CR-2679 VO4: ADVANCED REACTOR SAFETY RESEARCH,GUARTERLY REPORT, 4
2 OCTOBER-DECEMBER 1982.
- Sandia Laboratories.
April 1984.
207pp.
8406210433.
SAND 82-0904.
25100:097.
This report describes progress in a number of activities dealing i
with current safety issues relevant to both light water reactors (LWRs) and breeder reactors.
The work includes a broad range of i
experiments to simulate accidental c'onditions to provide the required 4
data base to understand important accident sequences and to serve as a j-basis for development and verification of the complex computer simulation models and codes used in accident analysis and' licensing l
reviews.
Such a program must include the development of analytical-j models, verified by experiment, which can be used to-predict' reactor and safaty system performance under a broad variety of abnormal conditions.
Current major emphasis is focused on providing information to NRC relevant to.(1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of'the proposed Clinch River Breeder Reactor.
l-i l
NUREO/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL
[
HEATER ROD DESIGN ON GUENCH BEHAVIOR. GOTTULA,R.C.
EG&G, Inc.
April 1994.
105pp.
8405220051.
EGG-2186.
24551:225.
j.
A separate effects experiment program was' conducted on a bundle E
of nine electrical heater rods in the Loss-Of-Fluid Test (LOFT) Test
[
Support Facility (LTSF).
The objective of the experiment program were to (a) evaluate the ef fect of cladding external.thermocouples on the guench (cooling) behavior of a cartridge-type nuclear fue11 rod simulator, (b) determine how accurately cladding external thermocouples measure cladding temperature during a high pressure i
quench, (c) provide a functional and reliability test for cladding-embedded thermocouples that are prototypes of a design to be 19 1
4 yy---,y9-,-.w ypry
.g*,-,.y--..,,v..,-.._.-.-.g....sg,,,y
-,.n,-..g..gw,--,,go,yww,wrm,&cw-,-.e-.,--,m,,,,-ic,,-,ew.ww-.y-,.wwww-+pe,,.
uOCd in the LOFT fuol redo, cnd (d) comporo tho quGnch bohcvior of a cartridge-type heater rod (which simulates a fuel pellet-cladding gap) with that of a solid-type heater rod (without a pellet-cladding gap) under thermal-hydraulic conditions that could occur during the blowdown phase (O to 10 s) of a large-break loss-of-coo lan t acc ident in a pressurized water reactor.
The prototype cladding-embedded thermocouples did not function correctly during the experiments however, useful data were obtained such that the objectives of the experiment program could be met.
NUREC/CR-2803: IMPROVED FIELD EXPERIMENTAL DESIGNS AND GUANTITATIVE J
EVALUATION OF AGUATIC ECOSYSTEMS. MCKENZIE,D.H.; THOMAS J.M.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May 1984.
31pp.
8405210607.
PNL-4138.
24534:242.
We used the paired-station concept and a log transformed analysis of variance methods to evaluate zooplankton density data collected during five years at an electrical generation station on Lake Michigan.
To discuss the example and the field design nec essary for a valid statistical analysis, we provide considerable bac kground on the questions of selecting 1) sampling station pairs, 2) experimentwise 4
error rates for multi species analyses. 3) levels of Tupe I and II error rates, 4) procedures for conducting the field monitoring program, and 5) a discussion of the consequences of violating statistical assumptions.
We include details for estimating sample sites necessary to detect changes of a specified magnitude.
Both statistical and biological problems with monitoring programs (as now conducted) are addressed; serial correlation of successive observations in the time series obtained was identified as one principal statistical difficulty.
Our procedure reduces this problem to a level where statistical methods can be used confidently.
NURE0/CR-2907 VO2: RADIOACTIVE MATERI ALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1981. TICHLER,J.s DENKOVITZ,C.
Brookhaven National Laboratory.
June 1984.
213pp.
8407170576.
BNL-NUREG-51581.
25631:033.
Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1981 have been compiled and reported.
Data on solid waste shipments as well as 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 1981 release data are compared with previous years' releases in tabular form.
Data covering specific radionuclides are summarized.
NURE0/CR-2921: CHEMICAL INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS. SALLACH,R.A.s GREENHOLT, C. J. s TAIG.A.R.
Sandia Laboratories.
April 1984.
70pp.
8405220180.
SAND 82-114 5.
24602:115.
The reaction of tellurium vapor with 304 stainless steel and Inconel-600 alloys in an as-received state and in a preonidized state was studied for the temperature range 500C to 800C.
Most reaction products were identified.
The reaction is fast and appears largely limited by tellurium transport through the surrounding gas phase.
Also studied are the reactions of tellurium vapor with silver Zirceloy-2.
Tellurium desorption rates from solid solutions of tellurium in nickel and 304 stainless steel were measured.
The 20
FLATDEP asdal for calculating tollurium deposition profilos is presented.
NUREC/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN BWRS-COMPUTER MODELING REGUIREMENTS. GREENE,S.R.
Oak Ridge National Laboratory.
April 1984.
237pp.
8405220029.
ORNL/TM-8517.
24557:017.
This report documents the results of an assessment performed at Oak Ridge National Laboratory to determine the reactor and containment hardware, systems, and phenomena which must be modeled in realistic boiling water reactor severe accident analysis computer codes.
The scope of the assessment is limited to BWR-4, 5,
and 6 reactors and Mark I,
II,.and III containment systems.
The report presents a concise review.of the subject reactor and containment designs, together with a description of the reactor and containment systems which have the capacity to impact the outcome of severe accidents.
the results of recent BWR probabilistic risk assessments are briefly discussed, and a detailed visualization of a BWR core melt accident is presented.
Recommendations are made regarding the type of phenomena uhich should be modeled and the level of modeling sophistication required form various stages of the core melt accident.
Finally, the current availability of the necessary models is discussed along with the associated model development priorities.
NUREO/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS. SPITZ,H.B.s SIMPSON,J.C.; ALDRIDGE,T.L.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May 1984.
50pp.
8405310117.
PNL-4550.
24736:241.
Uranium urinalysis and in vivo examination results obtained from workers at eleven uranium mills between 1978 and 1980 were evaluated by Pacific Northwest Laboratory at the request of the U.S.
Nuclear Regulatory Commission.
The main purpose of this evaluation was to determine the degree of the mills' compliance with bioassay monitoring recommendations given in the draft NRC Regulatory Guide 8. 22.
The effect of anticipated changes in the draft guidance, as expressed to PNL in May 1982, was also studied.
Statistical analyses of the data showed that the bioassay results did not reliably meet the limited performance criteria given in the draf t regulatory guide.
Furthermore, quality control measurements of uranium in urine indicated that detection limits a t alpha =be ta= 0. 05 rang ed from 13 miligrams/ to 29 miligrams/, whereas the draft regulatory guidance suggests 5 miligrams/ as the detection limit.
Recommendations for monitoring frequencies given in the draft guide were not followed consistently from mill to mill.
The results of these statistical analyses indicate a need to include performance criteria for accuracy,
~
precision, and confidence in revisions of the draf t regulatory guide.
Revised guidance should also emphasize the need for each mill to i
continually test the laboratory performing urinalysis by submitting quality control samples to insure that the performance criteria are being met.
NUREC/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE DED.
, TARBELL, W. W. s BLOSE,R.E.s ARELLANO,F.E.
Sandia Laboratories.
April 1984.
80pp.
8405220033.
SAND 82-2475.
24552:280.
The two particle bed tests employed 10-kg thermite melts (2700 degree K) teemed into a bed of iron oxide particles.
The objective of 21
- -. ~
the expericants was to :investigoto-bed s ponotration,. partic lo -
L Efloatation and fracturei_and_ heat' flux partitioning..'The results show-that the hydraulic forces exerted by the melt did not immediately displace the bed.
Bed. penetration was by melting and~ absorbing of the
-particles'with the, major portion of the displaced iron' oxide
. terminating in.the alumina phase of'the melt.
'The. movement of the
' penetration front: suggests the movement to be'a-series-of melt / freeze / remelt processes.
The large grain. structure of the iron phase indicates that the cooling.was slow and continuous.
A coherent-cm-thick. layer'ofKiron oxide in contact with the melt.was created bg sintering of the particles. -The particle-size of the unaffected portions 1of'the bed showed very'little. fracturing due'to thermal i
stress and slightly over 7% particle' growth due to sintering.
The calculated heat flux values to :the surrounding crucible structure suggest that:the bed is effective in delaying and reducing the-magnitude of the peak heat flux' values, i
NOREO/CR-3054: CLOSEQUT OF IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING WATER -TO ' SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM)-AND 4 -
MYTILUS SP.~(MUSSEL). R A I NS, 'd. H. s FOLEY,W.J.s HENNICK,A.
Parameter.
Inc.
June 1984.
-59pp.
8406270113.
25173:282.
On April 10, 1981, the-Office.of Inspection and Enforcement (IE) of the U.S.
Nuclear Regulatory Commission (NRC) issued Bulletin 81-03 j
requiring all nuclear generating unit licensees to' assess the potential ~for biofouling of safety-related system components as a j
result of Asiatic clams (Corbicula sp.) and marine: mussels (Mytilus sp.).
Issuance of the Bulletin was prompted by the shutdown of g
Arkansas Nuclear One, Unit 2 on September 3, 1980, as a result of flow blockage of. safety systems by Asiatic class.
Licensee responses to Bulletin 81-03 have been compiled and evaluated to determine the i
i magnitude of existing biofouling problems'and potential for future j.
problems.
An assessment of the real extent of Asiatic clam and marine mussel infestation has been made along with an evaluation of detection and control-procedures currently.in use by licensees.
Recommendations are provided with regard to adequacy of. detection, inspection and t
prevention practices currently in use, biocidal treatment programs, i
and additional areas of concern.
Safety implications and licensee responsibilities are. discussed.
Of 79 facilities licensed to operate, 3-17 have reported biofouling problems, 21 are judged to have high-biofouling potential, 17 are judged to have low or future potential, and 24 are Judged to have little or no potential.
For 49 facilities under construction,.the number of units for matching conditions of biofouling'are 3, 25, 15, and 6 in the same decreasing order of r
severity.
The Bulletin has been closed out for 85 of 129 current facilities. -Followup needed to close out the Bulletin for 21 operating facilities and 23 facilities under construction is proposed in Appendix C.
NUREQ/CR-3134: A SETS USER 'S MANUAL FOR VITAL AREA ANALYSIS.
STACK,D.W.s HILL.M.S.
Sandia' Laboratories.
June 1984.
108pp.
8407170560.
SAND 83-OO74.
25634:037.
This manual describes the use of the Set Equation Transformation System (SETS) f or vital area ana!'ysis.
Various techniques are j
presented for using SETS to solve vital area-analysis fault trees.
Depending on the input to SETS, the solution to the vital area analysis fault tree can be in terms of vital areas or primary events of the vital area analysis fault tree.
The techniques presented are also suitable and efficient for other kinds of common cause analysis.
22
NUTED/CR-32OO VOS: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
DODD.C.V.s DEEDS,W.E.; SMITH,J.H.; et al.
Oak Ridge National Laboratory.
May 1984.
18pp.
8406210100.
ORNL/TM-8796/V4.
25114:298.
Eddy-current inspection is the most suitable method for rapid boresid.e evaluation of steam generator tubing.
However, small flaws can be masked by the effects of harmless variables, such as tube supports.
To identify the critical properties accurately and reliably
)
in the presence of extraneous signals caused by variations of unimportant properties, sufficient information is needed to identify harmful variations and reject harmless ones.
For this rea son we have been developing instrumentation capable of meaturing both the amplitude and phase of the eddy-current signal at several different frequencies, as well as computer equipment capable of proc essing the data quickly and reliably.
Our probes and test conditions are also computer-optimized.
The most recent' probe design embodies an array of small flat " pancake" coils and improves the detection of small flaws and the rejection of tube support signals.
We have also esperimenta11y verified the accuracy of our computer programs for calculating the signals produced by defects in tubing and are adapting our new IBM System 9000 computer to take and process the larger amounts of data required by additional variables, such as copper coating and intergranular attack.
NUREG/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983. ROBERDS,W.; KLEPPE,J.s GONANO,L.
Golder Associates.
April 1984.
469pp.
8405220037.
813-1166.
24652:291.
This report includes the identification and subjective evaluation of alternative schemes for backfilling around waste packages and within emplacement rooms.
The aspects of backfilling specifically considered in this study include construction and testings costs have not been considered.
Houever, because construction and testing are simply implementation and verification of design, a design basis for backfill is required.
A generic basis has been developed for this study by first identifying qualitative performance objectives for backfill and then weighting each with respect to its potential influence on achieving th e 'r ep ository system perf ormanc e objectives.
Alternative backfill materials and additives have been identified and evaluated with respect to the perceived extent to which each combination can be expected to achieve the backfill design basis.
Several distinctly different combinations of materials and additives uhich are perceived to have the highest potential for achieving the backfill design basis have been selected for further study.
These combinations include zeolite /clinoptilolite, bentonite, muck and muck nized uith bentonite.
Feasible alternative construction and testing procedures for each selected combination have been discussed.
Recommendations have been made regarding appropriate backfill scheme for hard rock (i.e.,
domal salt on the Gulf Coast and generic bedded salt).
NUREO/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B & A533-B Reference Plates From PSF Simulated Surveillance & Through-Wall Irradiation Capsules. HAWTHORNE, J. R. s MENKE,B.H.s HISER,A.L.s et al.
Materials Engineering Associates, f
23
'i 7
Inc; April-19 4.
104pp.
-8405220006.
MEA-2017.. 24560:208.
'The NRC 's Ligh t Water Reactor-Pressure Vessel, Surveillanc e Dosimetry Program has irradiated Charpy-V (C(v), compact tension (CT) and tension test spe-imens of selected steels at 288 degrees centigrade.in a pressure vessel wall / thermal shield mock-up known as
'the. Pool Side Facility.
Objectives include the study of through-wall j
-toughness gradients produced by irradiation, the relative irradiation effect at surveillance capsule vs.
in-wall locations and the correspondence of C(v) vs. CT ' frac ture toughness test methods in their independentfdescriptions of radiation-induced embrittlement.
This report presents properties data developed for two steels:
the ASTM A302-B reference plate and the HSST Program A533-B Plate 03.
Irradiation at the simulated surveillance location reproduced reasonably'well the irradiation degradation developed at the vessel inner surface nd quarter wall. thickness locations.
The
_ radiation-induced toughness-gradient was-smalls the difference between j
transition temperatures at-the inner surface vs. mil-wall locations was 31 degrees centigrade or less, andspendent of the test method.
2 The temperature, elevation of the C(v) curve (41 J 1evel)
- ith irradiation was generally less than that defined by fracture toughness tests (100 MPa square root of m level) but greater than defined by
- Beta (Ic)-corrected" data.
NUREC/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL ~ SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
Tensile. Strength Determinations For PSF Simulated Surveillance &-
Through-Wall Specimen Capsules. HAWTHORNE,J.R.s MENKE,B.H.
Materials Engineering Associates, Inc.
- ENSA, Inc.
April 1984.
133pp.
8405220025.
MEA-2017.
24561:011.
The NRC's Light Water Reactor-Pressure Vessel Surveillance Dosimetry Improvement Program has irradiated Charpy-V (C(v) and tension test =pacimans of selected steels at 288 degrees centigrade-in a pressure vessel wall / thermal shield mock-up known as the Pool Side Facility.
Objectives include the study of through-wall toughness gradients produced by neutron irradiation and the relative irradiation effect at surveillance capsule vs.
in-wall locations.
This report presents properties data developed for six steels: the ASTM A 302-B reference plate, the HSST Program A 533-B Plate 03, 508-3 and 22NiMo-Cr37 forgings, and two submerged are weld deposits.
The radiation-induced toughness gradient between inner surface vs. mid-wall locations uas small (31 degrees centigrade or less) for five of the six indications.
Simulated surveillance capsule irradiations reproduced well the embrittlement observed for vessel inner surface and quarter wall thickness locations in almost all cases.
The primary excep tions to both trends were provided by a 0.23%
C u, 1.58% Ni weld deposit which showed the highest embrittlement sensitivity.
Material irradiation sensitivity levels are in accord with predictions based on copper and nickel contents.
NUREC/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC SAFETY STUDY: PLANT ANALYSIS. BERRY, D. L. s BRISBIN,N.L.s CARLSON,D.D.s et al.
Sandia Laboratories.
May 1984.
479pp.
8406070143.
SAND 83-1118.
24859:165.
This report describes the review of the internal and external event plant analyses of the Zion Probabilistic Safety Study (ZPSS).
The review was conducted by Sandia National Laboratories.
The purpose of the review was to search for areas in the ZPSS where omissions and 24
Lcritical Judga:nte scro mcdo which could impcct the quantitativo results.
The review identified several of these areas.
i NUREQ/CR -3303: USE OF NEUTRON NOISE FOR DIAGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
FRY,D.N.s. MARCH-LEUBA,J.s L
SWEENEY,F.J.
Oak Ridge National Laboratory.
May 1984.
100pp.
8405290438.
ORNL/TM-8774.
24696:256.
The value of neutron noise analysis for diagnosis of in-vessel anomalies in light-water reactors (LWRs ) was assessed by:
(1) analyzing ex-core neutron noise from seven pressurized-water reactors l:
. (PWRs) to determine the degree of similarity in the noise signatures and the sources of ex-core neutron noises (2) measuring changes in ex-core neutron noise over an entire fuel cycle at a commercial PWRs (3) applying PWR neutron noise analysis to diagnose a loose core barrel, to infer in-core coolant velocity, and to infer fuel assembly
- motions and (4) applying BWR ne.utron noise analysis to diagnose in-core instrument tube vibrations and bypass coolant boiling,'to' i
infer in-core two phase flow velocity and void fraction, and to infer stability associated with reactivity feedback.
This report summarizes these assessments and provides guidance for the acquisition and analysis of neutron noise in LWRs.
NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES. BURKETT.M.W.s IDAR,E.S.s CIDO,R.G.s et al.
Los Alamos Scientific Laboratory.
April 1984.
54pp.
8405220096.
i LA-9776-MS.
24594:313.
In this study, a more advanced "best-estimate" containment code, BEACON-MOD 3A, was used to calculate force and moment lead: re:ulking from a high-energy blowdown for two reactor cavity geometries previously analyzed with the licensing computer code COMPARE-MOD 1A.
The BEACON force and moment loads were compared with the COMPARE results to determine the saf ety margins provided by the COMPARE code.
The forces and moments calculated by the codes were found to be different, although not in any consistent manner, for the two reactor cavity geometries studied.
Therefore, generic summary statements regarding margins cannot be made because of the effects of the detailed physical configuration.
However, differences in the BEACON and COMPARE calculated forces and moments can be attributed to differences in the modeling assumptions used in the codes and the analyses.
NUREQ/CR-3307 V03: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report July-September 1983. EDLER,S.K.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984.
72pp.
8405220055.
PNL-4705-3.
24561:298.
l This document summarizes work performed by Pacific Northwest j
j Laboratory from July 1 through September 30, 1983, for the Division of
)
Accident Evaluation and the Division of Engineering Technology, U. S.
Nuclear Regulatory Commission.
Evaluations of nondestructive examination (NDE) techniques.and instrumentation include demonstrating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, and examining NDE reliability and probabilistic fracture mechanics.
Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of 4
fuel rod failure under normal operating conditions.
Experimental data
-and analgtical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in A
-+e ee
+i--ra e
-w e.--n*wwe-.e,,yr e
e 9
1w.
.y=.s.a.www m
..g,9-<
97yyeyigg mp.e.
99 py m q wygve e 4.e,_
gp pgeene egywr y, s i.m p W--
p o
-high-Onorgy fluid systcm piping.
Exporiaantal data validated sodols.
are.being used to determine a method for evaluating the acceptance of welded or weld-repaired. stainless steel piping.
Thermal-hydraulic models are.being developed to provide better digital codes. to compute the behavior of-full scale reactor-systems under postulated accident conditions.
High-temperature materials property tests are being conducted to provide data on severe core damage f uel' behavior.'
Severe 4
fuel damage accident tests are being conducted at the NRU reactor,
~
Chalk. River,' Canadas and an instrumented-fuel assembly irradiation program is being performed at Halden, Norway.
Fuel assemblies and analytical support are.being provided for experimental programs at,
{
other facilities, including the Super Sara Test Program, Ispra, Italy, and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho' Falls. Idaho.
NUREG/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report October-December 1983. EDLER S.K.
Battelle Memorial Institute, 3
l Pacific Northwest Laboratories.
May 1984.
38pp.
8406060432.
PNL-4705-4.
24668:347.
.This document summarizes work performed by Pacific Northwest
- Laboratory from October 1 through December 31,~1983, for the Division of Accident Evaluation and the Division of Engineering Technology,
. U. S.
Nuclear Regulatory Commission.
Evaluations of nondes tructive
' examination (NDE) techniques and instrumentation include i nvestigating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems and examining NDE reliability and probabilistic fracture mechanics.
Accelerated pellet-cladding l
interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions.
Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in l
]
high-energy fluid system piping.
Experimental data and validated models are being used to determine a methods for evaluating the j
acceptance of welded or weld-repaired stainless steel piping.
Thermal-hydraulic models are being developed to provide better digital r
codes to compute the behavior of fu11 scale. reactor systems under postulated accident conditions.
High-temperature materials property j
tests are being conducted to provide data on severe core damage fuel j
behavior.
Severe fuel damage accident tests are being conducted at j
the NRU reactor, Chalk River, Canadas an instrumented fuel assembly l
irradiation program is being performed at Halden, Norways and fuel assemblies and analytical support are being provided for esperimental I
programs at the Power Burst' Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho.
i NUREO/CR-3310: TESTING OF THE CONTAIN CODE. SCIACCA,F.W.s BEROERON,K.D.s MURATA,K.K.s et al.
Sandia Laboratories.
April 1994.
1 2OOp p.
8407020036.
SAND 83-1149, 25230:001.
-CONTAIN is a large computer code intended for use in the analysis of severe nuclear power plant accidents.
Many tests have been i
conducted on CONTAIN to assess its adequacy for dealing with j
. nuclear-accident problems.
This report describes the CONTAIN test program and summarises the results obtained to date.
These results l
are presented'so that users may be aware of the f eatures o f CONTAIN i
that have been checked and of the areas where problems have been j
identified.
In addition, this report provides information needed by users to repeat tests of interest in their specific work areas.
(
The test efforts have identified a substantial number of problems i
28 i
,.m
.--.e,-mwyr m,,m~.-m,-m.
r.
-e,,,,,,._.,____..,m-_m%w,~.
-,w w
.m, - -,
in the ccding or logic of the CONTAIN cedo.
Mact of thoco prob 1cma have been corrected.
These corrections have been included in the most recent versions of the code.
CONTAIN can accurately treat most of the phenomena expected to occur in containment atmospheres.
Some problems identified by the test program, involving pool-related phenomena, have prompted the development of a substantially new system of models for pool phenomena.
When completed, this new system will be sub Jected to intense testing of the type described here.
NUREO/CR-3316: VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TRANSPORT MODEL (SWIFT). WARDS, D, S. s REEVES.M.s DVDA,L.E.
Sandia Laboratories.
April 1984.
170pp.
8407060054.
SAND 83-1154.
25432:120.
The SWIFT Model has been developed and maintained by Sandia National Laboratories.
The Nuclear Regulatory Commission has sponsored this work under the high-level nuclear waste program.
SWIFT is a fully-coupled, transient, three-dimensional model.
It is implemented by a finite-difference code which solves the equations for flow and transport in geologic media and is used to evaluate repository-site performance.
This document represents an important part of the quality-assurance records for the code.
Here the process simulators for flow, heat and radionuclide transport are eramined using two different types of tests.
The analytical, verifications test SWIFT calculations against analytical solutions, and the field comparisons test SWIFT calculations against field data.
Both types of tests yield good agreement between the SWIFT computations and the comparative data.
NURE0/CR-3329 V04: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROORAM.Guarterly Report October-December 1983. THOMPSON, S. L.
Sandia Laboratories.
April 1984.
65pp.
8405220044.
SAND 83-1171.
24556:275.
The TRAC-PF1/ MOD 1 independent assessment program at Sandia National Laboratories (SNLA) is part of a multi-faceted effort sponsored by the Nuclear Regulatory Commission (NRC) to determine the ability of various system codes to predict the detailed thermal / hydraulic response of LWRs during accident and off-normal conditions.
This program is a successor to the RELAP5/ MOD 1 independent assessment project underway at Sandia for the last two years.
NURE0/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-3. OSBORNE.M.F.; LORENZ,R.A.s NORWOOD,K.S.s et al.
Oak Ridge National Laboratory.
May 1984.
69pp.
8405290450.
ORNL/TM-8793.
24711:141.
The third in a series of high-temperature fission product release tests was conducted for 20 min at 2000 degrees centigrade in flowing steam.
The test specimen, a 20-cm-long section of H.B.
Robinson fuel rod that had been irradiated to *25,200 mwd /t, was heated in an induction furnace in a hot cell.
Posttest examination showed that the Zircaloy cladding had melted, causing extensive disintegration of hte UO(2) fuel and formation of molten phases that appeared to be rich in uranium.
Analyses of test components revealed very high fractional releases of (85)Kr (59.0%), (137)Cs (58.8%), and (129)I (35.4%).
The releases if (125)Sb and (110m) Age however, were much less than those observed in test HI-2 at 1700 degrees centigrade, perhaps as a result of lower 27 1
s.
~
Jstcam fics roto in' tost HI-3.
The extont of.corosol f ormation,' as 7
- evidenced by masstof material. collected on filters,.~was:similar in the l
.two. tests.
1
'NUREO/CR-3350: 'LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL
~
JREACTOR. PROGRAM: Postirradiation Examination Results For The Third-Materials Experiment'(MT-3).:RAUSCH,W.N..Battelle Memorial Institute,. Pacific Northwest Laboratories.
April 1984.-
71pp.
8404300179.
PNL-4933J 24233:238.
A seriesyof;in-reactor; experiments were conducted by Pacific:
.. Northwest-Laboratory,'using full-length 32-rod pressurized water-reactor fuel bundles,' as'part of:the Loss-of-Coolant Accident (LOCA)
Simulation ~ Program.
-The' third materials-experiment (MT-3) was the
' sixth in the'seriesaof thermal-hydraulicLand materials
' deformation / rupture experiments' conducted in the National Research
-Universali(NRU) reactor, Chalk River, Ontario, Canada.
MT-3 was
' Jointly-funded by the U.S.
Nuclear Regulatory Commission and the United Kingdom' Atomic Energy Authority.
The experiment evaluated
. ballooning and rupture-during active.two-phase cooling in the
. temperature range from 1400 to 1500 fahrenheit.
The 12 test rods'in the center of.the 32-rod bundle were initially pressurized.to 550 psi to insure rupture in the correct temperature range.
All 12 of the-rods ruptured, with an average peak ' bundle strain of about 55%.
A hot cell postirradiation examination (PIE) of several of the ruptured rods was also conducted.
This report describes the work performed and
. presents,the PIE results.
Information obtained during-the PIE-s analysis included' cladding. thickness measurements, metallography, and particle size analysis of the cracked and' broken fuel pellets.
NUREC/CR-3360: COMPUTER PROGRAM CDCID: AN AUTOMATED QUALITY CONTROL PROGRAM USING CDC UPDATE. SINGER, C. L. J AGUILAR,F.
EG&G, Inc.
April 1984.
70pp.
8405220028.
EGG-2302.
24557:253.
A' computer program, CDCID, has been developed in coordination uith a quality control program to provide a highly automated method of documenting changes tc computer-programs at EG&G Idaho, Inc.
The method uses the standard CDC UPDATE program in'such a manner that updates and their associated documentation are easily made and retrieved-in various formats.
The method allows each card image of a source program to point to the document which describes it, who created.the card, and when it was created.
The method described is applicable to the quality control of computer programs in general.
The computer program described is esecutable only on.CDC computing systems, but the program could be modified and applied to any computing system with an adequate updating
- program, NUREO/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANIA-CDATED STEEL. POWERS,D.A.: ARELLANO,F.E.
Sandia Laboratories.
April 1984.
95pp.
8406230297.
SAND 83-1350.
25128:155.
Corium and Thermitic melts were teemed at various velocities onto bare steel plates and steel plates coated with urania.
An empirical correlation of the' penetration data is developed.
NUREO/CR-3378: VERIFICATION OF THE NETWORK FLOW AND 1 TRANSPORT / DISTRIBUTED VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
'DUDA,L.E.
LSandia Laboratories.
May 1984.
50pp.
8406190081.
28 m
~
CAND83-1466.
25029: 1C3.-
.The Network Flow and. Transport / Distributed Velocity Method (NWFT/DVM)^ computer. code was developed to. provide a computationally efficient ground-water-flow and contaminant transport capability for use'in risk analyses.
It is a semi-analytic, quasi-two-dimensional network' code that. simulates ground water flow and.the transport of
^
dissolved species.(radionuclides) in saturated porous medium.
This
_ code development was funded by'the U.S.
Nuclear Regulatory Commission as part of a methodology for assessing the risk from disposal of radioactive wastes in geologic formations.
A, separate project was funded to ensure that the codes developed are as error-free as
.possible:and include. verification and validation tests to represent the processes for uhich it is intended.
This document contains four
' verification problems for the NWFT/DVM computer code.
Two oP these l-problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions.
The other two are code-to-code verifications where results are compared to those of another computer
?
code.
The.NWFT/DVM results showed good agreement with both the analytical solutions and the results from the other code.
NUREG/CR-3379: SLAM - A SODIUM-LIMESTONE CONCRETE ABLATION MODEL.
SUO-ANTTILA A.
Sandia Laboratories.
April 1984.
77pp.
8405220176.
SAND 83-7114.
24601:082.
The Sodium-Limestone Ablation Model (SLAM) is described-in detail in this report.
SLAM is a three-region models containing a pool (sodium and reaction debris) region, a dry (boundary layer and
,t dehydrated concrete) region, and a wet (hydrated concrete) region.
The model includes a solution to the mass, momentum, and energy equations in each region.
A chemical kinetics model is included to i
provide heat sources due to chemical reactions between the sodium and the concrete.
Both isolated model as well as integrated "whole code"
=
l evaluations have been made with good results.
The chemical kinetics and water. migration models were evaluated separately, with good results.
Several small and large-scale sodium limestone concrete experiments were simulated with reasonable agreement between SLAM and i
the experimental results.
}
The SLAM code was applied to investigate the effects of mixing, 4
pool temperature, pool depth and fluidization.
All these phenomena were found to be of significance in the predicted response of the sodium concrete interaction.
Pool fluidization is predicted to be the j
nost important variable in large scale interactions.
l NUREG/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA. SWYLER K.J.s DODOE.C.J.s DAYAL,R.
Brookhaven National Laboratory.
April 1984.
{
86pp.
8405220086.
BNL-NUREG-51691.
24602:288.
The effects of external irradiation on anion, cation, and mixed 1
1 bed organic'lon exchangers have been investigated under conditions relevant to.radwaste storage and disposal.
Two effects are j
emphasized: (1) radiolytically induced release of acids, radionuclides or chemically aggressive species, and (2) radiolytic generation / uptake i
of corrosive or combustible gases.
For sulfonic acid cation resin, j
sulfate ion is produced in the radiolytic scission of the functional group.
The insensitivity"to external parameters may make the sulfate d
yield a convenient measure of radiation durability for regulatory considerations.
The acidity'which results from a given sulfate yield j
depends on the resin loading.
Acidity is substantially reduced for i
i 29 i
---r
,--..w..w-ym -m e,.
w
,,y,,.
yy,,-.(my,e__,w,,w,,w.~.
rm mp
,qw.,f-w-myg,w_,,-,_,,y..,.
l 1ccdings other.then H(+).
"or hoovy irradiction decon incorpsroting
. cation / anion resins in mi,sd bed' form, the presence of anions resin did not protect against radiolytic acidity formation.
The irradiated anion resin may also release substantial amounts of free liquid.
Radiolytic hydrogen gas yield data support the validity of accelerated
[
testing at high radiation dose rates.
Oxygen gas is removed from the environment of irradiated resins by an efficient radiolytic oxidation-process.
l NUREG/CR-3391 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report, April 1983 - June 1983.
LIPPINCOTT, E. P. s MCELROY,W.M.
Hanf ord Engineering Development Laboratory.
April 1984.
113p p.-
8404170026.
HEDL-TME 83-22.
24092:068.
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 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 (Mo l, Belgium),
EPRI (Palo Alto,-USA), KrA (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.
NUREG/CR-3391 V03: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGR AM. Annua l R ep or t,0c tob er 1,1982-Sep temb er 30,1983.
MCELROY,W.M.; KAM,F.B.s GRUNDL J.A.s et al.
Hanford Engineering Development Laboratory.
June 1984.
198pp.
8407180011.
HEDL-TME 83-23.
25652:102.
See NUREC/CR-3391,VO2 abstract.
NUREG/CR-3391 VO4: LWR PRESSURE VESSEL SURVE!LLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1983-December 1983. LIPPINCOTT E. P. s MCELROY, W. M.
Hanford Engineering Development Laboratory.
May 1984.
92pp.
8406080261.
24877:001.
See NUREG/CR-3391,VO2 abstract.
NUREG/CR-3410: CHMONE: A CNE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLCW AND CHEMICAL CONCENTRATIONS IN WATER BODIES, FISCHER,S.K.s HETRICK,D.M.s LIETZKE,M.H.s et al.
Oak Ridge National Laboratory.
April 1984.
257pp.
8404180403.
ORNL/TM-8786.
24107:098.
The computer code CHMONE simulates fast-transient, one dimensional hydrodynamic, t h erma l, and chemical-species-concentration conditions in controlled rivers and tidal estuaries.
The code is particularly designed for applications to actual site-specific problems that require accurate predictions of the chemical species concentrations for preliminary studies of the aggregate chemical ao
i.
impcct ente camesn watorbody ccused by chlorinaticn of the dischargo I
uater'from multiple'powertplant operations.
4
-The'CHMONE code can continuously simulate the hydrodynamic and
. thermal: conditions and concentrations of four chemical species for a 30-d period.
Because only'a sma11' amount of CPU time is necessary, CHMONE can.be readily utilized as a cost-effective tool in studying 4
l thermal and chemical' impacts of power plant discharges in controlled rivers and tidal estuaries.
.'NURE0/CR-3422 VO3: AEROSCL RELEASE AND TRANSPORT PROGRAM.Guarterly j,.
Progress Report For July-September 1983. ADAMS,R.E s TOBIAS,M.L.
Oak Ridge National Laboratory.
April 1984.
53pp.
8405290448.
ORNL/TM-8849/V3.
24711:090.
l This report summarizes progress for the Aerosol Release and Transport Program sponsored by the Nuclear Regulatory - Commission's Office of' Nuclear Regulatory Research, Division of Accident Evaluation, for July-September 1983.
Topics discussed include'(1) several capacitor discharge vaporization (CDV) experiments in the Fuel Aerosol' Simulant Test Facilitys (2) descriptions of mixed-aerosol experiments 611 and=612, which involved iron oxide and ' uranium oxide
]
in steams-(3) technical support work for the aerosol test program at' 1
Marviken, Swedens (4) core-meltJesperiment CM-35, in which tellurium and its oxide were used as additivess (5) progress in construction of 1
a 10-kg core-melt induction furnaces (6) finite-difference f
calculations.of energy deposition in CDV specimens; (7) a steam-only experiment in the NSPPs (8) code implementation activitiess and (9)
{
NAUA code validation studies.
l i
NURE0/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATCRI ALS USED FOR j
HIGH-LEVEL WASTE PACKAGING. Annual Report. April 1983 - April 1984.
j STAHL.D.s MILLER,N.E.
Battelle Memorial Institute, Columbus Laboratories.
June 1984.
282pp.
8407180206.
BMI-2113.
25665:001.
]
The effects on glass waste-form dissolution of' temperature, i
pressure, solution chemistry, and ratio of glass surface area to solution volume have been studied.
The glass-dissolution correlation is ready to be evaluated-by comparison with experiments.
The j
devitrification correlation has been completed.
In canister-corrosion j
studies, CFB alloy was found less susceptiele to glass attack than t
Type 304L stainless steel.
Limited experiments revealed no corrosion mechanism which would indicate that cast steel could not be used as a l
container materials additional tests with cracking agents are planned.
I In hydrogen-uptake studies, cast steel was found to absorb more I
hydrogen than wrought steel.
Parts of the general-corrosion correlation have been tested, and work continues on obtaining l
realistic experimental data as input for i t.
Gamma fluxes and dose i
rates in and near the waste package were calculated for CHLW and spent-fuel waste forms.
The current water-radiolysis model was found adequate when tested against existing data, and preliminary I
calculations'were perforned with the current water-chemistry models in j
both cases, additional chemical species are being incorporated.
4 i
l NUREQ/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF COPPER. HARRISON,F.L Lawrence Livermore National Laboratory.
April 1984.
62pp.
i 8406230213.
UCRL-53486.
25129:315.
This report provides a summary of research performed to determine
}
the physicochemical forms and fate of copper in effluents from power i
l l
31 I
.w
--- - - - -. - -. -. -. _. -. ~ -... -
f ototiano cdJccent uto oguatic ocongstems with wStor. that dif fore in
" salinity, p H, and concentrations of organic and inorganic
- constituents.
.In addition, research performed to evaluate responses t
of; selected, ecologically and economically important marine and i-freshwater organisms to increased concentrations of soluble copper is reviewed.
speciation showed that the quantities of
' Copper concentration and
~ copper associated with particles, colloids, and organic and inorganic.
ligands differed with the site, season and mode of operation of the station.
Under normal operating conditions, the differences between-
-influent and. effluent waters-were generally small, and most of the copper was in bound (complexed) species except when low pH water was
}
circulated..However, copper was high in concentration and present in j.
labile species during start-up of water circulation through some i
cooling systems and during changeover from open-cycle to closed-cycle I
operation.
i The toxic response to copper differed.with the species and life stage.of'the organism and with the chemical form of copper in the i
I water.
Our primary emphasis was on acute effects.
However.. sublethal
}
effects of copper on a population of bluegills living in a power j.
station cooling lake containing water of low pH and on a population.
j exposed to increased soluble copper in the laboratory were also assessed.
~
I NUREG/CR-3488 VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
l START. E. E. s CATE J.H.4 DICKSDN. C. R. s et al.
Commerce,. Dept. of, f
Natl. Oceanographic & Atmospheric Administration.
April 1984.
i 944pp.
8405220082.
24549:001.
+
i The 1991 Idaho Field Experiment was conducted in South East Idaho l
over the Upper. Snake River Plain.
Nine test-day case studies were measured between July 15 and 30, 1981.
Eight-hour releases of SF(6)
. gaseous tracer were made from 46 m above ground.
Tracer was sampled j
hourly, for 12 sequential hours at about 100 locations within an area i
24 km sguare.
Also, a single total integrated sample of about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> duration was collected at approximately 100-sites within an area j
48 by 72 km (using 6 kn spacings).
Extensive tower profiles of-
[
noteorology at the release point were collected.
RAWINSONDEE, RABALS l
and PIBALS were collected at 3 to 5 sites.
Ho r i r o.ita l, low-altitude
-l Winds were monitored using the INEL MESONET.
SF(6) tracer plumes were marked with co-located oil fog releases and bi-hourly sequential j
launches of tetroon pairs.
Aerial LIDAR observations of the oil fog.
j plume airbornw samples of SF(6) were collected.
High-altitude aerial 1-photographs of daytime plumes were also collected.
he Idaho Field i
Experiment is reported in three volumes, Volume II lists the data in tabular form or cites the special supplemental reports by other i
participating contractors.
While the primary user file and the data j
achieve-are maintained on 9 track /1600 cpi' magnetic tapes, listings of the individual values are provided for the user who either cannot utilize the' tapes or-wishes to preview the data.
The accuracies and quality of these data are described.
NUREQ/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEDLOGIC DISPOSAL DF NUCLEAR WASTE. KENDORSKI,F.S.s HAMBLEY,D.F.; WILKEY,P.L.
j Engineers International, Inc.
May 1984.
656pp.
8406210455, i
EI-1077.
25095:001.
l Currently, the most feasible alternative for permanent disposal i
of high level nuclear waste is storage in deep underground l
repositories in geologic media.
Uncertainties in investigation, I
i 32 l
V
-._~..--.._m..,_
m I
d30ign cnd canotruction nococcitato maintaining' tho retriovel optian until the isolation is proven likely.
Investigations were limited to concepts in geologic media' currently being investigated by DOE.
Retrieval in most concepts is not a simple reversal of waste emplacement.
This study identified several concerns.'
Technological concerns are associated with remining and monitoring radioactivity in backfilled storage rooms and retrieval of breached canisters.
Retrieval systems currently incorporated into DOE designs were found inadeguate for handling. breached canisters or those bound in the-l storage holes.-
Short holes containing single canisters could be
{
overcored but equipment must be developed to overcore large diameter holes..
Safety concerns common to all repository concepts are protection of personnel from heat, traffic congestion, and 4
j deterioration of ground support.
Concerns on radionuclide release i
were the radiation and radionuclides which would be released into the air and water.present in a storage room if there were a canister 4
i breach.
.The confinement ventilation circuit air-flows provided in the i
l'.
DOE conceptual designs are Just adeguate for retrieval and are inadeguate for retrieval from backfilled rooms.
NUREO/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
r j
CHER.F.F.s DOMANUS,H.M.s SHA, W. T. s et al.
Argonne National l
Laboratory.
May 1984.
53pp.
8407110019.
ANL-83-65.-
25546:001.
i The report describes the three additional. turbulence models 1
C 0-cguation (mixing-length), 1-eguation (k), and 2-eguation (k-E)3
}
recently implemented in the COMMIS-15 computer code.
.COMMIS-15 is a j
three-dimensional, steady-state / transient, single-phase computer code for thermal-hydraulic analysis of single / multicomponent systems under i
normal and off-normal operating conditions.
All three turbulence..
i models are provided as options, and a user can select the one that is t
most. appropriate for his or her application.
To validate these turbulence models, we have performed several
~
j numerical simulations and compared the results with experdmental~
j t
data.
Three of the simulations--turbulent flow in a pipe, flow in a j
circular duct with sudden expansion, and' thermal and fluid mixing in the cold leg and downcomer of a PWR--are presented here along with their comparisons with experimental data.
More analyses are needed L
for further validation.
Incorporation of the three turbulence models has expanded the range of application of the COMMIX code;
}
1 l
NUREO/CR-3505: A VOLUME-WE!QHTED SKEW-UPWIND DIFFERENCE SCHEME IN CDMMIX. MIAO,C.C.J LYCZK0WSK!, R. W s LEAF,G.K.s et al. -Argonne i
National Laboratory.
May 1984.
92pp.
8407180028.
ANL-93-66.
I 25683:140.
l A numerical difference scheme, called volume-weighted skew-upwind difference (VWSUD),-has been developed, and Raithby 's two-dimensional-t skew-upwind difference (SUD) scheme has been extended to three i
dimensions.. Both schemes have been implemented in'the energy eguation of the COMMIX-1B computer program.
The VWSUD scheme has the fo'11owing five major f eatures:
(1) it has the same order of accuracy as SUD, but eliminates all of the undershoots observed in SUDS (2) it retains I
i the' simplicity of SUD, without resorting to the artificial cut-offs l
l needed in SUDS (3) it significantly reduces numerical dif fusions (4) a linear stability analysis shows that VWSUD is numerically stables
{
and (5) a coarser mesh than f or the pure-upwind diff erence scheme can he used while obtaining results that arelof the same order of l
- accuracy, i
The assessment of SUD and VWSUD are accomplished by comparing i
I' l
33 I
t
--,my-,.,--ur-,-,r.-.e
,--wym y
--.m
.w,y
,creic,ypyww w m w ei.ny
,,ge-v,yr,.,-wee-wwwrvom e.e-e.
y m n e,qwww _
eut*w"
sov0rol sultidiohnsional thorcol mixing benchccrk computotians eith analytical solutions.
In addition, the analysis of two thermal mixing experiments shows ~ that use of the VWSUD scheme substantially -improves l
agreement with thermocouple response data in regions with highly angled. flows.
i-NUREO/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF WELDS. HISER,A.L.s LOSS,F.J.a MENKE.B.H.
Materials Engineering' i
Associates, Inc.
April 1984.'
616pp.
8405210598.
MEA-2028.
24531:001.
4 This investigation provides a. data base of J-R curve trends from irradiated A 508 and A 533-8 weld metals exhibiting low upper shelf Charpy-V (C(v) energy.
These welds were made with Linde 80 flum of the same lots used for vessels currently in service.
These materials j
exhibited postirradiation C(v) upper shelf energies of 58 J to 80 J.
Compact toughness (CT) specimens of four different sizes (0.5T-to 4T-CT) were characterited.
These specimens were irradiated to a
{
fluence of *1 x 10 (19) n/cm(2) > 1 MeV as part of the NRC-sponsored HSST program.
4 i
The J-R curves exhibited a power-law behavior for small crack i
extensions'(e.g...< 2 mm).
Irradiation decreased the level of the R-curve significantly in most cases.
The value of J-integral at the initiation of crack growth (J(Ic) decreased on average by *25% at 200 degrees centigrade and by *35% at 288 degrees centigrade.
.The average value of tearing modulus (T(avg) was a more discriminating indication l
of the degradation due to irradiation, as T (avg) decreased'on average l
6y '54% at 200 degrees centigrade and by '69% at 288 degrees centigrade.
A modest site ef fect associated with large sp ecimens was
=
indicted for tests in the unitradiated condition, while no size effect was apparent for tests in the irradiated condition.
4
{
These data compare favorably with corre1tions between C(v) upper i
shelf energy and J-R curve parameters observed from prior studies with j
1T-CT specimens.
These corre1tions could enhance the significance of 1
C(v) reactor surveillance data with respect to structural integrity.
}
NUREC/CR-3507: AN ANALYSIS OF THE NRC SAFETY GOALS FOR NUCLEAR POWER.
i FISCHHOFF.B.
Decision Research, Inc.
- Oak Ridge National Laboratory.
April 1984.
48pp.
8404300067.
ORNL/SUB-7576/2.
24230:196.
The document analyzes the proposed " safety goals" with the general theory of standard setting.
The analysis discusses the
)
concept of " acceptable risk" and the attempt to build policy instruments around it.
NOREQ/CR-3511 VO1: INTERIM RELI ABILITY EVALUATION PRDORAM: ANALYSIS OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume 1. Main Report.
PAYNE.A.C.
Sandia Laboratories.
May 1984.
273pp.
8405220017.
SAND 83-2086.
24552:007.
'This report presents the results of the Probabilistic Risk i
Assessment (PRA) of Calvert Cliffs Unit 1 Nuclear Power Plant.
The analysis was ~ performed as part of the Interim Reliability Evaluation Program (IREP).
The analysis used fault tree and event tree models as the primary tools to evaluate the risk due to a core melt 'at Calvert Cliffs.
Core melt seguences initiated by one of three break-size LOCAs or one of six ~ categories of transients were evaluated, and the dominant (i.e.,
highest frequency) seguences were further analyzed to estimate the magnitude of radionuclide release.
The accident i.
34 l
_.. - _ - _,, _ _ _ _ _ _ _ _ _ _ _. ~. _ _. ~ _ _ _ _
sequences were then placed into the release categories defined in the Reactor Safety Study to estimate this magnitude.
The most significant seguences contributing to the core melt frequency are (1) Anticipated Transients Without Scram (ATWS) (44%'of the total core melt frequency), (2) Small-small LOCAs (i.e.,
3" to 1.9" in diameter) with makeup system failure in the recirculation phase (19% of the total core' melt frequency), and (3) the loss of a DC bus followed by failure of secondary heat removal (14% of the total core melt frequency).
The estimated core melt frequency for Calvert Cliffs Unit 1 (CC-1) is similar to the values predicted by PRAs of other PWRs.
t NUREO/CR-3514: THE CHEMICAL BEHAVIOR OF IODINE IN AGUEOUS SOLUTIONS UP TO 150 C. An Experimental Study of Nonredon Conditions. TOTH.L.M.:
PANNELL,K.D.s KIRKLAND,0.L.
Oak Ridge National Laboratory.
April 1984.
47pp.
8405290434.
ORNL/TM-8664.
24711:001.
The chemical behavior of iodine, 12, in (pH = 6 to 10) aqueous solutions containing 2500 ppm boron as H3B03-(0.231 M) was studied at j
temperatures up to 150C, Absorption spectrophotometry was used to f
identify and monitor the iodine species present.
Three objectives l
were considered:
(1) species identification, with special attention i
g iven to " HOI": (2) the kinetics of reaction between iodine and water i
to produce iodide and todate ions: and (3) partition coefficients-i between liquid and vapor phases for individual iodine species.
4 Kinetic rate constants for the disproportionation of the " HOI" l
intermediate were measured.
A typical activation energy for this reaction was found to be 28.4 kJ/mol (6.8 kcal/mol).
No absorption bands can be assigned to the " HOI" intermediate even though it has been shown, in some cases, to be present at concentrations of >1 x
l 10(-3) M.
A very low molar absorptivity ( <10 M-1 cm-1) is probably responsible for its undetectability.
A partition coefficient of >1 x
i 10(4) has been estimated for " HOI" 1
I NUREO/CR-3513: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR j
DEVELOPING CRITERIA. K0ZINSKY,E.J.s GRAY,L.H.s BEARE,A.N.s et al.
Dak Ridge National Laboratory.
April 1984, 166pp.
8405210559.
4 ORNL/TM-8942.
24527:123.
4 This report presents a methodology for developing criteria for i
design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base.
It is the i
eleventh and final NUREO/CR Neport on the Safety-Related Operator 4
Actions Program, conducted by Oak Ridge National Laboratory for the U. S. Nuclear Regulatory Commission.
The operator performance data l
were developed from training simulator experiments involving operator j
responses to simulated scenarios of plant disturbancess from field
(
data on events with sinitar scenarios 3 and from task analy tic data.
A l
conceptual model was run, using the SAINT modeling language.
Proposed is a quantitative predictive model of operator performance, the
}
" Operator Personnel Performance Simulation (OPPS) Model," driven by i
task requirements, information presentation, and system dynamics.
The I
model outpute a probability distribution of predicted time to l
correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria.
4 f
NUMEO/CR-3533: RADON ATTENUATION HANDBOOK FOR URANIUM-MILL TAILINGS COVER DESIGN. ROGERS,V.C.s NIELSON,K.K.
Rogers & Associates 3
Engineering Corp.
KALKWARF.D.R.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984.
89pp.
8405210555.
l f-35
^
' PNL-4870.'
2CS29: 119.
This handbook has been prepared to facilitate.the. design of
. earthen covers to. control radon-emission'from uranium mill' tailings.
Redon emissions from bare and covered uranium mill tailings can be estimated by equations based on diffusion theory.. Basic equations are presented for~ calculating surface radon fluxes ~from covered-tailings, er alternatively, the cover thickness required.to satisfy a given o.
redon flux criterion. -Procedures are also given-for measuring diffusion coefficients for radon, or for estimating them from empirical correlations.
Since long-term soil moisture content-is.a critical parameter in determining the value of the diffusion i
coefficiente. methods.are given for estimating the -long-term moisture contents of-soils.
The effects of cover defects or advoction are also discussed and guidelines are given for determining if they are significant.
For most practical cases, advection and cover-defect effects on redon flux can be neglected.
Several examples are given.to demonstrate cover design calculations, and an extensive list of references-is included.
i 1
NUREC/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS.FOR SELECTED BDNE-SEEKING RADIONUCLIDES. CRISTY,M.J LEGGETT,R.W.; DUNNINO D..E.s et 1.
al.
Oak Ridge National Laboratory.
May 1984.
79pp.
8405210611, ORNL/TM-8929.
24534:274.
The transuranic elements and the radiostrontiums are bone-seekers and are potentially important contributors to bone dose from releases i
from a breeder reactor such as the Clinch River Breeder Reactor.
Currently available age-specific dose-conversion factors for these nuclides are based on methods of ICRP Publication 2,. published in 1959.
ICRP Publications 25.and 30, published in 1977 and 1979, i
outline methodology. incorporating new models and new concepts of risk, including consideration of dose to endosteal. surfaces and active bone marrow rather than dose to whole bone.
This report gives j
dose-conversion factors for acute intake of a given radionuclide by i
ingestion or' inhalation at various ages from birth to adulthood, using I
the methodology of,ICRP 26 and 30, but modified and extended as j
appropriate to include age-dependence.
Results for 32 isotopes of 4
strontium, plutonium, americium, and curium are tabulated.
I NUREC/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT f
RISK. HERMANN,0.W.s BURNS,T.J.
Oak Ridge National Laboratory.
April 1984.
23pp.
8405210566.
ORNL/TM-8964.
24526:257.
The conseguences, or risks, from light-water reactor accidents.
j have been evaluated as a function of containment building leakage
~
rates.
The analysis used the set of generic source terms and frequencies of occurrence developed as representative of the range of 4
postulated types of accidents currently applied in reactor safety i
research, and the calculated' result was.the. variable M(sp)., defined as I
the accident-spectrum weighted impact fraction rate from containment building leakage.
Explicitly, M(sp) was formulated as the sum of fractional increases in conseguences, due to the building leakage, for I
each type of accident weighted by its frequency of> occurrence.
The l
base case common to similar' types of analyses was applied.
The.
computed result was M(sp) less than or equal to 1.5 10(-3) fractional increase in the accident spectrum risk per X/ day containment building
{
1eakage rate.
t f
i 38 J
.1-_r,--
- r%
,,.w_,_,_m
_m.,
m-.,.,m,-ne,.,_,y__,.,,-
,,,,,,,_.m,,-
.g-.
.~,_,y
_o
,,,e-_-.,
i l
'NUREO/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIQUE CRACK OROWTH RATES OF A 351 CFSA CAST STAINLESS STEEL IN LWR ENVIRONMENT. CULLEN W.H.'s TAYLOR.' R. E. s TORRONEN,K.s et al.
Materials Engineering Associates.
Inc.
April 1984.
36pp.
8405220001.
MEA-2030.
24551:323.
The fatigue crack growth rates for A 351-CFBA cast stainless steel were determined over a range of temperatures from 95 degrees centigrade to 338 degrees centigrade (200 degrees to 640 d egrees fahrenheit).
The waveform was 17 mHz sinusoidal and the load ratio
(
. was 0.2.
The environment was borated and lithiated water with a i
dissolved oxygen content of
- 1 ppb.
The results show an easily measurable (factors of'2 to 8) increase in crack growth rates due to the environment.
However, thesa rates are well within the known band of results for low-alloy pressure vessel and low-carbon piping steels in LWR environments.
An extensive fractographic investigation shows fatigue fracture surfaces covered w'th brittle-like features.
This morphology is similar to that resulting from the environmental assistance mechanism producing increased crack growth rates due to i
stress-corrosion cracking.
NUREO/CR-3564: PRESSURIZED THERMAL. SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER 4
REACTOR. EYLER.L.L.
TRENT,D.S.
Battelle Memorial Institute. Pacific Northwest Laboratories.
April 1984.
92pp.
8404300363.
PNL-4?O9.
24233:144.
1 The TEMPEST computer program was used to simulate fluid and thermal' mixing in the cold leg and downcomer of a pressurized water i
reactor under emergency core cooling high-pressure injection (HPI),
t which is of concern to the pressurized thermal shock (PTS) problem.
l Application of the code was made in performing an analysis simulation of a full-Scale Westinghouse three-loop plant design cold leg and downcomer.
Verification / assessment of the code was performed and
]
analysis procedures developed using data from Creare 1/5-scale i
experimental tests.
Results of three simulations are presented.
The
)
first is a no-loop-flow case with high-velocitye low-negative-bougancy i
HPI in a 1/5-scale model of a cold leg and downcomer.
The second is a l
l no-loop-flow case with low-velocity, high-negative density (modeled l
with salt water) injection in a 1/5-scale model.
Comparison of i
j TEMPEST code predictions with experimental data for these two cases show good agreement.
The third simulation is a three-dimensional i
model of a one loop of a full size Westinghouse three-loop plant j
i design.
Included in this latter simulation are loop components extending from the steam generator to the reactor vessel and a one-third sector of the vessel downcomer and lower plenum.
No data i
j were availabis for this case.
i l
NUREO/CR-3566: SOCIDECCNOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
1 TAWILL.J.J.s CALLAWAY,J.W.s COLES.B.L.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratories.
June 1984.
206pp.
5 8406270117.
PNL-4911.
25171:138.
I This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at'a nuclear power plant.
The types of impacts that are addressed include economic impacts, health impacts.
l i
social / psychological impacts and institutional impacts.
These impacts are identified for each of several phases of a reactor accident--f rom
)
the warning phase through the post-resettlement phase.
The relative j_
importance of the impact during each accident phase and the degree to i
which the impact can be predicted are indicated.
The report also 37 4
,r.
,,._.._r_y
-,,,,. __~,-.._.,_- _,-
m._,.,-,
-,mm._
--n-.--
d
^
osaminas the methodsLthat are currently used for assessing nuclear reacter~ accidents, including development of accident scenarios end.the estimating of socioeconomic accident conseguences with various models, Finally ; a critical evaluation is made regarding the use of impact analyses in. estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk.
a T
[
NURE0/CR-3547: TRAC-PF1:AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR i
PRESSURIZED WATER REACTOR ANALYSIS. a Los Alamos Scientific Laboratory.
April.1984.
60pp.
5405220073.
LA-7744-MS.
24558:011.
4
- The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate t'
predictions of postulated accidents in light water reactors.
The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic esperimental facilities.
The
[
code features either a one-dimensional or a three-dimensional I
treatment of the pressure vessel and its associated internaiss -a -
two-phase, two-fluid nonequilibrium hydrodynamics model with a j
noncondensable gas fields flow-regime-dependent constitutive equation I
i treatments optional reflood tracking capability for both bottom flood j
and falling-film quench fronts: and consistent treatment of entire accident seguences including the generLtion of consistent initial-conditions.
A new numerical algorithm is used in the one-dimensional'i
+
if hydrodynamics that. permit this portion of the fluid dynamics to l
violate the material Courant condition.
This technique permits large time steps and, hence.-reduced running time for slow transients.
This report describes the thermal-hydraulic models and the j
numerical solution methods used in the code.
Detailed programming and j
user information also are provided.
A second Los Alamos report,
" TRAC-PF1 Developmental Assessment," presents the results of the developmental assessment calculations.
i 4
L NUREO/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR-HLW j
j DOSIMETRY (ICRP-30)
I.
ECKERMAN, K. F. s LEGGETT, R. W. s MEYER, R. s ' e t al.
t a
Dak Ridge National-Laboratory.
May 1984.
75pp.
5405290437.
ORNL/TM-9939, 24696:176.
This report provides an initial evaluation of the dependence on chemical forms of estimates of health effects from radionuclides in high-level waste (HLW).
Discussion is limited mainly to a review of studies of plutonium, americium, neptunium, and strontium that may be l
j useful in identifying (a) chemical forms of these radionuclides that j
[
are likely to reach humans after migration from a waste repository and j
(b) differences in metabolism and organ doses that result from intake l
l of various chemical forms of these radionuclides; we also attempt to j
identify research needs in these two areas.
In addition to providing I
a limited review of the literature, this report identifies some of the problems involved in determining speciation of these radionuclides in the' environment and provides a general picture of the potential errors that may be involved in applying models assumed to be independent of chemical form to estimate metabolism and dose from exposure to
(
[
different chemical species of a radionuclide, f
i NURE0/CR-3593: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR f
OBTAINING SITE CHARACTERISTIC INFORMATION. ESTES,J.E r SCEPAN,J.s i
RITTER, L s et al.
California, Univ. of, Santa Barbara, CA.
April 1994.
79pp.
3407110004.
S-762-R.
25545:247.
l The Nuclear Regulatory Commission contracted with E060/EM and the l
l as L
~ University o8 California, Santa-Barbara to assess the potential of
. photographic remote sensing for demographic and. environmental monitoring.
Aerial infrared imagery and ground truth along with collateral data provided inf ormation on site area ' demograp hics and land use, land cover characteristics.
The ability to determine transient populations from remotely sensed data was also evaluated.
Both manual and machine-assisted techniques for extracting these data from reflectance infrared images were gualitatively assessed.
The i
NASA Aircraft Programs
'U-2' acquired color infrared imagery at scales of 1: 45,000 and 1: 130,000, and Keystone Aerial Surveys (Ph ilad e lp h ia, j
Pennsylvania) using a Lear-Jet acguired color infrared imagery at scales of 1: 36, 000, 1:48.000, 1:60, 000, and 1:50,000.
Data on residence types and counts, industrial facilities types and location.
[
transient facilities. transportation networks, and the location of 7
water bodies were generated specifically for the study site i
surrounding the Limerick Power Station in Pottstown, Pennsylvania.
Of the three technigues of population estimations examined, the " Dwelling Unit" method was evaluated for respective utility and accuracy within NRC guidelines.
The level of spatial and classification accuracy of i
the derived products depended on both scale and image guality.
Area ~
[
l weighed thematic accuracy from manual analysis was 96X, while
~
~
I j
hy-category accuracies ranged from 71X to 100X.
1 r
i i
NURE0/CR-3500: THE EWECT 'OF LOCA S!MULATION PROCEDURES ON CROSS-LINKED j'
POLYOLEFIN CABLE'S PERFORMANCE. BUSTARD L.D.
Sandia Laboratories.
l 4
April 1994.
100pp.
4407060059.
SANDS 3-2406.
25441:190.
l Electri"a1 and mechanical properties of three commercial t
cross-linked polyolefin (XLPO) materials, typically used as electrical l
cable insulation, have been monitored during three simulations of j
nuclear power plant aging and accident stresses.
For one XLPO cable l_
we first performed accelerated thermal aging, then irradiated the i
samples to the combined aging and LOCA total dose.
Finally, we j'
applied a steam esposure.
For a second and third set of XLPD cables l
we'used simultaneous radiation and steam exposures to simulate a LOCA
[
l environment.
[
j Dur measurement parameters during these tests included:
dc q
insulation resistance, ac leakage current, ultimate tensile strengthe
[
t ultimate tensile elongation, percentage dimensional changes, and i
j percentage moisture absorption.
We present test results for three
[
1 XLPO materials.
i i
j NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN Oak Ridge National Laboratory.
April 1994.
130pp.
j' FY 1993-1997, e l
0405220027.
OPNL/TM-9000.
24603:202.
I l
The first in an annual series of five year program plan documents
[
is presented for the Heavy-Section Steel Technology program.
The program is carried out by the Oak Ridge National Laboratory for the i
Material Engineering Branch, Division of Engineering Technology, i
Office of Nuclear Regulatory Research of the U.S.
Nuclear Regulatory i
Commission.
The' program is aimed at advancing the understanding and validation of materials and structures behavior as they relate to I
light water reactor pressure vessel integrity.
The program has nine technical tasks and a management function.
A background statement and I
a plan-of-action is given for each.
The nine technical tasks address l
fracture methodology and analysis, materials characterization, crack i
growthe crack arrest,-irradiation effects, cladding evaluations, l
intermediate-vessel testing, thermal-shock testing, and pressurized thermal-shock esperiments.
se i-
[
NURE0/C>3596: SEVERE ACCIDENT SEQUENCE ANALYO!O (SASA) PT.00 RAM
- t SESUENCE EVENT TREE: BOILING WATER REACTOR ANTICIPATED TRANS!ENT i
WITHOUT SCRAM. BRUSME S.Z.s WRIGHT,R.E.
E0&G. Inc.
April 1984.
22pp.
8405220031.
E00-22SS.
24601:001.
L The United States Nuclear Regulatory Commission is sponsoring an en-going safety research. program to assess dominant risk events in boiling water reactors.
As part of this programe a seguence event tree for a boiling water reactor anticipated transient without scram i
(ATWS) has been developed and guantified.
The goal of the sequence I
event tree is to provide a logical representation of the systems that must respond to an ATWS. the required operator response to the event, operater actions that could be performed in response to multiple j
failures, and the phenomenological concerns.
The purpose of the sequence event tree is to provide a basis upon which to perform 4
additional deterministic thermal-hydraulic and core damage analyses in the most cost effective manner based on the most likely sequence of events that will lead to containment / core damage.
The ATWS seguence r
event tree is based on the General Electric Dwners Group emergency l
procedure guidelines and on preliminary deterministic l
thermal-hydraulic analyses performed by E060 !dahoe Inc. personnel at the Idaho National engineering laboratory under direction of the i
Severe Accident Sequence Analysis Program.
}
t NURE0/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST H I-4.
OSSORNE.M.F.s COLLINS J.L.1 LORENZ.R.A.s et al.
Oak Ridge National Laboratory.
June.1984.
70pp.
5407020154.
ORNL/TM-9011.
2S275:328.
i The fourth in a series of high-temperature fission product release tests was conducted in which a 20.3-cm-long fuel specimen from i
the Peach Bottom-2 reacter was heated for 20 min at a mesimum I
temperature of *1950 degrees centigrade in a flowing steam-helium f
j atmosphere.
The test specimen was part of a fuel rod which was i
irradiated to *10,10 mwd /kg.
3 Posttest metallographic esamination of the fuel specimen revealed
)
evidence of cladding melting at each of the transverse cuts that were made.
Oas analysis during the test indicated that '34% of the cladding was naidised.
Total oxidation did not occur because of the i
low steam flew which was used.
)
Gamma sepctrometry (GS) and neutron activation (NA) analyses of I
i test components revealed the following releases: (1) 05 - 21.1%
(SS) Men ' 31. 7X (137)Ces and (2) NA - 24.7% (129)! (percentages of-the total calculated segment inventories).
A value of 33. 8% c esium release was determined by counting the fuel red segment before and after the test, If the pellet-clad gap fission gas inventory had also been available for release in the teste the (SS)Mr release would have l
been 31.3%.
i Significant releases of radiogenic Rbe Cde Age and Bra as well as l
trace amounts of Tee La. Sa. Sr. and Eu, were detected by spark-source l
mass spectrometric analysis.
I NURE0/CR-3603: MINET VALIDATION SURVEY USING E90-!! TEST DATA. VAN I
TUYLE O.J.
Brookhaven National Laboratory.
May 1984.
39pp.
5403210371.
BNL-NURE0-51733, 24329:311.
A natural circulation test transient performed at EBR-!! facility
[
is simulated using the MINET computer code, and calculated results are compared against data from the plant.
The MINET EBR-!! representation F
includes much of the intermediate loop and the steam generater system.
l
l i
and corresponds to the portion of the plant usually represented by MINET when it is. executed with SSC, the-Super System Code.
,MINET calculations agreed,well with the plant transient data,_ with-discrepancies well within uncertainties in thermocouple time constants and boundary. conditions.
l NUREO/CR-3604: BOLTING APP'LICATIONS. C ZAJKDWSKI, C. J.i Brookhaven National Laboratory.
May-1984.
303pp.
8406120535.
BNL-NUREG-51735..
24893:004..
An' investigation of bolting practices specific to the nuclear industry was performed.
The report covered -a large spectrum of topics
- e. g.
bolts embedded in concrete, specifications, inspection of bolting,'both at. receipt and inservice.
Plots of preload versus gield
. strength for different bolting materials in different environments are
' presented as well as information relative to the stress corrosion cracking. resistance of the more recent reactor internals bolting materials A286 and Incone l X-750.
Partaof the report contains input by Standard Pressed Steel Inc. (a bolting consultant) relative to bolting standards, cottering methods and potential areas fer bolting improvement.
NUREO/CR-3606: NUCLEAR. POWER PLANT CONTROL ROOM CREW TASK' ANALYSIS DATABASE: SEEK SYSTEM.. (Users Manual). BURGY,D.s SCHROEDER,L.
General Physics Corp.
May.1984.
134pp.
8406190517.
GP-R-212106.
4 i.,
25029:052.
.The Crew Task Analysis SEEK Users Manual was prepared for.the 4
Office of Nuclear Regulatory.Research of the U.S.
Nuclear Regulatory Commission.
It is designed for use with the existing computerized i
Control Room Crew Task Analysis Database.
The SEEK system consists of
.a PRIME computer with its associated peripherals and software l
augmented by General Physics Corporation SEEK database management
~
software.
The SEEK sof tware programs provide.the Crew Tas k Database user with rapid access to any number of' records desired.
The software uses English-like sentences to allow the user to construct logical i
sorts and outputs of the task data.
Given the multiple-associative nature -of the database, users can directly access the data at the i
p l an t', op'Erating sequence, task, or elosent level - or any combination of these levels.'
A complete descriptio6 of the crew task data i
contained in the database is ' presented ' in NUREG/CR-3371, " Task Analysis of Nuclear Power Plant Control Room Crews (Volumes 1 and 2)."
l
>+
{A~
NUREG/CR-3608: RELAP5 ASSESSEMENT: LOFT Large Break L2-5. THOMPSON, S.'L. ;
i KMETYK', L. N.
Sandia Laboratories.
April 1984.
115pp.
8405220255.
.s SAND 83-2549.
24602:001.
.The RELAP5 independent assessment project at Sandia National j';,
. 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 LWRs during accidents and off-normal s
( d)'
'c'o n d i t i o n s.
The RELAP5 code is being assessedat SNLA' against test
[~7 '.
_ data from.various integral and separate effects test ~ facilities.
As part' of this assessment matrix, a large break transient perf ormed at the LOFT facility has'been analyzed.
e The results show that RELAP5/ MODI <co'r're'et1g calculates many of the' major system variables (i.e.,' pressure, break flows, p eak clad temperature) earig in a large break LOCA.
The major problems
!m encountered in the analyses were incorrect pump coastdown and loop in seal yclearing earig in the calculation,. excessive pump = speedup later
~
~wt '
n 1A' 41 bI.
!i+'
y
<a x
_.. - -,., -.. _... -. ~
'in the transient.(probably due to too much condensation-induced pressure drop at the ECC injection point), and excess ECC bypass-calculated throughout the later portions of the. tests only the latter problem significantly affected the overall:results.
This excess ECC r
bypass through the downcomer and vessel-side break resulted in too-large' late-time break flows and high system pressure due to prolonged choked flow condivions.
It also resulted in a second core heatup being calculated-after:the accumulator emptied, since water was not being. retained in the vessel.
Analogous calculations with a split-downcomer nodal 12ation deliveredLsome ECC water to the lower
. plenum,' which was then swept up the core and upper plenum and out the other - (pump-side) breaks thus no significant differences in long-term overall behavior were evident.
L NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983. ATTERIDGE,D.G.;
BRUEMMER,S.M.; AGE. R. E..
Battelle Memorial Institute, Pacific Northwest Laboratories.
June 1984.
55pp.
8406280214.
PNL-4941.
25192:277.
Pacific Northwest Laboratory (PNL), under a program sponsored by the Division of Engineering Technology of the U.S.
Nuclear Regulatory Commission (NRC), is conducting a. program to determine a method for i
evaluating the. acceptance lof welded and repair-welded stainless steel (SS) piping for light water reactor (LWR) service.
Validated models, based on experimental data, will be developed to predict the degree of sensitization (DOS) and the intergranular stress corrosion cracking (IGSCC) susceptibility in the heat affected zone (HAZ) of the SS weldments.
IGSCC'is caused by a combination of a sensitized microstructure, an aggressive environment, and tensile stress.
Control of any of these three factors can eliminate IGSCC in'most practical situations.
l This program will measure and model the development of a sensitized microstructure as it pertains to welded ~and repair-welded
}
SS pipe.
An empirical correlation between a material's DOS and its susceptibility to IGSCC uill be determined using constant extension rate tests (CERTs).
The successful completion of these tasks will result in a cethod for assessing the effects of weld / repairing-parameters on the IGSCC susceptibility of component-specific nuclear reactor welds / repairs.
t NOREO/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CABLE FAILURE WITH MECHANICAL DEGRADATION. STUETZER,0.
Sandia Laboratories.
April 1984.
-90pp.
8406250272.
SAND 83-2622.
25139:151.
An attempt is being made to assess complete electrical failure of I
signal and lou-power cables typically used in nuclear power plant containments and to correlate failure modes with the mechanical deterioration of the elastomeric cable material.
Work over the past 24 months, although limited to one cable configuration, has identified creep shortout and insulator cracking, both aggravated by mechanical stresses, as the phenomena most likely to cause electrical breakdown.
Comprehensive tests have been run for six months and are continuing.
Preliminary conclusions'can be drawn and are reported.
NUREG/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S GUIDE FOR THE GENERATION OF LATIN - HYPERCUBE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
LIMAN,R.L.s SHORTENCARIER Sandia Laboratories.
June 1984.
67pp.
8407110012.
SAND 83-2365.
25545:178.
r l~
u
,This docunent has been designed for users of the computer program
~ developed by.the cuthors at Sandia National Laboratories for the generation of either Latin hypercube or random multivariate samples-The Latin hypercube technique employs a constrained sampling scheme, whereas' random sampling. corresponds to a simple Monte Carlo technique.
Tha-generation of these samples is based on information supplied to the program by the user describing.the variables or parameters used as input to the computer model.
The actual sampled values are used to form vectors of variables commonig used as input to computer models for purposes of sensitivity and uncertainty. analysis studies.
The present program replaces the previous Latin hypercube sampling program developed at Sandia National Laboratories (Iman, Davenport, and Zeigler, 1980).
The present version is written using FORTRAN 77 and greatig extends the program while making the program portable and user j
friendig.
NUREO/CR-3626 VO1: MAINTENANCE ~ PERSONNEL PERFORMANCE SIMULATION (MAPPS)
MODEL:
SUMMARY
DESCRIPTION. SIEGEL A.I.s BARTTER,W.D.; WOLF.J.J.i et al.
Oak Ridge National Laboratory.
May 1984.
52pp.
8407060056.
ORNL/TM-9041/V1.
25452:250.
A summary description is presented of-the rationale for and the content and structure of the Maintenance Personnel Performance Simulation ~(MAPPS) model.
-The MAPPS model is a generalized stochastic computer simulation model developed to simulate the performance of naintenance personnel in nuclear' power plants.
The MAPPS model i
considers workplace, maintenance technician, motivation, human factors, and task-oriented variables to yield predictive information about the ef fects of these variables on successful maintenance personnel requirements.
The model, which is drawn from a firm research analytic base, uas examined for disqualifying defects from'a number of viewpoints and its sensitivity was extensively tested.
The MAPPS model is believed-to be ready for initial and' controlled applications which are in conformity with its purposes.
NUREO/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
GINIBURG,T.3 BOCCIO.J.L.s HALL,R.E.
Brookhaven National Laboratory.
June 1984.
151pp.
8406270111.
BNL-NUREG-51738.
25174:001.
This report deals with practical applications of the FRANTIC II code in analyzing the reliability of standby safety systems.
Time-dependent unavailability models such as FRANTIC II have two important advantages over more simplistic time-independent models:
(1) accountability for the " burn-in" and " wear out" effects in describing component failure distributions and (2) distinguishability betueen two systems having the same average unavailability, but with different periods of high risk.
.Thus, studies can be performed to assess the percentage of time the system spends with unavailability above a prescribed threshold level.
This report demonstrates the capability of FRANTIC II to evaluate the. standby safety system unavailability on a more realistic basis and perform a detailed examination of period testing policies.
Once the requisite input parameters to FRANTIC have been described and j
interpreted, and estimates made-from the available data, the code is l
applied to the three systems:
Emergency Feedwater System ( P WR ) s ^
Automatic Depressurization System (BWR): and High Pressure Coolant Injection System (BWR),
l The analysis includes system description, fault tree quantification, unavailability calculation, and error propagation P
43 l
u ovoluution.
Suggooticns.cro also mndo en how.to optimize gathering plant reliability' data.
NOREO/CR-3628: PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT
' STRUC TURES. ELLINGWOOD,B.
Brookhaven National Laboratory.
Commerce, Dept. Hof, National. Bureau of Standards.
May 1984.
73pp.
-8405210585.
BNL-NUREG-51737.
24535:181.
This report describes the basis for the development of. practical Probability-based design criteria for nuclear plant structures.
A
- brief. critical review of existing criteria is provided to highlight desirable. features of probability-based-safety checking.
A specific.
deterministic design _ criteria format is then recommended.
- Finally, the selection of a set of structures to test the validity of the probability-based checking equations is described.
Statistical data on structural loads are summarized in an appendix.
NUREO/CR-3629: THE EFFECT OF THERMAL AND: IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES. BUSTARD L.D.3 MINOR,E.s CHENION,J.; et al.
Sandia Laboratories.
May 1984.
81pp.
8405210589.
SAND 83-2651.
24532:257.
Prior to initiating a qualification test on safety-related equipment, the testing sequence for thermal and irradiation aging exposure must by chosen.
Likewise, the temperature during irradiation must be selected.
Typically, U. S.
qualification efforts employ ambient temperature irradiation while French qualification efforts employ 70 degree C irradiations.
For several polymer _ materials, the influence of the thermal and irradiation aging sequence.has been investigated in preparation f or Loss-Of-Coolant Accident s imulated tests.
Ultimate-tensile properties at completion of aging are presented for three XLPO and XLPE, five EPR and EPDM, two CSPE-(HYPALON), one CPE, one VAMAC,.one polydiallyphtalate, and one PPS material.
Bend test results at completion of aging are presented for two TEFZEL materials.
Permanent set after compression results are presented for-three EPR, one VAMAC, one BUNA N, one Silicone, and one Viton material.
NUREC/CR-3630: EQUIPMENT GUALIFICATION METHODOLOGY RESEARCH: TESTS OF PRESSURE SWITCHES. SALAZAR,E.A.
Sandia Laboratories.
April 1984.
180pp.
8406210432.
SAND 83-2652.
25097:223.
Pressure switches, two each of five different models from two manufacturers, were tested in baseline evaluation tests typical of IEEE-323 (1974) suggested profiles as part of the.NRC-sponsored Equipment Gualification Methodology Research Test Program (A-1355).
The tests incorporated generic seismic and loss-of-coolant accident (LOCA) environments to assess the functional capabilities of unaged equipment.
During the baseline evaluation tests, the seismic environment did not affect the functionality of the pressure switches, but the LOCA environment caused numerous functional failures and extensive physical damage in four of five models tested.
As a result, eight other switches of the same make and model as those used in the baseline evaluation tests were tested in a follow-up test.
In the follow-up test (a discrete-step pressure ramp LOCA environment) erratic functional behavior or complete failure was observed in all the equipment early in the test.
44
W
~
- NUREO/CR-3632: IMETHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING PROCEDURES.' MYERS.L.B.s BELL,A.J.
Battelle Memorial Institute,. Columbus Laboratories.
- Battelle Memorial Institute, f?acific Northwest Laboratories.
May 1984.
38pp.
8405210600.
~
PNL-4927.
24534:200.
~ Island 1(TMI)' accident, the U.S.
In response 1to'the Three-Mile Nuclear. Regulatory-Commission (NRC) has published the TMI Action Plan.
The TMI Action Plan Item I.C.1 called for the upgrading of Emergency Operating Procedures (EOPs) _at nuclear power plants.
The
-program developed.from this Action Plan. item has resulted in utility-efforts to 1) revise EOPs, 2) train personnel in the use of the EOPs, and'3) ~ implement the revised EOPs.
The.NRC supported the study presented in this report to identify factors which-influence the effectiveness of training and implementation of revised EOPs.
The NRC's major concern was the possible effects of negative' transfer of training.
The report includes a summary of existing methods' for' implementing revisions'to procedures based on interviews _of plant personnel, a review of the
-training ' literature applicable to the effect of previously learned procedures on the learning of and-performance with revised procedures (i.e.,
negative transfer) and recommendations of methods and. schedules for. implementing revised EOPs.
While the study found that the concern over negative transfer of training was not as great as anticipated, several recommendations were1made.
These include (1) overtraining of operators to reduce the effect of observed negative trcnsfer, and (2) implementation of the revised EOPs as soon as possible after training to minimize-the time operators must rely upon the old EOPs after i
having been trained on the revised EOPs.
The results of the study i
should be useful both to the utilities end the NRC-in the development
.and review of CDP impismantation programs.
i NUREC/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGR.J M FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model Description. TAYLOR,D.D.; MOHR,C.M.
EG&G, Inc.
Apri1~1984.
231pp.
)
8405210576.
EGC-2294.
24528:001.
The TRAC-BD1/ MOD 1 computer program provides a best-es timate analysis capability for the analysis of the full range of postulated accidents in Boiling Water Reactor (BWR) systems and related f
experimental facilities.
The program is described in four volumes:
Volume 1,- Code Description, Volume 2.
User 's Guides Volume 3, Code Structure and Programming Informations and Volume 4, Developmental Assessment.
Volume 1 describes the thermal-hydraulic models, numerical methods, and component models available.
Volume 2 describes the input and output of the TRAC-BD1/ MOD 1 code and provides guidelines for use of the code modeling of BWR systems.
Volume 3 is designed for the programmer or model developer who needs to, implement or modify the 3
TRAC-BD1/ MOD 1_ program.
Volume 4 discusses the results of the development assessment calculations performed with TRAC-BD1/ MOD 1.
U NUREO/CR-3633 VO2: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER f
PROGRAM FOR BOILING W.^.TER REACTOR TR ANSIENT ANALYSIS. Volume 2: Users-
~I i
Guide. SCHUMWAY,R.W.s MOHR C.M.
EG&G, Inc.
April 1984.
117pp.
8405210578.
EGG-2294.
24528:232.
See NUREG/CR-3633,VO1 abstract.
NUREO/CR-3633 VO3: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER "ROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code 3
i.
g
Structure and-Programming Information. SINGER G.L.3 MOHR,C.M.
EG&G, Inc.
April 1984, 110pp.
8405210579.
EGG-2294.
24528:357.
See NUREG/CR-3633,VO1 abstract.
i
' NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM. HILL.J.R.; HEGER,A.S.; KDEN.B.V.s et al.
April 1984.
42pp.
8405220179.
EGG-2295.
24594:269.
This report is the result of a preliminary feasibility study of the applicability of-Stein and related parametric empirical Bayes (PEB) estimators to the Nuclear Plant Reliability Data System (NPRDS).
A new estimator is derived for the means of several independent Poissor. distributions with different sampling times.
This estimator is applied-to data from NPRDS in an attempt to improve failure rate estimation.
Theoretical and Monte Carlo results indicate that the new PEB estimator can perform significantly better than the standard maximum likelihood estimator if the estimation of the individual means can be combined through the loss function or through a parametric class of prior distributions.
NUREG/CR-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION. DOBRANICh.D.J BUXTON,L.D.
Sandia.
Laboratories.
May 1984.
67pp.
8406040023.
SAND 84-OO40.
24805:001.
A series of best estimate thermal-hydraulic calculations was performed using_ TRAC-PF1 to simulate a hypothetical loss-of-coolant accident in Westinghouse two-loop pressurized water reactors.
Those reactors are equipped for low pressure injection of emergency coolant direct 1g into the upper plenum of the reactor vessel.
This type of I
injection is referred to as upper plenum injection (UPI).
The calculations uere performed to evaluate the effectiveness of UPI compared to injection into the vessel downcomer, referred to as downcomer injection (DI).
The TRAC results indicated that some channeling of upper plenum injected liquid down the core periphery occurreds however, a large percentage of that liquid was vaporized as it drained toward the lower plenum.
This vaporization degraded the bottom-flood quench front compared to that seen in TRAC calculations in which downcomer injection was assumed.
For the case of upper plenum injec tion, counter-current flow limiting conditions at the upper core support plate led to formation of a large subcooled liquid pool in the upper p lenums part of this subcooled liquid was entrained into the hot legs and steam generators.
Onig a small saturated liquid pool formed in the case of downcomer injection.
Overall, the calculations show that higher peak clad temperatures are produced when the low pressure injection is into the upper plenum instead of the vessel downcomer.
NUREO/CR-3641: RELI ABILITY ASSESSMENT OF INDI AN POINT UNIT 3 CONTAINMENT STRUCTURE. KAWAKAMI,J.; HWANG,H.s CHANG,M.T.1 et al.
Brookhaven National Laboratory.
May 1984.
51pp.
8405310082.
BNL-NUREG-51740, 24737:019.
In the current design criteria, the load combinations specified or design of concrete containment structures are in the deterministic formats.
However, by applying the probability-based reliability method developed by BNL to the concrete containment structures designed according to the criteria, it is possible to evaluate the reliability levels implied in the current design criteria.
For this 46 e
purpose, :the reliability analysis is applied to the Indian-Point Unit' No. 3 containment.
The details of the. containment structure (such as the geometries
.and'the rebar arrangements, etc., are taken from the working drawings and.the final safety analysis reports.
.Three kinds of loads are considered in the reliability analysis.
They are, dead' load accidental pressure due to a-large LOCA (P), and earthquake ground acceleration (E).
Reliability analysis of the containment subjected to all combinations of loads is performed.
The results are presented i
in this report.
f-NUREG/CR-3644: REVIEW OF PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS. JU F.D.
BUTLER,[T. A.
Los' Alamos Scientific Laboratory.
April.1984.
36p p.'
8405220015.
LA-10007-MS.
24554:308.
In this reporte failure criteria for structural components constituting'the primary coolant-system boundary of a Liquid Metal Fast Breeder-Reactor (LMFBR) are reviewed.
Because the materials
' b e'ing considered, mild ferritic steel and austenitic s,t a i n l e s s steels, are ductile, especially under LMFBR normal operating and-accident conditions, onig ductile criteria are considered.
The ductile F
criteria must be used in combination with true stress and strain.
neasures of deformation and internal. load.
. Specific criteria reviewed j
. include maximum stress and strain or' plastic instability based on
[
uniaxial tensile-test data and a hole growth theory based on coalescence of naighboring voids under load.
Criteria based onig on l
maximum stress or strain are not recommended for general use because they are not appropriate under general multiaxial stress conditions.
The plastic instability criterion, because it leaves a large unused toughness region before fracture, is recommended where considerable conservatism is warranted.
The hole growth criterion is recognized'as j
being analytically sounds however, it has not been extended to general three-dimensional geometry and multiaxial stress conditions.
The theory needs to'be substantiated with experimental data for specific materials being considered.
NUREC/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARI ABILITY - Some Preliminary Results. MARTZ, H. F. s WHITEMAN,D.E.
Los Alamos Scientific Laboratory.
April 1984.
55pp.
i 8405220072.
LA-10014-MS.
24557:315.
In-Plant Reliability Data System.(IPRDS) pump failure data on over 60 selected pumps in four nuclear power plants are statistically analyzed using the Failure Rate Analysis Code (FRAC).
A major purpose of the analysis is to determine which environmental, system, and operating factors adequately explain the variability in the failure data.
Catastrophic, degraded, and incipient failure severity categories are. considered for both demand-related and time-dependent failures.
i For catastrophic demand-related pump failures, the variability is explained by the following factors listed in their order of importance: system application, pump driver, operating mode, reactor type, pump type, and unidentified plant-specific influences.
Quantitative failure rate adjustments are provided for the effects of these factors.
j In the case of catastrophic time-dependent pump failures, the failure rate variability is explained by three factors: reactor type,
-pump driver, and unidentified plant-specific influences.
d' Finally, point and confidence interval failure rate estimates are
.provided for each selected pump by considering the influential
~
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47
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-v.
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-w.y..-~nw.,
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,-,,,w,.e,,,m---
- -. -., - -..,.. -, -. ~,. - - -. - -
[facQors.
Both types of estimates represent an~ improvement over the estimates computed exclusively from the data oro each pump.
NUREQ/CR-3652: EVALUATION 'OF INSTRUMENTATION FOR DETECTION OF INADEGUATE CORE COOLING IN BOILING WATER REACTORS. LEWIN,T.
.0ak.
Ridge National Laboratory.
April.1984.
31pp.
8405210602.
'ORNL/TM-9029.
24526:278.
-This report is a review.of.the APPROACH TO INADEGUATE CORE COOLINO issue in Boiling Water Reactors.
The report consists of seven sections.
The principal conc ~ usion is that the condition of the 1
reference leg, and operator awareness of that condition are of primary importance in level indication reliability for safety.
An indication of reference leg level and temperature displayed to the op erators.
would be a useful enhancement of reliability and a guide to further operator action in~all. circumstances.
.6he conclude th a t th e BWR practice of multiple, redundant coolant. level measurements..with overlapping ranges, can be a reliable basis for indication of approach to an ICC condition, and, in correlation with the other control and safety systems of modern BWRs, will prevent unsafe conditions.
NUREO/CR-3653: CONTAINMENT ANALYSIS TECHNIQUES. A State-Of-Th e-Art i
Summar y.. GREIMANN, L. 's FANOUS,F.s BLUHM,D.
Ames Laboratory, Energy &
Mineral Resources Research Institute.
April 1984 167pp.
8406210097.
SAND 83-7463.
25097:055.
The purpose of the work contained herein is to review the state-of-the art for the analysis of LWR nuclear containments with 4
j uniform internal pressure.
This includes:
(a)
A review of calculated static failure pressure of various containments.
(b)
A review of the different failure criteria used for i
predicting containment failure.
}
(c)
Comments on possible uncertainties associated with analysis i
techniques, material and geometric models, and other analysis features.
A state-of-the-art containment analysis is a finite element solution of an axisymmetric model.
Material and geometric j
nonlinearities are included.
Nonsymmetric features may be analyzed on an individual basis but are omitted in the axisymmetric model.
State-of-the-art models of the material constitutive relationships are used.
Deformation predictions are generally regarded as reliable, assuming the containment configuration is accurately described,
- e. g.,
known geometry, material and loads.
Predictions of -leakage are much nore uncertain.
There is no general agreement on when and where leakage will occur.
NUREO/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER. WOODLEY, R. E.
Hanford Engineering Development Laboratory.
June 1984.
35pp.
8407190060.
HEDL-TME 84-14.
25693:247.
The performance under dry storage conditions of LWR fuel rods containing water was analyzed to determine if radionuclide containment by the fuel rod cladding would be adversely affected.
Fuel rod and storage canister pressurization as well as cladding and fuel oxidation were examined.using " worst case" conditions.
The results of this study are presented.
48 l
~
NUREG/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD. O' CYCLE 4:
tA COMPUTER' CODE'WITH THREE-DIMENSIONAL' NEUTRON KINETICS FOR BWR-SYSTEM - TR ANSIENTS.~ ' WULFF, W.-; ' CHENG, R. S. a DIAMOND,D.J.i et 'a1.
Brookhaven National Laboratory..May 1984.
428pp-8405210615.
1 B NL-NUR EG-51746, 24533:001.'
i
..This report documents the' physical models and the numerical c
(
snethods employed in the BWR systems code RAMONA-3B.
Th e R AMONA-3B i
. code' simulates three-dimensionalfneutron kinetics and multichannel' L
core hydraulics ofl nonhomogeneous, nonequilibrium two phase flows.
RAMONA-3B is programmed to-calculate the steady and transient'
= conditions'in the_ main steam supply for normal and abnormal operational transients, including-the performances of plant control H
and protection ~ systems.
Presented are code capabilities and: limitations, models and solution techniques, the results-of' developmental code assessment and suggestions for improving the code in the future.
NUREG/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT: A FINAL REPORT ON THE. INVESTIGATION:OF FISSION PRODUCT CHEMICAL FORMS.
HENSLEY,W.K.* ROGERS,L.A.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984.
48pp.
8405020442.
PNL-5003.
24298:075.
In September'of 19654 an intentionally defective' fuel. rod failed in the Plutonium Recycle Test Reactor (PRTR), causing the rupture of the surrounding. pressure _ tube and the release of superheated cooling-
-water.into a region of the reactor core.
The Pacific Northwest t
Laboratory (PNL) has-reviewed the PRTR incident to assemble and update i
all the available information regarding the incident.
A' principal goal of the' review was to' analyze any remaining clues that may i
indicate the stoichiometry or most probable chemical and physical forms of the released fission products.
The review confirmed the role 4
of water in limiting iodine release.
About 97% of the iodine released during the accident was subsequently found in tanks containing the i
reactor / rupture-loop coolant.
Although the chemical form of the i-released radiciodine cannot be stated unambiguously, the'available evidence suggests that it was released in the form of* cesium iodide.
Most of the remaining 3% was found in the condensate collected from air cooling systems.
The chemical form of this scrubbed iodine.
remains undefined.
NUREO/CR-3670: VIOLENT TORNADO CLIMATOGRAPHY, 1800-1982. GRAZULIS,T.P.
Environmental Films, I n c.'
Battelle Memorial Institute, Pacific-Northwest Laboratories.
May 1984.
188pp.
8406040071.
PNL-SOO6.
24806:001.
All known information sources, ranging from newspaper reports to the-University of Chicago (DAPPL) and National Oceanic and Atmospheric Administration (NOAA/NSSFC) data lists, were utilized to produce a i
self-consistent compilation and description of violent tornado
_ occurrences in the contiguous United States for the years 1880 through-1982.
The 969 F-scale 4 and 5 tornadoes comprise the most complete
- and rational data base available for studies related to violent tornado risk assessments tho' data provide improved bases for. licensing i
decisions and development of standards in safety at nuclear facility sites.
Reconciliation of the DAPPL and NSSFC data lists'for violent tornadoes has been achieved.
Analysis of the data shows geographical and' temporal variability of tornado. occurrences 3 suggestions are given F
to help account for the nonuniform distributions, and other i
suggestions are made for needed future research.
4 i
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rT-=P--**=~%+=-+'---wt-arr--est--
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NOREG/CR-3672: ' EXAMINATION OF THE SIZETEFFECTS AND DATA SCATTER OBSERVED IN SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
.MERKLE J.G.
Oak Ridge National Laboratory.
April 1984.
87pp.
8405220026.
ORNL/TM-9038.
24565:013.
.In the transition range of temperature, the cleavage fracture toughness of steel rises-steeply with temperature, often necessitating the use.of elastic-plastic methods for calculating toughness. values with small specimens.
This usually leads to size effects, whereby measured toughness values increase with decreasing specimen size and data scatter becomes,laste.
Existing literature pertaining to the physical aspects of the onset of cleavage fracture and the occurrence of size effects is examined. and it is concluded that the primary cause of size effects is loss of triasial contstraint due to crack tip
.gielding and transverse contraction along the crack front.
The implications of an existing semiemperical equation, known as the Irwin B(Ic) equation, for removing size effects from small specimen cleavage fracture toughness data are examined by developing the equation based on reasonable' assumptions' including the conditions specified by ASTM E399 for valid K(Ic) testing.
The applicability of the Irwin B(Ic) adjustment equation to pressure vessel steels is evaluated by app 1ging it to several sets of small specimen fracture toughness data, and it is found that the equation consistently eliminates apparent size effects and significant1g reduces data scatter.
The B(Ic) adjustment appears to be applicable to dynamic as well as to static initiation toughness data, but only to the cleavage fracture toughness and not to the ductile tearing resistance.
NUREC/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
BURKE,R.P.s ALDRICH,D.C.s RASMUSSEN,N.C.
Sandia Laboratories.
Nag 1984.
250pp.
8406230091.
SAND 84-0178.
25131:005.
Models to be used for analyses of economic risks from events which occur during U.S.
LWR plant operation are developed in this study.
The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment.
The models have been developed for potential use by both the nuclear power industry and regulatory agencies in cost / benefit analyses for decision-making purposes.
The newly developed economic consequence models are applied in an e> ample to estimate the economic risks from operation of the Surry #2 plant.
The analyses indicate that economic risks from LWR operation in constrast to public health risks, are dominated by relatively high-frequency forced outage events.
The im'p l i c a t i ons of this conclusion for U.S.
nuclear power plant operation and regulation are discussed.
The sensitivities and uncertainties in economic risk estimates are also addressed.
NUREG/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION EXPOSURE RATES AND SOIL 266RA CONCENTRATIONS. YOUNG,J.A.s THOMAS,V.W.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984.
25pp.
8405020380.
PNL-5016.
24287:321.
Radon fluxes and contact gamma-radiation exposures rates were neasured at the grid points of rectangular grids on three properties in Edgemont, South Dakota that were known to have deposits of residual radioactivity relatively near to the surface.
The coefficient of determination, r(2), between the radon fluxes and the contact gamma-radiation-exposure rates varied from 0.89 to 0.31 for the three 50
h f
l b
pecportios'.
Corrolations betwoon fluxes and (226)Ra concentrations measured in boreholes that varied in depth from 60 to 195 cm were
-generally lower than those between fluxes and exposure rates, j
indicating that exposure rates are better than (226)Ra mea surements for detecting elevated' radon fluxes from near-surface deposits.
Measurements made on one property at two different times indicated that'if the average flum were determined from a large number (40) of naasurements at-one time, the average f. lux at a later time could be estimated from a few measurements using the assumption that the change in the flux at individual locations will be equal to the change in the
-average flux.
FluxLmeasurements around two buildings showing elevated indoor redon-daughter concentrations, but around which.no residual
)
radioactivity had been discovered by (226)Ra and gamma-radiation l
neasurements, provided no clear indication of the presence of such material.
NUREG/CR-3680: RELATIONSHIP BETWEEN THE GAS CONDUCTIVITY AND GEOMETRY OF A NATURAL FRACTURE. SCHRAUF T.W.s EVANS,D.D.
Arizona, Univ. of, Tucson, AZ.
April 1984.
140pp.
8405220089.
24561:156.
In recent_ gears considerable interest in determining the relationship between the hydraulic conductivity of a rock fracture and its average aperture has developed.
The present study involved both theoretical and experimental studies of the. geometrical factors which influence gas conductivity of rock fractures.
Theoretical analysis of parallel plate gas flow revealed that the gas conductivity of a fracture is the same as.for incompressible fluids and can be expected to follow a cubic law relationship.
Application of the cubic law to e
practical. field test situations, however, was found to be limited by uncertainties in flow boundary conditions, nonlinearity of flow 1
behavior, and effects of fracture surface roughness.
Quantitative l
assessment of uncertainties in flow boundary conditions including
}
elliptical injection boundaries, secondary intersecting fractures, and estimation of efJective radius was performed.
Nonlinear flow behavior uas also analyzed and the results applied to measurements of gas flow 1
rate through a single natural fracture.
Evaluation of<these results suggested a general flow equation of the form: -(dp/dx)= av + bv(2),
uhere a and b are constant coefficients defined by a fracture's i
average aperture and surface roughness.
NUR EG/CR-3681: MITIGATIVE TECHNIQUES AND ANALYSIS OF GENERIC SITE CCNDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS. SHAFER,J.M.; OBERLANDER,P.L.; SKAGGS,R.L.
Battelle Memorial Institute,-Pacific Northwest Laboratories.
April 1984.
367pp.
8405220069.
PNL-5072.
24562:001.
j The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide 4
i migration following a' severe commercial nuclear power reactor accident.
The two types of severe commercial reactor accidents investigated are: 1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and 2) containment basemat penetration of sump water uithout full penetration of the core mass.
Six generic hydrogeologic i
site classifications are developed from an evaluation of reported data pertaining ~to the hydrogeologic properties of all existing and proposed commercial reactor sites.
One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment.
Ground-water contaminant 51
_ _ ~,
k
. mitigation technig'uos..thet may be suitable._doponding.on specific site
~
4
[and accident conditions,.For.. severe power plant accidents are
- identified and evaluated.'. Feasible mitigative techniques and associated constraints on' feasibility are determined for.each of the (six hydrogeologic site classifications.
The fir,st of ~ three case studies is conducted on a~ site located on the Tesas Gulf Coastal Plain..
Mitigative strategies are evaluated for their impact on containment transport and results show that the techniques evaluated U
-significantly. increased' ground-water travel times.
1 NUREO'/CR-3682: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation of Existing Methods. PELTO, P. J. s. -RHOADS R. E. ; VESELY,W.E.s yet al.
)
Battelle Memorial. Institute, Pacific Northwest Laboratories.
May 1984.
133pp.
8405210594.
PNL-4990..
24534:069.
The U.-S.; Nuclear. Regulatory Commission' initiated the Fuel Cycle Risk Assessment Program to provide risk assessment-methods for use in-the regulatory process for nuclear fuel cycle facilities other than,
, reactors.. The first report from this program, NUREG/CR-2873, defined and; described fuel cycle elements-considered in the program.
The second report, NUREG/CR-2933, described the survey and compilation of fuel cycle-risk-related-literature.
This. report presents a review of the state-of-the-art of risk assessment methods for nuclear fuel cycle facilities-and;an evaluation'of the adequacy of these methods to meet NRC's needs for risk assessment information.
The approach used to.
perform this uork included:. identification of potential uses of fuel cycle -risk assessments at NRCs review of currently available fuel cycle risk assessment methodss_and identification of. potential methods development:needs.
NUREG/CR-36R3: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Program Summary i
Through Fiscal Year 1983. GEFFEN,C.A.s PELTO,P.J.s RHOADS,R.E.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May i
1984.
52pp.
8405210582.
PNL-4991.
24535:127.
The U. S.
Nuclear Regulatory Commission initiated the Fuel Cycle Risk-Assessment Program to provide risk assessment methods for use in i
the regulatory process for nuclear fuel cycle facilities other than l
reactors.
This report presents a summary of the work completed in the Fuel Cycle Risk Assessment Program through fiscal year 1983.
These j
efforts include descriptions of representative non-reactor facilities (NUREG/CR-2873),
- a. survey and computer compilation'of risk-related l
literature (NUREG/CR-2933), a preliminary relative ranking of fuel cycle facilities on the basis of risk, and an assessment of the adequcy of existing risk assessment methods (NUREG/CR-3682).
Further work in the program has been postponed at this point in time because of funding constraints and higher priorities for other ongoing l
programs within the NRC.
This program summary document will serve as a reference for use in future fuel cycle risk assessment research.
NUREG/CR-3684: NUCLEAR POWER PLANT ALARM PRIORITIZATION (NPP AP) PROGRAM STATUS REPORT. January 1,1983 to September 31,1983. ROSCOE,B.J.
Sandia Laboratories.
April 1984.
75pp.
8405220042.
SAND 84-0140.
24540:292.
This report describes the status of a research project directed I
toward nuclear power plant alarm prioritization.
Criteria for modified / alarm activation are being developed and studied.
Also bein'g developed are measures to regulate the alarm rate at some desired level.
The problem of alarm prioritization based upon maintenance of 52 l
e-
-_.-,..,----.,_--.n,.
~ -
L-
critico1 cofoty functions whilo maintcining comploto elcrm coverego of accidents is being addressed.
The plant information needed to support the associated technical development areas is being compiled for a specific plant, categorized, and entered into a computer data base.
Near term recommendations for regulatory action on plant annunciator systems are presented.
NUREG/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report. POWELL,G.H.
California, Univ. of, Berkeley, CA.
- Lawrence Livermore National Laboratory.
June 1984.
12pp.
8407110060.
UCRL-15597.
25542:050.
WIPS (Whip and Impact of Piping Systems) is a special purpose computer code for the structural analysis of pipe whip dynamic effects following a postulated pipe rupture.
WIPS has been developed primarily to provide support for the pipe whip analysis procedures described in Section 3.6.2 of the U.S.
Nuclear Regulatory Commission Standard Review Plan.
This report summarizes the purpose and scope of the WIPS development effort, identifying those clauses in the Standard Review Plan which refer to pipe whip analysis, and indicating how the WIPS code can be used to provide supporting data.
Detailed information on use of the code is contained in accompanying reports which cover (1) use instructions, (2) theory, (3) programming procedures, and (4) verification examples.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual. POWELL,G.H.; HOLLINGS,J.P.;
ROW,D.G.; et al.
California, Univ. of, Berkeley, CA.
June 1984.
227pp.
8407110034.
UCRL-15597.
25546:053.
See NUREC/CR-3686, Summary abstract.
NUREO/CR-3686 VO2: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual. POWELL,G.H.; HOLLINGS,J.P.;
ROW,D.G.s et al.
California, Univ. of, Berkeley, C A.
June 1984.
182pp.
8407110090.
UCRL-15597.
25538:001.
See NUREC/CR-3686, Summary abstract.
NUREC/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual. POWELL,G.H.s HOLLINGS,J.P.s ROW,D.G.; et al.
California, Univ. of, Berkeley, CA.
June 1984.
84pp.
8407110076.
UCRL-15597.
20537:248.
See NUREG/CR-3686, Summary abstract.
NUREC/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual. POWELL,G.H.s HU,F-C.
California, Univ. of, Berkeley, CA.
- Lawrence Livermore National Laboratory.
June 1984 247pp.
8407110052.
UCRL-15597.
25537:001.
See NUREG/CR-3686, Summary abstract.
NUREG/CR-3687: LOOSE-PART MONITORING PROGRAMS AND RECENT OPERATIONAL EXPERIENCE IN SELECTED U.S.
AND WESTERN EUROPEAN COMMERCIAL NUCLEAR POWER STATIONS. KRYTER,R.C.
Oak Ridge National Laboratory.
April 1984.
60pp.
8405290447.
ORNL/TM-9107.
24711:030.
Technical personnel at thirteen nuclear power stations (ten in 53
tho U.S.A.
and three in Western Europe) were interviewed to escortain their collective experience with acoustic-based loose-part monitoring systems (LPMSs).
Subjects receiving special attention were the number and location of accelerometers required to reliably detect and locate loose parts in both pressurized-and boiling-water reactor typesi detection sensitivity to loose objects in both primary and secondary coolant loops; false alarm e x p erienc e; calibration procedures; day-to-day monitoring system operationi premature failure of in-containment components of the LPMS; and overall success to date in detecting the presence of potentially damaging loose parts and in assessing their operational and safety implications.
The individual utilities' responses to questions addressing these issues are provided, along uith the author's summary and in terpretation of what the information gathered means in a collective sense.
It is. concluded that the technology of loose part detection and assessment is moving slowly toward increased acceptance by the utility industry but, at the same time, the full potential benefits of loose-part monitoring systems are not presently being realized and, furthermore, probably will not be unless actions are taken in four recommended areas.
NUREC/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit 1 Nuclear Reactor. HUTTON, P. H. ; DAWSON,J.F.s FRIESEL,M.A.J et al.
Battelle Memorial Institute, Pacific Northwest Laboratories.
June 1984.
52pp.
8407020165.
PNL-5022.
25275:181.
Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report.
The report deals with background, methodology, and results.
The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws.
The sub Ject work demonstrated that anticipated problem areas can be overcome.
Work is continuing toward AE monitoring during reactor operation.
NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER PLANTS. HARTLEY,C.S.s LEVY,I.S.; FECHT,B.A.
Battelle Memorial Institute, Pacific Northuest Laboratories.
April 1984.
134pp.
8405220088.
PNL-4952.
24565:098.
The Pacific Northuest Laboratory has completed a project to identify human factors deficiencies in safety-significant control stations outside the control room of a nuclear power plant and to determine whether NUREG-0700, " Guidelines for Control Room Design Reviews," would be sufficient for reviewing those local control stations (LCSs).
The project accomplished this task by first, reviewing existing data pertaining to human factors deficiencies in LCSs involved in significant safety actions; second, surveying LCSs environments and design features at several operating nuclear power plants: and third, assessing the results of that survey relative to the contents of NUREG-0700.
The study 's conclusions are 1) a definitive list of safety-significant local control stations cannot be specified because power plant designs vary significantlys 2) most, if not all, local control stations have design deficiencies that could be corrected by applying human factors engineering principless and 3)
NUREG-0700 is generally applicable to LCSs but that guidance is needed to address the design of manually operated valves and the design 54
- roguiremonts of LCSs in extrsma environmsnt conditions.
Finally, the study recommends.an approach 'for: improving present LCSs to reduce the likelihood that operator error will occur.
NUREO/CR-3697: LABORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS. ELMORE,M.R.s HARTLEY,J.N.
Battelle Memorial Institute, Pacific Northwest Laboratories.. April 1984.
53pp.
8405220076.
PNL-5025.
24560:324.
Pacific Northwest Laboratory,. under contract to the U.S.
Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, is investigating techniques to control fugitive dust emissions from
-active uranium oill tailings piles.
This report describes-laboratory tests conducted to evaluate 45 commercially available enemical stabilizers.
Tests were conducted in a wind tunnel to evaluate the effectiveness and durability of the stabilizers under similar conditions.
The effects of application rate, temperature j
(freeze / thaw) cycling, wet / dry cycling, and wind speedcwere determined.
In addition, tests were conducted to determine the effects of ultraviolet light and. water erosion on the durability of the stabilizers.
Permeability tests were also conducted to determine the potential effect of each stabilizer on the overall stability-of the tailings pile.
Results of these laboratory tests indicated that 16.of the stabilizers were. equally effective and more durable than the others.
f.I NUREO/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS). THEDFANOUS.T.Gi; NOURB AKHSH, H. P. s GHERSON,P.; et al.
Purdue Univ., West Lafayette, I N.
May 1984.
214pp.
8406060373.
24669:022.
This report consists of two parts.
In Part I physically based 2
calculational models are proposed for predicting (a) conditions for stratification due to HPI in a circulating reactor. loop 1-(stratification model) and (b) cooldown transients due to HPI in a j
stagnated primary reactor fluid (thermal mixing model).
The integral 4
aspects of these models are confirmed by comparison to the CREARE 1/5-scale data.
In Part II the thermal mixing model is assessed in an integral as well as in a local sense by comparison to the first round of data from Purdue's 1/2-scale facility.
These data are the only 4
available large-scale data at this time and they are an important complement to CREARE's 1/5-scale results in constructing a basis for.
s scale-up to reactor conditions.
Facility construction, 4
instrumentation, data reduction techniques and detailed experimental results are also included in Part II.
j l
NUREO/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT i
FLUID-THERMAL MIX:NG IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT 1
USING SOLA-PTS. DALY,B.J.
Los Alamos Scientific Laboratory.
April 1984.
86pp.
8406230301.
LA-10039-MS.
25131:319, i
The SOLA-PTS code has been used to analyze transient fluid-thermal mixing in a 180 degree sector of the downcomer and a cold leg of the Calvert Cliffs-1 plant for three assumed accident scenarios.
The inlet boundary conditions for these calculations were obtained from mass flou rates and temperatures that were computed. in systems code studies.
The results of the three-dimensional SOLA-PTS i
calculations indicated that a pressurized thermal shock risk was mitigates for these accident scenarios as the result of the particular o
i circulation patterns that developed in the downcomer.
i N
i-
.e
.. ~ _
~.
= +
. NUREd/CR-3713: : GROUPINO OF: LIGHT WATER REACTORS FOR EVALUATION OF DECAY.
l i'
' HEAT REMOVAL CAPABILITY. KAROL,R.s FRESCO,A.'s1 PERKINS,K.R.
- Brookhaven National Laboratory.
June.1984.
82PP.
8407110022.
L~,
BNL-NUREG-51752.
25547:001.
This grouping report.provides a compilation of decay. heat removal systems (DHRS) data for-operating commercial-light water reactors.
-The reactors have:been divided into<12 groups based.on similarity of
- the.DHRS Jand related l systems as-part of the NRC Task Action Plan on-
' Shutdown Decay-Heat Removal Reg'irements.
u i'
.NUREG/CR-3718:-RELIABILITY ANALYSIS OF STIFF VERSUS. FLEXIBLE. PIPING
~
STATUS REPORTJ ~ LU,S.C.A CHOU,C.K.
Lawrence Livermore Natianal Laboratory.
April-1984.. ;44pp.
8404250006.
.UCID-19722.
24202:295.
A confirmatory piping. reliability assessment for stiff versus flexible. piping. systems indicated that removing rigid supports tends, in general,.to. reduce thermal stress buttto increase seismic stress in
_the pipe.
As a result, piping design can jbe made'more reliable by some. reduction'of. rigid supports.
We also observed.that piping design 4
Using' snubbers among. support devices.may not exhibit the intended reliability because snubbers-often fail to perform'the desired a
function.
.It was-demonstrated that certain piping' systems with
. snubbers removed actually exhibit higher reliability than do those of' 1
-the original design.
The Steering and Technical Committees on Piping Systems
],
established by the Presrure Vessel Research Committee (PVRC) have investigated changes to be implemented in Regulatory Guide (RG) 1.61 l
and RG 1.122 aimed,et more flexible piping design.
An independent
-impact assessment-conducted by this-project concluded that:
. (1) PVRC proposed. changes substantially reduce calculated piping responses (2) e
{
proposed changes allow piping. redesigns with significant reduction in
[
number of supports and. snubbers without violating ASME code I
requirementss and (3) the more flexible piping redesigns are capable
~
of exhibiting. reliability levels equal to or higher than the original stiffer design.
i NUREG/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN l
STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN. HOWARD, G. E.
ANCO l
Engineers, Inc.
April 1984.
4pp.
8404110013.
22995:304.
. This report consists of comparisons of prediction and experiment for German Standard Problem 4a, a blowdown experiment involving structural dynamic response.
The comparisons presented herein are of i
the time histories of displacen.ent, bending stress, and. bending axis angle.
The reasons for error in the predictions are discussed.
The i
t d
structural model is improved to obtain a better match with the experimental natural frequencies.
NUREC/CR-3722: DAMPING TEST RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND 8-INCH UNPRESSURIZED PIPES. WARE,A.G.s THINNES G.L.
EG&G, Inc.
May 4
1984.
~68pp.
8406070132.
EGG-2305.
24850:280.
EO&G Idaho is-assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final p osition on revised damping; values for structural analyses of nuclear piping j-
-systems.
As part of this program,-a series of. vibrational tests on j
unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to j
various parametric effects.
The 33-ft straight sections of piping were supported at the ends.
Additionally, intermediate supports e
i l
l j.
56 c
_---.___._-.__.-._-._._.~m_, _. _ _ _ _ _. _... -. _
l.
lC e
esmprising spr'ing, ro'd, and constant-force heng'ers, as w211 as a sway brace and snub 5ers, were used. ~ Excitation was provided by lou-force-level' hammer, impacts,'a hydraulic shaker, and a 50-ton overhead crane for snapbackitesting.
Data was recorded using acceleration,; strain,fandidisplacement time histories.
This report (presents' test results showing the effect of stress level and type of
. supports"on.structu'ral damping in piping.
I NUREG/CR-3725i ' NUCLEAR ' POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:Part I - The Need For Plant-Reference Simulatorss Part
'II - The Use Of Plant-Reference Simulators. RANKIN.W.L.s B ALTON, P. A. ;
-SHIKIAR,R.i et al.
Battelle Human Affairs Research Centers.
~May 1984.
126pp.
8405310090.
.PNL-5049.
24734:223.
Part I.of.this report presents technical Justification for the use'of plant reference simulators in the-licensing and training of nuclear.. power plant' operators and examines alternatives to,the use of plant-reference simulators.
The technical rationale is based on research on the use'of simulators in other industries, psychological learning and testing principles, expert opinion, and user opinion.
Strong? technical Justification exists for requiring plant-reference simulators.for operator licensing purposes.. Technical Justification for the use'of plant-reference. simulators for operator training is less well grounded empirically, although expert opinion is that plant-reference simulators,' when properly used result'in the most effective training.
.Part II discusses the central considerations in using plant-reference simulators for licensing examination of nuclear power plant operators and for incorporating simulators-into nuclear.
power plant training programs.
Recommendations are presented for the administration of simulator examinations in operator licensing that reflect the goal of-maximizing both reliability and validity in the examination process.
A series of organizational. tasks. that promote the acceptance, use, and effectiveness of simulator tr'aining as part of the onsite training program is delineated.
NUREO/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS:A COMPREHENSIVE BIBLIOGR'APHY WITH SELECTED ANNOTATIONS. BOLTON,P.A.s FAIGENBLUM,J.M.; HOPE,A.M.; et al.
Battelle Human Affairs Research Centers.
May 1984.
70pp.
8405300079.
PNL-4765.
24713:302.
This document contains a comprehensive bibliography on the topic of simulator fidelity and training effectiveness, prepared during the preliminary phases of work on an NRC-sponsored project on the Role of Nuclear Power Plant Simulators in Operator Licensing and Training.
Section A of.the document is an annotated bibliography consisting of articles and reports with relevance to the psychological aspects of simulators in a variety of settings, including military.
The annotated items 'are oraen from a more comprehensive bibliography, presented in Section B, listing documents treating the role of simulators in operator training both in the nuclear industry and elsewhere.
NUREG/CR-3727: FISSION PROLUCT REMOVAL IN ENGINEERED SAFETY FEATURE
.(ESF) SYSTEMS.' Data Base Assessment And Suggested Experimental Program. ZALOUDEK,F.R.; POSTMA,A.K.s WINEGARDNER,W.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984.
49pp.
8405220035.
PNL-5050.
24560:159.
The available data base of the fission product removal capabilities of nuclear reactor Engineered Safety Feature (ESF) 57
=.
-cyctcas was1rovicwad end essessed. -The systems considered included l pressure suppression pools,. ice condenser systems, containment sprays.
filter systems.and containment' air coolers.. Based.on'this assessment,
'a'research program was recommended to expand,this data base to' support
-the development of mechanistic models and. computer codes for the-prediction of ESF system fission product removal.
This research
_ program included experimental efforts.to better define the performance of ice ' condenser ' systems, expand the range of data available on water spray systems and to investigate the behavior of containment air coolers, demisters and fans in the presence of aerosols typical of those expected following a severe accident.
1'
. NUREC/CR-3740: J-INTEGRAL TEARING INSTABILITY ' ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE. VASSILAROS,M.G.s HAYS R. A. i OUDAS,J.P.s et al.
David W. ' Taylor Naval Research.& Development Center.
April 1984.
.100pp.
8404300225.
24254:135.
An experimental. investigation was performed to evalua te the applicability of using J-Integral tearing instability analysis to vescribe the fracture behavior of 8-inch (203 mm). diameter, nuclear grade, ASTM A106 steel pipe.
Pipe sections measuring 48-inches (1219 mm) in length and 8.60 inches-(219 mm) in diameter with i
circumferential fatigue precracks were loaded in four point bending using a_ variable compliance test arrangement.
J-Integral tests were performed on.1/2 T,.1 T,.and.2 T plan compact specimens machined from the pipe.. These J-Integral resistance curves (J1-R curves) were i
compared.to the J1-R curves from the pipe bend t e s t s.'
Two different J-Integral analyses were used to describe f rac ture behavior.
In one analysis, the material was modelled by assuming elastic-perfectly measurements of mechanical. response of the loaded s truc tur e including hardening of the steel.
The evaluation of the J-Integral tearing instability analysis was performed using J versus T plots of each test.
The results of the investigation indicate that compact j'
speciment JI-R curve test results appear to agree with the JI-R curves
'from full size pipe bend tests.
Further, J-Integral tearing instability analysis can accurately describe the ductile tearing behavior of 8-inch ASTM A106 steel pipe provided the actual load, displacement, crack length and hardening behavior is available, Additionally, the results indicated that such an analysis with assumed elastic fully plastic behavior appears to produce conservative
[
results.
i l
NUREO/CR-3741 VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS l
CAPABILITIES. Phase 2 Top ical Report, Volume 1: Data Evaluation.
COLEMAN,D.R.
Control Data Corp.
April 1984.
137pp.
8405220084.
24565:230.
The second phase of acquisition, review, analysis, and processing j
of power reactor fuel performance data resources is described in this i
report.
These data resources are characterized to support subsequent i
evaluations of the NRC-sponsored fuel rod behavior code, FRAPCON.
Application of the Fuel Performance Data Base is shown to provide the basic data files which are sorted, processed, and restructured to establish key paraneters of interest on an individual rod basis.
The
-design, operational, and performance parameters are analyz ed to determine the data populations and the representation of various fuel design types in the data sample.
Also presented are the performance data distribution and trends relative to operational parameters such as power and burnup, and description of the data processing methods.
Significant amounts of power reactor fuel performance data are 48
~.. -.
. - -. =
~-.
t l
fovoilchlaftoisupport'high burnup codo evaluation studios.
The data clearly L indicates.the. cumulative ef f ects of rod -def orma tion, - fission L+
gas release, and corrosion which tend to alter the1as-built fuel rod l
thermal and mechanical conditions.
The availab'le' data' reflect the
[
-current status of commercial fuel utilization in that' incumbent p
designsgare gradually being' replaced by high burnup designs, but the l
newer fuel types do not yet dominate the data sample.
l l
- NUREG/CR-3743: THE IMPACT OF NDE UNRELIADILITY ON PRESSURE' VESSEL FRACTURE. PREDICTIONS. SIMONEN,F.A.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May 1984.
28pp.
8406060417.
I PNL-5062.
24669:236.
This report reviews the significant variables of' flaw depth, l
' length location and orientation required for fracture mechanics
^
i evaluations.of pressure vessel integrity.
Results of calculations are presented which emphasize pressurized thermal shock (PTS) and the 1significanceHof flaws" located at or near the inside surface of the
{'
vessel.
For PTS conditions, previous studies have shown that vessel f ailure : probability is relatively insensitive to flaw depth.
In this
-study the impact of flaw length is also evaluated, indicating the
+
importance of fully characterizing all flaw dimensions by NDE.
Results of'other evaluations are presented, showing the importance of 4
accurately locating flaws by NDE.
The. influence of vessel cladding is j
emphasized, with the relative significance of flaws through the clad and at various depths below the clad being addressed.
4 NOREG/CR-3745: BIOLOGICAL CHARACTERIZATION OF RADI ATION EXPOSURE AND
{'
DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annua l Progress Report: April li1982 - March 31,1983. EIDSON,A.F.
Inhalation Toxicology Research Institute.
- Lovelace Biomed & Environmental Research Institute.
May 1984.
53pp.
8406010536.
LMF-108.
24764:175.
j A quantitative infrared absorption method for yellowcake allowed the fraction of ammonium diuranate in a mixture to be determined accurately within 7% and the U(3)O(8) fraction within 13%.
The composition of yellowcake from six operating mills ranged from nearly pure ammonium diuranate to nearly pure U(3)D(8).
A study of retention and translocation of uranium after subcutaneous implantation in rats l
uas done.
The results showed that 49% of the' implanted yellowcake cleared from the body with a half-time in the body of 0.3 days, and the remainder was cleared with a half-time of 11 to 30 day s.
Twenty dogs exposed to a more soluble yellowcake form inhaled a e r o s o l s.
l _
producing an estimated initial lung burden of 130 micrograms of U per i
kilogram of body weight.
Aerosols inhaled by dogs exposed to a less j
soluble yellowcake form averaged an estimated initial lung burden of i
140 micrograms of U per kilograms of body weight.
Biochemical indicators of kidney dysfunction that appeared in blood and urine 4 to i
8 days after exposure to the more soluble yellowcake showed j.
significant changes in dogs, but levels returned to normal ~by 16 days after exposure.
No biochemical evidence of kidney dysf unc tion was observed in dogs exposed to the less soluble yellowcake form.
NUREQ/CR-3749: COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5-TEST.
BIAN,S.H.; THURGOOD,M.J.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984, 102pp.
8405210570.
PNL-5065.
24529:208.
The computer code COBRA / TRAC was used to simulate a Small Break 50
m--
Less-of-Coolant Accident (SBLOCA) test performsd at the-Semiscale'
~ MOD-2A Test Facility operatsd by the Idaho National Engineering Laboratory.
The'results of the simulation were compared with the:
.results of the' actual _ test.
.The comparison showed that the code-has the capability.to mode 11small-break accidents in an integrated coolant n
system of a~ pressurized water reactor.(PWR).
y NUREC/CR-3749: COBRA-NC POST-TEST PREDICTIONS.FOR HDR CONTAINMENT STEAM BLOWDOWN TEST V44 (INTERNATIONAL: STANDARD PROBLEM 16). THURGOOD,M.J.
Battelle Memorial. Institute, Pacific.Northwsst Laboratories. ' May 1984.
127pp.
8405210604.
PNL-5066.
24535:001.
COBRA-NClis a digital computer program written in FORYRAN IV that sim'ulates.the response;of multicompartment light water reactor containment systems to: postulated loss-of-coolant accidents.
It has been used to perform, post-test predictions German project HDR containment system to a ~ of the response of the.
simulated steam line break' blowdoun: transient.. Predictions were made of compartment pressures, gas temperatures, structural temperatures, and differential pressures-between rooms.
In.this report,.these predictions are compared with the experimentally measured values...The agreement with the data is reasonable.
Improvements in s the prediction can be made by more-carefully modeling the flow openings between' rooms or by-using a finer
- mesh, i
NUREG/CR-3753: AN EVALUATION OF MANUAL ULTRASONIC INSPECTION OF CENTRIFUGALLY CAST STAINLESS STEEL PIPING. - TAYLOR, T. T.
Ba ttelle l
Memorial Institute, Pacific Northwest Laboratories.
May 1984.
22pp.
l 8406080265.. PNL-5070.
24877:094.
3 This work was perforced as a portion of a NRC research program entitled " Integration of Nondestructive. Examination and Fracture Mechanics" (FIN. B2289).
The NRC technical monitor is 'Dr.
Joe Muscara.
Two studies have attempted to provide an answer to the i
degree of inspectability of _ Centrifuga11g : Cast Stainless Steel (CCSS) pipe..One study was an NRC-sponsored. Pipe Inspection Round Robin (PIRL) test conducted at Pacific Northwest Laboratory (PNL).
Another
).
study was conducted by Westinghouse.
-The PNL study reported that less i
than 30% detection was achieved on thermal fatigue cracks ranging from i.
5% to 50%.through-wall.
The Westinghouse study reported that 80%
detection was achieved for 20% through-wall mechanical fatigue cracks.
A cooperative program between PNL~and Westinghouse was conducted to. resolve the differences between-the two studies.
The program was designed as a limited round robin.
The data reported here indicates that flaw type -(thermal fatigue versus mechanical fatigue)
)
was a significant factor in detection.
Mechanical fatigue cracks were
}
more easily detected than t h erma l fatigue cracks.
The data conclusively shows that manual ultrasonic inspection cannot size flaws in cast stainless steel pipe' be continued because some failure mechanisms (i.e.,
mechanical fatigue cracks) have proven to be 4
detectable.
t NUREC/CR-3754: FAILURE-EVALUATION OF GENERAL ELECTRIC SB-1 AND SB-9 REACTOR MODE SWITCHES.-BACANSKAS,V.P.
Franklin Institute / Franklin Research Center.
April 1984.
37pp.
8405220046.
24559:250.
As a result.of reactor mode switch malfunctions at operating nuclear power plants (IE Inf ormation Notice 83-42), the NRC. requested that the Franklin Research Center perform a failure evaluation of the 80 m
v,--
.m er
, N, -. r a,
,.,,n,-
.n,,o,,,,
-.-.,-.-.,,_,,,,-s
,,,.n.+--,,.--,-...rn,~,-,-.w,
-~,eu
,,,.,.,,,,n,
- -, - -, ~, - -,
~ - -. -. -
CE SB-il reac tor ~ mode suitch.
The objectives of the program were to identify the failure mechanisms for the SB-1 switch and determine:if the failure mechanisms l
'were the result of age-related, conditions, defects of a particular switchi or:~ design..-In addition,-the. vendor proposed SB-9 replacement switch was-evaluated for susceptibility to similar failure mechanisms.
~
The SB-1-reactor mode switch that malfunctioned at Guad Cities Unit 1 was evaluated along with new SB-1 and SB-9 switches.
The SB-1 reactor mode switch malfunctions were most probably the
.resultLof the switch being placed in a false detent position (an intermediate switch position just-prior to the actual deten position) which allowed several of'the' contacts-required to be closed to remain open.
The false detent noted in the SB-1 switch operation is a result of the indexer mechanism design and not age-related conditions or a defect of a particular suitch.
The inderer mechanism for'the'SB switch is of a different design and is not susceptible to a similar failure mechanism.
NOREG/CR-3755: STRONG GROUND MOTION STUDIES FOR SOUTH CAROLINA EARTHGUAKES. NUTTLI,0.W.
RODRIGUES. R. s - HERRMANN, R. B. i et al.
Stc Louis Univ., St. Louis, MO.
April 1984.
96pp.
8405020377.
UCRL-15394.
24297:026.
This report is concerned with estimating the strong ground motion that will result from damaging earthquakes that occur in South Carolina, varying in size from those that can produce only minor damage'to those as large as the 1886 event.
The report is divided into three parts.
Part I discusses acceleration, velocity and displacement modeling, using available observational data (accelerograms and non-strong motion seismographic) and response spectra obtained from those data.
Part'II uses MM intensity data for estimating strong ground motion.
Part III surf ace-wave focal mechanism studies of South Carolina earthquakes.
NUREG/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED '
STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
BERNREUTER,D.L.4 SAVY,J.B.s MENSING, R. W. s e t al.
Lawrence Liver. sore National Laboratory.
April 1984 542pp.
8405220095.
UCRL-53527.
24554:354.
The EUS Seismic Hazard Characterization Project (SHC) is the outgrowth of an earlier study performed as part of the U. S. Nuclear Regulatory Commission's~(NRC) Systematic Evaluation Program (SEP).
The objectives of the SHC are: 1) to develop a seismic hazard characterization methodology for the region east of the Rocky Mountains 4 and 2) the application of the methodology to ten sites to assist the NRC staff in their assessment of the implications in the clarification of the U.S.
Geological Survey (USGS) position on the Charleston earthquake.
As in the-SEP, the fundamental characteristics of the methodology used in SHC consists in using expert opinions for all the input data.
The most important improvement over the-methodology'used in the SEP led to an estimate of the distribution of the hazard rather than Just point estimates.
An important aspect of eliciting expert opinion consists in holding feedback meetings in order to fine tune the methodology and the input data.
At this-point, the feedback process has not been completed.
'Our methodology and preliminary input from the expert panels is presented.
Estimates of the hazard (PGA and spectral velocity) at ten representative sites are' discussed including 61
o scnsitivity analysis and,e comparison with-the SEP results at four
- s i t e s.'
NUREG/CR-3759: LIGHTNING STRIKE DENSITY FOR THE CONTIGUOUS UNITED j
STATES FROM THUNDERSTORM DURATION RECORDS. MACGORMANiD.~R.J MAIER,M.W.4 RUST,l W.L D.
Commerce, Dept. of, Natl. Oceanographic &
Atmospheric Administration.
May 1984.
52pp.
8406010535.
24764:227.
An improvedLlightning ground strike climatology; has b een obtained from thunderstorm duration data recorded by 450 air weather stations.
From lightning strike location data collected in Florida and Oklahoma, it was found.that strike' density could be estimated from thunderstorm duration by,the~ equation N(s) _= 0.054H(1.1), where N(s) is.the number
,of strikes per square kilometer and H is thunderstorm-duration in hours.
This relationship was applied to thunderstorm duration data from the aviation stations to obtain lightning strike density for the
~
contiguous United ~ States.
NOREC/CR-3762: IDENTIFICATION OF. EGUIPMENT AND COMPONENTS PREDICTED' AS 4
{.
' SIGNIFICANT CONTRIBUTORS TO SEVERE CORE DAMAGE. HEISELMANN,H.W.
EG&C, Inc.
May 1984'.. ~60pp.
8405290430.
EGG-2311.
24696:106.
The Nuclear, Regulatory Commission, Office of Nuclear ~ Regulatory Research, sponsored,the Equipment Gualification Research Program which performed a survey of applicable severe accident study reports to aid in. focusing the program efforts.
The objective of the survey was to identify, where possible, equipment and components that have been predicted to be significant contributors to high probability accident
. sequence resulting in severe core damage.
A survey of.the results of the survey is presented in Tables 1 and 2 of this report.
Future updates of this report are anticipated as applicable risk study reports become available.
NUREG/CR-3768; NEW' MADRID SEISMOTECTONIC STUDY: Activities During Fiscal I
Year.1982. BUSCHBACH,T.C.
St. Louis' Univ., St. Louis, MO.
April i
1984.
180pp.
8405220039, 24564:035.
4
.The purpose of the New Madrid Seismotectonic Study is to identify j
the earthquake mechanisms within a 2OO-mile radius of New Madrid, Missouri.
Fiscal year 1982 marked the beginning of geological and studies aimed-at;better definition of the. east-west trending fault systems --
the Rough Creek and Cottage Grove systems -- and the northuest-trending Ste. Genevieve faulting.
A prime ob Jec tive -is to-l deteroine the nature and history of faulting and to establish the relat69nship with that faulting and the northeast-trending faults of the Wahash Valley and New Madrid areas.
One question.to be answered i
is whether or not the 38th-Parallel Lineament decouples the structural features to.the north from those south of the lineament.
Th ere were 222. earth qua kes located by the Saint Louis University 1
microearthquake network in 1982.
In addition, an earthquake swarm occurred in north-central Arkansas, and more than 17,000 events were recorded there during the year.
A seismic surveying program in the Wabash Valley area was completed in.1982, and the acquired data are being processed.
Early interpretations suggest that;there is a trough filled.with bedded units that are apparently1 pre-Mt. Simon sediments or volcanics.
l Studies of recent-fault movement suggest that there may have been some Post-Pleistocene movement along the Kentucky River Fault Zone but
$2
s none along the Shawneetoun, Illinois Fault Zone.
Researchers at Washington University postulate the existence of a Precambrian rift extending northwest-southeast through the state of Missouri -- and beyond -- based.on subtle gravity anomalg patterns and digital image processing.
- NUREC/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE. STEARNS, R. G. s TOWE,S.K.s HAGEE,V.L.3 et al.
Vanderbilt Univ., Nashville, TN.
April 1984.
49pp.
8405020505.
24298:124.
Gravity surveys at various levels of detail have been made at approximately 1200 stations in the Reelfoot Lake region of northwest Tennessee and adjacent portions of Missouri and Arkansas.
Individual features were surveyed in detail.
At Reelfoot Scarp, six lines of stations having a 100-500 feet spacing with close elevation control were measured.
Anomalies on these lines are caused by near-surface geology (faulting, clay-filled channel of abandoned course of Mississippi River).
A survey of less accuracy discovered an anomaly along a - f ault at Henning in the Ripleg South Guadrangle.
In the Reelfoot Lake Region the area of abundant earthquake occurrence is related to the gravity anomaly pattern.
The. earthquake area is sharply limited on the South by an abrupt change in anomaly trends, and the earthquakes _ diminish in number at a similar change in trend to the north.
Some positive gravity anomalies appear to mark j
plutons where they coincide with positive magnetic anomalies.
Gravity is useful in the region as a main component in a combined geophysical search for faults.
The search at Henning was successful, i
using gravity, combined with earth resistivity, es the main search technique.
l NUREC/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT MATERIAL. DOMIAN,H.A.
Babcock & Wilcox Co.
Oak Ridge National Laboratory.
May 1984.
101pp.
8407020266.
25279:029.
Pieces of Type SA508-2 steel, specially tempered to produce a high-impact-transition temperature, were welded in the side walls of l
Intermediate Test Vessels V-7 and V-8.
These vessels are to be tested by the Oak Ridge National Laboratory (ORNL) in the Pressurized-Thermal-Shock (PTS) Project of the Heavy-Section Steel Technology (HSST) Program.
A comparable piece of forging taken from the same source and heat treated with the vessels was characterized for its mechanical properties to provide data for use in the PTS tests.
NUREG/CR-3773: VARIATION OF PLANETARY BOUNDARY LAYER DISPERSION PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS. HICKS,B.B.
- Commerce, j
Dept. o f, Natl. Oceanographic & Atmospheric Administration.
May 1984.
50pp.
8406190078.
25029:238.
Recent developments in surface boundary layer and planetary boundary layer meteorology 'are combined to evaluate the height dependency of the dispersion parameters standard deviation z and standard deviation y of the familiar Gaussian plume relationships.
l Recommendations are based on analyses of surface boundary layer data, such as are collected at industrial sites under existing NRC guidelines.
63
NURED/CR-3774 VO1: ALTERNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL ~
RADIOACTIVE WASTES. Task 1: Description of Methods And Assessment Of-Criteria. BENNETT, R. D. s MILLER,W.O.s WARRINER,'J.B.s et al.
Army,-
Dept. of, Army Engineer Waterways Experiment Station.
April 1984.
82pp.
8405220068.
24559:289.
The studg reported herein contains the results of Task i of a
-four-task studg. entitled " Criteria for Evaluating Engineered
' Facilities."
The overall ob Jective.of this study is.to ensure that the'criter.ia needed to evaluate five alternative low-level radioacive waste (LLW) disposalimethods are available'to the Nuclear Regulatory-
~ Commission'(bMC) and the Agreement States.
-The' alternative methods considered are belowground. vaults, aboveground vaults, earthmounded concrete bunkers, mined cavities,.and augered holes.
Each of these alternatives is either being used by other countries for low-level i
radisactive waste (LLW) disposal or is being considered by other countries or U.S.
agencies.
In this. report the p erf ormance requirements 'are listed, each alternative is described, the experience gained with its use is
+
discussed, and the performance capabilities of each method are addressed.
.Next, the existing 10 CFR Part 61 Subpart D criteria with
. respect to paragraphs 61.50.through 61.53, pertaining to site suitability, design,. operations and. closure, and monitoring are assessed for applicability to evaluation of each~ alternative.
Preliminary conclusions and recommendations are offered on each l
nothod 's suitability as an LLW disposal alternative, the applicabil'ity of the criteria, and the need for supplemental or modified criteria.
NUREG/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION. EISENHOWER,E.H.
Commerce, Dept. of, National Bureau of Standards.
June 1984.
163pp.
8407170566.
25632:001.
d 3
This report describes results of a three-year program that will enable the Nuclear Regulatory Commission to improve, demonstrate, and document traceability of its measurements to the national physical i
neasurement standards for ionizing radiation.
The principal actions i
taken were:
(a) characterization of the response of a thermoluminescence dosimetry system used for routine surveillance of j
nuclear facilitiess (b) characterization of the response of six models of portable survey instruments; and (c) implementation of routine quality assurance services that will demonstrate that laboratories a
which calibrate survey instruments for the NRC are sufficiently t
consirtent (in agreement) with national measurement standards.
Tests i
of the TLD system were performed as specified in American National Standard N545-1975, plus several additional tests not-contained in that document.
Measurement assurance tests were conducted for the NRC l
Region-1 laboratory.
The response of the survey instruments was determined for photon energies as high as 6.5 MeV, and for beta 3
i particles of various energies, including those emitted by (133)Xe gas.
The basic principles under which the long-range interactive MGA program will operate were developed and documented, and the
-feasibility of the program was demonstrated.
NUREC/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draft Report Of The USNRC PCI Review Group.
j MACDONALD P.E.
EG&G, Inc.
TOKAR, M. 's VAN HOUTEN,R.
NRC - No Detailed Affiliation Given.
June 1984.
112pp.
8407020360.
EQQ-2313.
25274:148.
Because fuel failure estimates are used as input to radiological t
dose calculations, the U.S.
Nuclear Regulatory Commission has formed a i-64
tock forco of fuol behevior experts to study pellot-cladding interaction (PCI), due to concerns that existing rod overheating criteria might be inadequate for evaluating transient severity in this regard.
This report includes preliminary. findings for reactor events of the type addressed by Chapter 15 of the NRC Standard Review Plan.
Specifically,.the BWR turbine trip without bypass, PWR control rod withdrawal error, subcritical PWR control rod withdrawal error. BWR control blade withdrawal error, and the PWR steamline break are analyzed,on the Joint bases of peak rod power, power increase, ramp rate, and duration at elevated power.
These Chapter 15 events are compared to numerous test reactor rosults and to other relevant investigations, and tentative conclusions on transient severity and data base adequacy are presented.
Progress in developing computer i
codes,for predicting PCI-induced fuel rod failure is also discussed.
NUREO/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON SHIFT.- OLSON,J.) SCHREIBER,R.E.; MELDER,B.D.
Battelle Memorial Institute, Pacific Northwest Laboratories.
May.1984.
61pp.
8406080258.
PNL-5087.
24865:181.
This report represents the conclusions of a project studying the role of engineering expertise on shift in nuclear power plants.
Using the present shift technical advisor (STA) position as the base case, several alternatives were analyzed.
On-shift alternatives include the STA, the shift supervisor (SS), and the shift engineer (SE).
The SE is degreed, experienced, trained and licensed as a Senior Reactor Operator.
Some non-shift alternatives were also studied.
These included a cadre of on-call engineers and specialists within continual contact and easy reach of the plant, a technical system of phone and data lines linking the plant with a facility similar to an on-site technical support center, and a safety parameter display system (SPDS) to augment technical upgrading of operator aids presently available.
Potential problems considered in the analysis of implementation of these alternatives included job content constraints. problems of crew acceptance, and problems.of labor supply and retention.
Of the considered alternatives, the SE and SS options appear superior to the current STA approach.
The SE option appears the easiest to implement and the most effective under varied plant conditions.
The SE may also serve as liaison to off-site support facilities.
NUREG/CR-3797: DIGMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS OR MIGRATION. SIMMONS,M.A.; SKALSKI,J.R.) SWANNACK,R.; et al.
Battelle Memorial Institute, Pacific Northwest Laboratories.
April 1984.
38pp.
8406040139.
PNL-5028.
24805:265.
DIGMAN is an interactive computer program which allows the user to sample a hypothetical waste site.
Using sample results, the user is then required to determine the area contaminated by a waste spill or migration.
The report contains instructions for running the program and a sample session to aid the novice user.
DIGMAN is programmed for an Apple II computer with a minimum of 64K RAM and one d isk drive.
A disk containing a copy of the program is available from the authors.
4 NUREC/CR-3800: REFCO-83 USER'S MANUAL. DELENE J.G.)
HERMANN, 0. W.
Oak Ridge National Laboratory.
June 1984 76PP.
8407110018.
ORNL/TM-9186.
25547:087.
The computer code REFCO-83 utilizes a discounted cash flow (DCF) 65
cnolysis proceduro to celculato botch, cyclo, cnd lifetia? lovolized nuclear fuel cycle costs.
This code is an updated version of the REFCO computer code originally written in the early 1970s.
The basic nethodology and-procedures were retaineds however, extensive modifications were made to the input and data handling procedures.
Several computational procedures were updated to make the code more versatile and to simulate recent events such as the provisions of the Nuclear. Waste Policy Act of 1982.
'This report is a user's guide for the revised REFCO code.
It contains a description of the code methodology, a cost data base, a discussion of the general code structure, the code input instructions, j
and sample cases.
NUREG/CR-3305: ENGINEERING CHARACTERIZATION OF GROUND MOTION.. Task I: Ef f ects OF Characteristics Of Free-Field Motion On Struc tural Response. KENNEDY,R.P.s SHORT,S.A.s MERZ,K.L.s et al.
Structural i
Mechanics Associates.
May 1984.
389pp.
8406210448.
25098:044.
i
.This report presents the results of the first task of a two-task study on the engineering characterization of earthquake ground motion for nuclear power plant design.
_The overall objective of this. study is to develop recommendations for methods for selecting design response spectra or acceleration time histories to be used to characterize motion at the foundation level of nuclear power plants.
Task I of the study, presented herein, develops a basis for selecting design response spectra, taking into account the characteristics of free-field ground motion found to be significant in causing structural damage.
Task II of the study, to be completed later in 1984, will provide recommendations for methods for selecting response spectra and time histories incorporating wave passage and soil-structure interaction effects and Task I results.
NUREO/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report January-March 1984. EDLER.S.K.
Battelle Memorial Institute, Pacific Northwest Laboratories.
June 1984.
35pp.
8407180005.
P NL-5106-1.
25665:287.
This document summarizes work performed by Pacific Northwest Laboratory from January 1 through March 31, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, U. S.
Nuclear Regulatory Commission.
Results from an instrumental fuel assembly irradiation program being performed at Halden, Norway, are reported.
Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions.
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.
Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho.
High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior.
Thermal-hydraulic models are b eing developed to provide better digital codes to compute the behavint of f ull-scale reactor systems under postulated accident conditions.
Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada.
NUREO/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Quarterly Report:
86
~ _ _
i i
October 1983 - March 1904.Vols 1 & 2..HUTTON,P.H.1: KURTZ, R. J.
Battelle Memorial. Institute, Pacific Northwest Laboratories.
June 1984.
47pp.
8407120539.
PNL-5125.
25556:187.
This report describes. technical progress on a. program to appig
. acoustic emission for continuous integrity surveillance of nuclear reactor pressure boundaries.
The period is October 1983-March.1984.
Test data from the completed intermediate scale vessel (ZB-1) test is 2
being analgred to isolate AE from crack growth for the purpose of refining AE signal identification and.AE interpretation methods.
Fatigue crack growth in the ZB-1 vessel is being characterized by destructive ~ examination.
Acoustic data obtained from the No. O inlet nozzle during hot functional testing at Watts Bar Unit 1 reactor i
i showed a source concentration.
A. cooperative ef f ort be twe en TVA and PNL is planned to evaluate the significance of the data.
' Identification of crack growth AE by pattern recognition is showing much' improved results.
Fatigue testing of A106B ferritic pipe material is showing mived AE results related to previous relationships developed for A3338 steel.
Development of an ASTM Standard Practice for continuous AE. monitoring of pressure boundaries has been initiated.
A NUREG document on results from AE monitoring at Watts 4
Bar, Unit 1 reactor during hot functional testing has been completed.
f NUREQ/CR-3838: AN INITI AL REVIEW OF SEVERAL METEOROLOGICAL MODELS SUITABLE FOR LOW-LEVEL WASTE DISPOSAL FACILITIES. CULKOWSK I, W. M.
Commerce, Dept. of, Natl. Oceanographic & Atmospheric Administration.
j June 1984.
21pp.
8407110180.
25536:296.
Several mathematical models of the meteorological asp ects of 5
}
effluent releases have been examined for Dames and Moore, Inc.,
Science Applications, Inc., Argonne National Laboratory, and Oak Ridge l
National Laboratory, contain provisions for various combinations of l
wind erosion, area, and point source configurations as wall as i
deposition and elevated releases.
Methods employed by these models are compared for. relevance, availability of supporting data and f
i potential benefit.versus cost, a
}
NURE0/CR-3839: AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR 1
A 6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
4 CAMPBELL K.W.
Lawrence Livermore National Laboratory.
June 1984.
66pp.
8407180329.
UCID-2OO83.
25654:203.
To help assess the impact of the current U.S.
Geological Survey position on the seismic safety of nuclear power plants in the Eastern United States (EUS), several techniques f or estimating near-source 4
strong ground motion for a Charleston size earthquake were evaluated.
The techniques for estimating the near-source strong ground motion for i
a 6.6 mb (7.5 Ms) in the Eastern United States which were assessed are methods based on (1) site specific analyses, (2) semi-theoretical j
scaling techniques, and (3) intensity-based estimates.
Each method different1g approaches the problem of estimating near-source strong 6
ground motions.
The results and limitations of each technique are j
discussed and recommendations-made to correct for bias in the methods.
Suggestions for future work are also presented.
NUREO/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA: A User 's Guide For The Applicable Computer Programs In The Statistical Pac kag e For Social Scientists (SPSS). AR I 30, R. s HOWE.S.E.s WEBB,T.s et al.
Brown j
Univ., 'rovidence, RI.
June 1984 39pp.
8407020174 25275:235.
Radiocarbon-dated pollen records are a source of quantitative 2
i 87
y,
-.., ~. - -
--.. - -. - ~
L obtimatos for climatic variablos.for tho(post 9000 years.
Multiple t
regression is'the main method for calculation.of these estimates and
~
!=
.reguires a series of steps to gain equations that meet the statistical assumptions,of tho' analysis, cThis manual describes these stepstwhich include (1) selection of the region for analysis,'(2) selection of the 4
pollen l types for1statistica1' analysis, (3). deletion-of univariate
,outliners, (4) transformation,to produce linear. relationships, (5) c selection of the regression eguation, and-(6), tests of.the regression F
- residuals.
The:grput commands'and the output from a series of SPSS (Statistical Package.for Social Scientists) programs are illustrated and, described, and, as an example, modern pollen,and climatic data from lower Michigan are'used to calculate a regres*1on equation for
. July mean temperature.
4 A
2 1
NUREQ/CR-3848: EXPERIMENTAL INVESTIGATION OF UNSTEADY-TORNADIC WIND LDADS ON STRUCTURES..JISCHKE, M. C. s MOSLEHI.F.
Oklahoma, Univ. of, 1
Norman, OK.
June-1984 34pp.
8407120632.
25556:233.
j Wa rd 's tornado simulator was used to model the effects of a i
tornado-like vorten; on cylindrical model structure.
The esperiment-i
[
was conducted at swirl. angles of O and 45 degrees.
Pressure j
coefficients were measured at different locations on the model for j
steady and unsteady cases, corresponding to situations where the relative velocity - between the vortex and model is rero and nonzero.
Results are presented in the forms of sectional pressure coefficient
}
profiles, and sectional. force coefficients.
Pressure profiles show j
that there are significant differences between the steady and unsteady
[
results.
. Translation of the model through the simulator produces a more symmetric pressure distribution,.and also results in a more l
substantial pressure drop on=the model.
I I
It is observed.that in a flow with' swirl angle at 45 degrees,
{
translation causes a significant increase in the horizontal sectional 4
force coefficient.
Outside of the core region.. translation causes an i
increase in the axial sectional force coefficient.
The formation or very low pressure regions over the top section~of the structure leads i
c to very strong axial force coefficients.
This may cause the failure j
to first appear on the roof, and then propagate throughout the
].
structure and cause total failure.
3 i
j NURE0/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL I
HEAT FLUX BOILING. TUZLA,K.s UNAL C.s BADR,0.A.s et al.
Lehigh Univ., Bethlehem, PA.
June 1994.
57pp.
8407060340.
TS-843.
l 25432:290.
]
This report describes the rod bundle post-CHF tests in progress j
and the test facility at Lehigh University.
The mechanical and.
j electrical design of the experimental facility and the iterative 7
j process used to arrive at the choices made for the design are i
described in detail.
The test facility consists of a nine (3 x 3) rod l
bundle in a sguare shroud which form the test section together with the hot patches at the top.and bottom ends.
The rods and the hot
.t patches are electrically. heated while the shroud is radiatively heated.
The test-section includes instrumentation to measure the 4
vapor superheat temperature and pressure drop upstream and downstream l
of a rod gap spacer.
This is the first application of the hot patch i
technique for generating post-CHF conditions in a rod bundle and thus-f quasi-steady-state tests are being thought of as a backup procedure for conducting these post-CHF heat transfer tests.
The test section is part of a wel1~ instrumented recirculating i
f loop.to generate the desired post-CHF conditions.
The other major
.t I
se i
i
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I
'y(
\\,,
w '-
, e %.
5
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- ccapenento'of tho hodt transfer loop includo tho surgo tenk,-purpo, boiler,. separation tartk and condenser.
The test facility-also includes a~ versatile one hundred channel data, acquisition system.
The
- mechanical and electrical components in the facility have been chosen to have sufficient accuracy to yield meaningfulsresults for the heat
~
transfer coefficient in the rod bundle under various post-CHF conditions. '.
.-w
,,a. '
I NUREC/CR-3875: THE USE OF 'IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS. SADIK,S.; ARENDTS.J.G.s DIXON,B.W.4 et al.
EGreG, Inc.
June 1984.
186pp.
8407180218.
c EGG-EA-6650.
25654:015.
l This report supports the Nuclear Regulatory Commission (NRC)
Unresolved Safety Issue A-46,
" Seismic Qualification of Equipment in Operating Plants.,"J The report is divided into four distinct
. sections.
Part A identifies the basic technical approaches for using in-situ test procedures as a tool in alternate methods for the seismic
~
qualification of equipment in operating plants.
Part D includes the development of imprhved methods of developing structural models using the results of in-situ procedures,'and predicting ~ structural response
- during seismic events using methods of random vibrations.
Thorough technical justification for these methods of analysis is provided to support the related guidance and acceptance criteria presented in Part C.
Part D contains.a cost estimate for using the various alternative nethods for seismic. qualification of equipment.
~
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Contractor Report Number index 2
This index lists, in alphabetical order, the contractor-issued report codes for the NRC contractor reports in this compilation. Each contractor code is cross-referenced to the NUREGICR for the report and to the 10-digit NRC Document Control System accession number.
SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER
=_
313-1162C NUREG/CR-2614 GP-R-212106 NUREG/CR-3606 813-1166 NUREG/CR-3218 HEDL-TME 83-22 NUREG/CR-3391 VO2 ANL-83-65 NUREG/CR-3504 HEDL-TME 83-23 NUREG/CR-3391 VO3 ANL-83-66 NUREG/CR-3505 HEDL-TME 84-14 NUREG/CR-3658 BMI-2113 NUREG/CR-3427 VO4 IEB-81-03 NUREG/CR-3054 BNL-NUREG-51581 NUREG/CR-2907 VO2 LA-10007-MS NUREG/CR-3644 BNL-NUREG-51691 NUREG/CR-3383 LA-10014-MS NUREG/CR-3650 DNL-NUREG-51733 NUREG/CR-3603 LA-10039-MS NUREG/CR-3704 BNL-NUREG-51735 NUREG/CR-3604 LA-9776-MS NUREG/CR-3305 BNL-NUREG-51737 NUREG/CR-3628 LA-9944-MS NUREG/CR-3567 BNL-NUREG-51738 NOREG/CR-3627 LMF-108 NUREG/CR-3745 BNL-NUREG-51740 NUREG/CR-3641 MEA-2017 NUREG/CR-3295 VO2 BNL-NUREG-51746 NUREG/CR-3664 MEA-2017 NUREG/CR-3295 VO2 BNL-NUREG-51752 NUREG/CR-3713 NEA-2028 NUR EG /C R-3506 EGG-2164 NUREC/CR-2531 RO2 MEA-2030 NUREG/CR-3546 EGG-2186 NUREG/CR-2691 ORNL/NSIC-2OO NUREG/CR-2OOO VO3 N3 ECG-2288 NUREG/CR-3596 ORNL/NSIC-2OO NUREG/CR-2OOO VO3 N4 EGG-2294 NUREG/CR-3633 VO1 ORNL/NGIC-2OO NUREG/CR-2OOO VO3 N5 EGG-2274 NUREG/CR-3633 VO2 ORNL/SUD-7576/2 NUREG/CR-3507 EGG-2294 NUREG/CR-3633 VO3 ORNL/TM-8517 NUREG/CR-2940 EEG-2295 NUREG/CR-3637 ORNL/TM-8664 NUREG/CR-3514 E"aG-2302 NUREG/CR-3360 ORNL/TM-8774 NUREG/CR-3303 EEG-2305 NUREG/CR-3722 ORNL/TM-8786 NUREG/CR-3410 EGG-2311 NUREG/CR-3762 ORNL/TM-8793 NUREG/CR-3335 EGG-2313 NOREG/CR-3781 DRFT ORNL/TM-8796/V4 NUREG/CR-32OO VO4 EIG-EA-6650 NUREG/CR-3975 ORNL/TM-8849/V3 NUREG/CR-3422 VO3 EI-1077 NUREG/CR-3489 ORNL/TM-8929 NUREG/CR-3535 EPRI NP-3546 NUREG/CR-3504 ORNL/TM-8939 NUREG/CR-3572 EPRI NP-3547 NUREC/CR-3505 ORNL/TM-8942 NUREG/CR-3515 F-C5896-OO2 NUREG/CR-3754 ORNL/TM-8964 NUREG/CR-3539 71 1
SECCNDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER ORNL/TM-9008 NUREG/CR-3595 S-762-R NUREG/CR-3583 ORNL/TM-9011 NUREG/CR-3600 SAND 82-0342 NUREG/CR-2552 ORNL/TM-9029 NUREG/CR-3652 SAND 82-0904 NUREG/CR-2679 VO4 ORNL/TM-9041/V1 NUREG/CR-3626 VO1 SANDB2-1145 NUREG/CR-2921 ORNL/TM-9088 NUREG/CR-3672 SAND 82-2475 NUREG/CR-3023 ORNL/TM-9107 NUREG/CR-3687 SAND 83-OO74 NUREG/CR-3134 I
PNL-4241 ORNL/TM-9186 NUREC/CR-3900 SAND 83-1118 NUREG/CR-33OO VO1 PNL-4138 NUREG/CR-2803 SAND 83-1149 NUREG/CR-3310 NUREG/CR-2675 VO4 SAND 83-1154 NUREG/CR-3316 PNL-4550 NUREG/CR-2955 SAND 83-1171 NUREG/CR-3329 VO4 PNL-4705-3 NUREG/CR-3307 VO3 SAND 83-1350 NUREG/CR-3366 PNL-4705-4 NUREG/CR-3307 VO4 SAND 83-1466 NUREG/CR-3378 1
PNL-4765 NUREG/CR-3726 SAND 83-2086 NUREG/CR-3511 VO3 PNL-4878 NUREG/CR-3533 SAND 83-2365 NUREG/CR-3624 PNL-4909 NUREG/CR-3564 SAND 83-2406 NUREG/CR-3588 PNL-4911 NUREG/CR-3566 SAND 83-2549 NUREG/CR-3608 PNL-4927 NUREG/CR-3632 SAND 83-2622 NUREG/CR-3623 PNL-4933 NUREG/CR-3350 SAND 83-2651 NUREG/CR-3629 PNL-4941 NUREG/CR-3613 SAND 83-2652 NUREG/CR-3630 PNL-4952 NUREG/CR-3696 SAND 83-7114 NUREG/CR-3379 PNL-4990 NUREG/CR-3682 SAND 83-7463 NUREG/CR-3653 PNL-4991 NUREG/CR-3683 SAND 84-OO40 NUREG/CR-3639 PNL-5003 NUREG/CR-3669 SAND 84-0140 NUREG/CR-3684 PNL-5006 NUREG/CR-3670 SAND 84-0178 NUREG/CR-3673 PNL-5016 NUREG/CR-3677 TS-843 NUREG/CR-3849 PNL-5022 NUREG/CR-3693 UCID-19722 NUREG/CR-3718 PNL-5025 NUREC/CR-3697 UCID-2OO83 NUREG/CR-3839 PNL-5028 NUREG/CR-3797 UCRL-15594 NUREG/CR-3755 PNL-5049 NUREG/CR-3725 UCRL-15597 NUREG/CR-3686 VO!
PNL-5050 NUREG/CR-3727 UCRL-15597 NUREG/CR-3686 VO4 PNL-5062 NUREG/CR-3743 UCRL-15597 NUREG/C R-3686 PNL-5065 NUREG/CR-3748 UCRL-15597 NUREG/CR-3686 VO3 PNL-5066 NUREG/CR-3749 UCRL-15597 NUREG/CR-3686 VO2 PNL-5070 NUREG/CR-3753 UCRL-53486 NUREG/CR-3476 PNL-5072 NUREG/CR-3681 UCRL-53527 NUREG/CR-3756 PNL-5087 NUREG/CR-3785 PNL-5088 NUREG/CR-2424 VO2 PNL-5088-1 NUREG/CR-2424 VO1 PNL-5106-1 NUREG/CR-3810 VO1 PNL-5125 NUREG/CR-3825 VO1-02 72
Personal Author index This index lists the personal authors of NRC staff and contractor reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by that author. If further information is needed, refer to the main citation by the NUREG number.
3 ACKERMANN,G.R.
NUREG/CR-3488 VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
ADAMS,R.E.
NUREG/CR-3422 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM.Guarterly Progress Report For July-September 1983.
AGE,R.E.
NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
AGUILAR,F.
NUREG/CR-3360: COMPUTER PROGRAM CDCID: AN AUTOMATED QUALITY CONTROL PROGRAM USING CDC UPDATE.
ALBA,C.
NUREC/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
ALDRICH,D.C.
NUREC/CR-2552: CRAC2 MODEL DESCRIPTION.
NUREC/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
ALDRIDGE,T.L.
NUREG/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS.
ALEXANDER,D.H.
NUREC/CP-OOS2: NRC NUCLEAR WASTE MANAGEMENT' GEOCHEMISTRY
'83.
ALPERT,D.J.
NUR EG/CR-2552: CRAC2 MODEL DESCRIPTION.
ALTMAN,W.
NUREG-1055: IMPROVING GUALITY AND THE ASSURANCE OF GUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCI AL NUCLEAR POWE'i PLANTS. A Report To Congress.
ANDERSON,F.D.
NUREG-1058: TECHNICAL SPECIFICATIONS FOR CALLAWAY PLANT, UNIT NO.
1.
Docket No. STN 50-483.(Union Electric Company)
ANKRUM,T.
NUR EG-1055: IMPROVING GUALITY AND THE ASSURANCE OF GUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCIAL NUCLEAR POWER PLANTS.A Report To Congress.
ANTONNEN,G.
NUREC/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY' DESIGN - TUFF.
73
ARELLANO,F.E.
NUREG/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
NUREG/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANI A-COATED STEEL.
ARENDTS,J.G.
NUREG/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EGUIPMENT IN CURRENTLY OPERATING PLANTS.
ARIGO,R.
NUREG/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA: A User 's Guide For The Applicable Computer Programs In The Statistical Pac kag e For Social Scientists (SPSS).
ATTERIDGE,9.G.
NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
AXELSON,W.
NUREG-1028: RUPTURED CESIUM-137 WELL-LOGGING SOURCE AT SHELWELL SERVICES,INC., HEBRON, OHIO.
BACANSKAS V.P.
NUREG/CR-3754: FAILURE EVALUATION OF GENERAL ELECTRIC SB-1 AND SB-9 3
REACTOR MODE SWITCHES.
BADR,0.A.
NUREG/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
BALTON,P.A.
NUREG/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING: Part I - The Need For Plant-Reference Simulators; Part II - The Use Of Plant-Reference Simulators.
BARKS,D.B.
NUREG/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
BARTTER,W.D.
NUREG/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS)
MODEL:
SUMMARY
DESCRIPTION.
BEARE,A.N.
NUREG/CR-3515: SAFETY-RELATED OPER ATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
BELL,A.J.
NUREG/CR-3632: METHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING PROCEDURES.
BENKOVITZ,C.
NUREG/CR-2907 VO2: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Ann'Ja1 Report 1981.
BENNETT,R.D.
NUREG/CR-3774 VO1: ALTEHNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTES. Task 1: Description of Methods And Assessment Of Criteria.
BERGERON,K.D.
NUREG/CR-3310: TESTING OF THE CONTAIN CODE.
BERNREUTER.D.L.
NUREG/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
BERRY,D.L.
NUREG/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC SAFETY STUDY: PLANT ANALYSIS.
BIAN,S.H.
NUREC/CR-3748: COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
BIRCHARDeG.F.
NUREG/CP-OOS2: NRC NUCLEAR WASTE MANAGEMEN1 GEOCHEMISTRY
'83.
BLOND,R.M.
NUREG/CR-2552: CRAC2 MODEL DESCRIPTION.
74
~-
1 i
BLOSE,R.E.
NUREG/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
BLUHM,D.
NUREG/CR-3653: CONTAINMENT ANALYSIS JECHNIGUES. A State-Of-Th e-Art Summary.
BOCCIO,J.L.
NOREG/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
=BOLTON,P.A.
i NUREG/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A COMPREHENSIVE 3IBLIOGRAPHY WITH SELECTED ANNOTATIONS.
BORELLA,H.M.
NUREG/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING. TECHNIQUES FOR p
OBTAINING SITE CHARACTERISTIC INFORMATION.
BRACH,W.
NUREG-1055: IMPROVING GUALITY AND THE ASSURANCE OF GUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCIAL NUCLEAR POWER PLANTS. A Report 3
To Congress.
E BRISBIN N.L.~
NUREG/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC SAFETY STUDY: PLANT ANALYSIS.
BRUEMMER,S.M.
NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED i
-STAINLESS STEEL FOR LWR SERVICE. Annual Rept For 1983.
BRUSKE S.Z.
(
NUREG/CR-3596: SEVERE ACCIDENT SEGUENCE ANALYSIS (SASA) PROGRAM SEQUENCE EVENT TREE: BOILING WATER REACTOR ANTICIPATED TRANSIENT WITHOUT SCRAM.
i BURGY,D.
i NUREG/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATABASE: SEEK SYSTEM. (Users Manual),
i BURKE,R.P.
NUREC/CR-2552: CRAC2 MODEL DESCRIPTION.
NUREG/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
BURKETT,M.W.
NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
BURNS,T.J.
NUREG/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT
}
RISK.
BUSCHBACH,T.C.
g:
NUREC/CR-3768: NEW MADRID SEISMOTECTONIC STUDY: Activities During Fiscal i
Year 1982.
I BUSTARD,L.D.
NOREC/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDtlRES ON CROSS-LINKED POLYOLEFIN CABLE'S PERFORMANCE.
j NUREG/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
BUTLER,T.A.
NUREC/CR-3644: REVIEW 0F PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS.
4 BUXTON,L.D.
NUREC/CR-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH l
UPPER-PLENUM INJECTION.
CADWELL,L.L.
NUREG/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
CALLAWAY,J.W.
NUREG/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
l CAMPBELL K.W.
NUREG/CR-3839: AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR l
75
?
.. _.. _ ~.., _..
i A'6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
CARLIN,F.
iNUREG/CR-3629: 'THE EFFECT OF THERMAL.AND IRRADIATION AGING SIMULATION PROCEDURES ON-POLYMER PROPERITIES.'
.CARLSON,D.D.
NUREG/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC.
SAFETY STUDY: PLANT ANALYSIS.
- C ATE J.~ H. '
NUREG/CR-3488 VO2: IDAHO. FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
.CHA,B.K.
NOREO/CR-3505: A VOLUME-WEIGHTE,D SKEW-UPWIND DIFFERENCE' SCHEME'IN
' C OMMI X.-
)
CHAMNESS,M.
NUREG/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DONAL SALT.
CHANG,M.T.-
NUREC/CR-3641: RELIABILITY ASSESSMENT OF INDIAN POINT UNIT 3 CONTAINMENT STRUCTURE.
CHEN,B.C.
NOREG/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN -
COMMIX.
~CHEN,F.F.
NUREG/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN i:
COMMIX.
i.
CHEN,J.C.
NUREC/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL I-HEAT FLUX BOILING.
CHEN,P.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF' PIPING SYSTEMS. Part A - User 's Manual.
l_
NUREO/CR-3686 VO2: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NUREC/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Prog rammer 's Manual.
CHENG,H.S.
NUREO/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD. O CYCLE 4:
A COMPUTER CODE'WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
CHENION,J.
NUREO/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
CHER F.F.
}
NUREO/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
CHOU C.K.
NUREG/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING -
STATUS REPORT.
CHUNG,D.H.
NUREO/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED
- STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
COHEN,L.
NUREG-0837 VO3 NO4: ' P#tC TLD DIRECT R ADIATION MONITORING NETWORK. Progress Report September-December 1983.
COLEMAN,D.R.
NUREG/CR-3741 VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 2 Topical Report Volume 1: Data Evaluation.
COLES.B.L.
-NUREO/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
COLLINS,J.L.
NUREC/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST
'HI-4.
76 I
~.
- C00TELLO,F.
.NUREC-08371VO3 NO4: NRC TLD DIRECT RADIATION MONITORING
. NETWORK. Progress Report September-December 1983.
CRISTY,M.
. NUREO/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED
-BONE-SEEKING RADIONUCLIDES.
CRONIN,F.J.
NUREG/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
. CULKOWSKI,W.M.
NUREO/CR-3838: AN INITIAL REVIEW OF SEVERAL METEOROLOGICAL MODELS SUITABLE FOR LOW-LEVEL WASTE DISPOSAL FACILITIES.
-CULLEN,W.H.
I NUREC/CR-3546:. THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES I-.
OF A 351 CF8A CAST STAINLESS STEEL IN LWR ENVIRONMENT.
CURRIE,J.W.
NUREC/CR-3566: SOCIOECONOMIC CONSEGUENCES OF NUCLEAR REACTOR ACCIDENIS CZAJKOWSKI,C.J.
NUREG/CR-3604: BOLTING APPLICATIONS.
DALY,B.J.
NUREG/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT i-FLUID-THERMAL. MIXING 'IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT i
USING SOLA-PTS.
DAWSON,J.F.
NUREC/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts.Dar Unit 1 Nuclear Reactor.
}
DAYAL,R.
3 NUREC/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF i
RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
DEEDS,W.E.
NURE0/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
i DELENE,J.G.
j NUREG/CR-3800: REFCO-83 USER'S MANUAL.
l DIAMOND D.J.
NUREG/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD. O CYCLE 4:
1 A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR j
SYETEM TRANSIENTS.
DICKSON,C.R.
l NUREG/CR-3488 VO2:. IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
i DIXON,B.W.
NUREC/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC QUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
3 DOBRANICH,D.
-NURE0/CR-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH j
-UPPER-PLENUM INJECTION.
DODD,C.V.
I NOREG/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING 1
PROGRAM ANNUAL PROGRESS REPORT' FOR PERIOD ENDING DECEMBER 31, 1983.
DODOE,C.J.
I
' -IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF NUREC/CR-3383:
RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
DOMANUS,H.M.
NUREC/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
NUREC/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
DOMIAN H.A.
NURE0/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT MATERIAL.
DUDA.L.E.
NUREG/CR-3316: VERIFICATION AND FIELD COMPARISON OF THE SANDIA f
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. NUREG/CR-3378: VERIFICATION OF THE. NETWORK FLOW AND TRANSPORT /DISTRIBUTEU VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
h DUNNING,D.E.
i
! NUREC/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED l-BONE-SEEKING RADIONUCLIDES.
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EBERHARDT,L.E.
NUREO/CR-2675 VO4: RELEVANCE OF BIOTIC ' PATHWAYS TO THE LONG-TERM L
REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
' ECKERMAN, K. F.
NUREG/CR-3535: AGE-DEPENDENT' DOSE-CONVERSION FACTORS FOR SELECTED BONE-SEEKING RADIONUCLIDES.
NUREG/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR HLW DOSIMETRY'(ICRP-30),I.
EDLER,S.K.
j
, NUREC/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS.'Guarterly Report July-September 1983.
NUREG/CR-3307 VO4: REACTCR SAFETY RESEARCH PROGRAMS. Guarterly Report October-December 1983.
NUREO/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly J Report January-March 1984.
EIDSON,A.F.
NUREC/CR-3745: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress
' Report: April 1,1982 - March 31,1983.
EISENHOWER,E.H.
NURE0/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION.
ELLINOWOOD,B.
NUREG/CR-3628: PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT STRUCTURES.
ELMORE,M.R.
NUREG/CR-3697: LABORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
EMEIGH.C.W.
NUREO-1065: ACCEPTANCE CRITERIA FOR THE LOW ENRICHED URANIUM REFORM AMENDMENTS.
ERASLAN,A.H.
NUREG/CR-3410: CHMONE:A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
ESTES,J.E.
NURE0/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR OBTAINING SITE CHARACTERISTIC INFORMATION.
EVANS,D.D.
NUREC/CR-3680: RELATIONSHIP BETWEEN THE GAS CONDUCTIVITY AND GEOMETRY OF A NATURAL FRACTURE.
EYLER,L.L.
NUREO/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF ' A PRESSUR IZED WATER REACTOR.
FABRY,A.
NUREO/CR-3391 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
FAIGENBLUM J.M.
NUREO/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A COMPREHENSIVE BIBLIOGRAPHY WITH SELECTED ANNOTATIONS.
FANDUS,F.
NUREG/CR-3653: CONTAINMEN T ANALYSIS TECHNIGUES. A Sta te-Of-Th e-Art Summary.
FECHT,B.A.
78
NURE3/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESICN OF
' LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER PLANTS.
FINDLEY,D.
NUREG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
FISCHER,S.K.
l NUREG/CR-3410: CHMONE: A DNE-DIMENSIONAL' COMPUTER CODE FOR SIMULATING l
TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
FISCHHOFF,B.
NUREG/CR-3507:.AN. ANALYSIS OF THE NRC SAFETY GOALS FOR NUCLEAR POWER.
FOLEY,W.J.
NUREG/CR-3054: CLOSEOUT OF IE BULLETIN 91-03: FLOW BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.-
(MUSSEL).
FOTIAS,A.
t NUREG-0980: NUCLEAR REGULATORY LEGISLATION.
FRANK.L.
' NUREG-1063: STEAM GENERATOR OPERATING EXPERIENCE UPDATE 1982-1983.
[
FRESCO,A.
NUREG/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY 4
HEAT REMOVAL CAPABILITY.
FRIESEL,M.A.
l NUREG/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit 1 Nuclear Reactor.
a FRY,D.N.
l NUREG/CR-3303: USE OF NEUTRON NOISE FOR DI AGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
FULLWOOD,R.R.
NUREG/CR-3682: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation I
of Existing Methods.
GAUSSENS,G.
l NUREG/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
GEFFEN,C.A.
NUREG/CR-3683: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Program Summary Through Fiscal Year 1983.
GHERSON,P.
{
NUREG/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH l
APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
I GIDO,R.G.
l NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR C AVITY.
}
SUBCOMPARTMENT ANALYSES.
GINZBURG,T.
1 NUREC/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
j GOMER,F.E.
NUREG/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
CONANO,L i
NUREG/CR-3218: EVALUATION OF-ENGINEEHING ASPECTS OF BACKFILL PLACEMENT l
FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final
{
Report (Task 5) June 1981 - February 1983.
I COTTULA,R.C.
NUREG/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL j
HEATER ROD DESIGN ON QUENCH BEHAVIOR.
GR AY, L. H.-
4 NUREG' CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR
/
1 DEVELOPING CRITERIA.
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GRAZULIS,T.P.
NUREG/CR-3670: VIOLENT TORNADO CLIMATOGRAPHY, 1880-1982.
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t GREENE,S.R.
NUREO/CR-2940:-REALISTIC' SIMULATION OF SEVERE ACCIDENTS IN
~
- BWRS-COMPUTER MODELING REQUIREMENTS.
GREENHOLT,C.J.
NUREC/CR-2921i CHEMICAL ~ INTERACTIONS OF TELLURIUM VAPORS WITH' REACTOR GREIMANN,L.
MATERIALS.
.NUREC/CR-3653: ' CONTAINMENT ANALYSIS TECHNIGUES. A State-Of-Th e-Art 1
Summary.
GRUNDL.J.A.
NUREC/CR-3391 - VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY I
. IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
GUDAS,J.P.
NUREO/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR-8-INCH 4
DI AMETER ASTM A106 STEEL PIPE.
GUNDERSEN, G. E..
l NUREC-1065: ACCEPTANCE CRITERI A FOR THE LOW ENRICHED URANIUM REFORM j
AMENDMENTS.
HAAS,P.M.
NUREG/CR-3626 VO1: MAINTENANCE, PERSONNEL PERFORMANCE SIMULATION 1MAPPS)
MODEL:
SUMMARY
DESCRIPTION.
~
l HAGEE,V.L.
NOREC/CR-3769:. DESCRIPTION AND SIGNIFICANCE.0F THE GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
l HALL,R.E.
j-NUREG/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
HAMBLEY,D.F.
4 NUREC/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
HARDY,H.A.
{
NUREC/CR-2531 RO2: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
HARRIS,JJC.
NURE0/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL j
TESTINC. Watts Bar. Unit 1 Nuclear Reactor.
HARRISON,F.L.
i NUREG/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF COPPER.
HARTLEY,C.S.
NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER PLANTS.
HARTLEY,J.N.
i NUREC/CR-3697: LABORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
i HAWTHORNE,J.R.
NUREC/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE
- DOSIMETRY. IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B & A533-B Ref erence Plates From PSF Simulated Surveillance & Through-Wa11' Irradiation Capsules.-
NUREO/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
l Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
1 HAYS R.A.
l NURE0/CR-3740: J-INTEGRAL' TEARING INSTABILITY ANALYSIS FOR 8-INCH
' DIAMETER ASTM A106 STEEL PIPE.
HEGER,A.S.
NUREO/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA 1
a s
~CYSTEM.
HEISELMANN,H.W.
NUREO/CR-3762: IDENTIFICATION -OF EQUIPMENT AND COMPONENTS PREDICTED AS SIGNIFICANT CONTRIBUTORS -TO SEVERE CORE DAMAGE.
HENNICK,A.
NUREO/CR-3054: CLOSEOUT OF IE BULLETIN 81-03: FLOW' BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.
(MUSSEL).
HENSLEY,W.K.
-NUMEG/CR-3669: PLUTONIUM RECYCLE TEST REACTOR'(PRTR) ACCIDENT: A FINAL
' REPORT ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
- HERMANN,0.W.
NUREC/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT
- R ISK.
-NUREG/CR-3800:.REFCO-83 USER'S MANUAL.
HERNAN,R.
NUREG-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REGUIREMENTS ON CPERATING PLANTS ~ DESIGNED BY BABCOCK AND WILCOX.
- HERRMANN R.B.'
NUREO/CR-3755: STRONG GROUND MOTION STUDIES FOR SOUTH CAROLINA EARTHGUAKES.
HETRICK,D.M.
NUREQ/CR-3410: CHMONE: A ONE-DIMENSIONAL COMPUTER ' CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL' CONCENTRATIONS IN WATER BODIES.
HICKS,B.B.
NUREG/CR-3773: VARIATION OF PLANETARY BOUNDARY LAYER DISPERSION PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS.
HILL J.R.
NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED P ARAMETR IC' EMPIRICAL' BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.
HILL.M S.
NUREG/CR-3134: A SETS USER'S MANUAL FOR VITAL AREA ANALYSIS.
HISER,A.L.
NUREC/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL. SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility.& Fracture Toughness Degradation of A302-B & A533-B Reference Plates From PSF Simulated Surveillance & Through-Wall Irradiation Capsules.
NUREG/CR-3506: J-R CURVE CHARACTERIZATION OF ' IRRADIATED LOW UPPER SHELF-WELDS.
HOFMANN.R.
NUREO/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN-TUFF; HOLLINGS J P.
NUREC/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NUREC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theorg Manual.
NUREC/CR-3696. VO3: WIPS--COMPUTER CODE FOR WHIP AND' IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual.
HOPE,A.M.
NURE0/CR-3726: ' SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A COMPREHENSIVE BIBLIOGRAPHY WITH SELECTED ANNOTATIONS.
HOWARD,G.E.
NURE0/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
HOWE. S. E.-
NUREO/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA: A User 's Guide For 81
Tho Applicablo Camputer Progrens In The Statistical Pac kago For Social _ Scientists _(SPSS).
HU,F-C.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR - WHIP AND IMPACT ANALYSIS OF P IPING SYSTEMS.' Part A - User 's Manual.
.NOREC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF i
PIPING SYSTEMS.Part B -' Theory Manual.
NUREO/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF -
PIPING SYSTEMS.Part C - Programmer's Manual.
NUREC/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
HUKARI,N.F.
NUREO/CR-3488 VO2: IDAHO FIELD EXPERIMENT 1981..Vol 1: Measurement Data.
HUTTON,P.H.
NUREG/CR-3693:. ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit 1 Nuclear Reactor.
NUREO/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Guarterly Report:
October 1983 - March 1984.Vols 1 & 2 HWANG H.
NUREG/CR-3641: RELIABILITY ASSESSMENT OF INDIAN POINT UNIT 3 CONTAINMENT STRUCTURE.
IDAR E.S.
- NUREO/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
IDRISS.I.M.
NUREC/CR-3805: ENGINEERING CHARACTERIZATION OF QROUND MOTION. Task I: Effects Of Characteristics of Free-Field Motion On Struc tural Response.
IMAN,R.L.
NUREG/CR-3624: A FORTRAN 77 PROGRAM AND USER'S GUIDE FOR THE GENERATION OF LATIN HYPERCUBE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
IMHOFF,K.L.
NUREO/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
IYER.K.
NUREG/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
JISCHKE,M.C.
NUREO/CR-3848: EXPERIMENTAL INVESTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES, JOHNSON,J.D.
NUREO/CR-2552:. CRAC2 MODEL DESCRIPTION.
JONES.K.
NUREO/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
JONES,R.
NUREO/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN-TUFF.
JOYCE.J.A.
NUREO/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE.
JU,F.D.
NUREO/CR-3644: ' REVIEW OF PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS.
KADAMBI,N.P.
'NUREO-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REGUIREMENTS ON OPERATINO PLANTS DESIGNED BY BABCOCK AND WILCOX.
KALKWARF,D.R.
NUREO/CR-3533: RADON ATTENUATION HANDBOOK FOR URANIUM-MILL TAILINGS COVER DESIGN.
82
KAM,F.B.
NUREG/CR-3391 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
KAROL,R.
NUREG/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY.
KAWAKAMI,J.
NUREQ/CR-3641: RELIABILITY ASSESSMENT OF INDI AN POINT UNIT 3 CONTAINMENT STRUCTURE.
KEMPPAINEN,M.
L NUREC/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES OF A 351 CFSA CAST STAINLESS STEEL IN LWR ENVIRONMENT.
KENDORSKI F.S.
NUREC/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
KENNEDY,R.P.
NUREG/CR-3805: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task I:Ef fects Of Characteristics Of Free-Field Motion On Struc tural Response.
KENNEDY,W.E.
NUREQ/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REQULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
KHATIB-RAHBAR NUREQ/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-38 MOD. O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
KIM,K.H.
NUR EQ/CR-3410: CHMONE: A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE. FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
KIRKLAND,0.L.
NUREO/CR-3514: THE CHEMICAL BEHAVIOR OF IODINE IN AGUEDUS SOLUTIONS UP TO 150 C.An Experimental Study of Nonredon Conditions.
KLEPPE.J.
NUR EO/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
KMETYK,L.N.
NUREO/CR-3608: RELAPS ASSESSEMENT: LOFT Larg e Break L2-5.
KNEE.H.E.
NUREG/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS)
MODEL:
SUMMARY
DESCRIPTION.
KOEN B.V.
NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETR IC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.
KDESTEL A.
NUREC/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
K0ZINSKY,E.J.
NUREQ/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
KRYTER,R.C.
NUREG/CR-3687: LOOSE-PART MONITORING PROGRAMS AND RECENT OPERATIONAL EXPERIENCE IN SELECTED U.S.
AND WESTERN EUROPEAN COMMERCI AL NUCLEAR POWER STATIONS.
KURTZ,R.J.
NUREC/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Guarterly Report:
October 1993 - March 1984.Vols 1 & 2.
LAATS.E.T.
83
NUREO/CR-2531 R:2: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA DANK.
LEAF.G.K.
NUREG/CR-3505: A VOLUME-NEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
LEGGETT.R.W.
l NUREG/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED BONE-SEEKING RADIONUCLIDES.
NURE0/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRI ATE FOR HLW DOSIMETRY (ICRP-30),I.
LEMEUR.M.
NUREG/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
LEVY,I.S.
NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND GPERATOR INTERFACES IN NUCLEAR POWER PLANTS.
LEWIN,T.
NUREG/CR-3652: EVALUATION OF INSTRUMENTATION FOR DETECTION OF INADEGUATE CORE COOLING IN DOILING WATER REACTORS.
LEWIS P.M.
NUREG/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENIS.
LIETZKE.M.H.
NUREG/CR-3410: CHMONE: A DNE-DINENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE. FLOW AND CHEMICAL CONCENTRATIONS IN WATER DODIES.
LINE.J.F.
NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
LIPPINCOTT,E.P.
NUREO/CR-3391 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSINETRY INPROVEMENT PROGRAM.Guarterly Progress Report, April 1903 - June 1983.
NUREO/CR-3391 V04: LWR PRESSURE VESSEL SURVEILLANCE DOSINETRY IMPROVEMENT PROGRAM.Ouarterly Progress Report,0ctober 1983-December 1983.
LORENZ,R.A.
NUREO/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-3.
NUREO/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
LOSS.F.J.
NURE0/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF WELDS.
LU.S.C.
NURE0/CR-3710: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE P! PING -
STATUS REPORT.
LYCZK0WSKI.R.W.
NUREO/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN
- COMMIX, MACDONALD,P.E.
NURE0/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draft Report OF The USNRC PCI Review Group.
MACQORMAN D.R.
NURE0/CR-3759: LIGHTNING STRIKE DENSITY FOR THE CONTIGUOUS UNITED STATES FROM THUNDERSTORM DURATION RECORDS.
MAHASUVERACHAI NUREO/CR-3686 V01: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NURE0/CR-3686 VO2: WIPS-~ COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Theory Manual.
NURE0/CR-3686 V03: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF 84
PIPING SYGTEMS. Port'C - Progrcamor*o M:nual.
MA>ER, W.
NUREO-1090:
U. S.
NUCLEAR REGULATORY COMMISSION 1983 AltiUAL REPORT.
MAIER,M.W.
NUREO/CR-3759: LIGHTNING STRIKE DENSITY FOR THE CONTIQUOUS UNITED STATES FROM THUNDERSTORM DURATION RECORDS.
MALONE.P.O.
NUREC/CR-3774 VO1: ALTENNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTES. Task 1: Description of Methods And Assessment Of Criteria.
MARCH-LEUSA J.
NURE0/CR-3303: USE OF NEUTRON NOISE FOR DI AGNOSIS OF -!N-VESSEL ANOMALIES IN LIGHT-WATER REACTONS.
MARGULIES,T.S.
NUREG-1042: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
MARINELLI,F.
NURE0/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
NUREG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN-TUFF.
MARTIN,J.A.
NUREG-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENAR 108.
MARTZ,H.F.
NURE0/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results.
MCANENY.C.C.
NURE0/CR-3774 VO1: ALTkRNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTES Task 1: Description of Methods And Assessment Of Criteria.
MCCANN M.W.
NURE0/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBA8ILISTIC SAFETY STUDY: PLANT ANALYSIS.
MCCLUNG,M.W.
NUREO/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR STEAM OENERATOR TUBING PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDINO DECEMBER 31, 1983.
MCELROY,W.M.
NURE0/CR-3391 VO2: LWR PRESSURE VESSEL SURVE!LLANCE DOSIMETRY IMPROVEMENT PROGRAM. Guarterly Progress Report, April 1983 - June 1983.
NUMEO/CR-3391 V03: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0c tober 1,1982-September 30,1983.
NUME0/CR-3391 V04: LWR PRESSURE VESSEL SURVE!LLANCE DOSIMETRY IMPROVEMENT PROGRAM. Ouarterly Progress Report,0c tober 1983-December 1983.
MCGARRY, E. D.
NURE0/CR-3391 V03: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0ctober ~1,1982-September 30,1983.
McKENZIE,D.H.
NUMEO/CR-2675 V04: RELEVANCE OF S! OTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
NUREC/CR-2803: IMPROVED FIELD EXPERIMENTAL DES!ONS AND QUANTITATIVE EVALUATION OF AGUATIC ECOSYSTEMS.
MELBER.S.D.
NORE0/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERi!SE ON SHIFT.
MENKE B.H.
NURE0/CR-3295 V01: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness
- Degradation of A302-3 & A533-3 Ref erence Plates From PSF Simulated Surveillance & Through-Wall treadiation Capsules.
NURE0/CR-3295'V02: LIGHT WATER REACTOM PRESSURE VESSEL SURVEILLANCE 85
DOSIMETRY IMPROVEMENT PROGRAM:. Pootirrediction Natch Ductility &
-Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall. Specimen Capsules.
L
. NUREO/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF d
WELDS.
MENSING,R.W.
NUREO/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
MERKLE J.G.
i NUREG/CR-3672: EXAMINATION OF-THE SIZE EFFECTS AND DATA SCATTER
. OBSERVED IN SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
-MERZ,K.L.
NUREC/CR-3805: ENGINEERING CHARACTERIZATION OF OROUND MOTION. Task I:Ef fects Of Characteristics Of Free-Field Motion On Structural l
Response.
MEYER,R.
NUREO/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRI ATE FOR HLW DOSIMETRY (ICRP-30),I.
MIAO,C.C.
NURE0/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
MILLER,N.E.
NUREQ/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERI ALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, April 1983 - April 1984.
1 MILLER,W.O.
NUREO/CR-3774 VO1: ALTERNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL R ADIOACTIVE WASTES. Tas k 1: Description of Methods And Assessment Of Criteria.
1 MINER,5.
NUREG-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REQUIREMENTS ON OPERATINO PLANTS DESIONED BY BABCOCK AND WILCOX.
[
MINOR,E.
(
NUREO/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION I
PROCEDURES ON POLYMER PROPERITIES.
MOHR C. M.
i NUREQ/CR-J673 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOH BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Mode!
Description.
l NUREO/CR-3< 33 VO2: TRAC-BD1/ MOD 1: AN ADVANCED BEST EST! HATE COMPUTER j.
PROGRAM. OR BO! LING WATER REACTOR TRANSIENT ANALYSIS. Volume 2: Users j
Ouide.
l NUREC/CR-3633 VC3: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROCRAM FOR BO! LINO WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code Structure and Programming Information.
MOSADDAD B.
i I
NUREC/CR-3686 VO1: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
l NUME0/CR-3686 VO2: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF P! PING SYSTEMS.Part 8 - Theory Manua!.
NUREO/CR-3684 VO3: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF j
P! PING SYSTEMS.Part C - Programmer's Manua!.
MOSLEH!,F.
NUREG/CR-3848: EXPERIMENTAL INVESTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES.
MURATA,K.K.
NURE0/CR-3310: TESTING OF THE CONTAIN CODE.
I MYERS,L.B.
2 NUREO/CR-3632: METHODS FCR IMPLEMENTING REVISIONS TO EMERGENCY
}
OPEHATING PROCEDURES.
NESSE.R.J.
t i
08
i i
NURE!/CR-3566: SOCIOECONOMIC CONSEGUENCES OF NUCLEAR REACTOR ACCIDENTS.
'METI,8.
NOREO/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
NICKLIN,P.
NUREC/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NORE0/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part 5 - Theory Manual.
NUREO/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual.
NIELSON,K.K.
NURE0/CR-3533: RADON ATIENUATION HANDBOOK FOR URANIUM-MILL. TAILINGS COVER DESION.
4 NORWOOD,K.S.
NURE0/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-3.
NUREC/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
NOUR-0MID,S.
1 NURE0/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NURE0/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part 5 - Theory Manual.
NURE0/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual, NOUR9AKHSH,H.P.
i NURE0/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
I NOVA,8.J.
NURE0/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE j
REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
l NUTTLI,0.W.
NURE0/CR-3755: STRONG QROUND MOTION STUDIES FOR SOUTH CAROLINA EARTHOUAKES.
J O'KELLEY,0.D.
I NUME0/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRI ATE FOR HLW DOSIMETRY (ICRP-30),I.
03ERLANDER,P.L.
NURE0/CR-3681: MITIGATIVE TECHN!GUES AND ANALYSIS OF OENERIC SITE l
CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE i
ACCIDENTS.
I OLSON J.
NURE0/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERINO EXPERTISE ON SHIFT.
ONISHI,Y.
NURE0/CR-2424 V01: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /
l j
Radionuclide Transport Model FETRA.
NUREO/CR-2424 V02: MATHEMATICAL S!MULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing f or FETRA.
OSEOR NG, M. F.
NUREC/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-3.
NURE0/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
OSTMEYER,R.M.
NURE0/CR-2552: CRAC2 MODEL DESCRIPTION.
l OUGHOURLIAN,C.
NURE0/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF 87
PIPING ' SYSTEMS. Part A - User 's Manual.
NUREC/CR-3686 V02: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theorg Manual.
NOREG/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP - AND IMPACT ANALYSIS OF-1 PIPING SYSTEMS. Part C ' - Programmer 's Manual.
PANNELL,K.D.
NORE0/CR-3514: THE CHEMICAL BEHAVIOR OF IODINE IN'AGUEDUS SOLUTIONS UP TO 150 C.An Experimental Study of Nonredox Conditions.
PAPPAS,R.A.
NURE0/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL -
TESTING. Watts ~Bar Unit 1 Nuclear Reactor.
PATTERSON.M.R.
NUREG/CR-3410: CHMONE: A DNE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE. FLOW AND CHEMICAL -CONCENTRATIONS IN WATER BODIES.
PAYNE A.C.
NUREG/CR-3511 VC1: INTERIM RELI ABILITY EVALUATION PROGRAM: ANALYSIS OF
.THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume 1. Main-Report.
PELOGUIN,R.A.
NUREG/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
PELTD,P.J.
NUREO/CR-36B2: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation of Existing Methods.
NURE0/CR-3683: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Program Summary Through Fiscal Year 1983.
PENTZ,D.
NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN-TUFF.
PERKINS,K.R.
NORE0/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY.
POSTMA,A.K.
NORE0/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
POWELL,0.H.
NUREG/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.
NUREQ/CR-3686 V01: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NUREC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NUREC/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF P! PIN 3 SYSTEMS. Part C - Programmer 's Manual.
NUREO/CR-3686 V04: WIPS--COMPUTER CODE FOR ' WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
POWER,M.S.
NUREG/CR-3805: ENGINEERING CHARACTERIZATION OF QROUND MOTION.' Task I: Effects Of Characteristics of Free-Field Motion On Structural Response.
POWERS.D.A.
NUME0/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANIA-COATED STEEL MAHL,T.'E.
NUREO/CR-3875: THE USE OF IN-SITU PROCEDURES.FOR SEISMIC QUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
RAINS,J.H.
NURE0/CR-3054: CLOSEOUT OF IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.
(MUSSEL).
30
2-
' R ANK I N, W. L. :
NUREG/CR-3725: NUCLEAR. POWER-PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:Part I - The. Need For Plant-Reference Simulatorss Part
'II.
'The'Use.Of Plant-Reference simulators.-
NUREG/CR-3726: SIMULATOR FIDELITY AND _ TRAINING EFFECTIVENESS: A.
- COMPREHENSIVE BIBLIOGRAPHY WITH SELECTED ANNOTATIONS.
RASMUSSEN,N.C.
- NUREC/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
>RAUSCH,W.N.
NUREC/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (NT-3).
"RAWLINGS,G.
NUREO/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE-REPOSITORY DESIGN - DOMAL SALT.
NUREG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
REED,J.W.
NUREO/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC.
SAFETY STUDY: PLANT ANALYSIS.
REEVES,M.
NUREO/CR-3316: VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATIDN FLOW ANO TRANSPORT MODEL (SWIFT).
REICH,M.
NUREO/CR-3641:. RELIABILITY ASSESSMENT OF INDIAN POINT UNIT 3 CONTAINMENT STRUCTURE.
REXROTH,P.E.
NUREO/CR-3310: TESTING OF THE CONTAIN CODE.
RHOADS R.E.
NUREO/CR-3682:. NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation of Existing Methods.
NUREG/CR-3683: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Program Summary
.Through Fiscal Year 1983.
RIAHI,A.
NUREO/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
- NUREO/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NUREO/CR-3686 VO3: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual.
RITCHIE.L.T.
NUREO/CR-2552: CRAC2 MODEL DESCRIPTION.
RITTER.L.
NOREQ/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR OBTAINING SITE CHARACTERISTIC INFORMATION.
p ROBERDS.W.
NUREC/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
NOREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE
' REPOSITORY DESIGN-TUFF.
NUREO/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR.HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITOR IES. Final Report.(Task 5) June 1981 - February 1983.
RODRIGUES,R.
NUREG/CR-3755: STRONG GROUND MOTION STUDIES FOR SOUTH CAROLINA EARTHOUAKES.
ROGERS,L.A.
i NORE0/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT: A FINAL
- REPORT ON = THE INVESTICATION OF FISSION PRODUCT CHEMICAL FORMS.
ROGERS,V.C.
l es
NUREN/CR-3533: ' RADON ATTENUATION HANDBOOK FOR URANIUM-MILL TAILINGS
)
' COVER DESIGN.
4 ROSCOE.B.Ji NUREC/CR-3684: NUCLEAR POWER PLANT ALARM PRIORITIZATION (NPPAP). PROGRAM STATUS REPORT. January 1,1983 to September-31,1983.
. ROW D. C.
NUREG/CR-3686 VO1: ' WIPS--COMPUTER CODE FOR WHIP AND. IMPACT ANALYSIS OF f
. PIPING. SYSTEMS. Part A - User 's Manual.
q
'NUREG/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF '
PIPING' SYSTEMS.Part B - Theory Manual.
NUREC/CR-3686 V03: WIPS--COMPUTER CODE FOR WHIP. AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual.
ROWE,J.
NUREG/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE
~
REPOSITORY DESIGN - DOMAL SALT.
NUREC/CR-2614: IDENTIFICATION DF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN-TUFF.
RUSSELL.M.J.
NUREO/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION
'OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
RUST,W.D.
NURE0/CR-3759: LIGHTNING STRIKE DENSITY FOR THE CONTIGUOUS ' UNITED STATES FROM THUNDERSTORM DURATION RECORDS.
SAARI,L.M.
NUREC/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:PART I'- The Need For Plent-Reference Simulatorss Part II - The Use Of Plant-Reference Simulators.
SADIGH,K.
NURE0/CR-3805: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task I: Effects Of Characteristics of Free-Field Motion On Structural Response.
SADIK,S.
NOREO/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
SAGENDORF,~ J. F.
NUREG/CR-3488 V02: IDAHO FIELD EXPERIMENT 1981. Vol 1: Measurement Data.
SALAZAR,E.A.
NOREO/CR-3630: EQUIPMENT GUALIFICATION METHODOLOGY RESEARCH: TESTS OF PRESSURE SWITCHES.
SALLACH,R.A.
NURE0/CR-2921 : CHEMICAL INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS.
SAVY,J.B.
NUREC/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED STATES: METHODOLOGY AND INTERIM RESULTS FOR ' TEN SITES.
SCEPAN,J.
NUREC/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNICUES FOR OBTAINING SITE CHARACTERISTIC INFORMATION.
SC HR AUF, T. W.
NUREC/CR-3680: RELATIONSHIP BETWEEN THE GAS CONDUCTIVITY AND GEDMETRY OF A NATURAL FRACTURE.
SCHREIBER,R.E.
NURE0/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON SHIFT.
SC HROEDER, L.
NUREC/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATABASE: SEEK SYSTEM. (Users Manual).
SCHUMWAY,R.W.
NURE0/CR-3633 VO2: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER -
PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 2: Users i
90
Guido.
SCIACCA,F.W.
' NUREG/CR-3310: TESTING OF THE CONTAIN CODE.
CCDFIELD,N.R.
NUREC/CR-2531 RO2: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REQULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
SERKIZ,A.W.
NUREO-0800 03.9.3 R1: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 1 To
- Section 3.9.3, Appendix A.
NUREG-OBOO 03.9.4 R2: STANOARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2,To
'Section 3.9.4,
" Control Rod Drive Systems."
NOREO-OSOO 05.4.6 R3: STANDARD' REVIEW PLAN FOR'THE REVIEW OF SAFETY i
ANALYSIS REPORTS FOR NUCLEAR. POWER PLANTS. LWR Ed ition. Revi sion 3 To i
Section 5.4.6,
" Reactor Core Isolation Coolir; System ( B WR ). "
NUREG-OBOO 05.4._7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3-To Section 5.4.7,
" Residual Heat Removal (RHR) System."
NUREG-0000 06.-3 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY 3
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Ed ition. Revi sion 2 To Section 6.3,
" Emergency Core Cooling System."
3.
i NUREG-0000 09.2.1 R3: STANDARD REVIEW PLAN FOR THE. REVIEW OF SAFETY i.
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No.
3.
To Section 9.2.1,
" Station Service Water System."
NUREO-0800 09. 2. 2 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY j
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 9.2.2,
" Reactor Auxiliary Cooling Water Systems."
}
' NUREC-0800 10. 3 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No.3 To Section 10.3,
" Main Steam Suppig System."
l NUREO-OGOO 10. 4. 7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY i
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 10.4.7, "Cond ensa te And Feedwa ter Sy s tem" And BTP ASB 10-2,
" Design Guidelines For Avoiding Water Hammer...."
SHA.W.T.
NUREC/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
j NUREC/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN j
COMMIX.
SHAFER,J.M.
NURE0/CR-3681: MITIGATIVE TECHNIQUES AND ANALYSIS OF QENERIC SITE j
CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH S3 VERE ACCIDENTS.
I CHAH,V.L.
NUREO/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
j NOREO/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
.SHIKIAR.R.
i NUREG/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSINO t
j AND TRAINING:Part I - The Need For Plant-Ref erence Simulators: Part II - The Use Of Plant-Reference Simulators.
i I
SHORT S.A.
NUREQ/CR-3805: ENGINEERING CHARACTER!ZATION OF QROUND MOTION. Task i
I:Ef fec ts Of Characteristics Of Free-Field Motion On Struc tural Response.
SHORTENCARIER I
NUREO/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S CUIDE FOR THE GENERATION OF LATIN HYPERCUBE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
SIEGEL,A.I.
l NUREO/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS) i i
91
~
_.m
.MODEL[
SUMMARY
' DESCRIPTION.
- -SIMMONS,M.A.
(NUREC/CR-2675 VO4: RELEVANCE.0F BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
1 2 NUREO/CR-3797: DIONAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES '
.IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS OR MIGRATION-
'SIMONEN,F.A.
~
NUREG/CR-3743: -THE IMPACT OF NDE UNRELIABILITY ON PRESSURE VESSEL FRACTURE PREDICTIONS.
SIMPSON.J.C.
NUREO/CR-2955: ' ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT.
RESULTS FROM ELEVEN PARTICIPATING' URANIUM MILLS.
I SINGER C.L.
NUREO/CR-3360i COMPUTER PROGRAM 'CDCID: AN AUTOMATED GUALITY CONTROL PROGRAM USING CDC UPDATE.
NUREO/CR-3633 VO3: ' TRAC-BD1/ MOD 1i AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING. WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code Structure and Programming'Information.
)
SKAQQS.R.L.
NUREC/CR-3681: MITIGATIVE TECHNIGUES AND ANALYSIS-OF QENERIC SITE CONDITIONS FOR OROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE JACCIDENTS.
SKALSKI,J.R.
NUREO/CR-3797: DIGMAN:A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPL'NQ~ COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS
-OR MIGRATION.
SMITH,J.H.
NUREO/CR-32OO VO4: EDDY-CUHRENT INSPECTION FOR' STEAM QENERATOR TUBING i
j PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
SOTO, C.
.NUREC/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE
(.
REPOSITORY DESIGN - DOMAL SALT.
NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN-TUFF.
1 SPITZ,H.B.
j NUREO/CR-2955:. ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT j
RESULTS FROM ELEVEN PARTICIPATINO URANIUM MILLS.
i i
STACK,D.W.
j
.NUREO/CR-3134: A SETS USER 'S MANUAL FOR VITAL AREA ANALYSIS.
STAHL.D.
NUREO/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERI ALS USED FOR j
- l HIGH-LEVEL-WASTE PACKAOING. Annual Report, April 1983 - Apri l 1984.
START,E.E.
l NUREO/CR-3488 VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measur ement Data.
STEARNS R.G.
4 a
NOREO/CR-3769: DESCRIPTION ' AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
STRENGE,D.L.
l j
NUREO/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
j STUETZER,0.
3 NUREO/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CABLE FAILURE I
WITH~ MECHANICAL. DEGRADATION.
SUO-ANTTILA.A.
NUREO/CR-3379: SLAM - A SODIUM-LIMESTONE. CONCRETE ABLATION MODEL.
'SWANNACK R NUREO/CR-37971 DIONAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES i
IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS
.OR MIGRATION.
l SHEENEY,F.J.
k N
i 2
,..-..m---
-,.-.-.,m.-_.-,-.-,_-,_,,_,-,,----,_-...._,,-.._,..-_.-,_--,--......_.m.
m S
NUREC/CR-3303: _ USE. 0F NEUTRON NOISE FOR DI AGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
i
~ SWYLER,K.J.
l NUREC/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF L
RADWASTE'CONTAINING DRGANIC ION-EXCHANGE MEDIA.
I TAIG,A.R.
' NUREC/CR-2921: CHEMICAL INTERACTIONS 0F TELLURIUM VAPORS WITH REACTOR MATERIALS.
TARBELL,W.W.
NUREC/CR-3023: MOLTEN-THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
i-TAWILL,J.J.
NUREC/CR-3566: SOCIDECDNOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
TAYLOR,D.D.
NUREC/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER -
PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model Description.
TAYLOR M.
f NOREC-0837 VO3 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report September-December 1983.
TAYLOR,R.E.
t NUREO/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES j
OF A 351 CFSA CAST STAINLESS. STEEL IN LWR ENVIRONMENT.
TAYLOR,T.T.
NOREO/CR-3753: AN EVALUATION OF MANUAL ULTRASONIC INSPECTION OF CENTRIFUGALLY CAST STAINLESS STEEL PIPING.
THEDFANOUS,T.G.
NUREO/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
j THINNES,0.L.
j NUREO/CR-3722: DAMPING TEST RESULT 5 FOR STRAIGHT SECTIONS OF 3-INCH AND l
i 8-INCH UNPRESSURIZED PIPES.
THOMA J.O.
j NUREG-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN l
REGUIREMENTS ON OPERATING PLANTS DESIGNED BY BABCOCK AND WILCOX.
I THOMAS,J.M.
NUREC/CR-2003: IMPROVED FIELD EXPERIMENTAL DESIGNS AND QUANTITATIVE j.
EVALUATION OF AGUATIC ECOSYSTEMS.
NOREC/CR-3797: DICMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES l'
IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS I
OR MIGRATION.
THOMAS,V.W.
NUREC/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADI ATION EXPOSURE 5
RATES AND SOIL 266RA CONCENTRATIONS.
I-THOMPSON.F.L j.
NUREO/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCL!DE i
TRANSPORT IN COASTAL WATERG.Vol 1: Testing Of The Sediment /
}
Radionuclide Transport Model FETRA.
j NUREG/CR-2424 VO2: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE j
TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Lis ting f or FETRA.
1 THOMPSON.S.L.
j NUREO/CR-3329 VO4: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report October-December 1983 l
NUREC/CR-3608: RELAPS ASSESSEMENT: LOFT Large Break L2-5.
1 THOMPSON,T.
I NOREG-0837 VO3 NO4: NRC TLD DIRECT RADIATION MONITORING 4
NETWORK. Progress Report September-December 1983.
THURGOOD,M.J.
}
NURE0/CR-3748: COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
NUREC/CR-3749: COBRA-NC POST-TEST PREDICTIONS FOR. HDR CONTAINMENT STEAM BLOWDOWN TEST V44 (INTERNATIONAL STANDARD PROBLEM-16).
t
~
f 1
93
-TICHLERid.
NUREC/CR-2907 VO2: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER
. PLANTS. Annual Report 1981.
TOBIAS,M.L.
NUREG/CR-3422 VO3:-AEROSOL RELEASE AND TRANSPORT PROGRAM. Quarterly Progress Report'For' July-September 1983.
TOMAR,M.
NUREC/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draf t Report Of The USNRC PCI Review Group.
TDMARZ,F.J.
NUREG/CR-3805: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task i
I: Effects Of Characteristics of Free-Field Motion On Struc tural Response.
TORR 0NEN,K.
NUREC/CR-3546: THE. TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES OF A 351 CF8A CAST STAINLESS STEEL IN LWR ENVIRONMENT.
TOTH,L.M.
l NUREG/CR-3514: THE CHEMICAL: BEHAVIOR OF IODINE IN AGUEDUS SOLUTIONS UP TO 150 C.An Experimental Study of Nonredox Conditions.
TOWE.S.K.
NUREC/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE I
REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
TRAVIS.J.R.
NUREO/CR-3335: DATA
SUMMARY
REPORT FOR-FISSION PRODUCT RELEASE TEST HI-3.
NUREC/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST z
HI-4.
TRENT,D.S.
i-NUREC/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE I
SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER REACTOR.
TUZLA,K.
NUREG/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL j
HEAT FLUX BOILING.
UNAL.C.
NUREC/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
VAN HOUTEN.R.
NUREO/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draf t Report Of The USNRC PCI Review Group.
VAN TUYLE,0.J.
NURE0/CR-3603: MINET VALIDATION SURVEY USINC EBB-II TEST DATA.
i VAN VLIET,J.
NUREG-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REGUIREMENTS ON OPERATINO PLANTS DESIGNED BY BABCOCK AND WILCOX.
I VASSILAROS.M.G.
I NUREG/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH i
DIAMETER ASTM A106 STEEL PIPE.
VESELY,W.E.
NUREC/CR-3682: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation l
of Existing Methods.
VISSING.S.
NUREO-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REGUIREMENTS ON OPERATING PLANTS DESIGNED BY BABCOCK AND WILCOX.
i.
WARDS,D.S.
NUREC/CR -3316: VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TRANSPORT MODEL (SWIFT).
i WARE,A.G.
NUREG/CR-3722: DAMPING TEST RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND l
8-INCH UNPRESSURIZED PIPES.
l 94
WARRINER.J.D.
NUREG/CR-3774 VO1: ALTERNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTES. Task 1: Description of Methods And Assessment Of Criteria.
WEBB,T.
NUREG/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA: A User 's Guide For The Applicable Computer Programs In The Statistical Package For Social Scientists (SPSS).
WEBSTER,C.S.
NUREO/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST l
HI-3.
NUREO/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
WELDEN,C.
NUREO/CR-2613: INDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
WHITEMAN,D.E.
NUR EO/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results.
WILKEY,P.L.
NUREO/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
WILSON,S.L NUREO/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAu!TY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
WINEGARDNER W.
NUREO/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
WITHEE,C.J.
NUREG-1065: ACCEPTANCE CRITERI A FOR THE LOW ENRICHED URANIUM REFORM AMENDMENTS.
WOLF,J.J.
NUREO/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS)
MODEL:
SUMMARY
DESCRIPTION.
WOODLEY,R.E.
NUREC/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER.
WRIGHT,R.E.
NUREO/CR-3596: SEVERE ACCIDENT SEQUENCE ANALYSIS (SASA) PROGRAM SEGUENCE EVENT TREE: BOILING WATER REACTOR ANTICIPATED TRANSIENT WITHOUT SCRAM.
WULFF W.
NUREO/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD.O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR DWR SYSTEM TRANSIENTS.
YOUNG,J.A.
NUREO/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION EXPOSURE RATES AND SOIL 266RA CONCENTRATIONS.
ZALOUDEK,F.R.
NUREO/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESP) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
96
Subject index e
~,
This index was developed from keywords and word strings in titles and ab-st acts. During this development period, there will be some redundancy, which will be removed letor when a reasonable thesaurus has been developed
' through experience.' Suggestions for improvements are welcome.
/
1983 Annual eport
./'
NUREG-1090:
U. S NUCLEAR REGULATORY COMMISSION 1903 ANNUAL REPORT.
97th Congress,-2nd Session NUREG-0980: NUCL,EAh REGULATORY LEGISLATION.
ATWS 1
NUREO/CR 3396: SEVERE ACCIDENT SEQUENCE ANA'.YSIS (SASA) PROGRAM SEGUENCE EVENT TREE: DOILING WATER REACTOR ANTICIPATED TRANSIENT WITHOUT SCRAM.
NURCO/CR-3633 V01: TRAC-BD1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER PROGRAM FOR DOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model Descpfption.
NUREC/GR-3633 V02: THAC-DD1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 2: Users Guide.
NUREO/CR-3633 V03: TRAC-DD1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER PROGRAM FOR.DOILING WATER REACTOrt TRANSIENT ANALYSIS. Volume 3: Code StruIture,$nd H ogramming Information.
Abnormal Occurrence NUREO-0090 V06 NO3; REPORT TO CONGRESS ON ADNORMAL OCCURRENCES. July-September 1983.
NUREO-0030 vo6 N04: REPORT TO CONGRESS ON ADNORMAL OCCURRENCES. October
-Decembir 1993.
Accelerograms NURE0/CR-3755; STRONG GROUND -MOTION STUDIES FOR SOUTH CAROLINA EARTHOUAKES.
Accident Sequence NUREC/CR-376?: IDENTIFICATION OF EQUIPMENT AND COMPONENTS PREDICTED AS SICNIFICANT CONTRIDUTURS TO SEVERE CORE DAMAGE.
Accident NURE0/CR-2940: HEALISTIC SIMULATION OF SEVERE ACCIDENTS IN DWRS-COMPUTER MODELING REQUIREMENTS.
NURE0/CR-3310: TESTING OF THE CONTAIN CODE.
NUREC/CR-3539* IMPACT OF CONTAINMENT DUILDING LEAKAGE ON LWR ACCIDENT RICK.
NURE0/CR-3566: SOCIDECCNOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
j NURE0/CR-3608: RELAPS ASSEC3EMENT: LOFT Large Dreak L2-5.
NURE0/CR-3633 V01: TRAC-DD1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER PROGRAM FOR DOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model Description.
f NURE0/CR 3633 V02: TRAC-ED1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER 97
PROGRAM FOR DOILING WATER REACTOR 1RANS!ENT ANALYSIS. Volume 2: Users Guide.
NURE0/CR-3633 V03: TRAC-DD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR DOILING WATEN REACTOR TRANSIENT ANALYSIS. Volume 3: Code Scructure and Programming Information.
NURE0/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT: A FINAL REPORT ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
NURE0/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS, NUPEC/CR-3681: MIT!GATIVE TECHNIQUES AND ANALYSIS OF GENCRfC SITE CONDITIONS FOR OROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS.
Acoustic Emission NUREO/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit 1 Nuclear Reactor.
NURE0/CR-3825 V01-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Guarterly Report:
October 1983 - March 1984.Vols 1 & 2.
Aerial Imagery NURE0/CR-3503: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR ODTAINING SITE CHARACTERISTIC INFORMATION.
Aerosol Release NURE0/CR-3422 V03: AEROSOL RELEASE AND TRANSPORT PROGRAM. Quarterly Progress Report For July-September 1983.
Aerosol NURE0/CR-3745: DIOLOGICAL CHARACTER!ZATION OF RADI ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URAN!UM MILLING EFFLUENTS. Annual Progress Report: April 1 1982 - March 31,1983.
1 Agenda NUREG-0936 VO3 N01: NRC REGULATORY AGENDA. Quarterly ReportoJanuary-March 1984.
Aging NURE0/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDURES ON CROSS-LINKED POLYOLEFIN CADLE'S PERFORMANCE.
NURE0/CR-3627: FRANTIC !! APPLICATIONS TO STANDDY SAFETY SYSTEMS.
NUREO/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION A0!NO S!MULATION PROCEDURES ON POLYMER PROPER! TIES.
Air Coolers NURE0/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Dase Assessment And Suggested Esperimental Program.
Alarm Prioritization NURE0/CR-36S4: NUCLEAR POWER PLANT ALARM PRIORITIZATION (NPP AP) PROGRAM STATUS REPORT. January 1,198") to September 31,1983.
Alternative Disposal Methods NURE0/CR-3681: MITICATIVE TECHN! QUES AND ANALYSIS OF OENERIC SITE CONDITIONS FOR OROUND-WATER CONTAMINATION ASSOCI ATED WITH SEVERE ACCIDENTS.
Anomaly NURE0/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE ORAVITY FIELD IN THE REELFOOT LAKE RE0!ON OF NORTH 4EST TENNESSEE.
Anticipated Transients Without Scram NURE0/CR-3596: SEVERE ACCIDENT SEQUENCE ANALYSIS (SABA) PROGRAM SEQUENCE EVENT TREE:00! LING WATER REACTOR ANTICIPATED TRANSIENT WITHOUT SCRAM.
NURE0/CR-3633 V01: TRAC-DD1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER PROGRAM FOR D0! LING WATER REACTOR TRANS!ENT ANALYSIS. Volume 1: Model Description, NURE0/CR-3633 V02: TRAC-BD1/ MOD 1: AN ADVANCED DEST ESTIMATE COMPUTER t
PROGRAM FOR 00! LING WATER REACTOR TRANS!ENT ANALYSIS. Volume 2: Users Guide.
I 90
T NUREO/CR-3633 VO3: TRAC-DD1/ MOD 1:AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code-Structure and Programming Information.
Aquatic Ecosystems i
NUREG/CR-2803: IMPROVED FIELD EXPERIMENTAL DESIGNS AND GUANTITATIVE EVALUATION OF AGUATIC ECOSYSTEMS.
Aqueous Solution NUREG/CR-3514: THE CHEMICAL BEHAVIOR OF IODINE IN AOUEOUS SOLUTIONS UP TO 150 C.An Experimental Study of Nonredox Conditions.
Asiatic Clam NUREC/CR-3054: CLOSEOUT OF IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.
(MUSSEL).
Atomic Energy Act NUREG-0980: NUCLEAR REGULATORY LEGISLATION.
Attentuation NUREG/CR-3839: AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR A 6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
BEACON-MOD 3A NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR C AVITY SUBCOMPARTMENT ANALYSES.
BEACON NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR C AVITY SUBCDMPARTMENT ANALYSES.
Backfill NUREG/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
Bioassay Monitoring NUREG/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS.
Biotic Pathways NUREG/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE L3NG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
Blockage NUREG/CR-3054: CLOSEOUT OF IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.
(MUSSEL).
Plowdoun NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR C AVITY SUBCOMPARTMENT ANALYSES.
NUREG/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
Bolts NUREG/CR-3604: BOLTING APPLICATIONS.
Bone Doso NUREG/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED BONE-SEEKING RADIONUCLIDES.
Durial Sites NUREG/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
CDC NUREG/CR-3360: COMPUTER PROGRAM CDCID: AN AUTOMATED GUALITY CONTROL PROGRAM USING CDC UPDATE CDCID NUPEG/CR-3360 COMPUTER PROGRAM CDCID: AN AUTOMATED QUALITY CONTROL PROGRAM USING CDC UPDATE.
CHMONE NUREG/CR-3410: CHMONE: A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE. FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
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ICOBRA-NC NUREO/CR-3749: COBRA-NC POST-TEST PREDICTIONS FOR HDR CONTAINMENT STEAM 8 LOWDOWN TEST V44 (INTERNATIONAL STANDARD PROBLEM 16).
COBRA NUREO/CR-3307 VO3:. REACTOR SAFETY RESEARCH PROGR AMS. Guart erl y Report 3
July-September 1983.
NUREO/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report October-December"1983.
NUREG/CR-3810 VOI: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report
)
' January-March 1984.
COBRA / TRAC NUREC/CR-3748: COBRA / TRAC SIMULATION OF. SEMISCALE S-UT-5 TEST.
COMMIX-1B.
' NUREO/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
NOREG/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE ' SCHEME IN COMMIX.
COMMIX' NUREO/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
7 NUREO/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
COMPARE-MOD 1A NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT' ANALYSES.
COMPARE NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT' ANALYSES.
CONTAIN NUREO/CR-2679 VO4: ADVANCED REACTOR' SAFETY RESEARCH,GUARTERLY REPORT, OCTOBER-DECEMBER 1982.
NUREG/CR-3310: TESTING OF THE CONTAIN CODE.
CRAC NUREO-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
NUREO/CR-2552: CRAC2 MODEL DESCRIPTION.
NUREG/CR-2552: CRAC2:MODEL DESCRIPTION.
Calculation Of Reactor Accident Consequences NUREQ/CR-2552: CRAC2 MODEL DESCRIPTICN.
Calibration NOREO/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LNR SERVICE. Annual 'ept for 1983.
NUREG/CR-3775: GUALITY ASSURANCE FOR MEASL TMENTS OF IONIZING
. RADIATION.
Cast Stainless Steel Pipe NUREG/CR-3753: AN EVALUATION OF MANUAL ULTRASONIC INSPECTION OF CENTRIFUGALLY CAST STAINLESS STEEL-PIPING.
- Characterization NUREG/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT MATERIAL.
Chemical Behavior NUREG/CR-3514:.THE CHEMICAL BEHAVIOR OF IODINE IN AGUEOUS SOLUTIONS UP
-TO 150 C.An Experimental' Study of Nonredox Conditions.
Chemical Interactions NUREO/CR-2921: CHEMICAL INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS.
Chemical Reactions NOREO/CR-3379: SLAM - A SODIUM-LIMESTONE CONCRETE ABLATION MODEL.
Chemical Stabilizers NUREO/CR-3697: LABORATORY TESTING OF CHEMICAL-STABILIZERS FOR CONTROL OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
Chlorination NUREO/CR-3410: CHMONE:A DNE-DIMENSIONAL COMPUTER CODE FOR SIMULATING 100 L
~.
l TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
Circumferential Fatigue Precracks NUREG/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH
-DIAMETER ASTM A106 STEEL PIPE.
Cladding NUREG/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL HEATER ROD DESIGN ON GU6NCH BEHAVIOR.
NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN FY 1983-1987.
NUREG/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER.
Cleavage Fracture Toughness j
NUREG/CR-3672: EXAMINATION OF THE SIZE EFFECTS AND DATA SCATTER OBSERVED IN SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
i I
Climatic Variables NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
Climatography 1880-1982.
NUREC/CR-3670: VIOLENT TORNADO CLIMAT0 GRAPHY, Code NUREG/CR-2531 RO2: INTRODUCTORY USER'S MANUAL FOR THE U.S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
NOREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
NUREC/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-38 MOD.O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
NUREC/CR-3741 VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 2 Topical Report. Volume 1: Data Evaluat ion.
NUREG/CR-3608: RELAP5 ASSESSEMENT: LOFT Large Break L2-5.
Color Infrared NUREC/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR ~
OBTAINING SITE CHARACTERISTIC INFORMATION.
Comparison NUREG-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REGUIREMENTS ON OPERATING PLANTS DESIGNED BY BABCOCK AND WILCOX.
4 NOREG/CR-3720: PREDICTION AND EXPERIMENT COMPARISDNS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
Competer Code l
NUREC/CR-2552: CRAC2 MODEL DESCRIPTION.
NUREG/CR-2679 VO4: ADVANCED REACTOR SAFETY RESEARCH, GUARTERLY REPORT, OCTOBER-DECEMBER 1982.
NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
NUREG/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report July-September 1983.
NUREC/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report October-December 1983.
NUREC/CR-3310: TESTING OF THE CONTAIN CODE.
NUREG/CR-3360: COMPUTER PROGRAM CDCID: AN AUTOMATED GUALITY CONTROL PROGRAM USING CDC UPDATE.
NUREG/CR-3378: VERIFICATION OF THE NETWORK FLOW AND TRANSPORT / DISTRIBUTED VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
NUREC/CR-3410: CHMONE: A DNE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
NOREG/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
NUREG/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN i
COMMIX.
NUREG/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER 101
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REACTOR.
~NUREG/CR-3603: MINET VALIDATION SURVEY USING EBB-II TEST DATA.
NUREG/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
NUREO/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT' ANALYSIS OF PIPING; SYSTEMS. Summary Report.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NUREC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NUREG/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Programmer 's Manual.
NUREO/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
NUREG/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT FLUID-THERMAL MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT USING SOLA-PTS.
NUREG/CR-3748: COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
NUREC/CR-3800: REFCO-83 USER 'S MANUAL.
NUREC/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS. GuarterJ y Report January-March 1984.
Computer Model NUREO/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND RADIONUCLIDE TRANSPORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /
Radionuclide Transport Model FETRA.
NUREO/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN BWRS-COMPUTER MODELING REGUIREMENTS.
NUREG/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S GUIDE FOR THE GENERATION OF LATIN HYPERCUBE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
Computer Program NUREC/CR-2424 VO2: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing f or FETRA.
NUREO/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
NUREG/CR-3567: TRAC-PF1: AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED' AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
NUREG/CR-3749: COBRA-NC POST-TEST PREDICTIONS FOR HDR CONTAINMENT STEAM BLOWDOWN TEST V44 (INTERNATIONAL STANDARD PROBLEM 16).
NUREG/CR-3797: DIGMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS OR MIGRATION.
Computer Storage NUREC/CR-2531 RO2: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
Constructibility NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
i l
Construction NUREG-1055: IMPROVING QUALITY AND THE ASSURANCE OF GUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCIAL NUCLEAR POWER PLANTS.A Report To Congress.
NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT l
FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final 1
Report (Task 5) June 1981 - February 1983.
Containment Spray NURE0/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
{
Containment 102 l
l NOREO/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN BWRS-COMPUTER MODELING REGUIREMENTS.
NUREC/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT RISK.
NUREO/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CABLE FAILURE WITH MECHANICAL DEGRADATION.
NUREG/CR-3641: RELIABILITY ASSESSMENT OF INDIAN POINT UNIT 3 CONTAINMENT STRUCTURE.
NUREG/CR-3653: CONTAINMENT ANALYSIS TECHNIQUES. A State-Of-Th e-Art
. Summary.
NUR EO/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
NUREG/CR-3749: COBRA-NC POST-TEST PREDICTIONS FOR HDR CONTAINMENT STEAM
. BLOWDOWN TEST V44 (INTERNATIONAL STANDARD PROBLEM 16).
Contaminant NUREG/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS. Vol 1: Testing Of The Sediment /
Radionuclide Transport Model FETRA.
NURE0/CR-2424 VO2: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing f or FETRA.
Control Room NUREC/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATAB ASE: SEEK SYSTEM. (Users Manual ).
NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER PLANTS.
Cooling Systems NUREG/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF COPPER.
Cooling Water NUREO/CR-3054: CLOSEOUT OF IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.
(MUSSEL).
Cepper NUREG/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION. FATE AND EFFECTS OF COPPER.
Core Cooling NUREG/CR-3652: EVALUATION OF INSTRUMENTATION FOR DETECTION OF INADEGUATE CORE COOLING IN BOILING WATER REACTORS.
-Core Melt NUREC/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BEc.
NUREC/CR-3511 Vol: INTERIM RELI ABILITY EVALUATION PROGRAM: ANALYSIS OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume
- 1. Main Report.
NUREG/CR-3681: MITIGATIVE TECHNIQUES AND ANALYSIS OF GENERIC SITE CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS.
Corium NUREG/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANI A'-COATED STEEL.
Cost NUREG/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
NUREO/CR-3800: REFCO-83 USER 'S MANUAL.
Crack Arrest NUREO/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN FY 1983-1987.
Cross-Linked Polyolefin Cable NUREC/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDURES ON CROSS-LINKED POLYOLEFIN CABLE'S PERFORMANCE.
Cylindrical Model 1CG
_ _ _ _ _ _ _ _ ~...,
NURE3/CR-38 08: EXPERIMENTAL INVESTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES.
DHRS NUREG/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY.
DIGMAN NUREG/CR-3797: DIGMAN:A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS OR MIGRATION.
Damping Values NUREG/CR-3722: DAMPING TEST'RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND B-INCH UNPRESSURIZED PIPES.
Data Access Software NUREC/CR-2531.RO2: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
Data Bank I
NUREG/CR-2531 RO2: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA BANK.
Data NUREG/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
Database NUREC/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATABASE: SEEK SYSTEM. (Users Manual).
Decay Heat Removal Systems NUREC/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY.
Decay NUREG/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
Deep Geologic Repositories NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
Deficiencies NUREC/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER PLANTS.
Degradation NUREG-1056: REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM GENERATORS.
Degree Of Sensitization NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rep.t for 1983.
Design NUREG-1055: IMPROVING GUALITY AND THE ASSURANCE OF QUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCI AL NUCLEAR POWER PLANTS. A Report To Congress.
NUREG/CR-2803: IMPROVEO FIELD EXPERIMENTAL DESIGNS AND QUANTITATIVE EVALUATION OF AGUATIC ECOSYSTEMS.
NUREC/CR-3628: PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT STRUCTURES.
NUREC/CR-3641: RELIABILITY ASSESSMENT OF INDI AN POINT UNIT 3 CONTAINMENT STRUCTURE.
Discharge Water NUREG/CR-3410: CHMONE: A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
Dispersion NUREG/CR-3773: VARIATION OF PLANETARY BOUNDARY LAYER DISPERSION PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS.
Disposal 104 e-.~.-
e n...
l NOREO/CR-3383i IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF RADWASTE'CONTAINING ORGANIC ION-EXCHANGE MEDIA.
NUREC/CR-3489: ! ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
i NUREG/CR-3681: MITIGATIVE TECHNIGUES AND ANALYSIS OF GENERIC SITE CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCI ATED WITH SEVERE ACCIDENTS.
Distributed Velocity NUREG/CR-3378: VERIFICATION. 0F Tl!E. NETWORK FLOW AND '
TRANSPORT / DISTRIBUTED VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
Damal Salt NUREC/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE
~ REPOSITORY DESIGN - DOMAL. SALT.
Dase Conversion Factor NUREC/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED BONE-SEEKING RADIONUCLIDES.
Dase NUREC-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
NUREC/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM i
REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
NUREG/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDURES ON CROSS-LINKED 4
l POLYOLEFIN CABLE'S PERFORMANCE.
NUREG/CR-3745: BIOLOGICAL CHARACTERI1ATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED. URANIUM MILLING EFFLUENTS. Annual Progress Report: April 1,1982 - March 31,1983.
l
(-
NUREO/CR-3781 ' DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL
)
EVENTS-DRAFT: Draf t Report Of The USNRC PCI Review Group.
Desimetry 4
i NUREG/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness l
Degradation of A302-B & A533-B -Ref erence Plates From PSF Simulated Surveillance L Through-Wall Irradiation Capsules.
NUREG/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE l
i DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
l Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen _ Capsules.
NUREG/CR-3391 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report, April 1983
_ June 1983.
j NUREG/CR-3391 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
I NUREG/CR-3391 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Guarterl y Prog ress Report,0c tober 1983-December 1983.
Downcomer NUREG/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT FLUID-THERMAL MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT 4
USING SOLA-PTS.
Draft Environmental Statement NUREG-1074: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF HOPE CREEK GENERATING STATION. Docket No. 50-354.-(Public Service Electric And Cas Co And Atlantic City Electric Co)
Dry Storage NUREC/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER.
Ductile Material NUREC/CR-3644: REVIEW OF PROPOSED FAILURE CRITERIA FOR DUCTILE 4
MATERIALS.
Dust Emission NUREG/CR-3697: LABORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL OF FUGITIVE DUST EMISSIONS FRO" URANIUM MILL TAILINGS.
106
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' EB -II Toot ' Dsta' NUREO/CR-3603: MINET. VALIDATION SURVEY USING EBB-II. TEST DATA.
'ESF; NOREO/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base-Assessment And Suggested Experimental Program.
NUREC/CR-3755: STRONG GROUND. MOTION STUDIES FOR SOUTH CAROLINA
-EARTHGUAKES.
i
- NOREG/CR-3756
- t SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN ' UNITED :
STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
NUREO/CR-3768; NEW MADRID SEISMOTECTONIC STUDY: Activities During Fiscal Year 1982.
NUREC/CR-3769: DESCRIPTION AND SIGNIFICANCE-OF THE GRAVITY FI' ELD IN THE F
. REELFOOT LAKE REGION OF-NORTHWEST TENNESSEE.
NUREC/CR-3805: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task I: Ef f ec ts Of, Characteristics Of Free-Field Motion On Struc tural R e s p o r.s e.
-NUREG/CR-3839i AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR A 6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
Economic Risks
-_ NUREO/Ch-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
Ecosystems NUREO/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF COPPERS Eddy-Current Inspection NUREC/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR. STEAM. GENERATOR TUBING PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
Effluent Release NUREO/CR-3838: AN INITIAL REVIEW OF SEVERAL METEONOLOGICAL MODELS i
SUITABLE FOR' LOW-LEVEL WASTE DISPOSAL FACILITIES.
Effluent 4
NUREC/CR-2907 VO2: -RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1981.
- NUREO/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF COPPER.
I Electric Cable 1
NUREO/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDURES ON CROSS-LINKED POLYOLEFIN CABLE'S PERFORMANCE.
NUREO/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL. CABLE FAILURE WITH MECHANICAL DECRADATION.
Embrittlement ~-
NUREC/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE i
DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B & A533-B Ref erence Plates From PSF Simulated Surveillance L Thrcugh-Wall Irradiation Capsules.
l NUREC/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
1 Tensile Strength Determinations For. PSF Simulated Surveillance &
.Through-Wall. Specimen Capsules.
[
Emergency Coolant NU9EO/CR-3639: LARGE BREAM LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION.
-Emergency Core Cooling NUREO/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE.
SIMULATION OF THERMAL-MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER i
REACTOR.
Emergency Operating Procedure NUREC/CR-3632: METHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING PROCEDURES.
l i
106 a.
'Enstgcncy Plcnning NUREC-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
Enforcement Actions NUREG-0940 VO3 NO1: ENFORCEMENT ACTIONS:SIGNIFICANT ACTIONS RESOLVED.Guarterly Progress Report (January - March 1984).
Engineered Barriers NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
Engineered Safety Feature NUREC/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
Engineering Characterization NUREO/CR-3805: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task I: Ef f ects Of Characteristics Of Free-Field Motion On Structural Response.
Engineering Expertise NUREG/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON SHIFT.
Environmental Impact Appraisal NUREG-1071: ENVIRONMENTAL IMPACT AFPRAISAL FOR RENEWAL OF SOURCE MATERIAL LICENSE NO. SUB-526. Docket No. 40-3392.(A111ed Chemical Company UF6 Conversion Plant)
NUREG-1077: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-21. Docket No. 70-25.
(Energy Systems Croup Rockwell International Corporation)
NOREG-1078: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1097. Docket No. 70-1113. (General Electric Company,Wilmington Manufacturing Department)
Environmental Impact NUREG-0974: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF LIMERICK GENERATING STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-353.(Philadelphia Electric Company)
Equipment Gualification NOREG/CR-3630: EQUIPMENT GUALIFICATION METHODOLOGY RESEARCH: TESTS OF PRESSURE SWITCHES.
NUREO/CR-3375: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
Error NUREC/CR-3720: PREDICTION AND EXPERIMENr COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO ELOWDOWN.
Evaluation NUREG/CR-2803: IMPROVED FIELD EXPERIMENTAL DESIGNS AND GUANTITATIVE EVALUATION OF AGUATIC ECOSYSTEMS.
Event Tree NUREG/CR-3596: SEVERE ACCIDENT SEQUENCE ANALYSIS (SASA) PROGRAM SEGUENCE EVENT TREE: BOILING WATER REACTOR ANTICIPATED TRANSIENT I
WITHOUT SCRAM.
Ex-Core NUREG/CR-3303: USE OF NEUTRON NOISE FOR DIAGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
Exposure NUREG-1028: RUPTURED CESIUM-137 WELL-LOGGING SOURCE AT SHELWELL SERVICES,INC., HEBRON, OHIO.
NUREG/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION EXPOSURE RATES AND SOIL 266RA CONCENTRATIONS.
NUREO/CR-3745: BIOLOGICAL CHARACTERIZATION OF RADI ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report: April 1,1982 - March 31 1983.
107
FETRA-
-NUREG/CR-2424'VO1: NATHEMATICAL SIMULATION OFISEDIMENT AND R ADIONUCLIDE TRANS* ORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /.
Radionuclide Transport Model FETRA.
NUREG/CR-2424 VO2: : MATHEMATICAL. SIMULATION OF SEDIMENT AND R ADIONUCLIDE
- TRANSPORT ; IN. COASTAL' WATERS. V 2 Us er 's M CP Li s ting f or.FETRA.
FORTRAN IV.
- NUREC/CR-3749: COBRA-NC POST-TEST PREDICTIONS FOR 'HDR ' CONTAINMENT STEAM
-BLOWDOWN TEST-V44 (INTERNATIONAL STANDARD PROBLEM 16).
' FORTRAN NUREO/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S GUIDE FOR THE GENERATION
-OF LATIN HYPERCUBE'AND RANDOM SAMPLES FOR USE WITH COMPUTER.MODELS.-
FRAC NUREG/CR-3650:
A' STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results.
FRANTIC II NUREC/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
FRAPCON NUREO/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterl y Report July-September.1983.
I
' NUREO/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS.'Guarterly Report October-December 1983.
NUREO/CR-3741 VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS-
. CAPABILITIES. Phase 2 Top ical Report, Volume 1: Data Evaluat ion.
NUREG/CR-3810.VO1: ~-REACTOR SAFETY RESEARCH PROGR AMS. Guarterl y Rep or t :
Ja n uar y-Mar c h -- 1984.
Failure Evaluation NUREG/CR-3754: FAILURE EVALUATION OF GENERAL ELECTRIC SB-1 AND SB-9 REACTOR MODE SWITCHES.
Failure-NUREC-1055: IMPROVING GUALITY AND THE ASSURANCE OF QUALITY IN.THE-DESIGN AND CONSTRUCTION OF COMMERCI AL NUCLEAR POWER PLANTS. A Report To Congress.
i.
NUREC/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CABLE FAILURE WITH MECHANICAL DEGRADATION.
NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATEO PARAMETRIC
' EMPIRICAL BAYES ESTIMATORS TO THE NUCLEA9 PLANT RELIAI1ILITY DATA SYSTEM.
NUREO/CR-3644: REVIEW OF PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS.
NUREC/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results,
[
NUREG/CR-3653: CONTAINMENT ANALYSIS TECHNIQUES. A State-Of-Th e-Art l
Summarg.
NUREC/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL-EVENTS-DRAFT: Draft Report Of The USNRC PCI Review Group.
Fatigue Crack NUREO/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES HOF A 351 CFOA CAST STAINLESS STEEL IN LWR ENVIRONMENT.
Fault Tree NUREO/CR-3134: A SETS USER 'S MANUAL FOR VITAL AREA ANALYSIS.
NUREO/CR-3511 VO1: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF
=THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume
- 1. Main Report.
Fault NUREC/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE i
- REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
Ferrite Vessel NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGPAM - FIVE YEAR PLAN FY 1983-1987.
i Field Experiment 108 f-
. -,,, ~.
PAGE NURE9/CR-34 8 VO2: IDAHO FIELD ' EXPERIMENT 1981. Vol 1: Msasursm:nt Data.
Filter System NUREO/CR-3727: FISSION PRODUCT: REMOVAL IN ENGINEERED SAFETY FEATURE
-(ESF) SYSTEMS. Data Base Ascessment And Suggested Experimental Program.
' Final Environmental Statement NUREC-0974: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF LIMERICK GENERATING STATION, UNITS 1 AND 2. Docket Nos.- 50-352 And 50-353..(Philadelphia Electric Company)
NOREG-1026: FINAL ENVIRONMENTAL-STATEMENT RELATED TO THE OPERATION OF BRAIDWOOD STATION UNITS 1 AND 2. Docket Nos. STN 50-456 And STN 50-457.(Commonwealth Edison Company)
Fission NUREC/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
NUREG/CR-3669: PLUTONIUM. RECYCLE TEST REACTOR (PRTR) ACCIDENT: A FINAL REPORT ON THE-INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
Flaws NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE' YEAR PLAN FY 1983-1987.
NUREC/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit 1 Nuclear Reactor.
NUREG/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Guarterly Rep ort:
October 1983 - March 1984.Vols 1 & 2.
Fluid-Thermal Mixing
- NUREC/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT FLUID-THERMAL MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT USING SOLA-PTS.
Fracture Mechanics NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN FY 1983-1987.
Fracture NUREC/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
NUREG/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE.
Fuel Bundles NUREG/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PRCGRAM: Postirradiation Examination Results For The-Third Materials Experiment (MT-3).
Fuel Cladding Interaction NUREC/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
Fuel Cycle NUREC/CR-3422 VO3; AEROSOL RELEASE AND TRANSPORT PROGRAM.Guarterly Progress Report For July-September 1983.
NUREC/CR-3682: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation of Existing Methods.
Fuel Damage NUREC/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
Fuel Performance Data Base NUREG/CR-3741 ' VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 2 Topical Report, Volume 1: Data Evaluat ion.
Fuel Rod NUREG/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST l-HI-4.
I NUREC/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER.
NUR EG/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT:A FINAL l
1(2 t-l
~.
(REPORT ON THE.INVESTIGATIONLOF FISSION PRODUCT CHEMICAL' FORMS.
NUREG/CR-3741 VO1: ' EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 2 Topical Report. Volume 1: Data Evaluat ion.
c Fuel.
NUREG/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draf t Report Of The USNRC PCI. Review Group.
GPU-vfB&W-NUREG-1020LD VO1: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al.'v.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1,
Docket 50-289.
NUREG-1020LD.VO2: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et 41.
v.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit'1, Docket 50-289.
Gamma-Radiation Exposure NUREG/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION EXPOSURE RATES AND SOIL 266RA CONCENTRATIONS.
Gas Conductivity-NUREG/CR-3680: RELATIONSHIP BETWEEN THE GAS CONDUCTIVITY.AND GEOMETRY
'OF A NATURAL FRACTURE.
Geochemical Respcnse NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
~
Geochemistry NUREC/CP-OOS2: NRC NUCLFAR WASTE MANAGEMENT GEOCHEMISTRY
'83.
Geologic: Disposal NUREG/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
Geologic Repository t
NUREC/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPGSITORY DESIGN - TUFF.
Geology j-NOREG/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
German Standard Problem 4A NUREC/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PRCBLEM 4A: PIPING RESPONSE TO BLOWDCWN.
Gravity F191d NOREC/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN TnE' REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
I Ground Motion NOREG/CR-3755: - STRONG GROUND M01 ION STUDIES FOR SOUTH CAROLINA EARTHGUAKES.
NUREG/CR-3805: ENGINEENING CHARACTERIZATION OF GROUND MOTION. Task I: Ef f ec ts Of Characteristics Of Free-Field Motion On Struc tural Response.
l NUREG/CR-3839: AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR l'
A 6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
Ground-Water Contamination NUREC/CR-3681: MITIGATIVE TECHNIQUES AND ANALYSIS OF GENERIC SITE l
CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE I
ACCIDENTS.
HLW Disposal NUREC/CP-OO52: NRC NUCLFAR WASTE MANAGEMENT GEOCHEMISTRY
'83.
Health Effects l
NOREG/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR HLW DOSIMETRY (ICRP-30),I.
l C
l 110 I
o 1
~.mm,_m
.,_,,_,,,._,,~,__..m...__.....m.
._.,.~...y
H2ct Flux NUREG/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
Heater Rod NUREG/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL HEATER ROD DESIGN ON GUENCH BEHAVIOR.
Heavy-Section Steel Technology Program NOREG/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN FY 1983-1987.
Heissdampfreaktor
. NUREG/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
High Pressure Injection NUREG/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER REACTOR.
High-Energy NUREC/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
High-Level Waste NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
NUREC/CR-3316: VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TRANSPORT MODEL (SWIFT).
NUREG/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, April 1983 - April 1984.
NUREC/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR HLW DOSIMETRY (ICRP-30),I.
Human Error NUREG/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
Human Factors NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATGR INTERFACES IN NUCLEAR POWER PLANTS.
Human Reliability NUREC/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS)
MODEL:
SUMMARY
DESCRIPTION.
Hydraulic Conductivity NUREG/CR-3680: RELATIONSHIP 3ETWEEN THE GAS CONDUCTIVITY AND GEOMETRY OF A NATURAL FRACTURE.
Hy o r o d g r.ami c Conditions NUREC/CR-3410: CHMONE: A ONE-DIMENGIONAL COMPUT ER CODE FOR SIMULATING TEMr)ERATURE. FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
Hydrogeologic Site Classifications NU3EC/CR-3681: MITIGATIVE TECHNIGUES AND ANALYSIS OF GENERIC SITE CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS.
Hydrological Response I
NUR EG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
IE Information Notice 83-42 NUREG/CR-3754: FAILURE EVALUATION OF GENERAL ELECTRIC SB-1 AND SB-9 REACTOR MODE SWITCHES.
IGECC NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED l
STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
l IPRDS NUREG/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results.
111
Ico Candensor NUREO/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Base Assessment And Suggested Experimen tal Program.
Implementation NUREG/CR-2803: IMPROVED FIELD EXPERIMENTAL DESIGNS AND GUANTITATIVE EVALUATION OF AGUATIC ECOSYSTEMS.
In Situ Procedure NUREC/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EQUIPMENT IN CURRENTL Y OPERATING PLANTS.
In Vivo Examination NUR EC/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS.
In-Plant. Reliability Data System NUREG/CR-3650: A STATISTICAL ANALYfiIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results.
In-Vessel Anomalies NUREC/CR-3303: USE OF NEUTRON NOISE FOR DIAGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
Independent Assessment Program NUREG/CR-3329 VO4: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report October-December 1983.
Inspection NUREC-OO40 VO8 NO1: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report January 1984 - March 1984.(White Book)
NUREG/CR-3604: BOLTING APPLICATIONS.
Intergranular. Stress Corrosion Cracking NUREG/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
Iodine NUR EG/CR-3514: THE CHEMICAL BEHAVIOR OF IODINE IN AGUEOUS SOLUTIONS UP TO 150 C.An Experimental Study of Nonredox Conditions.
Ion Exchange Media NUREG/CR-3383: IRRADIATICN EFFECTS ON THE STORAGE AND DISPOSAL OF Rt.DWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
Iontring Radiation NUR EG/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATICN.
Iron Oxide Particles NUREC/CR-3023: MOLTEN THCRMITE TEEMING INTO AN IRON OXIDE PARTICf.E BED.
t Irradiated Reactor Fuel NUREC-0725 RO4: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
?rradiation NUREC/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE COSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B & A333 3 Ref erence Plates From P3F Simulated Surveillance & Through-Wall Irradiation Capsules.
NUREG/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
NUREC/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF WELDS.
NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN FY 1983-1987.
NUREC/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
J-Integral Tearing Instability NUR EC/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE.
J-R Curve 112
NUREG/CR-3506: > J-R CURVE. CHARACTERIZATION 'OF IRRADI ATED LOW UPPER SHELF
- WELDS.
Joint Licensing Process
-NUREO-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
- LER' l
NUREG/CR-2OOO VO3 N3: LICENSEE EVENT REPORT'(LER) COMP'LATION: For Month Of March 1984.
NUREG/CR-2OOO VO3 N4: LICENSEE EVENT REPORT (LER) COMPILATION:For Month
~
di Of April 1984.
NUREO/CR-2OOO ' VO3 N5: LICENSEF EVENT REPORT-(LER) COMPILATION: For Month Of May 1984.
LOCA.
NUREG/CR-3350: LOCA. SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PROGRAM: Postirradiation Examination Results For The Third
, Materials Experiment (MT-3).
_ NUREG/CR-3511 VO1: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume 1. Main Report.
NOREC/CR-3588: THE EFFECT.0F.LOCA SIMULATION PROCEDURES ON CROSS-LINKED POLYOLEFIN CABLE'S PERFORMANCE.
I NUREG/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: -Model i
Description.
NUREG/CR-3633 VO2: TRAC-BD1/ MOD 1; AN ADVANCED BEST ESTIMATE ~ COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 2: Users Cuide.
NUREG/CR-3633 VO3: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TR ANSIENT ANALYSIS. Volum'e 3: Code 3
~
Structure and Programming Information.
1 NUREG/CR-3639: LARGE BREAM LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION.
1 I
LOFT I
NUREG/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL k
HEATER ROD DESIGN ON OUENCH BEHAVIOR.
j NUREG/CR-3608: PELAP5 ASSESSEMENT: LOFT Large Break L2-5.
l' Large Break LOCA 7
NUREG/CR-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH
{
}-
UPPER-PLENUM INJECTION.
j Large Break Transient l
NUREG/CR-3608: RELAPS ASSESSEMENT; LCFT Large Break L2-5.
O Lawsuit i
NUREO-1020L D VO1: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON i
TMI-1. Ceneral Public Utilities Corporation, et al.
v.
The Babcock &
Wilect Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket i
50-289.
}
NUREG-1020LD VO2: CPU V.
B&W LelSCIT REVIEW AND ITS EFFECT pN TMI-1. General Public Utilitie s C 2rporatione et al.
v.
The Babcock &-
Wilcox Company,et al.Three Mi.e Island Nuclear Station, Unit 1, Docket i
50-289.
Leak l-NUREC-1056:-REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM GENERATORS.
NUREC/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT 2
{
' R I SK.
l 4
Legislation-NUREG-0980: NUCLEAR REGULATORY LEGISLATION.
i
-Licensed Operating' Reactors NUREC-OO2O VOB NO3: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of February 29,1984.(Greg Book) r i
NUREC-OO20. VOS N05: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of April 30,1984.(Grey Book) l NUREC/CR-3604: BOLTING APPLICATIONS.
f 4
113 1
.-__-..___....__-.--~._________...___._,__-~..-...-.,;
t i
.Li c anoso : Evsnt - Rep ort NUREC/CR-2OOO.VO3 N3: LICENSEE EVENT REPORT (LER)' COMPILATION: For Month
-Of March.1984.
NOREO/CR-2OOO' VO3 N4: LICENSEE EVENT REPORT-(LER) COMPILATION: For Month Of April-1984.
NUREO/CR-2OOO VO3 N5: LICENSEE' EVENT REPORT.(LER) COMPILATION:For Month Of May 1984.
-Licensing NUREO/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING 2AND TRAINING:Part I - The Need For Plant-Reference Simulatorst Part II - The Use Of Plant-Reference Simulators.
. L'ightning
-NUREG/CR-3759: LIGHTNING STRIKE DENSITY FOR THE CONTIGUOUS UNITED
, STATES FROM THUNDERSTORM DURATION RECORDS.
Liquid Metal Fast. Breeder Reactor NUREO/CR-3644: REVIEW OF PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS.
Local' Control, Stations-1 NUREG/CR-3696: POTENTI AL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND ' OPERATOR' INTERFACES IN NUCLEAR POWER.
i
. P LANTS.
Loose-Part Monitoring NUREG/CR-3687: LOOSE-PART MONITORING. PROGRAMS AND RECENT OPERATIONAL
^
EXPERIENCE IN SELECTEn U.S.
AND WESTERN EUROPEAN COMMERCIAL NUCLEAR POWER STATIONS.
Loss-Of-Coolant-Accident NUREG/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR ~ PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (MT-3).
NUREG/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER i
PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model j-Description.
NUREC/CR-3633 VO2: TRAC-BD1/ MOD 1:AN ADVANCED BEST ESTIMATE COMPUTER j
PROGRAM FOR BOILINO WATER REACTOR TR ANSIENT ANALYSIS. Volume 2: Users Guide.
NUREG/CR-3633 VO3: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE CGMPUTER i
PROORAM FOR BOILING WATER REACTOR TRANSIENT Ar.ALYSIS Volume 3: Code Structure and Programming Information.
j NUREO/CP-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION.
?
NUREG/CP-3749: COBRA-NC POST-TEST PREDICTIONS FOR HDR CONTAINMENT STEAM BLOWDOWN TEST V44 (INTERNATIONAL STANDARD PROCLEM 16).
F Loss-Of-Fluid Test NUREO/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND FLECTRICAL HEATER ROD DESIGN ON GUENCH BEHAVIGR.
l Low Altitude Photography NUREG/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR OBTAINING SITE CHARACTERISTIC INFORMATION.
[
lt Low Enriched Uranium NUREO-1065: ACCEPTANCE CRITERI A FOR THE LOW ENRICHED URANIUM REFORM AMENDMENTS.
l Low Upper Shelf Energy l
NUREO/CR -3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF l
WELDS.
Low-Level Waste i
l NUREO/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM 1
REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
NUREO/CR-3681: MITIGATIVE TECHNIGUES AND ANALYSIS OF GENERIC SITE i
CONDITIONS FOR OROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE
).
ACCIDENTS.
4 I
114 i
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NORE@/CR-3797: DIGMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES l
IN SAMPLING COMMERCIAL LOW-LEVEL. WASTE SITES FOR RADIONUCLIDE SPILLS
-OR MIGRATION.-
NUREC/CR-3838: AN INITIAL REVIEW OF SEVERAL METEOROLOGICAL MODELS SUITABLE FOR LOW-LEVEL WASTE DISPOSAL FACILITIES.
MAPPS NUREO/CR-3626 VO1:. MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS)
[
MODEL: '
SUMMARY
DESCRIPTION.
MINET.
-NUREG/CR-3603: MINET VALIDATION SURVEY USINO EBB-II TEST DATA.
MOD 1 l
. NUREG/CR-3329 VO4:. THERMA'L/ HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report October-December 1983.
NUREG/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER l
PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model i
Description, l-NUREC/CR-3633 VO2: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER l
PROGRAM FOR BOILILO WATER REACTOR TR ANSIENT ANALYSIS. Volume 2: Users
. Ou i d e.
NUREO/CR-3633 VO3: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code
. Structure and Programming Information.
Maintenance Personnel. Performance Simulation NUREC/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS) l f
MODEL:
SUMMARY
DESCRIPTION.
I
' Malfunctions NUREC/CR-3754: FAILURE EVALUATION OF GENENAL ELECTRIC SB-1 AND SB-9 REACTOR MODE SWITCHES.
Management System NUREO-1055: ' IMPROVING QUALITY AND THE ASSURANCE OF QUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCI AL NUCLEAR POWER PLANTS. A Report To Congress.
r j
Mark 1 j
j NUREG/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN 4
BWRS-COMPUTER MODELING REGUIREMENTS.
NUREG/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN
]-
BNRS-COMPUTER MODELINO REGUIREMENTS, Mark III r
l NOREO/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN BWRS-CGMPUTER MODELING REQUIREMENT 3.
j Materials Deformation NOREC/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL
}
REACTOR PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (MT-3).
I Measurement
.NUREC/CR-3488 VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
NUREO/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING 3
RADIATION.
4 Mechanical Degradation NUREC/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CABLE FAILURE WITH MECHANICAL DECRADATION.
i Mechanical 3esponse-i NUREO/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
I Meteorological Model i
NUREO/CR-3838: AN INITIAL REVIEW OF SEVERAL METEOROLOGICAL MODELS SUITABLE FOR LOW-LEVEL WASTE DISPOSAL FACILITIES.
Meteorology NUREO/CR-3488. VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
i NUREG/CR-3773: VARIATICN OF PLANETARY BOUNDARY LAYER DISPERSION i
115 I
~
_=
_ _ PROPERTIES-WITH-HEIGHT.IN UNSTABLE CONDITIONS.
M3thodology NUREC/CR-3630: EGUIPMENT QUALIFICATION METHODOLOGY RESEARCH: TESTS OF.
. PRESSURE SWITCHES.
NUREG/CR-3756:. SEISMIC HAZARD CHARACTERIZATION OF THE ' EASTERN UNITED i
STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
Migration.
NUREC/CR-3572: DETERMINATION OF METABOLIC. DATA APPROPRI ATE FOR HLWI DOSIMETRY:(ICRP-30),I.
]
Mitigation
~
L NUREO/CR-3681: 'MITICATIVE TECHNIQUES AND ANALYSIS OF GENERIC SITE
' CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE
-ACCIDENTS.
NUREO/CR-3797:.DIGMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES:
IN SAMPLING COMMERCIAL LOW-LEVEL' WASTE SITES FOR RADIONUCLIDE' SPILLS OR MIGRATION.
Monitoring
. ACOUSTIC l EMISSION MONITORING OF HOT. FUNCTIONAL :
NUREG/CR-3693:
TESTING. Watts Bar Unit.1 Nuclear Reactor.
NUREG/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR 4.
IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report:
October 1983 - March 1984.Vols 1 & 2.
Multichannel Core Hydraulics 4
NUREC/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD. O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR j ;
SYSTEM TRANSIENTS.
1 Mussel NUREO/CR-3054: CLOSEOUT OF IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING '
WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND
=MYTILUS SP.
(MUSSEL).
Mutual Jurisdiction l
NUREG-1052: FEDERAL / STATE COOPERATION IN THE LICENSING OF A NUCLEAR POWER PRCJECT. A Jo in t P r o c e s s Cetueer. Thc U.O.
Nuclear Reg ulatory l
Comm1rsion And The Washington State Energy Facility Sita Evaluation
- Council, i
NDE NUREC/CR-3337 VO3: REAC TOR SAFETY RESEARCH PROGRAMS. Guarterl y Report July-Se p ternt er 1983.
i NUREC/CR-3GO7 VCC REACTOR SAFETY RESEAPCH PROGRAMD. Guar terly Report October-becember 1933.
4 NUP EG /C R-3,'43.
THE Ir. PACT OF NDE UNRELIABILITY ON PRESSURE VESSEL FRACTURE PREDICTIONS.
NEPA 1
NUREG-1052: FEDERAL / STATE COOPERATION IN THE LICENSING OF A NUCLEAR 7
POWER PROJECT.A Joint Process Between The U.S.
Nuclear Regulatory Commission And The. Washington State Energy Facility Site Evaluation 1
Council.
f NPPAP l
NUREO/CR-3684: NUCLEAR POWER PLANT ALARM PRIORITIZATION (NPPAP) PROGRAM STATUS. REPORT. January 1,1983 to September 31,1983.
NPRDS I
NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.
NWFT/DVM NUREG/CR-3378: VERIFICATION OF THE NETWORK FLOW AND l
TRANSPORT / DISTRIBUTED VELOCITV METHOD (NWFT/DVM) COMPUTER CODE.
i National Environmental Policy Act NUREG-1052: FEDERAL / STATE COOPERATION IN THE LICENSING OF A NUCLEAR l
POWER PROJECT.A Joint Process Between The U.S.
Nuclear _ Regulatory e
116 1
...__._-u__,__.
Ccemincion-And The Washington Stato Energy Fccility-Site Evaluation Council.
Natural Fracture NUREC/CR-3680: RELATIONSHIP BETWEEN THE GAS CONDUCTIVITY.AND GEOMETRY l
OF A NATURAL FRACTURE.
Network Flow & Transport NUREC/CR-3378: VERIFICATION OF THE NETWORK. FLOW AND-TRANSPORT / DISTRIBUTED VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
Neutron Exposure NURE0/CR-3391 VO2: LWR PRESSURE VESSEL SURVEILLANCE. DOSIMETRY IMPROVEMENT PROGRAM. Quarterly Progress Report, April 1983 - June 1983.-
Ne0 tron Irradiation NUREG/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY ~ IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
Neutron Kinetics l
NUREG/CR-3664: A DESCRIPTION AND ASSESSMENT OF R AMONA-3B MOD. O. CYCLE 4:
I A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
i Neutron Noise NUREG/CR-3303: USE OF NEUTRON NOISE FOR DIAGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
New Madrid Seismotectonic Study 1
NUREG/CR-3768: NEW MADRID SEISMOTECTONIC STUDY: Activities During Fiscal i'
Year.1982.
Nondestructive Examination NUREG/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Rep ort July-September 1983.
NUREC/CR-3307 VO4: REACTOR 9AFETY RESEARCH PROGRAMS. Guarterl y Report October-December 1933.
NUREG/CR-3743: THE IMPACT OF NDE UNRELIABILITY ON PRESSURE VESSEL FRACTURE PREDICTIONG.
1 I
Notch Ductility NUREC/CR-3295 V01: LIGHT WATER REACTOR PF7SEURE VESDEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility &-Fracture Toughness i
j Degradation of A302-B & A533-E Ref erence Plates From PSF Simulated i
Surveillance & Through-Wall Irraoxation Capsules.
NUREC/CR-3296 VO2: LIGHT HATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: oostirradiation Notch Ductility &
Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
Nuclear Plant Reliability Data System NUREG/CR-3637-THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.
Nuclear Waste Disposal NURE0/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report, Nuclear Waste NUREG/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEDLOGIC DISPOSAL OF NUCLEAR WASTE.
Numerical Diffusion NUREC/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
Operating Experience NUREC-1063: STEAM GENERATOR OPERATING EXPERIENCE UPDATE 1982-1983.
Operating Reactors Licensing Actions NUREG-0748 VO4 NO2: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As-Of February 29,1984.(Orange Book) 117
. _ - - ~
'NUREO-0748.VO4 NO3: OPERATING REACTORS -LICENSING ACTIONS
SUMMARY
iData
-As.Of-March 31,1984.L(Orange Book).
- NUREO-0748 VO4 NO4: ~ OPERATING REACTORS LICENSING ACTIONS
SUMMARY
.. Data -
As-Of, April _-OO,1984.
(OrangefBook)
-Operators 4
NUREC/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR
. DEVELOPING' CRITERIA.
NUREG/CR-3725: NUCLEAR POWER PLANT. SIMULATORS FOR. OPERATOR LICENSING j
AND TRAINING:Part I - - The Need For Plant-Reference Simulators Part j
II~
.The-Use Of Plant-Reference Simulators.
1
- 0rgan. Doses
'NOREG/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR'HLW
. DOSIMETRY-(ICRP-30)
I.'
PEB NUREO/CR-3637: THE APPLICATION OF. STEIN AND RELATED PARAMETR IC :
. EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.-
)
.PRA NUREG/CR-3511 VO1: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF -
THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume 1. Main Report.
t-PRIME-NUREC/CR-3606: NUCLEAR POWER PLANT CONTROC ROOM CREW TASK' ANALYSIS -
DATABASE: SEEK SYSTEM. (Users Manual).
PTS NUREG/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER REACTOR.
NURE0/CR-3743i THE IMPACT OF NDE UNRELIABILITY ON PRESSURE VESSEL FRACTURE. PREDICTIONS.
Parametric Empirical Bayes.
I NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC j
EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY ' DATA SYSTEM.
Pellet-Cladding Interaction NUREC/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterig Report July-September 1983.
NUREG/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report Octooer-December 1983.
NUREG/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT; Draf t Report Of Th e USNRC PCI Review Group.
- ~
NUREO/CR-3810_VO1: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report January-March 1984.
Penetration NUREC/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
Petitions For Rulemaking r
NUREG-0936 VO3 NO1: NRC REGULATOPY AGENDA.Guarterly Report, January-March 1984.
Pipe. Inspection NUREG/CR-3753: AN EVALUATION OF MANUAL ULTRASONIC INSPECTION OF CENTRIFUGALLY CAST STAINLESS STEEL PIPING.
i
' Pipe Whip Analysis P
NUREC/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.'
NUREC/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF
. PIPING SYSTEMS. Part A - User 's Manual.
NUREG/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF -
PIPING. SYSTEMS.Part B - Theorg Manual.
Piping Analysis
}
NUREG/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING ' SYSTEMS. Part C - Programmer 's Manual.
118
f-NURE@/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - verification Manual.
l Piping System NUREG/CR-3722: DAMPING TEST RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND i
8-INCH UNPRESSURIZED PIPES.
Piping NUREG/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES OF A 351 CF8A CAST STAINLESS STEEL IN LWR ENVIRONMENT.
NUREC/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NUREC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NUREG/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part C - Progranner 's Manual.
NUREG/CR-3686 VO4: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual, NUR EG/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING -
STATUS REPORT.
NUREG/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
NUREG/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE.
Planetary Boundary Layer NUREG/CR -3773: VARIATION OF PLANETARY BOUNDARY LAYER DISPERSION PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS.
Plugging NUREG-1056: REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM GENERATORS.
Plutonuim Recycle Test Reactor NOREC/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT:A FINAL l
REFORT ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
Pollen NURE0/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
Post-Critical Heat Flux NUREG/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
Postirradiation Examination NUREC/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report January-March 1984.
Postirradiation NUREG/CR-3295 VO1: LIGHT WATER REACTOR FRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B & A533-D Ref erence Plates From PSF Simulated Surveillance & Through-Wall Irradiation Capsules.
NUREC/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
Postulated Accidents NOREG/CR-3567: TRAC-PF1: AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
Pressure Boundary NUREC/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit i Nuclear Reactor.
Pressure Switches NUREG/CR-3630: EQUIPMENT GUALIFICATION METHODOLOGY RESEARCH: TESTS OF PRESSURE SWITCHES.
i Pressure Vessel 119
d NURED/CR-3295.VO1:-LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture. Toughness -
- Degradation of A302-B~& A533-B Ref erence Plates From PSF Simulated Surveillance &~Through-Wall 1 Irradiation Capsules.
NUREQ/CR-3295.VO2: LIGHT-WATER REACTOR PRESSURE VESSEL' SURVEILLANCE.
DOSIMETRY) IMPROVEMENT; PROGRAM: Postirradiation Notch Ductility &
Tensile ' Strength ' Determinations For PSF Simulated. Surveillance ik Through-Wall Specimen Capsules.
NUREC/CR-3391 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY 41 IMPROVEMENT PROGRAM. Guarterly Progress Report, April 1983 - June 1983.
NUREC/CR-3391 - VO3:. LWR PRESSURE VESSEL SURVEILL'ANCE DOSIMETRY.
i IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
i NUREG/CR-3391'VO4: LWR ~ PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1983-December.
1983.:
H NUREO/CR-3567: TRAC-PF1: AN ADVANCED BEST-ESTIMATE. COMPUTER ' PROGRAM FOR
. PRESSURIZED WATER ' REACTOR ANALYSIS.
NUREO/CR-3595: ' HEAVY-SECTION STEEL TECHNOLOGY PROGRAM'- FIVE YEAR PLAN FY'1983-1987.
NUREG/CR-3743: -THE IMPACT OF NDE UNRELIABILITY ON PRESSURE VESSEL-l FRACTURE PREDICTIONS.
NUREC/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR,
IN-SERVICE MONITORING OF. NUCLEAR PRESSURE VESSELS. Guarterly Report:
i-October-1983 - March 1984.Vols 1 & 2.
= Pressure
.NUREG/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERIALS USED FOR l
HIGH-LEVEL WASTE PACKAGING. Annual Report, April 1983 - Ap r i l 1984.
i NUREC/CR-3949: TWO-PHASE 3X3 ROD BUNDLE. TEST FACILITY FOR POST-CRITICAL i-HEAT FLUX BOILING.
Pressurized Thermal' Shock-NUPEO/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST' COMPUTER CODE SIMULATION OF THERMAL MIXING. IN THE DOWNCOMER OF ' A PRESSURIZED WATER
{
REACTOR.
NUREO/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERE W7TH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
NUREG/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT-FLUID-THERMAL' MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT g
USING SCLA-PTS.
NUREG/CR-3743-THE IMPACT OF NDE UNRELIABILITY ON PRESSURE VESSEL i
FRAcrVRE PREDICTIONS.
l Primary Coolant System Boundary NUREG/CR-3644: REVIEW-OF PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS.
j.
Probabilistic Risk Analysis NUREG/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC SAFETY STUDY: PLANT ANALYSIS.'
NUREC/CR-3507: AN ANALYSIS OF THE NRC SAFETY GOALS FOR NUCLEAR POWER.
Probabilistic Risk Assessment NUREC/CR-3511 VO1: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF
.THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume 1. Main' Report.
E Probability Theory NUREG/CR-3628: PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT i
STRUCTURES.
- Procedures NOREC/CR-3391 VO2: LWR PRESSURE '/ESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report. April 1983 - June 1983.
NUREC/CR-3391 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
NUREG/CR-3391 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY
[
IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1983-December 1.
k^
120 1
2 u.
.u___,..____._..__
l l
1983.
_1 L
Pump Failure NUREC/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE. RATE VARIABILITY - Some Preliminary Results.
Quality. Assurance NUR EC/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION.
Guality Control NUREG/CR-3360: COMPUTER PROGRAM CDCID: AN AUTOMATED GUALITY CONTROL PROGRAM USING CDC UPDATE.
Quality and Assurance NUR EG-1055: IMPROVING GUALITY AND THE ASSURANCE OF GUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCIAL NUCLEAR POWER PLANTS.A Report To Congress.
Guench Behavior NUREG/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL
-HEATER ROD DESIGN ON GUENCH BEHAVIOR.
RAMONA-3B NUREC/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD.O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
REFCO-83 NUREG/CR-3800: REFCO-83 USER 'S MANUAL.
RELAPS NUREG/CR-3329 VO4: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report October-December 1983.
NUREC/CR-3608: RELAPS ASSESSEMENT: LOFT Large Break L2-5.
RELAP5/ MOD 1 NUREG/CR-3608: RELAPS ASSESSEMENT: LOFT Large Break L2-5.
Radiation Exposure NUREG/CR-2675 VO4: RELFVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
Radiation Monitoring NUREG-OS37 VO3 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORV.. Progress Report September-December 1983.
Radioactive Material NUREC/CR-2907 VO2: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1981.
4 Radioactive Particles NUR EG/CR-3697: LABORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL i
OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
Radiological Emergency NUREG-1028: RUPTURED CESIUM-137 WELL-LOGGING SOURCE AT SHELWELL SERVICES,INC., HEBRON OHIO.
Radionuclide Migration NUREC/CR-36 31: MITIGATIVE TECHNIQUES AND ANALYSIS OF GENERIC SITE CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS.
Radionuclides NUREG/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /
Radionuclide Transport Model FETRA.
NUREG/CR-2424 VO2: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE I
TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing f or FETRA.
NUREG/CR-2675 VO4: RELFVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
. NUR EG /CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED BONE-SEEKING RADIONUCLIDES.
NUREG/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR HLW DOSIMETRY (ICRP-30),I.
121
1 NOREG/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL ;
RODS CONTAINING WATER.
NUREG/CR-3797: DIGMAN:A. COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES-IN SAMPLING COMMERCIAL' LOW-LEVEL WASTE. SITES FOR RADIONUCLIDE SPILLS OR MIGRATION.
Radon Emission NUREC/CR-3533: RADON ATTENUATION HANDBOOK FOR URANIUM-MILL TAILINGS
. COVER DESIGN.
j Radon Fluxes 1
+
NUREG/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION EXPOSURE RATES AND SOIL 266RA CONCENTRATIONS.
Radwaste t
. NUR EG/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
Random-Sampling NUREG/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S GUIDE FOR THE GENERATION 1
OF LATIN HYPERCUBE AND RANDOM St.MPLES FOR USE WITH COMPUTER MODELS.
Reaction Products r
NUREG/CR-2921: CHEMICAL' INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS.
Reactor Cavity NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE' REACTOR C AVITY SUBCOMPARTMENT ANALYSES.
Reactor Core NUREG/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT: A FINAL REPORT.ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
Reactor Mode Switches NUR EG/CR-3754: FAILURE EVALUATION OF GENERAL ELECTRIC SB-1 AND SB-9 9
REACTOR MODE EWITCHES.
Reactor Pressure Boundary l
NUREC/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report July-September 1983.
NUREO/CR-3025 V01-02: ACOU 3 TIC EMISFION/ FLAW RELATIONSHIP FOR i
IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Gearterly Report:
October 1983 - March 1984.Vols 1 & 2.
o Reactor Safety Research NUREG/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Heport j
October-December 1033.
l NUREC/CR-3810 VO1: HEACTOR SAFETY RESEAPCH PROGRAMS. Guerterly Report January-March 1984.
I Reactor Safety NUREG-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
NUREG/CR-2552: CRAC2 MODEL DESCRIPTION.
4 NUREG/CR-2679 VO4: ADVANCED REACTOR SAFETY RESEARCH, GUARTERLY REPORT, OCTOBER-DECEMBER 1982.
NUREG/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report July-September 1983.
Reelfoot Lake NUREG/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE i
REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
Reform Amendments i
NUREG-1065: ACCEPTANCE CRITERIA FOR THE LOW ENRICHED URANIUM REFORM AMENDMENTS.
Regulatory And Technical Report NUREG-0304 VO9 NO1: REGULATORY AND TECHNICAL REPORTS. Comp ila tion For First Guarter 1984.
Release NUREG/CR-2907 VO2: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1981.
NUREG/CR-34E9 VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement Data.
122
NUREC/CR-3600i ' DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST
-HI-4.
NUREG/CR-3669; PLUTONIUM RECYCLE TEST' REACTOR (PRTR) ACCIDENT: A FINAL REPORT. ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
i.
.Roliability RELIABILITY ASSESSMENT OF' INDIAN POINT UNIT 3 NUREG/CR-3641:
CONTAINMENT STRUCTURE.
'NUR EG/CR-3652: EVALUATION OF ' INSTRUMENTATION FOR DETECTION OF INADEGUATE CORE COOLING IN BOILING WATER REACTORS.
"NUR EG/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING -
~
STATUS REPORT.
Remote. Sensing
- NUREC/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIGUES FOR J;
OBTAINING SITE CHARACTERISTIC INFORMATION.
Ropair-Welded Stainless Steel 3
NOREG/CR-3613:. EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rep t for 1983.
l
. Repair-NUREG/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT l
. MATERIAL.
Repository Design NOREG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TLFF.
Repository j_
NUREG/CR-3316: VERIFICATION AND FIELD COMPARISON CNr THE SANDIA 1
WASTE-ISOLATION FLOW AND TRANSPORT MODEL'(SWIFT).
]
Residual Radioactivity
{
NUREG/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADI ATION EXPOSURE fr RATES AND SOIL 266RA CONCENTRATIONS.
Rotrieval NUREC/CR-3489-. ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC j
DISP 0 SAL OF NUCLEAR WASTE.
t-Risk Assessn.ent l
NUREG/CR-3422 VO3: AEROSCL RELEASE AND TRANSPORT PROGRAM.Guerterly f.
Progress Report ~or July-September 1983.
NUR EC/CR-3682: NUCLEAR FUEL CYCLE RI3K ASSESSMENT: Review and Evaluation of Existing Methods.
i NUREC/CR-3713: GRCUPING OF LIGHT WATER REA0 TORS FOR EVALUATION OF DECAY i
HEAT REMOVAL CAPABILITY.
i l
Risks l
NUR EC/CR-3507: AN ANALYSIS OF THE NRC EAFETY GOALS FOR NUCLEAR POWER.
i NUREC/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT t
- RISK, NUREG/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
Rock Fracture NUREG/CR-3680: RELATIONSHIP'BETWEEN THE GAS CONDUCTIVITY AND GEOMETRY iI.
OF A NATURAL FRACTURE.
~ Rod Bundle NUREG/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL i
HEAT FLUX BOILING.
Rulemaking NUREC/CP-OO52: NRC NUCLEAR WASTE MANAGEMENT GEOCHEMISTRY
'83.
Rules-
[
NUREG-0936 VO3 NO1: NRC REGULATORY AGENDA.Guarterly Report, January-March 1984.
Rupture i
NUREG-1028: RUPTURED CESIUM-137 WELL-LOGGING SOURCE AT SHELWELL l
SERVICES,INC., HEBRON, OHIO.
NOREC-1056: REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM GENERATORS.
NUREG/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL i
L 123 L
REACTOR PROGRAM: Poetirradiction Excmination Romults For Tha Third Materials Experiment (MT-3).
NUREC/CR-3669: PLUTONIUM RECYCLE TEST REACTOR (PRTR) ACCIDENT: A FINAL REPORT ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
SA508-2 Steel NUREG/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT MATERIAL.
SAINT NUREC/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
SDLOCA NUREG/CR-3748: COURA/ TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
SEEK NUREG/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATABASE: SEEK SYSTEM.(Users Manual).
SETS NUREG/CR-3134: A SETS USER 'S MANUAL FOR VITAL AREA ANALYSIS.
SLAM NUREG/CR-3379: SLAM - A SODIUM-LIMESTONE CONCRETE ADLATION MODEL.
SOLA-PTS NUREG/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT FLUID-THERMAL MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT USING SOLA-PTS.
SSC NUREG/CR-3603: MINET VALIDATION SURVEY USING EDD-II TEST DATA.
SSEL NUREG-0525 RO9: SAFEGUARDS
SUMMARY
EVENT LIST (SSEL).
STA NUREG/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON SHIFT.
SWIFT NUREG/CR-3316 VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TRANSPORT MODEL (SWIFT).
Safeguard Summary Event List NUREG-0525 RO9: SAFEGUARDS
SUMMARY
EVENT LIST (SSEL).
Safety Evaluation Report NUREG-0420 GOS-SAFETY EVALUATION REFORT RELATED TO THE OPER ATION OF SHOPEHAM NUCLEAR POWER STATION, UNIT NO.
- 1. Docket No. 50-322.(Lcng Island Lighting Company)
NUREG-0675 S23: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2.Oocket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)
NUREG-0776 SO7: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SUSGUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-387 And 50-388.(Pennsylvania Power And Light Company, Allegheny Electric Cooperative,Incorporateo)
NUREG-0787 SO6: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF WATERFORD STEAM ELECTRIC STATION, UNIT 3. Docket No.
50-382.
(Louisiana Power And Light Company)
NUREC-0830 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF CALLAWAY PLANT, UNIT NO.1. Doc k e t No. 50-483.(Union Electric Company)
NUREG-0853 S03: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF CLINTON POWER STATION, UNIT NO.1. Docket No.
50-461.(Illinois Power Company,et al)
NUREG-0876 SO4: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF THE BYRON STATION, UNITS 1 AND 2. Docket Nos. STN 50-454 And STN 50-455.(Commonwealth Edison Company)
NUREG-0892 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF WPPSS NUCLEAR PROJECT NO.
- 2. Docket No. 50-397.(Washington Public Power Supply System) 124
j
~
y y
/
?-
~ :
NUREG-0954 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPER ATION OF C ATAWBA ' NUCLEAR STATION, UNITS 1-AND 2. Doc k e t Nos. 50-413 And 50-414.
(Duke Poweiv Company,. et :al._)
NOREC-0989: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF RIVER BEND STATION. Docket No.-
50-458. (Gulf States Utilities Company, Cajun Electric Power Cooperative)
NUREC-1038 SOi: SAFETY EVALUATION REPORT-RELATED TO THE OPER ATION OF SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1. Docket No. STN 50-400.
~
(Carolina Power And Light Company, North Carolina Eastern Municipal Power Agency)
- NUREG-1051 :. SAFETY EV'A UATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR :THE RESEARCH REACTOR AT THE. UNIVERSITY OF KANSAS. Docket No. 50-148.
.NUREG-1059: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE
-OPERATING LICENSE FOR THE UNION CARBIDE. SUBSIDIARY B,INC. RESEARCH REACTOR.DocketrNo. 50-54.
Safety Goals NUREC/CR-3507A AN ANALYSIS OF THE.NRC SAFETY GOALS FOR NUCLEAR POWER.
Safety. System. Components
,# 7 NUREG/CR-3054: CLOSEOUT OF. IE BULLETIN 81-03: FLOW BLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS BY CORBICULA SP.
(ASIATIC CLAM) AND-
_MU3SEL).
(
MYTILUS SP.
Safety-Related< Equipment NUREC/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POL /MER;PROPERITIES.
- Safety
,f e-NUREG-0828: INTEGSATED PLANT SAFETY ASSESSMENT REPORT SYSTEMATIC EVALUATION PROQMAM. Big Rock Point Plant. Docket No. 50-155.
(Consumers Power Company)
NUREG/CR-36T2: EVALUATION OF INSTRUMENTATION FOR DETECTION OF
~
INADEGUATE CORE COOLINC IN BOILING WATER REACTORS.
Sa6dia Waste-Isolation Flou And Tpansport NUREC/CR-3314: VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TR.ANSPORT MODEL (SWIFT).
Sediment c
NUREG/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS.Vol 1: Testing Of The ' Sediment /
Radionuclide-Transport Model FETRA.
NbREG/CR-2424 V02: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT -IN C0ASTAL WATERS. V 2 User's M CP Listing f or FETRA.
~
Seismic Hazard Characterization NUREC/CR-3756f SEISMIC HAZARD CHARACTERIZATION OF THE CASTERN UNITED STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
Seismic Stress NUREO/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING.-
STATUS REPORT.
Seismographic; NUREC/CR-3755: STRONG GROUND MOTION STUDIES FOR SOUTH CAROLINA-EARTHOUAKES.
Set Equation Transformation System NUREC/CR-3134:
A' SETS USER'S MANUAL FOR VITAL AREA ANALYSIS.
Severe Accident Sequence Anal'gsis NUREG/CR-3S96: SEVERE ACCIDENT SEGUENCE ANALYSIS (SASA) PROGRAM SEGUENCE EVENT TREE: BOILING WATER REACTOR ANTICIPATED TRANSIENT WITHOUT SCRAM.
Severe Core Damage NOREC/CR-3762: IDENTIFICATION OF EQUIPMENT AND COMPONENTS PREDICTED AS SIGNIFICANT CONTRIBUTORS TO SEVERE CORE DAMAGE.
Severe Fuel Damage 125
NUREC/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly R pert July-S3ptonbGr 1983.
NUREC/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report October-December 1983.
NUR EO/CR-3810. VO1 : REACTOR SAFETY RESEARCH PROGRAMS. Guarterl y Report January-March 1984.
Severe Reactor Accidents
. NUR EC-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
Shift Engineer NUREG/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON 3HIFT.
' Shift Technical Advisor NUREC/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPER TISE ON SHIFT.
Shipment NUREG-0725 RO4: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
Shutdown Decay Heat Removal NUREC/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY, Simulation NUREG/CR-2940: REALISTIC SIMULATION 'F SEVERE ACCIDENTS IN BWRS-COMPUTER MODELING REQUIREMENTS.
NUREC/CR-3748: COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
Simulator Fidelity NUREG/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:Part I - The Need For Plant-Reference Simulators; Part II - The Use Of Plant-Reference Simulators.
NUREC/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A COMPREHENSIVE BIBLIOGRAPHY WITH SELECTED ANNOTATIONS.
]
Sinulator NUREC/CR-3848: EXPERIMENTAL INVEGTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES.
Sma l l -Br eal. Lo s s-Of-Co o laat-Ac c i d en t NU7EC/CR-3748; COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
Snubbers NUREC/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING -
STATUS REPORT.
Socioeconomic Consequences NUREC/CR-3566: SOCIOECONOMIC CONS 7GUENCES OF NUCLEAR REACTOR ACCIDENIS.
Sodium-Limestene Ablation Model NURE0/CR-3379: SLAft - A SODIUM-LIMESTONE CONCRETE ABLATION MODEL.
Solution Chemistry NUREG/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERI ALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report. April 1983 - April 1984.
Specifications 4
NUREC/CR-3604: BOLTING APPLICATIONS.
Spent Fuel NUREG-0725 RO4: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
Stainless Steel NURE0/CR-3546: THE TEMPERATURE der'ENDENCE OF FATIGUE CRACK GROWTH RATES OF A 351 CFBA CAST STAINLESS STEEL IN LWR ENVIRONMENT.
Standard Review Plan r
NUREG-0800 03.9.3 R1: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 1 To Section 3.9.3,Appendir A.
NUREG-0800 03. 9. 4 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 3.9.4,
" Control Rod Drive Systems 126
E
-w NUREC-02OO 05.4.'6'R3: ! STANDARD; REVIEW PLAN FOR 'HE REVIEW OF SAFETY
. ANALYSIS REPORTS FOR: NUCLEAR POWERLPLANTS. LWR Edition.'Rovision 3 To b
'Section 5;4.6;1" Reactor-CoreLIsolation Cooling' System ~ ( B WR ). "
l NUREG-0800-05.'4.'7 R3: STAN0ARD REVIEW PLAN FOR THE REVIEW OF SAFETY
(
. ANALYSIS: REPORTS FOR' NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To l
Section15.4.7,. " Residual Heat Removal.(RHR). System."
NUREG-0800 06. 3 R2: STANDARD REVIEW PLAN'FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To
-Section 6.3,
" Emergency Core Cooling' System.'"
NUREG-0800 :09. 2.1 R3: STANDARD REVIEW PLAN FOR'THE REVIEW OF SAFETY ANALYSIS REPORTS FOR. NUCLEAR POWERLPLANTS. LWR Edition. Revision No.
3' To Section 9.2.1,
" Station Service Water System."
NUREG-0800 09. 2. 2 R2: STAN0ARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR' POWER PLANTS. LWR Edition. Revision 2 To
- Section 9.2.2,
" Reactor Auxiliary Cooling. Water Systems."
NOREC-0800 10. 3 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No.3 To Section 10 3i " Main Steam Supply System."
' NUREC-0800 10. 4. 7 R3:. STANDARD REVIEW PLAN FOR THE REVIEW. 0F SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR-Edition Revision 3 To Section.10.4.7,." Condensate And Fe edwater Sy s tem" And BTP. ASB 10-2,
-" Design Guidelines For Avoiding Water' Hammer.. ' "
Standby' Safety Systems
_,II APPLICATIONS TO' STANDBY SAFETY SYSTEMS.-
NUREC/CR-3627: FRANTIC Steam Ge'nerator NUREG-1056: REPORT ON U.'S. -JAP AN 1983 MEETINGS ON STEAM GENERATORS.
NUREC-1063i' STEAM GENERATOR OPERATING EXPERIENCE UPDATE 1982-1983.
~
NOREG/CR-32OO VO4:. EDDY-CURRENT' INSPECTION FOR STEAM GENERATOR TUBING PROGRAM ' ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMDER 31, 1983.~
NUREG/CR-3603: MINET VALIDATION SURVEY USING EBB-II TEST, DATA.
Steel. Pipe NOREC/CR-3740: J-INTEGRAL TEARING INSTABILITY-ANALYSIS FOR 8-INCH DIAMETER ASTM:A106 STEEL PIPE.
Steel NUR EG /CR-3366:- HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANI A-COATED STEEL.'
NUREC/CR-3672: EXAMINATION OF THE SIZE EFFECTS AND DATA SCATTER OBSERVED IN-SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
Stein Estimates
'NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.
Stratification NUREO/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
Stress Analysis NUREG/CR-3653: CONTAINMENT ANALYSIS TECHNIGUES. A State-Of-Th e-Art Summary.
- Stress' Corrosion Cracking NUREG/CR-3604: DOLTING APPLICATIONS.
~StructuralEAnalysis NUREC/CR-3686CWIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF.
' PIPING SYSTEMS. Summary Report.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF.
PIPING SYSTEMS. Part A - Us'er 's Manual.
NUREG/CR-3686'VO2: ~ WIPS--COMPUTER' CODE' FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B,
-Theory Manual.
NUREC/CR-3686~VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING ' SYSTEMS. Part C - Programmer 's Manual.
WIPS--COMPUTER CODE FOR WHIP AND IMP ACT ANALYSIS.0F
.y NUREG/CR-3686 VO4:
2 127 4
~
PIPING SYSTEMS.Part D
. Verification Manual; p.
- NUREC/CR-3722: DAMPING TEST RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND 8-INCH:UNPRESSURIZED PIPES.
- Structural' Dynamic Response NUREC/CR-3720: PREDICTION AND EXPERIMENT COMPARISGNS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
- m Structure NUREC/CR-3628:. PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT
-STRUCTURES.
Summary Information 1
L NUREG-0871;VO3 NO1:
SUMMARY
INFORMATION REPORT. Data As Of De c ember 31,1983.'(Brown Book)-
r Super System Code-NUREG/CR-3603: MINET VALIDATION SURVEY USING EBB-II TEST DATA.
Surface Boundary Lager j
NUREG/CR-3773: VARI ATION OF' PLANETARY. BOUNDARY LAYER L DISPERSION PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS.
Surveillance NUREG/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL-SURVEILLANCE DOSINETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation ' of A302-B & A533-B Ref erence Plates From PSF Simulated Surveillance & Through-Wall Irradiation Capsules.
NUREG/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE
' DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
t-
. Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
NUREG/CR-3391 VO2:. LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY i
IMPROVEMENT PROGRAM. Guarterly Progress Report, April 1983 June.1983.
NUREG/CR-3391 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY s'
IMPROVEMENT PROGRAM. Annua l ' Rep or t,0c tober 1,1982-September 30,1983.
NUREG/CR-3391 VO4:. LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1983-December i
1983.
Survey NUREC/CR-3769: DESCRIPTION AND SIGNIFICANCE OF.THE. GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
i l
. Systematic Evaluation Program NUREG-0828: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM. Big' Rock Point Plant. Docket No.
50-155.'
(Consumers Power Company)
P TEM EST NUREG/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE j
SIMULATION OF THERMAL MIXING IN THE.DOWNCOMER OF-A PRESSURIZED WATER REACTOR.
TLD 4
NUREG-0837 VO3 NO4: NRC.TLD DIRECT RADIATION MONITORING NETWCRK. Progress Report September-December ~ 1983.
TMI Action Plan Requirements NUREG-1066: COMPARISON OF-IMPLEMENTATION OF SELECTFD TMI ACTION PLAN REGUIRFMENTS' ON OPERATING PLANTS DESIGNED BY BABCOCK AND WILCOX.
' TRAC-BD1 NUREG/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER l-PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model
. Descrip tion.
' NUREG/CR-3633 VO2: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 2: Users Guide.
NUREC/CR-3633 VO3: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER
}-
PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code 4
. Struc ture and Programming - Inf ormation.
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---*n--
- v--
TRAC-PF1 NUREG/CR-3329 VO4: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM. Quarterly Report October-December 1983.
NUREG/CR-3567: TRAC-PF1: AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
NUREG/CR-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION.
TRAC NUREG/CR-3567: TRAC-PF1:AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
Tesk Analysis NUREG/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATAB ASE: SEEK SYSTEM. (Users Manual).
Tochnical Specifications NUR EG-1058: TECHNICAL SPECIFICATIONS FOR CALLAWAY PLANT, UNIT NO.
1.
Docket No. STN 50-483.(Union Electric Company)
Tollurium vapors NUREC/CR-2921: CHEMICAL INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS.
Tamperature Melt NOREG/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANIA-COATED STEEL Tamperature NUREG/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (MT-3).
NUREC/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, April 1983 - April 1984.
NUREG/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES OF A 351 CF8A CAST STAINLESS STEEL IN LWR ENVIRONMENT.
NUREG/CR-3600: DATA
SUMMARY
REPORT.FOR FISSION PRODUCT RELEASE TEST HI-4.
NUREG/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
NUREG/CR-3652: EVALUATION OF INSTRUMENTATION FOR DETECTION OF INADEGUATE CORE COOLING IN BOILING WATER REACTORS.
NUREG/CR-3672: EXAMINATION OF THE SIZE EFFECTS AND DATA SCATTER OBSERVED IN SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
NUREC/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
Test NUREG/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
NUREG/CR-3310: TESTING OF THE CONTAIN CODE.
Thernal Conditions NUREG/CR-3410: CHMONE:A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
Thermal Response NUREG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
Thermal Shield NUREG/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B & A533-B Reference Plates From PSF Simulated Surveillance & Through-Wall Irradiation Capsules.
NUREG/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
129
4
.Tcnoilo' Strength Datorminctions For PSF Simulated Surveillence-&
Through-Wall Specimen Capsules.
. Thermal-Shock-NUREC/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR. PLAN FY 1983-1987.
U Thermal Stress NUREO/CR-3718: RELIABILI'TY' ANALYSIS OF STIFF.VERSUS FLEXIBLE PIPING -
STATUS REPORT.
i
. Thermal-Hydraulic
'NUREC/CR-3350: LOCA SIMULATION-IN THE NATIONAL'RESEARCH UNIVERSAL
+
- REACTOR PROGRAM: Postirradiation Examination Results For The Third-
. Materials Experiment (MT-3).
NUREO/CR-3567i - TRAC-PF1: AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ~ ANALYSIS.
V Thermal NUREC/CR-3629: 1 THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION l
-PROCEDURES ON POLYMER PROPERITIES.
' Thermal / Hydraulic Response 3
NUREC/CR-3608: RELAPS ASSESSEMENT: LOFT Large Break L2-5.
F
- Thermal / Hydraulic j
NUREC/CR-3329 VO4: THERMAL / HYDRAULIC. ANALYSIS RESEARCH PROGRAM.Guarterly Report October-December 1983.
NUREC/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.-
NUREG/CR-3505: A VOLUME-NEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
Thermite Melt NUREC/CR-3023:' MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE-BED.
i Thermite
'NUREC/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANIA-COATED STEEL.
Thermocouples E TER ROD E GN O GUENCH BE V OR
- Thernoluminescence Dosimetry System i
i NUREC/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION.
Thermoluminescent Dosimeter j
' NUREC-0837 VO3 NO4: NRC'TLD DIRECT RADIATION MONITORING i
NETWORKJProgress Report September-December 1983.
.Thermomechanical History NUREC/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
Through-Wall Toughness NUREG/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & ' Fracture Toughness Degradation of A302-B & A533-B Ref erence Plates From PSF Simulated i.
Surveillance & Through-Wall Irradiation Capsules.
i NUREC/CR-3295. VO2: LIGHT WATc;R REACTOR PRESSURE VESSEL SURVEILLANCE I
DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
j Tensile Strength Determinations For PSF Simulated Surveillance &
l Through-Wall Specimen Capsules.
l Tidal Estuaries NUREG/CR-3410: CHMONE: A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
I' LTime-Dependent Reliability Analysis l
NUREC/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
l.
Title List i
NUREG-0540 VO6 NO1: TITLE. LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. January 1-31,1984.
NOREC-0540 VO6 NO2; TITLE LIST OF DOCUMENTS MADE PUBLICLY 130-
. :AVAILABLE.Fobrucry 1-29,1984.
NUREG-0540 VO6 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY l-AVAILABLE. March 1-31, 1984.
' NUREG-0540 VO6 NO4: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. April 1-30,1984.
Tornado-NUREC/CR-3670: VIOLENT TORNADO CLIMATOGRAPHY, 1880-1982.
- NUREC/CR-3848: EXPERIMENTAL INVESTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES.
. Toxic Response NUREC/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION. FATE AND EFFECTS OF COPPER Training Simulator I
NURE0/CR-3515: SAFETY-RELATED OPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
Training NURE0/CR-3632: METHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING PROCEDURES.
NUREC/CR-3725: NUCLEAF POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:Part I - The Need For Plant-Reference Simulators; Part II'- The Use Of Plant-Reference Simulators.
NUREC/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A COMPREHENSIVE BIBLIOGRAPHY WITH SELECTED ANNOTATIONS.
Transient Reactor Analysis Code NURE0/CR-3567: TRAC-PF1: AN ADVANCED DEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
Transient NUREG/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD.O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
Transport NUREC/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /
Radionuclide Transport Model FETRA.
NUREC/CR-3422 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM.Guarterly Progress Report For July-Sep tember 1983.
Tube Degradation NUREC-1063: STEAM CENERATOR OPERATING EXPERIENCE UPDATE 1982-1983.
d Tubes NUREG-1056: REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM CENERATORS.
NUREG/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
Tuff NUREC/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH -INFLUENCE REPOSITORY DESIGN - TUFF.
Turbulence Models NUREC/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
Two-Phase Cooling NUREC/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (MT-3).
j Underground Test Facility NUREC/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
'nderpressurized Pipe UREO/CR-3722: DAMPING TEST RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND 8-INCH UNPRESSURIZED PIPES.
' Unresolved Safety Issue A-46 NUREC/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
131
Unrocolved Safety Issuos NUREC-0606 VO6 NO2: UNRESOLVED SAFETY IS3UES
SUMMARY
. Data As10f May 18,
'1984.
(Aqua' Book)
Upper-Plenum Injection NUREC/CH-3639: LARGE BREAK LOCA ANALYSES FOR.TWO-LOOP PWRS WITH' UPPER-PLENUM ' INJECTION.
Urania-Coated Steel NUREC/CR-3366:'HIGH TEMPERATURE MELT ATTACK ON. STEEL AND URANIA-COATED STEEL.
. Uranium Mill Tallings F
NUREO/CR-3533: RADON ATTENUATION HANDBOOK FOR URANIUM-MILL TAILINGS I
COVER DESIGN.
E-NOREC/CR-3697: LABORATORY TESTING OF CHEMICAL' STABILIZERS FOR. CONTROL OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
-Uranium Urinalysis NOREC/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN. VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS.
Uranium NUREG/CR-3745:. BIOLOGICAL-CHAR ACTERIZATION OF RADI ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress R e p o r t: Ap r i l ' 1,,1982 - Mar c h 31,1783.
User's Guide NUREG/CR-3624:. A FORTRAN 77 PROGRAM AND USER 'S GUIDE FOR THE GENERATION OF LATIN HYPERCUBE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
User's Manual-l NUREC/CR-3134:. A SETS USER 'S MANUAL 'FOR VITAL AREA ANALYSIS.
NUREC/CR-3800: REFCO-83 USER 'S. MANUAL.
Vessel Cladding NUREC/CR-3743: THE IMPACT OF NDE UNRELIABILITY ON PRESSURE VESSEL FRACTURE PREDICTIONS.
b Vessel
{'
NUREC/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF p
WELDS.
Vital Area Analysis NUREG/CR-3134: A SETS USER'S MANUAL FOR VITAL AREA ANALYSIS.
WIPS NOREC/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.
NUREG/CR-3686 VO1: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's. Manual.
NUREG/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF i
PIPING' SYSTEMS.Part B - Theorg Manual.
1 NUREC/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part C - Programmer's Manual.
NUREC/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND. IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
Washington State Environmental Policy Act NUREC-1052: FEDERAL / STATE COOPERATION IN THE LICENSING OF A NUCLEAR POWER' PROJECT.A Joint Process Between The U.S.
Nuclear Regulatory-Commission And The Washington State Energy Facility Site Evaluation Council.
Waste Management l
NUREC/CP-OOS2: NRC NUCLEAR WASTE MANAGEMENT GEOCHEMISTRY
'03.
Waste Package NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
NOREC/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERIALS USED FOR j
. HIGH-LEVEL WASTE PACKAGING. Annual Report, April 1983 - April 1984.
j Waste. Repository i
i 132
r NUREC/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT
'FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES.. Final Report (Task 5)-June 1981 - February 1983.
NUREC/CR-3572:. DETERMINATION OF METADOLIC DATA APPROPRIATE FOR HLW DOSIMETRY (ICRP-30),I.
W3ste Site NUREC/CR-3797: DIGMAN:A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR'RADIONUCLIDE SPILLS
.OR' MIGRATION.
Water Bodies NUR EC /CR -3410: CHMCNE: A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHEMICAL CONCENTRATIONS IN WATER BODIES.
Water Hammer NUREC-0800 03.9.3 R1: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 1 To Section 3.9.3, Appendix A.
NUREC-0800 03. 9. 4 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY.
ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 3.9.4,
" Control Rod Drive Systems."
NUREC-0800 05.4.6 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section S. 4. 6,
" Reactor Core Isolation Cooling System (BWR)."
NUREC-0800 05. 4. 7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 5. 4. 7,
" Residual Heat Removal (RHR) System."
NUREC-0600 06.3 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 6. 3,." Emergency Core Cooling System."
NUREC-OBOO 09.2.1 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No. 3 To Section 9.2.1,
" Station Service Water System."
NUREC-0800 09. 2. 2 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Ed i t ion. R evi s ion 2 To Section 9. 2. 2,
" Reactor Auxiliary Cooling Water Systems."
NUREC-0800 10.3 R3: STANDARD REVIEW PLAN FOR - THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No.3 To Section 10.3,
" Main Steam Supply System."
NUREC-0800 10. 4. 7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 10.4.7,
" Condensate And Feedwater System" And DTP ASB 10-2,
" Design Guidelines For Avoiding Water Hammer.
Weather NUREG/CR-3759: LIGHTNING STRIKE DENSITY FOR THE CONTIGUOUS UNITED STATES FROM THUNDERSTORM DURATION RECORDS.
Weldment NUREC/CR-3593: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN FY 1983-1987.
Welds NUREC/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF WELDS.'
Whip And Impact Of Piping Systems NUREC/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.
NUREC/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part A - User 's Manual.
NUREC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NUREC/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part C - Programmer's Manual.
NUREC/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF 133
PIPING SYSTEMS. Part D -- Verifiection Manual.
!Yellowcake NUREC/CR-3745s BIOLOGICAL CHAR ACTERI ZATION ' OF RADI ATION. EXPOSURE AND DOSE' ESTIMATES FOR INHALED. URANIUM MILLING EFFLUENTS. Annua 1. Progress
' Report:. April 1.1982 - March 31,1983.
1 l
134 l
l
NRC Originating Organization Index (Staff Reports)
This index lists those NRC organizations that have published staff reports. The index is arranged alphabetically by major NRC organizations (e.g., program offices) and then by subsections of these (e.g., divisions, branches) where ap-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.
OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
REGION 1, OFFICE OF DIRECTOR NUREG-0837 VO3 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report September-December 1983.
DIVISION OF RADIOLOGICAL & MATERIALS SAFETY PROGRAMS NUREG-1028: RUPTURED CESIUM-137 WELL-LOGGING SOURCE AT SHELWELL SERVICES,INC., HEBRON, OHIO.
REGION 4.
OFFICE OF DIRECTOR NUREG-OO40 VO8 NO1: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report, January 1984 - March 1984.(White Book) 54 EDO - OFFICE OF ADMINISTRATION DIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL NUREG-0304 VO9 NO1: REGULATORY AND TECHNICAL REPORTS. Compilation For First Guarter 1984.
NUREG-0540 VO6 NO1: TITLE LIST OF DOCUt'ENTS MADE PUBLICLY AVAILABLE. January 1-31,1984.
NUREG-0540 VO6 NO2: TITLE LIST OF DOCUf1ENTS MADE PUBLICLY AVAILABLE. February 1-29,1984.
NUREG-0540 VO6 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. March 1-31, 1984.
NUREG-0540 VO6 NO4: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. April 1-30,1984.
NUREG-0750 V17: NUCLEAR REGULATORY COMMISSION ISSUANCES. January-June 1983. Pages 1-1,196.
NUREG-0750 V18 102: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. July-December 1983.
NUREG-0750 V18 N06: NUCLEAR REGULATORY COMMISSION ISSUANCES. Dec emb er 1983.Pages 1,303-1,482.
NUREG-0750 V19 NO1: NUCLEAR REGULATORY COMMISSION ISSUANCES. January 1984. Pp 1-485.
NUREG-0750 V19 NO2: NUCLEAR REGULATORY COMMISSION ISSUANCES. February 1984. Pp 487-554.
DIVISION OF RULES AND RECORDS 135
NUREG-0936 VO3 NO1: NRC REGULATORY AGENDA. Quarterly Report, January-March 1984.
EDO - OFFICE OF EXECUTIVE LEGAL DIRECTOR OFFICE OF THE EXECUTIVE LEGAL DIRECTOR i
NUREG-0980: NUCLEAR REGULATORY LEGISLATION.
EDO - OFFICE FOR-ANALYSIS & EVALUATION OF OPERATIONAL DATA DIRECTOR 'S OFFICE -
NUREG-OO90 VO6 NO3: - REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. July-September 1983.
NUREG-OO90 VO6 NO4: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. October -December 1983.
OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)
~ DIRECTOR'S OFFICE, OFFICE OF INSPECTION AND ENFORCEMENT NUREG-0940 VO3 NO1: ENFCRCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED.Guarterly Progress Report (January - March 1984).
GA BRANCH NUREG-1055: : IMPROVING GUALITY AND THE ASSURANCE OF GUALITY IN THE DESIGN AND CONSTRUCTION OF COMMERCIAL NUCLEAR POWER PLANTS. A Report To Congress.
OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS DIVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG-1071: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SOURCE MATERIAL LICENSE NO. SUD-526. Docket No. 40-3392.(A111ed Chemical 5
Company UF6 Conversion Plant)
NUREG-1077: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-21. Docket No. 70-25.
(Energy Systems Group Rockwell International Corporation)
NUREG-1078: ENVIRONMENTAL IMPACT APPRAISAL FOR RENEWAL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-1097. Docket No. 70-1113. (General Electric Company.Wilmington Manufacturing Department)
DIVISION OF SAFEGUARDS NUREG-0725 RO4: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
NUREG-1065: ACCEPTANCE CRITERIA FOR THE LOW ENRICHED URANIUM REFORM AMENDMENTS.
LICENSING POLICY & PROGRAMS BRANCH NUREG-0525 RO9: SAFEGUARDS
SUMMARY
EVENT LIST (SSEL).
U. S.
NUCLEAR REGULATORY COMMISSION NRC - NO DETAILED AFFILIATION GIVEN NUREC/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draft Report Of The USNRC PCI Review Group.
OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81) 136
i.
DIVISION OF HEALTH,~SITIN3 & WASTE MANAGEMENT NUREG/CP-OO52: NRC NUCLEAR WASTE MANAGEMENT CEOCHEMISTRY
'83.
DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429) j NUREG-1062: DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS.
EDO-RESOURCE MANAGEMENT OFFICE OF RESOURCE MANAGEMENT, DIRECTOR NUREG-1090:
U. S.
NUCLEAR REGULATORY COMMISSION 1983 ANNUAL REPORT.
DIVISION OF BUDGET & ANALYSIS NUREG-OO2O VOB NO3: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of February 29,1984.(Grey Book)
NUREG-OO2O VOB NO4: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As of March 31,1984.(Grey Book)
NUREG-OO2O VO8 N05: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of April 30,1984.(Grey Book)
MANAGEMENT INFORMATION BRANCH NUREG-0748 VO4 NO2: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of February 29,1984.(Orange Book)
NUREG-0748 VO4 NO3: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of March 31,1984.(Orange Book)
NUREG-0748 VO4 NO4: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of April 30,1984. (Orange Book)
NUREG-0871 VO3 NO1:
SUMMARY
INFORMATION REPORT. Data As Of December 31,1983. (Brown Book)
OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)
OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR NUREG-1020LD VO1: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation et al. v.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket 50-289.
NUREG-1020LD VO2: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al.
v.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1, Docket 50-289.
NUREG-1052: FEDERAL / STATE COOPERATION IN THE LICENSING OF A NUCLEAR POWER PROJECT.A Joint Process Between The U.S.
Nuclear Regulatory Commission And The Washington State Energy Facility Site Evaluation Council.
NUREG-1056: REPORT ON U.S.-JAPAN 1983 MEETINGS ON STEAM CENERATORS.
NUREG-1074: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF HOPE CREEK GENERATING STATION. Docket No. 50-354.(Public Ser.vice Electric And Oas Co And Atlantic City Electric Co)
DIVISION OF ENGINEERING NUREG-1063: STEAM GENERATOR OPERATING EXPERIENCE UPDATE 1982-1983.
DIVISION OF LICENSING NUREG-0420 S05: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SHOREHAM NUCLEAR POWER STATION, UNIT NO.
- 1. Docket No. 50-322.(Long Island Lighting Company)
NUREO-0675 S23: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 137
r And:50-323.(Pocific Gas An'd ElectricJCompeny)
.NUREC-0776 SO7: SAFETY-EVALUATION REPORT RELATED TO THE OPERATION -OF SUSGUEHANNA STEAM _ EL ECTRIC STATION, UNITS 1 AND 2. Doc k et Nos. 387
.And~50-388.(Pennsylvania: Power And Light Company, Allegheny
. Electric Cooperative, Incorporated)
. NUREG-0787 SO6: SAFETY EVALUATION REPORT RELATED "TO THE OPERATION OF WATERFORD STEAM ELECTRIC STATION, UNIT 3. Docket No. 50-382.
~,
'(Louisiana Power AndLLightLCompany)
{
y NUREG-0828: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC H
- EVALUATION PROGRAM. Big Roc k Point Plant. Doc ket No. 50-155.
(Consumers Power Company)
- NUREG-0830 SO3: - SAFETY EVALUATION REPORT. RELATED TO THE. OPERATION OF.
CALLAWAY PLANT, UNIT NO.~1. Docket No. 50-483.(Union Electric ~ Company)
NUREG-0853 SO3: : SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF
'CLINTON POWER STATION, UNIT PE). -l. Doc ket No. 50-461. (Illinois Power
)
Company,et'al) l NUREG-0876 -SO4: SAFETY EVALUATION REPORT RELATED. TO THE OPERATION OF THE BYRON STATION, UNITS.1_ AND 2. Dock et Nos. - STN 50-454 And : STN 50-455.(Commonwealth Edison Company)
NUREG-0892 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WPPSS NUCLEAR PROJECT-NO.
- 2. Docket No. 50-397.(Washington Public Power Supply System)
NUREG-0954 502: SAFETY EVALUATION. REPORT RELATED TO THE OPERATION OF CATAWBA NUCLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-414.
(Duke Power Company,et al.)-
NUREG-0974: FINALLENVIRONMENTAL STATEMENT-RELATED TO THE OPERATION OF' LIMERICK GENERATING STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-353.(Philadelphia Electric Company)
NUREG-0989: SAFETY ~ EVALUATION REPORT-RELATED TO THE OPERATION OF RIVER BEND STATION. Docket No. 50-458.(Gulf States Utilities Company, Cajun Electric. Power Cooperative)
NUREG-1026: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF BRAIDWOOD STATION, UNITS 1 AND 2. Docket Nos. STN 50-456 And STN 50-457.(Commonwealth Edison Company)
NUREG-1038 SO1: SAFEIY EVALUATION -REPORT RELATED TO THE OPERATION OF SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1. Docket No 'STN 50-400.
(Carolina Power And Light Company, North Carolina Eastern Municipal Power Agency)
NUREG-iO51: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR.THE RESEARCH REACTOR AT THE-UNIVERSITY OF' KANSAS. Docket No. 50-148.
NUREG-1058: TECHNICAL SPECIFICATIONS FOR CALLAWAY PLANT, UNIT NO.
1.
Docket No. STN 50-483.(Union Electric Company)
NUREG-1059: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE UNION CARBIDE SUBSIDIARY B, INC. RESEARCH REACTOR. Docket No. 50-54.
NUREG-1066: COMPARISON OF IMPLEMENTATION OF SELECTED TMI ACTION PLAN REQUIREMENTS ON OPERATING PLANTS DESIGNED BY BABCOCK AND WILCOX.
DIVISION OF SAFETY TECHNOLOGY NUREG-0606-VO6 NO2: UNRESOLVED SAFETY ISSUES
SUMMARY
. Da ta As Of May
' 18, 1984. (Aqua Book)
NUREG-0800 03.9.3 R1: STANDARD REVIEW PLAN FOR 'THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision'1 To Section 3.9.3, Appendix A.
NUREG-0800 03. 9. 4 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY
. ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS.' LWR Edition. Revision 2 To Section 3. 9. 4,' " Control Rod Drive Systems. " =
NUREO-OBOO 05. 4. 6 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY
' ANALYSIS REPORTS'FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 5.4.6,
" Reactor Core Isolation Cooling System (BWR). "
138 n~-.----
~ -. = - -. -. -.
. NURES-O!OO 05.- 4. 7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To Section 5.4.7,
" Residual Heat Removal.(RHR) System."
NUREG-0800 06.3 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To Section 6.3,
" Emergency Core Cooling System."
NUREG-0800 09.2.1 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY
.: ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No.
- 3 To Section 9. 2.1,. " Station Service Water System. "
NUREG-0800 09.2.2 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 2 To l
Section 9.2.2,
" Reactor Auxiliary Cooling Water Systems."
NUREG-OBOO 10.3 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision No. 3 To Section 10.3,
" Main Steam Supply System."
NUREG-OBOO 10.4.7 R3: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 3 To 1
Section 10.4.7,
" Condensate And Feedwater System" And BTP ASB 10-2,
" Design Cuidelines For Avoiding Water Hammer...
i 5
i l
)
139
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 FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA DIRECTOR'S OFFICE NUREG/CR-2000 V03 N3: LICENSEE EVENT REPORT (LER) COMP ILAT ION: For Month Of March 1984.
NUREG/CR-2000 V03 N4: LICENSEE EVENT REPORT (LER) CCMPILATION:For Month Of April 1984.
NUREG/CR-2000 V03 N5: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of May 1984.
OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)
DIVISION OF EMERGENCY PREPAREDNESS & ENGINEERING RESPONSE (POST 830103)
NUREG/CR-3054: CLOSEOUT OF IE DULLETIN 81-03: FLOW DLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS DY CORDICULA SP. (ASIATIC CLAM)
AND MYTILUS SP.
(MUSSEL).
NUREG/CR-3754: FAILURE EVALUATION OF GENERAL ELECTRIC SD-1 AND SD-9 REACTOR MODE SWITCHES.
OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS DIVISION OF WASTE MANAGEMENT NUREG/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - DOMAL SALT.
NUREG/CR-2614: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
NUREQ/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF DACKFILL PLACEMENT FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final Report (Task 5) June 1981 - February 1983.
NUREG/CR-3316: VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TRANSPORT MODEL (SWIFT).
NUREG/CR-3378: VERIFICATION OF THE NETWORK FLOW AND TRANSPORT / DISTRIBUTED VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
NUREG/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR THE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
NUREG/CR-3572: DETERMINATION OF METADOLIC DATA APPROPRI ATE FOR HLW DOSIMETRY (ICRP-30),I.
NUREG/CR-3774 V01: ALTERNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL 141
RADIOJ.CTIVE WASTES. Task 1: Description of Methods And Assessment Of Criteria.
OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)
OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR NUREG/CR-3769: DESCRIPTION AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
NUREG/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL EVENTS-DRAFT: Draf t Report Of The USNRC PCI Review Group.
DIVISION OF ACCIDENT EVALUATION NUREG/CR-2531 R02: INTRODUCTORY USER 'S MANUAL FOR THE U. S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA DANK.
NUREG/CR-2679 V04: ADVANCED REACTOR SAFETY RESEARCH.GUARTERLY REPORT, OCTOBER-DECEMBER 1982.
NUREG/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL HEATER ROD DESIGN ON QUENCH BEHAVIOR.
NUREG/CR-2921: CHEMICAL INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS.
NUREG/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS IN DWRS-COMPUTER MODELING REGUIREMENTS.
NUREG/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE BED.
NUREG/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report July-September 1983.
NUREG/CR-3307 VO4: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report October-December 1983.
NUREG/CR-3310: TESTING OF THE CONTAIN CODE.
NUREC/CR-3329 V04: THERMAL / HYDRAULIC ANALYSIS RESEARCH PROGRAM.Guarterly Report October-December 1983.
NUREG/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-3.
NUREG/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (MT-3).
NUREG/CR-3360: COMPUTER PROGRAM CDCID: AN AUTOMATED GUALITY CONTROL PROGRAM USING CDC UPDATE.
NUREG/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANIA-COATED STEEL.
NUREG/CR-3379: SLAM - A SODIUM-LIMESTONE CONCRETE ABLATION MODEL.
NUREG/CR-3410: CHMONE: A ONE-DIMENSIONAL COMPUTER CODE FOR SIMULATING TEMPERATURE. FLOW AND CHEMICAL CONCENTRATIONS IN WATER DODIES.
b NUREG/CR-3422 VO3: AEROSOL RELEASE AND TRANSPORT PROGRAM. Guarterly Progress Report For July-September 1983.
NUREG/CR-3504: TURDULENCE MODELING IN THE COMMIX COMPUTER CODE.
NUREG/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
NUREG/CR-3514: THE CHEMICAL DEHAVIOR OF IODINE IN AGUEOUS SOLUTIONS UP TO 150 C.An Experimental Study of Nonredox Conditions.
NUREG/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER REACTOR.
NUREG/CR-3567: TRAC-PF1: AN ADVANCED DEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
NUREG/CR-3596: SEVERE ACCIDENT SEGUENCE ANALYSIS (SASA) PROGRAM SEQUENCE EVENT TREE: DOILING WATER REACTOR ANTICIPATED TR ANSIENT WITHOUT SCRAM.
NUREG/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-4.
NUREG/CR-3603: MINET VALIDATION SURVEY USING EDD-II TEST DATA.
142
NUR20/CR-3608: RELAPS A22E!!EMENT: LOFT Lorgo Brock L2-5.
NUREQ/CR-3633 VO1: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PRDORAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1:
Model Description.
NUREQ/CR-3633 V02: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 2:
Users Guide.
NUREQ/CR-3633 V03: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 3: Code Structure and Programming Information.
NUREQ/CR-3664: A DESCRIPTION AND ASSESSMENT OF RAMONA-3B MOD. 0 CYCLE 4: A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
NUREQ/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
NUREQ/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TR ANSIENT FLUID-THERMAL MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT USING SOLA-PTS.
NUREG/CR-3741 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS C APAB ILITIES. Phase 2 Topical Report, Volume 1: Data Evaluation.
NUREG/CR-3748: COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
NUREQ/C R-3749: COBRA-NC POST-TEST PREDICTIONS FOR HDR CONTAINMENT STEAM BLOWDOWN TEST V44 (INTERNATIONAL STANDARD PROBLEM 16).
NUREG/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS. Quarterly Report January-March 1984.
NUREQ/CR-3839: AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR A 6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
NUREG/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
DIVISION OF FACILITY OPERATIONS NUREG/CR-3134: A SETS USER'S MANUAL FOR VITAL AREA ANALYSIS.
NUREQ/CR-3303: USE OF NEUTRON NOISE FOR DIAGNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
NUREQ/CR-3515: SAFETY-RELATED DPERATION ACTIONS: METHODOLOGY FOR DEVELOPING CRITERIA.
NUREQ/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS DATAB ASE: SEEK SYSTEM. (Users Manual ).
NUREG/CR-3684: NUCLEAR POWER PLANT ALARM PRIORITIZATION (NPPAP)
PROGRAM STATUS REPORT. January 1,1983 to September 31,1983.
NUREG/CR-3687: LOOSE-PART MONITORING PROGRAMS AND RECENT OPERATIONAL i
EXPERIENCE IN SELECTED U.S.
AND WESTERN EUROPEAN COMMERC I AL NUCLEAR POWER STATIONS.
DIVISION OF HEALTH, SITING & WASTE MANAGEMENT NUREG/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND RADIONUCLIDE TRANSPORT IN COASTAL WATERS. Vol 1: Testing Of The Sediment / Radionuclide Transport Model FETRA.
NUREG/CR-2424 V02: MATHEMATICAL SIMULATION OF SEDIMENT AND RADIONUCLIDE TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing for FETRA.
j NUREG/C R-2803: IMPROVED FIELD EXPERIMENTAL DESIGNS AND GUANTITATIVE j
EVALUATION OF AQUATIC ECOSYSTEMS.
NUREG/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
NUREQ/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF COPPER.
NUREQ/CR-3488 VO2: IDAHO FIELD EXPERIMENT 1981.Vol 1: Measurement 4
Data.
NUREG/CR-3533: RADON ATTENUATION HANDBOOK FOR URANIUM-MILL TAILINGS l
COVER DESIGN.
J 10 o
-NURE0/CR-3566: SOCIDECONOMIC CONSEGUENCE3 OF NUCLEAR-REACTOR ACCIDENTS.
. NUREG/C R-3583:- EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR OBTAINING SITE CHARACTERISTIC INFORMATION.
NUREG/CR-3670: VIOLENT TORNADO CLIMATOGRAPHY, 1880-1922.
NUREG/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION
-EXPOSURE RATES AND SOIL 266MA CONCENTRATIONS.
NUREG/CR-3680: RELATIONSHIP BETWEEN THE CAS CONDUCTIVITY AND GEOMETRY OF A NATURAL FRACTURE.
NUREG/CR-3681: MITIGATIVE TECHNIGUES AND ANALYSIS OF GENER IC SITE CONDITIONS FOR OROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS; NUREG/CR-3697: LABORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL DF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
NUREG/CR-3745: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual
' Progress Report: April 1,1902 - March 31,1983.
NURES/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED STATES:METHODOLDQY AND INTERIM RESULTS FOR TEN SITES.
NUMEO/CR-3759: LICHTNINC STRIKE DENSITY FDR THE CONTIQUOUS UNITED STATES FROM THUNDERSTORM DURATION RECORDS.
1 NUREQ/CR-3768: NEW MADRID SEISM 0 TECTONIC STUDY: Activities During Fiscal Year 1982.
NUREG/CR-3797: DIGMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPLING COMMERCIAL LOW-LEVEL WASTE SITES FOR ~
RADIONUCLIDE SPILLS OR MIGRATION.
NUREQ/CR-3800: PEFCO-83 USER 'S MANUAL.
DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429)'
NUREG/CR-2552: CRAC2 MODEL DESCRIPTION.
NUREG/CR-3507: AN ANALYSIS OF THE NRC SAFETY GOALS FOR NUCLEAR POWER.
NUREG/CR-3511 VO1: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume
- 1. Main Report.
NURE0/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S QUIDE FOR THE GENERATION OF LATIN HYPERCUBE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
NUREG/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
NUREG/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYSTEM.
NUREQ/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP FAILURE RATE VARIABILITY - Some Preliminary Results.
NUREG/CR-3653: CONTAINMENT ANALYSIS TECHNIQUES.A State-Of-The-Art Summary.
NUREQ/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
NUREQ/CR-3682: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation of Existing Methods.
NUREO/CR-3683: NUCLEAR ' FUEL CYCLE RISK ASSESSMENT: Program Summary Through Fiscal Year 1983.
DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST 840429)
NUREQ/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE ~ DISPOSAL: Phase I Final Report.
NUREQ/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS.
NURE9/CR-3427 VO4: LONO-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WAST 6 PACKAGING. Annual Report. April 1983 - April 1984.
NUREG/CR-3626 VO1: MAINTENANCE PERSONNEL PEEFORMANCE SIMULATION (MAPPS) MODEL:
SUMMARY
DESCRIPT10N.
NUREG/CR-3773: VARIATION OF PLANETARY BOUNDARY LAYER DISPERSION PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS.
t 144
i NUREG/CR-3775: QUALITY A3SURANCE FOR MEASUREMENTS OF IONIZING l
RADIATION.
NUREG/CR-3838: AN INITIAL REVIEW OF SEVERAL METEOROLOGICAL MODELS SUITABLE FOR LOW-LEVEL WASTE DISPOSAL FACILITIES.
NUREG/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA:A User's Guide For The Applicable Computer Programs In The Statistical Package For-Social Scientists (SPSS).
NUREG/CR-3848: EXPERIMENTAL-INVESTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES.
DIVISION OF ENGINEERING TECHNOLOGY NUREG/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR STEAM GENER ATOR TUBING PROGRAM AlJNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
NUREG/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE r
DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness Degradation of A302-B.& A533-B Reference Plates From PSF Simulated Surveillance'& Through-Wall Irradiation Capsules.
NUREG/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
NUREG/CR-3307 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report July-September 1983..
NUREG/CR-3391 VO2: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report, April 1983 - June 1983.
NUREQ/CR-3391 VO3: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Annual Report,0ctober 1,1982-September 30,1983.
NUREG/CR-3391 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Quarterly Progress Report,0ctober 1983-December 1983.
NUREG/CR-3506: J-R CURVE CHARACTERIZATION OF IRRADIATED LOW UPPER SHELF WELDS, NUREG/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT RISK.
NUREC/CR-3546: THE TEMPERATURE DEPENDENCE OF FATIGUE CRACK GROWTH RATES OF A 351 CFSA CAST STAINLESS STEEL IN LWR ENVIRONMENT.
NUREG/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDURES ON CROSS-LINKED POLYOLEFIN CABLE'S PERFORMANCE.
NUREQ/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REP AIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
NUREG/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CABLE FAILURE WITH MECHANICAL DEGRADATION.
NUREG/CR-3628: PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT I
STRUCTURES.
NUREQ/CR-3629: THE EFFECT OF THERMAL AND IRRADI ATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
NUREC/CR-3630: EQUIPMENT GUALIFICATION METHODOLOGY RESE#4CH: TESTS OF PRESSURE SWITCHES.
NUREG/CR-3641: RELIABILITY ASSESSMENT OF INDIAN POINT UNIT 3 CONTAINMENT STRUCTURE.
NUMEG/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER.
NUREG/CR-3672: EXAMINATION OF THE SIZE EFFECTS AND DATA SC ATTER OBSERVED IN SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
NUMEG/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.
NUREG/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT. ANALYSIS OF PIPING SYSTEMS.Part A - User's Manual.
NUMEC/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS i
0F PIPING SYSTEMS.Part B - Theory Manual, j
i um
\\
NUREQ/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS
- OF PIPING' SYSTEMS. Part C - Programmer 's Manual.
NUREG/CR-3686 VO4: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
NUREG/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL ~
TESTING. Watts Bar Unit 1 Nuclear Reactor.
NUREG/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING 2
STATUS REPORT.
NUREG/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
+
NUREG/CR-3722: DAMPING TEST RESULTS'FOR STRAIGHT SECTIONS OF 3-INCH AND.8-INCH UNPRESSURIZED PIPES.
NUREG/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE' l
(ESF) SYSTEMS. Data Base Assessment And Suggested Experimental Program.
j NOREG/CR-3743: THE IMPACT OF NDE UNRELIABILITY _0N PRESSURE VESSEL.
FRACTURE PREDICTIONS.
NUREG/CR-3753: AN EVALUATION OF MANUAL ULTRASONIC INSPECTION OF CENTRIFUGALLY CAST STAINLESS STEEL PIPING.
NUREG/CR-3762: IDENTIFICATION OF EQUIPMENT AND COMPONENTS PREDICTED AS SIGNIFICANT CONTRIBUTORS TO SEVERE CORE DAMAGE.
NUREG/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT MATERIAL.
NUREG/CR-3805: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task I:Ef f ects Of Charac teristics Of Free-Field Motion On Structural Response.
NUREG/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report January-March 1984.
NUREG/CR-3825 VO1-02: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR j
IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report:
October 1983 - March 1984.Vols 1 & 2.
}
EDU-RESOURCE MANAGEMENT
_,e i
DIVISION OF BUDGET & ANALYSIS NUREC/CR-2907 VO2: RADIDACTIVE MATERIALS RELEASED FROM NUCLEAR POWER
{
PLANTS. Annual Report 1981.
0FFICE OF NUCLEAR REACTOR RECULATION (POST 4/28/80)
I 0FFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR
^
NUREG/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL i
EVENTS-DRAFT: Draf t Report Of The USNRC PCI Review Group.
CLINCH RIVER BREEDER REACTOR PROGRAM OFFICE i
NUREG/CR-3644: REVIEW OF PROPOSED FAILURE CRITERIA FOR DUCTILE MATERIALS.
{
DIVISION OF ENGINEERING t
NUREG/CR-3604: BOLTING APPLICATIONS.
NUREG/CR-3755: STRONG OROUND MOTION ~ STUDIES FOR SOUTH CAROLINA EARTHGUAKES.
NUREG/CR-3756: SEISMIC HAZARD CHARACTERIZATION OF THE EASTERN UNITED STATES: METHODOLOGY 'ANO INTERIM RESULTS FOR TEN SITES.
. DIVISION OF HUMAN FACTORS SAFETY i
NUREG/CR-3632: METHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING Ph0CEDURES.
j NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER 2
'f i
148
r PLANTS.
NUREG/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND.TRAINIMO:Part I. - The ' Need For Plant-Ref erence Simulators; Part II - The Jse Of Plant-Reference Simulators, NUREG/CR-3/26: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A l'
COMPREHENSIVE BIBLIDORAPHY WITH SELECTED ANNOTATIONS.
I NUREG/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON SHIFT.
DIVISION OF SYSTEMS INTEORATION (POST 811005)
NUREG/CR-3305: COMPARI50N OF BEACON AND COMPARE REACTOR CAVITY SUBCOMPARTMENT ANALYSES.
NUREQ/CR-3535: AGE-DEPENDENT' DOSE-CONVERSION FACTORS FOR SELECTED BONE-SEEKING RADIONUCLIDES.
NUREC/CR-3639: LARGE BREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION.
l NUREG/CR-3652: EVALUATION OF INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING IN BOILING WATER REACTORS.
DIVISION OF SAFETY TECHNOLOGY i
NUREG/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC SAFETY STUDY: PLANT ANALYSIS.
NUREG/CR-3713: GROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY.
NUREG/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE.
NUREG/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC QUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
i 1
1 4
147
Contractor Index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports if further information is needed, refer to the main citation by the NUREG/CR number.
AMES LABORATORY, ENERGY & MINERAL RESOURCES RESEARCH INSTITUTE NUREG/CR-3653: CONTAINMENT ANALYSIS TECHNIGUES. A State-Of-Th e-Art Summary.
ANCO ENGINEERS. INC.
NUREC/CR-3720: PREDICTION AND EXPERIMENT COMPARISONS FOR GERMAN STANDARD PROBLEM 4A: PIPING RESPONSE TO BLOWDOWN.
ARGONNE NATIONAL LABORATORY NUREG/CR-3504: TURBULENCE MODELING IN THE COMMIX COMPUTER CODE.
NUREC/CR-3505: A VOLUME-WEIGHTED SKEW-UPWIND DIFFERENCE SCHEME IN COMMIX.
ARIZONA, UNIV. OF, TUCSON, AZ NUREC/CR-3680: RELATIONSHIP BETWEEN THE GAS CONDUCTIVITY AND GEOMETRY OF A NATURAL FRACTURE.
ARMY, DEPT. OF, ARMY ENGINEER WATERWAYS EXPERIMENT STATION NUREC/CR-3774 VO1: ALTERNATIVE METHODS FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTES. Task 1: Descrip tion of Methods And Assessment Of Criteria.
BABCOCK & WILCOX CO.
NUR EC/CR-3771: VESSEL V-7 AND V-8 REPAIR AND CHARACTERIZATION OF INSERT MATERIAL.
DATTELLE HUMAN AFFAIRS RESEARCH CENTERS NUREC/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:Part I - The Need For Plant-Reference Simulators; Part II - The Use Of Plant-Reference Simulators.
NUREO/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS:A COMPREHENSIVE BIDLIOGRAPHY WITH SELECTED ANNOTATIONS.
BATTELLE MEMORIAL INSTITUTE, COLUMDUS LADORATORIES NUREO/CR-3427 VO4: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAOING. Annual Report, April 1983 - April 1984.
7 NURE0/CR-3632: METHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING PROCEDURES.
BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORIES NURE0/CR-2424 VO1: MATHEMATICAL SIMULATION OF SEDIMENT AND RADIONUCLIDE TRANEPORT IN COASTAL WATERS.Vol 1: Testing Of The Sediment /
Radionuclide Transport Model FETRA.
NURE0/CR-2424 VO2: MATHEMATICAL SIMULATION OF SEDIMENT AND R ADIONUCLIDE TRANSPORT IN COASTAL WATERS. V 2 User 's M CP Listing f or FETRA.
NUREO/CR-2675 VO4: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL: Phase I Final Report.
NUREO/CR-2803: IMPROVED FIELD EXPERIMENTAL DESIGNS AND QUANTITATIVE 149
EVALUATION OF AGUATIC ECOSYSTEMS.
NUREC/CR-2955: ANALYSIS OF URANIUM URINALYSIS AND IN VIVO MEASUREMENT RESULTS FROM ELEVEN PARTICIPATING URANIUM MILLS.
NUREC/CR-3307 V03: HEACTOR SAFETY RESEARCH PROGRAMS. Guarterl y Report July-September 1983.
NUREC/CR-3307 V04: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report October-December 1983.
NUREC/CR-3350: LOCA SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR PROGRAM: Postirradiation Examination Results For The Third Materials Experiment (MT-3).
NURE0/CR-3533: RADON ATIENUATION HANDDOOK FOR URANIUM-MILL TAILINGS COVER DESIGN.
NUR EG/CR-3564: PRESSURIZED THERMAL SHOCK: TEMPEST COMPUTER CODE SIMULATION OF THERMAL MIXING IN THE DOWNCOMER OF A PRESSURIZED WATER REACTOR.
NUREO/CR-3566: SOCIOECONOMIC CONSEQUENCES OF NUCLEAR REACTOR ACCIDENTS.
NUREO/CR-3613: EVALUATION AND ACCEPTANCE OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Rept for 1983.
NUREC/CR-3632: METHODS FOR IMPLEMENTING REVISIONS TO EMERGENCY OPERATING PROCEDURES.
NUREC/CR-3669: PLUTONIUM REC"CLE TEST REACTOR (PRTR) ACCIDENT: A FINAL REPORT ON THE INVESTIGATION OF FISSION PRODUCT CHEMICAL FORMS.
NUREC/CR-3670: VIOLENT TORNADO CLIMATOGRAPHY, 1880-1982.
NUREO/CR-3677: COMPARISON OF RADON FLUXES WITH GAMMA-RADIATION EXPOSURE R ATES AND. SOIL 266RA CONCENTRATIONS.
NUREO/CR-3681: MITIGATIVE TECHNIQUES AND ANALYSIS OF GENERIC SITE CONDITIONS FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE ACCIDENTS.
NUREC/CR-3682: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Review and Evaluation of Existing Methods.
NUREC/CR-3683: NUCLEAR FUEL CYCLE RISK ASSESSMENT: Program Summary Through Fiscal Year 1983.
NUREC/CR-3693: ACOUSTIC EMISSION MONITORING OF HOT FUNCTIONAL TESTING. Watts Bar Unit 1 Nuclear Reactor.
NUREG/CR-3696: POTENTIAL HUMAN FACTORS DEFICIENCIES IN THE DESIGN OF LOCAL CONTROL STATIONS AND OPERATOR INTERFACES IN NUCLEAR POWER PLANTS.
NUREO/CR-3697: LADORATORY TESTING OF CHEMICAL STABILIZERS FOR CONTROL OF FUGITIVE DUST EMISSIONS FROM URANIUM MILL TAILINGS.
NURE0/CR-3725: NUCLEAR POWER PLANT SIMULATORS FOR OPERATOR LICENSING AND TRAINING:PART I - The Need For Plent-Reference Simulators: Part II - The Use Of Plant-Reference Simulators.
NURE0/CR-3726: SIMULATOR FIDELITY AND TRAINING EFFECTIVENESS: A COMPREHENSIVE DIDLIOGRAPHY WITH SELECTED ANNOTATIONS.
NURE0/CR-3727: FISSION PRODUCT REMOVAL IN ENGINEERED SAFETY FEATURE (ESF) SYSTEMS. Data Dase Assessment And Suggested Experimental Program.
NURE0/CR-3743: THE IMPACT OF NDE UNRELIADILITY ON PRESSURE VESSEL I
NUREC/CR-3748:
FRACTURE PREDICTIONS.
COBRA / TRAC SIMULATION OF SEMISCALE S-UT-5 TEST.
NUREO/CR-3749: COBRA-NC POST-TEST PREDICTIONS F3R HDR CONTAINMENT STEAM DLOWDOWN TEST V44 (INTERNATIONAL STANDARD PROBLEM 16).
NUREC/CR-3753: AN EVALUATION OF MANUAL ULTRASONIC INSPECTION OF CENTRIFUGALLY CAST STAINLESS STEEL PIPING.
NUREC/CR-3785: ALTERNATIVE APPROACHES TO PROVIDING ENGINEERING EXPERTISE ON SHIFT.
NUREG/CR-3797: DICMAN: A COMPUTER PROGRAM TO ILLUSTRATE THE COMPLEXITIES IN SAMPLING COMMERCI AL LOW-LEVEL WASTE SITES FOR RADIONUCLIDE SPILLS OR MIGRATION.
NUREC/CR-3810 VO1: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report 150
3 J nuOry-March 1983.
NUREO/CR-3825 V01-02: ' ACOUSTIC EMISSIGN/ FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Guarterly Report:
4-October'19837--March 1994.Vols 1 & 2.
BROOKHAVEN NAT10NAL LABORATORY NOREG/CR-2907 V02: RADIDACTIVE MATERIALS RELEASED FROM NUCLEAR POWER t
PLANTS. Annual:Meport-1981.
NUREO/CR-3383: IRRADIATION EFFECTS ON THE STORAGE AND DISPOSAL OF RADWASTE CONTAINING ORGANIC ION-EXCHANGE MEDIA.
NUREO/CR-3603: MINET-VALIDATION BURVEY USING EBB-II TEST DATA.
-NUREC/CR-3604:' BOLTING' APPLICATIONS.
NUREO/CR-3627: FRANTIC II APPLICATIONS TO STANDBY SAFETY SYSTEMS.
NUREO/CR-3628: - PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT STRUCTURES NUREC/CR-3641: RELIABILITY ASSESSMENT OF INDIAN POINT UNIT 3 CONTAINMENT' STRUCTURE.
NUREC/CR-3664: ' A DESCRIPTION AND ASSESSMENT OF R AMONA-3B MOD. O CYCLE 4:
A COMPUTER CODE WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR SYSTEM TRANSIENTS.
NUREC/CR-3713: OROUPING OF LIGHT WATER REACTORS FOR EVALUATION OF DECAY HEAT REMOVAL CAPABILITY.
BROWN UNIV., PROVIDENCE,'RI NUREC/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA: A User 's Guide For The Applicable Computer Programs In The Statistical Package For Social-Scientists (SPSS).
CALIFORNIA, UNIV.
OF,' BERKELEY..CA NUREC/CR-3686: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Summary Report.
NURE0/CR-3686 VO1: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS. Part A - User 's Manual.
NURE0/CR-3686 V02: WIPS -COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part B - Theory Manual.
NURE0/CR-3686 V03: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part C - Programmer's Manual.
NURE0/CR-3686-V04: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
CALIFORNIA, UNIV. OF, SANTA BARBARA, CA NOREO/CR-3583: EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR OBTAINING SITE CHARACTERISTIC INFORMATION.
COMMERCE, DEPT. OF, NATIONAL BUREAU OF STANDARDS' NUREQ/CR-3628: PROBABILITY BASED SAFETY CHECKING OF NUCLEAR PLANT STRUCTURES.
NUREO/CR-3775: GUALITY ASSURANCE FOR MEASUREMENTS OF IONIZING RADIATION.
COMMERCE. DEPT. OF, NATL. OCEAN 00RAPHIC & ATMOSPHERIC ADMINISTRATION NURE0/CR-3488 V02: IDAHO FIELD EXPERIMENT 1981. Vol 1: Measurement Data.
NUREQ/CR-3759; LIGHTNING STRIKE DENSITY FOR THE CONT!0VOUS UNITED STATES FROM THUNDERSTORM DURATION RECORDS.
NUREC/CR-3773: VARIATION OF PLANETARY BOUNDARY LAYER DISPERSIDN' PROPERTIES WITH HEIGHT IN UNSTABLE CONDITIONS.
NURE0/CR-3838: AN INITIAL REVIEW OF SEVERAL METEOROLOGICAL MODELS SUITABLE FOR LOW-LEVEL WASTE DISPOSAL FACILITIES.
CONTROL DATA CORP.
NURE0/CR-3741 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 2 Top ical Report, Volume 1: Data Evaluation.
DAVID W.
TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER NURE0/CR-3740: J-INTEGRAL TEARING INSTABILITY ANALYSIS FOR 8-INCH DIAMETER ASTM A106 STEEL PIPE.
DECISION RESEARCH, INC.
NURE0/CR-3507: AN ANALYSIS OF THE NRC SAFETY GOALS FOR NUCLEAR POWER.
i ist j
4
^
EGar3, - INC.
NUREG/CR-2531 RO2: -INTRODUCTORY USER 'S MANUAL FOR THE U..S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RESEARCH DATA. BANK.
NUREC/CR-2691: EFFECTS OF CLADDING SURFACE THERMOCOUPLES AND ELECTRICAL HEATER ROD DESIGN ON GUENCH BEHAVIOR.
-NUREG/CR-3360: COPPUTER ' PROGRAM' CDCID: AN AUTOMATED GUALITY CONTROL
-PROGRAM USING CDC UPDATE.
I NUREG/CR-3583: ' EVALUATION OF LOW-ALTITUDE REMOTE SENSING TECHNIQUES FOR f~
DBTAINTNG SITE CHARACTERISTIC INFORMATION.
NUREO/CR-3596:. SEVERE ACCIDENT SEGUENCE ANALYSIS (SASA) PROGRAM SEGUENCE EVENT TREE: BOILING WATER REACTOR ANTICIPATED. TRANSIENT WITHOUT SCRAM.
NUREG/CR-3633 VO1': TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 1: Model i
Description.
NUREG/CR-3633 VO2: TRAC-BD1/ MOD 1: AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume 2: Users Ouide.
NUREO/CR-3633 VO3: TRAC-BD1/ MOD 1; AN ADVANCED BEST ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS. Volume.3: Code l~
Structure and Programming Information.
I NUREC/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETR IC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA J
SYSTEM.
NUREO/CR-3722: DAMPING TEST RESULTS FOR STRAIGHT SECTIONS OF 3-INCH AND
)
8-INCH UNPRESSURIZED PIPES.
NOREO/CR-3762: IDENTIFICATION OF EQUIPMENT AND COMPONENTS PREDICTED AS SIGNIFICANT CONTRIBUTORS TO SEVERE CORE DAMAGE.
NUREO/CR-3781 DRFT: PCT-RELATED CLADDING FAILURES DURING OFF-NORMAL i
EVENTS-DRAFT: Draf t Report Of The USNRC PCI Review Group.
j NUREO/CR-3875: THE USE OF IN-SITU PROCEDURES FOR SEISMIC GUALIFICATION OF EQUIPMENT IN CURRENTLY OPERATING PLANTS.
ENGINEERS INTERNATIONAL, INC.
4 NUREG/CR-3489: ASSESSMENT OF RETRIEVAL ALTERNATIVES FOR 'THE GEOLOGIC i
DISPOSAL'0F NUCLEAR WASTE.
{
ENSA, INC.
g NUREO/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness
{
Degradation of A302-B & A333-8 Reference Plates From PSF Simulated i
Surveillance & Through-Wall Irradiation Capsules.
(
I NUREO/CR-3295 VO2: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE i
DOSIMETRY IMPROVEMENT PROGRAM: Postirradiation Notch Ductility &
i Tensile Strength Determinations For PSF Simulated Surveillance &
Through-Wall Specimen Capsules.
ENVIRONMENTAL FILMS, INC.
NUREO/CR-3670: VIOLENT TORNADO CLIMATOGRAPHY, 1980-1982.
FRANKLIN INSTITUTE / FRANKLIN RESEARCH CENTER j
NUREO/CR-3754: FAILURE CVALUATION OF GENERAL ELECTRIC SS-1 AND 85-9
)
REACTOR MODE SWITCHES.
j GENERAL PHYSICS CORP.
i NUREG/CR-3606: NUCLEAR POWER PLANT CONTROL ROOM CREW TASK ANALYSIS l-DATABASE: SEEK SYSTEM. (Users Manual).
GOLDER ASSOCIATES NUREO/CR-2613: IDENTIFICATION OF CHARACTERISTICS WHICH INFLUENCE I
REPOSITORY DESIGN - DOMAL SALT.
NURE0/CR-2614: IDENTIFICATION OF. CHARACTERISTICS WHICH INFLUENCE REPOSITORY DESIGN - TUFF.
i NUREO/CR-3218: EVALUATION OF ENGINEERING ASPECTS OF BACKFILL PLACEMENT J
l FOR HIGH LEVEL NUCLEAR WASTE (HLW) DEEP GEOLOGIC REPOSITORIES. Final l
Report (Task 5) June 1981 - February'1983.
j 4
l'
.s 1
HANFORD ENGINEERIND DEVELOPMENT LABORATORY'
- NUREC/CR-3391 ' VO2: LWR. PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Quarterly Progress Report, April 1983 - June 1983.
NUREC/CR-3391'VO3: LWR. PRESSURE VESSEL SURVEILLANCE DOSIMETRY '
IMPROVEMENT PROGRAM. Annua l Report,0c tober - 1 1982-Septemb er _ 30,1983.
NOREC/CR-3391 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOBIMETRY L
IMPROVEMENT PROGRAM.Ouarterly Progress Report, October 1983-December 1983.
NUREG/CR-3658: CONSIDERATIONS RELEVANT TO THE DRY STORAGE OF LWR FUEL RODS CONTAINING WATER.
i INHALATION T0XICOLOGY RESEARCH INSTITUTE i
NOREC/CR-3745: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report: April 1,1982 - March 31,1983.
JRB ASSOCIATES I
'NUREO/CR-33OO VO1: REVIEW AND EVALUATION OF THE ZION PROBABILISTIC SAFETY. STUDY: PLANT ANALYSIS.
LAWRENCE LIVERMORE NATIONAL LABORATORY NUREO/CR-3476: CHEMICALS IN EFFLUENT WATERS FROM NUCLEAR POWER
^
STATIONS: THE DISTRIBUTION, FATE AND EFFECTS OF CDPPER.
4 NUREC/CR-3686: WIPS--CDMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF l
PIPING SYSTEMS. Summary Report.
I -
NUREC/CR-3686 VO1:- WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF
. PIPING SYSTEMS. Part A - User 's Manual.
NUREG/CR-3686 VO2: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF P IPING SYSTEMS. Part B '-- Theorg Manual, t
NUREO/CR-3686 VO3: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF l-PIPING SYSTEMS.Part C - Programmer's Manual.'
NUREO/CR-3606 V04: WIPS--COMPUTER CODE FOR WHIP AND IMPACT ANALYSIS OF PIPING SYSTEMS.Part D - Verification Manual.
NUREC/CR-3718: RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PIPING -
l STATUS REPORT.
NUREC/CR-3755: STRONO GROUND MOTION STUDIES FOR SOUTH CAROLINA I
EARTHOUAKES.
NUREO/CR-3756: SEISMIC HAZARD CHARACTERIZATION CN THE EASTERN UNITED l
STATES: METHODOLOGY AND INTERIM RESULTS FOR TEN SITES.
NOREO/CR-3839: AN EMPIRICAL ASSESSMENT OF NEAR-SOURCE GROUND MOTION FOR
]
A 6.6 MB (7.5 MS) EARTHGUAKE IN THE EASTERN UNITED STATES.
4 LEHIGH UNIV., BETHLEHEM, PA NUR E0/CR-3849: TWO-PHASE 3X3 ROD BUNDLE TEST FACILITY FOR POST-CRITICAL HEAT FLUX BOILING.
LOS ALAMOS SCIENTIFIC LABORATORY NUREG/CR-3305: COMPARISON OF BEACON AND COMPARE REACTOR C AVITY SUBCOMPARTMENT ANALYSES.
NURE0/CR-3567: TRAC-PF1: AN ADVANCED BEST-ESTIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR ANALYSIS.
[
NUREC/CR-3644: REVIEW OF PROPOSED FAILURE CRITERI A FOR DUCTILE j
MATERIALS.
j NUREO/CR-3650: A STATISTICAL ANALYSIS OF NUCLEAR POWER PLANT PUMP l
FAILURE RATE VARIABILITY - Some Preliminary Results, j
NURE0/CR-3704: THREE-DIMENSIONAL CALCULATIONS OF TRANSIENT FLUID-THERMAL MIXING IN THE DOWNCOMER OF THE CLAVERT CLIFFS-1 PLANT USING SOLA-PTS.
LOVELACE BIOMED & ENVIRONMENTAL RESEARCH INSTITUTE I
NUREO/CR-3745: BIOLOGICAL CHARACTER!ZATION OF RADI ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URAN!UM MILLING EFFLUENTS. Annual Progress i
Report: April 1,1982 - March 31,1983.
1 MATERIALS ENGINEERING ASSOCIATES, INC.
l NUREG/CR-3295 VO1: LIGHT WATER REACTOR PRESSURE VESSEL SURVEILLANCE j
DOS! METRY IMPROVEMENT PROGRAM: Notch Ductility & Fracture Toughness I
I i-163
=
- D;gecdotion Lef-' A302-8 & 'A533-8.Roforcnco Platos Frco PSF Sinulat0d.
' Surveillance &;Through-Wall. Irradiation Capsules.
NUREG/CR-3295 VO2r. LIGHT WATER: REACTOR PRESSURE VESSEL SURVEILLANCE l DOSIMETRY !!MPROVEMENTl PROGRAM: Postirradiation: Notch Ductility &
R' -
17 ensile'Strengsh Determinations For PSF Simulated Surveillance &
Through-Wall Sp-Jcimen Cap sules.
NUREO/CR-3506:. J-R' CURVE CHARACTERIZATION.0F IRRADIATED LOW UPPER SHELF
+
-WELDS.
NUREG/CR-3546: ETHE-TEMPERATURE DEPENDENCE OF FATIQUE CRACK ORDWTH RATES OF A 351 CFSA CAST STAINLESS STEEL IN LWR' ENVIRONMENT.
- NATIONAL SCIENCE FOUNDATION.
-NUREG/CR-3847: CLIMATIC CALIBRATION OF POLLEN DATA: A. User 's Guide For
~The Applicable. Computer Programs In The Statistical Package For Social Scientists (SPSS).
OAK RIDGE NATIONAL LABORATORY NUREC/CR-2OOO VO3 N3: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of March 1984.
NUREG/CR-2OOO VO3 N4: LICENSEE EVENT REPORT (LER). COMPILATION:For. Month Of April 1984.
NUREG/CR-2OOO VO3 N5: LICENSEE EVENT REPORT,(LER) = COMPILATION: For Month Of May 1984.
NUREO/CR-2940: REALISTIC SIMULATION OF SEVERE ACCIDENTS'IN'
.BWRS-COMPUTER MODELING REGUIREMENTS.
NUREO/CR-32OO VO4: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING
)
- PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1983.
NUREO/CR-3303: USE OF NEUTRON NOISE FOR DIACNOSIS OF IN-VESSEL ANOMALIES IN LIGHT-WATER REACTORS.
. NUREG/CR-3335: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TEST HI-3.
NUREG/CR-3410: CHMONE:A ONE-DIMENSIGNAL COMPUTER CODE FOR SIMULATING TEMPERATURE, FLOW AND CHENICAL CONCENTRATIONS - IN WATER BODIES.
'NUREC/CR-3422 VO3: AEROSCL RELEASE AND TRANSPORT PROGRAM.Guarterly Progress Report For July-September 1983.
NUREC/CR-3507: AN ANALYSIS OF THE NRC SAFETY GOALS FOR NUCLEAR POWER.
(,
NUREG/CR-3514: THE CHEMICAL ~ BEHAVIOR OF 10 DINE IN AGUEDUS SOLUTIONS UP l
TO 150 C.An Experimental Study of Nonredox Conditions.
1 j
NOREG/CR-3515: SAFETY-RELATED ' OPERATION ACTIONS: METHODOLOGY FOR :
DEVELOPING CRITERIA.
i NUREO/CR-3535: AGE-DEPENDENT DOSE-CONVERSION FACTORS FOR SELECTED I
BONE-SEEKING RADIONUCLIDES.
3 NUREC/CR-3539: IMPACT OF CONTAINMENT BUILDING LEAKAGE ON LWR ACCIDENT I
RISK.
j NUREO/CR-3572: DETERMINATION OF METABOLIC DATA APPROPRIATE FOR HLW DOSIMETRY (ICRP-30),I.
i NOREO/CR-3595: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM - FIVE YEAR PLAN l
FY 1983-1987.
t NUREO/CR-3600: DATA
SUMMARY
REPORT FOR FISSION PRODUCT RELEASE TE3T I
HI-4.
NUREO/CR-3626 VO1: MAINTENANCE PERSONNEL PERFORMANCE SIMULATION (MAPPS) 1.
MODEL:
SUMMARY
DESCRIPTION.
)
NUREG/CR-3652: EVALUATION OF INSTRUMENTATION FOR DETECTION OF i
INADEGUATE CORE COOLING IN BOILING WATER REACTORS.
NUREQ/CR-3672: EXAMINATION OF THE SIZE EFFECTS AND DATA SCATTER 4
j.
OBSERVED IN SMALL SPECIMEN CLEAVAGE FRACTURE TOUGHNESS TESTING.
NUREO/CR-3687: LOOSE-PART MONITORING PROGRAMS AND RECENT OPERATIONAL EXPERIENCE IN SELECTED U.S.
AND WESTERN EUROPEAN COMMERCIAL NUCLEAR POWER STATIONS, NUMEO/CR-3771: VESSEL V-7 AND V-8 REPA!R AND CHARACTERIZATION OF INSERT MATERIAL.
j' NUREO/CR-3800: REFCO-83 USER 'S MANUAL.
i l-154
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OKLAHOMA, UNIV. OF, NORMAN, OK NUREG/CR-3848: EXPERIMENTAL INVESTIGATION OF UNSTEADY TORNADIC WIND LOADS ON STRUCTURES.
PARAMETER, INC.
NUREO/CR-3054: CLOSEOUT OF IE DULLETIN 81-03: FLOW DLOCKAGE OF COOLING WATER TO SAFETY SYSTEM COMPONENTS DY CORDICULA SP.
(ASIATIC CLAM) AND MYTILUS SP.
(MUSSEL).
PURDUE UNIV., WEST LAFAYETTE, IN NUREO/CR-3700: DECAY OF BUOYANCY DRIVEN STRATIFIED LAYERS WITH APPLICATION TO PRESSURIZED THERMAL SHOCK (PTS).
ROGERS & ASSOCIATES ENGINEERING CORP.
NUREC/CR-3533: RADCN ATIENUATION HANDDOOK FOR URANIUM-MILL TAILINGS COVER DE3IGN.
SANDIA LABORATORIES NUREO/CR-2552: CRAC2 MODEL DESCRIPTION.
NUREG/CR-2679 VO4: ADVANCED REACTOR SAFETY RESEARCH, GUARTERLY REPORT, OCTOBER-DECEMBER 1982.
NUREC/CR-2921; CHEMICAL INTERACTIONS OF TELLURIUM VAPORS WITH REACTOR MATERIALS.
NUREO/CR-3023: MOLTEN THERMITE TEEMING INTO AN IRON OXIDE PARTICLE DED.
NUR EC /CR-3134: A SETS USER 'S MANUAL FOR VITAL AREA ANALYSIS.
NUREG/CR-33OO VO1: REVIEW AND EVALUATION OF THE' ZION PROB ADILISTIC SAFETY STUDY; PLANT ANALYSIS.
NUREO/CR-3310: TESTING OF THE CONTAIN CODE.
NUREC/CR-3316; VERIFICATION AND FIELD COMPARISON OF THE SANDIA WASTE-ISOLATION FLOW AND TRANSPORT MODEL (SWIFT).
NUREG/CR-3329 VO4: THERMAL / HYDRAULIC ANALYbIS RESEARCH PROGRAM. Quarterly Report October-December 1983.
NUREO/CR-3366: HIGH TEMPERATURE MELT ATTACK ON STEEL AND URANIA-COATED STEEL.
NUREO/CR-3578: VERIFICATION OF THE NETWORK FLOW AND TRANSPORT / DISTRIBUTED VELOCITY METHOD (NWFT/DVM) COMPUTER CODE.
NUREO/CR-3379: SLAM - A SODIUM-LIMESTONE CONCRETE ADLATION MODEL.
NURE0/CR-3511 VO1: INTERIM RELI ADILITY EVALUATION PROGRAM: ANALYSIS OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. Volume 1. Main Report.
NUREC/CR-3588: THE EFFECT OF LOCA SIMULATION PROCEDURES ON CROSS-LINKED POLYOLEFIN CABLE'S PERFORMANCE.
NURE0/CR-3608: RELAPS ASSESSEMENT: LOFT Large Dreak L2-5.
NURE0/CR-3623: STATUS REPORT: CORRELATION OF ELECTRICAL CADLE FAILURE WITH MECHANICAL DEGRADATION.
NURE0/CR-3624: A FORTRAN 77 PROGRAM AND USER 'S GUIDE FOR THE GENERATION OF LATIN HYPERCUDE AND RANDOM SAMPLES FOR USE WITH COMPUTER MODELS.
NUREO/CR-3629: THE EFFECT OF THERMAL AND IRRADIATION AGING SIMULATION PROCEDURES ON POLYMER PROPERITIES.
NURE0/CR-3630; EQUIPMENT GUALIFICATION METHODOLOGY RESEARCH: TESTS OF PRESSURE SWITCHES.
NUREG/CR-3639: LARGE DREAK LOCA ANALYSES FOR TWO-LOOP PWRS WITH UPPER-PLENUM INJECTION.
NUREO/CR-3673: ECONOMIC RISKS OF NUCLEAR POWER REACTOR ACCIDENTS.
NUREO/CR-3604: NUCLEAR POWER PLANT ALARM PRIORITIZATION (NPPAP) PROGRAM STATUS REPORT. January 1,1983 to September 31,1983.
ST. LOUIS UNIV., ST. LOUIS, MO NUREQ/CR-3755: STRONG GROUND MOTION STUDIES FOR SOUTH CAROLINA EARTHGUAKES.
NUREO/CR-3768: NEW MADRID SEISMOTCCTONIC STUDY: Activities During Fiscal Year 1982.
STRUCTURAL MECHANICS ASSOCIATES NUR EO/CR-3805: ENGINEEHING CHARACTERIZATION OF GROUND MOTION. Task I:Ef f ects Of Charac teristic s Of Free-Field Motion On Struc tural Response.
156
E TEXAS, UNIV. OF, AUSTIN, TX NUREO/CR-3637: THE APPLICATION OF STEIN AND RELATED PARAMETRIC EMPIRICAL BAYES ESTIMATORS TO THE NUCLEAR PLANT RELIABILITY DATA SYS TEM.
VANDERBILT UNIV., NASHVILLE, TN NURE0/CR-3769: DESCRIPT!CN AND SIGNIFICANCE OF THE GRAVITY FIELD IN THE REELFOOT LAKE REGION OF NORTHWEST TENNESSEE.
WOODWARD-CLYDE CONSULTANTS,-INC.
NURE0/CR-3005: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Tas k I: Effects Of Characteristics of Free-Field Motion On Struc tural Response.
IM
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Licensed Facility Index
^;
This index liets the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabatical 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.
40-3392 Allied Chemical Corp., Norristown. NJ.
NUREG-1071 50-155 Big Rock Point Nuclear Plant. Consumers %wer Co.
NUREG-0828 STN-50-456 Braidwood Station. Unit 1. Commonwealth Edison Co.
NUREC-1026 STN-50-457 Braidwood Station. Unit 2.
Cor.mo nwe a l t h Edison Co NUREG-1026 STN-50-454 Byron Station. Unit 1. Commonweal 6h IIdison Co.
NUREG-0876 604
.N-50-493 Ca11away Plant. Unit 1. Union Electric Co.
NUREC-0830 SO3 STN-50-483 Callaway Plant, Unit 1.
Union Electric Co.
NUREG-1058 50-317 Calvert Clif f s Nuclear Pouer Plant. Unit 1.
Baltinore Gas & Electric NUREG/CR-3511 VO1 50-317 Calvert Clif f s Nuclear Power Plant. Unit 1.
Baltinere Gas & Electric NUREG/CR-3704 50-413 Catawba Nuclear Station. Unit 1 Duke Pouer Co.
NUREG-0954 S02 50-414 Catawba Nuclear Station. Unit 2.
Duke Pouer Co.
NUREO-0954 SO2 50-461 Clinton Pouer Statior. Unit 1.
Illinois Power Co.
NUREG-0853 SO3 50-275 Diablo Canyon Nuclear Power Plant. Unit 1, Pacific Gas & Electric Co NUREG-0675 S23 50-275 Diablo Canyon Nuclear Power Plant. Unit 1.
Pacific Gas & Electric Co NUREG/CR-3797 50-323 Diablo Canyon Nuclear Pouer Plant. Untt 2.
Pacific Gas & Electric Co NUREG-0675 S23 70-1113 Ceneral Elec tric Co.. Wilmington. NC.
NUREC-1078 3
50-351 Hope Creek Nuclear Station. Unit 1, Public Service Electric & Gas Co NUREG-1074 SC-206 Indian Point Station. Unit 3.
Power Authoaity of State of New York NUREG/CR-3641 50-352 Limerick Generating Station. Unit 1.
Philadelphia Electric Co.
NUREG-0974 50-353 Linerick Generating Station. Unit 2.
Philadelphia Electric Co.
NUREG-0974 50-459 River Bend Station. Unit 1.
Gulf States Utilities Co.
NUREG-0989 70-0025 Rockwell International Corp.. Canoga Park. C A.
NUREG-1077 50-400 Shearon Harris Nuclear Power Plant. Unit 1. Carolina Power & Light C NUREG-1038 S01 5C-322 Shoreham Nuclear Pouer Station. Long Island Lighting Co.
NUREG-0420 505 50-387 Susquehanra Steam Electric Station. Unit 1. Pennsgivania Power & Lig NUREG-0776 SO7 50-363 Susquehenna Steam Electric Station. Unit 2. Pennsgivania Power & Lig NUREG-0776 SO7 50-269 Three Nile Island Nuclear Station. Unit 1. Metropolitan Edison Co.
NUREG-102OLD VO2 50-289 Three Nile Island Nuclear Station. Unit 1. Metropolitan Edison Co.
NUREG-1020Ln VO1 50-54 Union Carbide Research Reactor. Union Carbida Corp.
NUREG-1059 50-148 Univ. of Kansas Research Reactor NUREG-1051 50-397 WPPSS Nuclear Project. Unit 2.
Washington Public Power Suppig System NUREG-0892 SOS 50-302 Waterford Generatsng Station. Unit 3.
Louisiana Power & Light Co.
NUREG-0787 SO6 50-390 Watts Bar Nuclear Plant, Unit 1.
Tennessee Valleg Authority NUREG/CR-3693 50-295 Zion Nuclear Power Station. Unit
- 1. Conmenwealth Edison Co.
NUREC/CR-33OO VO1 50-304 Zion Nuclear Power Station. Unit 2. Commonwealth Edison Co.
NVREG/CR-33OO VO!
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k I
157
t I
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U S. NUCLE AR REQULATORY COMMISSION i REPORT NuwsER (Aas.pned ey TfDC, eder Vet he, er seyi NRC POIM S3G (2 843
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BIBLIOGRAPHIC DATA SHEET NUREG-0304, Vol. 9, No. 2 j
3 su N51RUCTio 50 1,.EREvE,u a u ave. TAN.
l 1 riva Aho Sv.rau Regulatory and Technical Reports Compilation fo econd Quarter 1984
. oa,E,,E,OR,CO.,arto April - June g
.ON 1,.
YEAR 5 AUTMOmill 6 OATE RhPORT tS50E0 ONTH YEAR Aygus t 1984 S PR[ECTIT ASK! WORK UNIT NuusER
- 7. PERf 0 Ruin.ORGANi2 Af TON NAME ANo MAILING A R E SS flac8 esse te Ceest Division of Technical Info tion and Document Control
/
Office of Administration f"~ oa 'a
au""a U.S. Nuclear Regulatory Commis on
/
Washington, DC 20555
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5.O - OR..N,..,10NN..E. -.i.N... E55.
Quarterly Same as 7, above.
April - June 1984 j
,,A.5,RAe,,- -
This compilation lists all NRC regulatory,and techn cal reports published under the NUREG series during the second quart r of 1984.
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,4 DOCUMENT AN ALY5iS - e KE vwORD5/DESCRiPTORS STATEMENT
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,7 NuM8ER OF PAGES 18 PRICE
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NUCLEAR REGULATO2Y COMMISSION
'051*GEg't is raio
,g WASHINGTON, D.C. 20566 a sa o c nmN70!IaYlE'd!.saco Main Citations and Abstracts Contractor Report Number index 120555078877 1 1ANLA519T Personal Author index tjgM,0IVOF t4 TIDC 0
Y C PUB MGT BR-POR NUREG WASHINGTOtl DC 20555 Subject index NRC Originating Organization Index NRC Contractor SponsorIndex Contractor index Licensed Facility Index