ML20063H374
| ML20063H374 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1982 |
| From: | Savolainen A NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V07-N02, NUREG-304, NUREG-304-V7-N2, NUDOCS 8209010450 | |
| Download: ML20063H374 (155) | |
Text
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NUREG-0304 Vol. 7, No. 2 Regulatory and Technical Reports Compilation for Second Quarter 1982 April - June U.S. Nuclear Regulatory Commission Office of Administration i
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Available from NRC/GPO Sales Program Superintendent of Documents Government Printing Office Washington, D. C. 20402 l
A yea'r's subscription consists of 4 issues for this publication.
Single copies of this publication are available from National Technical information Service, Springfield, VA 22161 4
Microfiche of single copies are available from NRC/GPO Sales Program Washington, D. C. 20555 1
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NUREG-0304 Vol. 7, No. 2 Regulatory and Technical Reports Compilation for Second Quarter 1982 April - June Dr.te Published: August 1982 Division of Technical Information and Document Control office of Administration U.S. Nuclear Regulatory Commission Wzahington, D.C. 20665 f
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I CONTE.'.TS i
l Preface..................................
V I
J 1
l Index Tab W
Main Citation and Abstracts.....
...........1 Staff Reports......................
Conference Proceedings Contractor Reports..............
Contractor Report Nurnber index................................................... 2 Personal Author index.......
.3 S ubject Index........................
.4 NRC Originating Organization Index (Staff Reports)......
.5 NRC Contract Sponsor index (Contractor Reports).
.6 Contractor index.....
................................................ 7 Licensed Facility index.........................
.8 D..
lH.
PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to:
Division of TechnicalInformation and Document Control Attn: Ann W. Savolainen Landow 212 U.S. Nuclear Regulatory Commission Washington, D.C. 20566 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:
Contractor Report Number index Personal Author Index Subject index NRC Originating Organization index (Staff Reports)
NRC Contract Sponsor Index (Contractor Reports)
Contractor index Licensed Facility Index A detailed explanation of the entries precedes each index.
The bibliographic elements of the main citations are the following:
Staff Report NUREG-0508: MARK 11 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) repon title, (3) report author, (4) organizational unit of author, (5) date report was published, iC) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for intomal NRC use).
Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 00832:070.
Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC intemal use).
Contractor Report NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R.
Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 804A29. 08912:242.
Where the entries are (1) report nubber, (2) report title, (3) report authors, (4) organizational unit of i
authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC intomal use).
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The following abbreviations are used to identify the document status of a report:
ADD - addendum APP - appendix DRFT - draft ERR
- errata N - number R
- revision S
- supplement V
- volume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the NRC-GPO Sales 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. 20556 You may charge any purt hase 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 International Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents.
NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported.
In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-spons : red conference proceedings, All these report codes are controlled and assigned by the NRC Division of Technical Information and i
Document Control.
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Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is an NRC staff originated report, NUREG/CP-XXXX is an NRC sponsored conference report, and NUREG/CR-XXXX is an NRC contractor-prepared report. The bibliographic information (see Preface for details) is followed by a brief abstract of the report.
NUREG-OO2O VO6 NO2: LICENSED OPER ATING REACTORS STATUS
SUMMARY
REPORT. Data As Of January 31,1982.(Grey Dook)
Office of Management and Program Analysis.
April 1982.
260pp.
8205120469.
13060:001.
The OPER ATING UNITS STATUS REPORT - LICENSED OPERATING REACTQRS provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office of Management and Prt, Jam Analysis from the Headquarters staff of NRC's Office of Inspection and En f orc ement, from NRC's Regional Offices, and from utilities.
The three sections of the report are: monthly highlights and statistics for c ommerc ial op erating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC 's Re gional Of fic es. 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 hoped the rep or t is helpful to all agencies and individuals interested in maintaining an awareness of the U. S.
energy situation as a whole.
NUREG-OO2O VO6 NO3: LICENSED OPER ATING REACTORS. Status Summary Report. Februa ry 1982.(Deige Book)
- Management Information Branch.
June 1982.
259pp.
8207210026.
13994:001.
The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible.
This inf ormation i s collected by the Office of Management and Program Analysis from the Headquarters staff o f NRC 's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are: monthly highlights and statistics f or commerc ial op erating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC 's Re gional Offices, IE Headquarter s and the utilities; and an appendi x for miscellaneous information such as spent fuel storage capability, reactor years of e xperience and non power reactors in the U.S.
It is hoped the rep or t is helpful to all agencies nd individuals interested in maintaining an awareness of the U. 3.
energy situation as a whole.
NUREC-OO30 'VOS NO4: NUCLEAR POWER PLANTS-CONSTRUCTION STATUS REPORT. Data As Of December 31,1981.(Yellow Book)
- Office of 1
Management and Program Analysis.
April 1982.
400pp.
8204220533.
12827:005.
The " Construction Status Report," also referred to as the " Yellow Book," is a quarterly publication containing nuclear power plant construction data and actual progress.
The information contained in this report is supplied to the NRC by applicants with Construction Permits.
NUREG-OO30 VO6 NO1: NUCLEAR POWER PLANTS-CONSTRUCTION STATUS REPORT. Data as of March 31,1981.(Yellow Book)
- Management Information Branch, June 1982.
166pp.
8207220657.
14023:212.
The " Construction Status Report," also referred to as the " Yellow Book," is a quarterly publication containing nuclear power plant construction data and actual progress.
The information contained in this report is supplied to the NRC by applicants with Construction Permits.
NUREC-OO40 VO6 NO1: LICENSEE CONTRACTOR AND VENDOR INSPECTION S TATUS REPORT. Guarterly Report. January 1982-March 1982.(White Book)
- Region 4.
Office of Director.
April 1982.
194pp.
8205060032.
13003:014.
This periodical covers the results of inspections performed by the NRC 's Vendor Program Branch that have been distributed to the inspected organizations during the period from January 1982 through March 1982.
Also included in this issue are the results of certain inspections performed prior to January 1982 that were not included in previous issues of NUREG-OO40.
NUREG-OO90 VO4 NO4: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES.Guarterly Report. October-December 1981.
- Director's Office.
May 1982.
30pp.
8206290543.
13658:321.
Section 208 of the Energy Reorganization Act of 1974 identifies an 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 Congress.
This report covers the period October 1 to December 31, 1981.
During the report period, there were two abnormal occurrences at the nuclear power plants licensed to operate.
One involved a generic concern pertaining to blockage of coolant f low to saf ety-rela ted systems.
The other involved seismic design errors at Diablo Canyon Nuclear Power Plant with subsequent suspension of the fuel load and low power operating license during the report period; the Agreement States reported no abnormal occurrences to the NRC.
The report also contains information updating a previously reported abnormal occurrence.
NUREG-0304 VO3 SO1: REGULATORY AND TECHNICAL REPORTS. Compilation For l
1975-1978.
- Division of Technical Inf ormation & Document Control.
April 1982.
26pp.
8205110137.
13037:340.
This compilation lis ts f ormal staf f and contractor reports issued by the U.S.
Nuclear Regulatory Commission that were not listed in
" Regulatory and Technical Repor ts for 1975 - 1978," NUREG-0304, Vol.
3.
l This compilation contains a listing of reports and their abstracts and a keyword index.
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NUREG-0519 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LASALLE COUNTY STATION, UNITS 1 AND 2. Docket Nos. 50-373 And 50-374.
(Commonwealth Edison Company,et al.)
- Office of Nuclear Reactor Regulation, Director.
April 1982.
24pp.
8205040024.
12971:090.
Supplement No. 3 to the Safety Evaluation Report of Commonwealth Edison Company's application for licenses to operate its La Salle County Station, Units 1 and 2, located in Brookfield Township, La Salle County, Illinois has been prepared by the Of fice of Nuclear Reactor Regulation of the U.S.
Nuclear Regula tory C ommission.
This supplement addresses several items that have come to light since the previous supplement wa s issued.
NUREG-0540 NO2: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. Doc uments From October Through December 1974 For Dockets 50-334 Through STN 50-597.
- Division of Technical Information &
Document Control.
June 1982.
175pp.
B207070146.
13782:001.
This doc ument contains a descrip tion o f information received and generated by the U.S.
NRC.
This special ed ition contains Doc ket 50 material from 1978 that has not appeared in previous editions of the Title List.
The documents in this supplement are indexed by personal author, corporate source, and report number.
NUREG-0540 VO3 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. Dec emb er 1-31,1981.
- Division o f Technical Information &
Document Control.
April 1982.
575pp.
8204160034.
12712:063.
This doc ument is a monthly publication containing descriptions of inf ormation r eceived and generated by the US NRC.
This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory ag ency.
The docketed information includes information formerly issued through the U.S.
Department of Energy's Technical Inf ormation Center under the title Power Reactor Docket Information (PRDI).
This document replaces PRDI, which will no longer be prepared.
This document contains the following inde x e s: Personal Author Index, Corporate Source Index, and Report Number Index.
NUREG-0540 VO4 NO1: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
January 1-31,1982.
- Division of Technical Inf ormation & Doc ument Control.
May 1982.
450pp.
8205200296.
13201:001.
This doc ument is a monthly publication containing descriptions of information received and generated by the US NRC.
This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material rece ived and generated by NRC pert inent to its role as a regulatory agency.
The docketed information includes inf orma tion formerly issued through the U.S.
Department of Energy 's Technical Inf ormation C enter under the title Power Reactor Docket Information (PRDI).
This document replaces PRDI, which will no longer be prepared.
This document contains the following inde x e s: Personal Author Index, Corporate Source Index, and Rep ort Number Inder.
I NUREG-0540 V04 NO2: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
February 1-28, 1982.
- Division of Technical Information & Document Control.
May 1982.
402pp.
8205270478.
13291:237.
This doc ument is a monthly publication containing descriptions of 3
NUREG-0304 V06 N04: REGULATORY AND TECHNICAL REPORTS. Compilation For 1981. SAVOLAI NEN. A.
Division of Technical Information & Document Control.
May 1982.
499pp.
8205210507.
13216:278.
This compilation lists all NRC regulatory and technical reports published und er the NUREG series during 1981.
NUREG-0304 VO7 N01: REGULATORY AND TECHNICAL REPORTS. Compilation For First Guarter 1982. SAVOLAINEN,A.
Division of Technical Information
& Document Control.
May 1982.
143pp.
8206230080.
13593:001.
This compilation lists all NRC regulatory and technical reports published und er the NUREG series during the first quarter of 1982.
NUREG-0435 V04 N01: RESEARCH PROJECT CONTROL SYSTEM (RPCS) STATUS
SUMMARY
REPORT. Research Results Utilization. Data From July 1981-March 1982. (Buf f Bo o k )
- Office of Nuclear Regulatory Research, Director.
April 1982.
226pp.
8205270474.
13290:296.
This rep ort on "Research Results Utili zation" provides s tatus and control information concerning the utilization of research results in the regulatory policies and practices of the NRC.
Research Information Letters (RILs) are prepared by RES to transmit research results to NRC user offices upon completion of substantial. coherent and reasonably complete bodies of experimental and/or analytical research work.
Section 3. 0 o f this report lists the RILs issued to date, toge ther with an identifica tion of the research program manager and the research program element which generated the RIL.
The potential applicability cf each RIL to the regulatory process is also identified here, and comments from the cognizant RES and user office staff are summarized which relate to the expec ted impact of the reported RILs on the regulatory process.
NUREG-0485 VO4 NO3: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Da ta A s Of March 31,1982.(Buff Book)
- Management Information j.
Branch.
April 1982.
98pp.
8204220529.
12825:235.
The Systematic Evaluation Program is intended to examine many safety-related aspects of eleven of the old er light water reactors.
This document provides the existing status of the review process including individual topic and overall completion status.
NUREG-0485 V04 N04: SYSTEMATIC EVALUATION PROGRAM STATUS SUMMAR Y REPORT. Data A s Of April 30,1982.(Buff Book)
- Management Information Branch.
May 1982.
97pp.
8206100013.
13475:136.
The Systematic Evaluation Program is intended to examine many safety-related aspects of eleven of the older ligh t water rea c tors.
This document provides the existing status of the review proc ess including ind ividual topic and overall comp letion status.
NUREG-0485 V04 N05: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of May 31,1982. (Buff Book)
- Management Information Branch.
June 1982.
99pp.
8206240030.
13612:166.
The Systematic Evaluation Program is intended to examine many safety-related aspects of 11 of the older light water reactor s.
T h i s, document provides the existing status of th e review process including individual topic and overall completion sta tus.
4
inf ormation r eceived and generated by the US NRC.
This information includes (1) docketed material associated with civilian nuclear power plants and other uses of *adioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency.
The docketed information includes information formerly issued through the U.S.
Department of Energy's Technical Inf ormation Center under the title Power Reactor Docket Information (PRDI).
This document replaces PRDI, which will no longer be prepared.
This document contains the following inde x e s:
Personal Author Index, Corporate Source Index, and Report Number Inden.
NUREG-0540 VO4 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. March 1-31, 1982.(FOIA Supplement)
- Division of Technical Inf ormation & Document Control.
June 1982.
124pp.
8207060341.
13746:004.
This document is a monthly publication containing descriptions of inf ormation r eceived and generated by the U.S.
NRC.
This inf ormat f or.
includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory ag ency.
The docketed information includes in f orma ti on formerly issueo through the U.S.
Department of Energy's Technical Inf ormation C enter under the title, Power Reactor Doc ket Information (PRDI).
This document replaces PRDI, which will no longer be prepared.
This document contains the following indexes: Personal Author Index, Corporate Source Index, Repor t Numb er Inde x, and Cross Reference to Principal Documents Index.
NUREG-0540 VO4 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. March 1-31, 1982.
- Division of Technical Information &
Document Control.
June 1982.
683pp.
8207060342.
13742:001.
This document is a monthly publication containing descriptions of information received and generated by the U.S.
NRC.
This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency.
The docketed information includes information formerly issued through the U.S.
Department of Energy's Technical Inf ormation Center under the title, Power Reactor Docket Information (PRDI).
This document replaces PRDI, which will no longer be prepared.
This document contains the following indexes: Personal j
Author Index, Corporate Source Index, Repor t Number Index, and Cross j
Reference to Principal Documents Index.
NUREC-0566 VO2 NO2: STANDARDS DEVELOPMENT STATUS
SUMMARY
REPORT. Data As Of February 28, 1982. (Green Book)
- Internal Information Systems Branch.
April 1982.
224pp.
8205040046.
12969:064.
The Standards Development Status Summary Report is designed for i
scheduling, monitoring, and controlling the process by which Regulatory Standards, Guides, Reports, Petitions, and Environmental Statecsnts are written.
It is a summary of the current schedule plans for development of the above products.
NUREG-0580 V11 NO1-4: DRAFT REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data a s of January 1 - April 19,1982.(Blue Book)
- Management Information Branch.
May 1982.
75pp.
B206110008.
13492:196.
5
i Provide a review of the status of the progress of the licensing reviews for all construction permits, operating licenses, special project and non-power reactor renewals under review, as reported to Congress.
NUREG-0580 V11 N05: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of May 15,1982.(Blue Book)
- Manag ement In f ormation Branch.
June 1982.
53pp.
8206230346.
13595:292.
Provider
'eview of the status of the progress of the licensing reviews for all construction permits, operating licenses, special projects and non power reactor renewals und er review, as reported to Congress.
NUREG-0606 VO4 NO2: UNRESOLVED SAFETY ISSUES
SUMMARY
Data As Of May 21,1982. (Aqua Book)
- Office of Resource Management, Director.
June 1982.
51pp.
8206250018.
11361:363.
Provides an overview of the status of the progress and plans for resolution of the generic tasks addressing " Unresolved Safety Issues" as reported to Congress.
NUREG-0712 S06: SAFETY EVALUATION REPCRT RELATED TO THE OPERATION OF SAN ONOFRE NUCLEAR GENERATING STATION UNITS 1 & 2. Docket Nos. 50-361 And 50-362.(Southern California Edison Company)
- Office of Nuclear Reactor Regulation, Director.
June 1982.
69pp.
8207210141.
l 13993:075.
Supplement No. 6 to the Safety Evaluation Report for the application filed by Southern California Edison Company, et al for licenses to operate the San Onofre Nuclear Generating Station, Units 2 and 3 (Docket Nos. 50-361 and 50-362) located in San Diego County, California has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
This supplement updates the status of review with regard to certain items tha t were left unresolved in previous supplements and it evaluates several new review items.
NUREG-0725 RO2: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
- Office of Nuclear Material Safety &
Safeguards, Director.
June 1982.
88pp.
8207020050.
13696:169.
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 tequirements 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 May 1,
1982.
Spent fuel shipment routes, primarily for road transportation, but also including one rail route, are indicated on reproductions of DOT road maps.
Also included are the amounts of material shipped during the approximate three year period that safeguards regulations for sp en t 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 sh ipments as well as safeguards incidents regarding spent fuel shipments (of which none have been rep orted to date).
This additional information is furnished by the Commission in order to convey to the public a more complete picture of NRC regulatory 6
practices concerning the shipment of spent fuel than could be obtained by the publication of the shipment routes and quanti ti e s alone.
NUREG-0748 VO2 NO3: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of March 31,1982.(Orange Book)
- Management Information Branch.
April 1982.
400pp.
8205130235.
13076:341.
The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NHC) with an overview of licensing actions dealing with operating power and nonpower reactors.
NUREG-0748 VO2 NO4: OPERATING REACTOR LICENSING ACTIONS
SUMMARY
. Data As Of April 30,1982.(Orange Book)
- Mana g emen t Information Branch.
May 1982.
239pp.
8206100042.
13477:045.
The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NHC) with an overview of licensing actions dealing with operating power and nonpower reac tors.
NUREG-0748 VO2 N05: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of May 31,1982.(Orange book)
- Management Information Branch.
June 1982.
3OOp p.
8207060336.
13745:005.
The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors.
NUREG-0750 V13 102: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. January-June 1981.
- Division of Technical Information &
Document Control.
April 1982.
98pp.
8204150571.
12686:172.
Indexes to legal issuances of the Atomic Safety and Licensing Board and App eal Panels, the Commission, the Administrative Law Judge, and NRC Program Offices.
NUREG-0750 V14 101: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. July-September 1981.
- Division of Technical Information &
Document Control.
June 28, 1982.
70pp.
8206290546.
13658: 159.
Indexes to legal issuances of the Atomic Safety and Licensing
}
Board and App eal Panels, the Commission, th e Administrative Law Judge, and NRC Program Offices.
NUREG-0786 RO1: SITE-SUITADILITY REPORT IN THE MATTER OF THE CLINCH RIVER BREEDER REACTOR PLANT. Docket No.
50-537.
- Clinch River
[
Breeder Reactor Program Office.
June 1982.
70pp.
8206250025.
13628:017.
The Office of Nuclear Reac tor Regulation issued a Site Suitability Report (SSR) for the proposed Clinch River Dreeder Reartor Plant (CRBRP) in March 1977.
That report d ocumented the results of l
the staff's evaluation of the suitability of the proposed CRBRP sito for a facility of the general size and type as the CRBRP from the standpoint of radiological health and safety considerations.
The staff concluded in that report that the proposed CRBRP site was I
suitable for such a facility.
7
This rep ort supersed es the March 1977 report.
Although a number of changes have occurred since the March 1977 Site Suitability Report was issued, the staff's conclusion in this report remains unchanged.
The proposed CRBRP sit is suitable for a facility of the general size and type as the CRBRP frt the standpoint of radiological health and safety consid erations.
NUREG-0787 503: SAFETY EV.*.LUATION REPORT RELATED TO THE OPERATION OF WATERFORD STEAM ELECTRIC STATION UNIT NO.
- 3. Docket No.
50-382. (Louis iana Power & Ligh t Company)
- Office of Nuclear Reactor Regulation, Director.
April 1982.
2OOpp.
8205190042.
13169:260.
Supplement No. 3 to the Saf ety Evaluat ion Report for the application filed by Louisiana Power & Ligh t 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 Nuclear Regulatory Commission.
The purpose of this supplement is to provide the staff's evaluation of information submitted by the applicant since th e Saf ety Evaluation Report and Eupplement Nos. 1 and 2 were issued.
This supplement also includes a copy of the supp lemental report by the Advisory Committee on Reactor Saf eguards dated March 9, 1982.
NUREG-0793: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MIDLAND PLANT, UNITS 1 AND 2. Docket Nes. 50-329 And 50-330. (Consumers Power Company)
- Office of Nuc lear Reactor Regulation, Direc tor.
May 1982.
480pp.
8205190027.
13168:001.
The Saf e ty Evaluation Report f or the application filed by the Consumers Power Company, as app licant and owner, for a license to operate the Midland Plant, Units 1 and 2 (Doc ket Nos. 50-329 and 50-330), has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is located near the city of Midland in Midland County, Michigan.
Eubject to favorable resolution of the items discusses in this report, the staff concludes tha t the facility can be operated by the applicant without endangering the health and safety of the public.
1 NUREG-0793 SO1: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MIDLAND PLANT, UNITS 1 AND 2. Docket Nos 50-329 And 50-330.(Consumers Power Company)
- Office of Nuclear Reactor Regulation, Director.
June 1982.
50pp.
8207150635.
13860:093.
This rep ort supplements the Saf e ty Eva lution Report, NUR EG-0793, issued May 1982 by the Office of Nuclear Reactor Regulation o f the U. S.
Nuclear Regulatory Commission with respect to the application filed by Consumers Power Company, as applicant and owner, for licenses to operate the Midland Plant. Units 1 and 2 (Doc ket Nos. 50-329 and 50-330).
The facility is located in the city of Midland in Midland County, Michigan.
This supplement provides recent information regarding resolution of some of the open items identified in the Safety Evaluation Report and discusses recommendations of the Advisory Committee on Reactor Safeguards in its interim report dated June 8, 1982.
NUREG-0820 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATION PROGRAM FOR PALISADES PLANT. Doc k et No. 50-255. (Con sumers Power Company)
- Division of Licensing.
April 1982.
475pp.
8204160032.
12714:082.
8
The Integrated Plant Safety Assessment Report for the Consumers Power Company 's Palisades Plant (Docket No. 50-255) located in Covert Township, Van Buren County, Michigan, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission (NRC).
The report documents the review completed under the Systematic Evaluation Program (SEP).
The SEP was initiated by the NRC to review l
designs of older operating nuclear reactor plants to reconfirm and document their safety.
The review has provided for (1) an assessment of the significance of the difference between current technical positions on safety issues and those that e xisted when the Palisades Plant was licensed, (2) a tasis for deciding on how these differences should be resolved in an integrated p lant r eview, and (3) a documentated evaluation of plant safety.
Equipment and procedural l
changes have been identified as a result of the review.
It is expected that this rep ort will be one of the bases in considering the conversion of Palisades' provisional operating license to a full-term op erating license.
NUREG-0821 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATION PROGRAM FOR R. E.
GINNA NUCLEAR POWER PLANT. Doc k et No.
50-244.(Rochester Gas & Electric Corporation)
- Divs+ ion of Licensing.
May 1982.
400pp.
8206110322.
13481:302.
The Systematic Evaluation Program was initiated in February 1978 by the U.
S.
Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety.
The review provides (1) an assessment of how these p lants 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 rep ort documents the review of th e R.
E.
Ginna Nuclear Power Plant, owned and operated by Rochester Gas and Electric Corporation (located in Wayne County near Rochester, NY), one of ten plants reviewed under Phase II of this program, and indicates how 137 topics selected for review under Phase I of the program were address ed.
Equipment and procedural changes have been identified as a result oF the review.
It is expected that this repor t will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating lic en s e.
NUREG-0831 S02: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF GRAND GULF NUCLEAR STATION, UNITS 1 & 2. Docket Nos. 50-416 &
50-417.(Mississippi Power And Light Company)
- Office of Nuclear Reactor Regulation, Director.
June 1982.
100pp.
8207070141.
13781:018.
Supplement No. 2 to the Safety Evaluation Report for Mississippi Power and Light Company, et al, joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Por t Gibson, in Claiborne County, Mississippi, 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.
NUREG-0837 VO1 N01-2: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr ess Report, January-June 1981. COHEN,L.K.s SLOBODIEN,M.J.
Region 1,
Office of Director.
April 1982.
85pp.
9
8205130269.
13088:103.
This report provides the status and results of the NRC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network.
It presents the radiation levels measured in the vicinity of 55 NRC-licensed facility sites throughout the country for the first half of 1981. The program objectives, scope, and methodology are g iven.
The TLD system, dosimeter location, data processing scheme, and quality assurance program ari outlined.
NUREG-0837 VO1 NO3: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progress Report, July-December 1981. COSTELLO,F.;
THOMPSON,T.s COHEN,L.K.
Region 1,
Office of Director.
May 1982.
2OOp p.
8206090230.
13457:343.
This report provides the status and results of the NRC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network.
It presents the radiation levels measured in the vicinity of NRC-licensed facility sites throughout the country for the second half of 1981.
NUREG-0837 VO1 NO3-4: Errata To NUREG-0837 Volume 1, Numbers 3-4, Changing Number 3 & Dates Covered To July-September,1981, to NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report.
- Region 1, Office of Director.
June 23, 1982.
1p.
8207140002.
13847: 331.
This rep ort provides the status and results of the NRC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network.
It presents the radiation levels measured in the vicinity of NRC-licensed facility sites throughout the country for the second half of 1981.
NUREG-0837 VO1 NO4: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progress Report, October-December 1981. COSTELLO,F.;
THOMPSON, T. s COHEN,L.K.
Region 1, Office of Director.
June 1982.
170pp.
8207140214.
13844:209.
This report provides the status and results of the NRC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network.
It presents the radiation levels measured in the vicinity of NRC licensed facility sites throughout the country for the fourth quarter of 1981.
NUREG-0842: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE 0"ERATION OF ST. LUCIE PLANT, UNIT NO.
- 2. Docket No.
50-389. (Florida Pow 'r & Light Company)
- Of fice of Nuclear Reactor Regulation, Director.
April l
1982.
245pp.
8205110059.
13039:001.
The Final Environmental Statement related to the operation of the St. Lucie Plant Unit No. 2 by Florida Power and Light Company and Orlando Utilities Commission of the City of Orlando, Florida (Docket No. 50-389), located in St. Lucie County, Florida has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The statement reports on the staff's review of the impact i
of operation of the plant.
Also included are comments of state and l
federal government agencies and members of the public on the Draft Environmental Statement for this project and staff responses to these comments.
Th e NRC staf f has concluded, bas ed on a weighing o f environmental, technical and other factors, that an operating license l
could be granted.
l l
10 l
l
r NUREC-0847: SAFETY EVALUATION REPORT RELATED TO THE OPERAT. ION OF THE WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Doc k e t Nos. 50-390 And 50-391.(Tennessee Valley Authority)
- Of fice of Nuclear Reac tor Regulation, Director.
June 1982.
595pp 8207210019.
13995:001.
The Safety Evaluation Report for the application filed by the Tennessee Valley Authority, as applicant and owner, for a license to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391), has been prepared by the Offic e of Nuclear Reactor Regulation of the U.S.
Nuclear Regula tory C ommission.
The facility is located in Rh ea County, Tennessee, near the Watts Bar Dam of the Tennessee River.
Subject to favorable resolution of the items discussed in this report, the staff c onclud es that the facility can be operated by the applicant without endangering the health and safety of the public.
NUREG-0848: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF BYRON STATION UNITS 1 AND 2. Docket Nos. STN 50-454 And STN 50-455.(Commonwealth Edison Company)
- Office of Nuclear Reactor Regulation, Director.
April 1982.
400pp.
8204210626.
12797:150.
The information in this Final Environmental Statement is the second assessment of the environmental impact associated with the construction and operation of the Byron Sta tion, Units 1 and 2,
located in Rockvale Township, Ogle County, Illinois, approximately seventeen miles southwest of Rockford, Illinois.
The first assessment was the Final Environmental Statement related to construction issued in July 1974 prior to issuance of the Byron Construction Permits.
The present assessment is the result of the NRC s taf f r eview of the activities associated with the proposed operation of the plant, and includes the staff response to comments on the Draf t Environmental Statement.
NUREG-0849: STANDARD REVIEW PLAN FOR THE REVIEW AND EVALUATION OF EMERGENCY PLANS FOR RESEARCH AND TEST REACTORS. BATES E.F.;
GRIMES B.K.s RAMOS,S.L.
Direc tor 's Of fice. Office of Inspection and Enforcement.
May 1982.
37pp.
8206020104.
13332:315.
This doc ument provides a Standard Review Plan for the guidance of the NRC staff to assure that complete and uniform reviews are made of research and test reactor emergency plans.
The report is organized under ten planning standards which corresponds to the guidance criteria in Dra f t II of ANSI /ANS 15.16 at.
endorsed by Revision 1 to Regulatory Guide 2. 6.
The applicability of the items under each planning standard is indicated by subdivisions of the steady state thermal power levels at wh ich the reactors are licensed to operate.
Standard emergency classes and example action levels for research and test reac tors which should initiate these classes are given in an Appendix.
NUREG-0854: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF CLINTON POWER STATION, UNIT NO.
- 1. Doc k e t No.
50-461.
(Illinois Power Company, et al. )
- Office of Nuclear Reactor Regulation, Director.
May 1982.
400pp.
8206090137.
13457:051.
This Final Environmental Statement con tains the second assessment of the environmental impact associated with operation of the Clinton Power Station, Unit 1,
pursuant to th e Nati onal Environmental Policy Act of 1969 (NEPA) and 10 CFR Part 51, as amended, of th e NRC 's regulations.
This statement examines:
the purpose and need for the Clinton projects the affected environment, environmental consequentes 11
and mitigating actions, and environmental and economic benefits and costs.
The action called for is the issuance of an operating license for Unit 1 of the Clinton Power Station.
i NUREG-0857 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PALO VERDE NUCLEAR. GENERATING STATION, UNITS 1.2 & 3. Docket Nos. SlN 50-528, STN 50-529 & STN 50-530. ( Ari z ona Pub lic Servic e Compan y )
Office of Nuc lear Reactor Regulation, Direc tor.
May 1982.
31pp.
8206090227, 13457:311.
I Supplement No. 2 to the Safety Evaluation Report for the application filed by Arizona Public Service Company, et al, for licenses to operate the Palo Verde Nuclear Generating Station, Units 1,2 and 3 (Docket Nos. 50-528/529/530), located in Maricopa County, Arizona has b een prepared by the Office of Nuclear Reactor Regulation I
of the Nuclear Regulatory Commission.
Th e purpose of this supplement is to update the Safety Evaluation Report by providing (1) the evaluation of additional information submit ted by - the applicant since Supplement No. 1 to the Safety Evaluation was issued and (2) the evaluation of the matters that the staf f had under review when Supplement No. 1 was issued.
f NUREG-0061: TECHNICAL SPECIFICATIONS FOR LA SALLE COUNTY STATION, UNIT NO.
- 1. Docket No. 50-373 (Commonwealth Edison Company).
- Office of i
Nuclear Reactor Regulation, Director.
Ap r i l 1982.
484pp.
1 8205060007.
13000:004.
The La Salle County Station, Unit 1 Technical Specifications were prepared by the U.S.
Nuclear Regulatory Commission to set f or th the limits, operating conditions and other requirements applicable to a nuclear reactor facility as set forth an Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public.
1 NUREG-0863: SURVEY OF FOREIGN REACTDR OPERATOR QUALIFICATIONS, TRAINING, AND STAFFING REGUIREMENTS. AU,M.L.s D ISALVO, R. s MERSCHOFF,E.
Division of Facility Operations.
May 1982.
500pp.
8205200297.
13198:001.
This rep ort is a compilation of the data obtained from a survey of l
foreign nuclear power plant operator requirements.
Included among the considerations are shifting staffing, operator eligibility, operator i
training programs, operator licensing or certification, and operator retraining.
The data obtained from this survey are presented in matrix j
form and contrasted with U. S.
requirements.
i J
l NUREG-0868: A COLLECTION OF MATHEMATICAL MODELS FOR DISPERSION IN SURFACE WATER AND GROUNDWATER. CODELL,R.B.
Division of Engineering.
KEY,K.J.i WHELAN,C.
Battelle Memorial Ins t i tute, Pacific Nor thwest Laboratory. ' June 1982.
180pp.
8206180340.
13569:044.
This rep ort represents a collection of some of the manual procedures and simple computer programs used by the Hydrological l
Engineering Section of the Division of Engineering, Office of Nuclear Reactor Regulation, for computing the fate of routinely or j
accidentally released radionuclides in surface water and groundwater.
All models are straightforward simulations of dispersion with constant coefficients in simple geometries.
12
NUREG-0871 VO1 NO2:
SUMMARY
INFORMATION REPORT. January 1,1982-March 31,1982. (Brown Book)
- Office of Management and Program Analysis.
April 1982.
56pp.
8205190021, 13188:077.
Provides summary data concerning NRC and its licensees for general use by the Ch airman, other Commissioners and Commission staff offices, the Executive Director for Operations, and the Office Directors.
NUREG-0872: A FEASIBILITY STUDY OF USING LICENSEE EVENT REPORTS FOR A STATISTICAL ASSESSMENT OF THE EFFECT OF OVERTIME AND SHIFT WORK ON OPERATOR ERROR. DISALVO,R.s GERY,A.s P I TTMAN, J.
Division of Facility Operations.
June 1982.
94pp.
8207060004.
13741:121.
A study was made based upon the reported licensed operator errors from January 1981 to determine if a valid statistical determination could be made of the effects of shift work and overtime on op erator error.
The study concludes that the data reported in the Lic ensee Event Reports are inadequate to draw conclusions on the influence of overtime and shift work on operator error.
The analysis did show that the errors ar e not uniform over the hours of the day or the days of the weeks the causes of the non-uniformity could not be determined.
NUREG-0878: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF WOLF CREEK GENERATING STATION, UNIT 1. Doc k et No. STN 50-482. (Kansas Gas And Electric Company, et al. )
- Office of Nuclear Reactor Regulation, Director.
June 1982.
155pp.
8206290526.
13659:104.
The information in this Final Environmental Statement is the second assessment of the environmental impact associated with the construction and operation of the Wolf Creek Generating Station, Unit No.
1, located in Coffey County, Kansas.
The Draft Environmental Statement was issued in January 1982.
The first assessment was the Final Environmental Statement related to construction issued in October 1975 prior to issuance of the Wolf Creek Construction Permit.
The present a ssessment is the result of the NRC staff's review of the activities associated with the proposed operation of the plant, and includes the staff response to comments on the Draft Environmental Statement.
NUREG-0881: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WOLF CREEK GENERATING STATION, UNIT NO.
- 1. Docket No. STN 50-482.
(Kansas Gas And Electric Company, et al. )
- Office of Nuclear Reactor Regulation, Director.
April 1982.
800pp.
8204220539.
12826:001.
The Saf e ty Evaluation Report for the application filed by the Kansas Gas and Electric Company, as applicant and agent for the owners, for a license to operate the Wolf Creek Generating Station, Unit 1 (Docket No. STN 50-482), has been prepared by the Office of Nuclear Reactor Regulation of the U.
S.
Nuclear Reg ulatory Commission. The Facility is located in Coffey County, Kansas.
Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the ap plicant without endangering the health and safety of the public.
NUREG-0887: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PERRY NUCLEAR POWER PLANTS, UNITS 1 & 2. Docket Nos. 50-440 And 50-441.(Cleveland Electric Illuminating Company)
- Office of Nuclear Reactor Regulation, Director.
May 1982.
370pp.
8206170055.
13556:001.
This Safety Evaluation Report has been prepared by the Of fice of 13
1 Nuclear Reactor Regulation of the U.
S.
Nuc lear Regulatory Commission in response to an application filed by the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company and the Toledo Edison Company (the Central Area Power Coordination Group, CAPCO), as applicants and owners, for a license to operate the Perry Nuc lear Power Plant Units 1 and 2 (Docket Nos. 50-440 and 50-441).
The facility is located near Lake Erie in Lake County, Ohio.
Sub Ject to favorable resolution of the items discussed in this report, the NRC sta f f concludes that the facility can be operated by the Cleveland Electric Iluminating Company without endangering the health and safety of the public.
NUREG-0891: NUCLEAR PROPERTY INSURANCE: STATUS AND OUTLOOK. LONG,J.D.
Office of Sta te Programs, Dir e c t or.
May 1982.
115pp.
8206170048.
13543:233.
The report addresses the problem of the unavailability of adequate levels of property insurance for commercial power reactors to pay for decontamination and cleanup costs arising from accidents. The report is designed to answer six questions, as f ollows:
1.
What has been the development of each principal source o f nuclear property insurance used as of early 1982 by nuclear utilities in the United States?
2.
What are some of the distinguishing features of nuclear property insurance as offered by the principal sources?
3.
How much nuclear property insurance was offered by each of these sources as of January 1,
19827 4.
Assuming that present plans came to fruition, how much property insurance is likely to be offered by each of these sources as of Januarg 1-19A37 5.
What, if any, principei sources of nuclear property insurance are likely to emerge in the private sector by January 1,
1983?
6.
What problems serious enough to warrant action of the NRC exist with respect to nuclear property insurance and what actions should NRC ta ke in reponse to each problem?
NUREO-0894: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE CONSTRUCTION OF SKAGIT/HANFORD NUCLEAR PROJECT, UNITS 1 AND 2. Doc ket Nos. STN 50-522 And STN 50-523.(Puget Sound Power And Light Company, Pacific 1
Power And Light Company,et al.)
- Office o f Nuclear Reac tor Regulation, Director.
April 1982.
350pp.
8205120106.
13054:001.
This dra f t environmental statement contains an assessment of the environmental impact associated with the construction of Skag it/Hanford Nuclear Project, Units 1 and 2 (S/HNP) pursuant to the National Environmental Policy Act of 1969 (NEPA) and 10 CFR 51, as amended, of the NRC 's reg ulations.
This statement examines: the purpose and need for the S/HNP project, alternatives to the project, the af fec ted environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs.
No water-use impacts are expected from cooling-tower makeup withdrawn f
- om, or blowdown discharged into, the Columbia River.
Land-use and terrestria l-and aquatic-ecological impacts will be small.
Impacts to historic and prehistoric sites will be negligible with the development and implementation of the app licant 's cultural-resources management plan.
The risk associated with accidental radiation exposure is very low.
The net socio-economic effects of the project will be beneficial.
The action called for is the issuance of a construction permit for Skagit/Hanford Nuclear Project, Units 1 and 2.
14
NUREC-0895: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF SEA 3POOK STATION, UNITS 1 AND 2. Docket Nos. 50-443 & 50-444.
(Public Service Company Of New Hampshire, et al.)
- Office of Nuclear Reactor Regulation, Director.
May 1982.
400pp.
8205200290.
13200: 020.
The information in this statement is the second assessment of the environmental impact associated with the construction and operation of the Seabrook Station, Units 1 and 2, located in the town of Seabrook, New Hampshire.
The first assessment was th e Final Environmen tal Statement related to construction, issued in December 1974, prior tn issuance of the Seabrook construc tion permi ts.
The construction of Unit 1 is now 62% complete and commercial operation is scheduled for February 1984 The present assessment is the result of the NRC staff review of the activities associated with the proposed operation of the plant NUREG-0902: SITE SUITABILITY, SELECTION AND CHARACTERIZATION BRANCH TECHNICAL POSITION - Low Level Waste Licensing Branch. SIEFKEN,D.s PANGBURN,G.i PENNIFILL,R.s et al.
Division of Waste Management.
April 1982.
29pp.
8205060014.
13003:208.
The staf f provides an expanded interpretation of the site suitability requirements in the proposed rule 10 CFR Part 61, a description of the anticipated site selection process, and a detailed discussion of the site characterization program needed to support a license application and environmental rep or t.
The paper provides early-on guidance to prospective applicants in these three subject areas.
NUREG-0903: SURVEY OF INDUSTRY AND GOVERNMENT PROGRAMS TO COMBAT DRUG AND ALCOHOL ABUSE. ALTMAN,W.s BROWN,W.s BUSH,C.4 et al.
Director's Office. Office of Inspection and Enforcement.
June 1982.
76pp.
8206290552.
13658:357.
Report o f an NRC survey of the drug and alcohol programs of ten licensed nuclear utilities, two federal agencies, and two large corporations not in the nuclear industry.
Report contains management views on the extent of the drug and alcohol problem, policies on work-related use or possession of alcohol or drugs, and views on applicable proposed NRC regulatory initiatives.
Report describec practice and perceptions on: use of background investigations, psychological tests, supervisory training and behavioral observation, employee awar eness programs, employee assistance and rehabilitation programs, and use of chemical tests and oth er detec tion measures.
Report describes a recommended generic Daseline Program for combatting drug and alcohol problems in the nuclear industry, which includes: wri t ten drug and alcohol policy, company resolve to exercise the policy, employee awareness program, employee assistance or rehabilitation program, employee screening program, and drug and alcohol detec tion program.
NUREG-0904: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE DECOMMISSIONING OF THE RARE EARTHS FACILITY, WEST CHICAGO, ILLINOIS. Docket No. 40-2061. ( Kert-McGee Ch emical Corp ora tion )
- Office of Nuclear Material Safety & Safeguards, Director.
May 1982.
400pp.
8205200288.
13197:005.
This Dra f t Environmental Impact Statement is issued by the U. S.
Nuclear Regulatory Commission in response to the plan proposed by Kerr-McGee Ch emical Corporation for the dec ommissioning of their Have Earths Facility located in West Chicago, Illinois.
The statement 15
considers the Kerr-McGee preferred plan and various alternatives to that plan.
The action proposed by the Commission is the renewal of the Kerr-McGee license to allow safe storage of the radioactive waste onsite for a period of 5 years.
At the end of this period, the following alternatives will be evaluated:
1.
Renewal of the license f or an additional period uf 5 years and the possible imposition of additional conditions or remedial actions.
2.
Removal of the material to a lic en sed low-level waste disposal site.
3.
Termination of the license and transfer of the property to federal or state ownership.
NUREG-0909: NRC REPORT ON THE JANUARY 25,1982 STEAM GENERATOR TUDE RUPTURE AT R. E.
GINNA NUCLEAR POWER PLANT. MARTIN,T.T.
Division of Engineering & Technical Programs.
April 1982.
335pp.
8204210731.
12798:065.
This NRC Task Force report documents the circumstances surrounding the January 25, 1982, steam generator tube rupture event at the R.
ii.
Ginna Nuclear Power Plant.
It focuses on the period f rom 9: 25 a. m.
on January 25, when the tube rupture occurred, to 10: 45 a. m.
on January 25, when the plant entered the recovery pha se.
Information outside this period is recounted as nec essary to place the event in perspective.
The report is intended to describe factual information and significant findings associated with the event and, thereby, provide the required data base for appropriate detailed analysis and recommendations by various NRC of fices.
NUREG-0911: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE WASHINGTON STATE UNIVERSITY TRIGA REACTOR. Docket No. 50-27.
- Division of Licensing.
May 1982.
64pp.
8206240025.
13610:003.
This Safety Evaluation Report for the application filed by the Washington State University (WSU) for a renewal of operating license number R-76 to continue to operate a research reactor has been prepared by the Of fice of Nuclear Reactor Regulation of the U.
S.
Nuclear Regulatory Commission.
The facility is owned and operated by the Washington State University and is loca ted on the WSU campus in Pullman, Whitman County, Washington.
The s taff concludes tha t the TRIGA reactor facility can continue to be operated by WSU without endangering the health and safety of the public.
NUREG-0913: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE UNIVERSITY OF FLORIDA. Doc k e t No. 50-83.
- Of fice of Nuclear Reactor Regulation, Director.
May 1982.
90pp.
8206240021.
13612:080.
This Safety Evaluation Report f or the application filed by the University of Florida (UF) for a renewal of operating license number R-56 to continue to operate their Argonaut-type research reac tor has been prepared by the Of fice of Nuclear Reac tor Regulation of the U.
S.
Nuclear Regulatory Commission.
The facility is owned and operated by the University of Florida and is located on the UF campus in Gainsville, Alachua County, Florida.
The s taf f concludes tha t the reactor facility can continue to be operated by UF without endangering the health and safety of the public.
16
NUREG-0916: SAFETY EVALUATION REPORT RELATED TO RESTART OF R. E.
GINNA NUCLEAR POWER PLANT. Docket No. 50-244. (Roch ester Gas And Elec tric Corporation)
- Office of Nuclear Reactor Regulation, Director.
May 1982.
250pp.
8206100045.
13478:034.
This report documents NRC's evaluation of the tube rupture which occurred at the R.
E. Ginna Nuclear Power Plant on January 25, 1982 This plant, which is located in Wayne County, New York, is owned and i
operated by Rochester Gas and Electric Corporation.
In NUREG-0916, the staff has determined, based on conclusions reached in Section 10.0, that operation of the Ginna plant would be acceptable subject to the commitments contained in Section 9. 0 of that report which have been incorporated into the license as conditions.
NUREG-0916 ERR: SAFETY EVALUATION REPORT RELATED TO THE RESTART OF R. E.
GINNA NUCLEAR POWER PLANT. Doc ket No 50-244. (Rochester Gas And Electric Corporation)
- Office of Nuclear Reactor Regulation, Director.
May 26, 1982.
2pp.
8206110182.
13493:268.
This rep ort documents NRC's evaluation of the tube rupture which occurred at the R.
E. Ginna Nuclear Power Plant on January 25, 1982 This plant, which is located in Wayne County, New York, is owned and operated by Rochester Gas and Electric Corp oration.
In NUREG-0916, the staff has determined, based on conclusions reached in Section 10.0, that operation of the Ginna plant would be acceptable subject to the commitments contained in Section 9. 0 of tha t report which have been incorporated into the license as conditions.
NUREG-0923: ADVANCE NOTIFICATION OF SHIPMENTS OF NUCLEAR WASTE AND SPENT FUEL: Guidance.
- Office of Nuclear Material Safety &
Safeguards. Director.
June 1982.
22pp.
8206240002.
13607:281.
- p. S. Nuc lear Regulatory Commission reg ulations in 10 CFR 70.5b and 73.37(f) require NRC licensees to notify the governor of a state prior to making a shipment of nuclear waste or sp ent fuel within or through the state.
This guidance document was prepared to assist licensees in carrying out those requirements.
NUREG-0925: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF THE TETON PROJECT. Doc k et No. 40-8781. (Teton Exploration Drilling Compang Incorporated)
- Office of Nuclear Material Safety &
Safeguards, Director.
June 1982.
2OOp p.
8207140207.
13845: 010.
This Dra f t Environmental Impact Statement is issued by the U.S.
Nuclear Regulatory Commission in response to the request by Teton Exploration Drilling, Inc. for the issuance of an NRC Source and Byproduct Material License authorizing operation of the proposed Teton Project to mine uranium in situ by injecting a carbonate / bicarbonate lixiviant into the ore body.
The statement considers: (1) al ternative ofuno licensing action, (2) alternative er. orgy sources, and (3) alternatives if uranium ore is mined and refined on the site.
The proposed action is to grant a Source and Byproduct Material License to i
the applicant subject to the stipulated license condition.
NUREG-0926: TECHNICAL SPECIFICATIONS FOR GRAND QULF NUCLEAR STATION, UNIT NO.
- 1. Docket No. 50-416.(Mississippi Power and Light Company)
Division of Safety Technology.
June 1982.
210pp.
8207020058.-
13699:331.
The Grand Gulf Nuclear Station, Unit 1 Technical Specifi cations were prepared by the U.S.
Nuclear Regulatory Commission to set forth I
17 4
- - - - - - - - - - - - - - - - ~ ~ - - - ~
~ - ~ ~ ~ ' " ~ ~ - - ~ ~ ' '
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 saf ety of the public.
I NUREQ/CP-OO22: PROCEEDINGS OF THE SYMPOSIUM ON UNCERTAINTIES ASSOCIATED WITH THE REGULATION OF THE QEDLDQIC DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTE. KOCHER,D.C.
Oak Ridge National Laboratory.
April 1982.
600pp.
8205110116.
CONF-810372.
13049:103.
The primary purpose of this symposium was to provide a f orum for wide-ranging discussions on (1) technical aspects related to the development of standards for regulating geologic disposal of high-level radioactive waste, with particular emphasis on the sources and magnitudes of uncertainties associated with current methods f or predicting post-closure repository performance and potential health risks to future generations, (2) important licensing and regulatory issues involved in geolog1: waste disposal, and (3) the current social l
and political climate in which issues of high-level waste management are being deb ated.
Significant contributions to these discussions were provided by representatives from the U. S.
Nuclear Regulatory Commission (NRC),
U. S.
Department of Energy, U. S.
Environmental Protection Ag ency (EPA), various contractors of these three agencies, and other interested parties not affiliated with the Federal Government or its contractors.
The symposium was timed to coincide with the development and publication by the NRC of the proposed technical criteria for regulating the disposal of high-level radioactive wastes in geologic r epositories.
An additional subject of considerable interest at the symposium was the development of environmental radiation protection standards f or high-level radioactive was te by the EPA and the relationship of these standards to the NRC's prop osed technical criteria.
NUREC/CP-OO26: WORKSHOP ON PSYCHOLOGICAL STRESS ASSOCIATED WITH THE PROPOSED RESTART OF THREE MILE ISLAND, UNIT 1.
WALKER,P.s FRAIZE,W.E.s GORDON.J.J.s et al.
Mitre Corp.
Apr il 1982.
152pp.
8204210661.
MTR-82W26.
12799:032.
On 4 and 5 February 1982, eleven e xp er ts in the field of psychological stress and related fields met for a two-day Workshop at the MITRE. Corporation, McLean, Virginia.
The general purpose of the Workshop, sponsored by the Nuclear Regulatory Commission, was to assess the state-of-knowledge relevant to assessing psychological stress which may be associated with the restart of the nuclear power reactor Unit 1 at the Three Mile Island site of the Metrop olitan Edison Company 7
l (TMI-1).
Of particular interest was the extent to which existing concepts and studies might be used to extrapolate or infer th e range of stress responses likely to result from the proposed restart of TMI-1.
This report summarizes the discussions of the Workshop partic ipants.
NUREO/CP-OO31 VO1: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON
- Division of Technical OPERATOR TRA*NING AND GUALIFICATIONS.
Inf ormation & Document Control.
June 1982.
371pp.
8207210132.
CSNI REPT NO.63.
13992:053.
The events during the accident at TMI-2, along with others identified in retrospect at oth er nuc lear p lants, re-emphasized the critical role of the reactor operator.
Many countries are focusing greater attention on the capabilities of control room operating staff and on the problems they face.
In view of the importance to safety on the subject, the CSNI Subcommittee on Licensing decided that a j
18 l
l l
i specialist meeting should be held on the broad aspects of operator selection and training and the functions and organization of operating staff.
The meeting focused on the functions, role and organization of control room personnel as a crew and as individualss selection and qualifications of personnels operator training and requalifications evaluation of crew and individual performances professional and career alternatives f or control room p ersonnels and " concepts for the future" (e.g.,
implementation and impac t of computer technology, advanced simulator concepts, off-site monitoring and support).
Fourteen countries and three international organizations were represented.
This report consis ts of two volumes.
NUREQ/CP-OO31 VO2: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON OPERATOR TRAINING AND GUALIFICATIONS.
- Division of Technical Information & Document Control.
June 1982.
364pp.
8207210130.
CSNI REPT NO.63.
13991:076.
The events during th e ac c i d ent a t TMI-2, along with others identified in retrospect at oth er nuclear p lants, re-emphasized the critical role of the reac tor op erator.
Many countries are focusing greater attention on the capabilities of control room operating staff and on the problems they face.
In view of the importance to safety on the subject, the CSNI Subcommittee on Licensing decided that a specialist meeting should be held on the broad aspects of operator selection and training and the functions and organization of operating staff.
The meeting focused on the functions, role and organization of control room personnel as a crew and as individuals; selection and qualifications of personnels operator training and requalifications evaluation of crew and individual performances professional and career alternatives for control room p ersonnels and " concepts for the future"
( e. g.', implementation and imp a c t of computer technology, advanced simulator concepts, off-site monitoring and support).
Fourteen countries and three international organizations were represented.
This report consis ts of two volumes.
NUREG/CR-0169 V13: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES.
Volume XIII. Temperature Measurements. LASSAHN, G. D.
EG&G, Inc.
April 1982.
44pp.
8205110101.
EQG-2037.
13037:142.
Estimates of measurements for thermocouples and resistance thermometer used to measure temperatures in the Loss-of-Fluid Test (LOFT) system during experiments are provid ed.
The estimated uncertainties were obtained by evaluating the temperature measurements to determine possible errors and then combining the errors for each j
measurement. The evaluation showed that dif ferent uncertainity components are important for dif f erent temp erature measurements and that no one error source is a major source of uncertainty for all the LOFT temperature measurements.
i NUREQ/CR-02OO ERR: SCALE: MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LICENSING EVALUATION.
- Dak Ridge National Laboratory.
April 22, 1982.
1p.
8204230011.
12834:335.
This manual provides documentation for a new, mul t i-f ac e ted computational system called SCALE (Standardized Computer Analyses for i
Licensing Evaluation) that has been develop ed to provide a standard analysis tool for use by the NRC staff and licensees in evaluating nuclear fuel facility and package designs.
The SCALE system consists of several automated analytical sequences (control modules) which perform criticality, shielding. and/or heat transfer calculations with 19
a minimum of user-required input.
The computer codes (functional modules) used within each analytical sequence can also be run in a stand-alone fashion or coupled together in a sequence determined by the user.
NUREQ/CR-1030 VO2: SEDIMENT AND RADIONUCLIDE TRANSPORT IN RIVERS. Phase 2-Field Sampling Program For Cattaraugus And Buttermilk Creek s, New York. WALTERS,W.H.3 ECKERT.R.M.s ONISHI,Y.
Battelle Memorial Institute, Pacific Northwest Laboratory.
April 1982.
180pp.
8205060002.
PNL-3117.
13002:001.
A field sampling program was conducted on Cattaraugus and Buttermilk Cr eeks, New York during September 1978 to investigate the transport of radionuclides in surface waters as part of a continuing program to provide data f or application and verification of Pacific Northwest Lab oratory 's (PNL) sediment and radionuclide transp ort model, SERATRA.
Suspended sediment, bed sediment, and water samples were collected during low flow conditions over a 45 mile reach of stream channel.
Radiological analysis of these samples included primarily gamma ray emitters; however, six alpha and beta-emitting radionuclides were analyzed using radiochemical methods.
The Nuclear Fuel Services facilities are a possible source of two gamma-emitting radionuclides:
- 1) Cesium-134, and 2) C e s i u m-137.
The principal beta-emitter found was Strontium-90.
Elevated levels of both Cesium-137 and Strontium-90 were found at the sampling stations immediately downstream of the facilities.
Based on downstream trends of activity levels of other radionuclides, the Nuclear Fuel Field Services facilities may also be a possible source of Plutonium-238 and 239, 240, Americium-241, Curium-244, and Tritium.
This field sampling ef fort is the second of a three phase program to collect hydrologic and radiologic data at three different flow conditions.
NUREO/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION i
1, COMPUTATIONAL ANALYSIS PACKAGE. PAR ZI ALE, A. A. 3 PATENAUDE, C. J. 3 i
RENARD P.A.s et al.
Lawrence Livermore Laboratory.
April 1982.
145pp.
8206100031.
UCID-18146.
13474:172.
A methodology, called the Structured Assessment Approach (SAA),
has been developed to assess the effectiveness of material control and accounting safeguards systems at nuclear fuel cycle facilities. The methodology has been refined into a computational tool, the Version 1 analysis package, that has been used first to assess a hypothetical nuclear fuel cycle facility and more recently to assess operational nuclear plants.
The Version 1 analysis package is designed to analva s sa feguards systems that prevent the diversion of Special Nuclear Material from nuclear fuel cycle f acilities and to provid e assurance that d iversion has not occurred.
NUREQ/CR-1245 RO1: CORRECTIONS AND ADDITIONS TO USER 'S GUIDE FOR SNAP.
(NUREG/CR-124 5, SAND 80-0315). POLITO,J.
Sandia Laboratories.
May 1982.
95pp.
8205190005.
SANDB2-7017.
13186:259.
This doc ument contains corrections and additions to the " User's Guide for SNAP" (NURE0/CR-1245, SANDB2-0315).
These update the SNAP report so tha t it documents the most curren t version of SNAP.
An additional pr ogram, BATLE Statistics (BSTAT), is described here.
It provides a post-processing capability to analyze engagement data from SNAP simulations.
20
NUREG/CR-1594 VO4: ADVANCED REACTOR SAFETY RESEARCH OUARTERLY REPORT OCTOBER-DECEMBER 1980.
- Sandia Laboratories.
April 1982.
362pp.
8205040028.
SAND 80-1646.
12969:288.
The Advanced Reactor Safety Research Program, initiated in FY 1975, is a comprehensive research activity to assure that the necessary safety data and theoretical understanding emists to license and regulate the Liquid Metal Fast Breeder Reac tor (LMFBR) or oth er advanced converters, breeders or advanced light water reactors which may be commercialized in the United States. Recently the emph asis has shifted toward applying advanced reactor sa fety technology to LWR Class 9 accident concerns which have been of considerable interest following the accident at TMI-2.
For FY 1981 the program is organized in the following Tasks, progress on which is repor ted herein.
Task 1 Advanced Reactor Core Phenomenology, Task 2 Light Water Reactor (LWR) Severe Core Damage Phenomenology, Task 3 Core Debris Dehavior -- Inherent Retention, Task 4 Containment Analysis, Task 5 Elevated Temperature Design Assessment, Task 6 LMFBR Accident Delineation, and Task 7 Test and Facility Technology.
NUREG/CR-1622: FLOW MEASUREMENT DY PULSED-NEUTRON ACTIVATION TECHNIGUFS AT THE PKL FACILITY AT ERLANGEN (GERMANY). KEHLER,P.
Argonne National Laboratory.
April 1982.
140pp.
8205130241.
ANL-C T-81-35.
13087:001.
Flow velocities in the downcomer at th e PKL facility (in Erlangen, Germany) were measured by the Pulsed-Neutron activation (PNA) technique.
This was the first time that a fully automated PNA system, incorporating a dedicated computer for on-line data reduction, was used for flow measurements.
A prototype of a portable, pulsed, high-output neutron source, developed by the Sandia National Laboratories for the U. S.
Nuclear Regulatory Commission, was also successfully demonstrated during this test.
The PNA system was the primary flow-measuring device used at the PKL, covering the whole range of velocities of in teres t.
In this series, the PKL simulated s ma l l-b'r e a k accidents similar to the one that occurred at TMI.
The flow velocities in the downcomer were, therefore, very low, rang ing b e tween 0. 03 and 0. 35 m/sec.
Two additional flow-measuring methods were used over a smaller range of velocities.
Wherever comparison was possible, the PNA-derived velocity values agreed well with the measurements performed by the two more conventional methods.
NUREG/CR-1636 VO4: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: EFFECTS OF VARI ABLE HYDROLOGIC PATTERN 5 ON THE ENVIRONMENTAL TRANSPORT MODEL. DROWN,J.D.; HELTON,J.C.
Sandia Laboratories.
April 1982.
11dpp.
8205030642.
SAND 79-1909.
12949:033.
The Environmental Transport Model is a compartment model which represents radionuclide movement through a surface hydrologic system.
Some of the parameters in the model are bas ed on water and solid flow rates between various compartments in the system.
Mean yearly flow rates have been used in the calculation of these parameters, whereas the flow rates are (at best) periodic functions of time or (more realistically) periodic stochastic processes.
This report presents the results of an investigation into the ef fect s that these variable hydrologic pa tterns have on the Environmental Transport Model.
21
NUREG/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS PROGRAM: Calvert Cliffs No. 2 PWR Power Plant. HATCH,S.W.4 KOLB, 0. 0.
Sandia Laboratories.
CYB ULSKIS, P.
Battelle Memoeial Institute, Columbus Laboratories.
June 1982.
240pp.
8206240048.
SAND 80-1097 VO3.
13618:016.
This volume represents the results of the analysis of th e Calvert Clif fs Unit 2 Nuclear Power Plant which was performed as part of the Reactor Safety Study Methodology Applications Program ( RSSMAP ).
The RSSMAP was conducted to apply methodology developed in the Reactor Safety Study (RSS) to an additional group of plants with the following objectives: (1) identification of the risk dominating accident sequences for a broader group of reac tor designs; (2) comparison of these accident sequences with those identified in the RSS; and (3) based on this comparison, identification of design differences which have a significant impact on risk.
Significant use of RSS insights and results was made for the Calvert Cliffs analysis.
Loss o f coolant accidents (LOCAs) and transients were used as initiating events.
The release categ ories, human error, and c omp on en t failure data bases were the same as those used in the RSS.
The transient and LOCA event trees f or Calvert Clif f s differ somewhat from the RSS event trees due to dif f erent sys tems and interactions among sy stems at Calvert Clif Ps.
In addition, the RSSMAP transient and LOCA trees are interrelated in recognition that transient initiating events may ultimately lead to LOCA conditions.
A " Survey and Analysis" technique was used to identify the most likely failure modes of a system.
The determination of which accident sequences result in core melt and the subsequent containment response and release was made by the MARCH and CORRAL codes.
NUREG/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDIA: Tec hnical Review of NUREG/CR-1636, Vols 1,2 and 3, Dec emb er 1,1981-March 31,1982. STEVENS,C.A.s FULLWOOD,R.R.s AMIRIJAFARI,B.4 et al.. Science Applications, Inc.
June 1982.
98pp.
8207140108.
SAI-288-82-PA.
13846:125.
This project is an ongoing independent technical review of products from an NRC research program to develop a risk-based methodology for assessing the long-term risk of a nuclear waste repository in a geologic medium.
This report presents a review of three technical ceports on environmental transport modeling of the risk methodology.
NUREG/CR-1681: WRAP-PWR VER IFICATION STUDIES. GREGORY,M.V s BER ANEK, F. s AMES,P.L.s et al.
Savannah River Laboratory.
May 1982.
110pp.
8206090132.
DPST-80-4.
13442:199.
A modular computational system k nown a s the Water Reactor Analysis Package - Evaluation Model ( WR AP-E M ) was developed for the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor EM methods and computed results.
A subset of the system (WRAP-EM) provides the computational tools to perform a complete analysis of l oss-of-coolant acc idents ( LOC A 's ) in pressurized water reactors (PWR 's ).
A set of calculations modeling experimental tests in the Semiscale and LOFT facilities, and calculations of a larg e break in a typical four-loop Westinghouse PWR plant have verified that the WRAP-PWR-EM system is functioning as intend ed.
NUREG/CR-1820: STATUS REPORT ON THE FISSION-PRODUCT RESEARCH PROGRAM.
CUMMINGS,J.C.s SALLACK,R.A.s ELRICK,R.M.
Sandia Laboratories.
April 22
1982.
59pp.
8204290629.
SAND 80-2662.
12898:342.
This c=eliminary report discusses the status of fission product research conducted through September, 1980 as a part of a program entitled " Separate Ef f ects Test s for TRAP Code Development. "
We have used transpiration and microbalance techniques to measure vap or pressures, s t udy vapor-vap or an d vap or-wall reactions, and measure surface absorption / desorption rates of fission product species.
We are currently constructing a Fission-Product Reaction Facility (FPRF) to study the chemistry o f fi ssion-prod uc t 'sp ec ies in a high-temperature steam environment.
A Raman spectroscopy diagnostic setup, for use with the FPRF, is being tested and calibra ted wi th an interim Raman cell.
NUREC/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Volume 1: System Models And Numerical Methods. RANCOM,V.H.4 WAGNER,R.J.4 TRAPP,J.A 4 et al.
EG&G, Inc.
April 1982.
129pp.
8205060137.
EGG-2070.
13001:107.
The RELAP5/ MOD 1 code is described in three volumes: Volume 1, System Models and Numerical Methodss Volume 2, Users Guide and Input Requirements; and Volume 3, Checkout Problems Summary.
Volume 1 contains technical developments of the basic thermal-hydraulic model, constitutive relations, and solution scheme.
The adaptations of the basic model f or system components such as pumps, valves, accumulators, and branches are discussed with development of the core neutronics and control system models.
Volume 2 gives recommendations on code application and detailed input requirements.
Volume 3 summarizes the descriptions and results of example checkout problems to which the RELAP5/ MOD 1 code was applied.
The problems range from simple, separate-effects tests to integral LOFT exp eriment simulations.
Existing data are compared to code results.
NUREG/CR-1826 V02: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements. WAGNER,R.J.s CARLSON, K. E. s iGAPP,J.A.4 et al.
EG&G, Inc.
April 1982.
179pp.
8205060192. TGG-2070.
13000:001.
The purpose of Volume 2 of t h e R EL AP,5 d o c rem e ri ta t i o n ' i s to provide sufficient in f ormation to allow app li catio.n of RELAPS to thermal-hydra ulic systems.
This volume assumes that the user has some familiarity with the RELAP5 models described in Volume 1.
This volume has two principal parts.
The first describes the RELAPS prog ram Jrnm the user's viewpoint.
Each model.or feature is discussed with emphapis on how the user uses the feature to represent a physical system.
Input data requirements, user options, and. drscriptions of available output are included.
A description of the-programming features of RELAPS has not been prepared, so this volume ircludes some information regarding a
required files and use of the p r ogram. tra n s mi t ta l tape.
The second cart is a detailed description of the input data' requirements and format.
This information is maintained as a file of 72 character records and a copy of this file is included on the transmittal tape.
This information is formatted by TEXTJAB to a report form that can be printed on a Cyber printer with an upper / lower case print traAn:
The detailed input description is presented in Appendix A.
NUREC/CR-1851: REACTOR PHYSICS DESIGN CALCULATIONS FCh THE ACPR" UPGRADE. PICKARD P.S.; ODOM.J.P.
Sandia Laboratorios.
June 1982.
171pp.
8206250033.
SAND 80-0764.
13620:137.
This report describes the reactor physics calculations performed for the upgrade of the Annular Core Pulse Reactor.( ACPR).
The ACPR has been in operation since 1967 and has been utilt red for a variety of simulation and reactor safety esperiments involving both transient D
and steady-state operations.
The limitation in peforming such experiments in the ACPR has been the degree to which realistic reactor safety and nuclear effects simulation conditions could be created.
The motivation for the ACPR Upgrade was to increase pulse and steady-state performance with a sufficiently harder neutron energy spectrum to allow a wider range of tests to be performed.
NUREG/CR-1890: A3S,SRSS AND CDF RESPONSE COMBINATION EVALUATION FOR MARK III CONTAINMENT AND DRYWELL STRUCTURES. PHILIPPACOPOULO Brookhaven National Laboratory.
June 1982.
200pp.
8206240054.
BNL-NUREG-51328.
13611:001.
The behavior of a representative Mark III containment and its drywell is investigated with respect to their structural capacity when subjected to various load combinations that may be expected during their lifetime.
Mathematical models based on finite element idealization procedures are developed and verified.
These include three-dimensional finite element models and the so-called stick models of the Mark I II containment sy s tem.
The latter are employed for soil-structure interaction analysis.
Various BNL computer codes are utilized to evaluate structural responses.
A set of dynamic loads originating from LOCA, SRV and EARTHGUAKE are compiled from r eviewing the current literature.
The combinations are performed by employing both the Absolute Sum (ADS) and Square-Root-of-the-Sum-of-the-Squares (SRSS) methods.
In addition a probabilistic evaluation of th e combination outcome is carried out by using a Monte-Carlo technique.
This is done by generating cumulative distribution functions (CDF's) expressing th e nonexceedance probability (NEP) level of the maxima of the combinations.
The results from a large number of combination cases are demonstrated.
NUREG/CR-2000 VO1 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of February 1982.
- Dak Ridge National Laboratory.
April 1982.
61pp.
8205210512.
ORNL/NSIC-2OO.
13215:331.
This monthly report contains Lic ensee Event Report (LER) operational inf ormation that was processed into the LER data file oF the Nuclear Safety Information Center ( NSIC ) during the one month period identified on the cover of this document.
The LERs,from which this information is derived,are submitted to the Nuclear Regulatory Commission (NRC) by nuclear pouer plant licensees in accordance with federal regulations.
Procedures for LER reporting are described in detail in NRC Regulatory Guide 1.16 a n d NUR EG-0161, In s t r u c t i o n s for Preparation of Data Entry Sheets for Licensee Event Reports.
The LER summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility. Component, system,and keyword indexes follow the summaries. The components and systems are those identified by the utility when the LER form is initiatedJ the keywords are assigned by the NSIC staff when the summaries are prepared for computer entry.
NUREG/CR-2000 V01 N3: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of March 1982.
- Oak Ridge National Laboratory.
May 1982.
61pp.
8206110016.
ORNL/NSIC-2OO.
13492:101.
This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file oF the Nuclear Safety Information Center ( NSI C ) 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 Reg ulatory 24
~- - - - - - --
1 Commission (NRC) by nuclear power plant lic ensees in accordance with federal regulations.
Procedures for LER reporting'are described in detail in NRC Regulatory Guide 1.16 and NUREG-Ol61, Instructions for Preparation of Data Entry Sheets for Licens ee Event Reports. The LER summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility.
Component, system, and keyword indexes follow the summaries.
The components and systems are those identified by the utility when the LER form is initiateds th e keywords are assigned by the NSIC staf f when the i
summaries are prepared for computer entry, d
1 NUREC/CR-2OOO VO1 N4: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of April 1982.
- Oak Ridge National Laboratory.
June 1982.
120pp.
8206220025.
DRNL/NSIC-2OO.
13584: 001.
This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file of the Nuclear Safety Information Center (NSIC ) during the one month period identified on the cover of this document.
The LERs, from which this information is derived, are submitted to the Nuclear Regulatory Commission (NRC) by nuclear power plant licensees in accordance with federal regulations.
Procedures for LER reporting are described in detail in NRC Regulatory Guide 1.16 and NUREG-0161, Instructions For Preparation of Data Entry Sheets for Licensee Event Reports.
The LEN summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility.
Component,
+
system, and keyword indexes follow the summaries.
The components and j
systems are those identified by the utility when the LER form is l
initiateds the keywords are assigned by the NSIC staff when the summaries are prepared for computer entry.
NUREG/CR-2OOO VO1 N5: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of May 1982.
- Oak Ridge National Laboratory.
June 1982.
i 102pp.
8207060291.
ORNL/NSIC-2OO.
13744: 001.
This mon thly report contains Licen,see Event Report (LER) operational inf ormation that was processed into the LER data file oF l
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-0161. Instructions for Preparation of Data Entry Sheets for Licensee Event Reports.
The LER summaries in this report are arranged alphabetically by facility 'name and then chronologically by event date for each facility.
Component, system, and keyword indexes follow the summaries.
The compon ents and systems are those identiffed by the utility when the LER form is initiated; the keywords are assigned by the NSIC staff when the summaries are prepared for computer entry.
NUREG/CR-2015 VO4: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
JOHNSON.J.J.; MASENIKOV,0.R.s CHEN.J.C.s et al.
Lawrence Livermore Laboratory.
June 1982.
146pp.
8207060334.
UCRL-53021 VO4.
13746:116.
There were three objectives of the soil-structures interaction (SSI) project of the Seismic Safety Margins Research Program I
(SSMRP).
They were 1) to model SSI for system analysis, using 25 i
- - - - - -,. - ~ -..
~.
,,---c--n--,-
n.,
.----n,-
m_
1 d
state-of-the-art analysis techniquess 2) to identify important parameters in the SSI phenomena through sensitivity studies; and 3) to compare analysis techniques.
SSI was modeled in the systems analysis by the substructure approach, as implemented in the CLASGI family of computer programs.
The CLASSI formulation clearly separate,s the roles of earthquake, soil, and struc tures--a ba sic requirement o f the system analysis.
The calculative process is extremely efficient, as it must be to perform repeated deterministic analyses simulating earthquake occurrences.
The SSI input to the system analysis is detailed.
i SSI analysis of the Zion Nuclear Power Plant was examine d in relation to modeling decisions concerning the free-field ground motion and idealizing the soil-struc ture sy stem.
Specifying the free-field motion includes location of the control point, frequency j
characteristics of the control motion, and the spatial variation of motion.
Idealizing the soil-structure system entails modeling the soil configuration, dynamic soil behavior, foundations, and structures.
A comparison of linear approac hes to SSI analysi s was performed.
2 NUREG/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER j
TESTING. FOX,R.A.a HOOKER,C.D.s HOGAN,B.T.s et al.
Battelle Memorial Institute, Pa cific Northwest Laboratory.
May 1982, 25pp.
8206090211.
PNL-3762.
13457:285.
The experiences of industrial radiographers have indicated that electronic radiation-warning devices become inoperative when they are
[
used under some types of ambient conditions.
This report, as a followup to NUREG/CR-0554 and NUREG/CR-1452, documents the nature of tests performed on several additional commercially available models.
None of the four models tested passed the test for ruggedness and severe environmental conditions.
However, all models passed most of.
the requirements of a Health Physics Society draf t standard o f performance specifications for those devices.
The test procedures used l
in the projec t and the results obtained are discussed.
Conclusions i
from the tests and recommendations concerning potentially use ful modifications to existing devices are presented.
NUREG/CR-2022: TECHNICAL REVIEW OF THE DISPERSION AND DOSE MODELS USED IN THE MILDOS COMPUTER PROGRAM. HORST, T. W. ; SOLDAT,J.K.s BANDER,T.J.
Battelle Memorial Institute, Pacific Northwest Laboratory.
May 1982.
30pp.
8205260118.
PNL-3772.
13243:042.
This rep ort reviews the technical basis of the models used in the MILDOS computer code.
Two major areas are addressed: the mod els used for atmospheric dispersion, and the models used in the food chain and in human dosimetry.
The atmospheric dispersion review investigates relevant topics, such as diffusion meteorology, plume rise, deposition, and resuspension.
The environmental analysis review investigates the food chain model involving retention and translocation assumptions as applicable to the ingestion pathway in humans.
In addition, the human dosimetry model used in MILDOS is discussed in terms of all the appropriate p otential pathways for human exposure.
Suggested modifications are presented for possible revision of the MILDOS computer program.
k f
NUREG/CR-2039: DYNAMIC COMBINATIONS FOR MARK II CONTAINMENT STRUCTURES.
PHILIPPACOPOULOs REICH.M.
Brookhaven National Laboratory.
June f
I l
26 i
_~
1982.
170pp.
8206250015.
BNL-NUREG-51366.
13627:203.
The behavior of a representative Mark II containment is investigated with respect to its structural capacity when subjected to various load combinations that may be expected during its lifetime.
Mathematical models based on finite element idealization procedures are developed and verified.
These include three-dimensional finite models and the so-called stick models of th e Mark II containment system.
The latter are employed for soil-structure interaction analysis.
Various BNL computer codes are utilized to evaluate structural responses.
A set of loads are compiled from reviewing the current literature.
The combinations are performed by employing both the Absolute Summ (ABS) and Sqaure-Root-of-the-Sum-of-the-Sqaures (SRSS) methods.
In addition, a probabilistic evaluation of the combination outcome is carried out by using a Monte-Carlo technique.
This is done by generating cumulative distribution functions (CDF's) expressing the nonexceedance probability (NEP) level of the maxima of the combinations.
The results from a set of 800 combination cases are demonstrated.
NUREC/CR-2053: HEAT TRANSFER ANALYSIS OF THE LWR PRESSURE VESSEL STEEL IRRADI ATION C APSULES IN THE DAK RIDGE RESEARCH REACTOR-PRESSURE VESSEL BENCHM ARK FACILITY. SIMAN-TOV,I.I.
Oak Ridge National Laboratory.
April 1982.
135pp.
8205180123.
ORNL/NUREG/TM-4.
13136:317.
The purp ose of this study was to determine a design for irradiation capsules for the Light Water Reactor (LWR) Pressure Vessel Wall Simulation (PVWS) experiment in the Poolside Facility of the Oak Ridge Research Reactor.
The ex periment 's structural configuration is based on the actual configuration of an LWR PV wall, the surveillance specimen capsule, and the thermal shield.
The design tempera ture at which the metallurgical test specimens are to be maintained in the experiment is based on an LWR PV operating temperature of 288 degrees C.
A detailed investigation of the thermal behavior of the proposed design config uration was performed to arrive at an optimum flexibility design that will ensure a uniform temperature distribution of 280 degrees C plus or minus 10 degrees C for all the test specimens, while allowing for uncertainties in thermal behavior, component dependability, and nuclear heating rates in the iron.
The conc lusions of these studies determine the final design parameters for these irradiation capsules.
NUREC/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
CHAPMAN,R.L.; CHRISTENSEN,D.s DAFOE, R. E. 4 et al.
EG&G, Inc.
May 1982.
100pp.
8206100064.
EGG-CAAD-5629 13481:006.
This rep ort compiles data concerning known and suspected water hammer events reported by BWR and PWR nuclear power plants in the United States from January 1969 to May 1981.
This information is seemarized for each event and is tabulated for all events by plant, i
plant type, year of occurrence, type of water hammer, system affected, basis /cause for the event, and damage incurred.
Information is also included from other events not specifically identified as water hammer related.
The events involve vibration and/or system componen ts similar to those involved in the water hammer events.
These other events are included to ensure completeness of the report, but are not used to point out particular facts or trends.
Also, this report does not evaluate findings which can be abstracted from the data.
27
This rep ort shows a total of 81 DWR and 67 PWR occurrenc es having been reported as water hammer induced over a 12 year period.
Of these, approximately half occurred during preoperational testing, or the first year of commercial operation.
The remainder occurred during normal plant operation, operational surveillance testing and/or maintenance.
The report provides event summaries and corrective action tak en to prevent reoccurrence.
NUREG/CR-2099: COMMON CAUSE FAULT RATES FOR DIESEL GENERATORS: ESTIMATES BASED ON LICENSEE EVENT REPORTS AT U. S.
COMMERCIAL NUCLEAR POWER PLANTS, 1976-1978. ATWOOD,C.L.s STEVENSON,J.A.
EG&G, Inc.
June 1982.
87pp.
8207060337, 13744:106.
This rep ort presents estimates of common cause fault rates and related quantities, based on Licensee Event Reports for diesel generators in nuclear reactors.
The Licensee Event Report da ta base is described.
For estimating rates, the binomial failure rate model is used, extending to allow for the substantial observed plant-to-plant variability, and for shocks that by their nature make all the diesel generators in a plant inoperable.
Every quantity is estimated by both a point estimate and a 90% interval.
All rates are expressed per calendar hour.
NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4.4 - 30 SSTF DESCRIPTION DOCUMENT. B AR TON. J. E. s SCHUMACHER,D.G.s FINDLAY,J.A.s et al.
General Electric Co.
May 1982.
115pp.
8206140333.
EPRI NP-1584.
13508:065.
The 30 degrees Steam Sector Test Facility (SSTF), located at General Electric's Lynn, Massachusetts plan t, is a mockup of a 30 degrees sector of a GE boiling water reactor (DWR).
Its purp ose is to a data base for assessment of best estimate models and provide identification and evaluation of controlling phenomena during the refill phase of a hypothesized BWR loss-of-coolant accident.
lhis report describes the design, construction, and operation of the SSTF.
NUREG/CR-2141 VO4: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM.Guarterly Progress Report For October-December 1981. WHITMAN,G.D.s BRYAN,R.H.
Oak Ridge National Laboratory.
May 1982.
145pp.
8206100030.
ORNL/TM-8252.
13472:155.
The Heavy-Section Steel Technology Program is an engineering research activity conducted by the Oak Ridge National Laboratory for the Nuclear Regulatory Commission.
The program comprises studies related to all areas of the technology of materials fabricated into thick-section primary-coolant containment systems of light-water-cooled nuclear power reactors.
The investigation focuses on the behavior and structural integrity of steel pressure vessels containing crack-like flaws.
Current work is organized into six tasks: (1) program administration and procurement, (2) fracture mechanics analyses and investigations, (3) investigations of irradiated materials, (4) thermal-shock investigations, (5) pressure vessel investigations, and (6) steel cladding investigations.
Thermal strain modification to two-an d three-d imens iona l fracture mechanics were chec ked.
Subcontractors investigated fracture initiation and arrest toughness and the transition from cleavage to fibrous fracture.
Investigation of properties of irradiated steel included statistical analysis of Charpy data and continued irradiation of specimens.
Thermal-shock experiment TSE-6 was conducted.
Welds in intermediate test vessel V-BA and testing of material characterization 28
specimens wer e completed.
Further analyses of pressurized thermal-shock test concep ts were made.
Work commenced o.:
study of the ef f ects of weld overlay cladding on fracture behavior.
NUREG/CR-2172:
SUMMARY
AND BIBLIOGRAPHY OF 5AFETY-RELATED EVENTS AT BOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
MCCORMACK,K.E s GALLAHER,R.B.
Oa k Ridg e Na tional Laboratory, May 1982.
199pp.
8206110011.
ORNL/NSIC-195.
13490:352.
This doc ument presents a bibliography that contains 100-word abstracts of event reports submitted to the U.S.
Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980.
The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were esperienced at the facilities.
These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor.
Full-size keyword and p ermutsd-title inderes to facilitate location of individual abstracts are provided following the text.
Tables that summarire the inf ormation c ontained in the bibliography are also provided.
The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item.
Similar information is given for the various kinds of instrumentation and systems, causes of fa il ures, deficiencies noted, and the time of occurrence (i.e.,
during re f ueling, operation, testing, or construction).
Some of the more interesting events that occurred during the year are reviewed in detail.
NUREG/CR-2172 ERR:
SUMMARY
AND BIBIOLOGRAPHY OF SAFETY-RELATED EVENTS AT BOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
- Dak l
l Ridge National Laboratory.
May 10, 1982.
1p.
8206140327.
i 13509:048.
This doc ument presents a bibliography that contains 100-word abstracts of event reports submitted to the U. S.
Nuclear Regulatory Commission concerning operational events'that occurred at boiling-water-reactor nuclear power plants in 1980.
The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the f ac ili ties.
These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor.
Full-size keyword and p ermuted-ti tl e indexes to facilitate location of individual abstracts are provided following the text.
Tables that summavire the inf ormation c ontained in the bibliography are also provided.
The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item.
Similar information is given for the various kinds of instrumentation and systems, causes of fail ures, deficiencies noted, and the time of occurrence (i.e.,
during refueling, operation, testing, or cons truc ti on).
Some of the more interesting events that occurred during the year are reviewed in detail.
NUREC/CR-2173:
SUMMARY
AND DIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
MCCORMACK,K.E.; GALLAHER,R.B.
Oak Ridge National Laboratory.
May 1982.
270pp.
8206110001.
ORNL/NSIC-196.
13491:199.
This rep ort summarizes the data contained in reports submitted by licensees to the U.S.
Nuclear Regulatory Commission concerning operational events that occurred et pressurized-water-reactor power 29
plants in 1980.
A bibliography c on ta in in g 100-word abstracts of the I
event reports is included.
The 21666 abstracts describe the incidents, failures, and design or construction deficiencies experienced at the facilities.
They are arranged alphabetically by reactor name and then chronologically for each reactor.
Keyword and permuted-title indexes are provided to facilitate location of the abstracts of interest.
Tables summarizing the information contained in the bibliography are also presented and discussed.
Information listed in the tables includes instrument failures, equipment f ailures, system failures, causes of failures, deficiencies noted, and time of occurrenc e ( i. e.,
during refueling, operatton, testing, or construction).
Some of the more interesting events that occurred during the year are reviewed in detail.
NUREG/CR-2173 ERR:
SUMMARY
AND BIBLIOGR APHY OF SAFETY-RELATED EVENTS Al PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
MCCORMACK,K.E.4 GALLAHER,R.B.
Oak Ridge National Laboratory.
May 10, 1982.
1p.
8206110002.
ORNL/NSIC-196.
13493:284.
This report summarizes the data contained in reports submitted by licensees to the U.S.
Nuclear Regulatory Commission concerning operational events that occurred at pressurized-water-reactor power plants in 1980.
A bibliography containing 100-word abstracts of the event reports is included.
The 21666 abstracts describe the incidents, failures, and design or construction deficiencies experienced at the facilities.
They are arranged alphabetically by reactor name and then chronologically for each reactor.
Keyword and permuted-title indexes are provided to facilitate location of the abstracts of interest.
Tables summarizing the information contained in the bibliography are also presented and discussed.
Information listed in the tables includes instrument failures, equipment f ailures, system failures, causes of failures, deficiencies noted, and time of occurrence (i.e.,
during refueling, operation, testing, or construction).
Some of the more interesting events that occurred during the year are reviewed in detail.
NUREG/CR-2181 VO4: PHYSICS OF REACTOR SAFETY. Guarterly Argonne Na tional Laboratory.
April Report,0ctober-December 1981.
- 1982.
28pp.
8205030657.
ANL-81-29.
12928:215.
This quarterly progress report summari zes work done during the months of October - December 1981.
The wor k in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of th e Reactor Safety Appraisals Section.
Work on reactor core thermal-hydraulics is performed in ANL's Components Technolog y Division, emphasizing 3-dimensional code development for LMFBR accident under natural convection conditions.
An executive summary is provided including a statement of the findings and recommendations of the report.
NUREG/CR-2184: COMPARISON OF THE RADIOLOGICAL IMPACTS OF THORIUM AND l
URANIUM NUCLE AR FUEL CYCLES. MEYER,H.R.s WITHERSPOON.J.s NCBRIDE,J.P.4 et cl.
Oak Ridge National Laboratory.
April 1982.
33pp.
8205060091.
ORNL/TM-7868.
13002:238.
A study is being performed f or the Nuc lear Regulatory Commission (NRC) to determine whether the existing reg ulations for the uranium fuel cycles require modification and/or additions in order to regulate thorium fuel cycles.
This report was prepared during Phase 2 of the study and compares the radiological impacts of a fuel cycle in which 1
30
only uranium is recycled, as presented in the Final Generic Environmental Statement on the "Use of Recycled Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO)," with those of the light-water b reeder reactor (LWBR) thorium / uranium fuel cycle in the
" Final Environmental Statement, Light Water Breeder Reactor Program. "
The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing sp en t fuel, and reactor operations.
The major difference between the impacts from the two fuel cycles is the larger dose c ommitments associated with current uranium mining and milling opera tions as compared to thorium mining and milling.
Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses.
No significant d if f erences in environmental radiological impact are anticipated for reactors using either of the fuel cycles.
Radiological impacts assoc iated with routine releases from the operation o f either the thorium or uranium f ue? cyc les can be held to acceptably low levels by existing regulations.
NUREG/CR-2192 VO1 N2: EVALUATION OF ISOTOPE MIGRATION-LAND BURIAL. Guarte rly Progress Repor t, April-June 1981. CZYSCINSKI, K. S. ;
PIETRZAK,R.F.s WEISS.A.J.
Br ookhaven National Laboratory.
May 1992.
35pp.
8206100089.
13480:259.
Results are reported for radionuclide sorption experimen ts performed under anaerobic conditions and as a function of solution / solid ratio for trench shale and waters collected at the Maxey Flats disposal site in Kentucky.
The observed degree of sorp tion (equilibrium K(d)) varied unpredictably as a function of solution to solid ratio.
Measurements of pH and Eh were performed before and after the determina tions to determine if redox conditions were altered significantly during the experiments.
The experimental procedure appears capab le of maintaining anaerobic conditions during mo st of the d e termina ti on s.
Changes in solution / solid ratio appear to affect the observed equilibrium sorption more than any variations in redox state during the determinations.
However, our final evaluation of the proposed test procedure for measuring sorption of radionuclid es from anoxic groundwater is that the test is not completely reliable.
Sinc e further improvements in the experimental procedure are not planned, this type of batch sorption test for anoxic waters will be terminated.
Org an o-ra d i on uc l i d e complex stability e xp er iment s in controlled environment chambers were completed.
Controlled oxidation experiments using disposal site trench waters were initiated.
Preliminary results suggested that high contents of dissolved ferrous iron in trench watert.
can act as redox buffers to preserve low redox conditions during subsurface migration.
Data on coprecipitation of radionuclides on ferric oxyhydroxide will be reported when analyses are c omp le ted.
NUREG/CR-2193 VO1 N2: PROPERTIES OF RADIOACTIVE WASTES AND WASTE CONTAINERS. Guarterly Progress Report, April-June 1981. MORCOS,N.4 WEISS,A.J.
Brookhaven National Laboratory.
May 1982.
77pp.
8206100060.
BNL-NUREG-51410.
13480:292.
An empirical relationship has been developed to estimate the cumulative fractional releases of (137)Cs from simulated waste forme as a function of leaching time and the geometric surface-to-volume ratios.
Data from an ongoing leaching study were used.
The simulated waste forms consisted of organic cation exchange resins solidified in Portland I cement at a waste-to-cement ratio of 0.6 and water-to-cement ratio of 0.4.
The nominal sp ec imen dimensions were: 1-inch diameter x 31
1-inch high, 2-inch diameter x 2-inch high, 2-inch diameter x 4-inch high, 3-inch diameter x 3-ir' h high, 6-inch diameter x 6-inch high, 6-inch diameter x 12-inch
,.e, and 12-inch diameter x 12-inch high.
The waste forms were l e a c i. - a in detonized water using a modified 1 At A leaching procedure.
A study designed to evaluate the leachability of (137)Cs, ( 85 ) Sr.
and (60)Cs from simula**d boric acid waste solidified in Portland Ill compressive strength of the ensuing waste cement and to measuee s
forms before and after leaching was concluded.
Leaching data extending over 229 days are presented.
The simulated waste forms were leached in deionized water using a Lodified IAEA leach ing procedure.
The compressive strength of the specimens was measured initially and after their exposure to a leaching environment for 352 days.
NUREG/CR-2194: CONTAINMENT RESEARCH PRIORITIES. SCICCA,F.W.
Sandia i
Laboratories.
April 1982.
2OO p p.
8204150560.
SAND 81-1370.
12692:001.
This rep ort presents the results of efforts to establish priorities amony key areas of LMFBR c ontainment research.
The research areas are concerned primarily with those ph enomena and events that follow from whole-core accidents that can result in a challenge to the primary and/or secondary containment of an LMFBR.
They are concerned with the accident progression and key factors or aspects which may alter or mitigate the accident progression.
The evaluation was divided into two categories as follows:
A.
Primary containment areas; 1.
Fuel debris location and configura tion f ollowing core damage.
2.
In-vessel fuel debris coolability and characteristic s.
3.
In-vessel core retention structure effectiveness.
4.
Post-Accident Heat Removal ef f ectiveness: global heat removal from damaged fuel.
5.
Energetic recriticality outside of core region, 6.
Primary containment failure modes and failure characteristics.
B.
Secondary containment research areas:
1.
Debris bed coolability, location, and characteristic s.
2.
Fuel-steel-concrete in terac t ions.
3.
Sodium-concrete interactions, 4.
Core retention structure / sacrificial material effectiveness.
i 5.
Post-Accident Heat Removal ( PAHR ) effectiveness.
6.
Gas, vapor and aerosol conditions and behavior.
Three separate efforts were employed in the overall evaluation of containment research priorities.
Th i s wa s done to provide at least a partial check on the consistency and validity of the results obtained.
The most extensive evaluation process employed a weighted-score ranking scheme.
In this process, each of the candidate research areas was compared against three evaluation criteria.
NUREC/CR-2201: FOPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1978. PELOGUIN R.A.J SCHWAB J. D. s BAKER,D.A.
Battelle Memorial Institute, Pacific Northwest Laboratory.
June 1982.
123pp.
8207080063.
PNL-4039.
13795:001.
Population radiation dose commitments have been estimated from reported radionuclide releases from commerc ial power reactors operating during 1978.
Fifty year dose commitments f rom a one year exp osure were calculated from both liquid and atmospheric releases for four population groups (infants, child, teen-ager and adult) residing between 2 and 80 km from each site.
This r eport tabulates the results i
l 32 l
l
i 1
of these calc ulations, showing the dose commitments for both liquid and airborne pathways for each age group and organ.
Also included for each site is a his togram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways.
The total dose commitment from b oth liquid and air *orne pathways ranged from a high of 200 person-rem to o
a low of 0.0004 person-rem with an arithmetic mean of 14 person-rem.
The total population dose for all sites was estimated at 660 person-rem f or ti.e 93 million people considered at risk.
The average individual dose commitment from all pathways on a site basis ranged from a low of 3 x 10-6 mrem to a high o f 0. 08 mr e m.
No attempt was made in this study to determine the maximum dose commitment received by any one individual from the radionuclides released at any of the sites.
NUREC/CR-2204 VO4: ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM.Guarterly Progress Report,0ctober-December 1981. HARDY,J.E.s ROGERS,S.C.; MILLER,G.N.s et al.
Oak Ridge National Laboratory.
May 1982.
26pp.
8206100054.
ORNL/TM-8231.
13471:341.
The performance of the Wes tinghouse Rector Vessel Level Indicating System (RVLIS) during tests S-UT-6 and S-UT-7 (5% cold-leg br eak s in the Semiscale Test Facility) was analyred.
The RVLIS, a system employing dif ferential pressure (dP) cells, gave estimates of vessel level similar to those of Semiscale level instrumentation when measuring over equal spans.
Th ese RVLIS measurements are con servative to vessel coolant levels f or b o th S-UT-6 an d S-UT-7.
At times, the RVLIS indications are greater than th e vess el collapsed liquid level measured by Semiscale instrumentation.
During S-UT-6, level estimate dif f erences b etween RVLIS and Semiscale dPs of up to 215 cm (85 in.)
were observed.
These discrepancies may be explained by differences in Semiscale and Westinghouse pressurized-water reactor internal designs.
Excellent agreement was noted between Semiscale and Westinghouse vessel levels for S-UT-7, an upper-head injection test.
NUREG/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY. BARNTHOUSE,L.W.; VAN WINKLE,W.; GOLUMBEK,J.s et al.
Oak Ridge National Laboratory.
May 1982.
165pp.
8206100043.
ORNL/NUREC/ TM-3.
13476:023.
The purp ose of this three-volume repor t is to publish the individual pieces of testimony involving ORNL staf f in a thre e year adjudicatory hearing on the effects of elec tric power generation on the Hudson River.
Volume II contains four exhibits relating to impingement impacts and three critiques of certain aspects of the utilities' case.
The first exhibit is a quantitative evaluatica of four sources of bias (collection efficiency, reimpingement, impingement on inoperative screens, and impingement survival) affecting estimates of the number of fish killed at Hudson River power plants.
The following two contain, respectively, a detailed assessment of the impact of impingement on the Hudson River white perch population and estimates of conditional impingement mortality rates for seven Hudson River fish populations.
The fourth exhibit is an evaluation of the engineering feasibility and potential biological effectiveness of severat types of modified intake structures proposed as alternatives to cooling towers for reducing impingement impacts.
This volume also consists of critical evaluations of the utilities' empirical evidence for th e existence of density-dependent growth in young-of-the year striped bass and white perch, the es timate of ag e-comp osition of s triped bass spawning stock in the Hudson River, and their use of the Lawler, Natusky, and Skelly 33
1 (LMS) Real-Time Life Cycle model to estimate the impact of entrainment and impingement on the Hudson River striped bass population.
NUREG/CR-2220 VO3: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY. GOODYEAR,C.P.; KIRK,D.L.;
CHRISTENSEN,S.
Oak Ridge National Laboratory.
April 1982.
400pp.
8204290485.
ORNL/NUREG/TM-3.
12893:146.
The purp ose of this three-volume r ep or t is to publish the individual pieces of testimony involving ORNL staf f in a thre e year adjudicatory hearing on the effects of elec tric power generation on the Hudson River.
Volume III addresses the validity of the utilities' use of the Ricker stock-recruitment model to extrapolate the combined entrainment-impingement losses of young fish to reductions in the equilibrium p opulation si ze of adult fish.
In our testimony, a methodology was developed and applied to address a single fundamental question: if the Ricker model really did apply to the Hudson River striped bass population, could the utilities' estimates, based on curve-fitting, of the parameter alpha (wh ic h controls the impact) be considered reliable? The present Volume III includes, in addition, an analysis of the efficacy of an alternative means of estimating alpha, termed the t e c h n f. qu e of prior estimation of beta (used by the utilities in a report prepared for regulatory hearing s on the Cornwall Pumped Storage Project).
Our validation methodology should also be useful in evaluating inferences drawn in the literature from fits of stock-recruitment models to data obtained from other fish stocks.
NUREC/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report October 1-December 31,1981. DALL,S.J.s CLEVELAND,J.C.s HARRINGTON, R. M. s et al.
Oak Ridge National Laboratory.
June 1982.
23pp.
8206220101.
ORNL/TM-8260.
13585:023.
Report c overs progress during Oc t.
- Dec. 1981 under the High-Temperat ure Gas-Cooled Reactor ( HTOR ) Systems and Safety Analysis Program.
Work continued on code development and verification l
activities and included improvements in the ORTAP code steam line model and the ORECA code capabilities for long-term transients.
A preliminary severe accident sequence analysis exercise is presented that includes reactor building release source term, atmospheric dispersion, and radiation exposure calculat ions.
NUREG/CR-2223: AN EVALUATION OF THE SOLID ANGLE METHOD USED IN NUCLEAR CRITICALITY SAFETY. THOMAS,J.T.
Oak Ridge National Laboratory.
June 1982.
107pp.
8206230319.
ORNL/CSD/TM-158.
13603:232.
The solid angle method has long been used to establish safe l
spacings for suberitical units of fissile materials, especially uranium with a low (235)U content.
Analytic representation o f criticality in terms of the total solid angle subtended by th e unit nearest the c enter of an array has permitted an evaluation of the l
margin of subcriticality implicit in an allowable total solid angle, omega (A).
It is shown that the method cannot have general applicability but is dependent upon the typ e of fissile material, the number and sp ecific arrangement of the units in array, and th e array reflector conditions. The method is principally one of comparison.
The relative difference between the allowed total solid angle and the total solid angle corresponding to criticality is a measure o f the safety. This study demonstrates that the arbitrary application of an l
l l
34
omega (A) to an array of fissile material without having established the magnitude of the margin of subcriticality is questionable.
The method is usable provided the area of applicability is defined by a validated method.
NUREG/CR-2229 VO1: BUR LARGE BREAK SIMULATION TESTS--BWR BLOWDOWN / EMERGENCY CORE COOLING PROGR AM. LEE,L.S.s S0ZZI,G.L.s ALLISON,S.A.
General Electric Co.
April 1982.
185pp.
8205120130.
EPRI NP-1783.
13055:222.
The BD/ECC Program is an esperimentally based program jointly sponsored by the Nuclear Regulatory Commission, The Electric Power Research Institute, and The General Electric Company.
The BD/ECC 1A Test Phase of this program involves investigating the integral systems effects of emergency core coolant injection during a hypothetical LOCA.
Tests were conducted in a BWR system simulator, t h e Tw o-L o o p Test Apparatus (TLTA), which features a full-sized electrically heated bundle.
Fluid delivery systems were included to simulate emergency coolant injec tions.
Tests conducted under this program include large break (design basis acciden t), small break, and core uncovery under slow loss-of-coolant (boil-off) transients.
Three separate topical reports are issued, one for each type of test.
This topical covers the large break results.
NUREG/CR-2231: BWR LOW FLOW BUNDLE UNCCVERY TEST AND ANALYSIS.
SEELY,D.S.s MURALIDHARAN,R.
General Electric Co.
April 1982.
2OOpp.
8205200302.
EPRI NP-1781.
13196:004.
A series of separate effects tests was performed to evaluate bundle heat transfer and thermohydraulic flow conditions in a simulated BWR/6 core during a boil-off scenario.
The tests were conducted in the Two-Loop Test Apparatus.
The tests were run using constant bundle powers (near decay heat levels) and at constant pressures to determine the ef f ects o f power and pressures on bundle response.
The resultant measured and derived thermohydraulic quanti ties (such as axial void distributions, two-phase levels, end heat transfer coefficients) are compared to the predictions by current thermal hydraulic analysis methods.
In general, the predicted quantities agree closely with the test measurements.
i NUREC/CR-2238 VO1: ADVANCED REACTOR SAFETY RESEARCH.Ouarterly Report. January-March 19J1.
- Sandia Laboratories.
June 1982.
142pp.
8206100021.
SAN.81-1529 VO1.
13473:024.
Sandia Laboratories' Advanced Reactor Safety Research Program, initiated in FY 1975, is a comprehensive research activity conducted as part of th e NRC's confirmatory research effort to assure that the necessary safety data and theoretical understanding exist to license and regulate the Liquid Metal Fast Breeder Reactor (LMFBR) or other advanced converters, breeders or advanced light water reactors which may be commercialized in the United States.
A portion of the early effort in the program was directed toward obtaining data to support the licensing review of the Clinch River Breeder Reactor (CRBR) and the Fast Flux Test Facility (FFTF).
Recently the emphasis ha s shifted toward applying advanced reactor safety technology to LWR Class 9 accident concerns which have been of considerable interest following the accident at TMI-2.
For FY 1981, the program is crganized in the followng subtasks, progress on which is rep orted herein.
Task 1,
Core 35
Debris Behavior - Inherent Retention. Task 2,
Containment Analysis.
Task 3, Eleva ted Temperature Design Assessment, Task 4.
LMFBR Accident Delineation, Task 5,
Advanced Reactor Core Phenomenology, Task 6, Light Water Reactor (LWR) Severe Core Damag e Phenomenology, and Task 7,
Test and Facility Technology.
NUREG/CR-2279: WATER RELEASE FROM HEATED CONCRETES. KENT,L.A.
Sandia Laboratories.
May 1982.
30pp.
8206090115.
SANDB1-1732.
13442:309.
Water release from three concretes as a function of temperature has been determined experimentally.
Limestone concrete releases more water at a moderate temperature than do magnetite or basalt concretes.
The amount of water in the concrete is 6.2%,6.3%, and 5% by weight For limestone, ba salt and magnetite concretes respectively.
All of the concretes show three distinct weight losses as a function of temperature.
By 450K, 52 to 75 percent of the water is l o s t- -a l l oF the water is lost by 750-80rn.
NUREG/CR-2281 V02: NUCLEAR REACTOR SAFETY. Apr il 1-June 30,1981.
STEVENSON,M.G.
Los Alamos Scientific Laboratory.
April 1982.
Dupp.
8205110125.
LA-9209-PR.
13037:309.
The work that is highlighted here represents accomplishments for the period April 1 - June 30, 1981 by the groups at Los Alamos involved in reactor safety research for the Division of Accident Evaluation, Office of Nuc lear Regulatory Research of th e US Nuclear Regulatory Commission.
Presented are brief overviews compiled by project, along I
with a bibliography of Technical Notes and publications written during this quarter.
NUREG/CR-2281 V03: NUCLEAR REACTOR SAFETY. Jul y 1-September 30,1901.
STEVENSON,M.G.
Los Alamos Scientific Laboratory.
April 1982.
36pp.
8205060162.
LA-9229-PR.
12999:033.
This rep ort represents accomplishments for the period July 1-September 30, 1981 in the areas of Trac Cod e Development (TRAC).
Thermal-Hydra ulic Research for Reactor Saf e ty Analysis, Full-Length Emergency Cor e Heat Trans f er--Systems Effec ts and Separate Ef fects (FLECHT-SEASE T ) tests, TRAC Application to 2D/3D, SIMMER Model Development and Gualification Testing, Meth ods f or Saf ety Ana ly sis of Core Disruptive Accidents, Advanced Converter Safety Research on (HTOR), TRAC Calculation Assistance and User Liaison, and the Severe Accident Sequence Analysis Program for the Division of Accident Evaluation, Of fice of Nuclear Regulatory Re search.
Presented are brief l
overviews compiled by project, along with a bibliography of Technica]
Notes and pub lications written during this quarter.
NUREG/CR-2283: DIRECT OBSERVATION OF MELT BEHAVIOR DURING HIGH TEMPERATURE MELT / CONCRETE INTER ACTIONS. POWERS,D.A.s ARELLANO,F.E.
Sandia Laboratories.
April 1982.
119pp.
8204150561.
SAND 81-1754, 12692:212.
The feasibility of using a pulsed x-ray source and an x-ray image intensification system to provide continuous, real time data on high l
temperature melt behavior during interaction with concrete is l
demonstrated.
A test of the system using a 1972g metallothermically l
generated melt interacting with limestone / common sand concrete is l
described.
Analysis of the recorded x-ray image of the melt is used to l
determine the mode of melt contact with concrete, the time dependence l
3s
of pool swelling due to entrained gas, and the nature of gas injection into the melt.
Localized gas injection is found.
Swelling o f the pool increases with superficial gas velocity to an asymptotic limit.
Results are shown to be consistent with the gas film model of melt-to-concrete heat transfer.
The image data are used to assist interpretation of diagnostic data -- gas generation rate, gas composition and concrete temperatures -- ga thered in the test.
NUREC/CR-2297: SECURITY MANAGEMENT TECHNIQUES AND EVALUATIVE CHECKLIS1S FOR SECURITY FORCE EFFECTIVENESS. SCHURMAN, D. L. s DATESMAN C.H.s TRUITT.J.D.
Applied Science Associates, Inc.
April 1982.
125pp.
8204280027.
ASA NO. 635.
12877:111.
The report presents a system for evaluating and correcting deficiencies in security-force effectiveness in licensed nuclear facilities.
There are f our chec klists whic h security manager s can copy directly, or can use as guidelines for developing their own checklists.
The checklists are keyed to corrective-acti on guides f ound in the body of the report.
In addition to the corrective-action guides, the report gives bac kground information on the nature of security systems and discussions o f various special problems cf the licensed nuclear industry.
NUREG/CR-2299 V04: AEROSOL RELEASE AND TRANSPORT PROGRAM. Guarterly Progress Report For October-December 1981. ADAMS.R.E.s TOBIAS.M.L.
Oak Ridge National Laboratory.
June-1982.
43pp.
8206090119.
ORNL/TM-8307.
13442:309.
This rep ort summarizes progress for th e Aerosol Release and Transport Program sponsored by the Nuclear Regulatory Commission's Office of Nuc lear Regulatory Research, Division of Accident Evaluation, for the period October-December 1981.
Topics discussed include (1) i under-sodium tests in the Fuel Aerosol Simulant Test (FAST) Facility, (2) U(3)D(3) and Fe(2)O(3) in eteam (light-water reactor accident) aerosol experiments in the Nuclear Safety Pilot Plant, (3) generation and characterization of cadmium and CdO aerosols in the basic aerosol experimental program, (4) core-melt tests of Zircaloy-clad fuel capsules. (5) initial results of a piston-model bubble oscillation code allowing liquid bypass, and (6) calculations with the UVABUBL code to compare with underwater and und er-sod ium period measurements in FAST experiments.
NUREG/CR-2300 VO1 R1: DRAFT: PRA PROCEDURE CUIDE. A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants. HICKMAN,J.W.
American Nuclear Society.
Institute of Electrical & Electronic Engineers.
April 1982.
200pp.
8204070131 12593:001.
This procedures guide describes method s for performing probabilistic risk assessments (PRAs) f or nuclear power plants at four levels of scope:
(1) systems analysis; (2) systems and containment analysiss (3) systems, containment, and consequence analysiss and (4) full risk assessment, including external events.
After reviewing its I
objectives and limitations, this document describes the organization and managemen t of a PRA project and then presents procedures for accident-sequence definition and systems modeling, human-reliability analysis, the development of a data base, and the quantification of accident sequences.
Procedures for evaluating the physical processes of core meltd own are presented next, followed by guidance on the evaluation of radionuclide releases f rom th e containment as well as the 37
analysis of environmental transport and offsite consequences.
The analysis of external harctds is discussed next, including procedures for seismic, fire, and flood analyses.
The guide concludes with suggestions for the development and interpretation of results and the performance of uncertainty analyses.
NUREG/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS. WOO,H.H.
Los Alamos Scientific Laboratory.
LIM,E.Y.s DEDHIA,D.D.; et al.
Science Applications, Inc.
June 1982.
237pp.
8206230340.
UCRL-15490.
13595:001.
This rep ort summarizes the work performed during fiscal year 1901 by Science Ap plication, Inc. on the Piping Reliability Project for Lawrence Livermore National Laboratory.
The efforts concentrated on modifications of the stratified Monte Carlo code called PRAISE (Piping Reliability Analysis Including Seismic Events) to make it more widely applicable to probabilistic fracture mechanics analysis of nuclear reactor piping.
Pipe failures are considered to occur as the result of crack-like defects introduced during fabrication that escape detection during inspections.
The code mod ifications allow the following factors in addition to those considered in earlier work to be treated: other materials, failure criteria and subcritical crack growth charac teristics welding residual and vibratory stresses; and longitudinal welds (the original version considered only circumferential welds).
The fracture mechanics background for the code modifica tion is included, and details of the modifications themselves provided.
Additionally, an upda ted version of the PRAISF user's manual is included.
The revised code, known as PRAISE-D was then applied to a variety of piping problems, including various size lines subject to stress corrosion cracking and vibratory stresses.
Analyses including residual stresses and longitudinal welds were also performed.
The results of these analyses indicate that lines sub Ject to stress corrosion cracking (SCC) are more failure prone than ones subject to f a tigue.
NUREG/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR AND SHIELDING STUDIES. FORD,W.E.s DIGGS.B.R.s P ETRIE, L. M.
Oak Ridge National Laboratory.
June 1982.
95pp.
8207190039.
ORNL/CSD/TM-160.
13919:066.
A F(3) 227-neutron group cross-section library has been processed for the subsequent generation of problem-dependent fine-or broad group cross sections for a broad range of applications, including shipping cask calculations, general criticality saf ety analyses, and reactor core and shielding analyses.
The energy group structure covers the range 10(-5) eV - 20 MeV, including 79 thermal groups below 3 eV.
The 129-material library includes processed data for all materials in the ENDF/B-V General Purpose File, several data sets prepared from LENDL data, hydrogen with water-and polyethelene-bound thermal kernels, deuterium with D(2)O-bound thermal kernels, carbon with a graphite thermal kernel, a special 1/V data set, and a dose factor data set.
The library, which is in AMPX master
- format, is' de signated CSRL-V (Criticality Saf ety Reference Library based on ENDF/B-V data).
Also included in CSRL-V is a pointwise total, fission, elastic scattering, and (n gamma) cross-section library containing data sets for all ENDF/B-V resonance materials.
Data in the pointwise library 38
were processed with the infinite dilute approximction at a temperatur of 296 K.
NUREG/CR-2314: AGING WITH RESPECT TO FLAMMABILITY AND OTHER PROPERTIES IN FIRE-RETARDED ETHYLENE PROPYLENE RUBBER AND CHLOROSULFONATED POLYETHYLENE. SALAZAR,E.A.s BOUCHARD,D.A.s FURGAL,D.T.
Sandia Laboratories.
April 1982.
65pp.
8206100010.
SAND 81-1906.
13490:001.
The flammability characteristics of ethylene propylane and chlorosulfonated polyethylene rubbers containing fire-retardant additives, aged in different thermal and radiation environmen ts have been studied.
Flammability parameters for these materials (time to ignition, mass pyrolysis, burning rate and fuel consumption) when exposed to, and aged in thermal, radiation, and th ermal / rad ia ti on environments are discussed.
Two formulations of each type of rubber are compared.
The results are a direct contradiction to expected results based on small-scale flammability tests.
They show that the fire-retarding aEents used in this investigation do not reduc e, and in some cases, contribute to, rubber flammability when exposed to a full-scale fire environment.
In addition, the results show that for full-scale fire conditions, the energy required for ignition of chlorosulfonated polyethylene is lower than that required for ethylene propylene rub bers a complete reversal of expected results.
The effects of ag ing on the tensile-elongation properties have been determined.
Radiation dose-rate ef fects are also discussed.
Results.
show that the fire-retardant additives have a negligible influence on the tes ted ma terials ' tensile-elongtion properties and on material aging, regardless of the aging environment.
The data obtained, however, may be too limited to show significant dose-rate effects.
NUREG/CR-2317 VO1 N3: CONTAINER ASSESSMENT-CORROSION GTUDY OF HLW CONTAINER MATERIALS.Guarterly Progress Report, July September 1981.
AHN,T.M.s SOO, P.
Brookhaven National Laboratory.
May 1982.
21pp.
8206100034.
BNL-NUREG-51449 13474:320.
During this quarter work has been star ted on the corrosion and hydrogen embrittlement behavior of commercially pure titanium (ASTM Grade 2), tic ode-12 ( ASTM Grad e 12), and OFHC copp er, which are primary candidate materials for high level waste containers.
The test environment used is a simulated brine solution typical of bedded salt at 150 degrees C or room temperature.
The immersion test results for these materia ls are in reasonable agreement with previous screening test results of Sandia National Laboratory; electron beam welded titanium and T1 Code-12 samples show higher corrosion rates than the non-welded samples.
NUREG/CR-2317 VO1 N4: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATERI ALS. Ouarterly Progres s Report,0c tob er-December 1931.
AHN,T.M.s SOO, P.
Brookhaven National Laboratory.
June 1982.
4bpp.
8207190050.
BNL-NUREG-51449.
13919:309.
Efforts in this quarter have been conc entrated on the unif orm and crevice corrosion, and bydrogen embrittlement of TiCode-12, which are considered to be potential corrosion failure modes in high level wa t. t e container sys tems.
The weight gain of TiCode-12 in WIPP Brine A is in good agreement with previous results from Sandia National Laboratory.
The selective etching in weld heat-affected zones is considered to be responsible f or the slower weight gain in the welded TiCode-12 and commercially pure (CP) titanium.
The interaction of the oxide film 39
with a salt c ompound precipitated from the solution makes it difficult to correlate the weight gain with the thickness of the oxide film.
The crevice c orrosion of TiCode-12 in neutral brine solutions at 150 degrees C has been identified by the observation of corrosion products and oxygen effects.
The predominant oxide phase inside the crevice is TiO(2).
In order to understand the mechanisms involved, crevice corrosion of CP titanium has also been studied.
The effect o f the oxidizer (produced by radiolysis) on the op en circuit corrosion potential has been studied for TiCode-12 in WIPP Brine B.
For a
- ncentr*+ ion n* 33,000 ppm HC10(3), change in the potential has been observed, which is an indication of enhanced susceptibility to strent.
corrosion cracking in this material.
Fractographic analysis of TiCode-12 and titanium in the study of internal hydrogen effects has been carried out using Scanning Electron Microscopy (SEM).
A limited amount of theoretical work has been performed on the construc tion of potential-pH diagrams for copper and lead a t 100 degrees C.
Measurements have been made of the pressure build-up during gamma irrradiation of brine and the gases generated were analyzed.
NUREG/CR-2334: INTERPHASE TRANSPORT IN HORIZONTAL STRATIFIED CONCURRENT FLOW. JENSEN, R. J. ; YUEN,M.C.
Northwestern Univ.
May 1982.
197pp.
8206220005.
13583:128.
The prob lem of interfacial transport in concurrent, horizontal stratified ga s-liquid systems is considered. Local condensation heat transf er coef ficients and interface shear stress were obtained from mass and f orc e balances.
Based on concurrent stratified air-water flow data, the noncondensing interface shear stress was found to b e a function of the relative velocity between the phase and the liquid fraction.
Incorporated into Linehan's relation for condensing flow shear stress, the correlation was found to estimate the shear velocity for the condensation data considered.
Local condenstation heat transfer coef ficients and gas absorption mass transf er coefficients were found to be directly proportional to the shear velocity.
If the inner scales (ue) and (v/u*) are substituted into Lamont's models for the interface mass transfer coefficient, many features of the present correlation f or scalar transfer are predicted.
The correlations for interfacial shear stress and scalar transport can be combined to yield an iterative technique suitable for an engineering analysis of the interfacial shear, mass, and momentum transfer in a single driving force concurrent system.
I NUREC/CR-2343: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE l
WASTE: THE DNET COMPUTER CODE USER 'S MANUAL. CRANWELL R.M.s CAMPBELL.J.E.s STUCKWISCH.S.E.
Sandia Laboratories.
April 1982.
163pp.
8205030650.
SAND 81-1663.
12928:246.
This rep ort describes a network flow model (DNET) for use in simulating the process of salt dissolution in bedded salt formations.
l Included in the model are the capabilities for simulating processes such as salt creep, subsidence. and thermomechanical effects, all of l
which can effect the salt dissolution process.
The model was developed for use by th e Nuclear Regulatory Commission in the analysis of nuclear waste facilities in deep bedded salt formations.
This document is a user's manual and is intended to facilitate the use of the DNET simulator.
Mathematical equations, submodels and a description of the flow network are given.
A complete input data guide is included as well as four sample problems wi th input dec k descriptions and associated output.
40
NUREG/CR-2350: SENSITIVITY ANALYSIS TECHNIQUES: SELF-TEACHING CURRICULUM. IMAN,R.L.s CONOVER, W. D.
Sandia Laboratories.
June 1982.
163pp.
8207080061.
SAND 81-1978.
13795:102.
This rep ort contains discussions and exercises that illustrate the application of the sensitivity analysis techniques developed at Sandia National Laboratories for the Risk Methodology for Geclogic Disposal of Radioactive Waste Project.
With this report the user may familiarire himself with the application of the Latin Hypercube Sampling (LHS) program and the Stepwise Regression (STEP) program with the groundwater transport model NWFT/DVM to do sensitivity and uncertainty analyses.
The user may require the User's Guides for LHS (SAND 79-1473), STEP (SAND 79-1472), and NWFT/DVM (NUREC/CR-2081) to make full use of this self-teaching curriculum.
This report is one of a series of self-teaching curricula prepared under a technology transfer contract for the U.S.
Nuclear Regulatory Commission. Office of Nuclear Material Safety and Safeguards.
NUREQ/CR-2353 VO2: SPECIFICATION AND VERIFICATION OF NUCLEAR POWFR PLANT TRAINING SIMULATOR RESPONSE CHARACTER ISTICS. Part II:
Conclusions And Recommendations. HAAS,P.M.s KERLIN T.W.s SELBY,D.L.s et al.
Oak Ridge National Laboratory.
May 1982.
76pp.
8206090122.
ORNL/TM-7985/P2.
13456:185.
This r ep ort is the second volume of a two-volume report summarizing the findings, conclusions, and recommendations of a survty study for the U. S.
Nuclear Regulatory Commission Office of Nuclear Regulatory Research: (1) to gather in f ormat ion on standards and practices of the nuclear industry and certa in non-nuclear industries which might be used to specify and verify the response characteristics of nuclear power plant training simulatorss (2) to compare findings from the nuclear and non-nuclear industries in order to identify which of the non-nuclear practices might be profitably adopted by the nuclear industry; and (3) to recommend actions that could be pursued, by NRC or by the nuclear industry, to improve standards and practices.
The first volume of the report summarized the inf ormation gathered from the nuclear and non-nuclear industries.
This report presents the conclusions and recommendations from the survey study.
A general conclusion is: The nuclear industry should adopt and NRC regulatory and research actions should support the systems approach to training as a structured framework for development and validation of personnel training systems.
NUREG/CR-2356: UPDATED INPUT FOR THE WRAP-EM SYSTEM. REED,R.L.s GREGORY M.V.
Savannah River Laboratory.
April 1982.
180pp.
8204290497.
DPST-80-6.
12895:350.
The Water Reactor Analysis Package (WRAP) provides the capability to analyze loss-of-coolant accidents (LOCAs ) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluatiun models (ems).
The specifications for modules in the WRAP-EM system have been presented in previous documents.
This document pre sents revised and updated input speci fictions for the WRAP-EM modul es.
NUREG/CR-2359: ATMOSPHERIC STRUCTURE PRIOR TO TORNADOES AS DERIVED FROM PROXIMITY AND PRECEDENT UPPER AIR SOUNDINGS. TAYLOR G.E.s DAR KOW, 0. L.
Missouri, Univ. of. Columbia.
May 1982.
106pp.
8206110325.
13490:087.
The uniqueness of the thermodynamic and dynamic structure of the atmosphere in the area of imminent tori. ado-bearing storm development 41
)
is analyzed by comparing 115 tornado proximity soundings with upper air soundings made at the same location 6 and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> earlier (precedent soundings) and with soundings made simultaneously at neighboring upper air stations.
The c omp ar i s on s suggest that both the proximity station and the neighboring station upstream with respect to the mean flow in the low level moist air display very similar degreer.
of hydrostatic and potential-convective ins tability by late afternoon.
The principal difference is in the wind profiles at the two locations.
The tornado proximity station displays significantly stronger wind speeds above 1 km with the most striking difference being in the vertical shear of the wind in the layer from 1 to 3 km above ground level.
In this layer the winds at the proximity station show an average increase of about 3 m/sec while the upstream, nontornadic station shows a slight decrease of wind speed with height.
NUREC/CR-2362: RELATIONSHIPS BETWEEN CHARPY V-NOTCH IMPACT ENERGY AND FRACTURE TOUGHNESS. DOUGAN J.R.
Oak Ridge National Laboratory.
April 1982.
103pp.
8204290619.
ORNL/TM-7921.
12899:138.
The Fracture and Irradiation Eff ects Program has been concerned with the development of a better technical basis for the preparation oF regulatory guides regarding the prevention of fracture or excessive deformation under the expected environmental conditions in th e pressure boundaries of light-water reactors.
One program ob Jective ha s been the development of toughness estimates for frac ture analysis in the upper-shelf temperature range for beltline materials that have low upper-shelf Charpy energy values due to rad iation damage.
This report documents the investigation of correlations between Charpy-V-notch impact energy and fracture toughness.
NUREC/CR-2366 VO2: MULTIROD BURST TEST PROGRAM PROGRESS REPORT FOR JULY-DECEMBER 1981. CROWLEY,J.L.
Oak Ridge National Laboratory.
April 1982.
75pp.
8205110109.
DRNL /TM-8190.
13038:003.
The assembly of B-6, the final bundle of the Multirod Burst Test Program, was completed.
The bundle is now being connected to power and instrumentation for the burst test in January 1982.
The test condition heat rate and burst temperature have been revised slightly (5 K/s and 925 degrees C.
respectively), but the purpose remains that of deter.rining the large bundle effect on deformation in the a-B j
transition region.
t Posttest examination of B-5, including a water flow test, epoxy
(
casting, sectioning, and measurement of sec tions, has been completed.
Examination of the data obtained continues, and some preliminary observations are included in this report.
A special series of five single-rod burst tests was conducted at essentially identical conditions for the purpose of investigating statistical variations in deformation.
The variation of the sample standard deviation about the mean value for the burst-to-original circumference ratio was about 11%.
NUREC/CR-2377: TESTS & CRITERIA FOR FIRE PROTECTION OF CABLE PENETRATIONS. WILLIAMSGN.R.B.s FISHER,F.L.
Sandia Laboratories.
April 1982.
107pp.
8205060133.
SAND 81-7160.
13001:001.
A series of experiments are described which evaluate the effects of test furnace pressure differential and excess pyrolyrates on the fire resistance of cable penetrations installed in fire resistive walls.
It is shown that the measured fire resistance of pentrations can be strongly influenced by the pressure difference between the test 42
furnace and the unexposed face of the penetration, and, to a lesser degree, by th e presence or absence of exces s pyroly rates.
Methods for the local introduction of excess pyrolyrates into fire test furnace are discussed.
NUREG/CR-2381: GEOLOGIC AND HYDROLOGIC RESEARCH AT THE WESTERN NEW YORK NUCLEAR EERVICE CENTER, WEST VALLEY,NEW YORK.Pregress Report, August 1979-July 1981. ALBANESE,J.R.; DUNN L.A.4 R OGERS, W. B. ; et al.
New York, State o f.
May 1982.
113pp.
8205200277.
13197:238.
This is a report of the progress made during the first part of a proposed multi year program of geologic and hydrologic investigations at the Western New York Nuc lear Servi ce Cen ter.
The New York State Geological Survey previously worked (1975-1979) on a small part of this.
area, specifically that of the New York Sta te-licensed radioactive waste burial trenches.
During the latest reporting period a large scale topographic map of the 140 hectare si te immediately surrounding the nuclear f uel reprocessing plant has been produced, and th ree additional permanent stream stations have b een installed to allow I
monitoring of most runoff from the site.
Ten holes drilled in the North Plateau determined the geometry of the surfical gravel deposits there.
A system of groundwater monitoring wells was established in these holes.
The second phase of the geomorphic investigations of the Buttermilk Creek drainage basin and a study of the effect of submergence on the geotechnical properties of the burial till were completed.
NUREG/CR-2387: CREDIBLE ACCIDENT ANALYSES FOR TRIGA AND TRIGA-FUELED REACTORS. HAWLEY,S.C.; KATHREN,R.L.
Battelle Memorial Institute, Pacific Northwest Laboratory.
April 1982.
63pp.
8205110094.
PNL-4028.
13038:295.
Credible accidents were developed and analyzed for TRIGA and TRIGA-fueled reactors.
The only potential for offsite exposure appears to be from a fuel-handling accident that, based on highly conservative assumptions, would result in dose equivalents of less than or equal to 1 mrem to the total body from noble gases and less than or equal to
- 1. P rem to the thgroid from radiciodines.
Credible accidents from excest.
reactivity insertions, me ta l-wa ter reac tion s, lost, misplaced, or inadvertent experiments, core rearrangements, and changes in fuel morphology and ZrHx composition are also evaluated, and suggestions for further study provided.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980. CLEVELAND,J.C.; CONKLIN,J.C.;
HARRINGTON R. M. J et al.
Oa k Ri dg e Na tional Laboratory.
April 1982 58pp.
8205130231.
ORNL/TM-8073.
13090:106.
A summary is presented of the ma jor ac complishments of the research program on High-Temperature Gas-Cooled Reactor (HTGR) safety.
This report is intended to help the Nuclear Regulatory Commission establish goals for future research by comparing the status of the work here (as well as at other laboratories) with the perceived safety needs of the large HTGR.
The program includes ex tensive work on d ynami c s-rela ted safety code development, use of codes for studying postulated accident sequences, and use of experimental data for code verification.
Cooperative efforts with other programs are also described.
Suggestions for near-term and long-term research are presented.
43
i i
NUREG/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS. SMITH,A.M.s WRIGHT,A.L.s ROCHELLE, J. M. s et al.
Oak Ridge National Laboratory.
April 1982.
77pp.
8205130249.
DRNL/TM-8085.
13089:263.
4 This data record summarizes 34 uranium dioxide (UO(2))
vaporization experiments performed under water in the Fuel Aerosol Simulant Test (FAST) project.
The FAST pro ject is part of the Oak e
l Ridge National Laboratory Aerosol Release and Transport Program 4
sponsored by the Division of Accident Evaluation of the Nuclear Regulatory Commission.
The underwater tests were performed as a prelude to under-sodium tests and were done to permit characterization of the behavior of UO(2) vapor bubbles for various test conditions.
Included in the report are descriptions of test procedures along with tables and graphs summarizing the results.
I NUREG/CR-2393 ERR: Errata, changing rept number to NUREG/CR-2593,to A USER'S MANUAL FOR COMPUTER CODE RIBD/IRT. THAYER D.D.s LURIE,N.A.
Sandia Laboratories.
April 22, 1982.
Ip.
8205200254.
SAND 82-7013.
13202:355.
The comp uter code RIBD/IRT is a modified version of RIBD-II.
It j
is a grid processor that calculates isotopic concentrations resulting i
from two fiss ion sources with normal.down-c hain decay by beta emission i
and isomeric transfers and inter-chain coup ling resulting from n gamma i
reactions.
Calculations can be made to follow an irradiation history j
through an unlimited number of step changes of-unrestricted duration l
and variability including shutd own periods, restarts at different power levels and/or any other level changes.
Output information includes time-dependent inventories, activities, decay powers, and energy
^
releases for as many as 800 fission products.
Modifications to RIBD-II were necessitated by Loss-of-Coolant Accident (LOCA) studies conducted by IRT Corporation regarding fission produc t source term definition.
i These modific ations permit the user to trac k and modify the j
concentrations of individual elements as th ey decay with time following reactor shutdown.
In essence, one can determine time-dependent fission product source terms resulting from any reactor operating history which then can be used as input into fission prod uct transport code s.
Other modifications to RIBD-II expanded the outpu t information to a ssist the-user in analyring the source term.
This manual describes the modifications to RIBD/II, input requirements and a sample problem.
lhe appendicies give a listing of RIBD/IRT, sample output, and a listing oP a code called ZIP which prepares the library tape for input to RIBD/IRT.
Th e code is available in a UNIVAC 1100/81 version and a VAX 11/780 version.
NUREG/CR-2394 ERR: Errata, changing rept number to NUREG/CR-2594,to A USER 'S MANUAL FOR THE GAB AS SPECTRUM COMPUTER CODE. THAYER.D.D.s LURIE,N.A.
Sandia Laboratories.
April 20, 1982.
1p.
8205200265.
I SAND 82-7014.
13202:356.
l The Gamma and Beta Soectrum computer code (GABAS) was developed at IRT Corporation for calculating time-depend ent beta and/or gamma j
spectra from decaying fission products.
GABAS calculates composite i
fission product spectra based on the technique used by England, et al.,
i in conjunction with the CINDER family of fission product codes.
Multigroup beta and gamma spectra for individual nuclides are folded with their corresponding time-depend ent activities (usually generated by a fission product inventory co de) to produce a c omposite t ime-d e p end en t fission product spectrum.
This manual contains the methodology employed b y C AB AS, input r e quir ement s for proper execution, i
44
a sample prob lem and a FORTRAN listing compatible with a UNIVAC machine.
The code is available in a UNIVAC 1100/81 version and a VAX 11/700 version.
The former may be obtained fecm the Radiation Shielding Information Center (RSIC)s the latter may be obtained directly from IRT Corporation.
NUREG/CR-2403 SO1: SURVEY OF INSULATION USED IN NUCLEAR POWER PLANTS AND THE POTENTIAL FOR DEBRIS GENERATION. KOLBE, R. s GAHAN,E.
Durns &
Roe Co.
May 1982, 100pp.
8206110004.
SANDB2-0927.
13492: 273.
In support of Unresolved Safety Issue, USI A-43,
" Containment Emergency Sump Performance," 8 additional nuclear power plants
( r e p r e s en ta t i ve of dif f erent U. S.
reactors' manufacturers and arc h i tec t-eng in eer s ) were surveyed to identify and document the types and amounts of insulation used, location within containment, components insulated, material characteristics, and methods of installation and attachment.
The plants were selected to ob tain survey information on
" older" plants and supplements survey information previously reported in NUREC/CR-2403.
In addition, a preliminary assessment was made of the potential for migration of the insulation debris which might be generated as a result of the postulated los s-of-coolant accid ent (pipe break).
NUREC/CR-2412: HEAT REMOVAL FROM A STRATIFIED UO2-SODIUM PARTICLE BED.
MITCHELL,G.W.3 LIPINSKI,R.J.; SCHWARZ,M.L.
Sandia Laboratories.
May 1982.
120pp.
8205180110.
SAND 81-1622.
13136:045.
The D6 Debris Bed Experiment is one in a series of Post Accident Heat Removal (PAHR) Experiments being conducted to investigate the coolability o f debris beds which migh t exis t as a result of a severe nuclear reactor accident.
The D6 exp erimen t is the first in the series to investigate the effects of particle size stratification, which would likely exist for many accident scenarios, on debris bed coolability.
The D6 debris bed contained 4.87 kg of UO (2 ) particulate, which formed a bed 114 mm high and 102 mm in diameter.
At low power, heat removal could be described to the conduction equation, with effective bed conductivity in agreement with the Kampf-Karsten relation to within ten percent.
Single phase convection was not observed in the bed.
The power required to achieve dryout ranged from 0.28 to 0.45 W/g for overlying bulk sodium temperatures.
These powers are significantly below that wh ich would be predicted by curr ent models.
Based on evaluation of the data, it appears that stratification suppresses convection, reduces the power required to achieve dryout, and suppresses th e formation of vap or channels which would result in increased coolability.
NUREG/CR-2413: SURVEY OF REMOTE AREA MONITORING SYSTEMS AT U. S.
LIGHT-WATER-COOLED POWER REACTORS. KATHREN,R.L.
Battelle Memorial Institute, Pa c ific Ncrthwest Laboratory.
April 1982.
46pp.
8204280021.
PNL-4106.
12876:285.
A study was made of the capabilities and operating practices, including calibration, of remote area monitoring (RAM) systems at l i g h t-wa t er-c oo l e d power reactors in the United States.
The information was obtained by mail que s t i onna ir e.
Specific design capabilities, including range, readout and alarm features are documented along with the numbers and location rf detectors, calibration and operational procedures.
Comments of respondents regarding RAB systems are also included.
45
NUREG/CR-2416: INITI AL QUANTIFICATION OF HUMAN ERROR ASSOCI ATED WITH SPECIFIC INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS IN LICENSED NUCLEAR POWER PLANTS. LUCKAS,W.J.s LETTIERI,V.s HALL, R. E.
Brookhaven National Laboratory.
May 1982.
20pp.
8206100038.
BNL-NUREG-51480.
13475:030.
This rep ort provides a methodology for the initial quantification of specific categories of human error s made in conjunction wi th several instrumentation and control sys tem componen ts operated, maintained, and tested in licensed nuclear power plants.
The resultant human error rates (HER) provide the first real systems bases of comparison for the existing derived and/or best judgement equi valen t set of such rates or probabilities.
These calculated error rates also provide the first real indicati on of human performance as it related directiy to specific tasks in nuclear plants.
This work of developing specific HERs is both an extension of and an outgrowth of the generic HERs deve loped for safety system pumps and valves as reported in NUREG/CR-1880.
NUREG/CR-2417: IDENTIFICATION AND ANALYSIS OF HUMAN ERRORS UNDERLYING PUMP AND VALVE RELATED EVENTS REPORTED BY NUCLEAR POWER PLANT LICENSEES. SP EAKER D. M. s THOMPSON,S.R.; LUCKAS.W.J.
Brookhaven National Laboratory.
May 1982.
31pp.
8206100063.
DNL-NUREG-51481.
13475:355.
This rep ort provides a useful and adap table data base of human error associa ted with the operation, testing, and maintenance of reactor safety system pumps and valves in licensed nuclear power plants.
To produce this data base, a practical and workable methodology was developed and implemented on more than 3,000 Licensee Event Reports (LERs) which resulted in a human error data base six times larger than indicated by the LERs themselves.
This data base is intended to provide a realistic assessment of the appropriate human error populations required in NUREG/CR-1880.
NUREG/CR-2431: BURN MODE ANALYSIS OF HORIZONTAL CABLE TRAY FIRES.
SCHMIDT,W.H.
Sandia Laboratories.
April 1982.
55pp.
8204290610.
SAND 81-OO79.
12895:140.
Electrical cable fire tests have been conducted at the Sandia Fire Research Facility in Albuquerque, New Mexic o, in order to evaluate cable tray fire safety criteria for the Nuclear Regulatory Commission.
A burn mode c oncep t was developed in order to describe and classify the thermodynamic phenomena which occur in the presence of smoke and to compare the fire growth and recession of dif f erent cable type s under otherwise unchanged fire test conditions.
The importance of deep seated fires in cable trays from the standpoint of propagation, detection, and suppression is emphasized.
The cable tray fire tests demonstrate that fire recession and deep seated fires can result from a descending smoke layer and that reignition and secondary fire growth is possible by readmission of fresh air.
NUREO/CR-2432: A UNIGUE CONCEPT FOR LIGUID LEVEL AND VOID FRACTION DETECTION IN SEVERE FUEL DAMAGE TESTS. TOKAR Z, R. D. s CROWELL,S.L.;
PANISKO,F.E.
Battelle Memorial Institute, Pacific Northwest Laboratory.
May 1982.
47pp.
82052OO2B4.
PNL-4070.
13200: 263.
This rep ort describes a direct-c ontac t ing liquid level and void fraction detection system that is being developed by Pacific Northwest Laboratory.
The measurement technique could be used in the severe fuel damage tests that will be conducted at the Power Burst Facility, Idaho Falls, Idaho, and at the ESSOR reactor, Ispra, Italy.
The detection 46
system could also be retrofitted for commercial operating rea ctors to provide definitive thermal-hydraulic information.
The technique can provide unambiguous, real-time data on liquid level and voi" fraction during normal reactor operation as well as during shutdown and accident conditions.
NUREC/CR-2434: FRAC (FAILURE RATE ANALYSIS CODE): A COMPUTER PROGRAM FOR ANALYSIS OF VARIANCE OF FAILURE RATES. An Application User 's Guide. MARTZ,H.F.s BECKMAN R.J.4 MCINTEER, C. R.
Los Alamos Scientific Laboratory.
May 1982.
52pp.
8205180018.
LA-9116-MS.
13135:238.
Probabilistic risk assessments (PRAs) require estimates of the failure rates of various components whose failure modes appear in the event and fault trees used to quantify accident sequences.
Several reliability data bases have been designed f or use in providing the necessary reliability data to be used in constructing these estimates.
In the nuclear industry, the Nuclear Plant Reliability Data System (NPRDS) and the In-Plant Reliability Data System (IPRDS), among others, were designed for this purpose.
An important characteristic of such data bases is the selection and identification of numerous factors used to classify each component that is repor ted and the subsequent failures of each component.
However, the presence of such factors often complicates the analysis oF reliability data in the sense that it is inappropriate to group (that is, pool) data for those combinations of factors that yield significantly different failure rate values.
These types of data can be analyzed by analysis of variance.
Analysis of variance is a statistical data analysis methodology for use in addressing such questions as:
How do the factors af f ect th e failure rate? What are the estimated effects due to these factors?
Which factor combinations yield the largest failure rate estimates?
Are there factor interactions that significantly affect the failure rate? FRAC (Failure Rate Analysis Code) is a computer code that performs an analysis of variance of failure rates and provides information for estimates answering the above questions.
In addition, FRAC provides failure rate estimates.
NUREC/CR-2435: DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMP ARISONS.
YODER.G.L.s OTT,L.J.4 MORRIS.D.G.s et al.
Oak Ridge National Laboratory.
April 1982.
337pp.
8205130274.
ORNL-5822.
13087:126.
Assessment of six film boiling correlations and one single phace vapnr correla tion has been made using data from 22 steady state upflow rod bunale tests (series 3.07.9):
1.
Dougall-Rosenow, 2.
Dougall-Rosenow (wall Prandt1 number),
3.
Groeneveld 5. 7, 4.
Groenevend
- 5. 9, 5.
Condie-Bengston I V, 6.
Groeneveld-Delorme, and 7.
Dittus-Boelter.
Bundle fluid conditions were calculated using energy and mass cons ervation considerations.
Results of the steady state Palm boiling tests support the conclusions reached in the analysis of prior transient t e s t s 3. 03. 6AR, 3.06AR, 3.06.6B, and 3.08.6C.
Comparisons between experimentally determined and correlation predicted heat transfer coefficients indicate that the Dougall-Rohsenow correlation often overpredicts the heat transfer coefficient, while the Groeneveld
- 5. 7, Groeneveld
- 5. 9, and Condie-Bengston IV correlations tend to be in better agreement with the data.
The Groenevend-Delorme correlation underpredicts heat fluxes near dryout but improves as distance from dryout increa ses.
The Dittus-Boelter correlation, which tends to overpredict the heat transfer coefficient, was evaluated only when equilibrium qualities were greater than 1.
47
related to th e total core volumetric vapor generation rate.
Assessment of commonly used local void-fraction models indicated that of the correlations examined, the Yeh void correla tion was best suited For use under the subject test conditions.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPRDVING THE LICENSING REGULATORY BASE FOR SELECTED FACILITIES ASSOCI ATED WITH THE FRONT END OF THE FUEL CYCLE. C LARK, R. G. ; SCHRIEBER,R.E.s JAMISON,J.D.; et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
April 1982.
112pp.
8204280011.
PNL-4086.
12877:240.
Pacific Northwest Laboratory (PNL) has reviewed the health, safety and environmental regulatory base to assess its adequacy as a guide to applicants for licenses to operate UF(6) conversion f acilitie s and Fuel fabrication plants.
The regulatory base wa s defin,d as the body of documented requirements and guidance to licensees, including laws passed by Congress, Federal Regulations developed by the NRC to implement the laws, license conditions added to each license to deal with special requirements for that specific license, and Regulatory Guider.
The study concentrated on the renewal licensing accomplished in the last few years at five typical f acilities, and included analyses of licensing doc uments and interviews with individuals involved with dif f erent asp ects of the licensing process.
Those interviewed included the NMSS staff, Inspection and Enforcement (IE) officials, and selected licensees.
From the results of the analyses and interviews, the PNL study team concludes that the regulatory base is adequate but should be codified for greater visibility.
PNL recommends that NMSS c lari f y distinctions among legal requirements of the licensee, acceptance criteria employed by NMSS, and guidance used by all.
In particular, a prelicensing conference among NMSS. IE and each licensee would be a practical means of setting license conditions acceptable to all parties.
NUREC/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTI AL OR POISSON DATA.
BECKMAN,R.J.; MARTZ,H.F.i HARPER, M. D. s et al.
Los Alamos Scientific Laboratory.
April 1982.
40pp.
8205130261.
LA-9133-MS.
13088:326.
In conducting probabilistic risk analyses of nuclear power plants a suitable data base must be developed for use in estimating component unavailabilities which are required in quantification of accident sequences.
Often data exists on either the time to f ailure of certain components or the ncmber of component failures in a total operating or test time.
Frequently there is not a single underlying failure rate (lambda) for all of these data and the data represent a mixture of different pop ulations.
Techniques are developed in this manuscript which allow the analyst to classify data as coming from populations with failure rates that either do or do not differ by a specified amount such as an order o f magni tud e.
It is assumed that the failure data either follow an exponential (time to failure observed) or a Poisson (numb er of f ailures observed) distribution and that the true failure rate is itself a random variable with a specified prior i
distribution.
Several different prior distributions are considered in examining the performance of th e meth ods.
For both types of data, three classification schemes are presented.
The first is a c lassical scheme which ignores the prior data.
In th e second scheme, data are classified according to their maximum posterior probability, and the last method involves the minimization of an expected loss function.
48
l NUREC/CR-2442: RELI ABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Tech nical Report. Augus t 1980-S ep t e mb er 1981. GREIMANN,L.4 FANOUS,F.s WOLD-TINSAE.A.4 et a l.
Iowa Sta te Univ.
June 1982.
201pp.
8207190033.
13921:019.
J A best estimate and uncertainty assessment of the resistance of the St. Lucie. Cherokee, Perry, WPPSS and Browns Ferry containment vessels was p erf ormed.
The Monte Carlo simulation technique and second moment approach were compared as a means of calculating the probability distribution of the containment resistance.
A uniform static internal pressure was used and strain ductility was ta ken as the f ailure c riterion.
Approximate methods were developed and calibrated wi th finite element analysis.
Both approximate and finite element analy ses were performed on the axi-symmetric containment structure.
An uncertainty assessment of th e containment strength was then performed by the second moment reliability method.
Based upon the approxima te methods, the cumulative dis tribution for the resistance of five containments (shell modes only) is presented in the text.
NUREC/CR-2455: EXPERIMENTAL INVESTIGATIONS OF BUNDLE BOILOFF AND REFLOOD UNDER HIGH-PRESSURE LOW HEAT FLUX CONDITIONS. HYMAN,C.R.s ANKLAM.T.M.4 WHITE,M.D.
Oak Ridge Na tional Laboratory.
April 1982.
131pp.
8205110657.
ORNL-5846.
13036:038.
Results are reported from high pressure bundle boiloff and reflood tests run during the second series of p re s s ur i z e d-wa t er reactor small-break loss-of-coolant accident (SDLOCA) heat transfer experiments.
Tests were conduc ted at Oak Ridge National Laboratory in the Thermal Hydraulic Tes t Facility ( THTF ), a 64-rod full-length rod bundle heat transfer loop.
Tests discussed include five bundle boiloff tests and five reflood tests.
Tests were p erformed under conditions similar to those expected in an SBLOCA.
NUREG/CR-2456: EXPERIMENTAL INVESTIGATIONS OF UNCOVERED-BUNDLE HEAT TRANSFER AND TWO-PHASE MI XTURE-LEVEL SWELL UNDER HIGH-PRESSURE LOW HEAT-FLUX CONDITIONS. ANKLAM,T.M.; MILLER, R. J. s WHITE,M.D.
Oak Ridge National Laboratory.
April 1982.
311pp.
8205130229.
ORNL-5848.
13076:045.
Results are reported from a series of uncovered-bundle h eat transfer and mixture-level swell tests.
Experimental testing was performed in the Thermal Hydraulic Test Fac ility (THTF).
The THTF is an electrically heated bundle test loop configured to produce conditions similar to those in a small-break loss-of-coolant accident.
The objective of heat transfer testing was to acquire heat transfer coefficients and fluid conditions in a partially uncovered bundle.
Testing was p erf ormed in a quasi-steady-sta te mode with the heated core 30 to 40% unc overed.
Linear heat rates varied from 0.32 to 2.22 kW/m. rod (0.1 to 0.68 kW/ft. rod).
Under th ese conditions pea k clad temperatures in excess of 1050 K (1430 degrees F) were observed, and total heat transfer coefficients rang ed f rom 0. 0045 to 0. 037 W/cm(2). K (8 to 65 Btu /h.ft(2). degrees F).
Spacer grids were observed to enhance heat transfer at, and downstream of the grid.
Radiation heat transfer was calculated to account for as much as 65% of total transfer in low-flow tests.
It is recommended that a reference temperature correlation, based on the modified wall Rey nolds number, be used to predict convective heat transfer in the range 2000 less than or equal to Re(mw) less than or equal to 10,000.
Results of mixture-level swell testing showed that the relative expansion of the boiling length caused by the presence of vapor voids (mi x t ur e-l ev e l swell) was lineJrly 49
NUREG/CR-2469: AN ANALYSIS OF TRANSIENT FILM BOILING OF HIGH-PRESSURE WATER IN A ROD BUNDLE. MORRIS.D.G.; MULLINS.C.B.4 YODER,G.L.
Oak Ridge Nationa l Laboratory.
April 1982.
255pp.
8205120110.
DRNL/NUREG-85, 13053:003.
The following six dispersed-flow film-boiling correlations were assessed using data from three ORNL transient film-boiling experiments conducted in the THTF:
1.
Dougall-Rohsenow.
2.
Dougal-Rohsenow (with Prandtl number evaluated at the wall t emp e r a t ur e, as used in RELAP4-MOD 7),
3.
Groeneveld 5.9, 4.
Groeneveld 5.7, 5.
Groeneveld-Delorme, and 6.
Condie-Bengston I V.
The correlations were evaluated with the bundle fluid conditions calculated using a homogeneous two-phase flow and thermodynamic equilibrium thermal-hydra ulics code.
Comparisons made between experimentally determined heat transfer coefficients and the individual correlations' heat transfer coefficients indicate that (1) the Dougall-Rohsenow correlations of ten overpredict the heat transfer coef ficients; (2) the Groeneveld 5,7, Groeneveld 5. 9, and Condie-Bengston IV correlations tend to be in general agreement with the datas and (3) the Groeneveld-Delorme correlation underpredicts the data.
Equ il ib r ium bundle fluid conditions are reported along with fuel rod simulator surface temperature and heat flux.
Calculated experimental heat transf er coef ficients are also reported.
Uncertainties are reported for calculated heat transfer parameters for one of the transient tests.
Thermodynamic nonequilibrium in the three transient tests was examined with an advanced two-fluid thermal-hydraulics code.
NUREG/CR-2470: THERMONETRY IN THE MULTIROD BURST TEST PROGRAM.
ANDERSON.R.L.1 CARR,K.R.s KOLLIE,T.G.
Oak Ridge National Lab oratory.
June 1982.
91pp.
8206230031.
ORNL /Td-8024.
13593:146.
An impor tant objective of the MRBT program is to improve the understanding of the behavior of the Zircaloy cladding of nuc lear fuel rods under conditions postulated for large-break, loss-of-coolant accidents.
A temperature measurement error analysis was perf ormed For the Type S (C.25-mm-diam, bare-wire) and Type K (0.71-mm-diem, sheathed) thermocouple circuits used to measure the temperature of the Zircaloy-clad, electrically heated fuel-rod simulators in the Multirod Burst Test program (MRBT).
The analysis produced the following estimates for the total maximum errors ir. t h e rang e 300 to 1000 degrees C: Type K thermocouples (worst case of two test facilities) exclusive of thermal sh unting error, which recains to be estimated by mathematical modeling:
+12. 50 d egrees C in addition to
-1. 7 d egrees C l
due to thermocouple cold work.
Type S thermocouples: +10.6 degrees C l
in addition to -1.4 degrees C due to thermocouple cold work.
Eight categories of error sources were studied both analytically and experimentally: thermal shuntings electrical shunting and leakages l
thermocouple decalibration in services thermocouple properties of thermocouple extension wear, plugs, and Jac ks; thermocouple reference Junctions data acquisition systems and elec trical noise.
NUREG/CR-2473: SIMMER ANALYSIG OF PROMPT DURST ENERGETICS EXPERIMEN15.
HITCHCOCK,J.T.
Sandia Laboratories.
June 1982.
48pp.
8206250010.
SAND 81-0933.
13620:307.
The Prompt Burst Energetics experiments are designed to measure the pressure behavior of fuel and coolant a s working fluids during a hypothetical prompt burst disassembly in an LMFBR.
The work presented in this report consists of a parametric study of PBE-SS, a fresh oxide fuel experiment, using S I MMER-I I.
The var i ou s pressure sources in the experiment ar e examined, and the dominant source identified as 50
c s.
s s
3 incondensible contaminant gasses in the fuel.
The important modeling uncertainties and limitations of SIMMER-II as applied to these experiments are discussed.
NUREC/CR-2481: LIGHT WATEP4 REACTOR SAFINY RESEARCH PROGRAM.-Semiannual
Report, April-September 1981. BERMt.N, H.
Sandia' Laboratories.
Ap r i l' 1982.
300pp.
8404300030.
SANOS2_OOO6.
12919:009.
This rep ort covers progress in 5 main programs during April-September 1981.
1.
The Molten Fuel-Concrete Interac tions (MFCI) study presently consists of analytical investigations of the-chemica) and physical phenomena assoc (ated with interactions cetween molton core, materials and concrete.
Such interactirns are possable during hypotnetical fuel melt accidents in light water reactors (LWR s ).
2.
The two main purposes of the stead'eyplosion phenomena program are: (a) to identify expsrimentally the magnitudis and time characteristics of pressure pulses and other initial conditions necessary to trigger and propagate explosive interactions between water and molten light water reactor (LWR) materials; and (b) to assess the probability and consequences of steam explosions during postulated meltdown accidents in LWRs.
3.
The Containment Emergency Sump Performance (CESP) investigates the
~
reliability of ECCS sumps and has two main' purposet: (a) to p'rovide a containment-sump data base to NRC, andc(b) to p t. ovide ECCS sump design information to the nuclear industry.
4.
The goals of the Hydrogen Program are tol quanti f y the threat p oi, e d ;
by hydrogen released during LWR accidents and to gonerate information and equipment concepts which will prevent'or mitigate'that threat..
5.
The combustible Gas in Containment Pt ogram determines th e qsf anti ty and rate of g enerat, ion of hydrogen from the corrosion.of zine (in galvanized steel and in zinc-bearing paints)? located within light water reactor containment buildings.
NUREG/CR-2493: AGUEOUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And-Assessment. B ELL. J. T. s PALMER, R. A.
CAMPBELL,D.A.s et al.. Oa k R i d g e National Laboratory.
May 1982.
77pp.
8206100007.
ORNL -5824'.
i 13474:029.
Radioactive, iodine is among the most significant fission products with respect to pctential environmental insult in the event cP a serious nuclear reactor accident.
Th e potential environmentat,issult will depend on the chemical forms and radioactivities of the sodine inside the reactor containment.
Few publications report studies 67 -
aqueous iodine chemistry at conditions pertinent to a reactor
~
accident.
This raport assesses that chemistry under accident conditions, but based on results from studies of systems ~under convenient experimental conditions.
Several items of interest to iodine chemistry are summarized: (1) redox reacticns of iodine species, (2) hydrolysis and disproportionation reactions, (3) formation and reactions of org.enic iodide, (4) radiation effects on aqueous iodine species, (5) liquid gas phase partitioning, and (6) computer program (IGU) to calculate equilibrium concentrations of ten iodine species.
NUREG/CR-2494: OR-FLAW: A FINITE ELEMENT PRCGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFINED FLAWS IN PLATES, CYLINDERS AND i
PRESSURE-VESSEL NOZZLE CORNERS. ATLUR I, S. t?. s BRYSON,J.W.s BASS,B.R.s et al.
Oak Ridge National Laboratory.
May 1992.
82pp.
8206100022.
I
\\
51
l ORNL/CSD/TM-165.
13473:309.
This report describes the linear elastic finite element computer program OR-FLAW (Dak Ridge-FLAW).
The program directly calculates the mixed-mode stress intensity factors (K(I),K(II), and K(III)) along user-defined flaws in plates, cylinders, and pressure-vessel nozzle corners.
Special three-dimensional crack front elements are used to model the immediate vicinity of the flaw.
These crack front elements have the prop er square root and inverse square root variations for displacement and stresses, respectively.
Regular isoparametric elements are used away from the flaw front.
Interelement displacement compatibility between singular and regular elements is satisfied by assuming an independent boundary displacement field (hybrid-displacement procedures) and using a Lagrange multiplier technique to enforce the compatibility constraint.
The stress intensity f ac tors at various points on the crack front are solved directly along with the unknown nodal displacements.
The program provides for automatic generation of a finite element model incorporating either a mathematical or user-defined flaw.
Generation and analysis of the model are performed with program input consisting of 8 to 12 cards.
Applications of the program to a surface flaw in a flat plate and to a symmetrical corner crack in a plate-hole configuration are described.
NUREG/CR-2495: CHARACTERIZATION OF SOIL TO PLANT TRANSFER CDEFFICIENTS FOR STABLE CESIUM AND STRONTIUM. HOFFMAN,L.G.s KELLER,J.H.
Exxon Nuclear Co.,
Inc. (subs. of Exxon Corp.).
June 1982.
53pp.
8206230352.
13595:239.
Soil and vegetation samples were collected from seven counties in the United States in which commercial nuclear power reactors were sited.
Samples were analyzed f or stable cesium and strontium by atomic emission spectrometry.
In addition, soils were analyzed for major elements content, organic content, pH and ion exchange capacity using standard soil analytical methods.
Soils were classified using U. S.
Department of Agriculture (UEDA) soil. survey maps.
Soil to plant transfer coefficients were calculated for dry vegetation and dry soil and for fresh vegetation and dry soil.
The observed transfer coefficient values are higher than those reported in the U.S.
Nuclear Regulatory Commission's (USNRC) Regulatory Guide 1.109 f or b o th cesium and strontium.
The coefficients vary by a factor of 100 for cesium and by 1000 for strontium for corn.
Low ce sium concentrations in both the vegetation and soil resulted in some ambiguity in the transfer coefficients in some samples.
The soil extraction method used, a mineral acid leach, may result in transfer coefficients higher than those which would have resulted if a total dissolution technique had been used.
The limited number of samples collected at any site precluded any statistical treatment of the data.
NUREG/CR-2496: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS. D ANKS, W. W. s GERTMAN,D.I.s PETERSEN, R. J.
EG&G, Inc.
May 1982.
125pp.
8206100059.
13470:093.
The preliminary findings are that research is needed in the following areas of (CRT) generated displays in order to anchor l
regulatory guidelines and regulations to firm empirical data:
a.
Image Distortions b.
Display Formats c.
Work Surf ace Light Reflections d.
Cognitive Fid elitys e.
Response Times and f.
Phosphor.
l 52
NUREQ/CR-2497 VO1: PRECURSORS TO POTENTI AL SEVERE CORE DAMAGE ACCIDENTS: 1969-1979. A Status Report. Vol.
- 1. Main Report And Ap p.
A,C,D And E.
MINARICK,J.W.s KUKIELKA,C.A.
Science Applications, Inc.
June 1982.
350pp.
8207220676.
ORNL/NSIC-182.
14019:227.
Descriptions of 169 operational events reported as Licensee Event Reports, which occurred at commercial light-water reactors during.
1969-1979 and which are considered to be precursors to potential severe core damage, are presented, along with associated event trees and categorizations and subsequent analyses.
This report summarizes work in (1) the development of methods used to screen approximately 19,400 LER abstracts f or potential precursors, (2) the initial screening of those abstrac ts to determine which should be reviewed in detail, (3) the detailed review of those selected LERs that yielded the 169 events, (4) the categorization of the 169 events. (5) the calculation of function failure estimates based on precursor data, (6) the use of probability of severe core damage estimates to rank precursor events and estimate the frequency of severe core damage, (7) the identification of 52 events considered significante (8) trend s analyses of those significant events, and (9) the identification of th e other events of interest that occurred within 1 month of significant events.
NUREQ/CR-2497 VO2: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACC IDENT: 1969-1979. A Status Rep ort. Vol. 2 - Appendix B.
MINARICK,J.W.s KUKIELKA,C.A.
Science Applications, Inc.
June 1982.
725pp.
8207220678.
ORNL/NSIC-182.
14024: 062.
Descriptions of 169 operational events reported as Licensee Event Reports, which occurred at commercial light-water reactors during 1969-1979 and which are considered to be precursors to potential severe core damage, are presented, along with associated event trees and categorizations and subsequent analyses.
This report summari zes work in (1) the development of methods used to screen approximately 19,400 LER abstracts for potential precursors, (2) the initial screening of those abstrac ts to determine which should b e reviewed in deta il, (3) the detailed review of those selected LERs that vielded the 149 events, (4) the categorization of the 169 events, (5) the calculation of function failure estimates based on precursor data, (6) the use of probability of severe core damage estimates to rank precursor events and estimate the frequency of severe core damage, (7) the identification of 52 events considered significant, (8) trend s analyses of those significant events, and (9) the identification of the other events of interest that occurred within 1 month of significant events.
NUREO/CR-2505: ELECTRICAL IMPEDANCE STRING PROBES FOR TWO-PHASE VOIDS AND VELOCITY MEASUREMENTS. HARDY,J.E.s HYLTON,J.O.
Dak Ridge National Laboratory.
June 1982.
89pp.
8207190054.
ORNL/TM-8172.
13920:280.
Report c overs an instrumentation scheme developed to measure two-phase flow velocity and void fraction during refill /reflood stages of a loss-of coolant accident in experimental test facilities.
The principle operation was based on me4surement of the electrical impedance of two phase mixtures.
Two phase velocity estimated by time-of-fligh t analysis of signals from two spatially separated sensors.
Capacitive technique employed to measure void fraction.
The impedance sensor dubbed " string " prob e consists of a pair of stainles steel wires strung back and forth across a stainless steel frame and was deaigned to withstand temperatures of 350 degrees C, thermal transients of approximately 300 degrees C/s, and severe fluid-and i
condensation-induced shocks.
Void measurements f rom developed string i
53
probes were c ompared with gamma attenuation densitometer valuess velocity measurements by the string probe were compared with calculated phase velocities and turbine meter velocities.
In large open-flow areas (such as an upper plenum or end box), good agreement was found between densitometer void values and string sensor voids.
Flow velocities detemined by the string probe yielded reasonable agreement when compared with turbine and phase velocities.
Generally, the string probe instrumentation (1) proved to be durable in air / water and steam / water flows and (2) demonstrated an ability to measure a wide range of flow velocities (1 to 15 m/s) and void fractions (O to 0.99+).
NUREC/CR-2512: RADI ATION DOSE ESTIMATES AND HAZARDS EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Report July 1980-June 1981. MEWHINNEY,J.A.
Lovelace Biomed & Environmental Researc h Institute.
April 1982.
52pp.
8205060003.
LMF-92.
13002:181.
i The objective of this project is to conduct confirmatory research on aerosol characteristic s and the resulting radiation dose distribution in animals following inhalation and to provide prediction i;
of health consequences in humans due to airborne radioactivity which might be released in normal operations or under accident conditions during production of mixed oxide nuclear fuel.
Four research reports summarize the results of research being conducted.
The first presents results for several types of physical chemical characterizations of aerosol samples collected at an industrial facility during normal fabrication of mixed oxide fuel.
The second paper reports on the methods development process used for measurement of the specific surface area of aerosols, an important determinant in the rate of dissolution of particulates deposited in th e lung.
The third paper provides updated information on the retention, distrbution and excretien of Pu after inhalation by Beagles of aerosols of either 750 degrees C treated UO(2) plus pug (2), 1750 d egrees C treated (U,Pu)D(2) i or 850 degrees C treated " pure" pug (2) inc l ud ing the formulation of a biomathematical model useful in describing the results.
The fourth paper describes the early results from two studies in which Fischer-344 rats received inhalation exposure to aerosols of (U,Pu)O(2) or " pure" pug (2) to determine the relationship of radiation dose to biological response.
NUREQ/CR-2516 VO1 N1: CHARACTERIZATION OF TMI-TYPE WASTES AND SOLID PRODUCTS. Guarterly Progress Report, April-September 1981.
SWYLER, K. J. s WEISS,A.J.
Brookhaven National Laboratory.
May 1982.
48pp.
8206090216.
BNL-NUREG-51499.
13456:260.
Progress is reported on a research program to systematicalig characterize the type of radwestes which may be generated in cleanup procedures following off-normal reactor operations.
Specifically, the program is presently investigating how the properties of wastes containing ion-exchange media may be modified by heavy doses of irradiation from sorbed radionuclides.
Special effort is being devoted toward quantifying the ef f ects of fac tors such as radiation d ose rate, chemical loading on the ion exchangers, moisture content and composition o f external media, etc.,
which may influence the relation between laboratory test results and field p er f orman c e.
NUREC/CR-2518: THERMODYNAMIC PROPERTIES OF WATER FOR COMPUTER SIMULATION OF POWER PLANTS. KUCK I.Z.
Arizona, Univ. o f.
May 1982.
66pp.
8206090125.
13456:304.
54
Steam property evaluations may represent a significant portion of I
the computing time necessary for power system simulations. The iterative nature of the solutions for heat transfer and kinetic equations often requires thousands of steam property evaluations during the execution of a single program.
Considerable savings may be
{
realized by simplification of property evaluations.
Empirica l equations have been obtained for the thermodynamic properties of water in the region of interest.
To maintain thermodynamic consistency, the compressibility factor Z, in terms of pressure and temperature, was obtained by curve fitting, and the enthalpy, entropy, and internal energy were derived by standard relationships.
Formulations for heat capacity, saturation temperature as a function of saturation pressure, and specific volume of saturated water as a function of the saturation temperature were determined by i
curve fitting of independent equations.
Derivatives were obtained by differentiation of the appropriate formulations.
Evaporator and superheater components of a liquid metal fast breeder reactor power plant simulator were chosen as test cases for the empirical representations.
Results obtained using the empirical equations were comparable to those obtained using tabular values and requared 24% less computing time.
NUREG/CR-2521: METHOD FOR ESTIMATING WAKE FLOW AND EFFLUENT DISPERSION NEAR SIMPLE BLOCK-LIKE BUILDINGS. HOSKER,R.P.
Commerce, Dept. o f, National Oceanographic & Atmospheric Administration.
June 1982.
i 158pp.
8207190046.
ERL-ARL-108.
13919:155.
i This rep ort is intended as an interim guide for those who routinely f ac e air quality prob lems associated with near-buil ding exhaust stack placement and height, and the resulting concentration patterns.
Th e report consolidates available data and methods for estimating wa ke flow and ef fluent dispersion near isolated block-like structures.
The near-building and wake flows are described, and quantitative estimates for frontal eddy size, height and extent of roof and wake cavi ties, and far wake behavior are provided.
Concentration calculation methods for upwind, near-building, and downwind pollutant sources are g iven.
For an upwind source, it is possible to estimate the required stack height, and to place upp er limits on the likely near-building concentration.
The inflences of near-building source location and characteristics relative to the building geometry and orientation are considered.
Methods to estimate effective stack height, upper limits for concentration due to flush roof vents, and the effect of changes in rooftop stack height are summarized.
Current wake and wake cavity models are presented.
Numerous graphs of imp ortant expressions have been prepared to facilitate computations and quick estimates of flow patterns and concentration levels for specific simple buildings.
NUREC/CR-2522: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM PLAN. MILLER,R.L.s PAASCH,R.A.
United Nuclear Corp.
April 1982.
32pp.
8205110257.
13038:077.
This Program Plan describes a multi gear program initiated by the
[
Nuclear Regulatory Commission (NRC) to assess and evaluate th e methods, radiation exp osure and costs associated with decommissioning retired nuclear facilities.
The objective of this program is to provide the NRC licensing staff with comparative data that will allow assessment of decommissioning alternatives for regulatory and ALARA implementation of future decommissioning proposals.
The program is currentig limited to nuclear reactors.
55 l
Licensees currently decommissioning a facility or licensees who are planning decommissioning projects will be solicited for inclusion in the program.
An analysis will be parformed for each project and will include a comparison of the methods, costs and exposure usage with data contained in generic decommissioning studies.
NUREC/CR-2525 VO1: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
ANKLAM,T.M.s HUNT,D.F.s THOMPSON, M. S. s et a l.
Oak Ridge National Laboratory.
May 1982.
107pp.
8206090135.
ORNL/NUREC/TM-4.
13454:254.
The report presents experimental data and calculated steady-state and transient instrument uncertainties from Oak Ridge National Laboratory Small Break Loss of Coolant Accident (LOCA) Heat Transfer Test Series I.
The subject test series was composed of six high-pressure, low-flow, quas i-s t ea d y -s ta t e heat transfer tests and six high-pressure reflood tests.
The test series was designed to obtain data under conditions similar to those expected in a small br eak LOCA.
In addition to the experimental data, calculated inlet and outlet mass flows and rod powers are presented.
NUREG/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for Test 3.03.6AR - Transient Film Boiling In Upflow. MULLINS.C.B.s FELDE,D.K.4 SUTTON,A.C.s et al.
Oak Ridge National Laboratory.
May 1982.
249pp.
8206090129.
ORNL/NUREC/TM-4.
13455:164.
Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.03.6AR.
This test was conducted on May 21, 1980.
The objective of the program was to investigate heat transfer phenomena believed to occur in PWRs d ur in g accidents, including small and large break loss-of-coolant accidents.
Test 3.03.6AR was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions.
The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.03.6AR available.
Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.
l f
NUREC/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume l
3-Thermal-Hydraulic Test Facility Exp erimen tal Data Report For Test i
3.06.6B-Transient Film Boiling In Upflow. MULLINS,C.B.3 CDULD S.S.s l
FELDE,D.K.s et al.
Oak Ridge National Laboratory.
June 1982.
260pp.
8206240036.
ORNL /NUREG /TM-4.
13610:076.
Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.06.6B.
This test was conducted b y members of the Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT)
Separate-Effer.ts Program on August 29, 1980.
The objective of the program was to investigate heat transfer ph enomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents.
Te s t 3. 06. 6B wa s conduc ted to obtain transient film boiling data in rod bundle g eometry under reac tor accident-type conditions.
The primary purp ose of this report is to make the reduced instrument responses f or THTF Test 3. 06. 6B available.
Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod power.
56 l
i I
NUREG/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume i
5-Thermal-Hyd raulic Test Facili ty Exp erimen tal Data Rep ort For Test 1
3.08.6C-Transient Film Boiling In Upflow. MULLINS,C.B.s FELDE.D.K.s SUTTON,A.G.; et al.
Oak Ridge National Lab oratory.
June 1982.
259pp.
8207190013.
ORNL /NUREG /TM-4.
13922:004.
Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.08.6C.
This test was conducted b y members of the Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT)
Separate-Effects Program on October 1, 1980.
The o b'J e c t ive of the program was to investigate heat transf er ph enomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents.
T es t 3. 08. 6C was c onduc te d to obtain transient film boiling data in rod b undle geometry under reactor accident-type conditions.
The primary purpose of this report is to make the reduced instrument responses for THTF test 3.08.6C available. Included in the tsport are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.
NUREG/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Ser.es 3. 07. 9-Steady-State Film Boiling In Upflow. MULLINS, C. D. s FELDE, D. K. s SUTTON, A. G. s et al.
Oak Ridge National Laboratory.
June 1982.
101pp.
8207190021.
ORNL/NUREG/ TM-4.
13923:058.
Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was l
conducted by members of the Pre ssuriz ed-Wat er Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, l
September 18, and October 1,
1980.
The objective of the program was to investigate h eat transfer phenomena believed to occur in PWRs during accidents, including small-and large-break loss-of-coolant accidents. Test series 3. 07. 9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions.
This report presents the reduced instrument resp onses For THTF t e s t s er i e s 3. 07. 9.
Also included are uncertainties in the instrument re sponses, calculated mass flows, and calculated rod powers.
i NUREG/CR-2542: SENSITIVITY STUDY USING THE FRANTIC CODE FOR THE UNAVAILABILITY OF A SYSTEM TO THE FAILURE CHARACTERISTICS OF THE COMPONENTS AND THE OPERATING CONDITIONS. GINZBERG,T.; DICKEY,J.M.;
HALL,R.E.
Br ookhaven National Laboratory.
May 1982.
114pp.
8206100087.
BNL-NUREG-51504.
13400:145.
The purpose of this report is to show how the code FRANTIC II can be used to explore the sensitivity of the calculated unavailability of a system to various parameters.
Several generic systems are analyzed in detail.
The study illustrates the effec t of uncertainties in the i
empirical data and helps assess the relative importance of collecting more accurate data for particular components.
The study also helps in deciding whether it is worth improving a particular component or whether such improvement would have little effect.
In addition, the impac t of changing operational procedures can be assessed, and dif ferent operating strategies can be compared.
Change of some procedures may yield only slight imp r ovemen t, whereas in other instances a small change in timing may have significant effect.
- Thus, this report shows how FRANTIC II may be a useful and powerful tool in the analysis of the reliability of comples systems, and i n '. h e determination of the more significant factors.
l 57 4
~., _ _. -.,,,, _ _..
t NUREG/CR-2543: A STUDY OF THE FEASIBILITY OF MICROWAVE DIELECTRIC HEATING FOR LMFBR TRANSITION PHASE ACCIDENT SEQUENCE BOILING STUDIES.
MAKOWITZ,H.s GINSBERG,T.
Brook haven National Laboratory.
May 1982.
121pp.
8206100017.
BNL-NUREG-51506.
13475:234.
A study is reported on the feasibliity of the use of microwave dielectric heating to simulate the nuclear heat source in LMFDR
" transition phase" accident sequence volume-boiling simulation experiments.
The adequacy of microwave heating is judged based upon the criterion of heating uniformity p er uni t liquid volume and upun the ability to analytically characterize th e liquid power density distribution.
Two aspects of liquid power density uniformity are addressed.
First, the effect of liquid geometry on power density ic studied in order to determine whether millimeter-size droplets can be heated as efficiently as centimeter-scale masses which are exposed to the same source of radiation.
Both analyses and experiments were performed in this portion of the study.
Second, the spatial distribution of power density across liquid slabs is studied, in order to determine whether wave interference effects, which lead to severe power density gradients, can be minimized by choice of suitab le dielectric liquids.
The above analyses were carried out for a variety of wavelength s within the microwave radiation band, for several dielectric liquids and for a range of temperature.
NUREG/CR-2544: TWO-PHASE MASS FLUX UNCERTAINTY ANALYSIS FOR THERMAL-HYDRAULIC TEST FACILITY INSTRUMENTED SPOOL PIECES. CHEN, N. C. ;
FELDE,D K.
Oak Ridge National Laboratory.
June 1982.
40pp.
8206220019.
ORNL/TM-7859.
13584:324.
An analysis of two phase mass flux uncertainties for the Thermal-Hydraulic Test Facility (THTF) instrumented spool pieces is presented.
Comparisons are made between various homogeneous mass flux models based on hign-temperature and high-p ressure water mass flux data from steady-state upflow film boiling tests run at the THTF.
Subcooled flow at the test section inlet provides a well-defined standard for inplace evalution of the mass flux models at the high-quality, two-phase test section outlet.
Additionally, a transient two phase turbine meter model developed by Kamath and Lahey i
is applied to two transient THTF tests to a ssess the sensitivity of j
the calculated mass flux to uncertainties in two-phase flow parameters l
and transient response as applied specifically to THTF test conditions.
1 NUREG/CR-2545: DEFIGN CONCEPT AND TESTING OF AN IN-BUNDLE GAMMA DENSITOMETER FOR SUBCHANNEL VOID FRACTION MEASUREMENTS IN THE THTF ELECTRICALLY HEATED ROD BUNDLE. FELDE D.K.
Oak Ridge National Laboratory.
May 1982.
26pp.
8206110005.
13493:301.
A design concept is presented for an in-bundle gamma densitometer system for measurement of subchannel averag e fluid density and void fraction in rod or tube bundles.
This report describes (1) the l
application of the design concept to th e Th ermal-Hydraulic Te st Facility (THTF) electrically heated rod bundle and (2) results from tests conducted in the THTF.
l NUREG/CR-2546: REACTOR SAFEGUARDS AGAINST INSIDE SABOTAGE. BENNETT,H A.
Sandia Laboratories.
June 1982.
104pp.
8206290049.
SAND 82-0319.
13661:165.
A concep tual safeguards system is structured to show how both reactor opera tions and physical protection resources could be 58
l 4
integrated to prevent release of radioactive material caused by insider sabotage.
Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components.
Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed.
Recommendations for further development of safeguards system structures,, operational recovery, and sabotage prevention are suggested.
NUREC/CR-2551: RANK ORDERING OF VITAL AREAS WITHIN NUCLEAR POWER PLANTS. RICHARDSON.J.M.
Sandia Laboratories.
June 1982.
38pp.
8206250040.
SANDB2-0332.
13628:168.
The conceptual development of a methodology for rank order of vital areas within nuclear power plants based upon times associated with sabotage events and their consequences is discussed.
Th e important time parameters in the analysis include the time required to detect the perpetration of sabotage, the time required to repair or mitigate the consequences of the sabotage and the total time available to perform these functions before it is too late to reverse the damage.
These time interval parameters are incorporated into an interruption analysis importance measure that provides information on the ability oP the protection systems to cope with the results of the sabotage.
A consequence analysis that considers categories of release characteristics is the next step in the ranking scheme.
Results of the interruption and consequence analyses are combined to attain a risk index associa ted with each vital area.
The final ranking can be used to order upgrade priorities and to allocate scarce protection resources effectively.
NUREG/CR-2559: RESULTS OF PHASE ONE OF PLANT ELECTRICAL SYSTEM (PES)
STUDY. WYANT, F. J. s FURGAL,D.T.
Sandia Laboratories.
April 1982.
51pp.
8205130248.
SAND 82-0377.
13089:338.
This rep ort summariz es initial scoping study efforts assessing nuclear power plant electrical system performance.
Actual component failures and off-normal load and electrical power line conditions were determined.
Sources of information (data bases) are discussed.
A methodology for coding and classifying Plant Electrical System (PES) events is presented.
Data from 9 LER monthly reports is categorized by component.
This information is rank-ordered and cross-tabulated by frequency of occurrence, generic component, type of reactor, specific plant, component vendor and system interactive failure mode.
Recommendations for further study are presented.
NUREC/CR-2564: ENVIRONMENTAL FACTORS AFFECTING LONG-TERM STABILIZATION OF RADON SUPPRESSION COVERS FOR URANIUM MILL TAILINGS. YOUNG, J. K. s LONG.L.W.; REIF,J.W.
Battelle Memorial Institute, Pacific Northwest Laboratory.
Apr.' 1982.
110pp.
8205110103.
PNL-4193.
13037:186.
Pacific Northwest Laboratory is investigating the use of a rock armoring blanket (riprap) to mitigate wind and water erosion of an earthen radon suppression cover applied to uranium mill tailings.
To help determine design stresses for the tailings piles, environmental parameters are characterized for the five active uranium producing regions on a site-specific basis.
Only conventional uranium mills that are currently operating or that are scheduled to open in the mid 1980's are considered.
Available data indicate that flooding has the most potential for disrupting a tailings pile.
The arid regions of the Wyoming Basins and 59
the Colorado Plateau are subjec t to brief storms of high intensity.
The Texas Gulf Coast has the highest p o tent ia l for extreme precipitation from hurricane-related storms.
Wind data indicate average wind speeds from 3 to 6 m/sec for sites, but extremes o f 40 m/sec can be expected.
Tornado risks range from low to moderate.
The Colorado Plateau has the highest seismic po tential, with maximum acceleration caused by earthquakes ranging from 0.2 to 0.4 g.
Any direct effect from volcanic eruption is neg ligible, as all mills are located 90 km or more from an igenous or hy drothermal system.
NUREC/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED TORNADO CONDI TIONS. HORAK,H.L.; SMITH,P.R.
Los Alamos Scientific Laboratory.
GREGORY,W.S.; et al.
Northear t Missouri State Univ.
May 1982.
103pp.
8205180101.
LA-9197-MS.
13134:165.
This rep ort contains the results of structural tests to determine the response of High Efficiency Particulate Air filters to stimulated tornado condi tions.
The data include the structural limits of the filters, their resistance at high flow rates, and the effects of filter design features and tornado parameters.
Considering all the filters tested, th e mean break pressure or structural limit was Found to be 2.35 psi (16.2 kPa).
The maximum valve was 2. 87 psi (19.8 kPa), and the low value found was 1.31 psi (9.0 kPa).
The type of failure was usually a medium break of the downstream filter fold.
The type of filters that we evaluated were nuclear grade with design flow rates of 1000 cfm ( 0. 47. m(3)/s), standard separa tors, and folded medium design.
The parameters evaluated that are characteristic of the filter included manufacturer, separator type, faceguards, pack tightness, and aerosol loading.
Manuf actur er and medium prop erties were found to have a large effect on the structural limits.
The tests results are independent of tornado type.
The parameters we examined that are characteristic of tornados are pressurization rate and flow dura tion.
These two parameters did not have a major effect on the break pressures.
NUREG/CR-2567: FINAL DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-432. BR ADLEY, E. R. ; CUNNINGHAM,M.E.; LANNING.D.D.
Battelle Memorial Institute, Pacific Northwest Laboratory.
June 1982.
82pp.
8207060339.
PNL-4240.
13744:191.
This rep ort presents the in-reac tor da ta collec ted during the irradiation of the six-rod instrumented fuel assembly (IFA)-432 in the Halden (Norway) Boiling Water Reactor ( HBWR ) from June 1980 through June 1981.
This assembly (designed by PNL) was one of a series of NHC sponsored tests to obtain data for the deve lopment and assessment of steady-state fuel performance computer codes.
IFA-432 operated from December 1975 until June 1981, when it was removed from the r eoc tor.
Burnup levels in excess of 30,000 MWD /MTM were achieved.
Data collected prior to June 1980 were reported in NUREG/CR-0560 and NUREC/CR-1950.
Fuel centerline temperatures, cladding elongations, internal fuel rod pressures and local powers were monitored during the irra d ia ti on.
Detailed analysis of the data reported is not made.
NUREC/CR-2569: RESPONSE OF THE ZION & INDIAN POINT CONTAINMENT DUILDINGS TO SEVERE ACCIDENT PRESSURES. BUTLER,T.A.; FUGELSO,L.E.
Los Alamos Sc ientific Lab oratory.
May 1982.
41pp.
8206170051.
LA-9301-MS.
13555:240.
60
l l
The failure modes and associated failure pressures for two common generic types of pressuri zed wa ter reactor (PWR) containments are predicted.
One building type is a lightly reinforced, post-tension structure rep resented by the Zion nuc lear reactor containme.t.
The other is the normally reinforced Indian Point Containment.
Two-dimensional models of the buildings developed using the finite element method are used to predict the failure modes and failure pressures.
A three-dimensional finite model is used to evaluate the Zion building 's equipment hatch penetration.
Predicted failure modes for both containments involve loss of s truc tural integrity at the intersection of the cylindrical sidewall with the base slab.
The response of the Indian Point building to postulated detonation of a hydrogen-air mixture in the containment dome is also calculated.
NUREG/CR-2570: EXPERIMENTAL INVESTIGATION OF TEARING INSTABILITY PHENOMENA FOR STRUCTURAL MATERIALS. VASSILAROS,M.G.s GUDAS,J.P.s JOYCE,J.A.
David W.
Taylor Naval Research & Development Center.
April 1982.
40pp.
8205120122.
13053:263.
The objective o* this investigation wa s to extend the range of tearing instability validation experiments utilizing the compact specimen to include high toughness alloys.
J-Integral tests of ASTM A106s ASTM A516: Grade 70s ASTM A5334 HY-80s and HY-130 steels were performed in a variable compliant screw-driven test machine.
- Hesults, were analyzed with respect to the materials J(I)-R curves and various models of T(applied) for the compact s p ec im e n.
Tearing instability theory was validated for these high toughness materials.
For the cases of highly curved J(I)-R curves, it was shown that the actual value of T(material) at the point of instability should be employed ra ther than the average T(material) value.
The T(applied) analysis of Paris coworkers app lied to the c ompac t spec imen a ppears to be nonconservative in predicting the point of instabilitys whereas, the T(applied) analysis of Ernst and coworkers appears to be accurate, but requires precision beyond that displayed in th is pro gram.
The generalized Paris analysis applied to the compact specimen and evaluated at max imum load was most consistent in predicting instability.
NUREG/CR-2581: SOME EFFECTS OF ELECTRONS SLOWING DOWN IN MATERI ALS WITH APPLICATION TO SAFETY-RELATED EQUIPMENT GUALIFICATION. BUCKALEW,W.H.s WYANT,F.J.
Sandia Laboratories.
April 1982.
54pp.
8205120131.
SAND 82-0449.
13051:280.
Theoretical predictions have been made of the bremsstrah lung environments resulting from the slowing down of electrons in selected materials.
Several materials, material thicknesses, and electron energies were considered.
Parameters, of particular interest, obtained were transmitted photon energy and sp ec tra.
These data provide a means for estimating the effects of b eta-emi tting isotopes, released during a reactor loss of coolant accident (LOCA) or other accident scenario,on systems and components shielded by an enclosure or nousing.
NUREC/CR-2582: RADIATION CAPABILITIES OF THE SANDIA HIGH INTENSITY ADJUSTABLE COBALT ARRAY. BUCKALEW,W.H.s THOME.F.V.
Sandia Laboratories.
June 1982.
62pp.
8206160054.
SAND 81-2655.
13538:042.
The High Intensity Adjustable Cobalt Array radiation facility has been characterized for several source strengths and geometries using a three-dimensional array of self-biasing pho todiodes interfaced with automated data acquisition, reduction, and display equipment.
Maximum 61
dose rate ach ievable in a 24-in.-long a 22-in.-diameter volume is about 1. 5 Mr d /h.
Other source configurations can be selected also, e.g.,
fields 48 in.
long x 22 in, diameter produce dose rates on the order of 0.8 Mrd/h.
Even higher dose rates can be obtained by reducing the radiation volume.
NURdQ/CR-2584: METEROLOGICAL CONSIDERATIONS IN THE DEVELOPMENT OF A REAL-TIME ATMOSPHERIC DISPERSION MODEL FOR REACTOR EFFLUENT EXPOSURE PATHWAY. VAN DER HOVEN Commerce. Dept. of, National Oceanographic &
Atmospheric Administration.
May 1982.
20pp.
8206220029.
13583:329.
Meteorological considerations, as part of an overall emergency plan in tne event of an inadvertent atmosph eric release of radioactive j
i effluents from a nuclear reactor, are discussed in terms of the site meteorologica l measurement capability, the atmospheric transp ort and diffusion prediction requirements, the source term configuration, and i
the requirements posed by special site characteristics such a s coastal, valley, and mountainous locations.
NUREG/CR-2586: A SURVEY OF METHODS FOR IMPROVING OPERATOR ACCEPTANCE OF COMPUTERIZED AIDS. FREY,P.R.s KISNER,R.A.
Oak Ridge National Laboratory.
April 1982.
30pp.
8205130260.
ORNL/TM-8236.
13089:009.
The purp ose of this report is to draw from the literature factors related to user acceptance of computerized equipment that may also be applicable to the acceptance of computerized aids used in the nuclear power plant c ontrol room.
A review of the available literature i
revealed about seventy papers that deal with acceptance problems in computerized systems.
Two attempts to define and measure the characteristics of a user-acceptable system in nonnuclear industries j
form a basis for future work on this subjec t in the nuclear industry.
Operator acceptance of computerized aids can be influenced during design, operator training and system operation.
Design methods for improving acceptance include allowing the user to participate in the design process, considering acc eptanc e principles in the allocation of functions between the man and machine, minimizing the length and variation of the system response times, tailoring the dialogue to the task and use, integrating the system into the control room and providing usable system documentation.
During operator training, acceptance considerations include providing adequate detail on the purposes and limitations of the system, ensuring that the training siteations approximate the expected operational situations and i
providing training for subsequent generations of operators.
lhe primary accep tance considerations during op eration are system availability and system calibration.
1 NUREG/CR-2587: FUNCTIONS AND OPER ATIONS OF NUCLEAR POWER PLANT CREWS.
KISNER,R.A.J FREY,P.R.
Oak Ridge National Laboratory.
Mig 1982.
j 91pp.
8206100067.
ORNL/TM-8237.
13472:301.
J This rep ort summarizes the results of work performed to define the j
1 functions, op erations, and organization of nuclear power plant operating crews.
The primary information sources used were ANS and IEEE standards, normal and emergency operat ing procedures from nuclear i
power plants, interviews, and literature reviews.
The function and organization of operating crews for several plants are discussed generically.
The report covers a wide spec trum of topics including review of standards affecting human factors in the control room, 62
influences of automation on operator functions, classification of operator func tions, function of operator at onset of emergency, crew organization, work-induced stress, and operator acceptance of his role.
NUREG/CR-2588: SECURITY OFFICER RESPONSE STRATEGIES (SECURORS).
ROUNTREE,S.L.K.
Sandia Laboratories.
June 1982.
46pp.
8206240041.
SANDB2-0410.
13609:299.
The Security Officer Response Strategies (SECURORS) approach provides a method for deploying security officer.2 within a nuclear power plant subsequent to an adversary intr usion detec tion.
Under current nuclear power plant operating condi tions, the number of vital areas generally exceeds the number of security officers.
The SECURORS method allocates the available officers on the basis of numerical weights and ranking for each of the nuclear power plant vital areas and barriers.
It is assumed that the numerical weights have been obtained previously from readily available techniques or from expert opinion on th e vulnerability of vital areas and barriers.
This paper does not establish any methodology for the derivation of weights, but vital area characteristics related to the numerical weights and ranking are reviewed.
An e xamp le illustrates the integer programming problem formulation and solution process for several deployment strategies.
The SECURORS approach builds on the results of several procedures and analytic techniques.
It is assumed that the nuclear power plant has undergone a vital area analysis.
Additional results can be obtained from the Safeguards Automated Facility Evaluation (SAFE) method.
A glossary is provided to c larif y safeguards and mathematical programming terminology.
NUREG/CR-2589: A GROUND-PENETRATING RADAR SURVEY OF THE MAXEY FLATS LOW-LEVEL NUCLEAR WASTE DISPOSAL SITE, FLEMING COUNTY, KENTUCKY.
HORTON,K.A.
Geo-Centers, Inc.
June 1982.
48pp.
8207190065.
GC-TR-82-171.
13920:058.
A ground penetrating radar survey wa s conducted at the Maxey Flats Lcw-Level Nuclear Waste Disposal Site, Kentucky, to more accurately determine the location of burial trenches and pits, and to icentify loca tions and depths of any prominent subsurface f ea tures.
A geologic / electromagnetic model of the site was develop ed and utilized for analysis of the acquired data.
Depths of penetration derived from radar records correlated well with those calculated from the model.
A final interpretation of the radar data is presented.
NUREG/CR-2591: ESTIMATING THE POTENTIAL INDUSTRIAL IMPACTS OF A NUCLEAR REACTOR ACCIDENT. CARTWRIGHT,J.V.s BEEMILLER,R.M.; TROTT,E.A.s et al.
Commerce Dep t. of.
April 1982.
136pp.
8205190035.
13185:047.
This NUREG describes an industrial impact model that can be used to estimate the regional indus try-sp e c i fi c impacts of disasters, both natural and manmade.
Special attention is given to the impac ts of possible nuclear reactor accidents.
The report also presents three applications of the model.
The impacts estimated in the case studies are based on (1) general information and reactor-specific data, supplied by the U.S. Nuclear Regulatory Commission (NRC); regional economic models derived from the Regional Input-Output Modeling System (RIMS II) developed at the Bureau of Economic Analysis (BEA)s and (3) additional me thodology developed especially for taking into account the unique charac teristics of a nuclear reactor accident with respect to regional industrial activity.
63
NUREG/CR-2593: A USER 'S MANUAL FOR COMPUTER CODE RIBD/IRT. THAYER,D.D.;
LURIE,N.A.
Sandia Laboratories.
April 1982.
76pp.
8205200260.
SAND 82-7013.
13195:291.
The comp uter code RIBD/IRT is a modified s srsion of RIBD-II.
It is a grid processor that calculates isotopic concentrations resulting from two fission sources with normal down-c hain d ecay by beta emission and isomeric transfers and inter-chain coup ling resulting from n gamma reactions.
Calculations can be made to follow an irradiation history through an unlimited number of step changes of unrestricted duration and variability including shutdown periods, restarts at different power levels and/or any other level changes.
Output information includes time-dependent inventories, activities, decay powers, and energy releases for as many as 800 fission products.
Modifications to RIBD-II were necessitated by Loss-of-Coolar.t Accident (LOCA) studies conducted by IRT Corporation regarding fission product source term definition.
These modifications permit the user to track and modify the concentra tions of individual elements as they decay with time following reactor shutdown.
In essence, one can determine time-dependent fission product source terms resulting from any reactor operating history which then can be used as input into fission produc t transport cod es.
Other modifications to RIBD-II expanded the output information to assist the user in analyring the source term.
This manual describes the modifications to R IBD/ II, input requirements and a sample prob lem.
The ap pendic ies give a listing of RIBD/IRT, sample output, and a listing of a code called ZIP which prepares the library tape for input to RIDD/IRT.
The code is available in a UNIVAC 1100/81 version and a VAX 11/780 version.
NUREG/CR-2594: A USER 'S MANUAL FOR THE GABAS SPECTRUM COMPUTER CODE.
THAYER,D.D.s LURIE,N.A.
Sandia Lab ora tor ie s.
April 1982.
55pp.
8205200270.
SAND 82-7014.
13203:117.
The Gamma and Beta Spectrum computer c ode (GABAS) was developed at IRT Corporation for calculating time-dependent beta and/or gamma spectra from decaying fission products.
GADAS calculates composite fission product spectra based on the technique used by England et al.,
in conjunction with t h e C INDER family of fi ssion product codes.
Multigroup beta and g a mma spectra for individual nuclides are folded with their corresponding time-depend ent activities (usually generated by a fission product inventory code) to produce a composite time-dependent sission product spectrum.
This manual contains the methodology employed by CABAS, input r e qu ir emen t s for proper execution, a sample prob lem and a FORTRAN listing compatible with a UNIVAC machine The code is available in a UNIVAC 1100/81 version and a VAX 11/700 version.
The former may be obtained from the Radiation Shielding Inf ormation Center (RSIC); the latter may be obtained directly from IRT Corporation.
NUREG/CR-2597: STEADY-STATE PRESSURE LOSSES FOR MULTIROD BURST TEST (MRBT) BUNDLE B-5.
BAILEY,P.T.
Babcock & Wilcox Co.
May 1982.
107pp.
8206100053.
ORNL/SUD/80-404.
13472:004.
This rep ort describes the water-flow-test of 64-rod PWR fuel assembly simulation which was tested under loss-of-coolant-accident (LOCA) conditions.
The test, involving cladding deformation and rupture in th e temperature region of the Zircaloy alpha phase, was performed on May 30, 1980. The average of butst temperatures and pressure differentials were 773 degrees and 8,806 kPa.
1.
R.
H.
Chapman et al., Guick-look R ep or t on MRBT B-5 (8 x 8) 64
i Bundle Test, Internal Report ORNL/MRBT-5 (July 1980).
2.
R.
H.
Ch apman et al., Multirod Burst Test Program Prog. Rep.
January-June 1980, NUREG/CR-1883 (ORNL/NUREG/TM-426).
3.
A.
W.
Longest, Multirod Burst Test Program Prog. Rep.
January-June 1981, NUREG/CR-2366, Vol.
1, ORNL/TM-8058.
4.
J.
L.
Crowley, Multirod Burst Test Program Prog. Rep.
July-December 1981. NUREG/CR-2366, Vol. 2 ORNL/TM-8190.
This rep ort describes the work characterizing the hydraulic resistance of the B-5 bundle.
In addition to the flow test o f the deformed bundle, B&W assembled and flow tested an undeformed reference bundle (desig nated as B-5R) to provide comp arative data.
Magnetic tapes containing the raw test data, reduced test data, and calibration records of th e B&W flow tests are on file a t DRNL.
NUREG/CR-2600: END-OF-IRRADI ATION DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY ( IFA )-527. CUNNINGHAM,M.E.s LANNING,D.D.
Battelle Memorial Institute. Pa cific Northwest Laboratory.
May 1982.
116pp.
8205200293.
PNL-4201.
13203:001.
This rep ort presents data obtained during the irradiation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway.
This assembly is the last in a series of U.S.
Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance c omputer codes.
IFA-527 contains five identical rods with high-density stable fuel pellets and 230-um diametral gaps and one rod with similar fuel pellets but with a 60-um diametral gap.
All six rods were xenon-filled to simulate the eff ects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures.
The assembly operated successfully from July 1,
1980,to August 15, 1980s the reactor was then shut down until September 10, 1980.
During the shutdown, at least four o f the six rods suffered
[
pressure boundary failures.
Irradiation of the assembly continued with the failed rods from September 10, 1980, until April 8, 1981; the assembly was then removed from the reactor.
This report presents both pre-and postfailure data for IFA-527.
I NUREG/CR-2603: BUB 9LE BEHAVIOR IN LMFBR CURE DISRUPTIVE ACCIDENTS.
REYNOLDS A.B.
ERDMAN,C.A.3 BRADLEY,D.R.; et al.
Virginia, Univ. o f.
April 1982.
90pp.
8205040021.
12971:008.
Research performed at the University of Virginia during FY '81 for the Advanced Reactor Safety Research Division of the U.S.
Nuclear i
Regulatory Commission is reported.
The res earch is part of the LMFBR j
Aerosol Release and Transport Program.
Principal areas investigated were (1) analysis of ORNL FAST underwater tests, (2) pretest parametric analysis of ORNL under sodium tests, (3) axial motion of larg e expanding and collapsing bubbles, and (4) measurement of droplet sizes from flashing.
Analysis of the FAST tests with the UVABUBL c ode showed i
the strong influence of water vapor during the bubble expansions water vapor rapidly replaces UO(2) vapor as the vapor that drives the bubble.
In the case of the under sodium tests, it is expected that entrained sodium will vaporize and influence bubble behavior, but, unlike water, sodium will not be vaporized from the b ub b l e surface.
Earlier analyses of axial motion of the French EXCOBULLE exp eriments were impr oved.
Experimental methods in the experiment on droplet sizes from flashing I
were developed further.
06
NUREC/CR-2604: THE SNAP OPERATING SYSTEM (SOS) USER'S GUIDE.
SABUDA,J.D.s WALKER,J.L.4 POLITO,J.s et al.
Sandia Laboratories.
May 1982.
438pp.
8205190010.
SAND 82-7018.
13185:183.
The INAP Operating System (SOS) is a FORTRAN 77 program which provides ssistance to the saf eguards analy st who uses the Sa feguards Automated Facility Evaluation (SAFE) and the Safeguards Network Analysis Proc edure (SNAP) techniques.
Features offered by SOS are a data base sys tem f or storing a library of SNAP applications, computer graphics representation of SNAP models, a computer graphics editor to develop and modify SNAP models, a SAFE-to-SNAP interface, automatic generation of SNAP input data, and a computer graphics post processor for SNAP.
Th e SOS User 's Guide is designed to provide the user with the information necessary to use SOS effectively.
Examples are used throughout to illustrate the concepts.
The format of the user's guide follows the same sequence as would be used in executing an ac tual application.
NUREG/CR-2605: THE SNAP OPERATING SYSTEM REFERENCE MANUAL. SABUDA,J.D.;
POLITO,J.4 WALKER,J.L.
Sandia Laboratories.
May 1982.
268pp.
8206100058.
SAND 82-7019.
13467:137.
The SNAP Operating System (SOS) is a FORTRAN 77 program which provides assistance to the safeguards analyst who uses the Sa feguards Automated Fac ility Evaluation (SAFE) and the Safeguards Network Analysis Proc edure (SNAP) techniques.
Features of f ered by SOS are a data base sys tem for storing a library of SNAP applications, computer graphics representation of SNAP models, a computer graphics editor to develop and modify SNAP models, a SAFE-to-SNAP interface, automatic generation of SNAP input data, and a computer graphics post processor for SNAP.
The SOS Ref erence Manual provide s detailed applica tion information concerning SOS as well as a detailed discussion o f all GUS components and their associated command inp ut f ormats.
NUREG/CR-2610: RAGBEEF: A FORTRAN IV IMPLEMENTATION OF A TIME-DEPENDENT MODEL FOR RADIONUCLIDE CONTAMINATION OF BEEF. PLEASANT, J. C. s MCDOWELL-BOYERs KILLOUGH,G.G.
Oak Ridge Na tional Laboratory.
June 1982.
145PP.
8207190018.
ORNL/TM-8011.
13922:278.
RAGBEEF is a FORTRAN IV program that calculates radionuc lide concentrations in beef as a result of inges tion of contaminated feeds, pasture, and pasture soil by beef cattle.
The model implemented by RAGBEEF is dynamic in nature, allowing the user to consider age-and season-depend ent asp ec ts of beef cattle management in estimating i
(
concentrations in beef.
It serves as an auxiliary code to RAGTIME, previously documented by the authors, which calculates radionuclide concentrations in agricultural crops in a dynamic manner, but evaluates concentration in beef for steady-state conditions only.
1he t ime-d ep end en t concentrations in feeds, pasture, and pasture soil generated by RAGTIME are used as input to the RAGBEEF code.
- RAGBEEF, as presently implemented, calculates radionuclide concentrations in the muscle of age-based cohorts in a beef cattle herd.
Concentrations in the milk of lactating cows are also calculated, but are assumed age-independent as in RAGTIME.
This report describes the age-and season-dependent considerations making up the RAGBEEF model, as well as presenting the equations which describe the model and a documentation of the associated computer code.
Listing of th e RAGBEEE and updated R AGTIME codes are provided in appendices, as are the results of a sample run of RAG EEF and a description of recent modifications to RAGTIME.
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NUREG/CR-2611: MGO AND 70 WX UO2-30W% Y203: THERMOPHYSICAL AND TRANSIEN1 PROPERTIES. PILCH,M.
Sandia Laboratories.
April 1982.
28pp.
8205060079.
SAND 81-1230.
13007:286.
Interactions between a molten core simulant and MgO bric ks (Harklase) are being studied.
A molten core simulant, consisting of 70 w% U0(2) and 30 w% Y(2)O(3), has been proposed for use in large scale experiments (200 kg) at Sandia's Large Melt Facility.
This r eport documents the binary phase diagrams, thermophysical properties, and transport properties which are necessary for the analysis of these experiments.
NUREG/CR-2612: VARIABILITY IN DOSE ESTIMATES ASSOCIATED WITH THE FOOD CHAIN TRANSPORT AND INGESTION OF SELECTED R ADIONUCLIDES.
HOFFMAN.F.O.s GARDNER.R.H.. ECKERMAN,K.F.
Oak Ridge National Laboratory.
June 1982.
50pp.
8206240059.
ORNL/TM-8099.
13608:290.
Dose predictions for the ingestion of 90Sr and 137Cs, using aquatic and terrestrial f ood chain transpor t models similar to those in the Nuclear Regulatory Commission's Regulatory Guide 1.109, are evaluated through estimating the variability of model parameters and determining the effect of this variability on model output.
The variability in the predicted dose equivalent is determined using analytical and numerical procedures.
In ad dition, a detailed discussion is included on 90Sr dosimetry.
The overall estima tes of uncertainty are most relevant to conditions where site-specific data is unavailable and when model structure and parameter estimates are unbiased.
NUREG/CR-2618: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A NATURAL CIRCULATION TEST S-NC-7C. LARSON, R. A.
EG&G, Inc.
April 1982.
4Bpp.
8205130237.
EGG-2179.
13075:312.
This rep ort presents test data recorded for Test S-NC-7C of the Semiscale Mod-2A Natural Circulation Test Series.
This is one of several Semiscale tests that investigate th e thermal-hydraulic phenomena resulting from operational transients or small-break loss-of-coolant accidents (LOCAs) involving loss of mechanical primary coolant circulatien in a pressurized water reactor.
These tests produce experimental data to develop and assess the analytical capab lity of computer models used to predict the results of such small-break LOCAs and operational transients.
The primary objectives of Test S-NC-7C were to experimentally characterize the relationship of natural circulation flow to primary system inventory, and to examine the influence on system behavior of imbalancing the secondary side of one loop.
This rep ort presents the uninterpreted data from Test S-NC-7C for analysis.
Th e data, presented as graphs in engineering units, have been analyred only to the extent necessary to ensure that they are reasonable and consistent.
NUREG/CR-2622: ANALYSIS OF TRAC AND SCTF RESULTS FOR SYSTEM PRESSURE-EFFECTS TESTS UNDER FORCED FLOODING (RUNS 506,507 AND 508).
SUDO,Y.
Los Alamos Scientific Laboratory.
May 1982.
64pp.
8205180117.
LA-9258-MS.
13134:273.
The Transient Reactor Analysis Code (TRAC) and the Slab Core Test Facility (SCTF) results are compared for the three system pressure-effects tests (Runs 506, 507, and 508) with forced injection into the lower plenum.
The results show that TRAC can predic t well the 67 N
overall trans ients of core rod temperature, core differential pressure, and the liquid carryover into the hot leg, as well as in the upper plenum, effec ts that are strongly dependent on the system pressure.
Comparisons are also presented that show ma jor dif f erences be tween the SCTF test and the TRAC results that should be improved in the future.
NUREQ/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FRDM AGRO-ECOSYSTEMS. Annual Progress Rep or t,0c to b er 1980-September 1981.
SMITH,M.H.a ALBERTS,J.J.; ADR I ANO, D. C. ; et al.
Savannah River Laboratory.
April 1982.
50pp.
B205110077.
13038:245.
The research has as its ob Jective describing the fate and behavior in the environment of radionuclides f rom nuclear f uel reproce ssing.
Greenhouse radionuclide up take studies which examined factors possibly altering phytoavailability of radionuclides show only slight dif f erences among crop species or soil tr ea tment s (lime or lime plus chelate) in Pu or Cm upta ke.
The temporal effect on Pu and Cm uptake, from a partial data set, is inc onclus ive, with variable effects from crop species, radionuclides, and soil treatments.
Cesium uptake shows variable resp onse with crop species generally decreasing with time.
Field grown broadleaf crop s grown have differing Pu conc entrations observed in wheat and soybean crops.
In crops tending to trap aerially deposited Pu, washing removed more than 50% of the Pu.
Uranium contamination of a wheat crop grown near the separations facility appears to be strongly affected by root uptake.
This is in contrast to the behavior of Pu where superficial pathways are the dominant mode s of contamination, and is probably treated to (1) the ubiquitous pr esence of naturally ocurring U isotopes and (2) a greater concentration ratio for U than for Pu.
NUREG/CR-2629: INTERIM SOUTCE TERM ASSUMPTIONS FOR EMERGENCY PLANNING AND EGUIPMENT GUALIFICATION. NIEMCZYK,S.J.
Oak Ridge National Laboratory.
June 1982.
141pp.
8206290519.
ORNL/TM-8274.
13646:160.
The source terms recommended in the current regulatory guidance for considerations of light water reactor (LWR) accidents were developed a number of years ago when unders tanding of many of the phenomena pertinent to source term estimation was not well developed.
The purpose o f the work presented here was to review the literature on accident source term research and utilize the recent research to develop more realistic assumptions for calculation of accident source terms which could be used for regulatory purposes for two specific cons id eration s, namely, equipment qualifica tion and emergency planning.
The emphasis of this work was on developing appropriate assumptions for estimating the magnitude of the radionuclide releases for various g roups of accidents in each of the accident spectra of concern.
The overall approach taken was to adopt basic essumptions and models previously proposed for various aspects of source term estimation an d to modify those assumptions and models to reflect recently gained insights into, and data describing the release and transport of radionuclides during and after light water reactor accidents.
The report presents results of sample calculations of accident source terms and compares the results with the other published results.
The report also presents peer review comments on this study.
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NUREG/CR-2632: RESPONSE OF CENTRIFUGAL DLOWERS TO SIMULATED TORNADO TRANSIENTS. July-September 1981. IDAR, E. S. s MARTIN,R.A.s CREGORY,W.S.s et al.
Los Alamos Scientific Laboratory.
May 1982.
22pp.
8206100062.
LA-9276-SR.
13467:116.
During this quarter, quasi-steady and dynamic testing of the 24-in. centrifugal blower was completed using the blowdown facility located at New Mexico State University.
Th e data were ob tained using a new digital d ata-ac quisi t ion sy s tem.
Software was developed at the Los Alamos National Laboratory to reduce the dynamic test data and create computer-generated movies showing the dynamic performance of the blower under simulated tornado transient pressure conditions relative to its quasi-steady state performance.
Currently, quadrant-four (ootrunning flow) data have been reduced for the most severe and a less severe tornado pressure transient.
The results indicate that both the quasi-steady and dynamic blower performance are very similar.
Some hysteresis in the dynamic performance occurs because of rotational in ertia effects in the blower rotor and drive system.
Currently quadrant-two (backflow) data are being transferred to the LTSS c omputer system at Los Alamos and will be reduced shortly.
I NUREG/CR-2633: CONTAINMENT REACTOR CAVITY SUBCOMPARTMENT ANALYSIS PROCEDURES FOR A BOILING WATER REACTOR. TUR K, W. V. s GIDO,R.G.s LI,C.Y.
Los Alamos Scientific Laboratory.
May 1982.
32pp.
8206110015.
LA-9277-MS.
13492: 163.
Procedures for the performance of Boiling Water Reactor (BWR) cavity subcompartment analysis are presented.
The purpose of this presentation is to normalize th e analysis procedures and to provide a standard approach for such analyses.
As a result, differences in the manner of per f orming subc ompartment analyse s can be minimized and more readily understood and evaluated by others.
The procedures were developed within the constraint of current code capability for the performance of such analyses and the current US Nuclear Regulatory Commission guidelines.
A wide range of the effects of input and modeling variations on calculated sacrificial shield-wall (SSW) forces and moments were studied.
The studies were for a representative DWR cavity geometry with the pipe break inside the SSW.
The COMP ARE subcompartment analysis code was used for the studies.
NUREG/CR-2636: EXPERIMENTAL DATA REPORT FOR AIR-WATER FLOODING TESTS OF THE FLECHT-SEASET PROGRAM SET FACILITY VESSEL UPPER PLENUM.
ANDERSON,J.L.3 FOGDALL,S.P.
EG&G, Inc.
June 1982.
99pp.
8206240066.
EGG-2183.
13609:007.
A test facility to investgate the flooding characteristics of the FLECHT-SEASET Program's SET Fac ility vessel upper plenum has been developed and installed in the Steam-Air-Wa ter Test Facility of the Idaho Nationa l Engineering Laboratory.
A series of c o un t er c urr en t-f l ow-l imi t ed tests were peformed in the test facility using air-water at low pressure and room temperature.
This report documents the experimental system and the testing program, presents tabulations of the data, develops the e xp er imental uncertainty analysis and discusses the results of the testing.
NUREG/CR-2637: EMERGENCY RESPONSE CAPABILITIES AND EXAMPLE ASSESSMENTS FOR AIRBORNE RADIONUCLIDE DISCHARGES. START. G. E. s CATE.J.H.s ACKERMANN.G.R.s et al.
Commerce, Dept. of, National Oceanographic &
Atmospheric Administration.
May 1982.
32pp.
8206100036.
l 69
13474:354.
An existing emergency response capability has been developed for the Idaho National Engineering Laboratory by the National Oceanic and Atmospheric Admnistration Air Resources Lab oratories Field Re search Offices.
The system consists of several existing computers and associated data collection facilities.
This equipment has been coordinated into a useful capability providing initial and ongoing analysis of meteorological and radiological information.
In the event of a radiological emergency, this inf ormation may be used to assist action plan formulations and decisions for the area in and around the Snake River Plain in Southeast Idaho.
NUREG/CR-2638: SNOW LOADS FOR THE DESIGN OF NUCLEAR POWER PLANT STRUCTURES. ELLINGWOOD.B.; HARRIS,J.R.
Commerce Dept.of, National Bureau of Standards.
April 1982.
51pp.
8204160038.
12714: 015.
This rep ort describes a research program to characterize snow loads on roofs of nuclear power plant struc tures and to develop recommendations for operating basis and extreme environmental loads.
Snow surveys were conducted to gather field data about the distribution of snow on plant roofs and to correlate roof and ground snow loads.
The survey data were integrated with data from similar studies to provide recommendations for structural design.
Load conbinations involving rain and snow were analyzed probabilistically to provide a basis of comparison with other design basis environmental loads.
NUREG/CR-2639: HISTORICAL EXTREME WINDS FOR THE UNITED STATES-ATLANTIC AND GULF OF MEXICO COASTLINES. CHANGERY,M.J.
Commerce, Dept. o f.
National Oceanographic & Atmospheric Administration.
May 1982.
156pp.
8206140312.
13508:239.
Annual fastest mile wind data were extracted for the complete period of record for 53 locations along the Atlantic and Gulf of Mexico coastlines.
Existing models were used to standardize the data to 10 meters for airport-type exposures and meters for city exposures.
Selected probability estimates were developed from applicaton of t h e-Fisher-Tippet Type I extreme value mode for non-tropical storms and the Weibil model for tropical storms.
A mixed distribution was used for locations with a significant percentage of annual extremes caused by tropical storms.
NUREC/CR-2642: LONG-TERM SURVIVABILITY OF RIPRAP FOR ARMORING URANIUM MILL TAILINGS AND COVERS: A LITERATURE REVIEW. LINDSEY,C.G.J LONG, L. W. s DEGEJ,C.W.
Battelle Memorial Institute, Pacific Northwest Laboratory.
June 1982.
140pp.
8207140120.
PNL-4225.
13845:347.
Pacific Northwest Laboratory (PNL) is investigating the use of a rock
(
armoring b1cnket (riprap) to mitigate wind and water erosion of an earthen radon suppression cover applied to uranium mill tailings.
Because the radon suppression cover and the tailings must remain intact for up to 1000 years or longer, the riprap must withstand natural weathering forces.
This report is a review of information on rock weathering and riprap durability.
Chemical.and physical weathering i
processes, rock characteristics related to durability, climatic I
conditions affecting the degree and rate of weathering, and testing procedures used to measure weathering susceptibilities have been revised.
Sampling and testing techniques, as well as analysis of physical and chemical weathering susceptibilities, are necessary to evaluate rock durability.
Many potential riprap materials may not be able to survive 1000 years of weathering.
Available techniques for 70
durability testing cannot adequately predict rock durability for the 1000 year per iod because they do not consider the issue of time (i.e.,
how long must riprap remain stable).
This report includes an Appendix, which discusses rock weathering, written by Dr. Richard Jahns of Stanford University.
NUREC/CR-2644: AN ASSESSMENT OF OFFSITE. REAL-TIME DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS. MAECK,M.J.s HOFFMAN,L.G.s STAPLES, B. A. J et al.
Exzon Nuclear Co.,
Inc. (subs. of Exxon Corp.).
April 1982.
67pp.
8205040015.
ENICO-1110.
12970:289.
An evaluation is made of the effectiveness of fixed, real-time monitoring systems around nuclear power sta tions in determining the magnitude of unmonitored releases.
The ef f ects of meteorolog ical conditions on the accuracy with which the magnitude of unmoni tored releases is d etermined and the uncertainties inherent in defining these meterological conditions are discussed.
The number and placement or fixed field detectors in a system is discussed, and the data processing equipment required to convert field detector output data into release rate information is described.
Cost data relative to the purchase and installation of specific systems are given, as well as the characteristics and information return for a system purchased at an arbitrary cost.
NUREQ/CR-2647: CRITICAL HEAT FLUX EXPERIMENTS UNDER LOW FLOW CONDITIONS IN A VERTICAL ANNULUS. MISHIMA,K.; ISHII,M.
Argonne National Laboratory.
April 1982.
43pp.
8204290604.
ANL-82-6.
1289 5: 23b.
An experimental study was performed on critical heat flux (CHF) at low flow conditions for low pressure steam-water upward flow in an annulus.
The test section was transparent, therefore, visual observations of dryout as well as varicus instrumentations were made.
The data indicated that a prema ture CHF occ urred due to flow regime transition from churn-turbulent to annular flow.
It is shown that the cri'ical heat flux observed in the experiment is essentially similar to a flooding-limited burnout and the critical heat flux can be well reproduced by a nondimensional correlation derived from the previously obtained criterion for flow regime transition.
The observed CHF values are much smaller than the standard high quality CHF criteria at low flow, corresponding to the annular flow film dryout.
This result is very significant, because the coolability of a heater surface at low flow rates can be drastically reduced by the occurrence of this mode of CHF.
I NUREC/CR-2648: EXPERIMENTAL DATA REPORT FOR SEMISCALE MOD-2A NATURAL CIRCULATION TEST SERIES (TESTS S-NC-8B AND S-NC-9). SACKETT,K.E.s CLEGG,L.B.
EQ&Q. Inc.
April 1982.
56pp.
8205130243.
EGO-2184.
13090:028.
This rep ort presents test data recorded for Tests S-NC-8B and S-NC-9 of the Semiscale Mod-2A Natural Circulation Test Series.
These tests are part of a series of Semiscale tes ts that investigate the thermal-hydraulic phenomena resulting from operational transients involving loss of mechanical primary coolant circulation in a pressurized water reactor.
The primary objective of Tests S-NC-8D and l
S-NC-9 was to experimentally characterize the thermal-hydraulic behavior of a system during single phase, and reflux natural circulation c onditions experienced in the course of an integral small break with and without the presence of emergency core cooling water.
Of special interest were the effects of single phase natural 71 I
l circulation flow caused by changes in core power, primary pressure, and external heater' power.
This rep ort presents the uninterpretated data from Tests S-NC-8B and S-NC-9 for future analysis.
The data, presented as graph s in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent.
NUREG/CR-2651: ACCIDENT GENERATED PARTICULATE MATERI ALS AND THEIR CHARACTERISITICS--A REVIEW OF B ACKGROUND INFORMATION. SUTTER,S.L.
Battelle Memorial Institute, Pacific Northwest Laboratory.
May 1982.
97pp.
8206170387.
PNL-4154.
13557:195.
Saf ety a ssessments and environmental impact statements for nuclear fuel cycle facilities require an estimate of the amount of radioactive particulate material initially airborne (source term) during accidents.
Pacific Northwest Laboratory (PNL) has surveyed the literature, gathering information on the amount and siz e of these particles that has been developed from limited experimenta l work, measurements made from operational accidents, and known aerosol behavior.
Information useful for calculating both liquid and powder source terms is compiled in this repor t.
Potential aerosol generating events discussed are spills, resuspension, aerodynamic entrainment, explosions and pressurized releases, communition, and airb orne chemical reac tions.
A discussion of liquid behavior in sprays, sparging, evaporation, and condensation as applied to accident situations is also discus sed.
i NUREQ/CR-2652: EVALUATION AND PERFORMANCE OF CLOSED-CIRCUIT BREATHING APPARATUS. HACK,A.s TRUJILLO,A.s CARTER,K.s et al.
Los Alamos l
Scientific Laboratory.
June 1982.
23pp.
8206230328.
LA-9266-MS.
13594:295.
Seven closed-circuit self-contained breathing apparatus were worn by a panel of anthropometrically selected test subjects to determine the protection provided by each.
The types included those that supply breathing gas continuously, or on demand, or a combination of both.
One unit main tained a positive pressure and provided higher protection than the others.
Device performance by facial size is discussed.
NUREC/CR-2653: EARTH RESISTIVITY AS A TOOL FOR SHALLOW EXPLORATION IN THE REELFOOT LAKE AREA. TENNESSEE. STEARNS,R.G.s HASELTON,T.M.J TSAY,J.
Vand erb ilt Univ.
May 1982.
131pp.
8206100048.
13478:330.
Surface earth recistivity techniques were successfully tested at a j
shallow (10's of feet) depth in the Reelf oo t Lake area of Mis sissippi 's i
alluvial plain.
Profiling, Barnes Layer se ctions, Wenner sounding, and i
circle soundings proved useful.
]
Features of abandoned river channels (a central low resistivity clay
'p lug ' and lateral high resistivity, sandy natural levees) were 1
i readily located and mapped by profiling, and were located within 10 feet or less by circle soundings.
Approximately true resistivity columns were made by measuring the resistivity of samples from small diameter holes.
For these columns, Wenner Array soundings gave nearly correct layer thickness estimated in contrast to erroneous Schlumberger soundings.
NUREG/CR-2664: SELECTED REVIEW OF FOREIGN LICENSING PRACTICES FOR NUCLEAR POWER PLANTS. STEVENSON,J.D.s THOMAS,F.A.
Structural Mechanics Associates.
April 1982.
151pp.
8205030645.
12928:065.
i A compilation and description of curre nt U. S.
and foreign 72
_ _, _ _. ~ _
licensing and regulatory practices are given.
Also included i t, a brief description o f nuclear power plant regulatory and licensing organizations involved.
The particular countries surveyed ar e Canada, France, Japan, Sweden, th e Uni t ed Kin g dom, the United States and the Federal Repub lic of Germany.
NUREC/CR-2671: THE MARVIKEN FULL SCALE CRITIC AL FLOW TESTS. Summary Report.
- MARVIKEN.
May 1982.
26 *ap p.
8205200278.
MXC-301.
13195:014.
The Marviken Full Scale Critical Flow Tests were conducted as a multi-national project at Marviken Power Station in Sweden.
The program sough t to provide the critical mass flow data necessary to form a link between the available small scale test data and full scale pipe geometries found in operating nuclear power stations.
This rep ort summarizes the program objectives, test facility, in s trumen ta t i on, procedure, matrix, data and error limits, and significant test results.
The summary report is reprinted by USNRC under the mul ti-national agreement that allows public dissemination of the data two years after the tests.
NUREG/CR-2681: ESTIMATED RECURRENCE FREGUENCIES FOR INITIATING ACCIDENT CATEGORIES ASSOCI ATED WITH THE CLINCH RIVER BREEDER REACTOR PLANT DESIGN. COPUS,E.R.
Sandia Laboratories.
June 1982.
170pp.
8206250048.
SANDB2-0720.
13627:031.
Estimated recurrence frequencies for each of twenty-five generic LMFBR initiating accident categories were quantified using the Clinch River Breeder Reactor Plant (CRBRP) design.
These estimates were obtained using simplified systems fault trees and functional event tree models from the Accident Delineation Study Phase I Final Report coupled with order-of-magnitude estimates for the ini tia t or-d ep end ent failure probabilities of the individual CRBRP engineered safety systems.
Twelve distinct protected accident categories where SCRAM is assumed to be successful are estimated to occur at a combined rate of 10(-3) times per year while thirteen unprotected accident categories in which SCRAM f ails are estimated to occur at a combined rate on the order of 10(-5) times per year.
These estimates are thought to be representative despite the fact that human performance f ac tor s, maintenance and repair, as well as input common cause unc er ta inties, were not trea ted explicitly.
The overall results indicate that for the CRBRP design no single accident category appears to be dominant, nor can any be totally eliminated from further investigation in the areas of accident phenomenology for in-core events and post-accident phenomenology for containment.
NUREG/CR-2682: CITADEL: A COMPUTER CODE FOR THE ANALYSIS OF IODINE BEHAVIOR IN STEAM GENERATOR TUBE RUPTURE ACCIDENTS. RAGHURAM,S.s BAYBUTT,P.s DENNING,R.S.; et al.
Battelle Memorial Institute, Columbus Laboratories.
April 1982.
141pp.
8205110659.
BMI-2093.
13037:001.
The comp uter code CITADEL was written to analyze iodine behavior during steam generator tube rup ture acciden ts.
The code models the transport and deposition of iodine from its point of escape a t the steam generator primary break until its release to the environment.
This report p rovides a brief description of the code including its input requirements and the nature and form of its output.
This rep ort is in the form of a user's manual for the code.
Only a brief discussion of the processes modeled in the code is pr ovided 73
herein.
The interested reader is referred to a companion rep ort for detailed technical description of the models that have been included in the code.
NUREG/CR-2683: IODINE BEHAVIOR IN STEAM GENER ATOR TUBE RUPTURE ACCIDENTS. RAGHURAM,S.; BAYBUTT.P.; DENNING.R.S.s et al.
Battelle Memorial Institute, Colu.nb us La boratori es.
April 1982.
114pp.
8205110662.
BMI-2094.
13036:169.
This rep ort identifies the results of a program aimed at developing a computer code for use in the analysis of the behavior oF iodine during steam generator tube rupture (SGTR) accidents in pressurized water reactors (PWR 's ).
The program was directed towards the identification of the several processes that play a role in the transport and deposition behavior of iodine from its point of escape at the primary system break to its point of release to the envir onmen t, the deselopment of models to describe these processes and the incorporation of these models into a comput er code.
Preliminary calculations performed using th e computer c ode indicate that iodine contained in the water droplets that are formed as the primary coolant flashes could be a major source of iodine released to the atmosphere during an SGTR accident.
Additionally, the assumed chemical form of the iodine, molecular or ionic, appears to be extremely important in determining the consequences of the accident.
NUREC/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES. WANG. P. C. s CURRERI, M. s SHOOMAN, M. s et al.
Brookhaven National Laboratory.
May 1982.
90pp.
8206170046.
BNL-NUREG-51529.
13543:001.
This rep ort deals with the problem of combining two or more concurrent responses which were induced by dynamic loads acting on nuclear power plant structures.
Sp ec i fical ly, the acceptability of using the SRSS (square root of the sum of the squares) value of peak values as the combined response is investigated.
Emphasis is placed on the establishment of a simplified criteron that is convenient and relatively easy to use by design engineers.
NUREG/CR-2686: REVIEW OF LOAD COMBINATIONS FOR NSSS AND BOP PIPING AND EQUIPMENT OF MARK III PLANTS. P HILIPP ACOPOULOs REICH,M.; WANG,P.C.
Brook haven Na tional Laboratory.
May 1982.
2OOp p.
8206100065.
BNL-NUREG-51530.
13479:204.
This rep ort describes a review conducted by the Structural Analysis Division of Brookhaven National La boratory (BNL) for the Mechanical Engineering Branch of the Nuclear Regulatory Commission (MEB/NRC) on combinations of dynamic responses related to Nuclear Steam Supply Systems (NSSS) and Balance-of-Plant (BDP) piping and equipment components of Mark III plants.
A total of 167 combination ca ses were considered.
The response combinations revi ewed in this repor t were compiled by Structural Mechanics Associates for the General Electric Company, using time-histories and other technical data supplied by various arc h i tec t-eng ine ering firms working for the Mark III containment owners.
The objective of the (DNL) review was to verify the results presented by SMA.
NUREC/CR-2692: AN INTEGRATED SYSTEM FOR FOREC ASTING ELECTRIC ENERGY AND LOAD FOR STATES AND UTILITY SERVICE AREAS. CHERN,W.S.;
GALLAGHER,C.A.; TEPEL,R.C.; et al.
Oak Rid ge National Labora tory.
May 1982.
58pp.
8206110006.
ORNL/TM-7947.
13493:016.
74
This rep ort documents the integrated system for forecasting electric energy and load.
In the system, the service area models oF electrical energy (kWh) and the load distribution (minimum and maximum loads and lead duration curve) are linked to the state-level model of electrical energy (kWh).
Thus, the service area forecasts are conditional upon the state-level forecasts.
Such a linkage reduces considerably the data requirements for modeling service area electricity d emand.
Four utilities are selected to provide examples of the integrated forecasting system.
The statistical results suggest that the use of selected, important demand determinants, such as price and income, to explain the differences in electricity demand growth between the service area and the remainder of the corresponding state is appropriate.
The forecasting results show that the forecasted growth rates of elec tricity demand, in some cases, differ substantially between the service area and the corresponding state.
NUREG/CR-2696: CALCULATIONS OF TWO SERIES OF EXPERIMENTS PERFORMED AT THE POOLSIDE FACILITY USING THE DAK RIDGE RESEARCH REACTOR.
MAERKER,R.E.; WILLIAMS,M.L.
Oak Ridg e Nati onal Laboratory.
June 1982.
36pp.
8206240078.
ORNL/TM-8326.
13607:244.
This rep ort contains two papers that were presented at the Fourth ASTM-EURATOM Symposium on Reactor Dosimetry in Washington, D. C.
on March 22-26, 1982 and serves as documentation of the analytical work performed by the Engineering Physics Division.
These papers describe discrete ordinate calculations of two series of experiments that were performed at the Poolside Facility as part of the Surveillance Dosimetry Improvement Program, and are very similar in scope.
NUREG/CR-2699: TRANSPORTATION OF RADIOACTIVE MATERI AL IN MARYLAND June 1980-June 1981.
- Maryland, State of.
April 1982.
89pp.
8205110072.
13038:156.
The Maryland Department of Health and Mental Hygiene, under a Joint U.S.
DOT and NRC contract, conducted a one year study bepinning June 6,1980 to assess the transport of radioactive materials in Maryland.
Highway surveillance indicated that less than one truck in 10,000 was ha uling radioactive materials and that Low Specific Activity wastes constituted the primary material being transported.
Routing data was developed from surveillance and industry-supplied in f orma t ion.
Highway inspection and enforcement activities revealed that the level of transport and violations of radioactive materials indicate a minimum exposure risk to workers.
Some vio la t i ons of labeling and placarding regulations were, however, noted.
NUREG/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DIS?OSAL OF LOW-LEVEL RADIOACTIVE WASTE. LUTTON,R.J.; MALONE,P.G.; MEADE,R.D.; et al.
Army, Dept. of. Army Engineer Waterways Experiment Station.
May 1982.
84pp.
8206100024 13475:051.
Sixty-seven site parameters and parameter groups are identified as important characteristics of sites for disposal of low-level radioactive waste and require detailed evaluation.
Several of the most important parameters are needed for hydrological analysis while others are needed for facility design, construction, and operation.
Still others are needed for baseline and detection stages of monitoring.
It is recommended that all parameters be evaluated by technically qualified personnel.
Appropriate tests and documentatien methods are discussed in a second report, which will follow.
- However, 75
l site-specific testing or elaborate field measurement will not always be necessary, i. e., where indicated to be unnecessary on a technical basis.
Much of this report, Appendices A through G, is direc ted to explaining the importance of parameters and to establishing site-specific limitions.
NUREG/CR-2704:
U. S.
REACTOR SPENT-FUEL STORAGE CAPABILITIES. LEE,W.J.s HOFFMAN,C.C.s CAVINESS,C.K.
Nuclear Assurance Co.p.
June 1982.
56pp.
8206290529.
13660:001.
This rep ort describes the spent-fuel storage situation at reactors in the United States.
The focus of the report is on the reactors that are developing a spent-fuel storage problem and the alternatives the utilities are utilizing and planning to use to minimize the problem.
The alternatives the utilities are using and/or considering are described in the report and include:
High-Density Storage Racks l
Double-Tiered Storage Racks Rod Consolidation Dry Storage Systems Fuel Transshipments At-Reactor Storage Pools All of these alternatives are not available to every reactor and utility that is faced with a spent-fuel storage problem.
Generally, utilities are veracking or are planning to verack those spent-fuel pools that can be reracked with higher-density racks or double-tiered racks.
Where veracking is not feasible, the fuel transshipments are being performed or considered.
Since none of these other alternatives have been fully approved and licensed, these alternatives are all being evaluated.
NUREG/CR-2711: PERFORMANCE AND DESIGN REGUIREMENTS FOR A GRAPHICS DISPLAY RESEARCH FACILITY. TILLITT D.N.s PETERSEN,R.J.s SMITH,R.L.
I EG&G, Inc.
June 1982.
62pp.
8207190007.
EGG-2194.
13921: 233.
f Performance and design requirements for a Graphics Display Research Facility (GDRF) are presented.
Th e GDRF is an evolu tionary, computer-based, human-engineering experimentation center that is specifically designed to address long-tern research issues as sociated with automation, human performance, and risk in the operation of nuclear facilities.
Research capabilities provided by this facility will directly support the licensing and regulations of nuclear facilities within the United States.
This report discusses: the requirements, specifications, and implementation considerations For the facilitys the necessary hardware, software, and personnel capabilitiess and the potential costs of construction and operation
(
for various levels of research activity.
Research provided by this facilitig is intended to satisfy NRC needs to: (a) confirm design adequacy of, and develop evaluation criteria for computerized graphic display and other information presentation mechanisms proposed for use in' nuclear power plants, and (b) assess the possible effects on operator performance of computer-based operator-support concepts.
lhe ultimate goal of this research is to suppor t regulatory direc tives For minimizing th e risk of human error in the operation of nuclear facilities.
NUREC/CR-2713: VAPOR DEPOSITION VELOCITY MEASUREMENTS AND CORRELATIONS FOR I(2) AND C sI. NICOLOSI.S.L.s DAYDUTT,P.
Battelle Memorial Institute, Columbus Laboratories.
May 1982.
42pp.
8206090117.
76
BMI-2091.
13457:008.
Vapor deposition velocities were measured for I(2) and Cs1 vapors depositing on prefilmed Type 304 stainless steel and Inconel 600 surfaces in s team atmosph eres.
This work was performed to extend the data base of the TRAP (Transport of Radioluclides in Primary systems) code.
Arrenh ius type vapor deposition velocity correlations were developed for I(2) and for CsI vapors d ep os i ting on these materials.
The 300-900 C correlation for Inconel 600 is V(d) = 3.49 x 10(-6) exp(3940/RT), and the 300-1130 correlation for Type 304 stainless steel is V(d) = 2.53 x 10(-3) exp(-6670/RT).
The I(2) vapor deposition velocity correlation for Inconel 600 should not be used for temperatures greater than 900 C since this correlation gives a decreasing trend with increasing temperature whereas our experiments showed some evidence that the temperature d ependence of the vapor deposition velocity for this system may change to an increasing trend at 900 C.
Th e 300-1130 C correlation for Type 304 stainless steel should not be used at lower temperatures since the low temperature vapor deposition velocities decrease with temperature whereas the hiDh temperature vapor deposition velocities increase with temperature for this system.
The correlation derived f or CsI vapor depositing onto Type 304 stainless steel surfaces is V(d) 1.65 x 10(-9) e x p (21600/RT )
=
for 550-1040 C.
The correlation derived for CsI vapor depositing onto Inconel 600 s urf aces is V(d) = 6.36 x 10(-8 ) exp(13670/RT) for 815-1040 C.
NUREG/CR-2717: EXPERIMENT DATA REPORT FOR LOFT ANTICIPATED TRANSIENT WITHOUT SCRAM EXPERIMENT L9-3.
BAYLESS,P.D.; DIVINE,J.M.
EG&G, Inc.
June 1982.
189pp.
8206240072.
EGG-2195.
13609:106.
Selected pertinent and uninterpreted data from the third anticipated transient with multiple failures experiment (Experiment L9-3) conducted in the Loss-of-Fluid Test (LOFT) facility are presented.
The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system th erma l-h y d rauli c and nuclear conditions.
The operation of the LOFT system is typical of large [approximately 1000 MW(e)], commercial PWR op era tion s.
Experiment L9-3 simulated a loss-of-feedwater anticipated transient without scram.
The loss-of-feeduater accident led to an increase in the primary coolant system temperature and pressure.
Both the experiment power-operat ed relief valve (PORV) and safety relief valve opened and were able to limit and control the pressure tran s i ent.
The plant was th en rec overed with the cor. trol rods still withdrawn by injecting 72OO ppm borated water, manually cycling t h e P ORV, and feeding and bleeding the steam generator.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY, MISSOURI. BOOTH,L.F.s GROFF,D.W.; MC DOWELL. G. S. s et al.
Radiation Management Corp.
May 1982.
139pp.
8206100069.
13480:003.
This rep ort presents the results of a radiological survey of the West Lake Landfill, St. Louis County, Missouri, performed by Radiation Management Corporation during the spring and summer of 1981.
Measurements were made to determine e x terna l radiation levels, concentrations of airborne contaminants and the identity and concentrations of subsurface deposits.
Results indicate that large volumes of uranium ore residues, probably originating from th e Hazelwood, Missouri, Latty Avenue site, have been buried at the West Lake Landfill.
Two areas of contamination, covering more than 15 acres and located at depths of up to 20 feet below the present surface, have 77
been identified.
There is no indication that significant quantities oF contaminants are moving off-site at this time.
NUREG/CR-2727 V01: ECOLOGICAL STUDIES OF WOOD-BORING BIVALVES IN THE VICINITY OF THE OYSTER CREEK NUCLEAR GENERATING STATION. Progr ess Report, September-November 1981. HOAGLAND, K. E. s CROCKET L.
Lehigh Univ.
June 1982.
52pp.
8207220666.
14024:214 The species composition, distribution, and population dynamics of wood-boring b ivalves are being studied in the vicinity of the Oyster Creek Nuclear Generating Station, Barnegat Bay, New Jersey.
Untreated wood test panels are used to collect organisms at 12 stations.
Physiological tolerances of 3 species are also under investigation in the laboratory.
Competition among the spec ies is being analyzed.
In the fall of 1981, Teredo bartschi remained in Oyster Creek despite continuous prolonged outages of the Oyster Creek Nuclear Generating Station.
It did not spread to Forked River or Waretown as it had done in other years when the effluent was present.
The peak in.la rval production and settlement of T.
bartschi occurred between Sep tember and October.
Settlement of shipworms occurred on no r..onthly panels except those in Oyster Creek during the period of this report.
Laboratory experiments r evealed that T.
bartschi becomes inactive at 5 d egrees C (24 parts /thousand ) and T.
navalis shows signs of osmotic stress below 10 parts /thousand at 18 degrees C.
The shipworms in Barnegat Bay do not show a pr eference for settling at the mudline when the substrate is not limited.
NUREG/CR-2732: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES.(Tests S-IB-1 And S-IB-2).
SACKETT.K.E.s CLEGG. L. H.
EG&G, Inc.
June 1982.
58PP.
8207190010.
EGG-2196.
13958:001.
This rep ort presents test data recorded for Tests S-IB-1 and S-IB-2 of the Semiscale Mod-2A Intermediate Dreak Test Series.
These tests are par t of a series of Semiscale tes ts that investigate the thermal-hydraulic phenomena resulting from a hypothesized loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR) system.
These tests provide experimental data for assessing the analytical capability of computer codes used in LOCA analysis.
lests S-IB-1 and S-IB-2 were conducted from initial conditions clos ely approximating the specified initial conditions of: 15.5-MPa system pressure, 557-K cold leg temperature, and
- 1. 95-MW c ore power level.
This report presents uninterpreted data from both tests for future analysis.
The data, presented as graphs in engineering units, have been analyzed only to the estent necessary to ensure that they are reasonable and consistent.
NUREC/CR-2736: TRANSPORTATION OF RADIOACTIVE MATERI AL IN MICHIGAN. September 1980-August 1981. MCCARTY,M.J.s HENNIGAN,J.M.s BRUCHMANN,G.W.
Michigan, State o f.
May 1982.
98pp.
8206100050.
13479:100.
Most of the radioactive material transported into and through the State of Mich igan is comprised of rad iopharmaceuticals.
The remainder includes rdioactive waste from nuclear power plants and hospitals, uranium ore concentrate (yellowcake) from Ontario, Canada, and periodic spent fuel shipments from a university research reactor.
Investigations have revealed that minor violations of packaging and shipping paper regulations persist but to a lasser degree than in previous year s.
Major operational problems associated with two courier i
l l
7s
e companies have substantially ~ improved but'still require improveeent.
Several minor transportation accidents,'are reported, none of which, resulted in significant radiation exposure.
Joint investigations with federal agenc ies were made, and some resulted in legal action te shippers.
Future work performed will be under a contract with the U.
S.
Department of leansportation.
This report describes the fourth y ear 's study by the state of Michigan of the transportation o8 rad ioac tive material in Michigan, during the period September 1, 1960 t o Aug u s t 31,.1'l81.
For the periodr.
September 1, 1979 to August 31, 1980 see N'JREG/CR-2034s for September 1,
1978 to August 31, 1979 see NU,1EG/ CR-119 43 and the first year in unpublished.
NUREG/CR-2737: EVALUATION OF BULK PROPER 11ES OF RADWASrE. CLASS AND CERAMIC CONTAINER HATERIALS TO DETERMINE LONG-TERM STABILITY.
MACEDO.P.B.s BARKATT.A.
Catholic Univ.
June 1982.
116pp.
8207140265.
13843:073.
The general ob Jective is to in've stigat e the characteristics of simulated HLW glass and ceramics with respect to' surface corrosion, network dissolution and subsequent leaching under an, aqueous environment.
Based on these characteristic s, a model has been proposed to predict the durability to these waste forms.
Epecific tasks are:
- 1. Leach ing properties under neutral pH with relatively high dilution 2.
Wet-dry cycling test 3.
Variable pH under high dilution 4.
Flow-rate dependency 5.
MCC-1 round rebin participation.
NUREQ/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH THE 30TH PARALLEL LINEAMENT.
Final ~ 2chnic al Report, June 1979-June 1781. BRAILE,L.W.s HINZE,W.J.4 SEXTON,J.L.s et al.
Purdue Univ.
June 1982.
06pp.
8207140097.
13846:227.
Gravity, magnetic, geologic, and seismicity data have been combined in a seismotectonic analysis of th e New Madrid seismic zone.
Previous stud ies have presented evidence for sv.eral rif t zones in this area (Upper Mississippi enbagment), including the Ree1 Foot ri f t, a late Precambrian-early Paleozoic f ailed ar.n whic h extc.1ds north-northeast from the anci ent continental margin.
4Je suggest that the northern terminus of the Reelfoot rift forms a rift complea, with arms extending northeast ino southwesten Indiana, northwe r t along the Missis sippi River, and east into western Kentucky, wha 5 appears.to correlate well with the seismicity in the area.
This correlation suggests that faulto associated with this rift complex are being reactivated in th e contemporary stress field (east-northeast compression).
If this interpretation is valid, it represents a seismotectonic model which can be used to predict the extent of. future seismicity in the New Madrid seismic zone.. The proposed rif t complex also provides a coherent model for the tectonic development of this region of the North American midcontinent.
NUREG/CR-2760: ASSESSMENT OF SCALE EFFECTS ON VCRTEXING, SWIRL, AND INLE1 LOSSES IN LARCE SCALE GbMP MODELS. P ADMANAB HAN, M., HECKER. C'. E.
Alden Research Laboratory.
June 1982.
81pp.
8207220641.
ARL 02.
14024:290.
79
To verify the use of reduced scale hydraulic models of large scale ratios to demonstrate the performance of containment emergency sumps, a test program involving two geometric scale models (1:2 and 1: 4) of a full site sump (1: 1) was undertaken as a part of the total test program towards the resolution of Unresolved Safety Issue A-43,
" Containment Emergency Sump Performance.d The test results substantiated that hy draulic models of large scale such as 1:2 to 1:4 reliably predicted the sump bydraulic performance.
No scale effects on vortening or air-withdrawal s were apparent within the tested prediction range for both models.
- However, a good predic tion of pipe flow swirl and inlet loss coefficient was found to require that the approach flow Reynolds number and p ipe Reynolds numb er be above certain limits.
Based on the results of these testse it is concluded that properly designed and operated, reduced scale hydraulic models of geometric scales 1:4 or larger can be used both by utilities and by regulatory agencies to p rove the satisfactory hydraulic perf ormance of sump designs.
NUREQ/CR-2772: HYDRAULIC PERFORMANCE OF PUMP SUCTION INLETS FOR EMERGENCY CORE COOLING SYSTEMS IN BOILING WATER REACTORS.
PADMANABHAN,M.
Alden Research Laboratory.
June 1982.
60pp.
8207220649.
ARL-398A.
14024:124.
This doc ument reports on the hydraulic performance of representative Boiling Water Reactor Residual Heat Removal ssction inlet configurationss Mark I and Mark II and III designs.
Parameters of interest were air-ingestion, vortex types, pipe swirl, and pressure loss coefficients.
Tests were conducted with nearly uniform and non-uniform inlet approach flows.
Flows and submergences ranged from 2000 to 12000 gpm per pipe and 2 to 5 ft, respectively, giving a Froude number range from 0.17 to 1.06.
Zero air-withdrawal was measured for both configurations for Froude numbers equal to or less than O.8 even under non-uniform approach flowss no air-core vortices were observed for the same flow conditions.
At a Froude number above ~1.0 and with non-unif orm approach flows, air-withdrawal up to 4% by volume was observed in the Mark I design and air-withdrawals up to 0.5% by volume were observed in the Mark II and III design.
Swirl levels in the pipe up to 7 degrees were measured for Mark Il and III designs and up to 3 degrees for Mark I design.
Inlet loss coefficients were about 1.7 for Mark II and III designs and about 1. 0 for Mark I design.
NUREC/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATE GEOMETRY. BANKOFF,S.C.s KIM.H.J.s TANKIN R. S. s et al.
Northwestern Univ.
June 1982.
51pp.
8207220672.
NU-82018.
14024:163.
The study of steam condensation in countercurrent stratified flow of steam and subcooled water has been carried out in a rectangular channel, with an inclination angle 33 degrees from the horizontal.
The variables in this experiment were the inlet water and steam flow rates and the inlet water temperature.
Condensation heat transfer coefficients were determined as functions of local steam and water flow rates and the degree of subcooling.
Correlations are given for the local Nusselt number for the smooth and for the rough surface regimes, and also for the dimensionless wave amplitude.
A turbulence-centered model is also considered.
It is shown that better agreement with the data can be obtained if the characteristic lengths in the turbulent Nusselt number and turbulent Reynolds number are taken to be wave 80
1 t
j i
i I
amplitude and the friction velocity, ra th er than the water lager i
thickness and 0. 3 times the mean water velocity.
A new correlation is presented based on the wave parameters.
t 4
NUREQ/CR-2788: STRENGTH AND STIFFNESS OF UNIAXIALLY TENSIONED
)
REINFORCED CONCRETE PANELS SUBJECTED TO MEMBR/NE SHEAR. HILMY,S.2.a j
WHITE,R.N.s GERGELY,P.
Cornell Univ.
June 1982.
223pp.
8207220670.
14020:170.
This report presents experimental and analytical results on internal pressurization effects and seismic shear effects in a concrete containment vossel that is cracked by tension in one direction onig.
The investigation was a continuation of research i
reported in NUREO Reports CR-1602 and CR-2049.
The experimental program, which was restricted to 6 in, thick flat specimens with two-way reinforcement, included estab lishment of (a) extensional stiffness for uniasially tensioned specimens stressed to 0.6f(g), and i
(b) shear strength and stiffness of these cracked specimens with i
tension levels ranging from 0 to 0.9f(g)s values were about 10 to 15 percent higher than in similar biaatally tensioned specimens.
Eleven (11) specimens were tested (6 in monotonic shear and 5 in reversing cyclic shear).
1 Results are correlated with earlier experimental results from studies on similar specimens and on other simpler specimens that were tested in many different labs (Cornell, PCA, Torontc, Japan, and elsewhere).
A finite element representation of behavior is developed l
for prediction of initial shear modulus.
The report concludes with design recommendations.
i NURE0/CR-2790: AUTOMOBILE IMPACT FORCES ON CONCRETE WALL PANELS.
CHIAPETTA,R.L s PANO,E.C.
Chiapetta, Welch & Associates,Ltd.
(one i
1982.
260pp.
8207060002.
CWA 4010-FR.
13716:001.
The ob ective of this study was to dev.alop force-time impact J
I signature data for use in the design or evaluation of nuclear power d
plant structures subject to tornado-borne automotive vehicle impact.
The approach was based on the use of analytical vehicle models to j
calculate impact forces.
To assess the significance of vehicle /struc ture intersc tion f or head-on impact force-histories, a lumped-mass model of a reinforced concrete wall panel was coupled to a one-dimensional vehicle model for numerous panel design configurations within the range of practical interest.
Vehicle-structure in terac ti on found to have relatively little effect on the force-histories.
was The sensitivity of structural response to variations in force signature characteristics was established and idealized impact i
force-time relations were developed for five distinct impact speeds ranging from 20-60 meters /sec.
The use of these relations produce less conservative estimates of structural deflection, for all impact speeds considered, than the currently accep ted design procedure.
I s1
l l
l t
Contractor Report Number Index This index lists, in alphabetical order, the contractor-issued report codes for o
the NRC contractor reports in this compilation. Each contractor code is cross-referenced to the NUREG/CR for the report and to the 10-digit NRC Document Control System accession number.
SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER ANL-81-29 NUREG/CR-2181 VO4 EGG-2196 NUREG/CR-2732 ANL-82-6 NUREG/CR-2647 EGG-CAAD-5629 NUREG/CR-2059 ANL-CT-81-35 NUREG/CR-1622 ENICD-1110 NUREC/CR-2644 ARL-398A NUREG/CR-2772 EPRI NP-1584 NUREG/CR-2133 ARL-48-82 NUREG/CR-2760 EPRI NP-1781 NUREG/CR-2231 ASA NO. 635 NUREG/CR-2297 EPRI NP-1783 NUREG/CR-2229 VO1 DMI-2091 NUREG/CR-2713 ERL-ARL-108 NUREG/CR-2521 DMI-2093 NUREG/CR-2682 GC-TR-82-171 NUREG/CR-2589 DMI-2094 NUREC/CR-2683 GEAP-24939 NUREQ/CR-2133 DNL-NUREG-51328 NUREC/CR-1890 GEAP-24962-1 NUREG/CR-2229 VO1 BNL-NUREG-51366 NUREC/CR-2039 GEAP-24964 NUREC/CR-2231 BNL-NUREG-51410 NUREG/CR-2193 VO1 N2 LA-9116-MS NUREC/CR-2434 DNL-NUREG-51449 NUREC/CR-2317 VO1 N3 LA-9133-MS NUREC/CR-2464 ENL-NUREC-51449 NUREC/CR-2317 VO1 N4 LA-9197-MS NUREG/CR-2565 BNL-NUREG-51480 NUREG/CR-2416 LA-9209-PR NUREG/CR-2281 VO2 DNL-NUREG-51481 NUREG/CR-2417 LA-9229-PR NUREG/CR-2281 VO3 DNL-NUREG-51499 NUREG/CR-2516 VO1 N1 LA-9258-MS NUREG/CR-2622 DNL-NUREG-51504 NUREG/OR-2542 LA-9266-MS NUREC/CR-2652 BNL-NUREG-51506 NUREG/CR-2543 LA-9276-SR NUREG/CR-2632 BNL-NUREG-51529 NUREG/CR-2685 LA-9277-MS NUREG/CR-2633 DNL-NUREG-51530 NUREC/CR-2686 LA-9301-MS NUREG/CR-2569 CONF-810372 NUREG/CP-OO22 LMF-92 NUREG/CR-2512 CSNI REPT NO.63 NUREG/CP-OO31 VO2 MTR-82W26 NUREG/CP-OO26 CSNI REPT NO 63 NUREG/CP-OO31 VO1 MXC-301 NUREG/CR-2671 CWA 4010-FR NUREG/CR-2790 NU-8201B NUREC/CR-2783 DPST-80-4 NUREG/CR-1681 ORNL-5822 NUREC/CR-2435 DPST-80-6 NUREG/CR-2356 ORNL-5824 NUREG/CR-2493 EGG-2037 NUREG/CR-0169 V13 ORNL-5846 NUREG/CR-2455 EGG-2070 NUREC/CR-1826 VO1 ORNL-5848 NURFG/CR-24E6 EGG-2070 NUREG/CR-1826 VO2 ORNL/CSD/Tri-158 NUREG/CR-2223 EGG-2179 NUREC/CR-2618 ORNL/CSD/TM-160 NUREG/CR-2306 EGG-2183 NUREG/CR-2636 ORNL/CSD/TM-165 NUREG/CR-2494 EGG-2184 NUREC/CR-2648 ORNL/NSIC-182 NUREG/CR-2497 VO1 EGO-2194 NUREC/CR-2711 ORNL/NSIC-182 NUREG/CR-2497 VO2 EGG-2195 NUREG/CR-2717 ORNL/NSIC-195 NUREC/CR-2172 83
SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER ORNL/NSIC-196 NUREG/CR-2173 PNL-4193 NUREG/CR-2564 ORNL/NSIC-196 NUREC/CR-2173 ERR PNL-4201 NUREG/CR-2600 ORNL/NSIC-2OO NUREC/CR-2OOO VO1 N2 PNL-4225 NUREG/CR-2642 ORNL/NSIC-2OO NUREG/CR-2OOO VO! N3 PNL-4240 NUREC/CR-2567 ORNL/NSIC-2OO NUREC/CR-2OOO VO1 N4 SAI-288-82-PA NUREC/CR-1672 VO3 ORNL/NSIC-2OO NUREG/CR-2OOO VO1 NS SAND 79-1909 NUREC/CR-1636 VO4 ORNL/NUREG-85 NUREG/CR-2469 SAND 80-0764 NUREC/CR-1851 ORNL/NUREG/TM-3 NUREC/CR-2220 VO3 SAND 80-1646 NUREG/CR-1594 VO4 ORNL/NUREG/TM-3 NUREG/CR-2220 VO2 SAND 80-1897 VO3 NUREC/CR-1659 VO3 ORNL/NUREG/TM-4 NUREG/CR-2053 SAND 80-2662 NUREC/CR-1820 ORNL/NUREG/TM-4 NUREC/CR-2525 VO2 SAND 81-OO79 NUREG/CR-2431 ORNL/NUREC/TM-4 NUREC/CR-2525 VO1 SAND 81-0933 NUREC/CR-2473 OHNL/NUREC/TM-4 NUREG/CR-2525 VO3 SAND 81-1230 NUREG/CR-2611 ORNL/NUREC/TM-4 NUREG/CR-2525 VOS SAND 81-1370 NUREC/CR-2194 ORNL/NUREQ/TM-4 NUREG/CR-2525 VO7 SAND 81-1529 VO1 NUREC/CR-2238 VO1 ORNL/SUB/80-404 NUREG/CR-2597 SAND 81-1622 NUREG/CR-2412 ORNL/TM-7859 NUREC/CR-2544 SAND 81-1653 NUREG/CR-2343 ORNL/TM-7868 NUREC/CR-2184 SAND 81-17J2 NUREG/CR-2279 ORNL/TM-7921 NUREC/CR-2362 SAND 81-1754 NUREG/CR-2283 ORNL/TM-7947 NUREG/CR-2692 SAND 81-1906 NUREC/CR-2314 ORNL/TM-7985/P2 NUREC/CR-2353 VO2 SAND 81-1978 NUREC/CR-2350 ORNL/TM-8011 NUREC/CR-2610 SAND 81-2655 NUREG/CR-2582 ORNL/TM-8024 NUREC/CR-2470 SAND 81-7160 NUREC/CR-2377 ORNL/TM-8073 NUREG/CR-2392 SAND 82-OOO6 NUREG/CR-2481 ORNL/TM-8085 NUREG/CR-2393 SAND 82-0319 NUREC/CR-2546 ORNL/TM-8099 NUREC/CR-2612 SAND 82-0332 NUREC/CR-2551 ORNL/TM-8172 NUREC/CR-2505 SANDB2-0377 NUREC/CR-2559 ORNL/TM-8190 NUREC/CR-2366 VO2 SAND 82-0410 NUREG/CR-2588 ORNL/TM-8231 NUREC/CR-2204 VO4 SAND 82-0449 NUREG/CR-2581 ORNL/TM-8236 NUREG/CR-2586 SANDB2-0720 NUREC/CR-2681 ORNL/TM-8237 NUREG/CR-2587 SAND 82-0927 NUREC/CR-2403 SO1 ORNL/TM-8252 NUREC/CR-2141 VO4 SANDB2-7013 NUREG/CR-2393 ERR ORNL/TM-8260 NUREG/CR-2221 VO4 SAND 82-7013 NUREC/CR-2593 ORNL/TM-8274 NUREG/C3-2629 SAND 82-7014 NUREG/CR-2394 ERR ORNL/TM-8307 NUREC/CR-2299 VO4 SAND 82-7014 NUREC/CR-2594 ORNL/TM-8326 NUREG/CR-2(96 SAND 82-7017 NUREG/CR-1245 ROI l
PNL-3117 NUREG/CR-1030 VO2 SAND 82-7018 NUREG/CR-2604 l
PNL-3762 NUREG/CR-2019 SAND 82-7019 NUREG/CR-2605 l
PNL-3772 NUREC/CR-2022 SANDB2-7063 NUREG/CR-2760 PNL-4028 NUREG/CR-2387 SAND 82-7064 NUREG/CR-2772 1
PNL-4039 NUREG/CR-2201 UCID-18146 NUREC/CR-1233 VO4 l
PNL-4070 NUREG/CR-2432 UCPL-15490 NUREG/CR-2301 l
PNL-4086 NUREC/CR-2460 UCRL-53021 VO4 NUREC/CR-2015 VO4 PNL-4106 NUREG/CR-2413 PNL-4154 NUREC/CR-2651 1
l l
84
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-2637: EMERGENCY RESPONSE CAPABILITIES AND EXAMPLE ASSESSMENTS FOR AIRBORNE RADIONUCLIDE DISCHARGES.
ADAMS,R.E.
NUREG/CR-2299 VC4: AEROSOL RELEASE AND TRANSPORT PROGRAM. Quarterly Progress Report For October-December 1981.
ADLER,J.J.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY, MISSOURI.
ADRIANO,D.C.
NUREC/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FROM AGRO-ECOSYSTEMS. Annual Progress Repor t,0c to ber 1980-Sep tember 1981.
AHN,T.M.
NUREG/CR-2317 VO1 N3: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATERIALS.Guarterly Progress Report, July-September 1981.
NUREG/CR-2317 VO1 N4: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATERIALS.Guarterly Progress Report,0ctober-December 1981.
ALBANESE,J.R.
NUR EG /CR-2381 : GEOLOGIC AND HYDROLOGIC RESEARCH AT THE WESTERN NEW YORK NUCLEAR SERVI CE CENTER, WEST VALLEY, NEW YORK. Progres s Rep ort, August 1979-Ju1y 1981.
ALBERTS,J.J.
NUREG/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FROM AGRO-ECOSYSTEMS. Annual Progress R ep or t, 0c t o b er 1980-September 1981.
ALLISON,S.A.
NUREG/CR-2229 VO1: BWR LARGE BREAK SIMULATION TESTS--BWR BLOWDOWN / EMERGENCY CORE COOLING PRG.'R AM.
ALTMAN,W.
NUREG-0903: SURVEY OF INDUSTRY AND GOVERNMENT PROGRAMS TO COMBAT DRUG AND ALCOHOL ABUSE.
AMES,P.L.
NUREG/CR-1681: WRAP-PWR VER IFICATION STUDIES.
AMIRIJAFARI,B.
NUREG/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDI A: Tec hnical Review of NUREG/CR-1636,Vols 1,2 and 3, Dec emb er 1,1981-March 31,1982.
ANDERSON,J.L.
NUREG/CR-2636: EXPERIMENTAL DATA REPORT FOR AIR-WATER FLOODING TESTS OF THE FLECHT-SEASET PROGRAM SET FACILITY VESSEL UPPER PLENUM.
ANDERSON,R.C.
85
NUREG/CR-2603: BUDDLE BEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
ANDERSON,R.L.
NUREG/CR-2470: THERMOMETRY IN THE MULTIROD DURST TEST PROGRAM.
ANKLAM,T.M.
NUREG/CR-2455: EXPERIMENTAL INVESTIGATIONS OF BUNDLE BOILOFF AND REFLOOD UNDER HIGH-PRESSURE LOW HEAT FLUX CONDITIONS.
NUREG/CR-2456: EXPERIMENTAL INVESTIGATIONS OF UNCOVERED-BUNDLE HEAT TRANSFER AND TWO-PHASE MI XTURE-LEVEL SWELL UNDER HIGH-PRESSURE LOW HEAT-FLUX CONDITIONS.
NUREG/CR-2525 VO1: ORNL ROD DUNDLE HEAT TRANSFER TEST DATA. Volume 1-ORNL Small Dreak LOCA Test Series I: Experimental Data Report.
ARELLANO,F.E.
NUREG/CR-2283: DIRECT OBSERVATION OF MELT DEHAVIOR DURING HIGH TEMPERATURE MELT / CONCRETE INTER ACTIONS.
ATLURI.S.N.
NUREG/CR-2494: OR-FLAW: A FINITE ELEMENT PROGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFINED FLAWS IN PLATES. CYLINDERS AND E
ATWOOD,C.L.
PRESSURE-VESSEL NOZZLE CORNERS.
NUREG/CR-2099: COMMON CAUSE FAULT RATES FOR DIESEL GENERATORS: ESTIMATES BASED ON LICENSEE EVENT REPORTS AT U. S.
COMMERCIAL NUCLEAR POWER PLANTS, 1976-1978.
AU, M. L.
NUREG-0863: SURVEY OF FOREIGN REACTOR OPERATOR GUALIFICATIONS, TRAINING, AND STAFFING REGUIREMENTS.
BAILEY,P.T.
NUREG/CR-2597: STEADY-STATE PRESSURE LOSSES FOR MULTIROD BURST TEST (MRDT) BUNDLE B-5.
DAKER D.A.
NUREG/CR-2201: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1978.
BALL,S.J.
NUREG/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report,0ctober 1-December 31,1981.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
BANDER,T.J.
NUREC/CR-2022: TECHNICAL REVIEW OF THE DISPERSION AND DOSE MODELS USED IN THE MILDOS COMPUTER PROGRAM.
DANKOFF,S.G.
NUREG/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATE l
GEOMETRY.
l DANKS,W.W.
NUREC/CR-2496: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
DARKATT,A.
I NUREG/CR-2737: EVALUATION OF BULK PROPERTIES OF RADWASTE GLASS AND l
CERAMIC CONTAINER MATERI ALS TO DETERMINE LONG-TERM STABILITY.
i DARNTHOUSE.L.W.
NUREC/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
BARTON J.E.
t I
NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4.4 - 30 SSTF DESCRIPTION l
DOCUMENT.
BASIN S.L.
l NUREG/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDI A: Tec hnical Review of NUREG/CR-1636, Vols 1,2 and 3 Dec ember 1,1981-March 31,1982.
BASS.B.R.
36 l
NUREC/CR-2494: OR-FLAW: A FINITE ELEMENT PROGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFINED FLAWS IN PLATES, CYLINDERS AND PRESSURE-VESSEL NOZZLE CORNERS.
DATES,E.F.
NUREG-0849: STANDARD REVIEW PLAN FOR THE REVIEW AND EVALUATION OF EMERGENCY PLANS FOR RESEARCH AND TEST REACTORS.
BAYBUTT,P.
NUREG/CR-2682: CITADEL: A COMPUTER CODE FOR THE ANALYSIS OF IODINE BEHAVIOR IN STEAM QENERATOR TUBE RUPTURE ACCIDENTS.
NUREG/CR-2683: IODINE BEHAVIOR IN STEAM GENER ATOR TUBE RUPTURE ACCIDENTS.
NUREG/CR-2713: VAPOR DEPOSITION VELOCITY MEASUREMENTS AND CORRELATIONS FOR I(2) AND C sI.
DAYLESS,P.D.
NUREG/CR-2717: EXPERIMENT DATA REPORT FOR LOFT ANTICIPATED TRANSIENT WITHOUT SCRAM EXPERIMENT L9-3.
DECKMAN,R.J.
NUREG/CR-2434: FRAC (FAILURE RATE ANALYSIS CODE): A COMPUTER PROGRAM FOR ANALYSIS OF VARIANCE OF FAILURE RATES. An Application User 's Guide.
NUREG/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTI AL OR POISSON DATA.
DEEMILLER,R.M.
NUREG/CR-2591: ESTIMATING THE POTENTIAL INDUSTRIAL IMPACTS OF A NUCLEAR REACTOR ACCIDENT.
BEGEJ,C.W.
NUREG/CR-2642: LONG-TERM SURVIVABILITY OF RIPRAP FOR ARMORING URANIUM MILL 'AILINGS AND COVERS: A LITERATURE REVIEW.
DELL,J.T.
NUREG/CR-2493: AGUEOUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And Assessment.
BENNETT,H.A.
NUREC/CR-2546: REACTOR SAFEGUARDS AGAINST INSIDE SABOTAGE.
DERANEK,F.
NUREC/CR-1681: WRAP-PWR VERIFICATION STUDIES.
BERMAN,M.
NUREG/CR-2481: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report, April-September 1981.
ULUHM,D.
NUREG/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Tech nical Report, Augus t 1980-September 1981.
BOOTH.L.F.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUJG COUNTY. MISSOURI.
DOUCHARD,D.A.
NUREC/CR-2314: AGING WITH RESPECT TO FLAMMABILITY AND OTHER PROPFRTIES IN FIRE-RETARDED ETHYLENE PROPYLENE RUDDER AND CHLOROSULFONATED POLYETHYLENE.
DRADLEY,D.R.
NUREC/CR-2603: DUBBLE DEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
DRADLEY,E.R.
NUREC/CR-2567: FINAL DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMitLY (IFA)-432.
DRADLEY,0.D.
NUREG/CR-2652: EVALUATION AND PERFORMANCE OF CLOSED-CIRCUIT DREATHING APPARATUS.
DRAILE.L.W.
NUREG/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH THE 38TH PARALLEL LINEAMENT.
Final Technical Report, June 1979-June 1981.
87
DRITE,D.W.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULATORY DASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE FUEL CYCLE.
DRONSON,F.L.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY, MISSOURI.
BROWN,J.D.
NUREG/CR-1636 VO4: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: EFFECTS OF VARI ABLE HYDROLOGIC PATTERNS ON THE ENVIRONMENTAL TRANSPORT MODEL.
DROWN,W.
NUREG-0903: SURVEY OF INDUSTRY AND GOVERNMENT PROGRAMS TO COMBAT DRUG AND ALCOHOL ABUSE.
BRUCHMANN,G.W.
NUREG/CR-2736: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MICHIGAN. September 1980-August 1981.
BRYAN,R.H.
NUREG/CR-2141 VO4: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM.Guarterly Progress Report For October-December 1981.
DRYSON,J.W.
NUREG/CR-2494: OR-FLAW: A FINITE ELEMENT PROGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFINED FLAWS IN PLATES. CYLINDERS AND PRESSURE-VESSEL NOZZLE CORNERS.
DUCKALEW,W.H.
NUREC/CR-2581: SOME EFFECTS OF ELECTRONS SLOWING DOWN IN MATERIALS WITH APPLICATION TO SAFETY-RELATED EQUIPMENT GUALIFICATION.
NUREG/CR-2582: RADI ATION CAPADILITIES OF THE SANDI A HIGH INTENSI1Y ADJUSTABLE CODALT ARRAY.
BUSH,C.
l NUREG-0903: SURVEY OF INDUSTRY AND GOVERNMENT PROGRAMS TO COMDAT DRUG AND ALCOHOL ADUSE.
DUTLER,T.A.
NUREC/CR-2569: RESPONSE OF THE ZION & INDIAN POINT CONTAINMENT DUILDINGS TO SEVERE ACCIDENT PRESSURES.
CADA,0.F.
NUREC/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
CAMPDELL,D.A.
NUREG/CR-2493: AGUEOUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And Assessment.
CAMPDELL,J.E.
NUREG/CR-2343: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: THE DNET COMPUTER CODE USER'S MANUAL.
CANNON J.D.
NUREG/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
CARLSON,K.E.
NUREG/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Vo lume 1: System Models And Numerical Methods.
NUREG/CR-1826 VO2: RELAPS/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
CARR,K.R.
NUREG/CR-2470: THERMOMETRY IN THE MULTIROD DURST TEST PROGRAM.
CARTER,K.
NUREG/CR-2652: EVALUATION AND PERFORMANCE OF CLOSED-CIRCUIT BREATHING APPARATUS.
CARTWRIGHT,J.V.
NUREG/CR-2591: ESTIMATING THE POTENTIAL INDUSTRIAL IMPACTS OF A NUCLEAR REACTOR ACCIDENT.
88
CARUSO,S.C.
NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4.4 - 30 SSTF DESCRIPTION DOCUMENT.
I CATE,J.H.
NUREG/CR-2637: EMERGENCY RESPONSE CAPABILITIES AND EXAMPLE ASSESSMENTS FOR AIRBORNE RADIONUCLIDE DISCHARGES.
CAVINESS,C.K.
NUREG/CR-2704:
U. S.
REACTOR SPENT-FUEL STORAGE CAPABILITIES.
CHANGERY M.J.
NUREC/CR-2639: HISTORICAL EXTREME WINDS FOR THE UNITED STATES-ATLANTIC AND GULF OF MEXICO COASTLINES.
CHAPMAN R.L.
NUREG/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
CHEN J.C.
NUREC/CR-2015 VO4: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
CHEN,N.C.
NUREC/CR-2544: TWO-PHASE MASS FLUX UNCERTAINTY ANALYSIS FOR THERMAL-HYDRAULIC TEST FACILITY INSTRUMENTED SPOOL PIECES.
CHERN,W.S.
NUREG/CR-2692: AN INTEGRATED SYSTEM FOR FOREC ASTING ELECTRIC ENERGY AND LOAD FOR STATES AND UTILITY SERVICE AREAS.
CHIAPETTA,R.L.
NUREC/CR-2790: AUTOMOBILE IMPACT FORCES ON CONCRETE WALL PANELS.
CHOU,C.K.
NUREG/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS.
CHOW,H.
NUREG/CR-1826 VO1: RELAPS/ MOD 1 CODE MANUAL. Vo lume 1: System Mode ls And Numerical Methods.
NUREG/CR-1826 VO2: RELAPS/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
CHRISTENSEN,D.
NUREG/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
CHRISTENSEN,S.
NUREC/CR-2220 VO2: THE IMPACT OF ENTRAI'NMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
NUREC/CR-2220 VO3: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
CHUN,R.C.
l NUREC/CR-2015 VO4: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
CLAPP,N.E.
NUREG/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STdDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report,0ctober 1-December 31,1981.
l CLARK,R.C.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULAlORY j
BASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE i
FUEL CYCLE.
CLEGG,L.B.
i NUREG/CR-2648: EXPERIMENTAL DATA REPORT FOR SEMISCALE MOD-2A NATURAL CIRCULATION TEST SERIES ( TESTS S-NC-8B AND S-NC-9).
NUREG/CR-2732: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A INTERMEDI ATE BREAK TEST SERIES. (Tests S-IB-1 And S-IB-2).
CLEVELAND,J.C.
NUREC/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progress 89
Report,0ctober 1-December 31,1981.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
COBB,L.
NUREG-0903: SURVEY OF INDUSTRY AND GOVERNMENT PROGRAMS TO COMBAT DRUC AND ALCOHOL ABUSE.
CODELL,R.B.
NUREG-0868: A COLLECTION OF MATHEMATICAL MODELS FOR DISPERSION IN SURFACE WATER AND GROUNDWATER.
COHEN,J.
NUREC/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDI A: Tec hnical Review of NUREG/CR-1636, Vols 1,2 and 3, Dec emb er 1,1981-Marc h 31,1982.
COHEN,L.K.
NUREG-0837 VO1 NO1-2: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr ess Report, January-June 1981.
NUREG-0837 VO1 NO3: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr ess Rep ort, July-December 1981.
NUREG-0837 VO1 NO4: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr es s Rep ort, October-December 1981.
CONKLIN J.C.
NUREG/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report,0ctober 1-December 31,1981.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
CONOVER,W.D.
NUREG/CR-2350: SENSITIVITY ANALYSIS TECHNIGUES: SELF-TEACHING CURRICULUM.
COPUS,E.R.
NUREG/CR-2681: ESTIMATED RECURRENCE FREGUENCIES FOR INITIATING ACCIDENT CATEGORIES ASSOCIATED WITH THE CLINCH RIVER BREEDER REACTOR PLANT DESIGN.
COSTELLO F.
NUREG-0837 VO1 NO3: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr ess Report, Ju ly-D e c emb e r 1981.
NUREG-0837 VO1 NO4: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr ess Rep ort, October-December 1981.
CRANWELL.R.M.
NUPEG/CR-2343: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: THE DNET COMPUTER CODE USER'S MANUAL.
CROCKET,L.
NUREG/CR-2727 VO1: ECOLOGICAL STUDIES OF WOOD-BDRING BIVALVES IN THE VICINITY OF THE OYSTER CREEK NUCLEAR GENERATING STATION. Progr ess Report September-November 1981.
CROWELL S.L.
NUREC/CR-2432: A UNIGUE CONCEPT FOR LIGUID LEVEL AND VOID FRACTION DETECTION IN SEVERE FUEL DAMAGE TESTS.
CROWLEY,J.L.
NUREC/CR-2366 VO2: MULTIROD BURST TEST PROGRAM PROGRESS REPORT FOR JULY-DECEMBER 1981.
CUMMINGS,J.C.
NUREG/CR-1820: STATUS REPORT ON THE FISSION-PRODUCT RESEARCH PROGRAM.
CUNNINGHAM.M.E.
NUREC/CR-2567: FINAL DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-432.
NUREG/CR-2600: END-OF-IRRADI ATION DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-527.
CURRERI,M.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
90
1 l
CYBULSKIS,P.
j NUREG/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS PROGRAM: Calvert Cliffs No. 2 PWR Power Plant.
CZYSCINSKI,K.S.
NUREG/CR-2192 VO1 N2: EVALUATION OF ISOTOPE MIGRATION-LAND BURIAL. Guarterly Progress Repor t April-June 1981.
DAFOE,R.E.
NUREC/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
DARKOW,0.L.
NUREG/CR-2359: ATMOSPHERIC STRUCTURE PRIOR TO TORNADOES AS DERIVED FROM PROXIMITY AND PRECEDENT UPPER AIR SOUNDINGS.
DATESMAN C.H.
NUREC/CR-2297: SECURITY MANAGEMENT TECHNIQUES AND EVALUATIVE CHECKLISTS FOR SECURITY FORCE EFFECTIVENESS.
DAVENPORT L.C.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULAlOHY BASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE FUEL CYCLE.
DEDHIA D.D.
NUREC/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS.
DELENE.J.G.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
DENNING,R.S.
NUREC/CR-2ed2: CITADEL: A COMPUTER CODE FOR THE ANALYSIS OF IODINE BEHAVIOR IN STEAM GENERATOR TUBE RUPTURE ACCIDENTS.
NUREC/CR-2683: IODINE BEHAVIOR IN STEAM GENER ATOR TUBE RUPTURE ACCIDENTS.
DICKEY,J.M.
NUREG/CR-2542: SENSITIVITY STUDY USING THE FR ANTIC CODE FOR THE UNAVAILABILITY OF A SYSTEM TO THE FAILURE CHARACTERISTICS OF THE COMPONENTS AND THE OPERATING CONDITIONS.
DICKSON,C.R.
NUREG/CR-2637: EMERGENCY RESPONSE CAPABILITIES AND EXAMPLE ASSESSMENTS FOR AIRBORNE RADIONUCLIDE DISCHARGES.
DIGGS,B.R.
NUREC/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR AND SHIELDING STUDIES.
DISALVO,R.
NUREG-0863: SURVEY OF FOREIGN REACTOR OPERATOR GUALIFICATIONS, TRAINING.
AND STAFFING REGUIREMENTS.
NUREG-0872: A FEASIBILITY STUDY OF USING LICENSEE EVENT REPORTS FOR A STATISTICAL ASSESSMENT OF THE EFFECT OF OVERTIME AND SHIFT WORK ON OPERATOR ERROR.
DIVINE,J.M.
NUREG/CR-2717: EXPERIMENT DATA REPORT FOR LOFT ANTICIPATED TRANSIENT WITHOUT SCRAM EXPERIMENT L9-3.
DOUGAN,J.R.
NUREG/CR-2362: RELATIONSHIPS BETWEEN CHARPY V-NOTCH IMPACT ENERGY AND FRACTURE TOUGHNESS.
DUNN,L.A.
NUREC/CR-2381: GEOLOGIC AND HYDROLOGIC RESEARCH AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER WEST VALLEY, NEW YORK. Progress Report, August i
1979-July 1981.
ECKERMAN K.F.
I j
NUREC/CR-2612: VARIABILITY IN DOSE ESTIMATES ASSOCI ATED WITH THE FOOD CHAIN TRANSPORT AND INGESTION OF SELECTED R ADIONUCLIDES.
91
ECKERT.R.M.
NUREG/CR-1030 VO2: SEDIMENT AND R ADIONUCLIDE TRANSPORT IN RIVERS. Phase 2-Field Sampling Program For Cattaraugus And Buttermilk Creeks,New York.
ELLINGWOOD.B.
NUREG/CR-2638: SNOW LOADS FOR THE DESIGN OF NUCLEAR POWER PLANT STRUCTURES.
ELRICK.R.M.
NUREG/CR-1820: STATUS REPORT ON THE FISSION-PRODUCT RESEARCH PROGRAM.
ERDMAN,C.A.
NUREC/CR-2603: BUBBLE BEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
FANQUS.F.
NUREC/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRE.;GTH. Tec h nical Repor t. Augus t 1980-September 1981.
FELDE,D.K.
NUREC/CR-2525 VO1: ORNL ROD SUNDLE HEAT TRANSFER TEST DATA. Volume 1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
NUREC/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic. Test Facility Experimental Data Report for lest 3.03.6AR - Transient Film Boiling In Upflow.
NUREG/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.06.6B-Transient Film Boiling In Upflow.
NUREC/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facility Experimental Data Report For Test
- 3. 08. 6C-Trans ient Film Boiling In Upflow.
NUREG/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Series 3. 07. 9-Steady-State Film Boiling In Upflow.
NUREC/CR-2544: TWO-PHASE MASS FLUX UNCERTAINTY ANALYSIS FOR THERMAL-HYDRAULIC TEST FACILITY INSTRUMENTED SPOOL PIECES.
NUREG/CR-2545: DESIGN CONCEPT AND TESTING OF AN IN-BUNDLE GAMMA DENSITOMETER FOR SUBCHANNEL VOID FRACTION MEASUREMENTS IN THE THTF ELECTRICALLY HEATED ROD BUNDLE.
FELKINS,L.
NUREG/CR-2353 VO2: SPECIFICATION AND VERIFICATION OF NUCLEAR POWER PLANT TRAINING SIMULATOR RESPONSE CHARACTERISTICS. Part II:
Conclusions And Recommendations.
FINDLAY,J.A.
NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4.4 - 30 SSTF DESCRIPTION DOCUMENT.
FISHER,F.L.
NUREG/CR-2377: TESTS & CRITERIA FOR FIRE PROTECTION OF CABLE PENETRATIONS.
FOGDALL,S.P.
NUREC/CR-2636: EXPERIMENTAL DATA REPORT FOR AIR-WATER FLOODING TESTS OF THE FLECHT-SEASET PROGRAM SET FACILITY VESSEL UPPER PLENUM.
I FORD,W.E.
NUREC/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR AND SHIELDING STUDIES.
FOX,R.A.
NUREG/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER TESTING.
FRAIZE,W.E.
NUREG/CP-OO26: WORKSHOP ON PSYCHOLOGICAL STRESS ASSOCIATED WITH THE PROPOSED RESTART OF THREE MILE ISLAND, UNIT 1.
FREDERICK,E.J.
NUREC/CR-2184: COMPARISON OF THE RADIOLOGICAL IMPACTS OF THORIUM AND URANIUM NUCLEAR FUEL CYCLES.
92
FREEMAN,D.W.
NUREG/CR-1233 V04: THE STRUCTURED ASSESSMENT APPROACH. VERSION
- 1. COMPUTATIONAL ANALYSIS PACKAGE.
FREY,P.R.
NUREG/CR-2586: A SURVEY OF METHODS FOR IMPROVING OPERATOR ACCEPTANCE OF COMPUTERIZED AIDS.
NUREG/CR-2587: FUNCTIONS AND OPERATIONS OF NUCLEAR POWER PLANT CREWS.
FUGELSO,L.E.
NUREG/CR-2569: RESPONSE OF THE ZION & INDIAN POINT CONTAINMENT BUILDINGS TO SEVERE ACCIDENT PRESSURES.
,FULLWOOD,R.R.
NUREC/CR-1672 V03: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDIA: Tec hnical Review of NUREC/CR-1636, Vols 1,2 and 3, Dec ember 1,1981-March 31,1982.
FURGAL,D.T.
NUREC/CR-2314: AGING WITH RESPECT TO FLAMMABILITY AND OTHER PROPERTIES IN FIRE-RETARDED ETHYLENE PROPYLENE RUBBER AND CHLOROSULFONATED POLYETHYLENE.
NUREG/CR-2559: RESULTS OF PHASE ONE OF PLANT ELECTRICAL SYSTEM (PES)
STUDY.
GAHAN,E.
NUREG/CR-2403 S01: SURVEY OF INSULATION USED IN NUCLEAR POWER PLANTS AND THE POTENTI AL FOR DEBRIS GENERATION.
GALLAGHER,C.A.
NUREC/CR-2692: AN INTEGRATED SYSTEM FOR FORECASTING ELECTRIC ENERGY AND LOAD FOR STATES AND UTILITY SERVICE AREAS.
GALLAHER,R.B.
NUREG/CR-2172:
SUMMARY
AND BIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT BOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2173:
SUMMARY
AND BIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2173 ERR:
SUMMARY
AND BIBLIOGR APHY OF SAFETY-RELATED EVENTS AT PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
GARDNER,R.H.
NUREG/CR-26I2: VARIABILITY IN DOSE ESTIMATES ASSOCIATED WITH THE FOOD CHAIN TRANSPORT AND INGESTION OF SELECTED R ADIONUCLIDES.
GERGELY,P.
NUREG/CR-2788: STRENGTH AND STIFFNESS OF UNIAXI ALLY TENSIONED REINFORCED CONCRETE PANELS SUBJECTED TO MEMBRANE SHEAR.
GERTMAN,D.I.
NUREG/CR-2496: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
GERY,A.
NUREG-0872: A FEASIBILITY STUDY OF USING LICENSEE EVENT REPORTS FOR A t
STATISTICAL ASSESSMENT OF THE EFFECT OF OVERTIME AND SHIFT WORK ON OPERATOR ERROR.
GIDO,R.C.
NUREG/CR-2633: CONTAINMENT REACTOR CAVITY SUBCOMPARTMENT ANALYSIS PROCEDURES FOR A BOILING WATER REACTOR.
GINSBERG T.
NUREG/CR-2543: A STUDY OF THE FEASIBILITY OF MICROWAVE DIELECTRIC HEATING FOR LMFBR TRANSITION PHASE ACCIDENT SEGUENCE BOILING STUDIES.
GINZBERG,T.
NUREC/CR-2542: SENSITIVITY STUDY USING THE FR ANTIC CODE FOR THE UNAVAILABILITY OF A SYSTEM TO THE FAILURE CHARACTERISTICS OF THE COMPONENTS AND THE OPERATING CONDITIONS.
COLD,M.R.
NUREG/CR-2603: BUBBLE BEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
COLUMBEK,J.
NUREG/CR-2220 V02: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH 93
i a
POPULATIONS IN THE HUDSON RIVER ESTUARY.
COODYEAR,C.P.
NUREG/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT DN FISH i
POPULATIONS IN THE HUDSON RIVER ESTUARY.
NUREG/CR-2220 VO3: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT DN FISH l
POPULATIONS IN THE HUDSON RIVER ESTUARY.
GORDON,J.J.
NUREQ/CP-OO26: WORKSHOP DN PSYCHOLOGICAL STRESS ASSOCIATED WITH THE PROPOSED RESTART DF THREE MILE ISLAND. UNIT 1.
GOULD,S.S.
NUREG/CR-2525 VO1: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
NUREG/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for Test 3.03.6AR - Transient Film Boiling In Upflow.
NUREG/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume i
3-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.06.6B-Transient Film Boiling In Upflow.
NUREG/CR-2525 VO5: DRNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.00.6C-Transient Film Boiling In Upflow.
GRANT,F.H.
NUREG/CR-2604: THE SNAP OPERATING SYSTEM (SOS) USER'S GUIDE.
GREGORY,M.V.
NUREG/CR-1681: WRAP-PWR VERIFICATION STUDIES.
NUREG/CR-2356: UPDATED INPUT FOR THE WRAP-EM SYSTEM.
GRECORY,W.S.
NUREG/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED i
TORNADO CONDITIONS.
NUREG/CR-2632: RESPONSE OF CENTRIFUGAL BLOWERS TO SIMULATED TDRNADO TRANSIENTS. July-September 1981.
GREIMANN,L.
NUREC/CR-2442: RELIABILITY ANALYSIS OF STEEL CDNTAINMENT l
STRENGTH. Tech nical Report, Augus t 1980-September 1981.
l GRIMES.B.K.
l NUREG-0049: STANDARD REVIEW PLAN FOR THE REVIEW AND EVALUATION OF EMERGENCY PLANS FOR RESEARCH AND TEST REACTORS.
GROFF,D.W.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY, MISSOURI.
GUDAS,J.P.
NUREG/CR-2570: EXPERIMENTAL INVESTIGATION OF TEARING INSTABILITY PHENOMENA FOR STRUCTURAL MATERIALS.
HAAS,P.M.
NUREG/CR-2353 VO2: SPECIFICATIDN AND VERIFICATION OF NUCLEAR POWER PLANT TRAINING SIMULATOR RESPONSE CHARACTERISTICS. Part II:
Conclusions And Recommendations.
HACK,A.
NUREC/CR-2652: EVALUATION AND PERFORMANCE OF CLOSED-CIRCUIT BREATHING APPARATUS.
HALL.D.C.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User's Guid e And Input Requirements.
HALL,R.E.
NUREG/CR-2416: INITIAL QUANTIFICATION OF HUMAN ERRDR ASSOCIATED WITH SPECIFIC INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS IN LICENSED NUCLEAR POWER PLANTS.
NUREG/CR-2542: SENSITIVITY STUDY USINO THE FR ANTIC CODE FOR THE UNAVAILABILITY OF A EYSTEM TO THE FAILURE CHARACTERISTICS OF THE COMPONENTS AND THE OPERATING CONDITIONS.
l l
1 l
94
HANNER,0.M.
NUREG/CR-2059:
COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WAlhlt HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981 ).
HARDY,J.E.
NUREG/CR-2204 VO4: ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM.Guarterly Progress Report,0ctober-December 1981.
i NUREC/CR-2505:
ELECTRICAL IMPEDANCE STRING PROBES FOR TWO-PHASE VOIDS AND VELOCITY MEASUREMENTS.
HARPER,M.D.
NUREC/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTIAL OR POISSON DATA.
HARRINGTON,R.M.
NUREC/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report,0ctober 1-December 31,1981.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
HARRIS,D.O.
NUREG/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS.
HARRIS.J.R.
NUREC/CR-2638: SNOW LOADS FOR THE DESIGN OF NUCLEAR POWER PLANT STRUCTURES.
HASELTON T.M.
NUREC/CR-2653: EARTH RESISTIVITY AS A TOOL FOR SHALLOW EXPLORATION IN THE REELFOOT LAKE AREA, TENNESSEE.
HATCH,S.W.
NUREG/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS PROGRAM: Calvert Cliffs No. 2 PWR Power Plant.
HATTA,M.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
HAWLEY,S.C.
NUREG/CR-2387: CREDIBLE ACCIDENT ANALYSES FOR TRIGA AND TRICA-FUELED REACTORS.
HECKER,G.E.
NUREG/CR-2760: ASSESSMENT OF SCALE EFFECTS ON VORTEXING. SWIRL, AND INLE1 LOSSES IN LARGE SCALE SUMP MODELS.
HEDRICK,R.A.
NUREC/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMB ER 1980.
HELTON,J.C.
NUREG/CR-1636 VO4: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: EFFECTS OF VARI ABLE HYDROLOGIC PATTERNS ON THE ENVIRONMENTAL TRANSPORT MODEL.
HENNIGAN,J.M.
NUREG/CR-2736: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MICHIGAN. September 1980-August 1981.
i HICKMAN,J.W.
NUREG/CR-23OO VO1 R1: DRAFT:PRA PROCEDURE GUIDE. A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
HILMY,S.I.
l NUREG/CR-2788: STRENGTH AND STIFFNESS OF UNIAXIALLY TENSIONED REINFORCED CONCRETE PANELS SUBJECTED TO MEMBRANE SHEAR.
HINZE,W.J.
NUREG/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH THE 38TH PARALLEL LINEAMENT.
Final Technical Report, June 1979-June 1981.
HITCHCOCK,J.T.
i 95
1 NUREG/CR-2473: SIMMER ANALYSIS OF PROMPT DURST ENERGETICS EXPERIMENlS.
HOAGLAND,K.E.
NUREG/CR-2727 VO1: ECOLOGICAL STUDIES OF WOOD-BORING BIVALVES IN THE VICINITY OF THE OYSTER CREEK NUCLEAR GENERATING STATION. Progress Report. September-November 1981.
HOFFMAN,C.C.
NUREQ/CR-2704:
U. S.
REACTOR SPENT-FUEL STORAGE CAPABILITIES.
HOFFMAN,F.O.
NUREQ/CR-2612: VARI ABILITY IN DOPE ESTIMATES ASSOCIATED WITH THE FOOD CHAIN TRANSPORT AND INGESTION OF SELECTED R ADIONUCLIDES.
HOFFMAN,L.G.
NUREG/CR-2495: CHARACTERIZATION OF SOIL TO PLANT TRANSFER COEFFICIENTS FOR STABLE CESIUM AND STRONTIUM.
NUREG/CR-2644: AN ASSESSMENT OF OFFSITE,REAL-TIME DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS.
HOGAN B.T.
NUREG/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER TESTING.
HOOKER,C.D.
NUREG/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER TESTING.
HORAK H.L.
NUREG/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED TORNADO CONDITIONS.
HORST,T.W.
NUREG/CR-2022: TECHNICAL REVIEW OF THE DISPERSION AND DOSE MODELS USED IN THE MILDOS COMPUTER PROGRAM.
HORTON,K.A.
NUREG/CR-2589: A GROUND-PENETRATING RADAR SURVEY OF THE MAXEY FLATS LOW-LEVEL NUCLEAR WASTE DISPOSAL SITE, FLEMING COUNTY, KENTUCKY.
HOSKER,R.P.
NUREQ/CR-2521: METHOD FOR ESTIMATING WAKE FLOW AND EFFLUENT DISPERSION NEAR SIMPLE BLOCK-LIKE BUILDINGS.
HUNT,D.F.
NUREG/CR-2525 VO1: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
HYLTON,J.O.
NUREG/CR-2505: ELECTRICAL IMPEDANCE STRING P"OBES FOR TWO-PHASE VOIDS AND VELOCITY MEASUREMENTS.
HYMAN,C.R.
NUREG/CR-2455: EXPERIMENTAL INVESTIGATIONS OF BUNDLE BOILOFF AND REFLOOD UNDER HIGH-PRESSURE LOW HEAT FLUX CONDITIONS.
IDAR,E.S.
NUREG/CR-2632: RESPONSE OF CENTRIFUGAL BLOWERS TO SIMULATED TORNADO TRANSIENTS. July-September 1981.
IMAN,R.L.
NUREG/CR-2350: SENSITIVITY ANALYSIS TECHNIGUES: SELF-TEACHING CURRICULUM.
ISHII,M.
NUREG/CR-2647: CRITICAL HEAT FLUX EXPERIMENTS UNDER LOW FLOW CONDITIONS IN A VERTICAL ANNULUS.
JAMES,S.W.
NUREQ/CR-1826 VO1: RELAPS/ MOD 1 CODE MANUAL. Vo lume I: System Models And Numerical Methods.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
JAMISON,J.D.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULATORY BASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE FUEL CYCLE.
96
1 i
1 JENSEN,R.J.
NUREC/CR-2334: INTERPHASE TRANSPORT IN HORIZONTAL STRATIFIED CONCURRENT i
FLOW.
JOHNSON,J.J.
NURE0/CR-2015 VO4: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
JOHNSON,L.G.
NUREQ/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTOR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
JOHNSON,M.N.
NUREO/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER TESTING.
JOHNSON,R.C.
1 NUREC/CP-OO26: WORKSHOP ON PSYCHOLOGICAL STRESS ASSOCIATED WITH THE l
PROPOSED RESTART OF THREE MILE ISLAND, UNIT 1.
JORDAN,H.
NUREQ/CR-2682: CITADEL: A COMPUTER CODE FO'R THE ANALYSIS OF IODINE BEHAVION IN STEAM QENERATOR TUBE RUPTURE ACCIDENTS.
NUREC/CR-2683: IODINE BEHAVIOR IN STEAM QENER ATOR TUBE RUPTURE ACCIDENTS.
JOYCE,J.A.
NURE0/CR-2570: EXPERIMENTAL INVESTIGATION OF TEARING INSTABILITY PHENOMENA FOR STRUCTURAL MATERI ALS.
KATHIRESAN,K.
NURE0/CR-2494: OR-FLAW: A FINITE ELEMENT PROGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFINED FLAWS IN PLATES, CYLINDERS AND PRESSURE-VESSEL N0ZZLE CORNERS.
KATHREN,R.L.
NUREQ/CR-2387: CREDIBLE ACCIDENT ANALYSES FOR TRIGA AND TRIGA-FUELED REACTORS.
NUREQ/CR-2413: SURVEY OF REMOTE AREA MONITORING SYSTEMS AT U. S.
LIGHT-WATER-COOLED POWER REACTORS.
KAUL,D.
NURE0/CR-1672 VO3: RISK ASSESSMENT ETHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDIA: Tec hnical Review of NUREQ/CR-1636, Vols 1,2 and 3, Dec emb er 1,1981-March 31,1982.
KEHLER.P.
NURE9/CR-1622: FLOW MEASUREMENT BY PULSED-NEUTRON ACTIVATION TECHNIGUES AT THE PKL FACILITY AT ERLANGEN (GERMANY).
KELLER,G.R.
NURE0/CR-2741:. A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH TE 38TH PARALLEL LINEAMENT.
Final Technical Report, June 1979-June 1981.
KELLER,J.H.
NURE0/CR-2495: CHARACTERIZATION OF SOIL TO PLANT TRANSFER COEFFICIENTS FOR STABLE CESIUM AND STRONTIUM.
NURE0/CR-2644: AN ASSESSMENT OF OFFSITE, REAL-TIE DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS.
KENT,L.A.
NURE0/CR-2279: WATER RELEASE FROM HEATED CONCRETES.
KERLIN,T.W.
NUREC/CR-2353 VO2: SPECIFICATION AND VERIFICATION OF NUCLEAR POWER PLANT TRAINING SIMULATOR RESPONSE CHARACTERISTICS. Part II:
Conclusions And Recommendations.
KETALAAR,D.
NUREO/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Tech nical Report, Augus t 1980-Sep tember 1981.
KEY,K.J.
NUREG-0868: A COLLECTION OF MATHEMATICAL MODELS FOR DISPERSION IN SURFACE WATER AND OROUNDWATER.
97
KILLOUGH,C.G.
NUREQ/CR-2610: RAGBEEF:A FORTRAN IV IMPLEMENTATION OF A TIME-DEPENDENT MODEL FOR RADIONUCLIDE CONTAMINATION OF BEEF.
KIM,H.J.
NUREG/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATE GEOMETRY.
KIRK,B.L.
NUREC/CR-2220 VO3: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
KISER,D.M.
NUREQ/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Vo lume 1: System Models And Numerical Methods.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
KISNER,R.A.
NUREG/CR-2586: A SURVEY OF METHODS FOR IMPROVING OPERATOR ACCEPTANCE OF COMPUTERIZED AIDS.
NUREG/CR-2587: FUNCTIONS AND OPERATIONS OF NUCLEAR POWER PLANT CREWS.
KOCHER,D.C.
NUREC/CP-0022: PROCEEDINGS OF THE SYMPOSIUM ON UNCERTAINTIES ASSOCIATED WITH THE REGULATION OF THE GEOLOGIC DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTE.
KOLB,G.O.
NUREG/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS PROGRAM: Calvert Clif f s No. 2 PWR Power Plant.
KOLBE.R.
NUREC/CR-2403 SO1: SURVEY OF INSULATION USED IN NUCLEAR POWER PLANTS AND THE POTENTIAL FOR DEBRIS GENERATION.
KOLLIE,T.G.
NUREQ/CR-2470: THERMOMETRY IN THE MULTIROD BURST TEST PROGRAM.
KORNEGAY,F.C.
NUREG/CR-2221 VO4: HIGH-TEMPERATURE GAS.-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report,0ctober 1-December 31,1981.
KRESS,T.S.
NUREQ/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
KUCK,I.Z.
NUREG/CR-2518: THERMODYNAMIC PROPERTIES OF WATER FOR COMPUTER SIMULATION OF POWER PLANTS.
KUEHN,N.H.
NUREC/CR-1681: WRAP-PWR VERIFICATION STUDIES.
KUKIELKA,C.A.
NUREQ/CR-2497 VO1: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1969-1979. A Status Report.Vol.
- 1. Main Report And App. A.C.D And E.
NUREG/CR-2497 VO2: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENT: 1969-1979. A Status Rep ort. Vol. 2 - Appendix B.
KUO,H.H.
NUREG/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Vo lume 1: System Mode ls And Numerical Methods.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
LANNING,D.D.
NUREG/CR-2567: FINAL DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-432.
NUREG/CR-2600: END-OF-IRRADIATION DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-527.
LARSON,R.A.
NUREQ/CR-2618: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A NATURAL 98
CIRCULATION TEST S-NC-7C.
LASSAHN,0.D.
NUREQ/CR-0169 V13: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES.
Volume XIII. Temperature Measurements.
LEE,D.W.
NUREQ/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
LEE L.S.
NUREQ/CR-2229 VO1: BWR LARGE BREAK SIMULATION TESTS--BWR BLOWDOWN / EMERGENCY CORE COOLING PROGR AM.
LEE,W.J.
NUREQ/CR-2704:
U. S.
REACTOR SPENT-FUEL STORAGE CAPABILITIES.
LETTIERI V.
NUREQ/CR-2416: INITIAL GUANTIFICATION OF HUMAN ERROR ASSOCI ATED WITH SPECIFIC INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS IN LICENSED t.'JCLEAR POWER PLANTS.
LI,C.Y.
NUREQ/CR-2633: CONTAINMENT REACTOR CAVITY SUSCOMPARTMENT ANALYSIS PROCEDURES FOR A BOILING WATER REACTOR.
LIDIAK.E.G.
NUREQ/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH THE 38TH PARALLEL LINEAMENT.
Final Technical Report, June 1979-June 1981.
LIETZKE,M.H.
NUREQ/CR-2493: AGUEOUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And Assessment.
LIM,E.Y.
NUREQ/CR-2301: FRACTURE NECHANICS MODELS DEVELOPED FOR PIPIT:0 RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS.
LIN,T.
NUREQ/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Technical Report August 1980-September 1981.
LINDSEY,C.G.
NUREQ/CR-2642: LONG-TERM SURVIVABILITY OF RIPRAP FOR ARMORING URANIUM MILL TAILINGS AND COVERS: A LITERATURE REVIEW.
LIPINSKI R.J.
NUREQ/CR-2412: HEAT REMOVAL FROM A STRATIFIED UO2-SODIUM PARTICLE BED.
LI TTLETON, P.' E.
NUREQ/CR-2632: RESPONSE OF CENTRIFUCAL BLOWERS TO SIMULATED TORNADO TRANSIENTS. July-September 1981.
LONG,J.D.
NUREG-0891: NUCLEAR PROPERTY INSURANCE: STATUS AND OUTLOOK.
LONG,L.W.
NUREQ/CR-2564: ENVIRONMENTAL FACTORE AFFECTING LONG-TERM STABILIZATION OF RADON SUPPRESSION COVERS FOR URANIUM MILL TAILINGS.
NUREC/CR-2642: LONG-TERM SURVIVABILITY OF RIPRAP FOR ARMORING URANIUM MILL TAILINGS AND COVERS: A LITERATURE REVIEW.
LUCKAS,W.J.
NUREC/CR-2416: INITIAL QUANTIFICATION OF HUMAN ERROR ASSOCIATED WITH SPECIFIC INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS IN LICENSED NUCLEAR POWER PLANTS.
NUREQ/CR-2417: IDENTIFICATION AND ANALYSIS OF HUMAN ERRORS UNDERLYING PUMP AND VALVE RELATED EVENTS REPORTED BY NUCLEAR POWER PLANT l
LICENSEES.
i I
LURIE.N.A.
NUREQ/CR-2393 ERR: Errata, changing rept number to NUREQ/CR-2593,to A USER'S MANUAL FOR COMPUTER CODE RIBD/IRT.
NUREQ/CR-2394 ERR: Errata, changing rept number to NUREQ/CR-2594,to A USER 'S MANUAL FOR THE GABAS SPECTRUM COMPUTER CODE.
NUREQ/CR-2593: A USER 'S MANUAL FOR COMPUTER CODE RIBD/IRT.
90
NUREQ/CR-2594: A USER 'S MANUAL FOR THE GABAS SPECTRUM COMPUTER CODE.
LUTTON,R.J.
NUREO/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTE.
MACEDO,P.B.
NUREO/CR-2737: EVALUATION OF BULK PROPERTIES OF RADWASTE OLASS AND CERAMIC CONTAINER MATERIALS TO DETERMINE LONG-TERM STABILITY.
MAECK,M.J.
NURE0/CR-2644: AN ASSESSMENT OF OFFSITE,REAL-TIME DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS.
MAERKER,R.E.
NUREQ/CR-2696: CALCULATIONS OF TWO SERIES OF EXPERIMENTS PERFORMED AT THE POOLSIDE FACILITY USING THE DAK RIDGE RESEARCH REACTOR.
MAKOWITZ,H.
NUREQ/CR-2543: A STUDY OF THE FEASIBILITY OF MICROWAVE DIELECTRIC HEATING FOR LMFBR TRANSITION PHASE ACCIDENT SEQUENCE BOILING STUDIES.
MALONE P.O.
NUREQ/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTE.
MARTIN,R.A.
NUREQ/CR-2632: RESPONSE OF CENTRIFUGAL BLOWERS TO SIMULATED TORNADO TR ANSIENTS. Ju l y-Sep t emb er 1981.
MARTIN,T.T.
NUREG-0909: NRC REPORT ON THE JANUARY 25,1982 STEAM GENERATOR TUBE RUPTURE AT R. E.
GINNA NUCLEAR POWER PLANT.
MARTZ,H.F.
NUREQ/CR-2434: FRAC (FAILURE RATE ANALYSIS CODE): A COMPUTER PROGRAM FOR ANALYSIS OF VARI ANCE OF FAILURE RATES. An Application User 's Guide.
NUREC/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTI AL OR POISSON DATA.
MASENIKOV,0.R.
NUREO/CR-2015 VO4: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
MCBRIDE,J.P.
NUREQ/CR-2184: COMPARISON OF THE RADIOLOGICAL IMPACTS OF THORIUM AND URANIUM NUCLEAR FUEL CYCLES.
MCCARTY,M.J.
NUREG/CR-2736: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MICHICAN. September 1980-August 1981.
MCCORMACK.K.E.
NUREG/CR-2172:
SUMMARY
AND BIBLIOGR APHY OF SAFETY-RELATED EVENTS AT BOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREQ/CR-2173:
SUMMARY
AND BIBLIOGR APHY OF SAFETY-RELATED EVENTS AT.
PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREC/CR-2173 ERR:
SUMMARY
AND BIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
MCDOWELL,0.S.
NUREC/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY, MISSOURI.
MCDOWELL-BOYER NUREC/CR-2610: RAGBEEF: A FORTRAN IV IMPLEMENTATION OF A TIME-DEPENDENT MODEL FOR RADIONUCLIDE CONTAMINATION OF BEEF.
MCINTEER,C.R.
NUREC/CR-2434: FRAC (FAILURE RATE ANALYSIS CODE): A COMPUTER PROGRAM FOR ANALYSIS OF VARIANCE OF FAILURE RATES. An Application User 's Guide.
NUREG/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTI AL OR POISSON DATA.
MCLEOD,K.W.
100
l l
NUREG/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FRDM AGRO-ECOSYSTEMS. Annual Progress Repor t,0cto ber 1980-September 1981.
MEADE,R.B.
NUREG/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTE.
MERSCHOFF,E.
NUREG-0063: SURVEY OF FOREIGN REACTOR OPERATOR GUALIFICATIONS, TRAINING, AND STAFFING REGUIREMENTS.
MEWHINNEY,J.A.
NUREC/CR-2512: RADIATION DOSE ESTIMATES AND HAZARDS EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Annual Progress Report July 1980-June 1981.
MEYER,H.R.
NUREG/CR-2184: COMPARISON OF THE RADIOLOGICAL IMPACTS OF THORIUM AND URANIUM NUCLEAR FUEL CYCLES.
MILLER,0.N.
NUREG/CR-2204 VO4: ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM. Guart erly Progress Report,0ctober-December 1981.
MILLER,R.J.
NUREC/CR-2456: EXPERIMENTAL INVESTIGATIONS OF UNCOVERED-BUNDLE HEAT TRANSFER AND TWO-PHASE MIXTURE-LEVEL SWELL UNDER HIGH-PRESSURE LOW HEAT-FLUX CONDITIONS.
MILLER,R.L.
NUREG/CR-2522: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM PLAN.
MINARICK,J.W.
NUREG/CR-2497 VO1: PRECURSORS TO POTENTI AL SEVERE CORE DAMAGE ACCIDENTS: 1969-1979. A Sta tus Report. Vol.
- 1. Main Report And Ap p.
A,C,D And E.
NUREC/CR-2497 VO2: PRECURSORS TO POTENTI AL SEVERE CORE DAMAGE ACCIDENT: 1969-1979. A Status Rep ort. Vol. 2 - Appendix B.
MISHIMA,K.
NUREC/CR-2647: CRITICAL HEAT FLUX EXPERIMENTS UNDER LOW FLOW CONDITIONS.
IN A VERTICAL ANNULUS.
MITCHELL,G.W.
NUREG/CR-2412: HEAT REMOVAL FROM A STRATIFIED UO2-SODIUM PARTICLE BED.
MORCOS,N.
NUREG/CR-2193 VO1 N2: PROPERTIES OF RADI'OACTIVE WASTES AND WASTE CONTAINERS. Guarterly Progress Report, April-June 1981.
MOREY,R.M.
NUREG/CR-2589: A GROUND-PENETRATING RADAR SURVEY OF THE MAXEY FLATS LOW-LEVEL NUCLEAR WASTE DISPOSAL SITE, FLEMING COUNTY, KENTUCKY.
MORRIS,D.C.
NUREG/CR-2435: DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMP ARISONS.
NUREG/CR-2469: AN ANALYSIS OF TRANSIENT FILM BOILING OF HIGH-PRESSURE WATER IN A ROD BUNDLE.
NUREG/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for lest 3.03.6AR - Transient Film Boiling In Upflow.
NUREG/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hydraulic Test Facility Exp erimen tal Data Rep ort For Test
- 3. 06. 6B-Trans ient Film Boiling In Upflow.
NUREG/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facili ty Exp erimen tal Data Report For Test 3.08.6C-Transient Film Boiling In Upflow.
NUREG/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Th er ma l-Hy d ra u l i c Test Facility Experimen tal Data Support For Test Series 3. 07. 9-St ead y-Sta t e Film Boili ng In Upflow.
MULLINS,C.B.
101
NUREG/CR-2435: DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMP ARISONS.
NUREC/CR-2469: AN ANALYSIS OF TRANSIENT FILM DO! LING OF HIGH-PRESSURE WATER IN A ROD BUNDLE.
NUREC/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for Test 3.03.6AR - Transient Film Boiling In Upflow.
NUPEG/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.06.68-Transient Film Boiling In Upflow.
NUREG/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facility Experimental Data Report For Test
- 3. 08. 6C-Trans ient Film Bo iling In Upflow.
NUREG/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Series 3.07.9-Steady-State Film Boiling In Upflow.
MURALIDHARAN,R.
NUREG/CR-2231: BWR LOW FLOW BUNDLE UNCOVERY TEST AND ANALYSIS.
NELSON,R.A.
NUREG/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Vo lume 1: System Mode ls And Numerical Methods.
NELSON,W.R.
NUREC/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
NICOLOSI.S.L.
NUREG/CR-2713: VAPOR DEPOSITION VELOCITY MEASUREMENTS AND CORRELATIONS FOR I(2) AND C sI.
NIEMCZYK,S.J.
NUREG/CR-2629: INTERIM SOURCE TERM ASSUMPTIONS FOR EMERGENCY PLANNING AND EQUIPMENT GUALIFICATION.
NYERGES,P.L.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY, MISSOURI.
ODOM,J.P.
NUREG/CR-1851: REACTOR PHYSICS DESIGN CALCULATIONS FOR THE ACPR UPGRADE.
ONISHI,Y.
NUREG/CR-1030 VO2: SEDIMENT AND R ADIONUCLIDE TRANSPORT IN RIVERS. Phase 2-Field Sampling Program For Cattaraugus And Buttermilk Creeks,New York.
OTT,L.J.
NUREG/CR-2435: DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMP ARISONS.
PAASCH,R.A.
i NUREC/CR-2522: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM PLAN.
PADMANABHAN,M.
NUREG/CR-2760: ASSESSMENT OF SCALE EFFECTS ON VORTEXING, SWIRL, AND INLET LOSSES IN LARGE SCALE SUMP MODELS.
NUREG/CR-2772: HYDRAULIC PERFORMANCE OF PUMP SUCTION INLETS FOR EMERGENCY CORE COOLING SYSTEMS IN BOILING WATER REACTORS.
PALMER,R.A.
NUREG/CR-2493: AGUEDUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And l
Assessment.
PANG,E.C.
NUREG/CR-2790: AUTOMOBILE IMPACT FORCES ON CONCRETE WALL PANELS.
(
PANGBURN,0.
NUREG-0902: SITE SUITABILITY, SELECTION AND CHARACTERIZATION BRANCH TECHNICAL POSITION - Low Level Waste Licensing Branch.
PANISKO,F.E.
102
NUREG/CR-2432: A UNIGUE CONCEPT FOR LIGUID LEVEL AND VOID FRACTION DETECTION IN SEVERE FUEL DAMAGE TESTS.
PARKS,P.B.
NUREQ/CR-1681: WRAP-PWR VERIFICATION STUDIES.
PARZIALE,A.A.
NUREG/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION 1, COMPUTATIONAL ANALYSIS PACKAGE.
PATENAUDE,C.J.
NUREG/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION 1, COMPUTATIONAL ANALYSIS PACKAGE.
PATRICK,D.M.
NUREC/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTE.
PECK,S.I.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOU 10 COUNTY, MISSOURI.
PELOGUIN,R.A.
NUREG/CR-2201: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1978.
PENNIFILL,R.
NUREG-0902: SITE SUITABILITY, SELECTION AND CHARACTERIZATION BRANCH TECHNICAL POSITION - Low Level Waste Licensing Branch.
PETERSEN R.J.
NUREC/CR-2496: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
NUREG/CR-2711: PERFORMANCE AND DESIGN REQUIREMENTS FOR A GRAPHICS DISPLAY RESEARCH FACILITY.
PETRIE L.M.
NUREG/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR AND SHIELDING STUDIES.
PHILIPPACOPOULO NUREG/CR-1890: ABS, SRSS AND CDF RESPONSE COMB INATION EVALUATION FOR MARK III CONTAINMENT AND DRYWELL STRUCTURES.
NUREG/CR-2039: DYNAMIC COMBINATIONS FOR MARK II CONTAINMENT STRUCTURES.
NUREC/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
NUREG/CR-2686: REVIEW OF LOAD COMBINATIONS FOR NSSS AND BOP PIP ING AND EQUIPMENT OF MARK III PLANTS.
PICKARD,P.S.
NUREG/CR-1851: REACTOR PHYSICS DESIGN CALCULATIONS FOR THE ACPR UPGRADE.
PIETRZAK R.F.
NUREG/CR-2192 VO1 N2: EVALUATION OF ISOTOPE MIGRATION-LAND BURIAL.Guarterly Progress Rep or t, April-June 1981.
PILCH,M.
NUREG/CR-2611: MGO AND 70 W% UO2-30W% Y203: THERMOPHYSICAL AND TRANSIENl PROPERTIES.
PINDER,J.E.
NUREG/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FROM AGRO-ECOSYSTEMS. Annual Progress R ep or t,0c to b er 19PO-September 1901.
PITTMAN,J.
NUREG-0872: A FEASIBILITY STUDY OF USING LICENSEE EVENT REPORTS FOR A STATISTICAL ASSESSMENT OF THE EFFECT OF OVERTIME AND SHIFT WORK ON OPERATOR ERROR.
PLEASANT,J.C.
NUREG/CR-2610: RAGBEEF: A FORTRAN IV IMPLEMENTATION OF A TIME-DEPENDENT MODEL FOR RADIONUCLIDE CONTAMINATION OF DEEF.
POLITO J.
NUREG/CR-1245 RO1: CORRECTIONS AND ADDITIONS TO USER'S GUIDE FOR SNAP.
( NUREG/ CR-124 5, SANDBO-0315 ).
103
- s NOREG/CR-2604: THE SNAP OPERATING SYSTEM (SOS) USER'S GUIDE.
NUREG/CR-2605: THE SNAP OPERATING SYSTEM REFERENCE MANUAL.
POTTER.S.M.
NUREG/CR-2381: GEOLOGIC AND HYDROLOGIC RESEARCH AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER, WEST VALLEY, NEW YORK. Progress Report, August 1979-July 1981.
POWERS,D.A.
NUREG/CR-2283: DIRECT OBSERVATION OF MELT BEHAVIOR DURING HIGH TEMPERATURE MELT / CONCRETE INTER ACTIONS.
RAGHURAM,S.
NUREC/CR-2682: CITADEL: A COMPUTER CODE FOR THE ANALYSIS OF IODINE BEHAVIOR IN STEAM CENERATOR TUBE RUPTURE ACCIDENTS.
NUREG/CR-2683: IODINE BEHAVIOR IN STEAM GENER ATOR TUBE RUPTURE ACCIDENTS.
RAMOS,S.L.
NUREG-0849: STANDARD REVIEW PLAN FOR THE REVIEW AND EVALUATION OF EMERGENCY PLANS FOR RESEARCH AND TEST REACTORS.
RANSON,V.H.
NUREG/CR-1826 VO1: RELAPS/ MOD 1 CODE MANUAL. Vo lume 1: System Models And Numerical Methods.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
REED,D.A.
NUREG/CR-2435; DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMP ARISONS.
REED,R.L.
NUREG/CR-2356: UPDATED INPUT FOR THE WR AP-EM SYSTEM.
REICH,M.
NUREG/CR-2039: DYNAMIC COMBINATIONS FOR MARK II CONTAINMENT STRUCTUREG.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
NUREG/CR-2686: REVIEW OF LOAD COMBINATIONS FOR NSSS AND BOP PIPING AND EQUIPMENT OF MARK III PLANTS.
REIS,J.W.
NUREC/CR-2564: ENVIRONMENTAL FACTORS AFFECTING LONG-TERM STABILIZATION OF RADON SUPPRESSION COVERS FOR URANIUM MILL TAILINGS.
RENARD P.A.
NUREG/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION 1, COMPUTATIONAL ANALYSIS PACKAGE.
REYNOLDS,A.B.
NUREG/CR-2603: BUBBLE BEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
RICHARDSON,J.M.
NUREG/CR-2551: RANK ORDERING OF VITAL AREAS WITHIN NUCLEAR POWER PLANTS.
RICKETTS,C.I.
NUREG/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED TORNADO CONDITIONS.
ROBINSON J.J.
NUREG/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for Test 3.03.6AR - Transient Film Boiling In Upflow.
NUREG/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.06.6B-Transient Film Boiling In Upflow.
NUREG/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.08.6C-Transient Film Boiling In Upflow.
NUREG/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Series 3.07.9-Steady-State Film Boiling In Upflow.
ROCHELLE,J.M.
104
l i
l NUREG/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
ROGERS,S.C.
NUREC/CR-2204 VO4: ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM.Guarterly Progress Report,0ctober-December 1981.
ROGERS,W.B.
NUREG/CR-2381: GEOLOGIC AND HYDROLOGIC RESEARCH AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER. WEST VALLEY, NEW YORK. Progress Report, August 1979-July 1981.
ROSS,D.J.
NUREG/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION 1, COMPUTATIONAL ANALYSIS PACKAGE.
ROUNTREE,S.L.K.
NUREG/CR-2588: SECURITY OFFICER RESPONSE STRATEGIES (SECURORS).
SABUDA,J.D.
NUREG/CR-2604: THE SNAP OPERATING SYSTEM (SOS) USER 'S GUIDE.
NUREG/CR-2605: THE SNAP OPERATING SYSTEM REFERENCE MANUAL.
SACKETT,K.E.
NUREG/CR-2648: EXPERIMENTAL DATA REPORT FOR SEMISCALE MOD-2A NATURAL CIRCULATION TEST SERIES ( TESTS S-NC-8B AND S-NC-9).
NUREG/CR-2732: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES.(Tests S-IB-1 And S-IB-2).
SACKS,I.J.
NUREG/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION 1 COMPUTATIONAL ANALYSIS PACKAGE.
SALAZAR,E.A.
NUREG/CR-2314: AGING WITH RESPECT TO FLAMMABILITY AND OTHER PROPERTIES IN FIRE-RETARDED ETHYLENE PROPYLENE RUBBER AND CHLOROSULFONATED POLYETHYLENE.
SALLACK,R.A.
NUREC/CR-1820: STATUS REPORT ON THE FISSION-PRODUCT RESEARCH PROGRAM.
SANDERS,J.P.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
SAVOLAINEN,A.
NUREG-0304 VO6 NO4: REGULATORY AND TECHNICAL REPORTS. Compilation For 1981.
NUREG-0304 VO7 NO1: REGULATORY AND TECHNICAL REPORTS. Compilation For First Guarter 1982.
SCHMIDT,W.H.
NUREG/CR-2431: BURN MODE ANALYSIS OF HORIZONTAL CABLE TRAY FIRES.
SCHRIEBER,R.E.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULATORY BASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE FUEL CYCLE.
SCHUMACHER,D.G.
NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4. 4 - 30 SSTF DESCRIPTION DOCUMENT.
SCHURMAN,D.L.
NUREG/CR-2297: SECURITY MANAGEMENT TECHNIQUES AND EVALUATIVE CHECKLISTS FOR SECURITY FORCE EFFECTIVENESS.
SCHWAB,J.D.
NUREG/CR-2201 : POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1978.
SCHWARZ,M.L.
NUREG/CR-2412: HEAT REMOVAL FROM A STRATIFIED UO2-SODIUM PARTICLE DED.
SCHWINKENDORF NUREC/CR-2525 VO5: ORNL ROD DUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facility Exp erimen tal Data Report For Test
- 3. 08. 6C-Trans ient Film Boiling In Upflow.
106
E SCICCA.F.W.
NUREC/CR-2194: CONTAINMENT RESEARCH PRIORITIES.
SEELY,D.S.
NUREG/CR-2231: BWR LOW FLOW BUNDLE UNCOVERY TEST AND ANALYSIS.
SELBY,D.L.
NUREG/CR-2353 VO2: SPECIFICATION AND VERIFICATION OF NUCLEAR POWER PLANT TRAINING SIMULATOR RESPONSE CHARACTER ISTICS. Part II:
Conclusions And Recommendations.
SEXTON,J.L.
NUREG/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH THE 38TH PARALLEL LINEAMENT.
Final Technical Report, June 1979-June 1981.
SHOOMAN,M.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
SIEFKEN,D.
NUREG-0902: SITE SUITABILITY, SELECTION AND CHARACTERIZATION BRANCH TECHNICAL POSITION - Low Level Waste Licensing Branch.
SIMAN-TOV,I.I.
NUREC/CR-2053: HEAT TRANSFER ANALYSTS OF THE LWR PRESSURE VESSEL STEEL IRRADI ATION C APSULES IN THE DAK R ADGE RESEARCH REACTO9-PRESSURE i
VESSEL BENCHM ARK FACILITY.
SLOBODIEN,M.J.
NUREG-0837 VO1 NO1-2: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progr ess Repor t, January-June 1981.
SMITH,A.M.
1 NUREG/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
SMITH,M.H.
NUREG/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FROM AGRO-ECOSYSTEMS. Annual Pr ogres s Repor t,0c to b er 1980-Sep tember 1981.
1 SMITH,P.R.
NUREC/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED i
TORNADO CONDITIONS.
SMITH,R.L.
NUREC/CR-2711: PERFORMANCE AND DESIGN REGUIREMENTS FOR A GRAPHICS s DISPLAY RESEARCH FACILITY.
i SOLDAT,J.K.
j NUREC/CR-2022: TECHNICAL REVIEW OF THE DISPERSION AND DOSE MODELS USED j
IN THE MILDOS COMPUTER PROGRAM.
SOO. P.
4 NUREG/CR-2317 VO1 N3: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATERIALS.Guarterly Progress Report. July-September 1981.
NUREG/CR-2317 VO1 N4: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATER I ALS. Guart er ly P rogre s s Rep o rt,0c tob er-Dec emb e r 1981.
l SOZZI,G.L.
NUREC/CR-2229 VO1: BWR LARGE BREAK SIMULATION TESTS--BWR BLOWDOWN / EMERGENCY CORE COOLING PROGR AM.
SPEAKER,D.M.
NUREG/CR-2417: IDENTIFICATION AND ANALYSIS OF HUMAN ERRORS UNDERLYING PUMP AND VALVE RELATED EVENTS REPORTED BY NUCLEAR POWER PLANT LICENSEES.
l SPULD,S.S.
NUREG/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hyd raulic Test Facili ty Exp erimen tal Data Support For Test Series 3. 07. 9-Steady-State Film Boiling In Upflow.
STAPLES,B.A.
NUREG/CR-2644: AN ASSESSMENT OF OFFSITE, REAL-TIME DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS.
STARMER,R.J.
NUREG-0902: SITE SUITABILITY, SELECTION AND CHARACTERIZATION DRANCH 10B
TECHNICAL POSITION - Low Level Waste Licensing Branch.
START,G.E.
NUREG/CR-2637: EMERGENCY RESPONSE CAPABILITIES AND EXAMPLE ASSESSMENTS FOR AIRBORNE RADIONUCLIDE DISCHARGES.
STEARNS.R.G.
l NUREQ/CR-2653: EARTH RESISTIVITY AS A TOOL FOR SHALLOW EXPLORATION IN THE REELFOOT LAKE AREA. TENNESSEE.
STEVENS,C.A.
NUREG/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDIA: Technical Review of NUREG/CR-1636, Vols 1,2 and 3, December 1,1981-March 31,1982.
STEVENSON,J.A.
l NUREG/CR-2099: COMMON CAUSE FAULT RATES FOR DIESEL GENERATORS: ESTIMATES BASED ON LICENSEE EVENT REPORTS AT U. S.
COMMERCIAL NUCLEAR POWER PLANTS, 1976-1978.
STEVENSON,J.D.
NUREG/CR-2664: SELECTED REVIEW OF FOREIGN LICENSING PRACTICES FOR NUCLEAR POWER PLANTS.
STEVENSON,M.G.
NUREC/CR-2281 VO2: NUCLEAR REACTOR SAFETY. Apr il 1-June 30,1981.
NUREQ/CR-2281 VO3: NUCLEAR REACTOR SAFETY. July 1-September 30,1981.
STUCKWISCH S.E.
NUREG/CR-2343: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: THE DNET COMPUTER CODE USER'S MANUAL.
SUBUDHI,M.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
SUDO, Y.
NUREC/CR-2622: ANALYSIS OF TRAC AND SCTF RESULTS FOR SYSTEM PRESSURE-EFFECTS TESTS UNDER FORCED FLOODING (RUNS 506,507 AND 508).
SUTTER.S.L.
NUREG/CR-2651: ACCIDENT GENERATED PARTICULATE MATERIALS AND THEIR CHARACTERISITICS--A REVIEW OF BACKGROUND INFORMATION.
SUTTON,A.C.
NUREC/CR-2525 VO1: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume l
1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
NUREG/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for Test 3.03.6AR - Transient Film Boiling In Upflow.
NUREG/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.06.6B-Transient Film Boiling In Upflow.
NUREG/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hydraulic Test Facility Experimental Data Report For Test 3.08.6C-Transient Film Boiling In Upflow.
NUREC/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Series 3.07.9-Steady-State Film Boiling In Upflow.
SWYLER,K.J.
NUREC/CR-2516 V01 N1: CHARACTERIZATION OF TMI-TYPE WASTES AND SOLID PRODUCTS. Guarterly Progress Report April-September 1981.
TANKIN,R.S.
NUREC/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATE GEOMETRY.
TAYLOR,0.E.
NUREC/CR-2359: ATMOSPHERIC STRUCTURE PRIOR TO TORNADOES AS DERIVED FROM PROXIMITY AND PRECEDENT UPPER AIR SOUNDINGS.
TEPEL,R.C.
NUREG/CR-2692: AN INTEGRATED SYSTEM FOR FOREC ASTING ELECTRIC ENERGY AND LOAD FOR STATES AND UTILITY SERVICE AREAS.
THAYER,D.D.
in
l
- s..,
NUREQ/CR-2393 ERR: Errata, changing rept number to NUREQ/CR-2593, to A USER'S MANUAL FOR COMPUTER CODE RIBD/IRT.
NUREQ/CR-2394 ERR: Errata, c hanging rep t nuenber to NUREQ/CR'-2594, to A
~
USER 'S MANUAL FOR THE QAB AS SPECTRUM COffUTER CODE.
NUREC/CR-2593: A USER 'S MANUAL FOR COMPUTER CODE RIBD/IRT.
NUREQ/CR-2594: A USER 'S MANUAL FOR THZ GABAS SPECTRUM COMPUTER CODE.
THOMAS,F.A.
w NUREQ/CR-2664: SELECTED REVIEW DF FOREIGN ' LICENSING PRACTICES FOR NUCLEAR POWER PLANTS.
THGMAS,J.T.
NUREC/CR-2223: AN EVALUATION OF THE SOLID ANGLE METHOD USED IN NUCLEAR CRITICALITY SAFETY.
THOME,F.V.
NUREQ/CR-2582: RADIATION CAPABILITIES OF THE SANCIA HIGH INTENSITY ADJUSTABLE COBALT ARRAY.
THOMPSON,M.S.
i NUPEQ/CR-2525 VO1: ORNL ROD BUNDLE HEAT TRANSFER TEST-DATA. Volume 1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
THOMPSON,S.R.
NUREQ/CR-2417: IDENTIFICATION AND ANALYSIS OF HUMAN ERRORS UNDERLYING PUMP AND VALVE RELATED EVENTS REPORTED BY: NUCLEAR POWER PLANT LICENSEES.
THOMPSON,T.
NUREG-0837 VO1 NO3: NRC TLD DIAECT RADI ATION MONITORING NETWORK. Progr ess Report, July-December 1981. '
NUREG-0837 VO1 NO4: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progress Report, October-December 1981.
TILLITT,D.N.
NUREQ/CR-2711: PERFORMANCE AND DESIGN REGUIREMENTS FOR A QRAPHICS DISPLAY RESEARCH FACILITY.
TOBIAS.M.L.
NUREC/CR-2299 VO4: AEROSOL RELEASE AND TRANSPORT PROGRAM. Guarterly Progress Report For October-December 1981.
l TOKARZ,R.D.
j NUREQ/CR-2432: A UNIGUE CONCEPT FOR LIGUID LEVEL AND VOID FRACTION l
DETECTION IN SEVERE FUEL DAMAGE TESTS.
TOTH,L.M.
NUREQ/CR-2493: AGUEDUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And I
Assessment.
TRAPP,J.A.
NUREQ/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Vo lume 1: System Models And Numerical Methods.
NUREQ/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo luma 2: User's Guid e And Input Requirements.
TRIMBLE,J.L.
NUREQ/CR-2692: AN INTEGRATED SYSTEM FOR FORECASTING ELECTRIC ENERQY AND LOAD FOR STATES AND UTILITY SERVICE AREAG.
TROTT.E.A.
NUREQ/CR-2591: ESTIMATING THE POTENTIAL INDUSTRIAL INFACTS OF A NUCLEAR REACTOR ACCIDENT.
TRUITT,J.O.
NUREQ/CR-2297: SECURITY MANAGEMENT TECHNIGUES AND EVALUATIVE CHECKLISTS FOR SECURITY FORCE EFFECTIVENESS.
TRUJILLO,A.
NUREQ/CR-2652: EVALUATION AND PERFCRMANCE OF CLOSED-CIRCUIT BREATHING APPARATUS.
TSAY,J.
NUREQ/CR-2653: EARTH RESISTIVITY AS A TOOL FOR SHALLOW EXPLORATION IN THE REELFOOT LAKE AREA TENNESSEE.
TURK W.V.
108 l
NUREG/CR-2633: CONTAINMENT REACTOR CAVITY SUDCOMPARTMENT ANALYS15 PROCEDURES FOR A DOILING WATER REACTOR.
VAN DER HOVEN NUREG/CR-2584: METEROLOGICAL CONSIDERATIOhS IN THE DEVELOPMENT OF A REAL-TIME ATMOSPHERIC DISPERSION MODEL FOR REACTOR EFFLUENT EXPOSURE PATHWAY.
VAN WINKLE,W.
NUREG/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
VASSILAROS,M.G.
NUREG/CR-2570: EXPERIMENTAL INVESTIGATION OF TEARING INSTABILITY PHENOMENA FOR STRUCTUR AL MATERI ALS.
WAGNER,R.J.
NUREG/CR-1826 VO1: RELAPS/ MOD 1 CODE MANUAL. Vo lume 1: System Models Ai.d Numerical Methods.
NUREG/CR-1826 VO2: RELAPS/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
WALKER,J.L.
NUREG/CR-2604: THE SNAP OPERATING SYSTEM (SOS) USER'S GUIDE.
NUREG/CR-2605: THE SNAP OPERATING SYSTEM REFERENCE MANUAL.
WALKER,P.
NUREG/CP-OO26: WORKSHOP DN PSYCHOLOGICAL STRESS ASSOCIATED WITH THE PROPOSED RESTART OF THREE MILE ISLAND, UNIT 1.
WALTERS,W.H.
NUREG/CR-1030 VO2: SEDIMENT AND RADIONUCLIDE TRANSPORT IN RIVERS. Phase 2-Field Sampling Program For Cattaraugus And Buttermilk Creeks,New Yor k.
WANG,P.C.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
NUREG/CR-2686: REVIEW OF LOAD COMBINATIONS FOR NSSS AND BOP PIPING AND EQUIPMENT OF MARK III PLANTS.
WANG,Y.K.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
WEBSTER,C.C.
NUREG/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR AND SHIELDING STUDIES.
WEISS,A.J.
NUREC/CR-2192 VO1 N2: EVALUATION OF ISOTOPE MIGRATION-LAND BURIAL. Quarte rly Progress Repor t, April-June 1981.
NUREG/CR-2193 VO1 N2: PROPERTIES OF RADIOACTIVE WASTES AND WASTE CONTAINERS. Quarterly Progress Report, April-June 1981.
NUREC/CR-2516 VO1 N1: CHAR ACTERIZ ATION OF TMI-TYPE WASTES AND SOL ID PRODUCTS. Gua rterly Progress Report. April-Sep temb er 1981.
WELLS,M.E.
NUREG/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
WESTFALL,R.M.
NUREC/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR AND SHIELDING STUDIES.
WHELAN,G.
NUREG-0868: A COLLECTION OF MATHEMATICAL MODELS FOR DISPERSION IN SURFACE WATER AND GROUNDWATER.
WHITE,M.D.
NUREG/CR-2455: EXPERIMENTAL INVESTIGATIONS OF BUNDLE BOILOFF AND REFLOOD UNDER HIGH-PRESSURE LOW HEAT FLUX CONDITIONS.
NUREC/CR-2456: EXPERIMENTAL INVESTIGATIONS OF UNCOVERED-BUNDLE HEAT TRANSFER AND TWO-PHASE MIXTURE-LEVEL SWELL UNDER HIGH-PRESSURE LOW HEAT-FLUX CONDITIONS.
109 l
WHITE,R.N.
NUREG/CR-2788: STRENGTH AND STIFFNESS OF UNIAXI ALLY TENSIONED REINFORCED CONCRETE PANELS SUBJECTED TO MEMBRANE SHEAR.
WHITMAN,G.D.
NUREC/CR-2141 VO4: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM.Guarterly Progress Report For October-December 1981.
WILLIAMS M.L.
NUREG/CR-2696: CALCULATIONS OF TWO SERIES OF EXPERIMENTS PERFORMED AT THE POOLSIDE FACILITY USING THE OAK RIDGE RESEARCH REACTOR.
WILLIAMSON,R.B.
NUREC/CR-2377: TESTS & CRITERIA FOR FIRE PROTECTION OF CABLE PENETRATIONS.
WITHERSPOON,J.
NUREG/CR-2184: COMPARISON OF THE RADIOLOGICAL IMPACTS OF THORIUM AND URANIUM NUCLEAR FUEL CYCLES.
WOLD-TINSAE.A.
NUREG/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Tech nical Report, Augus t 1980-September 1981.
WOO,H.H.
NUREG/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS.
WOOTON,R.D.
NUREG/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS PROGRAM: Calvert Cliffs No. 2 PWR Power Plant.
WRIGHT,A.L.
NUREC/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
WYANT,F.J.
NUREG/CR-2559: RESULTS OF PHASE ONE OF PLANT ELECTRICAL SYSTEM (PES)
STUDY.
NUREG/CR-2581: SOME EFFECTS OF ELECTRONS SLOWING DOWN IN MATERI ALS WITH APPLICATION TO SAFETY-RELATED EQUIPMENT GUALIFICATION.
YODER,G.L.
NUREC/CR-2435: DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMPARISONS.
NUREG/CR-2469: AN ANALYSIS OF TRANSIENT FILM BOILING OF HIGH-PRESSURE WATER IN A ROD DUNDLE.
YOUNG,J.K.
NUREG/CR-2564: ENVIRONMENTAL FACTORS AFFECTING LONG-TERM STABILIZATION OF RADON SUPPRESSION COVERS FOR URANIUM MILL TAILINGS.
YOUNGER,J.M.
NUREG/CR-2591: ESTIMATING THE POTENTI AL INDUSTRIAL IMPACTS OF A NUCLEAR REACTOR ACCIDENT.
YUEN,M.C.
NUdEG/CR-2334: INTERPHASE TRANSPORT IN HORIZONTAL STRATIFIED CONCURRENT FLOW.
NUREG/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATE GEOMETRY.
ZABRISKIE,W.L.
NUREG/CR-2204 VO4: ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM.Guarterly Progress Report,0ctober-December 1981.
110
Subject index This index was developed from keywords and word strings in titles and ab-stracts. During this development period, there will be some redundancy, which will be removed later when a reasonable thesaurus has been developed through experience. Suggestions for improvements are welcome.
ACPR.
.NURE0/CR-1851 ALARA.
.NUREG/CR-252P ANSI /ANS 15.16.
.NUREG-0849 Abnormal Occurrence.
.NUREG-0090 VO4 N04 Abstracts.
.NUREG-0304 VO3 Acceptance Criteria.
.NUREC/CR-2460 Accident Delineation Study.
.NUREG/CR-2601 Accident Generated earticulate Materials.
.NUREG/CR-2651 Accident Sequence.
.NUREG/CR-2543 Accident Sequences.
.NUREG/CR-2434 Accident Sequences.
.NUREC/CR-2464 Accident Source Terms.
.NUREG/CR-26P9 Accident.
.NUREC/CR-2591 Accident.
.NUREC/CR-241P Accident.
.NUREG/CR-2493 Accidents.
.NUREG/CR-2525 VOS Accidents.
.NUREG/CR-2194 Accidents.
.NUREG/CR-2525 V07 Advance Notification of Shipments.
.NUREG-0923 Adversery Intrusion Detection.
.NUREG/CR-2588 Aerosol Behavior.
.NUREG/CR-26bl Aerosol Release and Transport.
.NUREG/CR-2393 Aerosol Release and Transport.
.NUREG/CR-2603 Aerosol Release.
. NUREG/CR-2299 VO4 Aerosols.
.NUREG/CR-2299 VO4 Air Guality Problems.
.NUREG/CR-25P1 Airborne Pathways.
.NUREG/CR-2201 Airborne Radionuc1ide Discharges.
.NUREG/CR-2637 Airborne Radionuclides.
.NUREG/CR-2512 Analysis of the Reliability of a Complex System.
.NUREC/CR-2542 Annual Fastest Mile Wind Data.
.NUREG/CR-2639 Annular Core Pulsed Reactor.
. NUREG/CR-19 51 Anticipated Transient Without Scram.
.NUREG/CR-2717 Applications Program.
.NUREG/CR-1659 V03 Aqua Book.
.NUREG-0606 V04 NOP Argonaut-Type Research Reactor.
.NUREG-0913 At-Reactor Storage Pools.
.NUREG/CR-2704 Atmospheric Dispersion Model.
.NUREG/CR-2584 Atmospheric Dispersion.
.NUREG/CR-2022 Atmospheric Release.
.NUREG/CR-2504 111 1
Atmospheric Structure.
.NUREG/CR-2359 Atmospheric Transport and Diffusion.
.NUREG/CR-2504 Automation.
.NUREG/CR-2711 BATLE Statistics.
.NUREG/CR-1245 ROI BDHT.
.NUREG/CR-2525 VO7 BDHT.
.NUREG/CR-2525 VOS BDHT.
.NUREG/CR-2525 VO3 BSTAT.
.NUREG/CR-1245 401 Barnes Layer Profiling.
.NUREG/CR-2653 Barriers.
.NUREG/CR-2508 Basalt Concretes.
)
Bedded Salt Formations.
.NUREG/CR-2343 Beef Cattle Management.
.NUREG/CR-2610 Beltline Materials.
.NUREG/CR-2362 Bibliography of Sa f ety-Related Events.
.NUREG/CR-2173 Blockage of Coolan t Flow.
.NUREG-OO90 VO4 NO4
(
Blowdown Heat Transfer.
.NUREG/CR-2525 VO7 Blowdown Heat Transfer.
.NUREG/CR-2525 VO3 Blowdown Heat Transfer.
.NUREG/CR-2525 VOS Blue Book.
.NUREG-0580 V11 NOS Breathing Apparatus.
.NUREG/CR-2652 Breeders.
.NUREC/CR-1594 VO4 Brems-Strahlung Environments.
.NUREG/CR-2501 Broadleaf Crops.
.NUREG/CR-2625 Bubble Behavior.
Bundle Boiloff and Reflood Tests.
.NUREG/CR-2455 Bunale Heat Transfer.
.NUREC/CR-2231 Burial Trenches.
.NUREC/CR-23H1 l
l Burial Trenches.
CESP.
CINDER.
CITADEL.
. NUREC/CR-26 82 CLASSI Family of Computer Programs.
.NUREG/CR-2015 VO4 COMPARE.
.NUREG/CR-2633 CORRAL.
. NUREG/CR-16 59 VO3 CRBR.
.NUREG/CR-2238 VO1 CRBRP Site.
.NUREG-0786 CRBRP.
.NUREG-0786 CRBRP.
.NUREG/CR-2681 Cable Penetrations.
.NUREC/CR-2377 Cab le Tray Fire Sa f ety.
. NUREG/CR-24 31 Cathode Ray Tube-Generated Displays.
.NUREG/CR-2496 Cavity Sub-Compartment Analysis.
.NUREG/CR-2633 Centrifugal Blower.
.NUREG/CR-2632 Cladding Deformation and Rupture.
.NUREG/CR-2597 Class 9 Accident.
.NUREG/CR-1594 VO4 Class 9 Accident.
.NUREG/CR-2238 VOJ Cleanup Costs.
.NUREG-0891 C1eanup Procedures.
.NUREG/CR-2516 VO! NO1 -03 J
Clinch River Breed er Reactor.
.NUREG-0786 Clinch River Breed er Reactor.
.NUREG/CR-2238 VO!
Clinch River Breed er Reactor.
.NUREG/CR-2681 Closed-Circuit Breathing Apparatus.
. NUREG/CR-26 52 Cognitive Fidelity.
.NUREG/CR-2496 Combustible Gas in Containment.
.NUREG/CR-2481 Common Cause Fault Rates.
.NUREG/CR-2099 Compilation.
.NUREG-0304 VO7 NO1 Component Failure Data Bases.
.NUREG/CR-1659 VO3 Component Failures.
.NUREC/CR-2559 112
Component Unavailabilities..
..NUREC/CR-2464 Computer-Based Operator-Support Concepts.
.NUREC/CR-2711 Computerized Aids.
.NUREG/CR-2586 Computerred Graphic Displays.
. NUREG/CR-2711 Concrete Containment Vessel.
.NUREG/CR-2788 Concrete.
.NUREG/CR-2283 Condensation Heat Transfer.
.NUREG/CR-2783 Confirmatory Research.,
..NUREG/CR-2238 VO2 Construction Permi ts.
. NUREG-0580 Vil NO1-4 Construc tion Permi ts.
. NUREG-0580 Vil NOS Cons truc tion Permi ts..
.NUREG-OO30 VO6 NOI Construction Status.
.NUREG-OO30 VO5 NO4 Construction Status.
.NUREG-OO30 VO6 NO2 Construction..
.NUREG-0848 Construction.
.NUREC-0878 Construction.
.NUREG-0895 Container Materials.
.NUREG/CR-2737 Containment Emergency Sump Performance.
. NUREG/CR-24 81 Containment Emergency Sumps.
.NUREG/CR-2760 Containment Vessels.
.NUREG/CR-2442 Containments.
.NUREG/CR-2569 Control Room Opera ting Staf f.
.NUREG/CP-OO31 VOJ Control Room Opera ting Staf f.
.NUREG/CP-OO31 VO2 Oontrol Room.
.NUREG/CR-2586 Converters..
.NtJREG/CR-1594 VO4 Cooling Towers.
.NUREG/CR-2220 VO2 Core Damage...
Core Debris.
.NUREC/CR-1594 VO4
..NUREG/CR-1594 VO4 Core Disruptive Accidents.
Core Retention.
..NUREG/CR-2281 VO3
. NUREG/CR-2194 Core Th ermal-Hydra ulic s.
. NUREC/CR-2181 VO4 Core and Shielding Analyses.
. NUREG/CR-2306 Countercurrent Steam-Water Flow.
.NUREG/CR-2783 Crac k Front Elemen ts.
Credible Accidents.
. NUREG/CR-24 94
. NUREC/CR-2387 Crevice Corrosion.
.NUREG/CR-2317 VO1 NO4 Critical Heat Flux.
.NUREC/CR-2647 Critical Mass Flow.
. NUREG/CR-2671 Criticality Safety Analyses.
. NUREG/CR-2306 Criticality Safety.
.NUREC/CR-2223 Cross-Section Library.
.NUREG/CR-2306 Debris Bed Experiment.
.NUREG/CR-2412 Decommissioning.
.NUREG-0904 Decommissioning.
.NUREC/CR-2522 Decontamination.
.NUREG-0891 Demand Determinants.
.NUREG/CR-2692 Design Basis Accid ent.
. NUREG/CR-2229 VO!
Detectors.
.NUREG/CR-2413 Diesel Generators.
.NUREG/CR-2099 Diffusion Meteorology.
.NUREG/CR-2022 Direct Radiation Monitoring Network.
.NUREG-0837 VO1 NO2-2 Direct Radiation Monitoring Network.
. NUREG-0837 VO1 NO3-4 Direc t Radiation Monitoring Ne twork.
.NUREG-0837 VOI NO4 Disasters.
.NUREG/CR-2591 Display Format.
..NUREG/CR-2496 Docketed Material.
.NUREG-0540 VO3 NO1 Docketed Material.
.NUREG-0540 VO4 N12 Documents Made Pub licly Available.
.NUREG-0540 VO4 NO2 Documents Made Pub licly Available.
. NUREG-0540 VO4 NO3 Documents Made Pub licly Available.
.NUREC-054L SEO2 113
.NUREG/CR-2639 Dose Measurement.
.NUREC/CR-2612 Dose Predictions.
.NUREC/CR-2612 Dosimetry.
.NUREG/CR-2704 Double-Tiered Storage Rac k s.
.NUREG-0903 Drug and Alcohol Programs.
.NUREG/CR-2704 Dry Storage Systems.
.NUREG/CR-2685 Dynamic Loads.
.NUREC/CR-2392 Dynamic s-Related Saf ety Code.
EARTHGUAKE.
ENDF/B-V.
. NUREG/CR-23 06
.NUREC/CR-2603 EXCOBULLE.
Earthen Radon Supp ression Cover.
.NUREG/CR-2639 Earthen Radon Supp ression Cover.
.NUREG/CR-2727 VOf Ecological Studies.
.NUREG/CR-2521 Ef fluent Dispersion.
Electrical Cable Fire Tests.
.NUREC/CR-2431
.NUREG/CR-2505 Electrical Impedence String Probes.
Electrical System Performance.
NUREC/CR-2559 NUREG/CR-2692 Electricity Demand Growth.
Electronic Dosimeter Testing.
.NUREC/CR-2019 Embrittlement Behavior.
.NUREG/CR-2317 V01 NO3 Emergency Core Coo lant Injection.
..NUREG/aR-2229 VO1 Emergency Core Cooling Systems..
.NUREC/CR-2772
.NUREG/CR-2281 V03 Emergency Core Hea t Transf er.
.NUREG/CR-2584 Emergency Plan.
.NUREG/CR-2629 Emergency Planning.
..NUREG-Oe49 Emergency Response Capability.
.NUREG/CR-2637 Emergency Sump P e r f orman c e..
. NUREG/CR-2403 E01 Engineered Safety Systems.
. NUREG/CR-2681 Entrainment-Imping ement Losses.
.NUREG/CR-2220 V03
..NUREG/CR-2022 Environmental Analysis.
.NUREG-0904 Environmental Impact Statement.
.NUREG-0925 Environmental Impact Statement.
Environmental Impa ct Statements.
.NUREG-0895 Environmental Impact.
.. NUREG/CR-2493 Environmental Insu lt.
Environmental Load s..
. NUREG/CR-2638 Environmental Radiation Protection.
.NUREG/CP-OO22 Environmental Report.
.NUREG-0902 Environmental Statement.
.NUREG-0842 Environmental Stat ement.
.NUREG-0854 Environmental Statement.
Environmental Statement.
.NUREG-0878 Environmental Statement.
.NUREG-0894 Environmental Stat ement.
.NUREG-0895 Environmental Transport Model.
. NUREC/CR-1636 V04 J
Environmental Transport Modeling.
..NUREG/CR-1672 V03
Equipment Failures.
..NUREG/CR-2629 Equipment Gualific ation.
.NUREC/CR-2172 Equipment.
. NUREG/CR-26 81 Event Tree Models.
Exhaust Stac k Plac ement and Height.
.. NUREG/CR-2603 FAST.
.NUREG/CR-2393 FAST.
.NJEEC/CR-2238 Vol FFTF.
.NUREC/CR-2636 FLECHT-SEASET.
.NUREG/CR-2281 V03 FLECHT-SEASET.
.NUREG/CR-1820 FPRF.
114
FRAC.
. NUREG/CR-24 34 FRANTIC II.
....NUREC/CR-2542 Failure Modes.
.NUREC/CR-2569 Failure Pressures.
.NUREG/CR-2569 Failure Rate Analy sis Code.
..NUREG/CR-2434 Failure Rates.
.NUREG/CR-2434 Fast Flux Test Facility..
..NUREG/CR-2238 VO!
Fatigue.
.NUREG/CR-2301 Fault Trees.
.NUREG/CR-2434 Faults..
. NUREG/CR-26 53 Fifty-Year Dose Commitments.
.NUREG/CR-2201 Film-Boiling Corre lations.
.NUREG/CR-2469 Film-Boiling Correlations.
.NUREC/CR-2435 Fire Growth and Recession.
Fire Resistance.
.NUREG/CR-2377 Fire-Retardant Add itives.
.NUREG/CR-2314 Fissile Materials..
. NUREG/CR-2223 Fission Product Source Term.
.NUREG/CR-2593 Fission Product Sp ectra.
.NUREG/CR-2594 Fission Product Transport Codes.
.NUREG/CR-2593 Fission-Product Reaction Facility.
.NUREG/CR-1820 Fission-Produc t Re search.
.NUREG/CR-1820 Flammability Parameters.
.NUREG/CR-2314 Flammability Tests.
.NUREG/CR-2314 Flat Plate Geometry.
.NUREG/CR-2783 Flaws in Plates and Cylinders.
.NUREG/CR-2494 Flooding-Limited B urnout.
.NUREG/CR-2647 Flow Measurements.
.NUREG/CR-1622 Flow Velocities.
.NUREG/CR-1622 Flow-Measuring Device.
.NUREG/CR-1622 Food Chain Transport Models.
..NUREG/CR-2612 Food Chain.
.NUREG/CR-2022 Force Signature Ch aracteristic s.
.NUREG/CR-2790 Forced Flooding.
.NUREG/CR-2622 Forecasting Electr ic Energy and Load.
.NUREG/CR-2692 Foreign Lice' sing and Regulatory Practices.
. NUREG /CR-26 53 Fracture and Irradiation Effects.
.NUREG/CR-2362 Fuel Aerosol Simlant Test.
.NUREG/CR-2393 Fuel Aerosol Simulant Test.
.NUREG/CR-2299 VO4 Fuel Assembly.
.NUREG/CR-2597 Fuel Cycle Facilities.
. NUREG/CR-26 51 Fuel Cycle Facility..
. NUREC/CR-1233 VO4 Fuel Damage Tests.
.NUREG/CR-2432 Fuel Debris.
.NUREG/CR-2194 Fuel Fabrication Plants.
.NUREG/CR-2460 Fuel Handling Accident.
.NUREG/CR-2387 Fuel Performance.
.NURE3/CR-2600 Fuel Performance.
.NURtG/CR-2567 Fuel Reprocessing Plant.
.NUREG/CR-2301 Fuel Reprocessing.
.NUREC/CR-2625 Fuel Transshipments.
.NUREC/CR-2704 Fuel-Steel-Concrete Interactions.
.NUREG/CR-2194 Functions and Organization of Operating Staff.
.NUREG/CP-OO31 VOi Functions and Organization of Operating Staf f.
.NUREC/CP-OO31 VO2 GABAS.
.NUREC/CR-2594 GDRF.
.NUREC/CR-2711 GESMO.
.NUREG/CR-2104 Gamma Densitometer System.
.NUREG/CR-2545 Gamma and Deta Spectrum
.NUREG/CR-2594 Geologic Disposal of Radioactive Waste.
.NUREG/CR-2350 115
.NUREG/CR-1636 VO4 Geologic Disposal.
Geologic Disposal.
.NUREC/CR-1968
..NUREG/CP-OO22 Geologic Disposal.
Geologic Medium.
.NUREC/CR-1672 VO3 Geologic and Hydrologic Investigations.
.NUREG/CR-2381 Geologic / Electromagnetic Model.
.NUREG/CR-2711 Graphics Display R esearch Facility..
Greenhouse Radionuclide Uptake Studies.
.NUREG/CR-2625 Ground-Penetrating Radar Survey.
.NUREC/CR-2589 Groundwater Monitoring Wells.
..NUREC/CR-2381 Groundwater Transp ort Model.
.NUREG/CR-2350 HLW Glass and Ceramics.
.NUREG/CR-2737 HTCR.
.NUREG/CR-2281 VO3 Hazard Evaluations.
.NUREG/CP-OO22 Health Risks..
.NUREG/CR-2435 Heat Transfer Coefficient.
Heat Transfer Coefficients.
.NUREG/CR-2469 Heat Transfer Phenomena
.NUREG/CR-2525 VO7
.NUREG/CR-2525 VOS Heat Transfer Phenomena..
Heat Transfer Test.
.NUREG/CR-2525 VOR Heated Concretes.
.NUREG/CR-2279 High Efficiency Particulate Air Filters..
.NUREC/CR-2565 High Intensity Adjustable Cobalt Array.
.NUREG/CR-2582 High Level Waste Container Sys tems.
.NUREG/CR-2317 VO! NO4 High Level Waste Containers.
.NUREG/CR-2317 VOI NO3 High Toughness Alloys.
.NUREG/CR-2570 High-Density Storage Racks.
.NUREG/CR-2704 High-Level Radioac tive Waste.
.NUREG/CP-OO22 High-Temperature Gas-Cooled Reactor.
.NUREC/CR-2221 VO4
.NUREC/CR-2392 High-Temperature Gas-Cooled Reactor.
Historical Extreme Winds.
.NUREG/CR-2639 Historical Extreme Winds.
.NUREG/CR-2638 Horizontal Stratified Concurrent Flow.
.NUREC/CR-2334 Human Dosimetry.
.NUREG/CR-2496 Human Engineering Design Considerations.
Human Error Rates.
.NUREG/CR-1659 VO3 Human Error.
Human Error...
..NUREG/CR-2417 Human Errors.
.NUREG/CR-2416 Human Factors in the Control Room.
.NUREG/CR-2587 Human Performance.
.NUREC/CR-2711 Hydraulic Performa nce.
.NUREC/CR-2772 Hydrogen Program.
.NUREC/CR-2481 l
Hydrologic Engineering.
.NUREG-0868 Hydrologic Patterns.
.NUREG/CR-1636 VO4 Hydrological Analysis.
.NUREG/CR-2700 IFA.
.NUREC/CR-2567 IPRDS...
.NUREG/CR-2434 Image Distortion..
.NUREC/CR-2496 Impact Force-Time Relations.
.NUREG/CR-2790 Impact Forces.
.NUREG/CR-2790 Impingement Impacts.
.NUREG/CR-2220 VO2 In-Plant Reliabili ty Data System.
.NUREG/CR-2434 Industrial Impact.
.NUREG/CR-2591 Inf ormation Presen tation Mechanisms.
.NUREC/CR-2022 Ingestion Pathway.
Insider Sabotage.
.NUREG/CR-2546 Instrument Failures.
..NUREG/CR-2173 Instrumentation and Control System Components.
..NUREG/CR-2416 116
Instrumentation.
..NUREG/CR-2172 Instrumented Fuel Assembly.
.NUREG/CR-2567 Instrumented Fuel Assembly.
.NUREG/CR-2600 Instrumented Spool Pieces.
.NUREG/CR-2544 Insulation Debris.
Integrated Plant Review.
.NUREG/CR-2403 SO1
.NUREG-0821 Integrated Plant Safety.
.NUREG-0820 Interfacial Shear Stress.
.NUREG/CR-2783 Intermediate Break Test.
..NUREG/CR-2732 Internal Pressurization Effects.
.NUREG/CR-2788 Interphase Transport..
.NUREG/CR-2334 Iodine Behavior.
. NUREG/CR-2683 Iodine Behavior.
...NUREC/CR-2682 Irradiated Reactor (Spent) Fuel.
.NUREG-0725 RO2 Irradiation Capsul es.
. NUREG/CR-20 53 Isotope Concentrations..
.NUREG/CR-2593 Keyword Index.
.NUREG-0304 VO3 LER.
.NUREG/CR-2OOO VOI NOP LER.
.NUREG/CR-2OOO VO1 NO3 LER.
. NUREG/CR-2OOO VOI NO4 LER.
.NUREG/CR-2OOO VO1 NOS LER..
.NUREG/CR-2497 VO2 LER.
.NUREG/CR-2497 VOL LHS.
,NUREG/CR-23b0 LMFBR Containment.
LMFBR..
.NUREG/CR-2181 VO4
.NUREG/CR-1594 VO4 LMFBR.
.NUREG/CR-2238 VOI LMFDR.
..NUREC/CR-2473 LMFBR.
. NUREG/CR-26 81 l
LMFBR..
LOCA.
.NUREC/CR-2229 VOI
.. NUREG/CR-16 59 VO3 LOCAs.
. NUREC/CR-23 56 LOCAs.
..NUREG/CR-2618 LOFT.
.NUREC/CR-1681 LOFT.
LOFT.
.NUREG/CR-0169 VOS
.NUREG/CR-2717 LWBR.
.NUREC/CR-2184 Large Break.
.NUREG/CR-1681 Large Melt Facility.
.NUREG/CR-2611 Latin Hypercube Sampling.
. NUREG/CR-23 50 Leaching.
License Application.
.NUREG/CR-2193 VOJ NOP
.NUREG-0902 License Conditions.
.NUREG/CR-2460 Licensed Operating Reactors.
.NUREG-OO2O VO6 NOP Licensed Operating Reactors.
.NUREG-OO2O VO6 NO3 Licensee Contractor and Vendor Inspection.
.NUREG-OO40 VO6 NO2 Lic ensee Event Rep ort.
.NUREG/CR-2OOO VO1 NOP Licensee Event Rep ort.
.NUREG/CR-2OOO VOI NO3 Licensee Event Rep ort.
.NUREG/CR-2OOO VO! NO4 Lic ensee Event Rep ort.
.NUREG/CR-2OOO VO1 nob Licensee Event Rep orts.
.NUREG-0872 Licensee Event Rep orts.
. NUREG/CR-20 99 117
Licensee Event Rep orts.
.NUREG/CR-2497 VO2 Licensee Event Rep orts.
..,..NUREG/CR-2497 VO!
.NUREG-0748 VO2 NO3 Licensing Actions Summary.
.NUREG-0748 VO2 nob Licensing Actions.
Licensing Reviews.
..NUREG-0580 V11 N O2 -4
..NUREG-0580 V11 NOS Licensing Reviews.
.NUREG-0821 Licensing Saf ety R equirements...
Limestone Concrete...
..NUREG/CR-2279 Liquid Level and Void Fraction Detection.
.NUREG/CR-2432 Liquid Metal Fast Breeder Reactor.
.NUREG/CR-2238 VO!
Listing of Reports.
.NUREG-0304 VO3 Long-Term Risk.
.NUREG/CR-1672 VO3
.NUREG/CR-2133 Loss-of-Coolant Accident.
Loss-of-Coolant Accident.
.NUREG/CR-2470 Loss-of-Coolant Ac cident.
.NUREG/CR-2505 Loss-of-Coolant Accident.
.NUREG/CR-2597 Loss-of-Coolant Accident.
.NUREC/CR-2403 SO1
.NUREG/CR-2581 Loss-of-Coolant Accident.
Loss-of-Coolant Accident.
. NUREC/CR-24 55 Loss-of-Coolant Accident.
.NUREG/CR-2732 Loss-of-Coolant Accident.
.NUREC/CR-2456
. NUREG/CR-1659 VO3 Los s-of-Coolant Ac c id ent s.
Loss-of-Coolant Ac cidents..
.NUREG/CR-1681 Loss-of-Coolant Accidents.
.NUREC/CR-2525 VO7 Loss-of-Coolant Accidents.
.NUREG/CR-2618 Loss-of-Coolant Accidents..
.NUREC/CR-2525 VOS Loss-of-Coolant Accidents.
.NUREG/CR-2525 VO3
.NUREG/CR-2525 VO2 Loss-of-Coolant Accidents.
Loss-of-Feedwater Accident.
..NUREG/CR-2717 Loss-of-Fluid Test.
.NUREG/CR-0169 VOB
..NUREC/CR-2717 Lose-of-Fluid Test.
.NUREG/CR-2699 Low Specific Activity Wastes.
Low-Level Nuclear Waste Disposal.
.NUREG/CR-2589 Low-Level Waste Licensing.
.NUREG-0902 Lower.
.NUREG/CR-2314 MARCH.
. NUREC/CR-1659 VO3
.NUREC/CR-2022 MILDOS.
MRBT.
.NUREG/CR-2470 Magnetite Concrete.
.NUREG/CR-2279 Mark II Containmen t.
.NUREG/CR-2039 Mark III Containment.
.NUREC/CR-1890 Mark III Plants.
.NUREG/CR-2686 Mathematical Model s.
.NUREG-0068 Maxey Flats Disposal Site.
. NUREC/CR-2192 VO1 NO2
.NUREG/CR-2509 Maxey Flats.
Melt Dehavior.
.NUREC/CR-2283
.NUREG/CR-2639 Meteorological Conditions.
Meteorological Considerations..
.NUREG/CR-2584 Meteorological Information.
.NUREG/CR-2584 Meteorological Mea surement.
Microwave Dielectric Heating..
.NUREG/CR-2543 Mining and Milling.
.NUREG/CR-2184 Mixed Oxide Fuel.
.NUREC/CR-2184
.NUREG/CR-2512 Mi xed Oxide Nuclear Fuel.
Mixed-Mode Stress Intensity Factors.
.NUREC/CR-2494 l
Modified Intake Structures.
..NUREC/CR-2220 VO2 Molten Core Simulant.
..NUREG/CR-2611 Molten Fuel-Concrete Interactions.
..NUREC/CR-2481 Monitoring Systems.
..NUREG/CR-2639 Monitoring.
...NUREG/CR-2700 118 l
Multirod Burst Test..
.NUREG/CR-2470 Multirod Burst Test.
.NUREG/CR-2366 VO2 Multirod Burst Test.
..NUREC/CR-2434 NWFT/DVM.
.NUREG/CR-2350 Natural Circulation Test.
.NUREC/CR-2648 Natural Circulation Test.
.NUREG/CR-2618 Natural Weathering Forces.
.NUREG/CR-2639 New Madrid Seismic Zone.
Nondocketed Material.
.NUREG-0540 VO3 NO1 Nondocketed Material.
.NUREG-1540 VO4 N12 Nuclear Fuel Rods..
.NUREG/CR-2470 Nuclear Plant Reliability Data Systems.
.NUREG/CR-2434 Nuclear Propertv Insurance.
..NUREG-0891 Nuclear Saf ety Pil ot Plant.
.NUREG/CR-2299 VO4 Nuclear Waste Repository..
.NUREG/CR-1672 VO3 Nuclear Waste.
.NUREG40923 OR-FLAW.
.NUREG/CR-2494 ORECA.
..NUREG/CR-2221 VO4 ORTAP.
.NUREG/CR-2221 VO4 Operating Conditions.
.NUREG-0861 Operating Conditions.
.NUREC-0926 Operating Crews.
.NUREC/CR-2507 Operating License.
..NUREG-0580 V11 N05 Operating License.
.NUREG-0820 Operating License.
.NUREG-0821 Operating License.
.NUREG-0854 Operating License.
.NUREG-0911 Operating License.
.NUREG-0913 Operating Licenses.
Operating Reactors.
. NUREC-0580 V11 NO1 -4
.NUREG-0748 VO2 NO3 Operating Reactors.
Operating Reactors.
..NUREG-0748 VO2 NO4
.NUREG-0748 VO2 nob Operating Strategies.
.NUREG/CR-2542 Operating Units..
.NUREG-OO2O VO6 NO3 Operation.
.NUREG-0842 Operation.
..NUREG-0848 Operation.
.NUREG-0878 Operation.
.NUREG-0895 Operational Events.
.NUREG/CR-2173 Operational Events.
.NUREG/CR-2172 Operational Information.
.NUREC/CR-2OOO VO! NOD Operational Procedures.
.NUREG/CR-254P Operational Transients.
.NUREC/CR-2648 Operational Transients.
.NUREG/CR-2610 Operator Eligibility.
..NUREG-0863 Ooerator Errors.
.NUREG-0872 Operator Functions.
.NUREG/CR-2507 Operator Licensing or Certification.
.NUREG-0863 Operator Requirements.
.NUREG-0863 Operator Retraining.
.NUREG-0863 Operator Selection and Training.
.NUREG/CP-OO31 VO!
Operator Selection and Training.
.NUREG/CP-OO31 VOD Operator Training Programs.
.NUREG-0863 Operator Training.
.NUREG/CR-2506 Organo-Radionuclide Complex Stability Experiments.
.NUREG/CR-2192 VOJ NOP PAHR.
.NUREG/CR-2412 PRAISE.
.NUREG/CR-2303 PVWS.
.NUREG/CR-20b3 119
Particle Size Stratification.
.NUREG/CR-2412 Personnel Training Systems.
.NUREG/CR-2353 VO2 Phosphor.
.NUREG/CR-2546 Physical Protection.
.NUREG/CR-2727 VO!
Physiological Tolerances..
Pipe Break.
.NUREG/CR-2403 GOS NUREG/CR-2633 Pipe Break.
Pipe Failures.
.NUREC/CR-2301 Piping Reliability Analysis Including Seismic Events..NUREG/CR-2303
.NUREG/CR-2301 Piping Reliability..
Piping and Equipment Components.....
.NUREG/CR-2686 Plant Safety.
.NUREG-0820 Plant Safety.
.NUREG-0821 Plant Struetures.
.NUREC/CR-2638 Plume Rise.
.NUREG/CR-2022 Poolside Facility.
.NUREG/CR-2696 Population Radiation Dose.
.NUREG/CR-2201 Population.
Post-Accident Heat Removal.
.NUREG/CR-2194 Post-Accident Heat Removal.
.NUREG/CR-2412 Post-Closure Repository Performance.
.NUREG/CP-OO22 Postulated Accident Sequences.
.NUREG/CR-2392 Power Plant Structures.
.NUREG/CR-268D Power System Simulations.
.NUREC/CR-2497 VO2 Precursors.
Precursors..
.NUREG/CR-2497 VO1 Pressure Boundaries.
.NUREC/CR-2362 Pressure Doundary Failures.
.NUREC/CR-2600 Pressure Vessel Wall Simulation.
.NUREG/CR-20b3 Pressure-Effects Tests..
.NUREG/CR-2622 Pressure-Vessel Nozzle Corners.
.NUREG/CR-2494 Probabilistic Evaluation.
.NUREC/CR-2039 Probabilistic Frac ture Mechanics.
.NUREG/CR-2301 Probabilistic Risk Analyses.
.NUREC/CR-2464 Probabilistic Risk Assessments.
.NUREG/CR-2434 Probability of Severe Core Damage.
.NUREG/CR-2497 VO2 Probability of Severe Core Damage.
.NUREC/CR-2497 VOi Promp t Burst Energ etics.
.NUREC/CR-2473 Psychological Stress.
.NUREG/CP-OO26 Pulsed-Neutron Activation.
.NUREG/CR-2772 Pump Suction Inlets.
RAB.
..NUREG/CR-2413 RAGBEEF.
.NUREC/CR-2610
.NUREG/CR-2610 RAGTIME.
RAM.
.NUREG/CR-2413 RELAP/ MOD 1.
.NUREC/CR-1826 VO1
.NUREG/CR-1826 VO2 RELAP5.
.NUREG/CR-2593 RIDD/IRT.
RILs.
.NUREG-0435 VO4 NOJ RIMS II.
.NUREG/CR-2591 RSSMAP..
.NUREG/CR-1659 VO3 RVLIS.
.NUREC/CR-2204 VO4 Radiation Capabilities.
.NUREG/CR-2582 Radiation Damage.
.NUREG/CR-2362 Radiation Dose Estimates.
.NUREG/CR-2512 Radiation Dose-Rate Effects.
.NUREG-0894 Radiation Exposure.
Radioactive Effluents.
Radioactive Iodine.
Radioactive Materials.
.NUREC/CR-2699 l
120
Radioactive Waste...
.NUREG/CR-2381 Radioactive Waste.
.NUREG/CR-1636 VO4 Radioactive Waste.
.NUREG/CR-1968 Radioactive Waste.
.NUREC/CR-2736 Radioactive Waste.
.NUREG/CR-2700 Radioactive Wastes.
.NUREG-0904 Radioactive Wastes.
.NUREG/CR-2193 VO1 NO2 Radiological Emergency.
.NUREG/CR-2637 Radiological Health.
.NUREG-0786 Radiological Impac ts.
..NUREC/CR-2184 Radiological I n f or ma t ion.
...NUREG/CR-2637 Radiological Survey.
..NUREC/CR-2722 Radionuclide Concentrations.
.NUREG/CR-2610 Radionuclide Releases.
.NUREG/CR-2201 Radionuclide Releases.
.NUREC/CR-2629 Radionuclide Sorption.
. NUREG/CR-2192 VO! NOP Radionuclides in Surface Water and Groun/ water.
..NUREG-0868 Radionuclidet.
Radiopharmaceuticals.
.NUREG/CR-1030 VO2
..NUREG/CR-2736 Radwastes..
.NUREC/CR-2516 VO1 NO1 -03 Rail Route.
Rank Ordering Vital Areas.
.NUREG-0725 RO2
.NUREG/CR-2551 Rare Earths Facility.
.NUREG-0904 Reactor Dosimetry.
.NUREG/CR-2696 Reactor Effluent Exposure Pathway.
.NUREG/CR-2584 Reactor Operations..
.NUREC/CR-2184 Reactor Operations.
.NUREG/CR-2546 Reactor Operator.
.NUREC/CP-OO31 VO!
Reactor Operator.
.NUREC/CP-OO31 VO2 Reactor Physics.
.NUREG/CR-1851 Reac tor Saf ety Ex p eria.ients.
.NUREG/CR-1851 Reactor Safety Modeling.
.NUREG/CR-2181 VO4 Reactor Safety Research.
.NUREG/CR-2281 VO2 Reactor Safety Study Methodology.
.NUREG/CR-1659 VO3 Reactor Safety.
.NUREG/CR-1594 VO4 Reactor Safety.
.NUREG/CR-2603 Recycled Plutonium.
.NUREG/CR-2184 Reelfoot Rift.
.NUREC/CR-2741 Refill Phase.
.NUREG/CR-2133 Refill /Reflood Stages.
.NUREG/CR-2505 Reflood Tests.
.NUREG/CR-2525 VO1 Regional Economic Models.
.NUREG/CR-2591 Regional Input-Output Modeling System.
Regulatory Base.
.NUREG/CR-2460 Regulatory Guide 2.6.
.NUREG-0849 Regulatory Initiatives.
.NUREG-0903 Regulatory Licensing Status.
.NUREG-0580 V11 N05 Regulatory Standards.
.NUREG-0566 VO2 NO2 Regulatory and Technical Reports.
.NUREG-0304 VO3 Regulatory and Technical Reports.
.NUREG-0304 VO6 NO4 Regulatory and Technical Reports.
.NUREG-0304 VO7 NOJ Release Categories.
.NUREC/CR-1659 VO3 Release of Radioac tive Material.
.NUREG/CR-2546 Reliability Analysis.
.NUREG/CR-2442 Reliability Data Bases.
.NUREG/CR-2434 Ramote Area Monitoring.
.NUREG/CR-2413 Report to Congress.
.NUREG-OO90 VO4 NO4 Reprocessing Spent Fuel.
.NUREG/CR-2184 Research Information Letters.
.NUREG-0435 VO4 NO1 Restarch Results Utilization.
.NUREG-0435 VO4 NO1 121
.NUREG/CR-2772 Residual Heat Removal.
s.
Response Time.
.NUREG/CR-2496 Ricker Stock-Recruitment Model.
.NUREG/CR-2220 VO3
Rift Zones....
Riprap.
.NUREG/CR-2639 Riprap.
.NUREG/CR-2564 Risk Dominating Accident Sequences.
.NUREG/CR-1659 VO3
-Risk Methodologg..
~
.NUREG/CR-1636 VO4 Risk Methodology.
Risk Methodology.
Risk in Operation..
..NUREG/CR-1672 VO3 Risk-Based Methodology.
.NUREG-0725 RO2 i
Road Transportation..
..NUREG/CR-2564 Rock Armoring Blanket.
Rock Armoring - Blan ket..
.NUREG/CR-2639 Rock Weathering.
.NUREG/CR-2525 VO2 Rod Bundle Heat Transfer Test.
.NUREC/CR-2525 VO3 Rod Bundle Heat Transfer Test.
Rod Bundle Heat Transfer Test.
.NUREG/CR-2525 VO!
Rod Bundle Tests.
.NUREG/CR-2435 Rod Bundle.
.NUREG/CR-2545 Rod Consolidation.
.NUREG/CR-2704 Root Uptake.
.NUREC/CR-2625
.NUREG/CR-2314 Rubbers.
.NUREG/CR-1233 VO4 SAA.
.NUREC/CR-2508 SAFE.
.NUREC/CR-2605 SAFE.
SAFE.
.NUREC/CR-2622 SCTF.
.NUREG/CR-2588 SECURORS.
SEP.
.NUREG/CR-1030 VO2 SERATRA.
..NUREG/CR-2636 SET Facility.
.NUREG/CR-2473 SIMMER-11.
SIMMER..
.NUREG/CR-2281 VO3 SNAP Operating System..
.NUREG/CR-2605 SNAP Operating System.
.NUREC/CR-2604
.NUREG/CR-1245 ROI SNAP.
.NUREG/CR-2605 SNAP.
SNAP.
.NUREG/CR-2604 SOS.
505.
.NUREC/CR-260
.NUREC/CR-1890 SRV.
.NUREG/CR-2015 VO4 SSI.
SSTF.
. NUREC/CR-2133 STEP.
.NUREG/CR-1968 SWIFT.
.NUREG/CR-2551 Sabotage Events.
.NUREG/CR-2546 Sabotage Prevention.
.NUREG/CR-2551 Sabotage.
. NUREG/CR-26 33 Sacrificial Shield-Wall.
.NUREG/CR-2588 Safeguards Automated Facility Evaluation.
.NUREC/CR-2604 Safeguards Automated Facility Evaluation.
Safeguards Automated Facility Evaluation.
.NUREG-0725 RO2 Safeguards Inciden ts..
.NUREC/CR-2604 Safeguards Network Analysis Procedure.
Safeguards Network Analysis Procedure.
.NUREG/CR-2546 Safeguards System.
.NUREG/CR-1233 VO4 Safeguards Systems.
.NUREG/CR-2281 VO3 Safety Analysis.
122
Safety Assessments.
. NUREG/CR-26 51 Safety Considerations.
.NUREG-0786 Safety Evaluation Report.
.NUREG-0712 SO6 Safety Evaluation Report.
.NUREG-0787 SO3 Safety Evaluation Report...
.NUREG-0793 Safety Evaluation Report..
.NUREG-0793 SO1 Safety Evaluation Report.
.NUREG-0831 SO2 Safety Evaluation Report.
.NUREG-0847 Safety Evaluation Report.
.NUREG-0857 SO2 Safety Evaluation Report.
.NUREG-0887 Safety Evaluation.
.NUREG-0881 Safety Evaluation.
.NUREG-0911 Safety Evaluation.
.NUREG-0913 Safety Evaluation..
.NUREG-0916 Safety Evaluation.
.NUREG-0519 GO3 Safety Research..
.NUREG/CR-2230 VO!
Saf ety System Pump s and Valves.
.NUREG/CR-2416 Safety System Pump s and Valves.
.NUREG/CR-2417 Safety and Safeguards Regulations.
.NUREG-0725 RO2 Safety-Related Aspects.
.NUREG-0485 VO4 NO4 Sa f e ty-Related Asp ec ts.
.NUREG-0485 VO4 nob Safety-Related Equipment Oualification.
.NUREG/CR-2581 Salt Creep.
.NUREG/CR-2343 Salt Dissolution.
.NUREG/CR-2343 Schlumberger Sound ings.
.NUREG/CR-2653 Security Manag emen t.
.NUREG/CR-2297 Security Of ficer Response Strategies.
.NUREG/CR-2588 Security Systems.
. NUREG/CR-2297 Sediment and Radionuclide Transport Model.
.NUREG/CR-1030 VO2 Seismic Design Errors.
.NUREG-OO90 VO4 NO4 Seismic Saf ety Mar gins Research.
.NUREC/CR-2015 VO4 Seismic Shear Effects.
.NUREG/CR-2788 Seismotectonic Model.
.NUREG/CR-2741 Self-Teaching Curricula.
.NUREG/CR-1968 Semiscale Test Fac ility.
.NUREC/CR-2204 VO4 Semiscale Tests.
.NUREG/CR-2732 Semiscale Tests.
.NUREG/CR-2648 Semiscale.
. NUREG/CR-16 81 Sensitivity Analysis Techniques.
.NUREG/CR-2350 Separate Effects Tests.
.NUREC/CR-2231 Severe Accident Sequence Analy sis.
.NUREG/CR-2221 VO4 Severe Core Damage.
.NUREG/CR-2497 VO2 Severe Core Damage.
.NUREG/CR-2497 VO1 Shift Staffing.
.NUREG-0863 Shipping Cask Calc ulations.
.NUREG/CR-2306 Shi; worms.
.NUREG/CR-2727 VO!
Sirgle-Rod Durst Tests.
.NUREG/CR-2366 VO2 Sice Characteristics.
.NUREC/CR-2584 Site Location.
.NUREG-0902 Site Suitability.
.NUREG-0786 Site Suitability.
.NUREG-0902 Slab Core Test Facility.
.NUREG/CR-2622 Small-Break Accidents.
.NUREG/CR-1622 Small-Dreak LOCA.
.NUREG/CR-2525 VO!
Small-Dreak LOCAs.
.NUREC/CR-2618 Snow Loads.
.NUREG/CR-2638 Socio-Economic Effects.
.NUREG-0894 Sodium-Concrete In terac t ions.
.NUREG/CR-2194 Soil Treatments.
.NUREG/CR-2625 Soil to Plant Tran sf er Coef fic ients.
.NUREG/CR-2495 123 i
l l
m m
l
Soil-Structure Interaction Analysis.
.NUREC/CR-1890 Soil-Structure Interaction Anlaysis.
.NUREG/CR-2039 Soil-Structures In terac tion.
.NUREG/CR-2015 VO4 Solid Angle Method.
.NUREG/CR-22D3 Source Term.
.NUREC/CR-2584 Source Terms.
.NUREG/CR-2629 Source and Byproduct Material License..
.NUREG-0925 Special Nuclear Ma terial.
.NUREG/CR-1233 VO4 Special Project and Non-Power Reactor Renewals.
.NUREG-0580 Vil NOS Spent Fuel Shipment Routes.
.NUREG-0725 ROD Sp ent Fuel Shipmen ts.
.NUREC/CR-2736 Spent Fuel Storage Capability.
.NUREG-OO2O VO6 NO3 Spent Fuel Storage.
.NUREG/CR-2704 Spent Fuel.
.NUREG-0923 Standard Emergency Classes.
.NUREG-0849 Standard Review Plan.
.NUREG-0849 Standards Development.
.NUREG-0566 VO2 NO2 Status Report.
.NUREG-OO40 VO6 NO2 Steady State Film Boiling.
.NUREG/CR-2435 Steam Explosion Ph enomena.
..NUREG/CR-2481 Steam Generator Tube Rupture Accidents.
.NUREC/CR-2683 Steam Generator Tube Rupture Accidents.
.NUREG/CR-2682 Steam Generator Tube.
.NUREG-0909 Steam Property Evaluations.
.NUREC/CR-2518 Steam-Air-Water Test Facility.
.NUREC/CR-2636 Steel Containment Strength.
.NUREG/CR-2442 Stepwise Regression.
.NUREG/CR-2350 Stress Corrosion Cracking.
.NUREC/CR-2317 VO! NO4 Stress Corrosion Crac king.
... NUREG/CR-23 01 Striped Bass Population.
.NUREC/CR-2220 VO2 Striped Bass..
.NUREG/CR-2220 VO3 Structural Capacity.
.NUREG/CR-2039 Structural Materia ls..
.NUREG/CR-2570 Structured Assessment Approach.
.NUREG/CR-1233 VO4 Subchannel Void Fraction Measurements..
.NUREC/CR-2545 Subcompartment Ana lysis Code.
.NUREG/CR-2633 Subcritical Units.
.NUREG/CR-2223 Subsidence.
.NUREC/CR-2343 Summary Data.
.NUREG-0871 VOI NOP Summary Information Report.
.NUREG-0871 VO1 NOP Surface Earth Resi s tivity.
. NUREG/CR-26 53 Surveillance Dosimetry Improvement Program.
..NUREG/CR-2696 System Documentation.
.NUREC/CR-2586 System Failures.,
.NUREC/CR-2173 Systematic Evaluation Program.
.NUREG-0820 Systematic Evaluation Program.
.NUREG-0821 Systematic Evaluation Program.
.NUREG-0485 VO4 NO3 Systematic Evaluation Program.
.NUREG-0485 VO4 NO4 Systematic Evaluation Program.
..NUREG-0485 VO4 NOS j
..NUREG/CR-2353 VO2 Systens Approach to Training.
Systems Fault Trees.
.NUREG/CR-2681 Systems and Safety Analysis.
.NUREC/CR-2221 VO4 Systems..
.NUREG/CR-2172 THTF.
..NUREG/CR-2469 THTF..
.NUREC/CR-2525 VO7 THTF.
.NUREG/CR-2525 VO3 THTF..
.NUREC/CR-2455 THTF.
..NUREG/CR-2525 VOS THTF.
. NUREG/CR-24 56 THTF.
.NUREG/CR-2525 VO2 124 m
THTF..
..NUREG-0837 VO! NO1-2 TLD.
.NUREG-0837 VO1 NO3-4 TLD.
.NUREG-0837 VOI NO4 TLTA.
.... NUREG/CR-2229 VOI TMI-Type Wastes and Solid Products.
....NUREC/CR-2516 VO1 NO1-03 TRAC.....
. NUREG/CR-2281 VO3 TRAC....
. NUREG/CR-2622 TRAP.....
. NUREC/CR-2387 TRIGA..
....NUREG-0911 Tearing Instability.....
....NUREG/CR-2570 Technical Position s....
......NUREG-0820 Technical Specifications.
.NUREG-0861 Technical Specifications.
.NUREG-0926 Technology Transfer..
..NUREG/CR-1968 Tectonic Development.
.NUREG/CR-2741 Temperature Measurements.
..NUREC/CR-0169 VO8 Teredo Bartschi..
..NUREC/CR-2727 VO1 Teredo Navalis.
.NUREG/CR-2727 VO1 Teton Project.
...NUREG-0925 Thermal-Hydraulic Information.
.NUREG/CR-2432 Thermal-Hydraulic Model.
...NUREG/CR-1826 VO!
Thermal-Hydraulic Phenomena.
....NUREG/CR-2618 Thermal-Hydraulic Phenomena.
.NUREG/CR-2648 Thermal-Hydraulic Phenomena.
.NUREG/CR-2732 Thermal-Hydraulic Research.
..NUREG/CR-2281 VO3 Thermal-Hydraulic Systems.
..NUREG/CR-1826 VO2 Thermal-Hydraulic Test Facility.
..NUREG/CR-2525 VO7 Thermal-Hydraulic Test Facility.
..NUREG/CR-2544 Thermal-Hydraulic Test Facility.
. NUREC/CR-24 56 Thermal-Hydraulic Test Facility.
..NUREG/CR-2545 Thermal-Hydraulic Test Facility.
.NUREG/CR-2525 VO5 Thermal-Hydraulic Test Facility.
.NUREC/CR-2455 Thermal-Hydraulic Test Facility..
.NUREG/CR-2525 VO3 Thermal-Hydraulic Test Facility..
.NUREC/CR-2525 VO2 Thermal-Hydraulics Code.
.NUREG/CR-2469 Thermohydraulic Flow Conditions.
.NUREG/CR-2231 Thermoluminescent Dosimeter.
..NUREG-0837 VO1 NO2-2 Thermoluminescent Dosimeter.
.. NUREG-0837 VO1 NO3--4 Thermoluminescent Dosimeter.
..NUREG-0837 VOI NO4 Thermomechanical Effects.
.NUREG/CR-2343 Thorium Fuel Cycles.
.NUREG/CR-2104 Thorium / Uranium Fuel Cycle.
.NUREC/CR-2104 Title List.
.NUREG-0540 VO3 NO2 Title List.
.NUREG-0540 VO4 NOP Title List.
.NUREO-0540 VO4 NO3 TitIe List.
.NUREG-0540 Title List..
.NUREG-0540 VO4 N1P Tornado Conditions.
.NUREG/CR-2565 Tornado Proximity Soundings.
.NUREC/CR-2359 Tornado Transient.
.NUREG/CR-2632 Tornado-Borne Automotive Vehicle Impact.
.NUREG/CR-2790 Training Simulator Response Characteristics.
.NURFG/CR-23b3 VOD Training Simulators.
. NUREG/CR-23 53 VO2 Transient Film Boiling.
.NUREG/CR-2525 VO2 Transient Reactor Analysis Code.
.NUREG/CR-2622 Transient With Mul tiple Failures.
.NUREG/CR-2717 Transport of Radionuclides in Primary Systems..
.NUREG/CR-2713 Transportation of Radioactive Material.
.NUREG/CR-2736 125
l Transportation of Radioactive Materials.
... NUREG/CR-2699
.... NUREG/CR-2192 VO3 NOP Trench Shale.
Tube Bundles.
. NUREG/CR-2545 Tube Rupture.
. NUREG-0916 Tube Rupture.
. NUREG-0909 Turbulence-Centered Models.
. NUREG/CR-2229 VOI Two-Loop Test Apparatus.
. NUREG/CR-2204 VO4 Two-Phase Flow Ins trumentation.
Two-Phase Mass Flux Uncertainty Analysis..
NUREG/CR-2544 UF6 Conversion Fac ilities.
. NUREG/CR-2460 USI A-43.
. NUREG/CR-2403 SO1 UVASUBL.,
. NUREG/CR-2299 VO4
. NUREG/CR-2603 UVADUBL.,
Uncertainty Analyses..
. NUREG/CR-0169 VOR Unresolved Safety Issue A-43.
. NUREG/CR-2760 Unresolved Safety Issue.
. NUREG/CR-2403 SO1 Unresolved Safety Issues.
. NUREG-0606 VO4 NOP
. NUREG/CR-23 59 Upper Air Sounding s.
Uranium Contamination.
. NUREG/CR-2104 Uranium Fuel Cycles.
Uranium Mill Tailings.
. NUREG/CR-2639 Uranium Mill Tailings.
. NUREG/CR-2736 Uranium Ore Conc en tra t e.
Uranium-Producing Regions.
. NUREG/CR-2713 Vapor Deposition Velocities.
Vohicle/ Structure Interaction.
. NUREG/CR-2790 Vossel Level Indicating System.
. NUREG/CR-2204 VO4 Vibratory Stresses.
. NUREC/CR-2301 Vital Areas.
.. NUREG/CR-2508 Void-Fraction Models.
. NUREG /CR-24 56
. NUREG/CR-23 56 WRAP-EM.
WRAP-EM.
. NUREC/CR-1681 NUREG/CR-2521 Wcke Flow.
i Weste Containers.
. NUREG/CR-2193 VO1 NOP Weste Disposal Site.
. NUREG-0904 Waste Facilities.
. NUREG/CR-2737 Weste Forms.,
Weste Management.
NUREC/CP-OO22 Weste-Isolation Flow and Transport.
. NUREG/CR-2059 Water Hammer Events.
Weter Reactor Analysis.
NUREC/CR-23 56 NUREG/CR-2317 VO1 NO4 Wald Heat-Affected Zones.
W2nner Array Soundings.
. NUREC/CR-26 53 W2st Lake Landfill.
. NUREG/CR-2722 White Perch Popula tion.
. NUREG/CR-2220 VO2 Wind Profiles.
. NUREC/CR-2359
. NUREG/CR-2564 Wind and Water Ero sion.
Wood-Boring Bivalves..
NUREG/CR-2727 VOI
., NUREG/CR-2496 Work Surface Light Reflection.
Yellow Book.
.. NUREG-OO30 VO6 NO1 Yellowcake.
. NUREG/CR-2736 Zircaloy Alpha Pha se.
. NUREG/CR-2597 Zircaloy Cladding.
. NUREC/CR-2470 l
l l
l I
l 126 f
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 VO 1 NO1-2: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, Ja n u a r y-Je n e 1981.
NUREG-0837 VO1 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-Dec emb er 1981.
NUREG-0837 VO1 NO3-4: Errata To NUREG-0837 Volume 1, Numbers 3-4, Changing Number 3 & Dates Covered To July-Sep temb er,1981, to Nit (:
TLD DIRECT RADI ATION MONITORING NETWORK. P rogre ss Report.
NUREG-0837 VO1 NO4: NRC TLD DIRECT RADI ATION MONITORING NETWORK. Progress Report, October-December 1981.
DIVISION OF ENGINEERING & TECHNICAL PROGRAMS 5
NUREG-0909: NRC REPORT ON THE JANUARY 25,1982 STEAM GENERATOR TUDE RUPTURE AT R.E.
GINNA NUCLEAR POWER PLANT.
REGION 4.
OFFICE OF DIRECTOR NUREG-OO40 VO6 NO1: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report, January 1982-March 1982.(White Book)
EDO - OFFICE OF ADMINISTRATION DIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL NUREG-0304 VO3 SO1: REGULATORY AND TECHNICAL REPORTS. Comp ilat ion For 1975-1978.
NUREG-0304 VO6 NO4: REGULATORY AND TECHNICAL REPORTS. Comp ila t ion For 1981.
NUREG-0304 VO7 NO1: REGULATORY AND TECHNICAL REPORTS. Comp ilat ion For First Guarter 1982.
NUREG-0540 NO2: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. Documents From October Through December 1974 For Dockets 50-334 Through STN 50-597.
NUREG-0540 VO3 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1981.
NUREG-0540 VO4 NO1: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILADLE.
January 1-31,1982.
NUREG-0340 VO4 NO2: TITLE LIST OF DOC JMENTS MADE PUBLICLY AVAILABLE.
February 1-28, 1982.
127
NUREG-0540 VO4 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. March 1-31, 1982.(FOIA Supplement)
NUREG-0540 VO4 NO3: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. March 1-31, 1982.
NUREG-0750 V13 102: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. January-June 1981.
NUREG-0750 V14 IO1: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. July-September 1981.
NUREG/CP-OO31 VO1: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON OPERATOR TR AINING AND GUALIFICATIONS.
NUREG/CP-OO31 VO2: PROCEEDINGS OF THE CSNI SPECIALIST MEETING ON OPERATOR TR AINING AND GUALIFICATIONS.
EDO - OFFICE OF MANAGEMENT & PROGRAM ANALYSIS OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS NUREG-OO2O VO6 NO2: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of January 31,1982. ( Grey B oo k )
NUREG-OO30 VO5 NO4: NUCLEAR POWER PLANTS-CONSTRUCTION STATUS REPORT. Data As Of December 31,1981.(Yellow Book)
NUREC-0871 VO 1 NO2:
SUMMARY
INFORMATION REPORT. January 1,1982-March 31,1982. (B rown Book )
INTERNAL INFORMATION SYSTEMS BRANCH NUREG-0566 VO2 NO2: STANDARDS DEVELOPMENT STATUS
SUMMARY
REPORT. Data As Of February 28, 1982. (Green Book)
EDO - OFFICE OF STATE PROGRAMS OFFICE OF STATE PROGRAMS, DIRECTOR NUR EG-0891 : NUCLEAR PROPERTY INSURANCE: STATUS AND OUTLOOK.
]
EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA DIRECTOR'S OFFICE NUREG-OO90 VO4 NO4: REPORT TO CONGRESS ON ABNORMAL
^ ~ - ~
OCCURRENCES. Guarterly R ep ort, October-December 1981.
OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)
DIRECTOR'S OFFICE, OFFICE OF INSPECTION AND ENFORCEMENT NUREG-0849: STANDARD REVIEW PLAN FOR THE REVIEW AND EVALUATION OF EMERGENCY PLANS FOR RESEARCH AND TEST REACTORS.
l l
NUREG-0903: SURVEY OF INDUSTRY AND GOVERNMENT PROGRAMS TO COMBAT DRUC AND ALCOHOL ABUSE.
OFFICE OF NUCLEAR MATERI AL SAFETY & SAFEGUARDS OFFICE OF NUCLE AR MATERI AL SAFETY & SAFEGUARDS, DIRECTOR NUREG-0725 RO2: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
NUREG-0904: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE DECOMMISSIONING OF THE RARE EARTHS FACILITY, WEST CHIC AGO, ILL INOIS. Doc k e t No. 40-2061.(Kerr-McGee Chemical Corporation)
NUREG-0923: ADVANCE NOTIFICATION OF SHIPMENTS OF NUCLEAR WASTE AND SPENT FUEL: Guidanc e.
128
NUREG-0925: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE OPER ATION OF THE TETON PROJECT. Doc k e t No.
40-8781 (Teton Exploration Drilling Comp any, Inc orp ora ted )
DIVISION OF WASTE MANAGEMENT NUREG-0902: SITE SUITABILITY, SELECTION AND CHARACTERIZATION BRANCH TECHNICAL POSITION - Low Level Waste Licensing Branch.
l l
CONSOLIDATION OF THE OFFICES OF NUC. REG. RESEARCH & STANDARDS DEVELOPMEN l
l OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR NUREG-0435 VO4 NO1: RESEARCH PROJECT CONTROL SYSTEM (RPCS) STATUS
SUMMARY
REPORT. Researc h Resul ts Utilizati on. Lata From July 1981-March 1982.(Buff Book)
DIVISION OF FAC ILITY OPERATIONS NUREG-0863: SURVEY OF FOREIGN REACTOR OPERATOR GUALIFICATIONS, TRAINING. nND STAFFING REGUIREMENTS.
NUREC-0872: A FEASIBILITY STUDY OF USING LICENSEE EVENT REPORTS FOR A STATISTICAL ASSESSMENT OF THE EFFECT OF OVERTIME AND SHIFT WORK ON OPERATOR ERROR.
EDO-RESOURCE MANAGEMENT OFFICE OF RESOURCE MANAGEMENT, DIRECTOR NUREG-0606 VO4 NO2: UNRESOLVED SAFETY ISSUES
SUMMARY
. Da ta As Of May 21,1982. ( Aqua Book )
MANAGEMENT INFORMATION BRANCH NUREG-OO2O VO6 NO3: LICENSED OPERATING REACTORS. Status Summar y Rep or t. Febr uary 1982.(Beige Book)
NUREG-OO30 VO6 NO1: NUCLEAR POWER PLANTS-CONSTRUCTION STATUS REPORT. Data as of March 31,1981. ( Ye l l ow B o o k )
NUREG-0485 VO4 NO3: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
-(
REPOR T. Da ta As Of March 31,1982. (Bu f f Boo k )
NUREG-0485 VO4 NO4: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of April 30,1982. (Bu f f Boo k )
NUREG-0485 VO4 NOS: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of May 31,1982. (Buf f Book )
NUREG-0580 V11 NO1-4: DRAFT REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data as of January 1 - April 19,1982. (Blue Boo k )
NUREG-0580 V11 N05: REGULATORY LICENSING ST ATUS
SUMMARY
REPORT. Data As Of May 15,1982. ( B l u e Book)
NUREC-0748 VO2 NO3: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of March 31,1982.(Orange Book)
NUREG-0748 VO2 NO4: OPERATING REACTOR LICENSING ACTIONS
SUMMARY
. Data As Of April 30,1982.(Orange Book)
NUREG-0748 VO2 N05: OPERATING REACTORS LICENSING ACTIONS SUMM ARY. Data As Of May 31,1982. (Orange Boo k )
OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)
OFFICE OF NUCLEAR REACTOR REGULATION, D IRECTOR NUREG-0519 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LASALLE COUNTY STATION, UNITS 1 AND 2. Doc k e t No s.
50-373 And 50-374.
(Commonweal th Edison Company,et al.)
NUREG-0712 SO6: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 1 & 2. Docket Nos.
129
50-361 And 50-362. (Southern Calif ornia Ed ison Company )
NUREG-0787 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WATERFORD STEAM ELECTRIC STATION, UNIT NO.
- 3. Docket No.
50-382.(Louisiana Power & Light Company)
NUREG-0793: S AFETY EVALUATION REPORT RELATED TO THE OPERATION OF HIDLAND PLANT, UNITS 1 AND 2. Doc ket Nos. 50-329 And 50-330.
(Consumers Power Company)
NUREG-0793 501: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MIDLAND PLANT, UNITS 1 AND 2. Doc ket Nos 50-329 And 50-330. (C onsumers Power Company)
NUREG-0831 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF GRAND GULF NUCLEAR STATION, UNITS 1 & 2. Do c k et Nos. 50-416 &
50-417. (Mis sissippi Power And Light Compa ny )
NUREG-0842: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPER ATION OF ST. LUCIE PLANT, UNIT NO.
- 2. Docket No. 50-389.(Florida Power & Light Company)
NUREG-0847: S AFETY EVALUATION REPORT RELATED TO THE OPERATION OF THE WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Doc k et Nos. 50-390 And 50-391.(Tennessee Valley Authority)
NUREG-0848: F INAL ENVIRONMENTAL STATEMENT RELATED TO THE OPER AT ION OF BYRON STATION UNITS 1 AND 2. Docket Nos. STN 50-454 And STN 50-455.(Commonwealth Edison Company)
NUREG-0854: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPER ATION OF CLINTON POWER STATION, UNIT NO.
- 1. Docket No. 50-461. (Illinois Power Company,et al.)
NUREC-0857 SO2: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PALO VERDE NUCLEAR GENERATING STATION, UNITS 1,2 & 3. Doc ket Nos. S1N 50-528, STN 50-529 & STN 50-530. ( Ari zona P ublic Servic e Comp any )
NUR EG-0861 : TECHNICAL SPECIFICATIONS FOR LA SALLE COUNTY STATION, UNIT NO.
- 1. Doc k e t No.
50-373 (Commonwealth Edison Company).
NUREG-0878: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPER ATION OF WOLF CREEK GENERATING STATION, UNIT 1. Docket No. STN 50-482.
(Kansas Gas And Electric Company.et al.)
NUREG-0881: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WOLF CREEK GENER ATING STATION, UNIT NO.
- 1. Docket No. STN 50-482.
(Kansas Gas And Electric Company et al.)
NUREG-0887: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PERRY NUCLEAR POWER PLANTS, UNITS 1 & 2. Docket Nos. 50-440 And 50-441.(Cleveland Electric Illuminating Company)
NUREG-0894: DRAFT ENVIRONMENTAL STATEMENT RELATED TO THE CONSTRUCTION OF SKAGIT/H ANFORD NUCLEAR PROJECT, UNITS 1 AND 2. Docket Nos. STN 50-522 And STN 50-523.(Puget Sound Power And Light Company, Pacific Power And Ligh t Company, et al. )
NUREC-0895: DR AFT ENVIRONMENTAL STATEMENT RELATED TO THE OPER ATION OF SEA 3 ROOK STATION, UNITS 1 AND 2. Docket Nos. 50-443 & 50-444.
(Public Service Company Of New Hampshire,et al.)
NUREG-0913: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF TilF OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE UNIVERSITY OF FLORIDA. Doc k e t No. 50-83.
NUREG-0916: S AFETY EVALUATION REPORT RELATED TO RESTART OF R. E.
GINNA NUCLEAR POWER PLANT. Doc k et No.
50-244. (Ro chester Gas And El ec tric Corporation)
NUREG-0916 ER R: SAFETY EVALUATION REPORT RELATED TO THE RESTART OF R. E.
CINNA NUCLEAR POWER PLANT. Doc k et No 50-244.(Rochester Gas And Electric Corporation)
CLINCH RIVER BREEDER REACTOR PROGRAM OFFICE NUREG-0786 RO1: SITE-SUITABILITY REPORT IN THE MATTER OF THE CLINCH RIVER BREEDER REACTOR PLANT. Docket No. 50-537.
DIVISION OF ENGINEERING 1%
NUREG-0868: A COLLECTION OF MATHEMATICAL MODELS FOR DISPERSION IN SURFACE WATER AND GROUNDWATER.
DIVISION OF LICENSING NUREG-0020 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC l
EVALUATION PROGRAM FOR PALISADES PLANT. Do c ket No. 50-255. (C onsumers Power Company)
NUREG-0821 DR FT: INTEGRATED PLANT SAFETY ASSESSMENT SYSTEMATIC EVALUATION PROGRAM FOR R.E.
GINNA NUCLEAR POWER PLANT. Doc k e t No.
50-244. (Roc hester Gas & Elec tric Corporat ion) l NUREG-0911: S AFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THI-OPERATING LICENSE FOR THE WASHINGTON STATE UNIVERSITY TRIGA REACTOR. Doc k e t No. 50-27.
DIVISION OF SAFETY TECHNOLOGY NUREG-0926: TECHNICAL SPECIFICATIONS FOR GR AND GULF NUCLEAR STATION, UNIT NO.
- 1. Docket No.
50-416. (Missi ssipp i Power and Light Company) 131
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, refor to the main citation by the NUREG/CR number.
EDO - OFFICE OF STATE PROGRAMS OFFICE OF STATE PROGRAMS, DIRECTOR NUREG/CR-2699: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MARYLAND.
June 1980-June 1981.
NUREG/CR-2736: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MICHIGAN. September 1980-August 1981.
EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA DIRECTOR'S OFFICE NUREG/CR-2OOO VO1 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of February 1982.
NUREG/CR-2OOO VO1 N3. LICENSEE EVENT REPORT (LER) COMPILATION: For Month of March 1982.
NUREG/CR-2OOO VO1 N4: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of April 1982.
NUREG/CR-2OOO VO1 N5: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of May 1982.
OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS, DIRECTOR NUREG/CR-02OO ERR: SCALE: MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LICENSING EVALUATION.
NUREG/CR-2022: TECHNICAL REVIEW OF THE DISPERSION AND DOSE MODELS USED IN THE MILDOS COMPUTER PROGRAM.
NUREG/CR-2306: CSRL-V: PROCESSED ENDF/D-V 227-NEUTRON-GROUP AND POINTWISE CROSS-SECTION LIDRARIES FOR CRITICALITY SAFETY, REACTOR l
AND SHIELDING STUDIES.
DIVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG/CR-2184: COMPARISON OF THE RADIOLOGIC AL IMPACTS OF THOR IUM AND URANIUM NUCLEAR FUEL CYCLES.
l NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULATORY BASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE FUEL CYCLE.
j I
NUREG/CR-2704:
U. S.
REACTOR SPENT-FUEL STOR AGE C APADILITIES.
133 I
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, Gl.
LOUIS COUNTY, MISSOURI.
DIVISION OF WASTE MANAGEMENT NUREG/CR-2350: SENSITIVITY ANALYSIS TECHNIGUES: SELF-TEACHING CURRICULUM.
NUREG/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DISPOSAL OF LOW-LEVEL R ADIOAC TIVE WASTE.
CONSOLIDATION OF THE OFFICES OF NUC. REG. RESEARCH & STANDARDS DEVELOPMEN OFFICE OF NUCLE AR REGULATORY RESEARCH, DIRECTOR NUREG/CR-0169 V13: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES. Volume XIII. Temperature Measurements.
NUREG/CR-1245 RO1: CORRECTIONS AND ADDITIONS TO USER 'S GUIDE FOR SNAP. (NUREG/CR-1245. SAND 80-0315).
NUREG/CR-1594 VO4: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER-DECEMBER 1980.
NUREG/CR-1622: FLOW MEASUREMENT DY PULSED-NEUTRON ACTIVATION TECHNIGUES AT THE PKL FACILITY AT ERLANGEN (GERMANY).
NUREG/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLO3Y APPLICATIONS PROGRAM: Ca lvert Clif f s No. 2 PWR Power P lant.
NUREG/CR-1681: WRAP-PWR VERIFICATION STUDIES.
NUREG/CR-1820: STATUS REPORT ON THE FISSION-PRODUCT RESEARCH PROGRAM.
NUREG/CR-1826 VO1: RELAP5/ MOD 1 CODE MANUAL. Volume 1: System Models And Numerical Methods.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Volume 2: User 's Gu ide end Input Requirements.
NUREC/CR-1851: REACTOR PHYSICS DESIGN CALCULATIONS FOR THE ACPR UPGRADE.
NUREG/CR-2053: HEAT TRANSFER ANALYSIS OF THE LWR PRESSURE VESSFL STEEL IRRADIATION CAPSULES IN THE OAK RIDGE RESEARCH REACTOR-PRESSURE VESSEL DENCHMARK FACILITY.
NUREG/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
NUREC/CR-2141 VC4: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM. Gua rterly Progress Report For October-December 1981.
NUREG/CR-2172:
SUMMARY
AND DIDLIOGRAPHY OF SAFETY-RELATED EVENTS AT DOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2172 ERR:
SUMMARY
AND DIDIOLOGRAPHY OF SAFETY-RELATED EVENTS AT DOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2173:
SUMMARY
AND DIDLIOGRAPHY OF SAFETY-RELATED EVENTS AT 6
PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2173 ERR:
SUMMARY
AND DIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT PRESSURI ZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2181 VO4: PHYSICS OF REACTOR S AFET Y. Gua r t e r l y Report,0ctober-December 1981.
NUREG/CR-2192 VO1 N2: EVALUATION OF ISOTOPE MIGRATION-LAND DURI AL. Guar terly Progress Rep ort, Ap ril-June 1981.
NUREC/CR-2193 VO1 N2: PROPERTIES OF R ADIOACTIVE WASTES AND WASTE CONTAINERS. Guarterly Progres s Rep or t, Apr il-June 1981.
NUREG/CR-2194: CONTAINMENT RESEARCH PRIORITIES.
NUREG/CR-2204 VO4: ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM. Gua r terly Progress Report,0c tober-Decemb er 1981.
NUREG/CR-2220 VO2: THE IMPACT OF ENTR AINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
NUREC/CR-2220 VO3: THE IMPACT OF ENTR AINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
NUREG/CR-2221 VO4: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progress 134
l i
Report, October 1-December 31,1981.
NUREQ/CR-2231: BWR LOW FLOW BUNDLE UNCOVERY TEST AND ANALYSIS.
NUREG/CR-2238 VO1: ADVANCED REACTOR SAFETY RESEARCH.Guarterly Report, January-March 1981.
NUREG/CR-2279: WATER RELEASE FROM HEATED CONCRETES.
NUREG/CR-2281 VO2: NUCLEAR REACTOR SAFETY. April 1-June 30,1981.
NUREQ/CR-2281 VO3: NUCLEAR REACTOR SAFETY. July 1-September 30,1981.
NUREQ/CR-2203: DIRECT OBSERVATION OF MELT BEHAVIOR DURING HIGH TEMPERATURE MELT / CONCRETE INTERACTIONS.
NUREQ/CR-2299 VO4: AEROSOL RELEASE AND TRANSPORT PROGRAM. Guar ter1y Progress Report For October-December 1981.
NUREG/CR-23OO VO1 R1: DRAFT:PRA PROCEDURE QUIDE. A Guide To Th e Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
NUREQ/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-MEUTRON-GROLO AND i
POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALITY SAFETY, REACTOR l
AND SHIELDING STUDIES.
j NUREG/CR-2314: AGING WITH RESPECT TO FLAMMABILITY'AND OTHER j
PROPERTIES IN FIRE-RETARDED ETHYLENE PROPYLENE RUBBER AND CHLOROSULFONATED POLYETHYLENE.
NUREG/CR-2317 VO1 N3: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW
- ~
NUREG/CR-2317 V01 N4: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATERIALS. Guarterly Progress Report July-September 1981.
CONTAINER MATERI ALS. Guarterly Progress Report, October-December 1981.
NUREQ/CR-2353 VO2: SPECIFICATION AND VERIFICATION OF NUCLEAR POWER PLANT TRAINING SIMULATOR RESPONSE CHARACTERISTICS. Part II:
Conclusions And Recommendations.
NUREQ/CR-2356: UPDATED INPUT FOR THE WR/P-EM SYSTEM.
NUREG/CR-2362: RELATIONSHIPS BETWEEN CHN.*Y V-NOTCH IMPACT ENERGY AND FRACTURE TOUGHNESS.
NUREQ/CR-2366 VO2: MULTIROD BURST TEST PROGRAM PROGRESS REPORT FOR JULY-DECEMBER 1981.
NUREQ/CR-2377: TESTS & CRITERIA FOR FIRE PROTECTION OF CABLE PENETRATIONS.
NUREQ/CR-2392:
SUMMARY
OF ORNL WORK ON MIC-SPONGORED HTOR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
NUREG/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
NUREG/CR-2393 ERR: Errata, changing rept number to NUREQ/CR-2593,to A USER 'S MANUAL FOR COMPUTER CODE RIBD/IRT.
l NUREG/CR-2394 ERR: Errata, changing rept number to NUREG/CR-2594,to A USER'S MANUAL FOR THE GABAS SPECTRUM COMPUTER CODE.
NUREQ/CR-2412: HEAT REMOVAL FROM A STRATIFIED UO2-SODIUM PARTICLE BED.
NUREG/CR-2416: INITIAL GOANTIFICATION OF HUMAN ERROR ASSOCIATED WITH l
SPECIFIC INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS IN LICENSED 1
NUCLEAR POWER PLANTS.
NUREG/CR-2417: IDENTIFICATION AND ANALYSIS OF. HUMAN ERRORS UNDERLYING PUMP AND VALVE RELATED EVENTS REPORTED BY NUCLEAR POWER PLANT LICENSEES.
NUREQ/CR-2431: BURN MODE ANALYSIS OF HORIZONTAL CABLE TRAY FIRES.
NUREG/CR-2432: A UNIGUE CONCEPT FOR LIGUID LEVEL AND VOID FRACTION DETECTION IN SEVERE FUEL DAMAGE TESTS.
NUREG/CR-2434: FRAC (FAILURE RATE ANALYSIS CODE): A COMPUTER PROGRAM FOR ANALYSIS OF VARIANCE OF FAILURE RATES.An Application User's Guide.
NUREQ/CR-2435: DISPERSED FLOW FILM BOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMPARISONS.
i 135 s
,,,,,, -, ~ - -. - -.__
NUREQ/CR-2455: EXPERIMENTAL INVESTIGATIONS OF BUNDLE BOILOFF AND REFLOOD UNDER HIGH-PRESSURE LOW HEAT FLUX CONDITIONS.
NUREQ/CR-2456: EXPERIMENTAL INVESTIGATICNS OF UNCOVERED-BUNDLE HEAT TRANSFER AND TWO-PHASE MIXTURE-LEVEL SWELL UNDER HIGH-PRESSURE LOW HEAT-FLUX CONDITIONS.
NUREQ/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTIAL OR POISSON DATA.
NUREQ/CR-2469: AN ANALYSIS OF TRANSIENT FILM BOILING OF HIGH-PRESSURE WATER IN A ROD BUNDLE.
NUREQ/CR-2470: THERMOMETRY IN THE MULTIROD BURST TEST PROGRAM.
NUREQ/CR-2473: SIMMER ANALYSIS OF PROMPT BURST ENERGETICS l
EXPERIMENTS.
I NUREQ/CR-2481: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report, April-September 1981.
NUREQ/CR-2493: AGUEOUS IODIE CHEMISTRY IN LWR ACCIDENTS: Review And l
Assessment.
NUREQ/CR-2494: OR-FLAW: A FINITE ELEMENT PROGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFIED FLAWS IN PLATES, CYLINDERS i
AND PRESSURE-VESSEL N0ZZLE CORNERS.
NUREQ/CR-2496: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
NUREC/CR-2505: ELECTRICAL IWEDANCE STRING PROBES FOR TWO-PHASE VOIDS AND VELOCITY MEASUREMENTS.
NUREQ/CR-2516 V01 N1: CHARACTERIZATION CF TMI-TYPE WASTES AND SOLID PRODUCTS. Guarterly Progress Report, April-September 1981.
NUREQ/CR-2525 VO1: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Vo lume 1-ORNL Small Break LOCA Test Series I: Experimental Data Report.
NUREQ/CR-2525 VO3: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hydraulic Test Facility Experimental Data Report For Test 4
- 3. 06. 68-Tra nsient Film Boilin g In Upflow.
NUREQ/CR-2525 VO5: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Vo lume 5-Thermal-Hydraulic Tes t Facility Experimental Data Report For Test 3.08.6C-Transient Film Boiling In Upflow.
NUREQ/CR-2525 VO7: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Vo lume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Series 3.07.9-Steadt-State Film Boiling In Upflow.
NUREQ/CR-2542: SENSITIVITY STUDY USING THE FRANTIC CODE FOR THE l
UNAVAILABILITY OF A SYSTEM TO THE FAILURE CHARACTERISTICS OF THE COMPONENTS AND THE OPERATING CONDITIONS.
NUREQ/CR-2543: A STUDY OF TE FEASIBILITY OF MICROWAVE DIELECTRIC HEATING FOR LMFBR TRANSITION PHASE ACCIDENT SEGUENCE BOILING STUDIES.
NUREQ/CR-2544: TWO-PHASE MASS FLUX UNCERTAINTY ANALYSIS FOR THERMAL-HYDRAULIC TEST FACILITY INSTRUMENTED SPOOL PIECES.
NUREQ/CR-2545: DESIGN CCNCEPT AND TESTING OF AN IN-BUNDLE GAMMA DENSITOMETER FOR SUBCHANNEL VOID FRACTION MEASUREMENTS IN THE THTF ELECTRICALLY HEATED ROD BUNDLE.
NUREQ/CR-2546: REACTOR SAFEQUARDS AGAINST INSIDE SABOTAGE.
NUREQ/CR-2551: RANK ORDERING OF VITAL AREAS WITHIN NUCLEAR POWER PLANTS.
NUREQ/CR-2559: RESULTS OF PHASE ONE OF PLANT ELECTRICAL SYSTEM (PES)
STUDY.
NUREQ/CR-2565: STRUCTURAL PERFORMAlvCE OF HEPA FILTERS UNDER SIMULATED TORNADO CONDITIONS, NUREQ/CR-2581: SOME EFFECTS OF ELECTRONS SLOWING DOWN IN MATERIALS WITH APPLIC ATION TO SAFETY-RELATED EQUIPMENT GUALIFICATION.
NUREQ/CR-2582: RADIATION CAPABILITIES OF THE SANDIA HIGH INTENSITY ADJUSTABLE COBALT ARRAY.
NUREQ/CR-2586: A SURVEY OF ETHODS FOR IMPROVING OPERATOR ACCEPTANCE OF COMPUTER IZED AIDS.
138
{
/
NUREG/CR-2587: FUNCTIONS AND OPERATIONS OF NUCLEAR POWER PLANT CREWS.
NUREG/CR-2588: SECURITY OFFICER RESPONSE STRATEGIES (SECURORS).
NUREG/CR-2593: A USER 'S MANUAL-FOR COMPUTER COLE RIBD/IRT.
NUREG/CR-2594: A USER 'S MANUAL FOR THE C% BAS SPECTPUM COMPUTER col >E.
NUREG/CR-2597: STEADY-STATE PRESSURE LOSSES' FOR MULTIROD BURST TEST (MRBT) BUNDLE B-5.
NUREG/CR-2604: THE SNAP OPERATING SYSTEM (SOS) USER 'S GUIDE.
NUREG/CR-2605: THE SNAP OPERATING SYSTEM REFERENCE MANUAL.
NUREC/CR-2610: RAGBEEF: A FORTRAN IV IMPLEMENTATION OF A TIME-DEPENDENT MODEL FOR RADIONUCLIDE CONTAMINATION OF BEEF (9 REG /CR-2611: MGO AND 70 W% UO2-30W% Y203: THERMOPHYSICAL AND
.RAME'eNT PROPERTIES.
NUREG/CR-2618: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A NATURAL.
CIRCUL ATION TEST S-NC-7C.
NUREG/CR-2622: ANALYSIS OF TRAC AND SCTF RESULTS FOR SYSTEM PRESSURE-EFFECTS TESTS UNDER FORCED FLOODING (PUNS 506.507 AND 508).
NUREG/CR-2632: RESPONSE OF CENTRIFUGAL BLOWERS TO SIMULATED TORNADO TRANSIENTS. July-Sep temb er 1931.
NUREG/CR-2636: EXPERIMENTAL DATA REPORT FOR AIR-WATER FLOODING TESTS OF THE FLECHT-SEASET PROGRAM SET FACILITY VESSEL UPPER PLENUM.
NUREG/CR-2647: CRITICAL HEAT FLUX EXPERIMENTS UNDER LOW FLOW CONDITIONS IN A VERTICAL ANNULUS.
NUREG/CR-2649: EXPERIMENTAL DATA REPORT FOR SEMISCALE MOD-2A NATUR AL CIRCULATION TEST SERIES (TESTS S-NC-8B AND S-NC-9).
NUREC/CR-2652: EVALUATION AND PERFORMANCE OF CLOSED-CIRCUIT BREATHINO APPARATUS.
NUREG/CR-2671: THE MARVIKEN FULL SCALE CRITICAL FLOW TESTS. Summary Report.
NUREC/CR-2681: ESTIMATED RECURRENCE FREQUENCIES FOR INITI ATING ACCIDENT CATEGORIES ASSOCIATED WITH THE CLINCH RIVER BREEDER REACTOR PLANT DESIGN.
NUREC/CR-2692: AN INTEGRATED SYSTEM FOR FORECASTING ELECTRIC ENERGY AND LOAD FOR STATES AND UTILITY SERVICE AREAS.
NUREG/CR-2696: CALCULATIONS OF TWO SERIES OF EXPERIMENTS PERFORMED Al THE POOLSIDE FACILITY USING THE DAK RIDGE RESEARCH REACTOR.
NUREG/CR-2711: PERFORMANCE AND DESIGN REQUIREMENTS FOR A GRAPHICS DICPLAY RESEARCH FACILITY.
NUREG/CR-2717: EXPERIMENT DATA REPORT FOR LOFT ANTICIPATED TR ANSIFNT WITHOUT SCR AM EXPERIMENT L9-3.
NUREG/CR-2732: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A INTERMEDI ATE BREAK TEST SERIdS. (Tes ts S-I B-1 And S-IB-2).
DIVISION OF ACC IDENT EVALUATION NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4. 4 - 30 SSTF DESCRIPTION DOCLMENT.
NUREG/CR-2229 VO1: PWR LARGE BREAK SIMULATION TESTS--BWR BLOWDOWN / EMERGENCY CORE COOLING PROGR AM.
NUREG/CR-2518: THERMODYNAMIC PROPERTIES OF WATER FOR COMPUTER SIMULATION OF POWER PLANTS.
NUREG/CR-2567: FINAL DATA REPORT FOR THE INSTRUMENTED FUEL ASSFMDLY (IFA)-432.
NUREG/CR-2600: END-OF-IRRADIATION DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFAl-527.
NUREG/CR-2603: BUBBLE BEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
NUREG/CR-2602: CITADEL: A COMPUTER CODE FOR THE ANALYSIS OF IODINE BEHAVIOR IN STEAM GENER ATOR TUBE RUPTURE ACCIDENTS.
NUREG/CR-2683:.IODIyE BEHAVIOR IN STEAM GENERATOR TUBE RUPTURE ACCIDEN1S.
NUREG/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES NUREC/CR-2713: VAPOR DEPOSITION VELOCITY MEASUREMENTS AND 137
CORRELATIONS FOR I(2)AND CsI.
NUREG/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATF GEOMETRY.
DIVISION OF FAC ILITY OPERATIONS NUREG/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER TESTING.
NUREG/CR-2297: SECURITY MANAGEMENT TECHNIQUES AND EVALUATIVE CHECKLISTS FOR SECURITY FORCE EFFECTIVENESS.
DIVISION OF HEALTH, SITING & ENVIRONMENT NUREG/CR-2343: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIDACTIVl?
WASTE: THE DNET COMPUTER CODE USER 'S MANU AL.
NUREC/CR-2359: ATMOSPHERIC STRUCTURE PRIOR TO TORNADOES AF DERIVED FROM PROXIMITY AND PRECEDENT UPPER AIR SOUNDINGS.
NUREG/CR-2381: GEOLOGIC AND HYDROLOGIC RESE ARCH AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER WEST VALLEY, NEW YORK. Prog ress R eport, August 1979-July 1981.
NUREG/CR-2521: METHOD FOR ESTIMATING WAKE FLOW AND EFFLUENT DISPERSION NEAR SIMPLE BLOCK-LIKE DUILDINGS.
NUREG/CR-2564: ENVIRONMENTAL FACTORS AFFECTING LONG-TERM STABILIZATION OF RADON SUPPRESSION COVERS FOR URANIUM MILL TAILINGS.
NUREG/CR-2584: METEROLOGIC AL CONSIDER ATIONS IN THE DEVELOPMENT OF A REAL-TIME ATMOSPHERIC DISPERSION MODEL FOR REACTOR EFFLUENT EXPOSURE PATHWAY.
NUREG/CR-2589: A GROUND-PENETRATING R ADAR SURVEY OF THE MAXEY FLATS LOW-LEVEL NUCLEAR WASTE DISPOSAL SI TE. FLEMING COUNTY, KENTUCKY.
NUREG/CR-2625: CRITICAL P ATHWAYS OF R ADIONUCLIDES TO MAN FROM AGRO-ECOSYSTEMS. Annual Progre ss Rep ort,0c tober 1980-Sep temb er 1981.
NUREG/CR-2637: EMERGENCY RESPONSE CAP ADILIT IES AND EX AMPLE ASSESSMENTS FOR AIRDORNE RADIONUCLIDE DISCHARGES.
NUREG/CR-2630: SNOW LOADS FOR THE DESIGN OF NUCLEAR POWER PLANT STRUC TURES.
NUREG/CR-2639: HISTORICAL EXTREME WINDS FOR THE UNITED STATES-ATLANTIC AND GULF DF MEXICO COASTL INES.
NUREG/CR-2642: LONG-TERM SURVIVADILITY OF R IPRAP FOR ARMORING URANIUM MILL TAILINGS AND COVERS: A LITERATURE REVIEW.
NUREG/CR-2653: EARTH RESISTIVITY AS A TOOL FOR SHALLOW EXPLOR ATION IN THE REELFOO T LAKE AREA, TENNESSEE.
NUREG/CR-2727 VO1: ECOLOGICAL STUDIES OF WOOD-DORING DIVALVES IN THE VICINITY OF THE OYSTER CREEK NUCLEAR GENERATING STATION. Pro gress Report, September-November 1981.
NUREG/CR-2737: EVALUATION OF DULK PROPERTIES OF RADWASTE GLASS AND CERAMIC CONTAINER MATERI ALS TO DETERMINE LONG-TERM STADILITY.
NUREG/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH THE 38TH PARALLEL LINEAMENT.
Final Technical Report, June 1979-June 1981.
DIVISION OF RISK ANALYSIS NUREC/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDIA: Technic al Review of NUREG/CR-1636, Vuli.
1,2 and 3, December 1,1981-March 31,1982.
NUREG/CR-2099: COMMON CAUSE FAULT RATES FOR DIESEL GENERATORS: ESTIMATES DASED ON LICENSEE EVENT REPORTS AT U. S.
COMMERCIAL NUCLEAR POWER PLANTS, 1976-1978.
NUREG/CR-2497 VO1: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1969-1979. A Status Report.Vol.
- 1. Main Report And App.
A.C.D And E.
NUREG/CR-2497 VO2: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENT: 1969-1979 A Status Report.Vol. 2 - Appendix D.
NURE0/CR-2651: ACCIDENT QENERATED PARTICULATE MATERIALS AND THEIR CHARACTERISITICS--A REVIEW OF BACKGROUND INFORMATION.
DIVISION OF ENGINEERING TECHNOLOGY NURE0/CR-2015 VO4: SEISMIC SAFETY MARODE RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
I NUREO/CR-2223: AN EVALUATION OF THE SOLID ANGLE METHOD USED IN NUCLEAR CRITICALITY SAFETY.
NURES/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED JOR PIPING RELIABILITY ASSESSMENT IN LIGH'i WATER REACTORS.
NUREO/CR-2522: EVALUATION OF NUCLEAR FACILITY DECOM'11SSIONING PROJECTS PROGRAM PLAN.
NURES/CR-2570: EXPERIMENTAL INVESTIGATION OF TEARING INSTABILITY PHENOMENA FOR STRUCTURAL MATERIALS.
NURE0/CR-2788: STRENGTH AND STIFFNESS OF UNIAXIALLY TENSIONED REINFORCED CONCRETE PANELS SUBJECTED TO MEMBRANE SHEAR.
(
NUREQ/CR-2790: AUTOMOBILE INACT FORCES ON CONCRETE WALL PANELS.
EDO-RESOURCE MANAGEMENT DIVISION OF DATA AUTOMATION & MANAGEMENT INFORMATION NUREG/CR-2201: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1978.
l OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)
DIVISION OF ENGINEERING NUREQ/CR-1890: ABS, SRSS AND CDF RESPONSE COMBINATION EVALUATION FOR MARK III CONTAINMENT AND DRYWELL STRUCTURES.
NUREQ/CR-203i: DYNAMIC COMBINATIONS FOR MARK II CONTAINMENT STRUCTURES.
NUREQ/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Te chnical Report. Aug ust 1980-Sep tember 1981.
NUREQ/CR-2569: RESPONSE OF THE ZION & INDIAN POINT CONTAINMENT BUILDINGS TO SEVERE ACCIDENT PRESSURES.
NUREQ/CR-2591: ESTIMATING TE POTENTI AL INDUSTRIAL IMPACTS OF A NUCLEAR REACTOR ACCIDENT.
NUREQ/CR-2686: REVIEW OF LOAD COMBINATIONS FOR NSSS AND BOP PIPINC AND EQUIPMENT OF MARK III PLANTS.
DIVISION OF SYSTEMS INTEGRATION (POST 811005)
NUREG/CR-2387: CREDIBLE ACCIDENT ANALYSES FOR TRIGA AND TRIGA-FUELED REACTORS.
NUREQ/CR-2413: SURVEY OF REMOTE AREA MONITORING SYSTEMS AT U. S.
LIGHT-WATER-COOLED POWER REACTORS.
NURE0/CR-2495: CHARACTERIZATION OF SOIL TO PLANT TRANSFER COEFFICIENTS FOR STABLE CESIUM AND STRONTIUM.
NUREQ/CR-2612: VARI ABILITY IN DOSE ESTIMATES ASSOCIATED WITH THE FOOD CHAIN TRANSPORT AND INGESTION OF SELECTED RADIONUCLIDES.
NUREQ/CR-2629: INTERIM SOURCE TERM ASSUNTIONS FOR EMERGENCY PLANNINC AND EQUIPMENT GUALIFICATION.
NUREG/CR-2633: CONTAINMENT REACTOR CAVITY SUBCOMPARTMENT ANALYSIS PROCEDURES FOR A BOILING WATER REACTOR.
NUREG/CR-2644: AN ASSESSMENT OF OFFSITE.REAL-TIME DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS.
NUREG/CR-2760: ASSESSMENT OF SCALE EFFECTS ON VORTEXING, SWIRL,AND INLET LOSSES IN LARGE SCALE SUMP MODELS.
NUREQ/CR-2772: HYDRAULIC PERFORMANCE OF PUMP SUCTION INLETS FOR EMERGENCY CORE COOLING SYSTEMS IN BOILING WATER REACTORS.
139
DIVISION OF LICENSING NUREQ/CR-2664: SELECTED REVIEW OF FOREIGN LICENSING PRACTICES FOR NUCLEAR POWER PLANTS.
DIVISION OF SAFETY TECHNOLOGY NUREQ/CR-2403 SO1: SURVEY OF INSULATION USED IN NUCLEAR POWER PLANTS AND THE POTENTIAL FOR DEBRIS GEERATION.
140
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.
l l
ALDEN RESEARCH LABORATORY NUREG/CR-2760: ASSESSMENT OF SCALE EFFECTS ON VORTEXING, SWIRL,AND INLET LOSSES IN LARGE SCALE SUMP NODELS.
NUREC/CR-2772: HYDRAULIC PERFORMANCE OF PUMP SUCTION INLETS FOR EMERGENCY CORE COOLING SYSTEMS IN BOILING WATER REACTORS.
AMERICAN NUCLEAR SOCIETY NUREG/CR-23OO VO1 R1: DRAFT: PRA PROCEDURE GUIDE. A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
APPLIED SCIENCE ASSOCIATES, INC.
NUREC/CR-2297: SECURITY MANAGEMENT TECHNIQUES AND EVALUATIVE CHECKLISTS FOR SECURITY FORCE EFFECTIVENESS.
ARGONNE NATIONAL LABORATORY NUREC/CR-1622: FLOW MEASUREMENT BY PULSED-NEUTRON ACTIVATION TECHNIGUES AT THE PKL FACILITY AT ERLANGEN (GERMANY).
NUREC/CR-2181 VO4: PHYSICS OF REACTOR SAFETY. Guarterly Report,0ctober-December 1981.
NUREG/CR-2647: CRITICAL HEAT FLUX EXPERIMENTS UNDER LOW FLOW CONDITIONS IN A VERTICAL ANNULUS.
ARIZONA, UNIV. OF NUREG/CR-2518: THERMODYNAMIC PROPERTIES OF WATER FOR COMPUTER SIMULATION OF POWER PLANTS.
ARMY, DEPT. OF, ARMY ENGINEER WATERWAYS EXPERIMENT STATION NUREG/CR-2700: PARAMETERS FOR CHARACTERIZING SITES FOR DISPOSAL OF LOW-LEVEL RADIOACTIVE WASTE.
BABCOCK & WILCOX CD.
NUREC/CR-2597: STEADY-STATE PRESSURE LOSSES FOR MULTIROD BURST TEST (MRBT) BUNDLE B-5.
BATTELLC MEMORIAL INSTITUTE, COLUMBUS LABORATOR IES NUREC /CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS I
PROGRAM: Calvert Cliffs No. 2 PWR Power Plant.
NUREC/CR-2682: CITADEL: A COMPUTER CODE FOR THE ANALYSIS OF IODINE l
BEHAVIOR IN STEAM GENERATOR TUBE RUPTURE ACCIDENTS.
NUREC/CR-2683: IODINE BEHAVIOR IN STEAM GENER ATOR TUBE RUPTURE ACCIDENTS.
NUREC/CR-2713: VAPOR DEPOSITION VELOCITY PEASUREMENTS AND CORRELATIONS FOR I(2)AND CrI.
BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORY NUREG-0868: A COLLECTION OF MATHEMATICAL MODELS FOR DISPERSION IN SURFACE WATER AND GROUNDWATER.
I l
141 l
l
f NUREC/CR-1030 VO2: SEDIMENT AND R ADIONUCLIDE TR ANSPORT IN RIVERS. Phase 2-Field Sampling Program For Cattaraugus And Buttermilk Creek s New York.
NUREG/CR-2019: THIRD PHASE OF POCKET-SIZED ELECTRONIC DOSIMETER TESTING.
NUREG/CR-2022: TECHNICAL REVIEW OF THE DISPERSION AND DOSE MODELS USED IN THE MILDOS COMPUTER PROGRAM.
NUREG/CR-2201: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1978.
NUREQ/CR-2387: CREDIBLE ACCIDENT ANALYSES FOR TRIGA AND TRIGA-FUELED REACTORS.
NUREG/CR-2413: SURVEY OF REMOTE AREA NONITORING SYSTEMS AT U. S.
LIGHT-WATER-COOLED POWER REACTORS.
NUREG/CR-2432: A UNIGUE CONCEPT FOR LIGUID LEVEL AND VOID FRACTION DETECTION IN SEVERE FUEL DAMAGE TESTS.
NUREG/CR-2460: TECHNICAL SUPPORT FOR IMPROVING THE LICENSING REGULATORY BASE FOR SELECTED FACILITIES ASSOCIATED WITH THE FRONT END OF THE FUEL CYCLE.
NUREG/CR-2564: ENVIRONMENTAL FACTORS AFFECTING LONG-TERM STABILIZATION OF RADON SUPPRESSION COVERS FOR URANIUM MILL TAILINGS.
NUREG/CR-2567: FINAL DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-432.
NUREG/CR-2600: END-OF-IRRADIATION DATA REPORT FOR THE INSTRUMENTED FUEL ASSEMBLY (IFA)-527.
NUREG/CR-2642: LONG-TERM SURVIVABILITY OF RIPRAP FOR ARMORING URANIUM MILL TAILINGS AND COVERS: A LITERATURE REVIEW.
NUREG/CR-2651: ACCIDENT GENERATED PARTICULATE MATERIALS AND THEIR CHARACTERISITICS--A REVIEW OF B ACKGROUND INFORMATION.
BROOKHAVEN NATIONAL LABORATORY NUREG/CR-1890: ABS. SRSS AND CDF RESPONSE COMBINATION EVALUATION FOR MARK III CONTAINMENT AND DRYWELL STRUCTURES.
NUREG/CR-2039: DYNAMIC COMBINATIONS FOR MARK II CONTAINMENT STRUCTUREG.
NUREG/CR-2192 VO1 N2: EVALUATION OF ISOTOPE MIGRATION-LAND BURI AL. Guar te rly Progress Rep or t, April-June 1981.
NUREC/CR-2193 VO1 N2: PROPERTIES OF RADIOACTIVE WASTES AND WASTE CONTAINERS. Guarterly Progress Rep or t', April-June 1981.
NUREC/CR-2317 VO1 N3: CONTAINER ASSESSMENT-CORROSION STUDY OF HLW CONTAINER MATERI ALS. Guarterly Progres s Report July-Sep tember 1981.
NUREC/CR-2317 VO1 N4: CONTAINER ASSESSMENT-COHROSION STUDY OF HLW CONTAINER MATERIALS.Guarterly Progress Report,0ctober-December 1981.
NUREC/CR-2416: INITI AL GUANTIFICATION OF HUMAN ERROR ASSOCI ATED WITH SPECIFIC INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS IN LICENSED NUCLEAR POWER PLANTS.
NUREG/CR-2417: IDENTIFICATION AND ANALYSIS OF HUMAN ERRORS UNDERLYING PUMP AND VALVE RELATED EVENTS REPORTED BY NUCLEAR POWER PLANT LICENSEES.
NUREG/CR-2516 VO1 N1: CHARACTERIZATION OF TMI-TYPE WASTES AND SOLID PRODUCTS. Guarterly Progress Report, April-September 1981.
NUREG/CR-2542: SENSITIVITY STUDY USING THE FR ANTIC CODE FOR THE I
NUREG/CR-2543:
UNAVAILABILITY OF A SYSTEM TO THE FAILURE CHARACTERISTICS OF THE COMPONENTS AND THE OPERATING CONDITIONS.
A STUDY OF THE FEASIBILITY OF MICROWAVE DIELECTRIC HEATING FOR LMFBR TRANSITION PHASE ACCIDENT SEGUENCE BOILING STUDIEG.
NUREC/CR-2685: EVALUATION OF CONCURRENT PEAK RESPONSES.
NUREG/CR-2686: REVIEW OF LOAD COMBINATIONS FOR NSSS AND BOP PIP ING AND EQUIPMENT OF MARK III PLANTS.
BURNS & ROE CD.
NUREG/CR-2403 SO1: SURVEY OF INSULATION USED IN NUCLEAR POWER PLANTS AND THE POTENTI AL FOR DEBRIS GENERATION.
CATHOLIC UNIV.
l 1G
NUREG/CR-2737: EVALUATION OF BULK PROPERTIES OF RADWASTE CLASS AND CERAMIC CONTAINER MATERI ALS TO DETERMINE LONG-TERM STABILITY.
CHIAPETTA, WELCH & ASSOCIATES,LTD.
)
NUREG/CR-2790: AUTOMOBILE IMPACT FORCES ON CONCRETE WALL PANELS.
COMMERCE DEPT. OF, NATIONAL OCEANOGRAPHIC & ATMOSPHERIC ADMINISTR ATION NUREG/CR-2521: METHOD FOR ESTIMATING WAKE FLOW AND EFFLUENT DISPERSION NEAR SIMPLE BLOCK-LIKE BUILDINGS.
NUREG/CR-2584: METEROLOGICAL CONSIDERATIONS IN THE DEVELOPMENT OF A REAL-TINE ATMOSPHERIC DISPERSION MODEL FOR REACTOR EFFLUENT EXPOSURE PATHWAY.
NUREQ/CR-2637: EMERGENCY RESPONSE CAPABILITIES AND EXAMPLE ASSESSMENTS FOR AIRBORNE RADIONUCLIDE DISCHARGES.
NUREG/CR-2639: HISTORICAL EXTREME WINDS FOR THE UNITED STATES-ATLANTIC AND GULF OF MEXICO COASTLINES.
COMMERCE, DEPT.0F NUREG/CR-2591: ESTIMATING THE POTENTIAL INDUSTRIAL IMPACTS OF A NUCLEAR REACTOR ACCIDENT.
COMMERCE, DEPT.0F, NATIONAL BUREAU OF STANDARDS NUREG/CR-2638: SNOW LOADS FOR THE DESIGN OF NUCLEAR POWER PLANT STRUCTURES.
CORNELL UNIV.
NUREC/CR-2788: STRENGTH AND STIFFNESS OF UNIAXIALLY TENSIONED REINFORCED CONCRETE PANELS SUBJECTED TO MEMBRANE SHEAR.
DAVID W.
TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER NUREC/CR-2570: EXPERIMENTAL INVESTIGATION OF TEARING INSTABILITY PHENOMENA FOR STRUCTURAL MATERI ALS.
EG&G, INC.
NUREG/CR-0169 V13: LOFT EXPERIMENTAL MEASUREMENTS UNCERTAINTY ANALYSES.
Volume XIII. Temperature Measurements.
NUREC/CR-1826 VO1: RELAPS/ MOD 1 CODE MANUAL. Volume 1: System Models And Numerical Methods.
NUREG/CR-1826 VO2: RELAP5/ MOD 1 CODE MANUAL. Vo lume 2: User 's Guid e And Input Requirements.
NUREQ/CR-2059: COMPILATION OF DATA CONCERNING KNOWN AND SUSPECTED WATER HAMMER EVENTS IN NUCLEAR POWER PLANTS (CY 1969-MAY 1981).
NUREG/CR-2099: COMMON CAUSE FAULT RATES FOR DIESEL GENERATORS: ESTIMATES BASED ON LICENSEE EVENT REPORTS AT U. S.
COMMERCIAL NUCLEAR POWER PLANTS, 1976-1978.
NUREG/CR-2496: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
NUREG/CR-2618: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A NATURAL CIRCULATION TEST S-NC-7C.
NUREG/CR-2636: EXPERIMENTAL DATA REPORT FOR AIR-WATER FLOODING TESTS OF THE FLECHT-SEASET PROGRAM 1;ET FACILITY VESSEL UPPER PLENUM.
NUREC/CR-2648: EXPERIMENTAL DATA REPORT FOR SEMISCALE MOD-2A NATURAL CIRCULATION TEST SERIES (TESTS S-NC-8D AND S-NC-9).
NUREG/CR-2711: PERFORMANCE AND DESIGN REGUIREMENTS FOR A GRAPHICS DISPLAY RESEARCH FACILITY.
NUREG/CR-2717: EXPERIMENT DATA REPORT FOR LOFT ANTICIPATED TRANSIENT WITHOUT SCRAM EXPERIMENT L9-3.
NUREC/CR-2732: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2A INTERMEDIATE BREAK TEST SERIES.(Tests S-IB-1 And S-IB-2).
EXXON NUCLEAR CO.,
INC. (SUBS. OF EXXON CORP.)
NUREG/CR-2495: CHARACTERIZATION OF SOIL TO PLANT TRANSFER COEFFICIENTS FOR STABLE CESIUM AND STRONTIUM.
NUREC/CR-2644: AN ASSESSMENT OF OFFSITE. REAL-TIME DOSE-MEASUREMENT SYSTEMS FOR EMERGENCY SITUATIONS.
GENERAL ELECTRIC CO.
NUREG/CR-2133: BWR REFILL-RELOAD PROGRAM TASK 4. 4 - 30 SSTF CISCRIPTION DOCUMENT.
143
NUREQ/CR-2229 VO1: BWR LARGE BREAK SIMULATION TESTS--BWR BLOWDOWN / EMERGENCY CORE COOLING PROGRAM.
NUREG/CR-2231: BWR LOW FLOW BUNDLE UNCOVERY TEST AND ANALYSIS.
GEO-CENTERS, INC.
NUREG/CR-2589: A GROUND-PENETRATING RADAR SURVEY OF THE MAXEY FLATS LOW-LEVEL NUCLEAR WASTE DISPOSAL SITE FLEMIhD COUNTY, KENTUCKY.
INSTITUTE OF ELECTRICAL & ELECTRONIC ENGINEERS NUREG/CR-23OO VO1 R1: DRAFT:PRA PROCEDURE QUIDE. A Guide To The Performance Of Probabilistic Risk Assessments For Nuclear Power Plants.
IOWA STATE UNIV.
NUREG/CR-2442: RELIABILITY ANALYSIS OF STEEL CONTAINMENT STRENGTH. Tech nical Report. Augus t 1980-Sep te mb er 1981.
LAWRENCE LIVERMORE LABORATORY NUREG/CR-1233 VO4: THE STRUCTURED ASSESSMENT APPROACH. VERSION 1 COMPUTATION AL ANALYSIS PACKAGE.
NUREQ/CR-2015 VO4: SEISMIC SAFETY MARGINS RESEARCH PROGRAM PHASE I FINAL REPORT - SOIL STRUCTURE INTERACTION (PROJECT III).
LEHIGH UNIV.
NUREG/CR-2727 VO1: ECOLOGICAL STUDIES OF WOOD-BORING BIVALVES IN THE VICINITY OF THE OYSTER CREEK NUCLEAR GENERATING STATION. Progress Report, September-November 1981.
LOS ALAMOS SCIENTIFIC LABORATORY NUREG/CR-2281 VO2: NUCLEAR REACTOR SAFETY. Apr il 1-June 30,1981.
NUREG/CR-2281 VO3: NUCLEAR REACTOR SAFETY. July 1-September 30,1981.
I NUREG/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELI ABILITY ASSESSMENT IN LIGHT WATER REACTORS.
NUREG/CR-2434: FRAC (FAILURE RATE ANALYSIS CODE): A COMPUTER PROGRAM FOR ANALYSIS OF VARI ANCE OF FAILURE R ATES. An Application User 's Guide.
NUREG/CR-2464: METHODS FOR CLASSIFYING MIXTURES OF EXPONENTIAL DISTRIBUTIONS BASED ON EITHER EXPONENTI AL OR POISSON DATA.
NUREC/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED TORNADO CONDITIONS.
NUREG/CR-2569: RESPONSE OF THE ZION & INDIAN POINT CONTAINMENT BUILDINGS TO SEVERE ACCIDENT PRESSURES.
NUREG/CR-2622: ANALYSIS OF TRAC AND SCTF RESULTS FOR SYSTEM PRESSURE-EFFECTS TESTS UNDER FORCED FLOODING (RUNS 506,507 AND 500).
NUREG/CR-2632: RESPONSE OF CENTRIFUGAL BLOWERS TO SIMULATED TORNADO TRANSIENTS. July-September 1981.
NUREG/CR-2633: CONTAINMENT REACTOR CAVITY SUBCOMPARTMENT ANALYSIS i
PROCEDURES FOR A BOILING WATER REACTOR.
NUREG/CR-2652: EVALUATION AND PERFORMANCE OF CLOSED-CIRCUIT BREATHING APPARATUS.
LOVELACE BIOMED & ENVIRONMENTAL RESEARCH INSTITUTE f
NUREG/CR-2512: RADIATION DOSE ESTIMATES AND HAZARDS EVALUATIONS FOR 1
INHALED AIRBORNE RADIONUCLIDES. Annual Progress Report July 1980-June 1981.
MARVIKEN NUREC/CR-2671: THE MARVIKEN FULL SC ALE CRITIC AL FLOW TESTS. Summary Report.
MARYLAND, STATE OF NUREG/CR-2699: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MARYLAND. June 1980-June 1981.
MICHIGAN, STATE OF l
NUREG/CR-2736: TRANSPORTATION OF RADIOACTIVE MATERIAL IN MICHIGAN. September 1980-August 1981.
MISSOURI, UNIV. OF, COLUMBIA NUREG/CR-2359: ATMOSPHERIC STRUCTURE PRIOR TO TORNADOES AS DERIVED FRON PROXIMITY AND PRECEDENT UPPER AIR SOUNDINGS.
144
MITRE CORP.
NUREG/CP-OO26: WORKSHOP ON PSYCHOLOGICAL STRESS ASSOCIATED WITH THE PROPOSED RESTART OF THREE MILE ISLAND, UNIT 1.
NEW YORK, STATE OF NUREG/CR-2381: GEOLOGIC AND HYDROLOGIC RESEARCH AT THE WESTERN NEW YORK NUCLEAR SERVICE CENTER WEST VALLEY, NEW YORK. Progress Report, August 1979-July 1981.
NORTHEAST MISSOUR I STATE UNIV.
NUREG/CR-2565: STRUCTURAL PERFORMANCE OF HEPA FILTERS UNDER SIMULATED TORNADO CONDITIONS.
NORTHWESTERN UNIV.
NUREG/CR-2334: INTERPHASE TRANSPORT IN HORIZONTAL STRATIFIED CONCURRENT FLOW.
NUREG/CR-2783: COUNTERCURRENT STEAM-WATER FLOW IN A FLAT PLATE GEOMETRY.
NUCLEAR ASSURANCE CORP.
NUREG/CR-2704:
U. S.
REACTOR SPENT-FUEL STORAGE CAPABILITIES.
DAK RIDGE NATIONAL LABORATORY NUREG/CP-OO22: PROCEEDINGS OF THE SYMPOSIUM ON UNCERTAINTIES ASSOCIATED WITH THE REGULATION OF THE GEOLOGIC DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTE.
NUREC/CR-02OO ERR: SCALE: MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LICENSING EVALUATION.
NUREC/CR-2OOO VO1 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month OF February 1982.
NUREG/CR-2OOO VO1 N3: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of March 1982.
NUREC/CR-2OOO VO1 N4: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of April 1982.
NUREG/CR-2 COO VO1 N5: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of May 1982.
NUREG/CR-2053: HEAT TRANSFER ANALYSIS OF THE LWR PRESSURE VESSEL STEEL IRRADIATION CAPSULES IN THE DAK RIDGE RESEARCH REACTOR-PRESSURE VESSEL BENCHMARK FACILITY.
NUREG/CR-2141 VO4: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM. Guart erly Progress Report For October-December 1981.
NUREG/CR-2172:
SUMMARY
AND BIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT BOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREG/CR-2172 ERR:
SUMMARY
AND BIBIOLOGRAPHY OF SAFETY-RELATED EVENTS AT BOILING-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREC/CR-2173:
SUMMARY
AND BIBLIOGR APHY OF SAFETY-RELATED EVENTS AT PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREC/CR-2173 ERR:
SUMMARY
AND BIBLIOGRAPHY OF SAFETY-RELATED EVENTS AT PRESSURIZED-WATER NUCLEAR POWER PLANTS AS REPORTED IN 1980.
NUREC/CR-2184: COMPARISON OF THE RADIOLOGICAL IMPACTS OF THORIUM AND URANIUM NUCLEAR FUEL CYCLES.
NUREC/CR-2204 VO4; ADVANCED TWO-PHASE FLOW INSTRUMENTATION PROGRAM.Guarterly Progress Report,0ctober-December 1981.
NUREC/CR-2220 VO2: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
NUREC/CR-2220 VO3: THE IMPACT OF ENTRAINMENT AND IMPINGEMENT ON FISH POPULATIONS IN THE HUDSON RIVER ESTUARY.
NUREG/CR-2221 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF REACTOR SAFETY RESEARCH. Guarterly Progres s Report,0ctober 1-December 31,1981.
NUREC/CR-2223: AN EVALUATION OF THE SOLID ANGLE METHOD USED IN NUCLEAR CRITICALITY S AFETY.
NUREC/CR-2299 VO4: AEROSOL RELEASE AND TRANSPORT PROGRAM. Guarterly Progress Report For October-December 1981.
NUREG/CR-2306: CSRL-V: PROCESSED ENDF/B-V 227-NEUTRON-GROUP AND 146
POINTWISE CROSS-SECTION LIBRARIES FOR CRITICALI,TY SAFETY, REACTOR AND SHIELDING STUDIES.
NUREG/CR-2353 VO2: SPECIFICATION AND VERIFICATION OF NUCLEAR POWER PLANT TRAINING SIMULATOR RESPONSE CHARACTER ISTICS. Part II:
Conclusions And Recommendations.
NUREG/CR-2362: RELATIONSHIPS DETWEEN CHARPY V-NOTCH IMPACT ENERGY AND FRACTURE TOUGHNESS.
NUREG/CR-2366 VO2: MULTIROD DURST TEST PROGRAM PROGRESS REPORT FOR JULY-DECEMBER 1981.
NUREG/CR-2392:
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HTGR SAFETY RESEARCH, JULY 1974-SEPTEMBER 1980.
NUREC/CR-2393: FUEL AEROSOL SIMULANT TEST DATA RECORD REPORT:
UNDERWATER TESTS.
NUREC/CR-2435: DISPERSED FLOW FILM DOILING IN ROD BUNDLE GEOMETRY-STEADY STATE HEAT TRANSFER DATA AND CORRELATION COMP ARISONS.
NUREG/CR-2455: EXPERIMENTAL INVESTIGATIONS OF BUNDLE DOILOFF AND REFLOOD UNDER HIGH-PRESSURE LOW HEAT FLUX CONDITIONS.
NUREG/CR-2456: EXPERIMENTAL INVESTIGATIONS OF UNCOVERED-BUNDLE HEAT TRANSFER AND TWO-PHASE MIXTURE-LEVEL SWELL UNDER HIGH-PRESSURE LOW HEAT-FLUX CONDITIONS.
NUREG/CR-2469: AN ANALYSIS OF TRANSIENT FILM DOILING OF HIGH-PRESSURE WATER IN A ROD DUNDLE.
NUREG/CR-2470: THERMOMETRY IN THE MULTIROD BURST TEST PROGRAM.
NUREG/CR-2493: AGUEOUS IODINE CHEMISTRY IN LWR ACCIDENTS: Review And Assessment.
NUREG/CR-2494: OR-FLAW: A FINITE ELEMENT PROGRAM FOR DIRECT EVALUATION OF K-FACTORS FOR USER-DEFINED FLAWS IN PLATES, CYLINDERS AND PRESSURE-VESSEL NOZZLE CORNERS.
NUREG/CR-2505: ELECTRICAL IMPEDANCE STRING PROBES FOR TWO-PHASE VOIDS AND VELOCITY MEASUREMENTS.
NUREG/CR-2525 VO1: ORNL ROD DUNDLE HEAT TRANSFER TEST DATA. Volume 1-ORNL Small Dreak LOCA Test Series I: Experimental Data Report.
NUREG/CR-2525 VO2: ORNL ROD BUNDLE HEAT TRANSFER TEST DATA. Volume 2 -
Thermal -Hydraulic Test Facility Experimental Data Report for Test 3.03.6AR - Transient Film Doiling In Upflow.
NUREG/CR-2525 VO3: ORNL ROD DUNDLE HEAT TRANSFER TEST DATA. Volume 3-Thermal-Hyd raulic Test Facility Exp erimen tal Data Report For Test 3.06.6D-Transient Film Doiling In Upflow.
NUREC/CR-2525 VOS: ORNL ROD DUNDLE HEAT TRANSFER TEST DATA. Volume 5-Thermal-Hyd raulic Test Facility Exp erimental Data Report For Test
- 3. 08. 6C-Trans ient Film Doiling In Upflow.
NUREG/CR-2525 VO7: ORNL ROD DUND'.E HEAT TRANSFER TEST DATA. Volume 7-Thermal-Hydraulic Test Facility Experimental Data Support For Test Series 3.07.9-Steady-State Film Boiling In Upflow.
NUREG/CR-2544: TWO-PHASE MASS FLUX UNCERTAINTY ANALYSIS FOR THERMAL-HYDRAULIC TEST FACILITY INSTRUMENTED SPOOL PIECES.
NUREG/CR-2545: DESIGN CONCEPT AND TESTING OF AN IN-BUNDLE CAMMA DENSITOMETER FOR SUDCHANNEL VOID FRACTION MEASUREMENTS IN THE THTF ELECTRICALLY HEATED ROD BUNDLE.
NUREG/CR-2586: A SURVEY OF METHODS FOR IMPROVING OPERATOR ACCEPTANCE OF COMPUTERIZED AIDS.
NUREC/CR-2587: FUNCTIONS AND OPERATIONS OF NUCLEAR POWER PLANT CREWS.
NUREG/CR-2610: RAGBEEF: A FORTRAN IV IMPLEMENTATION OF A TIME-DEPENDENT MODEL FOR RADIONUCLIDE CONTAMINATION OF BEEF.
NUREG/CR-2612: VARI ABILITY IN DOSE ESTIMATES ASSOCI ATED WITH THE FOOD CHAIN TRANSPORT AND INGESTION OF SELECTED R ADIONUCLIDES.
NUREG/CR-2629: INTERIM SOURCE TERM ASSUMPTIONS FOR EMERGENCY PLANNING AND EQUIPMENT GUALIFICATION.
NUREG/CR-2692: AN INTEGRATED SYSTEM FOR FOREC ASTING ELECTRIC ENERGY AND LOAD FOR STATES AND UTILITY SERVICE AREAS.
146.
i L
NUREG/CR-2696: CALCULATIONS OF TWO SERIES OF EXPERIMENTS PERFORMED AT THE POOLSIDE FACILITY USING THE DAK RIDGE RESEARCH REACTOR.
PURDUE UNIV.
NUREG/CR-2741: A TECTONIC STUDY OF THE EXTENSION OF THE NEW MADRID FAULT ZONE NEAR ITS INTERSECTION WITH TFE 38TH PARALLIL LINEAMENT.
Final Technic al Report. June 1979-June 1981.
RADI ATION MANAGEMENT CORP.
NUREG/CR-2722: RADIOLOGICAL SURVEY OF THE WEST LAKE LANDFILL, ST. LOUIS COUNTY. MISSOURI.
l SANDIA LABORATORIES NUREG/CR-1245 RO1: CORRECTIONS AND ADDITIONS TO USER 'S GUIDE FOR SNAP.
( NUREG/CR-124 5, SAND 80-0315 ).
NUREC/CR-1594 VO4: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER-DECEMRER 1980.
NUREG/CR-1636 VO4: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: EFFECTS OF VARIABLE HYDROLOGIC PATTERNS ON THE ENVIRONMENTAL TRANSPORT MODEL.
NUREC/CR-1659 VO3: REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS; PROGRAM: Calvert Cliffs No. 2 PWR Power Plant.
NUREG/CR-1820: STATUS REPORT ON THE FISSION-PRODUCT RESEARCH PROGRAM.
NUREG/CR-1851: REACTOR PHYSICS DESIGN CALCULATIONS FOR THE ACPR UPGRADE.
NUREG/CR-2194: CONTAINMENT RESEARCH PRIORITIES.
NUREG/CR-2238 VO1: ADVANCED REACTOR SAFETY RESEARCH. Guarterly Report, January-March 1981.
I NUREG/CR-2279: WATER RELEASE FROM HEATED CONCRETES.
I NUREC/CR-2283: DIRECT OBSERVATION OF MELT BEHAVIOR DURING HIGH TEMPERATURE MELT / CONCRETE INTER ACTIONS.
NUREG/CR-2314: AGING WITH RESPECT TO FLAMMABILITY AND OTHER PROPERTIES IN FIRE-RETARDED ETHYLENE PROPYLENE RUBBER AND CHLOROSULFONATED POLYETHYLENE.
NUREG/CR-2343: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADIOACTIVE WASTE: THE DNET COMPUTER CODE USER'S MANUAL.
NUREG/CR-2350: SENSITIVITY ANALYSIS TECHNIGUES: SELF-TEACHING CURRICULUM.
NUREC/CR-2377: TESTS & CRITERIA FOR FIRE PROTECTION OF CABLE PENETRATIONS.
NUREG/CR-2393 ERR: Errata, changing rept number to NUREC/CR-2593,to A USER'S MANUAL FOR COMPUTER CODE RIBD/IRT.
NUREG/CR-2394 ERR: Errata, changing rept numb er to NUREG/CR-2594, to A USER 'S MANUAL FOR THE CAB AS SPECTRUM COMPUTER CODE.
NUREC/CR-2412: HEAT REMOVAL FROM A STRATIFIED UO2-SODIUM PARTICLE BED.
NUREC/CR-2431: BURN MODE ANALYSIS OF HORIZONTAL CABLE TRAY FIRES.
NUREC/CR-2473: SIMMER ANALYSIS OF PROMPT BURST ENERGETICS EXPERIMENTS.
NUREG/CR-2481: LIGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semi annual Report. April-September 1981.
NUREG/CR-2546: REACTOR SAFEGUARDS AGAINST INSIDE SABOTAGE.
NUREG/CR-2551: RANK ORDERING OF VITAL AREAS WITHIN NUCLEAR POWER PLANTS.
NUREG/CR-2559: RESULTS OF PHASE ONE OF PLANT ELECTRICAL SYSTEM (PES)
STUDY.
NUREC/CR-2581: SOME EFFECTS OF ELECTRONS SLOWING DOWN IN MATERI ALS WITH APPLICATION TO SAFETY-RELATED EQUIPMENT GUALIFICATION.
NUREG/CR-2582: RADIATION CAPABILITIES OF THE SANDIA HIGH INTENGITY ADJUSTABLE COBALT ARRAY.
NUREG/CR-2588: SECURITY OFFICER RESPONSE STRATEGIES (SECURORS).
NUREG/CR-2593: A USER 'S MANUAL FOR COMPUTER CODE RIBD/IRT.
NUREG/CR-2594: A USER 'S MANUAL FOR THE GABAS SPECTRUM COMPUTER CODE.
NUREG/CR-2604: THE SNAP OPERATING SYSTEM (SDS) USER 'S GUIDE.
NUREG/CR-2605: THE SNAP OPERATING SYSTEM REFERENCE MANUAL.
147
NUREC/CR-2611: MCO AND 70 WX UO2-30W% Y203: THERM 0 PHYSICAL AND TRANSIEN1 PROPERTIES.
NUREQ/CR-2681: ESTIMATED RECURRENCE FREQUENCIES FOR INITIATING ACCIDENT CATEGORIES ASSOCI ATED WITH THE CLINCH RIVER BREEDER REACTOR PLANT DESIGN.
SAVANNAH RIVER LABORATORY NUREG/CR-1681: WRAP-PWR VERIFICATION STUDIES.
NUREG/CR-2356: UPDATED INPUT FOR THE WR AP-EM SYSTEM.
NUREC/CR-2625: CRITICAL PATHWAYS OF RADIONUCLIDES TO MAN FROM AGRO-ECOSYSTEMS. Annual Progress Rep or t,0c to b er 1980-September 1981.
SCIENCE APPLICATIONS, INC.
NUREG/CR-1672 VO3: RISK ASSESSMENT METHODOLOGY DEVELOPMENT FOR WASTE ISOLATION IN GEOLOGIC MEDIA:Tec hnical Review of NUREG/CR-1636, Vols 1,2 and 3, Dec emb er 1,1981-March 31,1982.
NUREG/CR-2301: FRACTURE MECHANICS MODELS DEVELOPED FOR PIPING RELIABILITY ASSESSMENT IN LIGHT WATER REACTORS.
NUREQ/CR-2497 VO1: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1969-1979. A Status Report. Vol.
- 1. Main Report And App.
A,C,D And E.
NUREC/CR-2497 VO2: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENT: 1969-1979. A Status Rep ort. Vol. 2 - Appendix B.
STRUCTURAL MECHANICS ASSOCIATES NUREG/CR-2664: SELECTED REVIEW OF FOREIGN LICENSING PRACTICES FOR NUCLEAR POWER PLANTS.
UNITED NUCLEAR CORP.
NUREG/CR-2522: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS PROGRAM PLAN.
VANDERBILT UNIV.
NUREC/CR-2653: EARTH RESISTIVITY AS A TOOL FOR SHALLOW EXPLORATION IN THE REELFOOT LAKE AREA, TENNESSEE.
VIRGINIA, UNIV. OF NUREG/CR-2603: BUBBLE BEHAVIOR IN LMFBR CORE DISRUPTIVE ACCIDENTS.
t l
.148
Licensed Facility Index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are l
preceded by their Docket number and followed by the report number. If fur-l ther information is needed, refer to the main citation by the NUREG number.
40-2061 West Chicago Facility. Kerr-Mc Ge e Corp.,
MUREG-0904 40-8781 Teton Emploration Drilling Co.,
Inc..
NUREG-0925 50-000 Generic Docket NUREC/CR-2569 50-000 Generic Docket NUREG/CR-2569 50-000 Generic Docket NUREG/CR-2569 50-3 Indian Point Station, Unit 1. Consolidated Edison Co. of New York NUREC/CR-2569 50-27 Washington State Univ. Re s earc h Reactor NUREG-0911 50-83 Univ. of Florida Training Reactor NUREG-0913 50-219 Ogster Creet Nuclear Power Plant, Jersey Central Power & Light Co.
NUREG/CR-2727 VO!
50-244 Robert Emmet Ginna Nuclear Plant. Unit 1, Rochester Gas & Electric C NUREG-0821 DRFT 50-244 Robert Emmet Ginna Nuclear Plant. Unit 1.
Rochester Gas & Electric C NUREG-0909 50-244 Robert Emmet Ginna Nuclear Plant. Unit 1.
Rochester Gas & Electric C NUNEG-0916 50-244 Robert Emmet Ginne Nuclear Plant. Unit 1.
Rochester Gas & Electric C NUREG-0916 ERR 50-247 Indian Point Station. Unit 2. Consolidated Edison Co. of New York NUREC/CR-2569 50-235 Palisades Nuclear Plant. Consumers Power Co NUREG-OS2O DRFT 50-286 Indian Posnt Station. Unit 3. Power Authority of State of New York NUREC/CR-2569 50-289 Three Mile Island Nuclear Station, Unit 1 Metropolitan Edison Co.
%AREG/CP-OO26 50-289 Three Mile Island Nuclear Station. Unit 1,
Metropolitan Edison Co.
NUREG/CP-OO26 50-295 Za on Nuclear Power Station, Unit 1.
Commonwealth Edison Co.
NUREG/CR-2569 50-304 Zson Nuclear Power Station, Unit 2, Commonwealth Ed a son Co.
NUREG/CR-2569 50-318 Ca lvert Clif f s Nuclear Power Plant, Unit 2, Baltimore Gas & Elec tric NUREG/CR-1659 VO3 S0-318 Calvert Clif f s Nuclear Power Plant, Unit 2, Baltimore Gas & Llectric NUREG/CR-1659 VO3 50-329 Madiend Plant. Unit 1.
Consweers Power Co.
NUNEG-0793 50-329 Madland Plant, Unnt 1.
Consumers Power Co NUMEG-0793 SO1 50-330 Midland Plant. Unit 2, Consumers Power Co NudEG-0793 50-330 Madland Plant, Unit 2, Consumers Power Co NUREG-6793 SO!
S0-361 San Onof re Nuc lear Station, Unit 2. Southern California Ldison Co.
NUREG-0712 SO6 50-362 San Onof re Nuc lear Station, Unit 3. Southern California Edison Co NUREC-0712 S06 50-373 LaSalle County Station, Unit 1.
Commonwealth Edison Co NUREG-OS19 SO3 S0-373 LaSalle County Station. Unit 1 Commonwealth Edison to NUNEG-0861 S0-374 LaSalle County Station. Unit 2.
Commonwealth Edison C o.
NUHEG-0519 SO3 50-382 Waterford Generatang Station. Unnt 3. Lout s a ana Power & Light Co.
NUREC-0787 S03 j
50-3R9 St. Lucie Plant. Unit 2, Florida Power & Light Co NUREG-0842 50-390 Watts Bar Nuclear Plant, Unit 1.
Tennessee Valleg Authority NUREG-0847 50-391 Wa tts Bar Nuclear Plant. Unit 2.
Tennessee Valley Authoring NUNEG-0847 i
50-416 Gr and Gulf Nuc lose Station. Unit 1.
Mississippi Power & Light Co NUREG-0831 S02 50-416 Grand Gulf Nuclear Station. Unit 1.
Mississipp a Power & Light Co NUNEC-0926 50-417 Grand Gulf Nuclear Station, Unit 2, Mississippi Power & Light Co NUREG-0831 SO2 l
50-440 Perry Nuclear Power Plant, Unit 1.
Cleveland Electric Illumanating C NUREG-0087 l
50-441 Perry Nuclear Pcwer Plant. Unit 2.
Cleveland Electric illumahating L NUREG-0887
)
50-443 Se ab r oo k Nuclear Station. Unit 1.
Pubitt Service Co of New Hampshar NOHEG-0895 50-444 Seabrook Nuclear Station, Unst 2, Public Service Co.
of Nrw Hampshir NUNEG-0895 STN-SO-454 Byron Statson. Unit 1,
Commonwealth Ed a son Co NOREG-0848 STN-SO-4SS Byron Statson. Unit 2. Commonwealth Edison Co MMEG -0848 50-461 Clinton Power Stataan. Unit 1.
Illinois Power Co NUREG-0854 STN-SO-482 Wo l f Cre e n Generat in g St a t ion, Kansas Gas & Electrac NUREG-0078 STN-50-a82 Wolf Creen Generatang Station. Mansas Gas & Electric NUREG-0881 S0-522 Skagat Nuclear Power Project. Unnt 1.
Puget Sound Power & L aght NUREG-0894 50-523 Shagit Nuclear P owe r Project. Unnt 2, Puget Sound Power & Light NUREG-OB94 STN-SO-528 Palo Verde Nuclear Station. Untt 1.
Aragona Public Servite Co.
NUREG-OBS7 SO2 STN-SO-320 Palo Verde Nuclear Station, Unst 1.
Arigona Public Service to NUREG-0857 SO2 STN-SO-S29 Pa lo Verde Nuclear Station. Unit 2. Aragona Putlac Servate Co NUHEG-0857 SO2 STN-SO-530 Pa le Verde Nuc lear Station. Unit 3, Ararona Publac bervate to NUREC-0057 SO2 50-S37 Clinch River Breeder Reactor. Pr o j ec t Manag ement C ou p NUMEG-0786 ROI 50-537 Clinch River Breeder Reactor. Pr ojec t Management Cor p NUMEG/CR-2681 149
_~_m i
I I
l B
150
f,j,c rom 336
bf[ REG-03Yff "
u.s. NuCLsAa naruLATORY COMMISSION BIBLIOGRAPHIC DATA SHEET Vol. 7, No. 2
- 4. TITLE AND SUBTITLE (Aad volume Na, siappropr,esel
- 2. Reere 6/mAJ Regulatory and Technical Reports Compilation for Second Quarter 1982
- 3. RECsPIENT'S ACCESSION NO.
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLETED I"
A. Savolainen, Compiler and Indexer
- 9. PERFORMING ORGANIZATION N AME AND MAILING ADDRESS (/ncium lea Codel DATE REPORT ISSUED MONTH l YEAR Division of Technical Information and Document Control August 1982
.0ffice of Administration
- s. g,,,, y,,,,
U. S. Nuclear Regulatory Comission Washington, D. C.
20555
- 8. ne. e u=*s
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (include I,a Codel p
Same as 9, above.
- 11. FIN NO.
- 13. TYPE OF REPORT PE RIOD COVE RED (inclusere daars/
Reference April-June 1982
- 15. SUPPLEMENTARY NOTES
- 14. (Leave ofm*/
- 16. ABSTR ACT #00 words or lessJ This compilation lists all NRC regulatory and technical reports published under the series during the second quarter of 1982.
- 17. KE Y WORDS AND DOCUMENT AN ALYSIS 17a, DESCRIPTORS 17b IDENTIFIE RS!OPEN-ENDE D TE RYS 18 AVAILABILITY STATEMENT
- 19. SE CURITY CLASS ITh,s report) 21 NO. OF P AGES Unc1assified Unlimited 20 g iTvCg ga,,p,,i 22 Price NKC FORM 33S sit tu
UNITED STATES sover uaissait l
NUCLEAR REGULATORY COMMISSION "3'*[' ' 'c, i f s pa>o WASHINGTON, D C. 20566 was o Of FICIAL BU$1 NESS Main Citations
.="
and Abstracts B
Contractor Report Number Index
{
1 Personal Author index Subject index 120555078877 1
99999 US NRC ADM DIV 0F T10C POLICY E PUBLICATIONS MGT BR POR NUREG COPY NRC Originating h$3g}NvTON DC 20555 Organization index NRC Contractor Sponsor Index Contractor Index Licensed Facility Index
--