ML20083D093
| ML20083D093 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1983 |
| From: | NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V08-N03, NUREG-304, NUREG-304-V8-N3, NUDOCS 8312270202 | |
| Download: ML20083D093 (152) | |
Text
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NUREG-0304 Vol. 8, No. 3 I
Regulatory and Technical Reports i
Com allation for Thirc Quarter 1983 July - September U.S. Nuclear Regulatory Commission Office of Administration y+* "c <,,
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I 8312270202 831130 Q@4NREG N
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Available from N RC/GPO Sales Program Superintendent of Documents Government Printing Office Washington, D. C. 20402 A year's subscription consists of 4 issues for this pub!ication.
Single copies of this publication are available from National Technical Information Service, Springfield, V A 22161 Microfiche of single copies are available from NRC/GPO Sales Program Washington, D. C. 20555 I
l
NUREG-0304 Vol. 8, No. 3 Regulatory and Technical Reports Com ailation for Thirc Quarter 1983 July - September Date Published: November 1983 Division of Technical Information and Document Control Office of Administration U.S. Nuclear Regulatory Commission 1
Washington, D.C. 20555 pa =%9 l-i 3
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CONTENTS Preface.........................................................................'.....'v e-Index Tab
- Main Citation and Abstracts...............................................
- S taff R epo rts.......................................................................
Conference Proceedings.............................................................
Contractor Reports...........................
Contractor Report Number index.........................................................
Personal Author Index.............
.......................................2.
S u bject in dex......................................................................... 3 NRC Originating Organization Index (Staff Reports)......................................... 4 Ce ntractor Index......................eports)..............,...........
NRC Contract Sponsor index (Contractor R
.......... 6
- Licensed Facility index................................................................. 7
.............................................. 8 s
...ill
m 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 Technical Information and Document Control Policy and Publications Management Branch Publishing and Translations Section Woodmont 501 -
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 The main citations and abstracts in this compilation are listed in NtJREG number order: NUREG-XXXX, NUREG/CP-XXXX, and NUREG/CR-XXXX. These precede the following indexes:
Contractor Report Number Index o
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 NUHEG-0508: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITER ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048 09570:200.
Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for internal NRC use).
Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070.
Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled l
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 I
microfiche address (for NRC internal use).
Contrer.'or Report NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR L REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.: BENNETT, P.R.
i Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242.
Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC l
Document Control System accession number, (8) the report number of the originating organization (if j,
given), and (9) the microfiche address (for NRC internal use).
~
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l 1
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 Avai! ability 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 mcney order, payable to the Superintendent of Documents, to the following address:
U.S. Nuclear Regulatory Commission ATTN: Sales Manager Washingten, D.C. 20656 You may charge any purchase to your GPO Deposit Account, Master Charge card, or VISA charge card by calling the NRC-GPO Sales Office on (301) 492-9530. Non-U.S. customers must make payment in advance either by 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 numbe s that could not be correlated with NRC sponwrship of the work being reponed.
l l
In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference l
proceedings.
All these report codes are controlled and assigned by the NRC Division of Technical Information and Document Control.
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1 Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG XXXX is an NRC staff originated report, NUREG/CP-XXXX is an NRC sponsored conference report, and NUREG/CR-XXXX is an NRC contractor-prepared report. The bibliographic information (see Preface for details) is followed by a brief abstract of the report.
NUREG-0020 V07 NO3: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Greg Book)
- Management Information Branch.
July 1983.
382pp.
8308010034.
19865:001.
The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office of Management and Program Analysis from the Headquarters staff of NRC's Office of Inspection and Enforcement, f rom NRC 's Regional Offices, and from utilities.
The three sections of the report cre:
monthly highlights and statistics for commercial operating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC's Regional Offices, IF Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the U.S.
It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S.
energy situation as a whole.
NUREG-0020 VO7 NO4: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of March 31,1983. (Grey Book)
- Management Information Branch.
August 1983.
380pp.
8308250415.
20173:001.
The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the eperation of nuclear units as timely and accurately as possible.
This information is collected by the Of fice of Management and Program Analysis from the Headquarters staff of LRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are:
monthly highlights and statistics for commercial operating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC's Regional Of fices, IE Headquarters and the utilitiess and an appendix for miscellaneous information such is spent fuel storage capability, reactor-years of experience and non-pcwer reactors in the U.S.
It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S.
energy situation as a whole.
NUREG-0020 V07 NOS: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of April 30,1983.(Grey Book)
- Management Information 1
Branch.
August 1983.
375pp.
8309080026.
20321:001.
The OPERATING UNITS STATUS REPD3T - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office j
of Management and Program Analysis from the Headquarters staff of NRC 's Of fice of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are:
monthly highlights and statistics for commercial operating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC 's Regional Of fices, IE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the U. S.
It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S.
energy situation as a whole.
NUREG-0020 VO7 N06: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of May 31,1983.(Greg Book)
- Management Information Branch.
September 1983.
396pp.
8310050339.
20637:187.
The OPERATING UNITS STATUS REPORT - LICENSED OFERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible.
This information is collected by the Office of Management and Program Analysis from the Headquarters staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are:
monthly highlights and statistics for commercial operating units, and errata from previously reported datas a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of cxperience and non-power reactors in the U.S.
It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S.
energy situation as a whole.
NUREG-0040 VO7 NO2: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report, April 1983 - June 1983.(White Book) e Region 4, Office of Director.
July 1983.
275pp.
8307250591.
19716:057.
This periodical covers the results of inspections performed by the NRC's Vendor P.og am Dranch that have been distributed to the inspected organizations during the period April 1983 through June 1983.
Also included in this issue are the results of certain inspections performed prior to April 1983 that were not included in previous issues of NUREG-0040.
NUREG-0090 VO6 NO1: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. January-March 1983.
- Director's Office.
September 1983.
49pp.
8310130133.
20772:280.
Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as a6 anscheduled 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 January 1 to March 31, 1983.
During the report period, there were three abnormal occurrences at the nuclear power plants licensed by the NRC to operate.
The first involved a main feedwater line break due to water htmmer.
The second 2
involved management and procedural control deficiencies.
The third involved failure of the automatic reactor trip system.
There were no abnormal occurrences fer the other NRC licensees.
There were six abnormal occurrences at Agreement State licensees.
One involved an
'ndividual who ingested and was contaminated by radioactive material.
j Four involved lost or stolen radioactive sources.
One involved i
radioactive contamination of a metals production facility.
4 NUREG-0304 VO8 NO2: REQULATORY AND TECHNICAL REPORTS. Compilation For Second Guarter 1983.
- Division of Technical Information & Document Control.
August 1983.
146pp.
8308230729.
20145:096.
This compilation lists all NRC regulatory and technical reports published under the NUREG series during the second quarter of 1983.
l f
NUREO-0383 VO1 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR l
RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved j
Packages. *-
Division of Fuel Cycle & Material Safety.
September j
1983.
432pp.
8310130151.
20773:186.
This directory contains a Summary Report of NRC Approved Packages
{
(Volume 1), Certificate of Compliance (Volume 2), and a Summary Report
{
of NRC Approved Quality Assurance Programs for Radioactive Material i
Packages-(Volume 3).
The purpose of this directory is to make available a convenient of information on packaging which have been approved by the source U. S.
Nuclear Regulatory Commission.
To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance number is included at the back of each volume of the directory.
The Summary Report includes a listing of all users of each package design prior to the publication date of the directory.
Shipments of radioactive material utilizing these packagings must be in accordance with the provisions of 49 CFR 173.471 and 10 CFR Part 71, as applicable.
In satisfying the requirements of Section 71.12, it is the responsibility of the licensees to insure them that they have a copy of the currant approval and conduct their transportation activities f.n accordance with an NRC approved quality assurance program.
Co-les of the current approval may be obtained from the U.S.
Nuclear Regulatory Commission Public Document Room files (see Docket No. listed on each certificate) at 1717 H Street, Washington, DC 20555.
Note that the general license of 10 CFR 71.12 does not I
authorize the receipt, possession, use or transfer of byproduct l
source, or special nuclear materials such authorization must be obtained pursuant to 10 CFR Parts 30 to 36, 40, 50, or 70.
NUREG-0383 VO2 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR R ADIDACTIVE MATERIALS PACKAGES. Certificates Of Compliance.
l Division of Fuel Cycle & Material Safety.
September 1983.
618pp.
8310110532.
20710:001.
This directory contains a Summary Report of NRC Approved Packages (Volume 1 ), Certificates of Compliance (Volume 2), and a Summary Report of NRC Approved Guality Assurance Programs for Radioactive Material Packages (Volume 3).
The purpose of this directory is to make available a convenient of information on packagings which have been approved by the source U. S.
Nuclear Regulatory Commission.
To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance number is included at the back of each volume of the directory.
The Summary Report includes a listing of all users of each 3
package-design prior.to the publication date of the directory.
-Shipments of radioactive material utilizing these packagings must Hb e in accordance with the provisions of 49 CFR 173.471 and 10 CFR Part l
71, as applicable.
In satisfying the requirements of Section 71.12, i
it is the responsibility of the 12censees to insure them that they have a copy of the current approval and conduct their transportation activities in accordance with an NRC approved quality assurance program.
Copies of the current approval may be obtained from the U.S.
Nuclear Regulatory Commission Public Document Room Files (see Docket No. listed on each certificate) at 1717 H Street, Washington, DC 20555.
Note that the general license of 10 CFR 71.12 does not authorize the receipt, possession, use or transfer of byproduct source, or special nuclear materials such authorization must be obtained pursuant to 10 CFR Parts 30 to 36i 40, 50, or 70.
NUREG-0383 VO3 RO3: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR R ADIDACTIVE NATERI ALS PACKAGES. Summary Report Of NRC Approved Guali ty Assurance Programs For Radioactive Material Packages.
- Division oP Fuel Cycle & Material-Safety.
September 1983.
42pp.
8310110593.
20715:001.
This directory contains a Summary Report of NRC Approved Package (Volume 1), Certificates of Compliance-(Volume 2), and a Summary Report of NRC Approved Guality Assurance Programs for Radioactive Material Packages (Volume 3).
The purpose of this directory is to make available a convenient source of information on packaging which have been approved by the U. S.
Nuclear Regulatory Commission.
To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance number is included at the back of each volume of the directory.
The Summary Report includes a listing of all users of each package design prior to the publication date of the directory.
Shipments of radioactive material utilizing these packagings must be in accordance with the provisions of 49 CFR 173.471 and 10 CFR Part 71, as applicable.
In satisfying the requirements of Sections 71.12, it is the responsibility of the licensees to insure them that they have a copy of the current approval and conduct their transportation 4
activities in accordance with an NRC approved quality assurance program.
Copies of the current approval may be obtained from the U.S.
Nuclear Regulatory Commission Public Document Room files (see Docket No. listed on each certificate) at 1717 H Street, Washington, DC l
20555.
Note that the general license of 10 CFR 71.12 does not l
authorize the receipt, possession, use or transfer of byproduct l
source, or special nuclear materials such authorization must be obtained pursuant to 10 CFR Parts 30 to 36, 40, 50, or 70.
NUREG-0390 VO7 NO1: TOPICAL REPORT REVIEW STATUS.(Blue Book)
- Management Information Branch.
August 1983.
2OOp p.
8308230763.
20129:207 The primary purpose of this document is to provide periodic progress reports of on-going topical report reviews, to identify those topical reports for which the Nuclear Regulatory Commission (NRC) staff review has been completed, and to the extent practicable, to provide NRC management with sufficient information regarding the conduct of the topical report program to permit taking whatever actions deemed necessary or appropriate.
I 4
NUREG-0423 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF THC SHOREHAM NUCLEAR.*0WER STATION, UNIT NO.
- 1. DOCKET NO. 50-322.
(Long Island Lighting Company)
Division of Licensing.
September 1983.
71pp.
8310070116.
20671:122.
Supplement No. 4 to the Safety Evaluation Report of Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission.
This supplement addresses several items that have come to light since the previous supplement was issued.
6 NUREG-0430 V03 NO2: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. July - December 1982.
- Director's Office. Office of Inspection and Enforcement.
July 1983.
14pp.
8308230757.
20130:056.
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review ofthe information and completion of any related investigations.
Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium or uranium-233.
NUREG-0485 VO5 NO6: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
i REPORT. Data As Of June 30,1983.(Buff Book)
- Office of Resource
~
Management, Director.
Office of Nuclear Reactor Regulation.
Director.
July 1983.
74pp.
8307270366.
19811:266.
The Systematic Evaluation Program is intended to examine safety-related aspects of 11 of the older light water reactors.
This d ocument provides the existing status of the review process including individual ' topic and overall completion status.
i NUREG-0485 VO5 NO7: SYSTEMATIC EVALUATION PROGRAM STATU0
SUMMARY
REPORT. Data As Of July 31,1983.(Buff Book).
- Office of Resource Management, Director.
August 1983.
73pp.
8309060270.
20282:020.
The Systematic Evaluation Program is intended to examine many mafety-related aspects of 11 of the older light water reactors.
This document provides the existing status of the review process including I
individual topic and overall completion status.
4 i
NUREG-0485 VO5 NOB: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
f REPORT. Data As Of August 31,1983.(Buff Book)
- Office of Resource i
Management, Director.
September 1983.
74pp.
8310130149.
I 20773:109.
The Systematic Evaluation Program is intended to examine many safety-related aspects of 11 of the older light water reactors.
This document provides the existing status of the review process including 3
individual topic and overall coroletion status.
NUREG-0519 SO5: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LA SALLE COUNTY STATION, UNITS 1 AND 2. Docket Nos. 50-373 And
}
50-374.(Commonwealth Edison Company)
Division of Licensing.
August 1983.
106pp.
8308230747.
20129:100.
i Supplement No. 5 to the Safety Evaluation Report of Commonwealth Edison Company's application for a license to operate its La Salle 5
i County Station, Unit 2, located in Brookfield Township, La Salle County, Illinois, has been prepared by the Office of Nuclaar Reactor Regulation of the U.
S.
Nuclear Regulatory Commission.
This supplement is to update our evaluations on Unit 2 issues identified in the previous Safety Evaluation Report and Supplements that need resolution prior to issuance of the operating license for Unit 2.
I NUREG-0525 RO7: SAFEQUARDS
SUMMARY
EVENT LIST (SSEL), REVISION 7.
Division of Safeguards.
August 1983.
74pp.
8309200013.
20421:050.
l The Safeguards Summary Event List (SSEL) provides brief summaries of several hundred safeguards-related events involving nuclear 3
material or facilities regulated by the U.S.
Nuclear Regulatory Commission (NRC).. Events are described under the categories of bomb-related, intrusion, missing / allegedly stolen, transportation, vandalism, arson, firearms-related, radiological sabotage and miscellaneous.
The information contained in the event descriptions is derived primarily from official NRC reporting. channels.
i NUREG-0540 VO5 NO4: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. April 1-30, 1983.
- Division of Technical Information &
Document Control.
July 1983.
635pp.
8308000067.
19954:302.
This document is a monthly publication containing descriptions of l
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 j'
material received and generated by NRC pertinent to to its role as a regulatory agency.
The following indexes are included: Personal Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.
NUREG-0540 VO5 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
l May 1-31,1983.
- Division of Technical Information & Document Control.
August 1983.
644pp.
8309080029, 20323:258.
i This document is a monthig 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 mate-fal received and generated by the NRC pertinent to its role as a regulatory. agency.
The following indexes are included: Personal Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.
NUREG-0540 VOS N06: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
June 1-30,1983.
- Division of Technical Information & Document Control.
August 1983.
574pp.
8309200012.
20418:299.
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 l
regulatory agency.
The foAlowing indexes are included: Personal l
Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.
l 6
u NUREG-0540 VO5 NO7: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
July 1-31,1983.
- Division cf Technical Information & Document Control.
September 1983.
535pp.
8310200382.
20063:286.
This document is a monthly publication containing descriptions of information received and generated by the U.S.
NRC.
This information 1
includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency.
The following indexes are included:
Personal Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.
NUREG-0580 V12 N06: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of June 30,1983.(Blue Book)
- Management Information Branch.
July 1993.
65pp.
8307250332.
19737:221.
Provides a review of the status of the progress of the licensing reviews for all construction permits, operating licenses, special pruJects and non power reactor renewals under review, as reported to l
Congress.
i NUREG-0500 V12 NO7: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As F
Of July 31,1983 (Blue Book)
- Management Information Branch.
August 1983.
64pp.
8308230735.
20145:031.
Provides a review of the status of the progress of the licensing reviews for all construction permits, operating licenses, special projects and non power reactor renewals under review, as reported to Congress.
NUREG-0500 V12 NOO: REQULATORY LICENSING STATUS
SUMMARY
REPORT. Data As of August 31,1983 (Blue Book)
- Management Information Branch.
September 1983.
62pp.
8309280555.
20541:245.
Provides a review of the status of the progress of the licensing reviews for all construction permits, operating licenses, special j
projects and non power reactor renewals under review, as reported to Congress.
NUREG-0606 VO5 NOj: UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As Of August 19,1983. (Aqua Book).
- Management Information Branch.
September 1983.
50pp.
8309280589.
20542:089.
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-0675 S16: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYDN NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 i
And 50-323.(Pacific Gas And Electric Company)
- Division of I
Licensing.
August 1983.
49pp.
8309060116.
20282:133.
Cupplement No.
16 to the Safety Evaluation P port for Pacific Cos and Electric Company's application for licenses to operate the Diablo
+
Canyon Nuclear Power Plant (Docket Nos. 50-275 and 50-323) located in y
San Luis Obispo County, California has been prepared by the Office of Nuclear Reactor Regulations of the U.S.
Nuclear Regulatory Commission.
4 This supplement provides recent information concerning the Diablo Canyon component cooling wnter system.
i 7
NUREG-0675 S18: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYDN NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)
Division of Licensing.
August 1983.
200pp.
8308230751.
20130:071.
Supplement No. 18 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate the Diablu Canyon Nuclear Power Plant (Docket Nos. 50-275 and 50-323), located in San Luis Obispo County, California, has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
This supplement presents the staff's evaluation on matters related to a verification effort on Diablo Canyon Unit 1 that was the result of Commission Order CLI-81-30 and a letter to Pacific Electric Company, dated November 19, 1981.
NUREe- 0725 RO3: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTDR FUEL.
- Office of Nuclear Material Safety &
Safeguards, Director.
July 1983.
49pp.
8308010155.
19856:271.
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'(HEC), and to meet the requirements of Public Lac 96-295.
The NRC staff must approve such. routes prior to their first use in accordance with the regulatory provisions of Sectien 73.37 of 10 CFR Part 73.
The information included reflects NRC staff knowledge as of May 1, 1983.
Spent fuel shipment routes, primarily for road
^
transportation, but also including one rail route, are indicated on reproductions of DDT road maps.
Also included are the amounts of material shipped during the approximate three year period that safeguards regulations for spent fuel shipments have been wffective.
In addition, the Commission has chosen to provide information in this document regarding the NRC's safety and safeguards regulations for spent fuel shipments as well as safeguards incidents regarding spent fuel shipments (of which none have been reported to date).
This additional information is furnished by the Commission in order to convey to the public a more complete picture of NRC regulatory practices concerning the shipment of spent fuel than could be obtained by the publication of the shipment routes and quantities alone.
NUREG-0743 V03 N06: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of June 30,1983.(Orange Book)
- Management Information Branch.
July 1983.
333pp.
8303100400.
19997:226.
The Operating Reactors Licensing Actions Summary is designed to provide tha management of the Nuclear Regulatory Commission with an overview of licensing actions dealing with operating power and nonpower reactore.
NUREG-0748 V03 NO7: OPERATING REACTORS LICENSING ACTIONS SUMMARV. Data As Of July 31,1983.(Orange Book)
- Management Information Branch.
August 1983.
330pp.
8309200023.
20424:042.
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.
l NUREG-0821 SO1: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVAL.JATION PROGRAM - R.
E.
GINNA NUCLEAR POWER PLANT. Docket No.
i 8
l l
u p
50-244.(Rochester Cas And Electric Corporation)
- Division of Licensing.
August 1983.
46pp.
8309200476.
20452:050.
The Nuclear Regulatory Commission (NRC) has published its Supplement No. 1 to the Final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0821), under the scope of the Systematic Evaluation Program (SEP), for Rochester Gas & Electric Corporation's R.
E.
Ginna Plant located in Wayne County, near Rochester, New York.
The SEP wcs initiated by the NRC to review the design of older operating nuclear power plants to reconfiria and document their safety.
This report documents the review completed under the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations subsequent to issuing the Final IPSAR for the Ginna Plant.
The review has provided for (1) an assessment of the significance of dif ferences between current technical positions on selected safety issues and those that existed when the Cinna Plant was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) documented evaluation of plant safety when the supplement to the a
4 Final IPSAR and the Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued.
The Final IPSAR and its supplement will form part of the bases for considering the conversion of the existing provisional operating license to a full-term operating license.
NUREG-0828 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT REPORT SYSTEMATIC EVALUATION PROGRAM - BIG ROCK POINT PLANT. Docket No. 50-155.
(Consumers Power Company)
- Division of Licensing.
Septsmber 1983.
7bOpp.
8310200413.
20866:242.
The Systematic Evaluation Program was initiated in February 1977 by the U.S.
Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety.
The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a Jocumented i
evaluation of plant safety when all supplements to the Final Integrated Plant Safety Assessment Report and the Safety Evaluation Report for converting the license from a provision to a full-term license have been issued.
This report documents the reviet-of the Big Rock Point Plant, operated by Consumers Power Company located in Charlevoix, Michigan.
j Big Rock Point is one of ten plants reviewed under Phase II of this program.
This report indicates how 137 topics selected for review under Phase I of the program were addressed.
It also addresses a majority of the pending licensing actions for Big Rock Point, which include TMI Action Plant requirements and implementing criteria for resolved generic issues.
Equipment and procedural changes have been identified as a result of the review.
NUREG-0837 V02 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, October - December 1982, COSTELLO,F.s THOMPSON,T.s COHEN,L.
Region 1, Office of Director.
July 1983.
243pp.
8308010012.
19860:011.
This report provides the status and results of the TGC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network.
It presents the radiation levels measured in the vicinity of NRC licensed facility sites throughout the counbry for the fourth quarter of 1902.
9
-~
l l'
NUREG-0852 SO2: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN OF THE STANDARD NUCLEAR STEAM SUPPLY REFERENCE SYSTEM CESSAR SYSTEM
- 80. Docket No. 50-470.(Combustien Engineering, Incorporated)
- Office of Nuclear Reactor Regulation, Director.
September 1983.
163pp.
8310050232.
20635:151.
Supplement No. 2 to the Safety Evaluation Report for the
)
application filed by Combustion Engineering, Inc. for a Final Design Approval for the Combustion Engineering Engineering Standard Safety Analysis Report (STN 50-470) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission.
The purpose of this supplement is to update the Safety Evaluation by providing (1) the evaluation of additional information submitted by the applicant since Supplement No. 1 to the Safety Evaluation Report was issued and (2) the evaluation of the matters the staff had under review when Supplement No. 1 to the Safety Evaluation Report was issued.
NUREG-0871 VO2 NO3:
SUMMARY
INFORMATION REPDRT. Data As Of June 30,1983.(Brown Book)
- Management Information Branch.
August 1983.
55pp.
8308170187.
20085:075.
Provides summary data concerning NRC and its licensees for general use by the Chairman, other Commissioners and Commission staff i
offices, the Executive Director for Operations, and the Office Directors.
I NUREG-0881 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WOLF CREEK GENERATING STATION,UN7T 1. Docket No. 50-482.(Kansas Gas And Electric Company et al.)
- Division of Licensing.
August 1983.
31pp.
8309060272.
20282:097.
Supplement No. 3 to the Safety Evaluation Report related to operation of the Wolf Creek Generating Station. Unit No. 1 updates the information contained in the Safety Evaluation Report, dated April 1982 and Supplement Nos. 1 and 2, dated August 1982 and June 1983, respectively.
Supplement No. 3 contains resolutions of open issues and confirmatory items and addresses Board Notifications.
The Safety Evaluation and its supplements pertain to the i
application for a license to operate the Wolf Creek Generating Station, Unit No. 1 filed by Kansas Gas and Electric Company on February 19, 1980.
The Construction Permit CPPR-147 was issued on May 17, 1977.
The facility is located in Coffey County, Kansas.
NUR EG-0925: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF THE TETON URANIUM ISL PROJECT. Doc ket No. 40-8781. (Teton Exploration Drilling,Inc.)
- Office of Nuclear Material Safety &
Safeguards, Director.
August 1983.
25pp.
8308230517.
20127:182.
This Final 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 ProJoct to mine uranium in situ by injecting a carbonate / bicarbonate lixiviant into the ore body.
The statement considers:
(1) alternative of no licensing action, (2) alternative energy sources, and (3) alternatives if cranium ore is mined and refined on the site.
The proposed action is to grant a Source and Byproduct Material License to the applicant subject to the etipulated license condition.
10
NURED-0936 VO2 NO2: NRC REQULATORY AGENDA.Guarterly Report, March.- May 1983.
- Division of Rules and Records.
July 1983.
180pp.
8309030470. - 19910:001.
i The NRC Regulatory Agenda is a compilation of all rules on which the NRC has proposed or is considerins action and all petitfor.s fue 4
rulemaking which have been received by the Cuir. mission and are pending disposition by the Commission.
The Regulatory Agenda is updated and
-issued each guarter._ The Agendas for April and October are published in their entirety in the Federal Register while a notice of availability is published in the Federal Register for the January and July Agendas.
NUREG-0940 VO2 NO2: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED.Guarterly Progress Report, April-June 1983.
- Enforcement Staff.
July 1983.
341pp.
8308150492.
20068:023.
P This compilation summarizes significant enforcement actions that have been resolved during one guarterly period (April - June 1983) and
- =
includes copies of letters, notices, and orders sent by the Nuclear Regulatory Commission to the licensee with respect to the enforcement action and the licensee's response.
It is anticipated that the information in this publication will be widely disseminated to managers and empla-tes engaged in activities licensed by the NRC,-in the interest of promoting public health and safety as well as commen defense and security.
NUREG-0979 SO1: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN APPROVAL OF THE GESSAR II BWR/6 NUCLEAR ISLAND DESIGN. Docket No.
50-447.(General Electric Company)
Division of Licensing.
July 1983.
47pp.
8309200489.
20452:001.
l Supplement 1 to Safety Evaluation Report (SER) for the J
application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design (GESSAR II) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission.
This report supplements the GESSAR II SER (NUREG-0979) issued.in April 1983 summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design.
The staff concludes that, subject to approval of the balance-of plant 1'
design, applicants referencing GESSAR II can conform with the
. provisions of the Act and the regulations of the Nuclear Regulatory
- C ommi s s i on.
l NUR EG-0984: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE CORNELL UNIVERSITY TRIGA RESEARCH REACTOR. Docket No.50-157. BERNARD,H.
Division of Licensing.
August j
1983.
84pp.
8308170190.
20085:129.
This Safety Evaluation Report for the application filed by Cornell University for renewal of operating license number R-80 to continue to operate the TRIGA research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is owned and operated by Cornell University and is located on the university campus in New York.
The staff concludes that the TRIGA reactor facility can continue to be i
operated by Cornell University without endangering the health and safety of the public.
l 11 "e+-
.-,,s,-.~~a.e.
e..-_. -,
. - m
- - ~
l NUREG-0985 VO1:
U. S.
NUCLEAR REGULATORY COMMISSIDN HUMAN FACTORS PROGRAM PLAN.
- Division of Human Factors Safety.
August 1983.
75pp.
8309060100.
20282:186.
The Human Factors Program Plan has been developed to ensure consideration of human factors.in the design, operation and maintenance of nuclear facilities.
This initial plan addresses nuclear power plants and describes staff develcpment activities planned to (1) provide technical bases for resolution of the remaining human factors related tasks described in NUREG-0660 and NUREG-0737, and-(2) investigate and resolve additional human factors problems that have been identified since publication of NUREG-0660 and NUREG-0737.
The Plan has seven major program elements:
(1) Staffing and Gualifications,
. (2) Training, (3) Licensing Examinations, (4)
Procedures, (5) Man-Machine Interface, (6) Management and
+
Organization, and (7) Human Reliability.
Activities within these program elements are directed to providing technical bases for developing guidance to the nuclear industry, improving staff capabilities to perform licensing activities, and supporting staff recommendations regarding the degree of regulation required to resolve i
technical issces.
Appendixes to the Plan describe the status of TMI Action Plan human factors issues, staff responses to recommendations provided by the Human Factors Society, and schedules for the seven program elements.
This edition (Volume 1) of the Plan describes activities in progress and planned for Fiscal Years 1983 - 1985.
NUR EG-0988: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE OMAHA VETERANS ADMINISTRATION MEDICAL CENTER. Docket No. 50-131.
- Office of Nuclear Reactor Regulation, Director. ' July 1983.
70pp.
8308080731.
19971:158.
This Safety Evaluation Report for the application filed by the Omaha Veterans Administration Medical Center (DVAMC) for a renewal of operating license number R-57 to continue to operate a researth i
reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is owned and operated by the Veterans Administration and is located in its Medical Center in the city of Omaha, Nebraska.
The staff concludes that the TRIGA reactor facility can continue to be operated by OVAMC without endangering the health and safety of the public.
NUREG-0991: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LIMERICK GERNERATING STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-353.(Philadelphia Electric Company)
- Division of Licensing.
i August 1983.
665pp.
8309200017.
20422:001.
l The Safety Evaluation Report for the application filed by the Philadelphia Electric Company, as applicant and owner, for licenses to operate-the Limerick Generating Station, Units 1 and 2 (Docke t Nos.
l 50-332 and 50-353), has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
Nuclear Regulatory Commission.
The facility is located near Pottstown, Pennsylvania.
Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that r
the facility can be operated by the applicant without endangering the health and safety of the public.
NUR EG-0996: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE UNIVERSITY OF OKLAHOMA RESEARCH REACTOR. Docket No. 50-112.
- Division of Licensing.
September 1983.
12
i 57pp.
B310100071.
21826:174.
This Safety Evaluation' Report for the application filed by the University of Oklahoma for a renewal of operating license number R-53' to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.
S. Nuclear Regulatory Commission.
The facility is owned and operated by the University of Oklahoma and is located on the campus in Norman, Cleveland County, Oklahoma.
The staff concludes that the AON-211 reactor facility can continue to be operated by University of Oklahoma without endangering the health and safety of the public.
NUREG-0998:
U. S.
NUCLEAR REQULATORY COMMISSION 1982 ANNUAL REPORT.
DICKBON,J.H.
Office of Resource Management, Director.
July 1983.
224pp.
8307220626.
19697:293.
This report addresses all NRC activities, policies, and decisions made during the reporting period, complete with illustrations, cle r ts, and treatment of technical material in lag language for consumption by the lay public.
NUREG-1000 VO2: GENERIC IMPLICATIONS OF ATWS EVENTS AT THE SALEM NUCLEAR POWER PLANT. Licensee And Staff Actions.
- Office of Nuclear Reactor Regulation, Director.
August 1983.
37pp.
8308250377.
3016a: 082.
This report, Volume 2 of two volumes of NUREG-1000, describes the intermediate term actions to be taken by licensees and applicants of the U.S. Nuclear Regulatory Commission (NRC), on the one hand, and by NRC staff, on the other, to address the generic issues raised by two anticipated transients without scram (ATWS) at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983.
These actions came about as a result of the findings of NUREG-1000, Volume 1, and of reviews by the NRC Committee to Review Generic Requirements, the NRC Program Offices, and the Commission.
The actions to be taken by licensees and applicants have been detailed in a letter pursuant to 10 CFR 50. 54 ( f ).
l NUREG-1003: DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR REPAIR AT H.B.
ROBINSON STEAM ELECTRIC PLANT UNIT NO.
2.
Docket No.
50-261.(Carolina Power And Light Company)
Division of Licensing.
August 1983.
43pp.
8309200005.
20423:311.
The staff has considered the environmental impacts and economic costs of the proposed steam generator repair at the H.
B.
Robinson Steam Electric Plant Unit No. 2 along with reasonable alternatives to the proposed action.
The staff has concluded that the proposed repair will not significantly affect the quality of the human environment and that there are no preferable alternatives to the proposed action.
Furthermore, any impacts from the repair program are outweighed by its benefits.
I NUREG-1007: SAFETY EVALUATION RFr0RT RELATED TO THE LICENSE RENEWAL AND POWER INCREASE FOR THE NATIO'
. BUREAU OF STANDARDS REACTOR. Docket No. 50-184. BERNARD,H.
Divi
.on of Licensing.
September 1983.
116pp.
8310200384.
20863:118.
This Safety Evaluation Report for the application filed by the National Bureau of Standards l
(NBS) for renewal of the operating license number TR-5 for a period of 20 years and for an increase in p ower from 10 MWt to 20 MWt has been prepared by the Office of Nuclear 13 t
-~-,m-,
Reactor Regulation c of the U. S.
Nuclear Regulatory Commission.
The facility is owned and operated by the Bureau of Standards which is
_ located on the NBS site in Gaithersburg, Md.
The staff concludes that the NBS reacter facility can be operated without endangering the health ~and safety of the public.
NUREG-1010: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE ZERO-POWER REACTOR AT CORNELL UNIVERSITY. Docket No. 50-97.
(Cornell University)
- Division of Licensing.
September 1983.
72pp.
8309290436.
205S4:137.
This Safety Evaluation Report for the application filed by Cornell University for renewal of operating license number R-97 to continue'to operate the Zero Power Research (ZPR) reactor has been i
prepared by the Office of Nuclear Reactor Regulation of the U.S i
Nuclear Regulatory Commission.
The facility is owned and operated by
. Cornell University and is located on the university campus in Ithaca, New York.
The staff concludes that the ZPR reactor facility can continue to be operated by Cornell University without endangering the l
health and safety oV the public.
NUREG-1011: DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERAlDR REPAIR AT THE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
50-266.(Wisconsin Electric Power Company) COLBURN,T.G.
Division of Licensing.
July 1983.
43pp.
8307250323.
19725:284.
The staff has considered the environmental impacts and economic i
costs of the proposed steam generator repair at the Point Beach Nuclear Plant Unit No. I along with reasonable alternatives to the proposed action.
The staff has concluded that the proposed repair will not significantly affect the quality of the human environment and that there are no preferable alternatives to the proposed action.
1 Furthermore, any impacts from the repair program are outweighed by its benefits.
i NUR EG-1011: FINAL ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR REPAIR AT THE POINT BEACH NUCLEAR PLANT, UNIT 1. Doc ket No.
50-266.(Wisconsin Electric Power Company) COLBURN,T.G.
Division of Licensing.
September 1983.
60pp.
8310070439.
20672:299.
The staff considered the environmental impacts and economic costs of the proposed steam generator repair at the Point Beach Noclear Plant Unit No. 1 along with reasonable alternatives to the proposed l
action.
The staff has concluded that the proposed repair will not significant1g affect the quality of the human environment and that there are not preferable alternatives to the propcsed action.
Furthermore, any impacts from the repair program are outweighed by its benefits.
NUREG-1016: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE IOWA STATE UNIVERSITY. Docket No. 50-116.
- Division of Licensing.
September 1983.
83pp.
8310180169.
20825:089.
'This Safety Evaluation Report for the application filed by the Iowa State University (ISU) for a renewal of the Class 104 Operating
~
License R-59 to continue to operate its Argonaut-type research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.
S. Nuclear Regulatory Commission.
The facility is owned and operated by the Iowa State University, and is located on the ISU 14
l 1
campus in Ames, Story County, Iowa.
The staff concludes that the reactor facility can continue to be operated by ISU without i-endangering the health and safety of the public.
NUREG-1018: SEISMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS.
Status Report, Unresolved Safety Issue A-46. CHANG,T.Y.
Division of Safety Technology.
September 1983.
87pp.
8310110345.
20714:074.
This report summarizes the status of work accomplished on USI A-46 by the U.
S.
Nuclear Regulatory Commission staff and its contractors, Idaho National Engineering Laboratory (INEL), Southwest I
Research_ Institute (SWRI), Brookhaven National Laboratory (BNL) and Lawrence Livermore National Laboratory (LLNL) that is applicablu to USI A-46.
This assessment lead to the conclusion that the use of i
seismic experience data for equipment qualification provides the onig i
i reasonable alternative to current qualification criteria.
Consideration of seismic qualification by use of experience data was a specific task in USI A-46.
Several other A-46 tasks serve to support the use of an experience data base.
The status of continuing efforts to establish requirements for an experience data base is provided in this report.
NUREG-1020 VO1: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al V.The Babcock &
Wilcox Compang,et al.Three Mile Island Nuclear Station, Unit 1.
Docket No. 50-289.
- Office of Nuclear Reactor Regulation, Director.
Septemter 1983.
85pp.
8310070437.
20674:126.
This report documents a review by the Nuclear Regulatory Commission (NRC) staff of the General Public Utilities v.
Babcock &
Wilcox lawsuit record to assess whether any of the staff's previous conclusf or.s or their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in light of the information contained in the lawsuit record.
Details of the lawsuit record are provlded in the appendices contained in Volume 2 of this report.
NUREG-1020 VO2: GPU V.
B&W LAWSUIT REVIEW AND ITS ER'ECT ON TMI-1. General Public Utilities Corporation,et al V.
The Babcock &
Wilcox Compangeet al.Three Mile Island Nuclear Station, Unit 1.
Docket No.50-289.
- Office of Nuclear Reactor Regulation, Director.
September 1983.
832rp.
8310070441.
20677:001.
This report documents a review of the Nuclear Regulatory Commission (NRC) staff of the General Public Util, ties v.
Baucock &
Wilcox lawsuit record to assess whether any of the staff's previous conclusions or their principal bases presented et the Three Mile Island Unit-1 (TMI-1) restart hearing, support *.ng restart of TMI-1, should be amended in light of the inforr.:'2on contained in the lawsuit record.
Details of the lawsuit record are provided in the appendicos
-contained in Volume 2 of this report.
NUR EG-1022: LICENSEE EVENT REPORT SYSTEM. Description Of System And Guidelines For Reporting. HEBDON,F,J.
Director's Office.
September 1983.
BGpp.
8310130130.
20772:189.
On July 26, 1983, the Commission published in the Federal Register a final rule that modifies and codifies the Licensee Event Report (LER) system.
The rule becomes effective on January 1,
1984.
This NUREG provides supporting information and guidance that will be 15
I of interest to persons responsible for the preparation and review of i
LERs.
The information contained in this NUREG includes: (1) a brief I
description of how LERs are analyzed by the NRC, (2) A restatement of the guidance contained in the Statement of Consideration that accompani'ed the publication of the LER rule, (3) a set of examples of potentia 11g ' reportable events with staff comments on the actual reportability of each event, (4) guidance on how to prepare an LER, including the LER form, and (5) guidance on the submittal of LERs.
NURE0/CP-OO47: TRANSACTIONS OF THE ELEVENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.To Be Held At National Bureau of 1
Standards,Gaithersburg,Naryland,0ctober 24-28,1983.
- Office of j
Nuclear Regulatory Research,. Director.
September 1983.
37pp.
8310110171.
20702:001.
This report _contains summaries of papers on reactor safety research work to be presented at the lith Water Reactor Safety Research Information Meeting.
The meeting will be held at the National Bureau of Standards in Gaithersburg, Maryland, October 24-28, i
1983.
The summary reports highlight the progrars and results of nuclear safety research work sponsored by the Office of Nuclear l
Regulatory Research, USNRC.
Summaries of invited papers are also included.
The latter represent work on reactor safety research conducted by the electric utilities through the Electric Power R e s ear c.h Institute, the nuclear industry, and various government and industry organizations in Europe and Japan.
The summaries have been compiled in one report to facilitate discourse and the open exchange of information during the course of the meeting.
NUREQ/CR-0130 ADDO2: TECHNOLOGY, SAFETY AND COSTS DF DECOMMISSIONING A l
REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Effects On j
Decommissioning Of Interim Inability To Dispose Of Wastes Offsite.
HOLTER,0.M.s MURPHY,E.S.
Battelle Memorial Institute, Pacific l
Northwest Laboratory.
July 1983.
45pp.
8308170163.
20088: 288.
j The impacts on the previous 1g reported technology, safety and costs of decommissioning a reference pressurized water reactor power i
station that would result from an inability to dispose of decommissioning wastes offsite at the time of decommissioning are examined.in this addendum.
Three onsite waste storage alternatives j
are evaluated:
- 1) interim onsite storage of low-level waste (LLW), 2) l interim onsite storage of spent fuel with offsite disposal of LLW, and
- 3) interim onsite storage of both LLW and spent fuel.
Storage periods of up to 100 years are considered, after which the wastes are postulated to be disposed of offsite and decommissioning of the i
facility and site to unrestricted release conditions is completed.
Estimates of costs and occupational radiation dose are made for each of the alternatives studied.
NUREG/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND CDSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR POWER STATION. Ef f ects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
HOLTER,G.M.s MURPHY,E.S.
Battelle Memorial Institute, Pacific Northwest Laboratorg.
July 1983.
45pp.
8308180441.
20101:005.
The impacts on the previously reported technology, safety and costs of decommissioning a reference boiling water reactor power station that would result from an inability to dispose of decommissioning wastes offsite at the time of decommissioning are examined in this addendum.
Three onsite waste storage alternatives 16
cro oviluated:
- 1) interia ensite storage of Inw-level waste (LLW), 2)
-interim onsite storage of spent fuel with offsite disposal of LLW, and
- 3) interim onsite storage of both LLW and spent fuel.
Storage periods of up to 100 years are considered, after which the wastes are postulated to be disposed of offsite and decommissioning oh the facility and site to unrestricted release conditions is completed.
Estimates of costs and occupational radiation dose are made for each of the alternatives studied.
NUREQ/CR-1246 VO4: SAFE USERS MANUAL. Volume 4: Computer Programs.
GRADY,L.M.
Sandia Laboratories.
September 1983.
179pp.
8309200528.
SAND 79-2247/4.
20454:205.
Documentation for the Safeguards Automated Facility Evaluation (SAFE) computer programs is presented.
The documentation is in the form of subprogram trees, program abstracts, flowcharts, and listings.
Listings are provided on microfiche.
NUREG/CR-1756 ADDO1: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameters.
KONZEK,G.J.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
2OOp p.
8308080663.
19977:006.
Additional analyses of decommissioning at the reference research and test (R&T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial report (NUREG/CR-1756).
The parameters examined for decommissioning are:
- 1) the effect on costs and radiation exposure of plant size and/or t y p e; 2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and 3) the cos's of and the available alternatives for the disposal of nuclear R&T reactor fuel assemblies.
NUREQ/CR-2OOO VO2 N6: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of June 1983.
- Dak Ridge National Laboratory.
July 1983.
9Bpp.
-8308010026.
ORNL/NSIC-2OO.
29860:299.
This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file of i
the Nuclear Safety Information Center (NSIC) during the one month l
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 j-federal regulations.
Procedures for LER reporting are described in i
detail in NRC Regulatory Guide 1.16 and NUREG-0161, Instructions for Preparation of Data Entry Sheets for License 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, keywords, and component vendor indexes follow the summaries.
l The components, systems, and vendors are those identified by the d
utility when the LER form is initiateds the keywords afe assigned by the computer using correlation tables from the Sequence Coding and Search System.
NUREG/CR-2OOO VO2 N7: LICENSEE EVENT REPORT (LER) COMPILATION:For Month of July 1983.
- Dak Ridge National Laboratory.
August 1933.
252pp.
l 8309200025.
ORNL/NSIC-2OO.
20420:157.
j This monthly report contains Licensee Event Report (LER) 4 l
17
^
operational information that was processed into the LER data Pile of the Nuclear Safety Information Center (NSIC) during the one month period identified on the cover of this document.
The LERs, from which i
l 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 ReDulatory 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 data for each facility.
Component, system, keywords, and component vendor indexes follow the summaries.
The-components, systems and vendors are those identified by the utility when the LER form is initiateds the keywords are assigned by the computer using correlation tables from the Sequence Coding and i
Search System.
a NUREG/CR-2OOO VO2 N8: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of August 1983.
- Oak Ridge National Laboratory.
September 1983.
154po.
8310110023.
ORNL/NSIC-2OO.
20703:128.
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 i
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, keywords, and component vendor indexes follow the summaries.
The components,-system and vendors are those identified by the utility when the LER form is initiateds the keywords are assigned by the computer using correlation' tables from'the Sequence Coding and Search S y s t em.-
I NUREG/CR-2146 VO2: DYNAMIC ANALYSIS TO ESTABLISH NORMAL SHOCK AND VIBRATION OF RADIOACTIVE MATERIAL SHIPPING PACKAGES: Guarterly l
Progress Report. April 1,1981 - June 30,1981. FIELDS,S.R.
Hanford I
Engineering Development Laboratory.
July 1983.
57pp.
8308010030.
l HEDL-TME-03-8.
19859:320.
The CARDS (Cask Rail Car Dynamic Simulator) model was modified to simulate the cask-rail car systems used in Tests 13, 16 and 18 of the I
series of rail car coupling tests conducted at the Savannah River Laboratories (SRL) in July and August of 1978.
An assessment of how well CARDS simulates the behavior of these cask-rail car systems was made by comparing calculated and experimental values of four response variables.
This completes the development and validation of the CARDS I
model.
NUREG/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS
{
OF OXIDE FUEL RODS. SIEFKEN,L.J.s DERNA.G.A.s SHAH,V.N.i et al.
l EG&G, Inc.
July 1983.
73pp.
8308010007.
EGG-2104.
19866:023.
l FRAP-T6 is a computer code which is being developed to calculate I
the transient behavior of a light water reactor fuel rod.
This report is an addendum to the FRAP-T6/ MODO user's manual which provides the additional user information needed to use FRAP-T6/ MODI.
This includes 18 r-
-r
i csdol chcngoo, icprovcsonts, and additions, coding changes and improvements, changes in input and control language, and example i
r problem solutions to aid the user.
This information is designed to supplement the FRAP-T6/ MODO user's manual.
t f
NUREG/CR-2238 VO4: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEMBER 1981. Volume 20.
- Sandia Laboratories.
September i
1983.
300pp.
8309290140.
SAND 81-1529.
20584:265.
Sandia National Laboratories is conducting, under USNRC's sponsorship, phenomenological research related to the safety of commercial nuclear power reactors.
The overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues.
(2) understanding risk-significant accident sequences.
(3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in i
commercial service in the United States without appropriate consideration being given to their effects on health and safety.
i Together with other progra:ns, the Sandia effort as directed at l
assuring the soundness of the technology base upon which licensing decisions are made.
This report describes progress in a number of activities dealing with. current safety issues relevant to both light water and breeder reactors.
The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes L
used in accident analysis and licensing reviews.
Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions.
Current mmJor emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.
NUREO/CR-2254: A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS. BELL.B.J.s SWAIN,A.D.
Sandia Laboratories.
July 1983.
147pp.
8308100455.
SAND 81-1655.
19997:028.
This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the i
methods described in NUREQ/CR-127, " Handbook for Human Reliability l
Analysis with Emphasis on Nuclear Power Plant Applications" As overview of the procedure describing the major elements of a human r
reliability analysis is presented along with a detailed description of each element and an example of an acteal analysis.
An appendix consists of some sample human reliability analysis problems for further study.
t NUREQ/CR-2331 VO2 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report,0ctober l
-December 31,1982. WEISS,A.J.
Brookhaven National Laboratory.
July 1983.
173pp.
8308010085.
DNL-NUREG-51454.
19859:146.
l The Advanced and Water Reactor Safety Research Programs Guarterly Progress Reports have been combined and are included in this report entitled, " Safety Research Programs Sponsored by Office of Nuclear 19
l Regulatory Research Guarterly Progress Report."
This progress report will describe current activities and technical progress in the following projects:
HTGR Safety Evaluation, SSC Development, Validation and Application, CRDR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, LWR Plant Analyzer Development LWR Code Assessment and Applications Stress Corrosion Cracking of PWR Steam Generator Tubing, Dolting Failure Analysis, Probability Based Load Combinations for Design of Category 1 Structures, Mechanical Piping Benchmark Problems, Soil Structure Interactions Human Error Data for Nuclear Power Plant Safety Related Events, Criteria for Human Engineering Regulatory Guides and Human Factors in Nuclear Power Plant Safeguards.
The previous reports have covered the period October 1,
1976 through September 30, 1982.
NUREG/CR-2402: RISK ANALYSIS NETHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report. PEPPING,R.E.; CHU,M.S.Y.s WAHI,K.K.; et al.
Sandia Laboratories.
July 1983.
10Spp.
8308100487.
SAND 81-2409.
20007:130 A previously develop ed risk assessment methodology was modified and applied in the analysis of risks from the geologic disposal of spent unreprocessed fuel (SURF).
The methodology had initially been developed for application to the risk assessment from the geologic disposal of reprocessed, high-level waste (HLW).
A set of disruptive scenarios, chosen from a larger set of scenarios in the previous study, was analyzed in this study to demonstrate the methodology as applied to a conceptual SURF repository in bedded salt.
The results are expressed in terms of integrated discharges and conditional cancer risk.
A comparison with the proposed EPA Standard is also presented for demonstration purposes only.
The results of individual SURF scenario analynes are compared with the corresponding HLW results.
Based on the present analysis, it appears that over a period of 100,000 years the r isks associated with the disposal of SURF are higher than those for HLW for the assumed repository configuration at a hypothetical bedded salt site.
The study also demonstrates how the degree and order of importance of various parameters and radionuclides can change when the basis for estimating the risk is changed.
Whereas, the methodology could be extended to other media, one is cautioned against generalizing the conclusions of this study to repositories in media other than bedded salt.
NUREG/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS. BERLAD,A.L.s SIDULKIN,M.; YANG,C.H.
Brookhaven National Laboratory.
July 1983.
43pp.
8307250530.
BNL-NUREG-51492.
19739:228.
A knowledge of comburtion phenomena and their characteristics is necessary in accident analyses related to the release of hydrogen.
As l
a result of the accident at Three Mile Island, and from the results of related studies of hypothetical degraded core accidents, it is recognized that combustion of hydrogen may, under some circumstances, threaten the integrity of a reactor containment building.
In general, detailed combustion information is required in order to understand and characterire the combustion phenomena and processes which may occur in a containment building.
It is also important to identify the l
criticality or limiting conditions under which important combustion processes may be extinguished, initiated, or otherwise transformed.
Data and information is also required for analytic modeling of safety-related hypothetical accident scenarios and to allow modelers to predict with confidence the consequences of combustion processes.
20
Scfoty-roloted strotogion ciccd et mitigation of accident-related combustion also require a knowledge of combustion phenomena.
The report identifies areas where inadeguate understanding exists and distinguishes among the ranges of applicability of selected items and classes of combustion data, experiments, theories, and models.
In pursuit of these objectives, this report attempts to provide a perspective on combustions processes which may not otherwise be derived easily from the enormous 1g diverse combustion literature.
NUREG/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIM 1ZE INFILTRATION AT WASTE DISPDSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design. JOHNSON.T.M.s LARSON,T.H.s HERZOG, B. L. s et al.
Illinois, State of.
July 1983.
103pp.
8308250395.
20164:121.
Laboratory tests were conducted on geologic materials selected for use in covers for waste disposal sites:
the materials included fine-grained loess, glacial till, and various coarse grained materials used in drainage systems.
The physical, engineering, and hydrological propetties of these materials were determined in respect to their p erformance in trench covers, i
A' dual-energy, gamma-ray attenuation apparatus was constructed to measure simultaneously the moisture content and density of a laboratory soil column.
Initial experiments indicated that moisture movement in unsaturated materials is inhibited by the presence of an interface between materials of highly contrasting texture.
One dimensional and two-dimensional models were used for computer simulations of moisture movement through several cover designs.
The l
modeling indicated that a layer of coarse-textured, unsaturated material overlain by a layer of fine-grained material, could serve as a barrier to moisture movement.
The effectiveness of the barrier is i
related to the contrast in texture and saturated hydraulic j
conductivity between the layers.
Where moisture break-through occurs, the moisture content of the overlying layer is less than saturation and the pressure head at the interface is less than zero.
4 NUREC/CR-2402 V04: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1. 1 -
National Waste Package Program,0ctober 1982 - March 1983. SOO. P.
Brookhaven National Laboratory.
September 1983.
180pp.
8310070411.
BNL-NUREG-51494.
20675: 001.
This report is part of an ongoing effort to review the national high level waste package prograc.
Contributions of the reference waste form, container and packing material to containment and controlled release of radionuclides after emplacement of the waste package are addressed.
Tuff repository conditions are discussed to j
complement prior discussions on salt and basalt.
Failure /dogradation modes for borosilicate glass waste forms, carbon eteel containers and i
packing materials (bentonite-based and zeolite based) have been reviewed and the licensing information needed to demonstrate compliance with NRC performance objectives have been identified.
NUREC/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HICH-LEVEL HASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEGUENCES. KOCHER.D.C.s SJOREEN,A.L.s BARD,C.S.
Oak Ridge National Laboratory.
July 1983.
227pp.
8308010019.
ORNL-5038.
4 19864:001.
This report discusses uncertainties in predicting (1) radionuclide transport between a repository and the biosphere via 21
~_ _. _.
groundwater flow and.(2) the health risk to man following releases of long-lived radionuclides to the biosphere.
The discussion of groundwater transport focuses on the nature of the sources of uncertainity in predicting.the present geologic environment and its future evolution due to natural geologic processes and to repositorg development and waste emplacement, groundwater hydrology, radionuclide geochemistry, and the interactions among.these phenomena.
The treatment of these uncertainties is primarily qualitative rather than quantitative, because of the lack of evidence that current models can provide realistic predictions of radionuclide transport in groundwater i
for. expected repository environments.
The analysis of uncertainties in the long-term health risk par unit release to the biosphere considers radionuclide concentrations in the environment, radionuclide intake-bu exposed individuals, the dose per unit intake, the number of I
exposed individuals, and the health risk per unit dose.
Quantitative estimates of uncertainties in these factors are given, but the limitations of current models for realistic predictions are also emphasized.
NUREG/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC. ANDERSEN,J.G.s CHU,K.H.s j
SHAUG,J.C.
General Electric Co.'
September 1983.
113pp.
]
8310180085.
EPRI NP-2375.
20826:001.
l TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a reactor system.
The constitutive correlations for shear and heat transfer developed for the Boiling Water Reactor (BWR) version of TRAC are described.
A universal flow reg'ime map has been developed to tie the regimes for shear and the heat transfer into a consistent package.
New models in the areas of interfacial shear, interfacial heat transfer and i
thermal radiation have been introduced.
Improvements have also been F
made to the constitutive correlations and the numerical methods.
All the models have been implemented into the GE version TRACB02 and extensively tested against data.
t NUREG/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4. 7 - MODEL DEVELOPMENT l
TRAC-BWR COMPONENT MODELS. CHEUNG, Y. K. s PARAMESWARAN,V.s SHAUG,J.C.
General Electric Co.
September 1983.
120p p.
8310180076.
EPRI NP-2376.
20026:235.
TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis for the thermal-hydraulic conditions in a l
reactor system.
The development and assessment of the BWR component models developed under the Refill /Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report, i
These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model and upper plenum mixing moael.
These j
models have been implemented into TRAC-B02.
Also a single channel option has been developed for individual fuel channel analysis following a system response calculation.
NUREG/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TF.ST PLAN THOMPSON,J.E.
General Electric Co.
Electric Power Research Institute.
September 1983.
35pp.
8309280497.
EPRI NP-23J3.
20541:185.
A test plan for the BWR Full Integral Simulation Test (FIST)
Program is presented.
The test program will investigate BWR thermal 22
hydrculic roepenco under LOCA cnd ep'orational transient cenditions in scaled DWR-6/218 test apparatus.
In this document, test parameters a
are selected, individual tests are defined, and measurement and data utilization plans are presented.
NJR EG/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EGUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS. GALLUP,D.R.; VANNONI,H.G.
Sandia Laboratories.
September 1983.
32pp.
8310110174.
SANDB2-2726.
20709:297.
This report presents a logical and comprehensive approach to the identification of all plant functions that contribute to the protection of public health and safety from exposure to radiation released by ncclear power plants.
The work is the first task of a proposed three-task program to assist the NRC in developing future regulatory guidance and rulemaking for the quality assurance programr.
of nuclear plants.
The primary result of this task is the development of a logical structure for the identification and ranking of nuclear power plant equipment according to its importance to safety.
The overall measure of importance used in this structure is the risk of the plant endangering public health and safety.
In parallel with the logical structuring of the problem, general methodological approaches for ranking equipment important to safety are outlined The pt oposed second task will formalize the ranking methodology and provide a basis for the development of quality assurance requirements keyed to equipment importance to safety.
NUREG/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A REVIEW OF THE LITERATURE. CHAN M.K.W.;
MISHIMA,J.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
150pp.
8308150407.
PNL-4174.
20054:022.
To determine the effects of fires in nuclear fuel cycle facilities, Pacific Northwest Laboratory (PNL) has surveyed the literature to gather data on the characteristics of combustion products, this report discusses the theories of the origin of combustion with an emphasis on the behavior of the combustible materials commonly found in nuclear fuel cycle facilities.
Data that can be used to calculate particulate generation rate, sire, distribution, and concentration are included.
Examples are given to illustrate the application of this data to quantitatively predict both the mass and heat generated from fires.
As the final result of this review, information gaps are identified that should be filled for fire accident analyses in fuel cycle facilities.
NUR EG/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIADILITY MODEL. SIEGEL,A.I.; BARTTER,W.D.s WOLF,J.J.; et a l.
Oak Ridge National Laboratory.
September 1983.
153pp.
8310190243.
ORNL/TM-8300.
20843:263.
The front end analysis performed for thc nuclear power plant maintenance personnel reliability modeling program consisted of three primary tasks which are addressed within this report.
The first of these was a front-end user survey which investigated the need for and potential content of a structured methodology for nuclear power plant maintenance.
The second task was a literature review of existing human behavioral methodologies and an assessment of their applicability for this program.
The third task was the develooment of a comprehensive program plan for the maintenance reliability model.
Results of these tasks indicated that a computerized model would 23
be very useful end that the type of methodology to be developed should be of the simulation type.
This report also provides a 35-month program plan for its development.
NUR EQ/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL. CAMP, A. L. s CUMMINGS,J.C.; SHERMAN,M.P.s et al.
Sandia Laboratories.
September
.1983.
337pp.
8310190283.
SANDB2-1137.
20839:166.
A manual concerning the behavior of hydrogen in light water reactors has been prepared.
Both normal operations and accident situations are addressed.
Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation.
Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate.
A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful.
Different sections are written at different levels, according to the most likely audience.
The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procedures.
NUREG/CR-2739: DATA BASE FOR DASALT METHODOLOGY. GUZOWSKI, R. V. s CRANWELL,R.M.
Sandia Laboratories.
August 1983.
94pp.
8308180747.
SANDB2-1197.
20102:129.
Most basaltic magma originates in the mantle.
The extrusion of this magma onto the earth's surface produces lava flows.
Flow mo phology depends on the silica and gas content of the lava and the topography and wetness of the land.
An ideal lava flow has a dense, fractured interior with brecciated cooling rinds (interflows) at the base and the top.
Ground-water movement through a sequence of basalt flows is controlled by pressure gradients in the flow systems the aperture, spacing, and filling of fractures and the permeability of interflows and interbeds.
The composition of the water is controlled by the original water composition, mixing of different ground waters, rock mineralogy, duration of water-rock contact, and age of the water.
All of these factors influence the rate at which ions in solution react with the rock and are able to move through the flow sysbem.
Whereas intact basalt is a strong rock, fractures are common and may reduce the in-situ strength.
Most mechanical data are from labcratory experiments on intact rock.
Data on thermal properties are not always consistent for some parameters because of a lack of standardized testing procedures.
NUREG/CR-2805 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1952 - December 1982. LIPPINCOTT E.P.i MCELROY,W.N.
Hanford Engineering Development Laboratory.
July 1983.
63pp.
8308080670.
HEDL-TME-82-21.
19977:204.
The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) has been established by NHC to improve, test, verify, and standardire the physics-dosimetry-metallurgy, damage correlation, and the associated reactor analysis methods, procedures and data used to predict the integrated effect of neutron exposure to LWR pressure vessels and their support structures.
A vigorous research effort attacking the same measurement and analysis problema exists worldwide, and strong cooperative links between the US NRC-supported activities at HEDL, ORNL, NBS, and MEA-ENSA and those supported by CEN/SCK (Mol, Belgium),
24
I EPRI (Palo Alto, USA), KFA (Julich, Germany), and several UK laboratories have been extended:to a number.of other countries and t
laboratories.
These cooperative links are strengthened by the active membership of the scientific staff from many participating countries 1
and laboratories in the ASTM E10 Committee on Nuclear Tachnology and i
Applications.-
Several subcommittees of ASTM E10 are responsible for the preparation of LWR surveillance standards.
NUREG/CR-2835: REVIEW OF MODELS APPLICABLE TO ACCIDENT AEROSOLS.
GLISSMEYER,J.A.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
33pp.
8300080761.
PNL-4294.
19971:124.
Estimates of potential airborne particle releases are essential 1
in safety assessments of nuclear fuel facilities.
This report is a review of aerosol behavior models that have potential applications for predicting aerosol characteristics in compartments containing accident generated aerosol sources.
Such characterization of the accident-generated aerosols is a necessary step toward estimating j
.their eventual release in any accident scenario.
Existing aerosol models can predict the size, distribution, concentration, and composition of aerosols as they are acted on by ventilation, diffusion, gravity, coagulation and other phenomena.
Models developed i
in the fields of fluid mechanics, indoor air pollution and nuclear reactor accidents are reviewed with this nuclear fuel facility I
application in mind.
The various capabilities of modeling aerosol behavior are tabulated and discussed, and recommendations are made for applying the models to problems of differing complexity.
NUREG/CR-2844 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982. BALL,S.J.J CLEVELAND, J. C. ;
e HARRINGTON,R.M.; et al.
Oak Ridge National Laboratory.
August 1983.
24pp.
8309080010.
ORNL/TM-8443/V4.
20323:030.
Work continued on high-temperature gas-cooled reactor safety code development, including both the Fort St. Vrain and the 2240-MW(t) lead plant versions of the three-dimensional core code ORECA, the BLAST steam generator code, and a simplified core model code called SCORE.
Oak Ridge National Laboratory participated in the Nuclear Regulatory Commission siting study for the lead plant with three other J-laboratories.
Investigations continued to determine the status of fission product source-term methodology applicable to postulated severe accidents.
NUREC/CR-2850 VO2: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE j
RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980. BAKER, D. A. ;
PELOGUIN,R.A.
Battelle Memorial Institute, Pacific Northwest Laboratory.
August 1983.
122pp.
8310070442.
PNL-4221.
20674:001.
Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1980.
In addition doses derived from the shutdown reactors at the Three Mile Island site were included.
Fifty-year dose commitments from a one year exposure were calculated from both liquid i
and atmospheric releases for four population groups (infant, child, teenager and adult) residing between 2 and 80 km from each site.
This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age i
group and organ.
Also included for each site is a hictogram showing the fraction of the total population within 2 tn 80 km around each 25 1
v n-
-n n
-.,---. ~, -,,,,. - _. -, -,,,
--nm.,
n
---....,nn c=..
,-~,--.e
,,---,--r-------
~
_ _ ~
~_ -.
site receiving various average dose commitments form the airborne pathways.
The total dose commitment from both liquid and airborne-pathways. ranged from a high of 40 person-rem to a low of 0.02 i
' person-rem with an crithmetic mean of 4 person-rem.
The total i
population. dose from all sites was estimated at 80 person-rem for the 96 million people considered at risk.
The average individual dose commitment from all pathways on a site basis ranged from a low of 6 x 10(minus 6)-mrem to a high of 0.06 mrem.
No attempt was made in this study to determine the e.aximum dose commitment received by any one individual from the radionuclides released at any of the sites.
NUREC/CR-2874 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PRDGRESS REPDRT, OCTOBER 1 - DECEMBER 31,1982. BALL,S.J.; CLAPP.N.E.s CONMLIN,J.C.; et a l.
Oak Ridge National Laboratory.
August 1983.
24pp.
8309000010.
ORNL/TM-8443/V4.
20323:030.
Work continued on high-temperature gas-cooled reactor safety code development,. including both the Fort St. Vrain and the 2240-MW(t)~1ead plant versions of the three-dimensional core code ORECA, the BLAST steam generator code, and a simplified core model code called SCORE.
Oak Ridge National Laboratory participated in the Nuclear Regulatory Commission siting study for the lead plant with three other j
laboratories.
Investigations continued to determine the status of fission product source-term methodology applicable to postulated severe accidents.
NUREG/CR-2938: TAILINGS TREATMENT TECHNIGUES FOR URANIUM MILL WASTE: A Review Of Existing Information. SHERWOOD,D.R.s SERNE,R.J.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
82pp.
8307270360.
PNL-4453.
19802:244.
In this study, three general amelioration methods were identified: neutralization, fixation and specific constituent removal.
At present, neutralization processes appear to be best suited for 1
treating uranium mill tailings because they can, at a reasonable cost, limit the solution concentration of many contaminants, thus reducing the potential for ground-water contamination.
However, the effectiveness of the process depends on the reagent used as well as the waste being treated.
Of the six reagents studied (lime, limestone, caustic soda, and soda ash), combined limestone seems best, especially for tailings containing ferric iron as the limestone economically buffers the solution acidity while the lime takes the pH t o 8. 0, an optimum level for heavy metal removal.
For those tailings containing ferrous iron, lime alone works best.
In general, the costs for the lime / limestone or lime processes range from $0.20 to $1.00 per 3.
1000 gallons of treated water, excluding capital equipment costs.
- NUREG/CR-2982 RO1: BUOYANCY, TRANSPORT,AND HEAD LOSS OF FIBROUS REACTOR INSULATION. BROCARD.D.N.
Alden Research Laboratory.
Sandia Laboratories.
July 1983.
70pp.
8308010003.
SAND 82-7205.
19862:286.
In the event of a loss-of-coolant accident (LOCA) in a nuclear power plant, it is possible that insulation for pipes or other inside i
the containment building could be dislodged by the high energy break Jet.
This insulation debris could affect the recirculation of water from the-sump of the emergency core cooling system (ECCS) by collecting on the screen surrounding this sump.
To help in assess:.ig
-the possible effect of detached insulation on the ECCS, buoyancy, 26
w transport, and head loss characteristics of the insulation were studied experimentally.
Three types of insulation pillows were tested in undamaged, opened, broken up in pieces and shredded conditions.
Small samples of reflective metallic and closed cell insulations were aise issted for transport and buoyancy.
This report contains esperimental data which can be used to assess insulation debris transport and screen b1ccksge effects.
Revision 1 contains additional data and uncertainty analyses, f r.-
head loss characteristics of accumulated insulation fragments.
Filomat fiberglass and mineral wool fibrous insulations were tested.
l NUREG/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR j
POWER PLANTS. BATTLE,R.J.s CAMPDELL,D.J.
Oak Ridge National Laboratory.
July 1983.
395pp.
8308010038.
ORNL/TM-8545.
19861:034.
Reliability of emergency onsite ac power systems at nuclear power plants has been questioned within the Nuclear Regulatory Commission (NRC) because of the number of diesel generator failures reported by nuclear plant licensees and the reactor core damage that could result from diesel failure during an emergency.
This report contains the results of a reliability analysis Of the onsite ac power systems.
Included is a design and operating experience review.
Eighteen plants representative of typical onsite ac power systems and ten f
generic designs were selected to be modeled by fault trees.
Operating experience data were collected from the NRC files and from nuclear plant licensee responses to a questionnaire sent out for this project.
Important contributors to onsite power system reliability were identified, and the costs and improvement in reliability for modifications were estimated.
Sensitivity of the onsite system unreliability to independent diesel generator failure, common-cause failure, and scheduled maintenance unavailability were analyzed, and costs of decreasing the probabilities of failure for these contributors were estimated.
The important factors.affecting onsite 4
ac power system reliability were found to be dependent upon j
plant-specific features.
NUREC/CR-3OO3 VO2:
HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS. BANKS W.W.s GILMORE,W.E.s bLACKMAN,H.S s et al.
EG&G, Inc.
August 1983.
104pp.
8308250264.
EGC-2230.
20175:006.
The report, the second sponsored by the Nuclear Regulatorg Commission, identifies issues and variables related to operator 1
performance in nuclear reactor control rooms where cathode ray tube-generated graphic displays are used.
The first report, NUREC/CR-2469, examines hardware related issues.
This report investigates software related issues most relevant to the design of display screens (e.g.,
safety parameter display systems).
Each
' variable is defined and pertinent data are discussed to support or refute particular display design guidelines.
Recommendations are addressed to determine whether sufficient data exist in human I
performance literature to support a particular suggestion or guideline.
NUREC/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAR F
FACILITIES WITH SEISMIC EQUIPMENT GUALIFICATION IN NUCLEAR PLANTS (A-46). SMITH,P.D.s DONG,R.G.
Lawrence Livermore Laboratory.
Auguct 1983.
92pp.
8308300388.
UCID-19465.
20202:001.
27 l
~-__
A study on the feasibili.tv of using experience data on the performance of equipment in nonnuclear facilities during earthquakes found that such data would be quite useful, if they were developed, in addressing issues concerning the seismic qualification of equspment in operating nuclear power plants located in the eastern United States.
Guidelines and criteria for the use of these data were also developed.
Thirty possible issues associated with seismic qualification of equipment were identified and ranked,. and the current amismic squipment qualification requirements evaluated against each of these issues.
These results and this technique are also believed to be oF use in areas beyond experience data and seismic issues such as in general GA.
I NURE0/CR-3048: CLOSEOUT OF IE BULLETIN BO-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY. FOLEY,W.J.: HENNICK.A.
Parameter, Inc.
September 1983.
35pp.
8310110580.
PARAMETER IE-12.
20714:281.
Cracking of yokes in active valves in a safety-related system was discovered during preoperational testing of the Residual Heat Removal System at Susquehanna 1.
These defective yokes had been supplied to Anchor / Darling Valve Company'by Malcolm Foundry Ccmpany, Inc.
B e c a u r.e Malcolm Foundry had gone out of business, it was necessary to issue the Bulletin to utilities to determine directly whether Malcolm Foundry had provided parts to valve manufacturers other than Anchor / Darling. From results of the extensive survey of valve manufacturers caused by this Bulletin, it was determined that only Anchor / Darling had used Malcolm. Foundry as a source of safety-related valve parts.
On the basis of survey results and a search of Anchor / Darling records, it was found that only seven facilities had i
affected valves.
Utility responses and NRC inspection reports were l
reviewed for 130 current facilities.
Followup action items are proposed for the seven facilities identified by means of Anchor / Darling records.
+
NUREG/CR-3050: CLOSEOUT OF IE BULLETIN 81-02 AND SUPPLENENT: Failure Of Gate-Type Valve's To Close Against Differential Pressure. FOLEY,W.J.)
HENNICK,A.
Parameter, Inc.
September 1983.
51pp.
8310180079.
[
PARAMETER IE-12.
20826:119.
In NUREG-0737 of October 31, 1980, NRC formally provided the utilities with a proposed-program for full-scale qualification testing of gate-type power-operated relief valve (PORV) block valves.
Test results for the first seven valves tested by Electric Power Research Institute (EPRI) indicated that certain models of Westinghouse and Borg-Warner 3-inch valves. failed to close completely.
Later.
Westinghouse reported that its entire line of motor-operated gate valves had this potential deficiency.
In order to bring this potentia 1' problem to the attention of licensees and permit holders and to assure completion of any necessary corrective actions in safety-related systems, the NRC issued IE Bulletin 81-02 on April 9, 1981 and the Supplement covering additional Westinghouse valves on
. August 18, 1981.
Utility responses for 130 facilities with operating licenses, construction permits or licenses for low power testing have been evaluated.
Both the Bulletin and Supplement have been closed out for 110 of the 130 current facilities.
Followup for which followup of a
corrective action will be tracked by the regions as 10 CFR 50.55(e) items.
Investigation of fracture of Limitorque operator housings of modified Westinghouse gate valves at Oconee 2 and Watts Bar i and 2 is recommended.
28
=._
NUNEQ/CR-3060: A CRITICAL REVIEW DF FLDODING LITERATURE. BANKOFF,S.C.3 i
LEE S.C.
Northwestern Univ.
July 1983.
133pp.
8308150425.
20063:250.
1 A number of models have been proposed over the years for flooding, or countercurrent flow limitation (CCFL), for gas-liquid flow.
These may be roughly classified as surface wave stability theories, static equilibrium theories, envelope or limiting operating condition theories, and semiempirical correlations.
The scatter of the data is rather largs, partig because of the different definitions of flooding which have been employed, and partly because of the strong influence of the entrance and erit conditions.
Some' progress has been c
made recently in distinguishing between various critical conditions and locations for floodings these effects tend to distinguish steam-water flooding from air-water flooding.
The various theories are reviewed briefly and comparisons made with representative data.
Directions for future research are indicated.
NUREQ/CR-3083: EXPERIMENT DPERATIONS PLAN FOR THE TH-2 EXPERINENT IN THE NRU REACTOR. RUSSCHER,C.E.s WILSON,C.L.s PARCHEN L.J.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
40pp.
8308010021.
PNL-4559.
19860:254.
A series of thercal-hydraulic and cladding materials deformation i
experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LDCA) Simulation Program.
This report is the formal operations plan i
for TH-2--the second experiment in the series of thermal-hydraulic tests conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada.
The major ob ective of TH-2 was to J
develop the experiment reflood control parameters and the proceduret to be used in subsequent experiments in this program.
In this experiment, the data acquisition and control system was used to control the fuel cladding temperature during a simulated LOCA by using variable reflood coolant flow.
NUREC/CR-3085 V02: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol II - Appendix A And B.1 Thru B.4. Docket No. 50-245.(Northeast Nuclear Energy Company) p AMICO,P.J.
Science Applications, Inc.
- Sandia Laboratories.
August 1983.
375pp.
8308190741.
SANDB2-7212/2.
20099:321.
This report presents the results of the analysis of Millstone Unit One nuclear power plant, which was performed as part of the Interim Reliability Evaluation Program (IREP).
Two of the IREP objectives have been achieved by this analysis.
They are (1) the l
identification of those accident' sequences which can be expected to l
dominate the risk related to the operation of Mi)1 stone 1, more l
. extensive probabilistic risk assessments of Milletone 1.
i NUREC/CR-3005 V03: INTERIM RELIABILITY EVALUATION PRDORAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER. PLANT.Vol III Appendix i
L B. 5 Thru B.B. Docket No.
50-245.(Northeast Nuclear Energy) AMICO,P.J.
Science Applications, Inc.
- Sandia Laboratories.
July 1983.
l 772pp.
8308010053.
SANDB2-7212/3.
19869:001.
l-This report presents the results of the analysis of Millstone j
Unit Dne nuclear power plant, which was performed as part of tha Interim Reliability Evaluation Program (IREP).
p Two of the IREP
[
objectives have been achieved by this analysis.
They are (1) the
[
identification of those accident sequences which can be expected to l
29 1
decincto the riek related to the operation of Millstone 1, and (2) the
. development of syttem models that can be used for future, more extensive probabilistic risk assessments of Millstone 1.
NUREG/CR-3085 VO4: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol IV Appendix D.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
AMICO,P.J.
Science Applications, Inc.
Sandia Laboratories.
July 1983.
720pp.
8308010045.
SANDB2-7212/4.
19867:001.
This report presents the results of the analysis of Millstone Unit One nuclear power plant, which was performed as part of the Interim Reliability Evaluation Program (IREP).
Two of the IREP ob Jectives have been achieved by this analysis.
They are (1) the identification of those accident sequences which can be expected to dominete the vaak related to the operation of Millstone 1, and (2) the development of system models that can be used for future, more extensive probabilistic risk assessments of Millstone 1.
NUREG/CR-3091 VO2: REVIEW OF WASTE PACKAGE VERIFICATION TESTS. Semiannual Report Covering The Period October 1982 - March 1983. SOO. P.
Brookhaven National Laboratory.
August 1983.
130pp.
8309200019.
BNL-NUREG-51630.
20421:228.
This report is part of an ongoing task to specify tests that may be used to verify that engineered waste package / repository systems meet 10 CFR 60 performance objectives for containment and controlled release of high level radioactive waste.
Verification tests for borosilicate glass waste forms and bentonite-and zeolite-based packing materials are analyzed.
NUR EG/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC AIR. SUTTER,S.L.
Battelle Memorial Institute, Pacific Northwest Laboratory.
August 1983.
44pp.
8308230525.
PNL-4566.
20128:336.
Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an esticate of potential airborne releases caused by accidents.
Aerosols generated in accidents are being investigated to develop source terms for these releases.
An upper boundary accidental release event would be a pressurized release of powder or liquid in static air.
Experiments measured the mass airborne and the particle size distribution of aerosols produced by pressurized releases.
Two powder and two liquid sources were used: TiO(2) and depleted uranium dioxide (DUO)2 and j
aqueous uranine (sodium fluorescein) and uranyl nitrate solutions.
Pressurization level and source size were significant variables for the airborne powder releases.
For this experimental configuration, the liquid releases were a function of pressure only.
100 g and 350 g of DUO (1 m dia) and TiO(2) (1.7 m dia) powders and 100 cm(3) and 350 cm(3) of uranine and uranyl nitrate solutions were released at pressures ranging from 50 to 500 psig.
The average of the largest fractions of powder airborne was about 24%.
The maximum amount of liquid source airborne was significantly less, about 0.15TX.
Powders ranged from 5 to 19 m AEDs liquids ranged from 2 to 29 m.
All of the releases produced a significant fraction of respirable particles of 10 m and less.
30
NUREQ/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES.
CHU,M.S.Y.s ORTIZ,N.R.
Sandia Laboratories.
WAHI,K.K.s et al.
Science Applications, Inc.
August 1983.
190pp.
8308250552.
SANDO2-2969.
20174:029.
This report summarizes the work performod for the Nuclear Regulatory Commission (NRC) to assess its proposed Rule (10CFR60) for disposal of high-level radioactive wastes in geologic repositories.
The objectives of this project were to provide information that NRC could use to evaluate the rationale for the technical requirements in the Rule, to respond to public comments on the proposed Rule, and to analyze the benefits of alternative criteria for the final Rule.
Three of the numerical criteria of the Performance Objectives of the Rule were analyzed in detail in this study.
The three criteria pertain to: a containment period, a controlled release rate, and a pre-waste emplacen.ent ground-water travel time.
A series of parametric analyses were performed on the potential releases of radionuclides to the accessible environment in order to determine the impact of the three criteria on compliance with the draft EPA Standard.
The study also examined the achievability of the three numerical criteria based on an assessment of the existang technology.
Parameters and techniques that could be used to assess compliance were identified.
In addition to the three numerical criteria, certain other requirements were also addressed in varying degrees of detail.
The structure and content of the Rule were analyzed for completeness, consistency, and redundancy.
As part of this study, a panel of experts was formed to address and debate the issues relating to the requirements contained in the proposed Rule.
A number of recommendations have been made in the report based on the analyses performed and the outcome of the panel discussions.
NUREC/CR-3118 VO1: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 1: Executive Summary. ZANETOS,M.A.s WARLING,J.C.s MARSH,G.M.
Battelle Memorial Institute, Columbus Laboratories.
September 1983.
30pp.
8309280551.
20524:329.
This three part report describes the results of a research project intended to provide quantitative estimates of the risk of cancer and other diseases among persons exposed to hazardous substances in the workplace.
The risk estimates presented are based on a comprehensive review of recent epidemiologic studies.
Primary emphasis was placed on studies of workers exposed to hazardous chemicals under conditions typical of a given industry over a working lifetime.
Despite finding over 100 chemicals associated with increased incidence of disease, convincing dose-response trends existed for only a few.
Although there were notable exceptions (arsenic, asbestos, PAH's, etc), it was generally impossible to estimate risk per-unit of dose in a manner analogous to calculations which exist for radiation exposure.
The principal reason for tnis is the lack of adequate environmental monitoring data for the specific chemical, locations, and time periods needed to estimate individual cumulative doses.
In addition to analyzing risk in terms of increased incidence of specific diseases, we also examined life expectancy and gears of life lost due to cancer in four selected occupational groups via a life table model which allowed for competing risks.
The resulting estimates indicated that the overall life expectancy of these groups was generally greater than that of the general population but that the workers suffered greater loss of life expectancy (LLE) due to cancer than their counterparts in the general population.
31
Published estimates of LLE due to radiation induced cancer indicate that at doses of 0.5 rem /gr, radiation workers are projected to suffer I
less LLE than any of the four non-nuclear cohorts examined.
At 5 rem /gr (the MPD) or higher, LLE may be greater than one or more of the cohorts examined.
1 NUREG/CR-3118 VO2: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 2: Supporting Documentation And Review Of l
Epidemiologic Studies Of Specific Chemicals. ZANETOS,M.A.s 4
WARLING,J.C.s MARSH,G.M.
Battelle Memorial Institute, Columbus Laboratories.
September 1983.
264pp.
8309280547.
20523:001.
This three part report describes the results of a research project intended to provide quantitative estimates of the risk of cancer and other diseases among persons exposed to hazardous substances in the workplace.
The risk estimates presented are based j
on a comprehensive review of recent epidemiologic studies.
Primary emphasis was placed on studies of workers exposed to hazardous chemicals under conditions typical of a given industry over a working lifetime.
Despite finding over 100 chemicals associated with increased incidence of disease, convincing dose-response trends existed for only a few.
Although there were notable exceptions (arsenic, asbestos, PAH's, etc), it was generally impossible tc estimate risk per-unit of dose in a manner analogous to calculations which exist for radiation exposure.
The principal reason for this is the lack of adequate environmental monitoring data for the specific chemicals, locations, and time periods needed to estimate individual cumulative doses.
In addition to analyzing risk in terms of increased
?
incidence of specific diseases, we also examined life expectancy and years of life lost due to cancer in four selected occupational groups I
via a life table model which allowed for competing risks.
The resulting estimates indicated that the overall life expectancy of these groups was generally greater than that of the general populatien but that the workers suffered greater loss of life expectancy (LLE) due to cancer than their counterparts in the general population.
Published estimates of LLE due to radiation induced cancer indicate that at doses of 0.5 rem /yr, radiation workers are projected to supper less LLE than any of the non-nuclear cohorts examined.
At 5 rem /gr i
( the MPD) or higher, LLE may be greater than one or more of the 2
cohorts examined.
i NUREG/CR-3118 VO3: HEALTH EFFECTS DF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 3: Abstracts of NIOSH Criteria Documents.
ZANETOS,M.A.s WARLING J.C.
Battelle Memorial Institute, Columbus Laboratories.
September 1983.
185pp.
8309280536.
20524:001.
This three part report describes the results of a research project intended to provide quantitative estimates of the risk of cancer and other diseases among persons exposed to hazardous substances in the workplace.
The risk estimates presented are based on a comprehensive review of recent epidemiologic studies.
Primary emphasis was placed on studies of workers exposed to hazardous chemicals under conditions typical of a given industry over a working lifetime.
Despite finding over 100 chemicals associated with increased incidence of disease, convincing dose-response trends existed for only a few.
Although there were notable exceptions (arsenic, asbestos, PAH's, etc), it was generally impossible to estimate risk-per-unit of dose in a manner analogous to calculations 32
which exist for radiation exposure.
The principal reason for this is the lack of adequate environmental monitoring data for the specific chemicals, locations, and time periods needed to estimate individual cumulative doses.
In addition to analyzing risk in terms of increased incidence of specific diseases, we also examined life expectancy and years of life lost due to cancer in four selected occupational groups via a life table mode? which allowed for competing risks.
The resulting estimates indicated that the overall life expectancy of these groups was generally greater than that of general population but that the workers suffered greater loss of life expectancy (LLE) due to cancer than their counterparts in the general population.
Published estimates of LLE due to radiation induced cancer indicate that at
' doses of 0.5 rem /gr, radiation workers are projet*:ed to suffer less y
LLE than any of the four non-nuclear cohorts examined.
At 5 rem /gr (the MPD) or higher, LLE may be greater than one or more of the cohorts examined.
NUREG/CR-3122: PDTENTIAL DAMAGING FAILURE MODES OF HIGH-AND E
MEDIUM-VOLTAGE ELECTRICAL EGUIPMENT. HOY,H.C.
Oak Ridge National Laboratory.
August 1983.
90pp.
8310070426.
DRNL/NSIC-213.
p 20674:238.
The electrical equipment failures of both nuclear and nonnuclear public utilities were reviewed.
Those failures that could pose an
[
additional problem to surrounding and connected equipment were defined.
The literature was searcheds utilities, repair shops, and
+
large electrical equipment users were contacted for Failure i n f orma ti on.
The data were reviewed in detail, and failure modes were determined.
Sample cascade failures are discussed.
The failure rate of electrical equipment in utilities is historically quite low.
Nuclear plants record too few failures to be statistically valid, but failures that have been recorded show that good design usually restricts the failure to a single piece of equipment.
1 NUREG/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS l
FROM THE H.B.
RUBINSON STEAM ELECTRIC PLANT. HARRISON,F.L.s BISHOP,D.J.: RICE, D. W. s et al.
Lawrence Livermore Laboratory.
. August 1983.
38pp.
8309200523.
UCRL-53047, 20451:227.
Water and bedload sediments were sampled at the intake and discharge areas of the H.
B. Robinson Steam Electric Plant and at a control area upstream of the plant in November 1979 and May and July 1990.
Total copper concentrations in discharge waters ranged from 33 to 79 micro g/L, whereas those in intake waters ranged from 27 to 50 micro g/Ls copper in the water in the control area ranged from 3.1 to 5.8 micro g/L.
In most samples, copper in the soluble fraction appeared equally as labile and beund.
The concentration of dissolved organic carbon in the water samples from the impoundment area showed little variation with times the values range from 3.8 to 5.7 mg C/L.
The organic carbon in the particulate fraction was less than or equal t o mg C/L.
Determination of molecular size distribution of the copper species by ultrafiltration of discharge and control waters indicated that the largest concentration of copper appeared in the greater than 10,000, less than 100,000 molecular weight fraction both in untreated and UV-oxidized water.
Copper concentrations in intact bedload l
sediments of the intake area ranged from 4.4 to 260 micro g/g dry
-weights those in the discharge area ranged from 6.1 to 72 micro g/g.
Copper in the upstream sediments was very low, less than or equal to 3.7 micro g/g.
Copper distribution coefficients of intact bedload i
sediments in the intake and discharge areas ranged from 300 to 21,000.
33 l
NUREG/CR-3133: THE SENSITIVITY OF ADULT,EMBRYDNIC AND LARVAL CRAYFISH PROCAMBARUS CLARKII TO COPPER. RICE,D.W.s HARRISON,F.L.
Lawrence Liv 6rmore Laboratory.
July 1983.
29pp.
8307270319.
UCRL-53948 19811:341.
The copper sensitivity of adult, embryonic, and larval stages of the crayfish Procambarus clarkii was determined with flow-through bioassag methods.
From the family of curves of cumulative mortality versus duration of exposure, median lethal times were determined and used to construct comparative toxicity curves.
The 20-d (480-h) median lethal concentrations show the order of copper sensitivity of P.
clarkii life-history stages: larvae (120 microgram Cu/L) greater
'than adults (1300 microgram Cu/L) greater than embryos (3700 microgram Cu/L).
In addition, comparisons in percent hatching were made between P.
clarkii embryos exposed early and late during embryological development.
Embryos exposed to copper concentrations as low as 250 gram Cu/L 600 h prior to hatching showed onig 17% hatching, whereas
'l embryos exposed to as high as 2840 microgram Cu/L 250 h prior showed 100% hatching.
Larvae exposed to copper as embryos were less sensitive than those exposed after hatching.
_ Copper concentrations in adult crayfish tissues were analgred.
The remains of adult crayfish showed exposure-related increases in copper accumulat2on.
The P.
f
-clarkii muscle, gill, liver, and kidney showed no dose-related increases in copper accumulation, though gills showed significant increases in copper accumulation at doses of 480 microgram Cu/L and higher.
NUREG/CR-3138: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FOR A PWR GEOMETRY UNDEP SIMULATED REFLOOD CONDITIONS.
THOMAS,D.G.s CDMBS.S.K.
Oak Ridge National Laboratory.
August 1983.
489pp.
8308250449.
ORNL/TM-8204.
20176:001.
i The Instrument Development Loop (IDL) Program is part of the International 2D/3D Refill and Reflood Experimental and Analytical Research Program.
The two major objectives of the IDL Program were to simulate expected flows at the core / upper plenum interface during the reflood phase of a postulated LOCA and to develop and calibrate instrumentation systems for mass flow measurement at the core / upper plenum interface.
Three experimental facilities were used in these studies:
a one-bundle air / water loop a three-bundle air / water loop, and a one-bundle steam / water loop.
Three flow regimes were identified and studied:
(1) all liquid down, '(2)~ counter-current flow and (3) cocurrent flow in which both
]
gas (or vapor) and liquid go up.
Some of the significant achievements of the IDL program include:
the tie-plate drag body was developed and tested successfullys measurement with tie-plate drag body was shown to be equivalent to the delta P measurements the tie plate drag body gave a useful measurement in pure dowrflow situations the combination of drag-turbine correlates with mass flow for high upflows and empirical correlations were developed for countercurrent flow.
I j_
NUREG/CR-3148: INDUPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MODI CODES 1
AT BNL IN FY 1981. SAHA,P.s JO, J. H. ; NEYMOTIN,L.; et al.
Brookhaven l
National Laboratory.
July 1983.
133pp.
8308030459.
B NL-NUR EG-51645.
19910:195.
i This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAPS/ MODI codes during Fiscal Year i
1981.
A large variety of separate-effects experiments dealing with i
j 34 I.
_ - -, -,,.. _. ~,
(1). steady-state and transient critical flow, (ii) level swell, (iii) flooding and entrainment, (iv) steady-state flow boiling..(v) integral economizer once-through steam generator (IEDTSO) performance, (vi) bottom reflood and (vii) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2.
In addition, the early part of an over-cooling transient which occurred at the Rancho Seco Nuclear i
Power Plant on March 20, 1978 was also computed with an updated version of TRC-PD2.
Three separate-effects tests dealing with (i) transient critical flow, (ii) steady-state flow boiling, and (iii)
IEDTSG peformance were also simulated with RELAP5/ MODI code.
Comparisions between the code predictions and the test data are presented.
A number of areas where further model ng improvements are i
necespary have been identified for both codes.
'tveral suggestions have also been made to aid the code developers i making further improvements.
NUREG/CR-3152: CDRRELATION OF THE ANALYTICAL SOLUTION WITH EXPERIMENT AL DATA DN TME ELECTROLYSIS POTENTIAL PROBE. BETHMANN,W.F.
Oak Ridge National Laboratory.
July 1983.
62pp.
8308030473.
ORNL/TM-8650.
19913:001.
In response to a loss of coolant (LOCA) in a water cooled nuclear reactor, the reactor vessel is reflooded with water from standby storage tanks to carry off heat that would otherwise damage fuel rods, vessels and other structures.
During the later phases of the reflooding, a mixture of steam and water flows between the structures and a flowing film of water adheres to the surfaces. To gain some j
knowledge about the of these films and to aid heat transfer calculations in designing reactors, instruments have been developed for the measurement of film thickness, flow velocitios, and mass flow rates in recctors and their simulators.
The Electrolysis Potential (EP) Probe is one such instrument and is designed to measure average liquid velocity inside the films. It was disecvered that if the EP Probe was excited with a' constant voltage of sufficient magnitude to produce electrolysis, the current through the sensor decreased in a repeatable manner with increasing average film velocity.
The EP Probe was capable of measuring liquid film velocities over the range from 5 to 50 cm/s with an accuracy of approximately plus or minus 5% of reading under laboratory conditions.
It shoulo be noted, however, L
that field use proved to be impractical because of the extreme sensitivity to water-chemistry which could not be adequately measured or controlled.
NUREG/CR-3155: MODIFICATION OF DCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING DN THE INNER SURFACE. SAUTER,A.;
CHEVERTDN, R. D. s ISKANDER,S.K.
Oak Ridge National Laboratory.
July 1983.
37pp.
8308050058.
ORNL/TM-8649.
19943:230.
The computer code OCA-I calculates the temperature distribution through the wall of a cylinder during a thermal transient and then j
performs a two-dimensional linear-elastic fracture-mechanics analysis to obtain stress-intensity factors for long surface flaws, considering both pressure and thermal loads.
The. code has been particularly useful in evaluating flaw behavior in reactor pressure vessels during overcooling accidents, but it has not previous 1g treated the stainless steel cladding on the inner surface of the vessel as a discrete region.
Although the cladding is quite thin compared with the base material, the large difference in thermal conductivity and coefficient of thermal expansion between the two materials results in a c
significant effect of the cladding on stress-intensity factors for 35
surface cracks.
Thus, tha cladding was recently included as a discrete region in OCA-I.
NUREQ/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
DODD, C..V. s DEEDS,W.E.s MCCLUNG,R.W.
Oak Ridge National Laboratory.
July 1983.
17pp.
8308100482.
ORNL/TM-8741.
20000:328.
Eddy-current inspection is.the most suitable method for rapid boreside evaluation of steam generator tubing.
However, sen11 flaws can be masked by the effects of harmless variables such as tube supports.
To. identify the critical properties accurately and reliably in the presence of extraneous signals caused by variations of unimportant properties, sufficient information is needed to identify harmful variations and to reject harmless-ones.
For this reason we have been developing instrumentation capable of measuring both the amplitude and phase nf the eddy-current signal at several different frequencies as well as computer equipment capable of processing the data quickly and reliably.
Our probes and test conditions are also computer-optimized.
The most recent probe design embodies an array of small flat " pancake" coils and promises to improve the detection of small flaws and the rejection of tube support signals.
We have also verified the accuracy of our computer programs for calculating the signals produced by defects in tubing.
NUREG/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY GDALS FOR NUCLEAR POWER PLANTS. MORRIS.F.A.s HOOPER,R.L.
Battelle Human Affairs Research Centers.
- Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
66pp.
8300030455.
PNL-4641.
39908:313.
The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants.
The report emphasizes one particular category of institutional problems:
the possible use of safety goals as a basis for legal challenges _to NRC actions, and the resolution of such challenges by the courts.
Three types of legal issues are identified and analyzed.
These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof.
Second is the particular i
formulation of the goals.
Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism.
Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.
NUREG/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR. BLAKELY,K.D.s HOWARD,G.E.;
WALTON,W.B.; et al.
Anco Testing Laboratory, Inc.
July 1983.
479pp.
8300080771.
19971:229.
Response of a nonlinear finite element model of the Heissdampfreaktor recirculating piping loop (URL) was compared to measured data, representing the physical benchmarking of a nonlinear i
model.
Analysis-test comparisons of piping response are presented For i
snapback tests that induced extreme nonlinear behavior at the URL system.
In addition, seismic analyses are reported for a nonlinear 36 r.
y P
and a-linear piping model.
The nonlinear model predicted peak stresses that were 20% to 50% lower than the linear model's results.
NUREG/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON
)
SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
BEBAHANI,A.s MILLER.D.W.s TALNAGI J.W.
Ohio State Univ.
July 1983.
j 415pp.
8308180712.
20101:048.
This report describes a method for testing neutron sensor (compensated ion chamber) response time to meet the provisions of 13A G67.06, IEEE Std 338-1977, and NRC Regulatory Guide 1.118.
The method requires the perturbation of a high voltage detector (sensor) power supply and measurement of the sensor response.
The response to a l
perturbation of the power supply is related to the response of the sensor to a transient change in neutron flux.
The research program experimentally and analytically evaluated l
the relation between a power supply perturbation and a neutron transient flux change.
It also evaluated the change in sensor end signal respense times through changes in ion chamber fill gas and the use of different delag cables for a signal cable.
NUREG/CR-3188: THE STEADY STATE POROUS BODY ASSEMBLY CDDE (SPAC): Users Manual. VAN TUYLE,0.J.
Brookhaven National Laboratory.
July 1983.
56pp.
B308030440.
BNL-NUREG-51654.
19912:261.
The Steady State Porous Body Assembly Code (SPAC) was derived directly from the ENERGY III code, which was developed and reported by Dr.
E. U.
Khan.
The ENERGY III code was part of a code series, including the ENERGY I and ENERGY II codes, that Dr. Khan used as a test bed for the porous body models, on which his Doctoral thesis (MIT) was based.
The ENERGY III code itself, as initially written, required extensive computer core space to execute, was fairly costly (slow) to use, and suffered truncation errors that were quite noticeable.
In contrast, the SPAC code requires only 7% of the space, 3% of the run time, and 2.5% of the expense that are required to use ENERGY III.
The truncation errors have been eliminated and this report is intended to produce sufficient documentation.
With the improved run characteristics, SPAC also eliminates the need for ENERGY I and ENERGY II, they were alternate versions of ENERGY III with, as
. limiting assumptions made to save execution cost.
NUREG/CR-3196:
DRUG AND ALCOHOL ABUSE:THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY. RADFORD,L.R.s RANKIN,W.L.s l
BARNES V.s et al.
Battelle Human Affairs Research Centers.
July 1983.
135pp.
8308050202.
PNL-4679.
19943:085.
l This report describes the nature, prevalence, and trends of drug l
and alcohol abuse among members of the U.S.
adult population and among personnel in non-nuclear industries.
Analogous data specific to the nuclear utility industry are not available, so these data were gathered in order to provide a basis for regulatory planning.
The
- nature, prevalence, and trend information was gathered using a computerized literature search, telephone discussions with experts, and interviews with employee assistance program representatives from the Seattle area.
This report also evaluates the possible impacts that drugs and alcohol might have on nuclear-related Job performance, based on currently available nuclear utility Job descriptions and on the scientific literature regarding the impairing effects of drugs and alcohol on human performance.
Employee assistance programs, which can g
L r
L.
37
('
be used to minimize or eliminate Job performance decrements resultinp frem drug or alcohol abuse, are also discussed.
I NUREG/CR-32OO VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING l
_ PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
DODD, C. V. s DEEDS.W.E.s SMITH,J.H.s et al.
Oak Ridge National Laboratory.
August 1983.
Bp p.
8309080023.
ORNL/TM-0796/V1.
20323:079.
Eddy-current inspection is the most suitable' method for rapid boreside evaluation of steam generator tubing.
However..small flaws can be masked by the effects of harmless variables such as tube supports.
To identify the critical properties accurately and reliably in the presence of extraneous signals caused by variations of unimportant properties sufficient information is needed to identify harmful variations and to reject the amplitude and phase of the eddy-current signal at several different frequencies as well as computer equipment capable of processing the data quickly and j
reliably.
Our probes and test conditions are also computer-optimized.
The most recent probe design-embodies an array of small flat " pancake" coils and improves our ability to detect small flaws and reject tube support signals.
By using different formulas for calculating the sizes of flaws near a tube support, we have been able to increase the detectability of such flaws.
NUREC/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS. HSIEH P.A.s NEUMAN,S.P.s SIMPSON,E.S.
Arizona, Univ. o f, Tucson, AZ.
July 1983 i
193pp.
8308000609.
19976:085.
A comprehensive methodology of hydraulic testing in fractured rocks is presented.
The methodology utilizes geological and geophysical information as background.
It consists of conventional single-hole packer tests in. conjunction with a newlu developed cross-hole packer-testing procedure.
The cross-hole procedure consists of injecting fluid at a constant rate into. packed-off
~
interval in one borehole and monitoring pressure variations in packedroff intervals within neighboring boreholes.
If injection takes place at a constant pressure, the analysis considers only those data obtained after the flow rate stabilizes.
Borehole orientation is generally unrelated to the principal hydraulic conductivity directions which, therefore, need not be known a priori.
The method yields
.information about the directional nature of hydraulic conductivity in three dimensions on a scale comparable to the distance between the test boreholes.
In addition to providing all six components of the hydraulic conductivity tensor, the crois-hole method also vields the specific storage of the-fractured rock mass.
While the theory behind this method treats the rock as a uniform anisotropic porous medium
{
the test provides detailed information about the degree to which such assumptions may actually be valid in the field.
NUREG/CR-3214:
SUMMARY
OF NUCLEAR REGULATORY COMMISSION'S LOFT PROGRAM EXPERIMENTS. NALEZNY,G.L.
EG&G, Inc.
August 1983.
354pp.
8309080014.
EGG-2248.
20322:019.
This document is a summary of all the experiments conducted by EG&G Idaho, Inc. for the-United States Nuclear Regulatory Commission in the Loss-of-Fluid Test (LDFT) facility.
The LOFT facility is a 50-MW(t) pressurized water reactor.(PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear 38
..g_
core behavior during various types of accident conditions.
The LDFT Experimental Program included 10 mini-blowdown experiments to check j
out the LDFT system and 30 integral system experiments.
The integral j
systes; consisted of six experiment groups including nonnuclear large breaks, and nuclear large breaks, small breaks, intermediate breaks, operational transients, and anticipated transients with multiple I
failures.
The Program was successfully conducted, and all the 3
objectives were met which:
?!) Provided data required to evaluate the i
adeguacy of and to improve the analytical methods currently used to predict the response of larga PWRs to postulated accident conditions, the performance of engineered safety features (ESF) with particular j
emphasis on emergency core coolant systems (ECCS), and the guantitative margins of safety inherent in the performance of the EGF, Ide'tified and investigated any unexpected events (s)
(2) n or threshold (s) and developed analytical techniques that adequately describe and account for the unexpected behavior (s).
(3) Evaluated and developed methods to prepare, operate, and recover systems and
- plant f or and from reactor accident conditions.
(4) Identified and i.
investigated methods by wich reactor safety can be enhanced, with emphasis on the interaction of the operator with the plant.
l NUREG/CR-3215 V01: DRGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview. DSBORN,R.N.s OLSON,J.s SOMMERS,P.E.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
August 1983.
73pp.
8308310018.
PNL-4655.
20241:001.
E Report of organizational literature focusing on potential
~
relationships among organizational factors and safety indicators applicable in nuclear power plant safety assessments.
This report is the first of two volumes.
It presents an overview i?
of the organizational literature, and introduces a model for analgring, singly and in combination, nuclear utility industry organizational factors (e.g.,
environment, history, governance.
design), intermediate outcomes (e.g.,
quality, compliance, efficiency) and safety indicators (e.g.,
quality assurance, technical specifications, surveys, evaluations).
Finally, recommendations are j
made for additional research on safety assessment approaches for assessing nuclear utility and power plant performance from organizational perspectives.
The second volume of this report presents a detailed analysis of organizational factors,~ intermediate outcomes and safety indicators presumed to influence nuclear power plant safety.
.NUREG/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
[
OSBORN, R. N. s DLSON, J. s SOMMERS. P. E. s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
August 1983.
281pp.
8309080020.
.PNL-4655.
20310:029.
Tuo volume report investigating potential relationships among organizational factors and safety indicators, applicable in nuclear power plant safety assessments.
This report is the second of-two volumes.
It presents a detailed i
analysis of organizational factors, intermediate outcomes and safety related indicators presumed-to influence nuclear power plant safety.
Major elements of an organizational analysis approach to safety assessment are discussed.
Included are perceived links between organizational design, management, external demands, and safety.
L Additionally, empirical data currently available on nuclear utility 39
._.--_.-__.__7
_.--m-organizatians and measures of safety are examined for their relatooness.
Finally, regulatory experiences of selected federal agencies are described to show how organizational analyses are relevant to the regulatory needs to the NHC.
Volume I. presents a model for analyzing organizational factors and safety, and recommendations for incorporating organizational analyses.
NUREG/CR-3219 V01: DRAFT TECHNICAL POSITION SUBTASK 1.1: WASTE PACKAGE PERFORMANCE AFTER REPOSITORY CLOSURE. DAVIS,M.S.s SCHWEITZER,D.E.
Brookhaven National Laboratory.
July 1983.
50pp.
8308300291.
BNL-NUREG-51658.
20202:163.
This Draf t Technical Position pr ovides guidance to the DOE on the issues and information necessary for the NRC to evaluate the performance of high level nuclear waste packages after repository closure.
Options available to the DOE for~ complying with proposed 10 CFR 60 re described and advantages and disadvantages of each are discussed.
Examples of demonstrability, predictability and reasonable assurance are discussed.
The types of testing, modeling and statistical analyses that can be used to demonstrate performance are considered.
.NUREG/CR-3221: MC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION. STIRPE,D.s GOLDMAN,A.S.s GUTMACHER.R.G.
Los Alamos Scientific Laboratory.
July 1983.
150pp.
8309050119.
LA-9708-MS.
19942:292.
The United States Nuclear Regulatory Commission (NRC) has developed an Advanced Notice of Proposed Rulemaking for Materials Control and Accounting (MC&A) that requires licensees to use process monitoring data to provide timely (24 h) detection and resolution oP potential losses of strategic special nuclear materials (SSNM).
To furnish guidance to licensees in complying with the loss-detection requirements of this proposed MC&A Reform Amendment, (published September 10, 1981) the NRC has initiat6d two example-system development studies.
The second of these studies, entitled " Material Control and Accounting Upgrade Rule Example System Beta," is the subject of this report and differs from the first study by allowing modifications to the reference process design to develop a best case approach.
In addition, this second study addresses only the detection-provisions of the Reform Amendment and response features needed to achieve detection goals because these are the areas where advanced technologies and process changes will have an impact on system compliance.
This final report gives materials balance standard deviations for the candidate MC&A systems suggested for Example System Beta and also describes statistical tests used to investigate compliance of these suggested MC&A systems with the loss-detection requirements of the proposed Reform Amendment.
NUREG/CR-3223: BWR REFILL REFLDOD PROGRAM FINAL REPORT. MYERS,L.L.
General Electric Co.
- Electric Power Research Institute.
September 1983.
142pp.
8310180118.
EPRI NP-3093.
20825:176.
The BWR Refill-Reflood Program is part of the continuing Loss-of-Coolant Accident (LOCA) research in the United States which is Jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Com'pany.
The current program expanded the focus of this research to include full scale experimental evaJuations of multidimensional and multichannel 40
y effects during system refill.
The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients.
A summary description of the complete program is provided including the principal findings and main conclusions of the program.
The results of the progran have shown that multidimensional and parallel channel effefts have the potential to significant1g impreve the system response over that observeJ in single channel tests.
The best estimates have been shown to prcperly model physical phenomena l
and hava the capability to handle realistic BWR system interactions.
1 Sufficient data are presented to support these conclusions.
NUREG/CR-3234: THE POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH ELECTRICAL PENETRATION ASSEMBLIES UNDER SEVERE ACOIDENT CONDITIONS.
SEBRELL,W.
Sandia Laboratories.
August 1983.
61pp.
8308250559.
SAND 83-0538.
20175:283.
An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I),
indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies.
Because of the high temperatures, it was i
postulated ir a previous study that the sealants would fail and all the electrical penetretion assemblies would leak before structural failure would occur.
Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure.
The results of this study, however, show that this conclusion does not hold For PWRs because in the worst accident sequence, the long time containment gases stabilire to 350 degrees fahrenheit.
Many electrical penetration assemblies have been designed and tested to this temperature and, therefore, should have a very low potential for leakage.
BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakaLe.
To evaluate the physical integrity in each plant requires individual investigation because loss of physical integrity depends on the seal materials, which vary from plant to I
plant.
NUREG/CR-3238: PNASIM: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIGUE OF MASS FLOU MEASUREMENT. KEHLER,P.
Argonne National Laboratory.
August 1983.
75pp.
8308180718.
ANL-82-51.
20091:231.
This report contains a listing and a sample run of PNASIM, a computer program that simulates the processes involved in flow measurements by the Pulsed-Neutron Activation (PNA) technique.
The program is written in BASIC for the HP Model 9845 desk computer.
It is useful for the planning and optimization of PNA measurements, for the analytical derivation of correction factors to be applied to two-phase density measurements, for the analytical derivation of the proper exponent to be used in the velocity equation, and for the
[
prediction of the overall accuracy of PNA mass flow measurements.
l 1
l NUREG/CR-3241: PROBABILISTIC METHODS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SEGUENCES IN LOOP-TYPE LMFBRS. JAMALI,K.M.A.
Brookhaven i
National Laboratory.
August 1983.
328pp.
8308250257.
B NL-NOREG-51664.
20172:001.
-The aim of the Probabilistic Accident Progression Analysis (PAPA) l 41 LE
p described herein is to establish a framework for better use of the probability measures first, as a basis for deterministic calculations, and second, as a part of a comprehensive method for risk assessment in its own right.
The existing techniques related to (1) improvements in the existing approaches for acquisitions and analyses of accident s eguenc ess (2) defining a new measure of probabilistic importance that i
aids in the ranking of sequences of highly uncertain events; and (3) implementation of new techniques for quantification of dependent failures of similar components, are discussed and the state of the art is reviewed.
The PAPA spproach is described.
The techniques of PAPA are applied to a class of Protected Transients (transients in which the reactor is successfully shutdown) in the Clinch River Breeder l
Reactor Plant (CRBRP).
The results of the application of these
-techniques are described, and conclusions and suggestions for future work are presented.
NUREQ/CR-3257: RELAP5 ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
THOMPSON, S. L. s KMETYK L.N.
Sandia Laboratories.
August 1983.
104pp.
8308250437.
SAND 83-0832.
20174:219.
The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to I
determine the ability of various systems codes to predict the detailed thermal / hydraulic response of LWRs during accident and off-normal conditions.
The RELAP5/ MODI code is being assessed at SNLA against test data from various integral and separate effects test facilities.
As part of this assessment matrix, a turbine trip rapid cooldown transient performed at the LDFT test facility has been analyzed.
The results show that RELAPS/ MOD 1 can predict the experimental behavior of LDFT test L6-7/L9-2 in detail.
However, careful selection of modeling options and adjustment of boundary conditions within the experimental uncertainties is required.
It was also necessary to modify the LOFT pump descriptions in order to match the natural circulation flow data during L9-2.
If other options are selected and/or boundary conditions l
are not adjusted then the calculated results can deviate quite far l
from the test data in the.later stages of the transient.
A number of code problems were detected in this study.
The more serious included large mass and energy errors, and difficulties with some Junction models.
Modeling guidance for RELAP5/ MOD 1 is presented as are suggestions for code improvements.
NUREQ/CR-3258: RELAPS ASSESSMENT: SEMISCALE NATURAL CIRCULATION TESTS S-NC-2 AND S-NC-7. MCCLAUN,J.M.; KMETYK L.N.
Sandia Laboratories.
July 1983.
100pp.
8308030463.
SAND 83-OB33.
19912:161.
The RELAP5 independent assessment project at Sandia Nationc!
Laboratories is part of an overall_ effort funded by the NRC to determine the ability of-various systems codes to predict the detailed thermal-hydraulic response of LWRs during accident and off-normal
(
conditicns.
The RELAP5 code is being assessed at SNLA against test data from various integral and separate natural circulation tests performed at the Semiscale facility have been analyzed.
The results show that RELAP/ MOD 1 qualitatively describes all modes of natural circulation correctly, but with quantitative disagreement on both the absolute magnitude of the two-phase natural circulation (which is always overestimated), and an the mass inventories at which various mode transitions (from single phase to two-phase to reflux cooling) occur.
There is no censistent pattern to the inventory discrepancies seen.
The major difficulties encountered in the calculations were unphysical oscillations in two-phase and 42
r reflux natural circulation, which could be eliminated by reducing the time step (indicating problems in the time step control algorithm).
These observations agree very well with the conclusions drawn f rom earlier RELAP5 analyses of PKL natural circulation tests, and indicate possible problems in the interface drag model (affecting the amount of liquid entrained at any given inventory), and in the two-phase loss coefficients for abrupt area changes (affecting~the peak values in two-phase natural circulation).
NUREG/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS. MYERS,0.A.s TYLER,S.W.s MITCHELL. D. H. s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
September 1983.
53pp.
8310110347.
PNL-4694.
20714:023.
This study evaluated the performance of existing and alternative leak detection systems for lined uranium mill tailings ponds.
Existing systems for detecting leaks at uranium mill tailings conds investigated included groundwater monitoring wells, subliner drains, and Igsimeters.
Three alternative systems which demonstrated the ability to locate leaks in bench-scale tests were moisture blocks, soil moisture probes, and a soil resistivity system.
Several other systems in a developmental stage are described.
For proper performance of leak detection systems (other than groundwater wells and lysimeters), a subgrade is required to assure lateral dispersion of a leak.
Methods to enhance dispersion are discussed.
Cost estimates were prepared for groundwater monitoring wells, subliner drain systems, and the three experimental systems.
Based on the results of this report, it is suggested that groundwater monitoring systems be used as the primary means of leak detection.
However, iP a more responsive system is required due to site characteristics and groundwater quality criteria, subliner drains are applicable For ponds with uncovered liners.
Leak-locating systems for ponds with covered liners require further development.
Other recommendations are discussed in the report.
NUR EQ/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
BERNATOWICZ,H.s BORGONOVI.C.s EPPERSON,D.
et al.
Science Applications. Inc.
August 1983.
292pp.
8309190497.
SAIO1583-368LJ.
20418:003.
The purpose of the study was to investigate the effect of parameters on nondestructive techniques used for the assay of nuclear material, firstig by an assessment of the current knowledge of parametric ef7ects, and second1g by carrying out experiments on selectec parameters.
The techniques examined were passive gamma NDA, passive neutron NDA, active NDA densitometry, and calorimetry.
For each technique, operational, intrinsic, and external parameters were identified in a systematic way, and the need for experiments to provide additional information on the effect of selected parameters was also identified.
In the area of passive gamma NDA, experiments were conducted to investigate the effect of nonhomogeneity of nuclear material, and a theoretical approach for the description of the effect of nonhomogeneity was developed.
The approach can be used as a corrective technique.
In the area of passive neutron NDA, experiments were conducted to determine the effect of small amounts of matrix i
l having different moderating and absorbing characteristics.
In the area of active NDA, experiments were conducted to investigate the effects of matrix and spatial distribution on coincidence counting in 43
e a fission multiplicity detector, and to test methods for minimizing the effects.
NUREG/CR-3265: EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE l
MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS. HOBSON,D.O.;
j
' N ANSTA,D, R. K.
Oak Ridge National Laboratory.
July 1983.
27pp.
8308100491.
ORNL/TM-8661.
20008:001.
We examined the effects of off-specification procedures on the mechanical properties of half-bead weld repairs.
The name half-bead is derived from the specification that half the thickness of the initial weld lager be ground off before the second layer is deposited.
In this study the heat-affected zones of a weldment made with both all and none of the first lager removed were tested for toughness, hardness, and microstructural differences, and the results were compared with the properties of a prototypical half-bead repair made under ASME Boiler and Pressure Vessel Code. Sect. XI, guidelines.
The results.of this limited study showed no apparent Justification for the requirement to grind off half the first lager in this type of weld repair.
The graded electrode sizes used to make the welds probably had more to do with the weld properties than did the range of first-layer thicknesses used in this study.
i i
NUREG/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS. CURRERI,J.s COSTANTINO.C.; SUBUDHI,M.s et al.
Brookhaven National Laboratory.
September 1983.
106pp.
8309290558.
BNL-NUREG-51667.
20580:310.
A generic floor response spectra has been developed for use in the qualification of electrical and mechanical equipment in operating nuclear power plants.
The characteristics of 1000 floor response spectra were studied to determine the generic spectra.
A procedure for its application to any operating plant has been established.
The procedure uses as much or as little information that currently exists at the plant relating to the question of equipment qualification.
The general approach in the development of the response spectra was to study the effects on the dynamic characteristics of each of the elements ir. the chain of events that ~goes between the loads and the responses.
This includes the loads, the soils and the structures.
A free-field earthquake response spectra was used to generate horizontal earthquake time histories.
The excitation was applied through the I
soil and into the various structures to produce responses in equipment.
From all of the models of soils and structures, floor response spectra were generated at each floor level.
A horizontal generic floor response spectra is proposed for the top level of a generic structure.
Reduction factors are applied to the peak acceleration for equipment at lower levels.
NUREQ/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH,CO32-)
RECIPROCAL SYSTEM SALLACH,R.A.
Sandia Laboratories.
September 1983.
40pp.
8309280556.
SANDB2-1980.
20526:309.
Phase diagrams for the (Na(+), Ca(2+)) (OH(-), CO(3)(2-)
recaprocal system and the NaOH - Ca(OH)(2) system are described.
The solubility of CACO (3) in NaOH for temperatures up to 550 degrees centigrade is defined.
Results of the studies are used to interpret sodium / concrete interaction data.
44
l NUREQ/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS. WEED,J.E.s HUDSON,H.G.s WYTHE,D.M.s et al.
Lawrence Livermore Laboratory.
August 1983.
222pp.
8308300848.
UCID-19703.
20216:001.
Electromagnetic interference (EMI) levels were determined experimentally by passive measurement at various locations within two commercial nuclear power plants (NPPs).
A description of the measurement locations for each plant and a discussion of how the measurements were made are provided.
Photographs of the original time and frequency domain data are provided.
The results presented were obtained from one plant while it was operating at high power conditions and from a second plant that was at cold shutdown conditions following a refueling outage.
The high frequency currents measured were generally less than 0.5 amperes peak-to peak, however, currents over 5 amperes were measured.
Dominant frecuencies tended to be under one megahertz with some as high as 5.2 megahertz.
Also presented are levels of voltage pulse peaks occurring on buses which were monitored for several days using power line disturbance monitors.
These data show that the peak voltages tend to be under 100 volts for the 120 volt lines and some pulses over 400 volts for the 480 volt circuits.
Most of the radiated magnetic field data measured was dominated by the 60 Hertz line frequency.
The greatest magnetic field measured was 54.4 milliampere / meter in excess of 100 kilohertz.
These and the current data are presented in raw form and in a summary fashion to serve as an aid in establishing an EMI data base.
NUREG/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4). WILSON C.L.s MOHR,C.L.;
HESSON, G. M. s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
224pp.
8308010002.
PNL-4669.
19862:066.
A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program by Pacific Northwest Laboratory (PNL).
The fourth materials experiment (MT-4) was the seventh experiment in the series of thermal-hydraulic and materials deformation / rupture experiments conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada.
This experiment was funded by the U.S.
Nuclear Regulatory Commission (NRC) to evaluate ballooning and ruptura during adiabatic heatup in the temperature range of 1033 to 12OOK (1400 to 1700 degrees farenheit).
The 12 test rods in the center of the 32-rod bundle were l
initially pressurized to 4.62 MPa (670 psia) to insure rupture in the correct temperature range.
All 12 test rods raptured with an average strain of 43.7% at the maximum flow blockage elevation of 2.68 m (105.4 in).
Experimental data for the MT-4 transient experiment and
(
p ost-test measurements and photographs of the fuel are presented in l
this report.
NUREG/CR-3277: RELAP5 ASSESSMENT: SEMISCALE MdD-3 SMALL DREAK TEST.
SUMMERS.R.M.
Sandia Laboratories.
September 1983.
204pp.
8309280462.
SAND 83-!O38.
20525:001.
The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal-hydraulic response of LWR's during accident and of f-normal conditions.
The RELAP5/ MODI code is being assessed at SNLA against test data from various integral and separate effects test facilities.
l 45 l
t
~
l~
l' As part of.this assessment matrix, a series of small break experiments performed at the Semiscale Mod-3 facility have been analyzed.
The results show that:RELAPS/ MOD 1 will predict reasonably well J
many aspects'of the four small break transients analyzed.
Many i
problems'were encountered'in matching the initial primary and l
which in turn affected the 1
secondary steady state conditions,,the calculations predicted the J
transient calculations.
Although l
relative severity of the experiments in terms of mass inventories and distributions, quantitatively they generally predicted a more severe transientfthan was observed experimentally.
The code cannot correct 1g calculate'the break flow in situations which often arise in small break. scenarios where substantial amounts of liquid are pooled in the bottom of the piping but.the break orifice is uncovered.and
' discharging high quality steam.
This modeling deficiency may have contributed ta the' general under-prediction of mass inventories for these transients.
Also, the current rewetting criteria cause the clad thermal response after. core uncovery to be miscalculated in the presence of numerically-induced density oscillations.
Subcooling behavior in the steam generators generally was not calculated properly.
Some of the discrepancies between calculated results and experimental-data can be attributed to uncertainties in app 1ging boundary'conditionsi which in some cases (e.g.,
break flow rates, pump curves, environmental heat losses, and heat loss distribution) are quite large and can significantly affect the results.
' NUREQ/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
GIDA,R.G.s KDESTEL A.
Los Alamos Scientific Laboratory.
July 1983.
122pp.
8308010016.
LA-9749-MS.
19864:225.
Predicted are the Sequoyah and McQuire ice-condenser-contaznment atmosphere pressure and temperature responses-during' degraded-core accidents that involve the release and combustion cf large' quantities of h'gdrogen.
Predictions lare' based on (1) an appropriately modified version of the US Nuclear Regulatory Commission (NRC) subcompartment analysis'coJe COMPARE and (2) a Burning-Lean-with-Turbulence model for the estimation of burn parameters.
The COMPARE code modifications are i ~
described.
Also, COMPARE-calculated.results are compared with utility-calculated results submitted to the NRC.
i NUREQ/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT. BOYACK,B.E.
Los Alamos Scientific Laboratory.
September 1983.
506pp.
8310200389.
l LA-9704-MS,
'20065:096.
The Transient Reactor Analysis Code (TRAC) is being developed at the. Lost Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light-water reactors.
The TRAC-PF1~ program provides this capability for pressurized water L.-
-reactors and for many thermal-hydraulic experimental facilities.
This report presents the results of initial development assessment' calculations performed with TRAC-PF1 before its public l
release.
All calculations were performed with the same code version.
The assessa.ent' set consisted of 10 calculations, 9 of which. predicted integral effects in the Semiscale, Loss-of-Fluid Test Facility, and Primarkreislaufe facilities, while the remaining calculation predicted a single-effect test conducted in the Dartmouth air-water countercurrent flow apparatus.
Computer run times required to predict
-each integral-test are reported and compared with earlier TRAC versions'where possible.
l
NUREG/CR-3282: INTERNAL HYDROGEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT RDOM TEMPERATURE. AHN,T.M.s SOO, P.
Brookhaven National Laboratory.
August 1983.
40pp.
8308250274.
BNL-NUREG-51671.
20164:298.
Notched and unnotched specimens of hydrogenated TiCode-12 (Ti-0. 3 Mo-0.8 Ni) have been mechanically tested in tension at room temperature.
For hydrogen ~ levels in excess of the as-received value, internal hydrogen embrittlement is present.
A mechanism for the embrittlement is proposed.
- NUREG/CR-3283: FIRST EXCURSION PROBLEMS FOR GAUSSIAN VECTOR PROCESSES.
- K AKO, T. s SHINOZUKA. M. s TSURI,A.
Brookhaven National Laboratory.
August.1983.
65pp.
8308150409.
BNL-NUREG-51672.
20044:265.
Under the assumptions that the linear structural response analys.s is acceptable and that the under1ging random processes are Gaussian, the present study deconstrates (1) how a structural reliability analysis can accommodate some of the more realistic limit state conditions, particularly hyperpolyhedral and hyper-spherical i
limit state conditions in the terms of stress, (2) how a finite element analysis can be incorporated into the reliability analysis to be performed under these limit state conditions, and (3) how the random process idealization of loading conditions can be used in the q
dynamic structural reliability analysis in conjunction with such limit j
. state conditions, by taking the earthquake ground acceleration process as an example.
I NUREC/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS SHARMA,S.s REICH,M.; CHANG,T.Y.
Brookhaven National Laboratory.
September 1983.
137pp.
8309280569.
BNL-NUREG-51673.
20524:190.
This report presents an in-depth review of current modeling techniques-and nonlinear analysis procedures for reinforced concrete structural evaluations.
A detailed discussion is given of numerous L
mathematical models that have been proposed in the literature for modeling the nonlinear maierial behavior of concrete and steel and for representing-the complex concrete-steel interaction effects.
Finite element discretizations of reinforced concrete structures based on discrete, embedded and smeared approaches are described.
Special a
modeling techniques for representing such effects as cracking, crushing, tension stiffening, shear ~ transfer, concrete-steel bond degradation and loss of prestress are oiscussed in detail
- Further, various static.and dynamic nonlinear solution procedures are reviewed in terms of solution accuracy and computational efficiency.
Numerical examples are presented to demonstrate the extent of variations in analytical predictions that can result when different modeling approaches are used.
L NUREC/CR-3289: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES. Estimates Based on Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, 1976-1981. MEACHUM,T.R.s ATWOCD,C.L.
EC&G, Inc.
July 1983.
200pp.
8308030477.
EGG-2258.
19911:151.
. This report presents estimates of' common cause fault rates and related quantities, based on Licensee Event Reports of instrumentation and contro1' assemblies in nuclear reactors.
The Licensee Event Report data base is-briefly described, and-imperfections in the data are discussed.
The components are grouped into assemblies, for which rates are estimated.
For estimating rates, the binomial failure rate 47
model is used, extended to allow for the substantial observed plant-to plant variability, and for_ shocks that by their nature cause all the-assemblies in a sustem to fail.
Every quantity is estimated j
by both a point estimate and a 90% interval.
All rates are expressed per calender hour.
NUREG/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1903. SWANSON,D.O.s CASTLE,J.N.s ANDERSON,P.D.s et al.
Applied Science Associates, Inc.
September 1983.
297pp.
8310050224.
20635:318.
This final report describes work performed as part of consultive 2
support to the U.S.
Nuclear Regulatory Commission on a variety of topics related to the material interactions that would occur following a postulated core meltdown accident in a Light Water Reactor.
The main topics addressed include (1). an evaluation of candidate core retention system materials for the Zion and Indian Point (Z/IP) nuclear reactors, (2) an examination of various core retention system concepts with emphasis on the restrictions imposed by the conditions present in the Z/IP reactors. (3) development of a concept for retrofitting the Z/IP reactors with a molten core retention device, (4) reactor cavity conditions following vessel failure, including the existence of a hydraulic jump and molten core entrainment composition, strengthe gas pressure and surface tension, (6) a discussion of reactor containment system materials in the context of interactions with core debris, (7) reactions forming combustible hydrogen, carbon monoxide and hydrocarbons, (8) a discussion of the properties and behavior of urania and the prospects for decontamination of aerosols in a BWR suppression pool, and (9) the effect of water salinity and acidity on concrete.
NUREQ/CR-3306: LWR AND HTGR COOLANT DYNAMICS: THE CONTAINMENT OF SEVERE DYNAMICS. THEOFANOUS.T.G.s OHERSON,P.s NOURBAKHSH,H.P.s et al.
Purdue Univ., West Lafayette, I N.
July 1983.
279pp.
8300010031.
19863:001.
This is the final report of a project containing three major tasks.
Task I deals with the fundamental aspects of energetic fuel / coolant-interactions (steam explosions) as they pertain to LWRcore melt accidents.
Task II deals with the applied aspects of response, and includes consideration of energetic fuel / coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena.
Finally, Task III is concerned with HTCR loss of forced circulation accidents.
This report is organized into three major parts corresponding to these three tasks respectively.
Each part is presented as an autonomous unit with its own numbering system, table of contents, summary, nomenclature (if appropriate), referencesi Figures, and Appendices.
NUREQ/CR-3307 V01: REACTOR SAFETY RESEARCH PRDORAMS.Guarterly Report. January -March 1983. EDLER,S.K.
Battelle Memorial Institute, Pacific Northwest Laboratory.
July 1983.
70pp.
8308090027.
PNL-4705-1.
19969:298.
This document summarizes work performed by Pacific Northwest
. Laboratory from January 1 through March 31, 1983, for the Division of
. Accident Evaluation and the Division of Engineering Technology, U. S.
Nuclear Regulatory Commission.
Evaluations of nondestructive examination (NDE) techniques and instrumentation are reporteds areas 48 w.,
.-__-m
,se_,9 y
y
~
g.
l l
of investigation include demonstrating the feasibility of determining I
the strength of structural graphite, evaluating the feasibility of detecting and analgring flaw growth'in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing.the integrity of pressurized water reactor steam generator tubes where service-induced degradation has-been indicated.
Experimenta1' data and analytical models are being provided to aid in decision making regarding pipe-to pipe impacts following postulated breaks in high-energy fluid system piping.
Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions.
Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including severe fuel
~ damage tests at the NRU reactor, Chalk River, Canadas fuel rod deformation and severe fuel damage tests for the Super Sara Test 4
Program, Ispra, ItalyJ the instrumented fuel assembig irradiation
. program at Halden, Norways and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho.
NUREC/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTU.7AL LOADS. HWANG,H.s WANG,P.C.; SHOOMAN,M.s et al.
Brookhaven National' Laboratory.
August 1983.
314pp.
8308150474.
BNL-NUREG-51678.
20065:025.
For safety evaluation of seismic category I nuclear structures, probabilistic characteristics of the actual loads acting on the structures are needed.
A study was conducted to determine the probabilistic characteristics of the operational and accidental loads using the consensus estimation technique.
A study committee on nuclear power plant structural loads was formed and experts in the nuclear industry were invited to participate.
The concensus study was carried out using two rounds of questionnaires.
The responces to the second round questionnaire were statistically evaluated and the-results are documented in this report.
The results are Judged to be tentative because of the limiled numbers of responses for each category of the load surveyed.
Since very little probabilistic load l
data exists in the literature, the consensus study results supplemented with Judgements should be of considerable value in the 1
probability-based reliability analysis of nuclear structures.
NUREC/CR-3327: ANALYSIS OF STRONG-MOTION DATA FROM THE NEW HAMPSHIRE EARTHGUAKE OF 18~ JANUARY 1982. CHANG,F.K.
Army, Dept. o f, Army Engineer Waterways Experiment Station.
September 1983.
76pp.
4 I
8310070154, 20671:193.
This report provides and presents an analysis of the strong-motion data from the New Hampshire earthquake of 18 January 1982 latitude 43.5 degrees north, longitude 71.6 degrees west.
Thirty-six~accelerograms were recorded and digitized.
The digitized accelerograms integrated velocities and displacements are presented in graphic form.
The attenuation curves of the three.orthogonal components for peak acceleration, peak velocity, peak displacement,.and velocity spectrum intensity on the crest, abutment, and downstream sites of the five U.S. Army Corps of Engineers earth dams were plotted and j
compared.
The transverse component at the Franklin Falls Dam shows i
the. highest energy.
However, its attenuation rate is faster than the vertical and longitudinal components.
The transverse component of the accelerograph located on the right abutment of Franklin Falls Dam (8 49
km from the-epicenter) recorded a maximum acceleration of 0.55 g's, which is the highest acceleration'ever recorded in the eastern United States. -The integrated maximum velocity is 5.59 cm/sec, which is very low in comparison with other earthquake records in the western United States.
NUREG/CR-3329 VO1: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983. THOMPSON,S.L.
Sandia Laboratories.
August 1983.
78pp.
8309060108.
SAND 83-1171.
I 20282:265.
The RELAPS independent assessment project is part of a multi-faceted effort sponsored by the NRC to determine the ability of various system codes to predict the detailed thermal-hydraulic response of LWRs during accident and off-normal conditions.
The version used for the FYO2 essessment project was RELAP5/ MOD 1/ CYCLE 14, the latest publicly released version available at the time this project was begun.
The RELAPS code is being assessed at SNLA against test data from various integral and separate effects experimental test facilities.
The majority of the FY82 calculations using MOD 1 have been completed, and all are underweg.
Results from most of the MODI analyses were summarized in the last quarterly and are described in detail in topical reports.
The completed analyses include three LOB 1 large break tests, a PKL natural circulation test series, four Semiscale small break tests, a FLECHT SEASET steam generator separate effects test, and LOFT operational transients L6-7/L9-2 and L9-1/L3-3 end small break test L3-6/L8-1.
Results of MOD 1 analyses completed this quarter are summarized.
Brief individual status reports on these completed and other ongoing LOFT, Semiscale NC, B&W steam generator, Semiscale UT and BCL calculations (all being done with MOD 1) are given in the main body of this quarterly, as is the current status of the
' MOD 1.5 are given in the main body of this quarter 1g, as is the current status of MOD 1.5 code.
NUREG/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL. PULLIAM,R.s PRICE.H.E.:
BOGARRA, J. s et al.
Oak Ridge National Laboratory.
August 1983.
192pp.
8309080015.
ORNL/TM-8781.
20317:197.
This report describes a general' method for allocating control functions to man or machine during nuclear power plant (NPP) design, or for-evaluating their allocation in an existing design.
The research examined soms important characteristics of the systems design process.: and the results make it clear that allocation of control functions is an intractable problem, one which increases in severity with the_incrersing complexity.of systems.
The method is reported in terms of specific steps which should be taken during the early stages of a new system design, and which will lead to an optimal allocation at the functional design level of detail.
The procedures described are not-expected to provide an ultimate solution to the allocation-of-functions problem.
However, these procedures can at least assure that allocation of control functions is considered during design in an' orderly and rational way.
They should substantially i
advance the general understanding of this problem and the ability of l
the design community to allocate control functions to humans or automation in complex systems.
[
' NUR EC/CR-3333: NEUTRON SPECTRAL CHARACTERIZATION OF THE FOURTH NUCLEAR REGULATORY COMMISSION HEAVY SECTION STEEL TECHNOLOGY 1T-CT 7
50 L
--J i
IRRADIATIDN EXPERIMENTS: Dosimetry And Uncertainty Analysis.
STALLMAN, F. W. s BALDWIN,C.A.s KAM,F.B.K.
Dak Ridge National Laboratory.
August 1983.
52pp.
8308250309.
DRNL/TM-8789.
'20164:232.
In support of the fourth series of the Heavy Section Steel i
Technology metallurgical irradiation experiment, which is sponsored by the Nuclear Regulatory Commission, a complete spatial map of damage exposure parameter values has been determined for each of the four metallurgical capsules.
The values of fluence > 1.0 MeV, fluence >
0.1 MeV, and displacements per atom are determined using a least squares logarithmic (LSL) adjustment procedure with input values taken from extensive dosimetry measurements and neutron transport calculations.
Adjustments were made " globally," processing simultaneous 1g the dosimetry measurement at different locations and from different capsules.
Using this procedure, uncertainties for the i
damage exposure parameter values can be reduced to less than 8%
standard deviation.
~
NUREQ/CR-3334 VO1: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983. PUCH,C.E.
Oak Ridge National Laboratory.
September 1983.
191pp.
8310050217.
ORNL/TM-8787/V1.
20636:354.
The. Heavy-Section Steel Technology (HSST) Program is an engineering research activity conducted by the Oak Ridge National Laboratory for the Nucler Regulatory Commission.
The program comprises studies related.to all areas of the technology of materials fabrica'ed 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 cracklike flaws.
Current work is organized into seven 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, (6) stainless steel cladding investigations, and (7) environmentally assisted crack growth studies.
NUREO/CR-3336: EVALUATION OF NUCLEAR DECOMMISE;0NING PROJECTS. Summary Report.Ames Laboratory Research Reactor.nocket No. 50-116. LINK,B.W.;
MILLER R.L.
United Nuclear Corp.
July 1983.
48pp.
8308050105.
19942:076.
This document summarizes the available information concerning the decommissioning of the Ames Laboratory Research Reactor (ALRR), a i
five-megawatt heavy water moderated and cooled research reactor.
The data were placed in a computerized information retrieval / manipulation
-system which permits its future utilization for purposes of comparative analysis.
This information is presented both in detail in its computer output form and also as a manually assembled i
1
-summarization which highlights the more important aspects of the decommissioning program.
Some comparative information with reference to generic decommissioning data extracted from NUREQ/CR 1756,
" Technology, Safety and Costs of Decommissioning Nuclear Research and Test Reactors," is included.
NUREG/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC SIMULATION. DE JARLAIS,C.
Argonne National Laboratory.
July 1983.
122pp.
8308030467.
ANL-83-44.
19911:025.
51 l
ov,
W 4h Inverted annular flow was simulated adiabatically with turbulent water-Jets, issuing-downward from long aspect nozzles, enclosed in gas annuli._ Velocities, diameters, and gas species were varied, and core Jet length, shape, break-up mode, and dispersed-core droplet flow at high relative velocities.
For both of the above transitions from inverted annular flow, l
correlations for core Jet length were developed by extending work done on free liquid Jets to-include this coaxial, Jet disintegration 3
phenomenon.
Jet break-up length is. correlated as a function of jet diameter,. jet Reynolds number, Jet Weber' number, void traction, and gas Weber number-
= Correlations for core shape, break-up mechanisms and dispersed core droplet size were developed by extending free Jet stability, roll wave entrainment, and churn turbulent flow regime studies.
Due to-the usefulness of various Jet break-up studies in correlating the results of this odiabatic inverted annular flow experiment, this report also includes a review-section on Jet stability.
. NUREG/CR-3341: -PROBABILISTIC DESCRIPTIONS OF' RESISTANCE OF t
SAFETY-RELATED NUCLEAR STRUCTURES. ELLINGSWOOD,R.
Commerce, Dept.
o f, National Bureau of Standards.
August.1983.
57pp.
8308260041.
B NL-NUR EG-51681. '20187:017.
j Calculations of limit state probabilities for safety-related nuclear structures require a' knowledge of the probability distributions that describe their resistance.
This study considers the applicability of existing statistical data for describing the 3 -
resistance of steel and reinforced concrete nuclear plant structures.
Probability distributions are recommended that can be used to assess i
the-reliability of nuclear structures and to select appropriate resistance criteria for probability-based structural design.
NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DDE TO DEGRAD/IION OF-STEAM GENERATOR AND REACTOR COOLANT PUMP
- SUPPORT & Seismic Cefety Margins Research Program. BOHN, M. P. s WELLS,J.E.; SHIEH,L.C.; et al.
Lawrence Livermore Laboratory.
August.1983.
105pp.
8310110032.
UCID-19719.
20703:018.
i During the NRC licensing review for the North Anna Units 1 and 2 f
pressurized water reactors (PWRs), questions were raised regarding the potential for low fracture toughness of steam generator and reactor l
' coolant pump. supports.
Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues.
The work described in this report was performed to l
provide the NRC with a quantitative evaluation of the value/ impact implications of the various options of resolving the fracture i
[
toughness ouestions.
j.
This report presents an assessment of the probabilistic risk L
associated with nil-ductilit failures of steam generator and reactor l
coolant pump structural' support systems during seismic events, performed'using the Seismic Safety Margins Research Program codes and data bases.
Two cases were analyzed.
In Case i the support failure was assumed to occur at the safe-shutdown earthquake (SSE), while in Case 2 the systems were assumed to fail at.twice the SSE.
The public dose assuming reduced capacity supports was found to be about 32 man-REM / gear (Case 1) and 12 man-REM / gear (Case 2h compared to the base case (normal support capacity) of 6.6 man-REM / gear.
52
' NUREG/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 38). NOTLEY,F.
Los Alamos Scientific Laboratory.
August 1983.
78pp.
8309080013.
LA-9596-MS.
20323:176.
The results of a Transient Reactor Analysis Code posttest analysis of the Cylindrical Core Test Facility Evaluation-Model Test agree very well with the results of the experiment.
The good agreement obtained verifies the multidimensional analysis capability of the TRAC code.
Because of the steap radial power profile, the importance of using fine noding in the core region was demonstrated (as compared with poorer results obtained from an earlier pretest prediction that used a coarsely noded model).
NUREC/CR-3359 VO1: PHYSICS OF REACTOR SAFETY.Guarterly Report, January -
March-1983.
- Argonne National Laboratory.
July 1983.
54pp.
8308100448.
ANL-83-11 VO1.
19997:174.
This Guarterly Progress report summarizes work done during the months of January-March 1983 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Office of Nuclear 2
Regulatory Research of the U.S.
Nuclear Regulatory Commission.
The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by aembers of the Reactor Safety Appraisals Section.
Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.
An executive summary is provided including a statement of the findings and recommendations of the report.
NUREG/CR-3362: THE INFLUENCE OF LYMPHOHEMATOPOIETIC DISEASE IN ALTERING RADIATION INDUCED CANCER SENSITIVITY IN NICE. CAIN,C.R.
California, Univ. o f, Davis.
August 1983.
128pp.
8308230740.
20145:240.
Using murine models with the pre-existing 1gmphohematopoietic dyscrasias (RF, NZB, W/WY, _BXSB, Beige), we determined relative radiation sensitivity between these strains and hematologicalig normal controls.
The main lifetime' effects observed following protracted (approximately 20 R/ day for 28 days) whole body irradiation were:
1) higher incidence of neoplasia and decreased survival in irradiated diseased vs. irradiated healthy mice, and 2) decreased latency in diseased strains.
In many cases, mice with lymphohematopoietic i
dyscrasias had baseline alterations alterations in susceptible cellular populations (e.g.
increased lymphoid progenitor and effector cell populations) that may reflect targets for radiation-induced transformation.
Some diseased strains also exhibited enhanced recovery and overreplicative responses of Igmphohematopoiesis following radiation exposure.
This may have allowed for fixation of transformed cells.
Although enhanced sensitivity to protracted irradiation was observed in diseased mice, the magnitude of this response was relatively small compared to the high accumulated exposure.
Significant expression of life-shortening occurred relatively late in life in most cases.
It was concluded that although the diseased mice exhibited enhanced sensitivity, that at current occupational exposure limits latency would exceed lifespan, thus no significant effect would manifest.
NUREC/CR-3364: PRESENCE OF PATHOGENIC MICROORGANISMS IN POWER PLANT COOLING WATERS. TYNDALL,R.L.
Oak Ridge National Laboratory.
August 1983.
46pp.
8308100715.
ORNL/TM-8009.
20102:229.
The content of Legiannaires' Disease Bacteria (LDB) in air 53
p i
4 discharged from industriel and powar cooling towers was compared with the concentrations found in the cooling-tower basin water.
Concentrates of the power plant air samples were also tested for the presence of-free-living amoebae.
The content of LDB in basin water was generally 1 x 10(6) to 10 x 10(6) times greater than that found in comparable' volumes of discharge air.
The content of LDB in air samples correlates with other sampling data that indicate that-the amount of water discharged in cooling-tower plumes is generally one part of water by volume to 10 x 10(6) to 100 x 10(6) parts of air.
Nonpathogenic Naegleria were isolated from one air sample taken in the mist generated at the base of natural-draft cooling tower.
l NUREG/CR-3370: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS
SUMMARY
REPORT NORTH CAROLINA STATE UNIVERSITY RESEARCH AND TRAINING REACTOR.
I LINK,B.W.s MILLER,R.L.
United Nuclear Corp.
August 1983.
38pp.
8309200009.
20424:001.
Thiv document summarizes information from the decommissioning of the NCSUR-3 (R-3),
a 10 KWt university research and training reactor.
The decommissioning data were placed in a computerized information retrieval / manipulation system which permits future utilization of this i
information in pre-decommissioning activities with other university l
reactors of similar design.
The information is presented both in some j
detail in its computer output form and also as a manually assembled summarization which highlights the more significant aspects of the decommissioning project.
Generic decommissioning data extracted from j
NUREG/CR 1756, " Technology, Safety and Costs of Decommissioning Nuclear-Research and Test Reactors," and data from the decommissioning of the Ames Laboratory Research Reactor (ALRR), a 5 MWt research i
reactor, is also included for comparison.
- NUREG/CR-3371 Vol: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL RDOM CREWS. Volume 1: Project Approach And Methodology. BURGY,D.s LEMPGES,C.s MILLER,A.s et al.
General Physics Corp.
September 1983.
254pp.
8309200568.
GP-R-221020.
20453:308.
A task analysis of nuclear power plant control room crews was I
perfctmed by General Physics Corporation and Biotechnology, Inc. For i
the Office of Nuclear Regulatory Research, US NRC.
The task analysis II methodology used in the project is discussed and compared to traditional task analysis and Job analysis methods.
The objective of the project was to conduct a crew task analysis that would provide data for evaluating six areas: (1) human engineering design of control rooms, (2) the numbers and types of control room operators needed with requisite' skills and knowledges, (3) operator qualification and i
i training requirements, (4) normal, off-normal, and emergency operating l
procedures, (5) Job performance mids, and (6) communications.
A generic structural framework for assembling the large task data base was employed from observations and videotaping of crew behaviors during 44 operating sequences conducted at 8 power plant sites.
The results of the data collection effort were compiled in a computerized task database.
Six demonstrations for verifying the suitability of the analytical approach and for suitability analysis of each of the 6 areas were performed and Jescribed.
Volume 1 detafis the Project i
Approach and Methodology, and Volume 2 provides the Data Results including a description of the computerized task analysis data format.
j.
NUREQ/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results. BURGY,D.; LEMPGES,C.; MILLER,A.s et al.
I 54 i
n.--,,
---.n-..
.-w
_-_-.,------.--n
'00ncrol Physico Corp.
Septesbor 1983.
353pp.
8310070436.
GP-R-221020.
20673:001.
I A task analysis of nuclear power plant control room crews was performed by General Physics Corporation and Biotechnology. Inc. for the Office of Nuclear Regulatory Research, US NRC.
The task analysis methodology used in the project is discussed and compared with traditional task analysis and Job analysis methods.
The ob Jective of the projact was to conduct a crew task analysis that would provide data for evaluating six areas: (1) human engineering design of control rooms, (2) the numbers and types of control room operators needed with requisite skills and knowledges, (3) operator qualification and training requirements. (4) normal, off-normal, and emergency operating procedures, (5) Job performance aids, and (6) communications.
A
~
generic structural frame-work for assembling the large task data base was employed from observations and videotaping of crew behaviors during 44 operating sequences conducted at 8 power plant sites.
The results of the data collection effort were compiled in a computerized task database (SEEK) Six demonstrations for verifying the suitability of the analytical approach and for suitability analysis of each of the 6 avecs were performed and described.
Volume 1 details the Project Approach and Methodology, and Volume 2 provides the Data Results including a descrip tion of the computerized task analysis data format.
NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION. SONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES. CHENG,L.Y.s DREW,D.A.4 LAHEY,R.T.
Rensselaer Polytechnic Inst., Troy, NY.
July 1983.
322pp.
8308030452.
19909:015.
A wave dispersion relation has been derived from a one-dimensional, two-pressure, two-component, two-fluid model.
A modified Rayleigh's equation was used as the closure law which related the phasic pressures.
Comparison of analytical results with experimental data indicated that the virtual mass effect is very important in the determination of the speed of sound in a two phase mixture.
The propagation velocity of a pressure pulse was related to the group velocity which was evaluated from the wave dispersion relation for a propagating wave.
The effect of the differential terms on critical two phase flow due to virtual mass acceleration was assessed by comparision of model predictions with critical two phase flow data.
The relationship between critical flow and wave propagation was also established.
It has been demonstrated that both critical flow and wave propagation data can be used to deduce the functional form of the virtual volume coefficient, Cvm(alpha), although neither were able to provide data on the virtual mass acceleration parameter, lambda (alpha).
NUREQ/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EGUILIBRIUM GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE. STERNER,R.W.s LAHEY,R.T.
Rensselaer Polytechnic Inst., Troy, NY.
July 1983.
94pp.
8308080616.
19976:273.
Subchannel measurements were performed in order to determine the equilibrium quality and mass flux distribution in a four rod bundle, using air / water flow.
An isokinetic technique was used to sample the Flow in the center, side and corner subchannel were measured.
The observed data trends agreed with previous diabatic measurements in which the center subchannel had the highest quality and mass flux, while the corner subchannel had the lowest.
These data 55
clearly indicate a lateral " void drift" mechanism, and they are the only reliable data on fully developed subchannel distributions in existence.
NUREQ/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITONETER AND ITS APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION. OHKAWA,K.s LAHEY,R.T.
Rensselaer Polytechnic Inst., Troy, NY.
July 1983.
350pp.
8308080201.
19953:001.
This study discusses the development of a non-intrusive gamma-ray scattering measurement technique which was applied to measure the local density (i.e.,
void fraction) in two-phase (air / water flow in eccentric and concentric annular test sections.
The densitometer system consists of a shielded 10 ci(137) Cs photon source, an Na1 photon detector and a co111mation system, which defined the scattering volume within whi~ch the photons were scattered.
A numerical method was developed to allow the efficient and accurate inversion of the integral equation expressing the measured counting rate in terms of the geometric configuration, density and nuclear properties of the scattering volume.
A comparison of the eccentric and concentric density measurements taken in the annular test section indicated lateral void drift phenomena, in which a higher void fraction was seen in the wider part of the annulus than in the narrower part.
NUREC/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS. PARKiG.C.s PODOWSKI, M. s BECKER,M.s et a l.
Rensselaer Polytechnic Inst., Troy, NY.
July 1983.
400pp.
8308080781.
19970:003.
The state-of-the-art one dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR).
The model accounts for phasic slip, distributed spacers, subcooled boiling, space / time-dependent power distributions and distributed heated wall dynamics.
In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation.
In its final form, this model constitutes a multi-input, multi-aut (MIMO) linear system, which features a general nodal neutron kinetics model.
Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions.
The stability characteristics have been investigated with the Nyquist criterion.
The computer. implementation of this model, NUFREG-N, was used to study stability of a typical BWR/4.
Also, comparisons were made with existing in-core and out-of-core data.
It was found that NUFREG-N generally agreed well with out-of-core data.
The stability analysis of a typical BWR/4 revealed the importance of using accurate models of thermal-hydraulic and neutron kinetic phenomera.
Moreover, the accuracy of the nuclear input data is extremely important.
NUREQ/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A BWR/4. FAKORY,M.R.; LAHEY,R.T.
Rensselaer Polytechnic Inst., Troy, NY.
July 1983.
744pp.
'8308150438.
20066:001.
56
A cories of experiments were performed in the parallel channel effects (PCE) test-section at RPI.
This program simulated the short-term-and long-term core cooling of a BWR/4 subsequent to a hypothetical design basis LOCA, for conditions of both intact and bruken Jet pumps.
'Significant parallel channel effects were noted, and it was found that Jet pump breaks can have a detrimental effect on both the short and long-term core cooling. capabilities of a BWR/4.
NUREQ/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
PARK,J.K.s BECKER,M.s PARK,G.C.
Rensselaer Polytechnic Inst., Troy.
N Y.
July 1983.
167pp.
8308050110.
19942:125.
General space kinetics models have been developed for more accurate stability analysis utilizing nodal analysis, a commonly used technique for analyzing power distributions in large power reactors.
Kinetics parameters for use in these kinetics models have been properly derived by utilizing self-consistent nodal data and power distributions.
The procedure employed in the nodal code SIMULATE has been utilized for power distribution, since that methodology is general and includes various commonly used nodal methods as special cases.
Cross sections are correlated as functions of void fraction and exposure.
i A computer program investigating thermo-hydrodynamic stability, NUFREG has been modified to accommodate general spatial kinetics models with an improved thermal-hydraulics model.
The thermal-hydraulic models used are lumped and distributed parameter heated wall dynamics models with or without heat fluz perturbations in the two phase region.
The assumptions used in the thermal-hydraulic models are separated flow, non-uniform heat flux including spacer effects.
The Nyquist criterion has been utilized to obtain stability margins corresponding to the kinetics model.
NUREG/CR-3380: CURRENT METHODOLOGIES FOR ASSESSING SEISMICALLY INDUCED SETTLEMENTS IN SOIL. LEDBETTER,R.
Army, Dept. o f, Army Engineer Waterways Experiment Station.
August 1983.
57pp.
8308230746, 20130:289.
i Earthquake-induced surface settlements have ranged from 0.7 to JO
[
percent of lager thickness for the relatively few incidences where reliable estimates have been made of settlement magnitudes and soil conditions.
Standard penetration test results obtained from pre-earthquake and postearthquake conditions in Japan show that relative densities have changed from 188 percent increase to 44 l
percent decrease.
At present, there are no verified methods of i
seismic settlement analysis.
However, there are current methods of l
analysis ranging frem empirical to fully theoretical, which take into account a few to all of the major variables affecting seismically inductd settlement behavior.
This report reviews pertinent current knowledge and methodologies related to this subject.
NUR EG/CR-3381: EVALUATION OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS. DOUCHERTY, D. s ADAMS,J.W.
Brookhaven National Laboratory.
August 1983.
67pp.
8309300319.
BNL-NUREC-51689.
20202:097.
Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams.
Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, fon-exchange resins (Epicor-II) and 57
r
strippable coatings.
The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of-1'O CFR Part 61.
It appears that much of the Epicor-II ion-exchange resin being disposed of in commercial land burial will be class B.and require stabilization.
Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability.
Results indicated that both radionuclide contamination and chelating agents leach from strippable coating waste.
NOREC/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE. - STEVENS. D. L. s KLECKER, R. W. s SIMONEN,F.A.s et al.
Battelle Memorial Institute, Pacific Northwest Laboratory.
September 1983.
92pp.
8310070317.
PNL-4774.
20679:236.
The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NFO staff evaluation of pressurized thermal shock.
VISA uses Monte Carlo simulation to evaluate the failure probability of a pressur ize d water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user.
Linear elastic fracture mechanics are used to model crack initiation and propagation.
Parameters for initial crack size, copper content, initial RT(NDT), fluence, crack-initiation fracture toughness, and i
arrest fracture toughness are treated as random variables.
This j
report documents the version of VISA used in the NRC staff report (Policy Issue from J.W.
Dircks to NRC Commissioners, " Enclosure A:
NRC Staff Evaluation of Pressurized Thermal Shock, November 1982 "
SECY-82-465) and includes a user's guide for the code.
l NUREC/CR-3335: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
VESELY, W. E. s DAVIS,T.C.s DENNING,R.S.; et al.
Battelle Memorial Institute. Columbus Laboratories.
July 1983.
106pp.
8308150444.
BMI-2103.
20062:253.
The objectives of this work are to evaluate the importance of the containment and the different safety functions as assessed in probabilistic risk analyses.
Tu accomplish this ob Jective, risk importance measures are defined to evaluate a feature's importance in i
.further reducing the risk and its importance in maintaining the l
present risk level.
One defined importance measure, called the f eature 's risk reduction worth, is useful for prioritizing feature improvements which can most reduce the risk.
The other defined l
importance, called the feature's risk achievement worth, is useful for l
prioritizing features which are most important in reliability assurance and maintenance activities.
j NUREC/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL i
SITE HEMATITE, MISSOURI. BOOTH, L. F. s MCDOWELL,C.S.s PECK,S.I.; et al.
Radiation Management Corp.
July 1983.
70pp.
8308010022.
19863:280.
i This report presents the results of a radiological survey of the burial site adjacent to the Comiustion Engineering (C-E) plant in Hematite, Missouri, performed by Radiation Management Corporation l
]
(RMC) in the spring and summer of 1982.
Measurements were made to determine external radiation levels, surface and subsurface radionuclide concentrations and radioactivity in air and water.
[
ss I
Ronults shcw urcnium concontrations in burial pits as high as 38 and 21 pCi/g for U-238 and U-235 respectively.
Results also show uranium concentrations in surface soils as high as 4.7 and 1.1 pC/g for U-238 and U-235 respectively.
Based on an estimated U-234/U-238 activity ratio of about 10 to 1, the highest U-234 activity in the burial pits L
is estimated to be approximately 400 pCi/g, and in surface soils approximately 47 pCi/g.
Radium and thorium concentrations did not exceed background levels.
Radioactivity in water which exceeded EPA drinking watar standards was found in two onsite monitorino wells.
NUR EG/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION METHODOLDGY AND APPLICATION. HSIEH,B.J.s MOT, C. A. s SRINIVASAN,M.G.
Argonne National Laboratory.
August 1983.
108pp.
8308170170.
ANL-83-38.
20085:211.
As part of an assessment of the role of dynamic testing in the design of nuclear power plant (NPP) structures, the state of knowledge concerning system-identification methodologies and applications is evaluated.
In general, it is found that system identification is limited to parameter estimation for a priori chosen models.
The well developed parameter identification techniques assume linear system behavior.
For structural applications the most commonly used method is the modal analysis technique.
Nonlinear parameter estimation and generalized system identification are still in the developmental stage.
In applications to NPP buildings system identification can play a useful role in the verification and improvement of design methods.
Low-level testing of NPP structures is feasible and would suffice to determine many of the parameters of linear design models.
Since testing of NPP structures at design level excitations is not practical, the verification by system identification of design methods in this regime requires an appropriate databank.
To apply s y s t en.
identification methods to NPP structures the accuracy and reliability of these techniques must be assessed.
No systematic attempt to verifying even the current, well developed methods with real data en be documented.
j NUREG/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition. SULLIVAN,P.s SWARTZMAN,G.; BINDMAN,A.
Washington, Univ. o f, Seattle, WA.
July 1983.
193pp.
8308150484.
20044:075.
This notebook contains a detailed comparison of 16 models of fish and zooplankton growth, energetics, population dynamics and feeding.
It is a basic document for the evaluation of these models' usefulness for impact assessment.
Model equations are categorized into 18 l
subprocesses comprising the major processes of consumption, predation, metabolic processes, growth, fecundity and mortality.
The model equations are compared in a standard notation and the equation rationales are considered and put into a historical framework with historical precedence charts.
Model parameters are computed in
' standard units and data sources and techniques used for parameter estimation are identified.
A translator compares standard notation with the notation used in the models.
The major contribution of this work is that these models are arrayed clearly with their assumptions, and sheir parameter values are listed for easy comparison.
This allows clarification of data needs.
Thus the process notebook can be used as a base for further simulation comparison.
59 l
}
.~
NUREG/CR-3394 VO1: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Executive Summary And j
Significant Findings. WYSOCKI,J.J.
Burns & Roe Co.
Sandia Labor. tories.
July 1983.
2OOp p.
8308180726.
SAND 83-7116.
20091:011.
The failure of high energy reactor piping systems within the containment vessel mag lead to the destruction of large quantities of insulation If the resulting insulation debris is transported to the recirculation sump, it can impair the functioning of the emergency l
core cooling system during the recirculation mode.
l This report presents the results of a probabilistic risk assessment undertaken to estimate the probabilities of break i
occurrence, the quantities of fibrous debris which might be generated, and the resulting probabilities of sump screen blockage.
The results f
show that sump screen blockage is a function of pipe break I
probabilities, postulated break locations, pipe diameter, the assumed i
zone (s) of destruction, and allowable screen blockage head loss.
This study was performed to assist in the resolution of " Unresolved Safety Issue" (USI) A-43,
" Containment Emergency Sump Performance."
NUREG/CR-3394 VO2: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Compendium OF Calculated Break Probabilities, Sump Blockage Frequencies,And Calculational Program Listing. WYSOCKI,J.J.
Burns & Roe Co.
July 1983.
600pp.
8308180732.
SAND 83-7116.
20098:001.
The failure of high energy reactor piping systems within the containment vessel may lead to the destruction of large quantities of insulation.
If the resulting insulation debris is transported to the recirculation sump, it can impair the functioning of the emergency core cooling system during the recirculation mode.
This report presents the results of a probabilistic risk assessment undertaken to estimate the probabilities of break occurrence, the quantities of fibrous debris which might be generated, and the resulting probabilities of sump screen blockage.
The results show that sump screen blockage is a function cf pipe break probabilities, postulated break locations, pipe diameter, the assumed zone (s) of destruction, and allowable screen blockage head loss.
This study was performed to assist in the resolution of " Unresolved Safety Issue" (USI) A-43,
" Containment Emergency Sump Performance."
l NUR EC/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM l
THE NBS D(2)D-MODERATED (252)CF SOURCE. SCHWARTZ,R.B.s EISENHAVER C.M.s GRUNDL,J.A.
Commerce, Dept. o f, National Bureau of Standards.
August 1983.
34pp.
8309200582.
20453:119.
Approximately three years ago, two of the authors pointed out that of the great variety of coderated neutron sources used for calibration and testing, a 15 cm radius D(2)D-moderated (252) Cf source came about as close as was practical to producing a spectrum similar to that found inside a typical nuclear power reactor (NUREG/CR-1204).
It was suggested that such a source would he useful p
for calibrating neutron personnel dosimeters.
Much of the rationale for the development and use of this source has been based on characteristics of the neutron spectrum calculated by Ing and Cross.
Prior to this work there has been no accurate experimental l
verification of the calculated spectrum.
In this document NBS reports on'the experimental verification of the spectrum calculated for a (252)Cf neutron source moderated by a D(2)O sphere of 15 cm radius.
Using NBS double fission chambers as 00
l threshold detectors, excellent agreement is noted between NBS measurements and results derived from the calculated spectrum in the energy range below 10 kev.
Although the measurments suggest the existence of small, but significant, discrepancies above 600 keW the value of the fluence-to-dose-equivalent conversion factor derived from NBS measurements agrees with that derived from the calculated spectrum j
(9.0 x 10-(6) arem cm(2)) to within 2% with an estimated uncertainty of plus minus 4%.
NUREQ/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRDN DOSE EQUIVALENT AT NUCLEAR POWER PLANTS. EISENHAVER, C. M. s SCHWARTZ,R.B.
Commerce, Dept.
o f, Nationcl Bureau of Standards.
August 1983.
74pp.
8308230732.
20129:025.
Published data from measurements made by Pacific Northwest Laboratory (PNL) and those made Jointly by the Environmental Measurements Laboratory (EML) and by Rensselaer Polytechnic Institute (RPI) inside containment at nuclear power plants were examined for the purpose of determining the best method for estimating the neutron dose equivalent received by workers.
These data included measurements with TLD albedo dosimeters, 9-inch spherical remmeters, Anderson-Braun remmeters, multisphere sets, " Cutie Pie" gamma survey meters, (3)He spectrometers, and tissue equivalent proportional counters.
It was found that no single technique was clearly most accurate for determining neutron dose equivalent.
The continued use of TLD albedo dosimeters calibrated with a moderated (252)Cf source and corrected for variations in neutron spectrum is recommended.
The continued use of remmeters calibrated with bare or moderated Cf neutron sources and corrected for variations in neutron spectrum determined in the same way as for TLD albedo at each nuclear power plant with a full set of moderating spheres in order to determine neutron spectra.
Using measured gamma exposure rates to infer neutron dose equivalent rates is not a recommended option because these data do not show good correlation.
NUREG/CR-3401: LABDRATORY-SCALE SODIUM-CARBONATE AGGREGATE CONCRETE INTERACTIONS. WESTRICH,H.R.s STDCKMAN,H.W.s SUO-ANTTILA,A.
Sandia Laboratories.
September 1983.
64pp.
8310050253.
SAND 83-1502.
20635:094.
i.
L A series of laboratory-scale experiments was made at 600 degrees centigrade to identify the important heat producing chemical reactions between sodium and carbonate aggregate concretes.
Reactions between sodium and carbonate aggregate were found to be responsible for the bulk of heat production in sodium-concrete tests.
Exothermic reactions were initiated at 500 plus minus 30 degrees centigrade for limestone and dolostone aggregates was well as hydrated limestone concrete, 540 plus minus 10 degrees centigrade for dehydrated limestone concrete, and were ill-defined for dolostone concrete.
Major reactions products included cad, MgO, Na(2)D, NaOH, and elemental carbon.
Sodium hydroxide, which forms when water is released from cement phases, causes slow ero.sion of the concrete with little heat production.
The time-temperature profiles of these experiments have been modeled with a simplified version of the SLAM computer code, which has allowed derivation of chemical reaction rate coefficients.
61
NUREG/CR-3402: ALTERNATVE PROCEDURES FOR.J-R CURVE DETERMINATION.
HISER,A.L.s LOSS,F.J.
Materials Engineering Associates, Inc.
ENSA, Inc.
July 1983.
79pp.
8308080698.
MEA-2022.
19978:092.
This investigation evaluates alternative displacement techniques with the single specimen compliance (SSC) procedure which'do not require the common 1g used load-line displacement measurements in a compact toughness-(CT) specimen to determine the J-R curve.
The two techniques studied are a double clip gage technique involving both crack-mouth 115 placement and load-line displacement at the edges of the specimen and a technique which requires only crack-mouth displacement. -R-curves developed using these techniques are compared to R-curves developed with the normal load-line based deflections.
In addition.rescits of a round robin program using the EC procedure are discussed in terms of validating the double clip gage technique.
R-curves developed from crack-mouth displacements on the J-integral specimen from ASTM E 813 suggest that the standard ASTM E 399 specamen design can be used for both linear-elastic (e.g.,Kic) and elastic plastic (e.g.,
R-curve) toughness assessments.
This fact can be of particular value in the testing of reactor surveillance specimens which may not have been machined to permit load-line displacement measurements.
NUREG/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN
-URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH SYSTEM.Model Formulations And Preliminary Results. PETERSON,S.R.s FELMY,A.R.s SERNE,R.J.; et al.
Battelle Memortal Institute, Pacific Northwest LaboratorJ.
August 1983.
91pp.
8309200503.
PNL-4782.
20451:269.
An equilibrium thermodgnamic conceptual model consisting oF minerals and solid phases was developed to represent a soil column.
A computer program was used as a tool to solve the system of mathematical equations imposed by the conceptual chemical model.
The combined conceptual model and computer program were used to predict aqueous phase compositions of effluent solutions from permeability cells packed with geologic materials and percolated with uranium mill tailings solutions.
Initial calculations of ion speciation and mineral solubility and our understanding of the chemical processes occurring in the model.
The modeling predictions were compared to the analytically determined column effluent concentrations.
Hypotheses were formed, based on modeling predictions and laboratory evaluations, as to the probable mechanisms controlling the migration of selected contaminants.
An assemblage of minerals and'other solid phases could be used to predict the concentrations of several of the macro constituents (e.g.,
Ca, SO(4), A 1, Fe, and Mn) but could not be used to predict trace element concentrations.
These modeling conclusions are applicable to situations where uranium mill tailings solutions of low pH and high total dissolved solids encounter either clay liners or natural geologic materials that contain inherent acid neutralizing capacities.
NUREG/CR-3405 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FDR HIGH-LEVEL WASTE PACKAGING. Annual Report. March 1982 - April 1983.
STAHLiD.;-MILLER,N.E.
Battelle Memorial Institute, Columbus Laboratories.
July 1983.
2OOp p.
8308150491.
BMI-2105.
20064:188.
The Nuclear Regulatory Commission's Office of Nuclear Regulatory Research is developing predictive methodologies to assist in evaluating the Department of Energg's application for construction and operation of deep-mined geologic repositories for high-level 62
i radioactivo unnto dicposal.
As part of the NRC's effort, Batte11e's Columbus Laboratories is investigating the long-term performance of materials used For high-level waste packages.
This report documents
~
the results of our first year's investigations.
The topics addressed are internal and external corrosion and hydrogen embrittlement of the waste container, waste form degradation, accelerated testing and statistics,- and model development both for separate effects and for multiple effects on the integrity of the waste package.
Experimental work has been initiated to address modeling needs, either to fill in missing data, to confirm existing data at prototypic conditions, or to evaluate mechanisms at accelerated test conditions.
A first-generation system code using simple separate-effects
)
correlations is described.
NUREQ/CR-3416 VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
COLEMAN. D. R. s HARRISON,D.L.
Control Data Corp.
September 1983.
120pp.
8309290439.
20584:026.
The acquisition, review, analysis, and processing of power reactor fuel performance data resources is described in this report.
These data resources are characterized here to support subsequent evaluations of the NRC-sponsored fuel rod behavior code, FRAPCON.
Application of the Fuel Performance Data Base is shown to provide the basic data files which are sorted, processed, and restructured to establish key parameters of interest on an individual rod basis.
The design, operational, and performance parameters are analyzed to determine the data populations and the representation of various fuel design types in the data sample.
Also presented are the performance data distribution and trends relative to operational parameters such as power and burnup, and a description of the data processing methods.
Significant amounts of power reactor fuel performance data are available to support high burnup code evaluation studies.
The date clearly indicates the cumulative effects of rod deformation and fission gas release which tend to alter the as-built fuel rod thermal and geometric conditions.
The available data reflect the current status of commercial fuel utilization in that incumbent designs are gradually being replaced by high burnup design, but the newer fuel types do not get dominate the data sample.
NUREG/CR-3416 VO2: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
COLEMAN,D.R.
Control Data Corp.
September 1983.
84pp.
8309290455.
20583:270.
FRAPCON-2 (V1M4) was applied to generate fuel performance predictions for 60 rods of a recently evaluated power reactor data sample.
Rod design, operational, and performance data was obtained from the EPRI Fuel Performance Data Base.
The data was systematically processed to generate code input parameters.
FRAPCON was initially applied for scoping studies to identify the best estimate mechanical response and fission gas release modeling options.
Based on final scoping results, the balance of rods were analyzed with FRACAS-2 mechanics and FASTORASS gas release models.
Comparision between measured and calculated fuel and cladding deformation, fission gas release, internal pressure, and gas composition are presented and interpreted relative to code error magnitudes, distributions, and trends versus rod design and operating parameters.
The results indicate that FRAPCON-2 has best estimate capability for analysis of moderate duty fuel rod performance, 63 l
provided that fabrication parameters are well characterized, and the fu(1 is dimensional 11g stable.
Usa of FASTGRASS and FRACAS-2 options
?esults in better calculational accuracy compared to BEYER-HANN and PELET options.
Finer diagnosis of modeling needs requires improvement of fuel densification, swelling, gas release, clad creep, and irradiation growth correlations in NATPRO.
NUPEC/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2B POWER LOSS TEST SERIES (TEST S-PL-1,-2,AND-3). SACKETT,K.E.s CLEGG,L.B.
EC&G, Inc.
September 1983.
51pp.
8309200533.
EGG-2269.
20456:001.
This report presents test data recorded for Tests S-PL-1,
-2, and
-3 of the Semiscale Mod-2B Power Loss (PL) Test Series.
These tests are the first of a series of Semiscale tests that represent a broad spectrum of offsite ac power loss transients ranging from an offsite power loss with a normal recovery to a station blackout (failure of onsite ac power) with various primary and/or secondary system failures.
The objectives of these tests are to provide long and short term transient data for code assessment and development on events unique to offsite power loss transients: develop a data base on general plant response during loss-of-offsite power transients: and evaluate existing and potential recovery procedures such as feed and hieed for use when the steam generators are not available to remove decay heat.
Test S-PL-1 was a loss-of-offsite power transient with a normal plant recovery using the steam generators to remove decay heat.
Test S-PL-2 was a loss-of-offsite power transient with concurrent failures of emergency ac power and turbine driven auxiliary feedwater.
Plant recovery under real conditions.
Test sequences for Test S-PL-3 were repeats of Test S-PL-2 failures, but with altered initial conaitions for core power, core flow rate, cold and hot leg fluid temperatures, pressurizer liquid level, and secondary mass.
Particular attention was paid to high pressure injection system (HPIS) injection / charging pump flow rates (first use of feed and bleed for recoverg).
NUREG/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FDRCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATIDN LOOP DESIGN. KOCAMUSTAFAOCULs ISHII,M.
Argonne National Laboratory.
August 1983.
76pp.
8309200634.
ANL-83-61.
20456:064.
Scaling criteria for a natural and forced convection circulation loop under single-phase and two-phase flow conditions are derived from the fluid balance equations, boundary conditions, and solid energy equations.
For a single-phase flow case the continuity, intgegral i
momentum, and energy equations in one-dimensional area-averaged forms are used.
For a two-phase flow case the one-dimensional drift-flux model obtained from the short time temporal averaging and the I
sectional area averaging is used.
L The scaling criteria are applied to a conceptual design of a 2 x l
4 loop facility for simulating the Babcock and Wilcox 177 NSSS design.
l Feasibility of practical solutions are studied and some conclusions on
~
this proposed facility in terms of the proper scaling are obtained.
t NUREQ/CR-3423: THE SAFEQUARDS NETWORK ANALYSIS PROCEDURE SNAP: A User's Manual. POLITO,J.
Pritsker & Associates.Inc.
September 1983.
251pp.
8309280455.
SAND 83-7123.
20526:056.
The Safeguards Network Analysis Procedure (BNAP) is an analysis methodology'to evaluate safeguards systems at fixed-sites.
SNAP is a Monte Carlo simulation language specialized for safeguards systems.
64 t
Models of facilities are developed by arranging a special set of symbols which comprise the SNAP language.
Each of the symbols is translated into a data card which is read by an analysis program.
The program performs the required simulations and produces output reports.
NUREG/CR-3425: LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS. NELSON R.F.s CHILTON.A.B.
Illinois, Univ. o f, Urbana-Champaign.
September 1983.
60pp.
8310070323.
20669:245.
Monte Carlo computer codes were employed to calculate depth doses for parallel beams of low-energy, moncenergetic photons incident on slab and spherical phantoms.
Photon energies ranged from 10 to 150 kev.
Phantom compositions include:
ICRU 4-element issue, ICRU 10-element tissue, and a tissue-substitute plastic.
Tabular data is presented for 0.007, 0.3 and 1.0 g/cm 2 depths.
Results are compared with calculations of other workers and with experimental data available for the tissue-substitute plastic.
NUREG/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FDR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
STAHL,D.s MILLER.N.E.
Battelle Memorial Institute, Columbus Laboratories.
August 1983.
53pp.
8309190473.
20417:278.
This quarterly report describes the experimental program for investigating w'aste form degradation.
The mathematical description of waste glass dissolution has been refined.
Groundwater corrosion studies have been nearig completed for titanium and are under way for steel container materials.
Internal glass-stainless steel corrosion tests have been initiated and pitting chemical interactiens are being evaluated.
A first-cut physical description of general corrosion of container materials in contact with groundwater is near completion.
An improved description of fluid flow through a nuclear waste package was developed, along with a method for tracking radioisotopes.
Efforts continue in water-chemistry and accelerated testing.
NUREC/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EQUATIONS OF MULTIPHASE FLOW. SHA,W.T.i CHAO,B.T.s SDO, S. L.
Argonne National Laboratory.
September 1983.
48pp.
8309280595.
20523:269.
Local volume averaging of equations of conservation for a multiphase system vields equations in terms of local volume-averaged products of density, velocity, energy, stress and field forces, together with interface transfer integrals.
Time averaging reduces the volume-and area-averaged products to products of averages plus terms representing eddy diffusivities of mass, Reynolds stresses, and eddy conductivities of heat, etc.,
arising from high-frequency fluctuations.
The interface transfer integrals after time averaging give transfer integrals of low-and high-frequency components.
This procedure of time averaging after local volume averaging, but not in the reverse order, leads to a complete set of differential integral equations of vonservation.
For pure stratified flow, this set of equations can be further reduced to a set'of partial differential equations when the dissipative transfer integrals are presented by experimental correlations.
NUREC/CR-3407: DATA ANALYSIS USING BIONOMIAL FAILURE RATE COMMON CAUSE MODEL. ATWOOD,C.L.
EG&G, Inc.
September 1983.
28pp.
8310050331.
EGC-2271.
20639:069.
65
l This report explains how to use the Binomial Failure Rate (BFR)
{
method to estimate common cause failure rates.
The entire method is
-described, beginning with the conceptual model, and covering practical issues of data preparation, treatment of variation in the failure rates, Bayesian estimation of the quantities of interest, checking the j
model assumptions for lack of fit to the data, and the ultimate application of the answers.
NUREG/CR-3447 VO1: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methods. VESELY,W.E.s SHAFAGHI,A.s GRAY,I.; et al.
Battelle Memorial Institute, Columbus Laboratories.
August 1983.
86pp.
8309200576.
BMI-2106.
20453:157.
The_ goal of this work is to develop a systematic approach to assist in prioritizing research programs and research needs in the Office of Nuclear Regulatory Research within the Nuclear Regulatory Commission (NRC).
There are several approaches which can be applied to prioritize research programs and which satisfy to some degree the requirements desired.
Upon review of the possible suitable 4
approaches, the Analytic Hierarchy process, or AHP for short, developed by T.
L.
Saaty, was selected as being an approach adaptable to the Of fice of Research 's needs and capable of achieving the goals.
The report develops specific guidelines for utilizing the AHP to prioritize the research programs.
Specific examples are given to illustrate the steps in the AHP.
As part of the work, a computer code has been developed and the use of the code is described.
The code allows the prioritizations to be done in a codified and efficient manners sensitivity and parametric studies can also be straightforward 1g performed to gain better understanding of the prioritization results.
NUR EG/CR-3451: BENCHMARK PROBLEMS FOR RADIDLOGICAL ASSESSMENT CODES.
MILLS,M.s VOGT,D.s MANN,B.
Teknekron, Inc.
September 1983.
91pp.
8310130143.
20773:015.
This report describes benchmark problems to test computer codes used in the radiological assessment of high-level waste repositories.
The problems presented in this report will test two types of codes.
The first type of code calculates the time-dependent heat generation i
l and radionuclide inventory associated with a high-level waste package.
Five problems have been specified for this code type.
The second code type addressed-in this report involves the calculation of radionuclide transport and dose-to-man.
For these codes, a comprehensive problem and two subproblems have been designed to test the relevant capabilities of these codes for assessing a high-level waste repository setting.
NUR EG/CR-3464: THE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTIAL THROUGH WALL CRACKS. PARIS P.C.; TADA,H.
Del Research Corp.
EG&G, Inc.
September 1983.
203pp.
8310050335.
20638:223.
This report presents methods of applying tearing instability analysis to large circumferential through wall cracks in piping to demonstrate cases of adequate " ductility" to avoid " double ended break."
The first part gives results of applying simplified rigid plastic cracked section analysis to some typical piping systems and subsequently gives some additional refinements of this type of analysis.
The second part presents full elastic-plastic cracked 66 J
._._..,.m,
_,_._y.
V section analysis which demonstrates the adequacy of rigid plastic analysis for most cases and provides the methodology where elastic plastic analysis is required.
The results herein combined with adequate treatment of material property data and other deformation limiting and load limiting aspects of "as-built" piping systems, provides a complete analysis of sub J e c t.
NUREC/CR-3472 VO1: SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS. HENCH, L. L. s CLARK,D.E.
Florida State Univ.
September 1983.
99pp.
8310050239.
20636:250.
Leaching mechanisms of simulated HLW-borosilicate glasses are studied by various surface spectroscopic measurement and solution chemistry analysis.
The variables included in this annual report are:
degree of glass fracture, glass composition, devitrification, and l
Wet-dry cyclic effect.
NUREG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983. DAEMAN,J.J.K.s STORMANT,J.C.s COLBURN,N.I.s et al.
Arizona, Univ. o f.
Tucson, AZ.
September 1983.
298pp.
8310070414.
20676:006.
This report describes primarily experimental borehole plugging performance assessments completed, started, or planned during June 1982 - May 1983.
Flow testing of granite, basalt, and tuff cylindert.
with cement and bentonite plugs in coaxial holes demonstrates that hole sealing is possible to a quality where neither plug nor plugrock interface form preferential flowpaths.
Cement plugs in basalt and granite have sufficient interface strength to make axial failure improbable.
Flow and strength performance are impaired significantly when plugs dry.
Partial recovery occurs upon resaturation.
Cement plug interface strength is erratic, making extrapolations difficult.
A cement plug has been installed in a borehole running between c
two mine drifts.
Field work, instrumentation and design are completed for in-situ testing of plugs in basalt.
Static testing of cement plugs in granite, reference basis for dynamic testing, is complete.
Dynamic equipment has been assembled.
Studies of drilling damage in the walls of holes in basalt have started.
Equipment for studying the influence of in-situ ambient rock temperatures on plug performance is l
being assembled.
An assessment is given of rock mass features and l
construction procedures that might influence sealing of large size I
escavations.
NUREG/CR-3475: CRITICAL DI? CHARGE OF INITIALLY SUBCOOLED WATER THROUGH SLITS. AMOS,C.N.s SCHROCK,V.E.
Lawrence Berkeley Laboratory.
September 1983.
298pp.
8310070110.
LBL-16363.
20672:001.
This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits.
The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure.
Two new analytical models, which allow for the generation of a metastable liquid phase, are developed.
Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.
A total of 94 experiments are reported.
A parametric study of the effects of stagnation pressure, stagnation subcooling, and length-to-diameter ratio (L/D) of the slit on both the critical mass flux and the pressure profile within the slit was made.
Stagnation pressures range 67 l
n L
l between 4.1 and 16.2 MPa.
Subcoolings were in the range from zero to
)
65 degrees centigrade.
L/D of the slits were obtained 83 and 400, with the flow direction dimension fixed at 6. 35 cm.
NUR EO/CR-3486:
RADIDLDOICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
BOOTH,L.F.s NYERGES,P.L.s OROFF,D.W.s et al.
Radiation Management Corp.
September 1983.
54pp.
8310060079.
- 20664:001.
This report presents the results of a radiological survey of the buraal site on the Nuclear Fuel Services (NFS). Inc.
Erwin Plant grounds.
This site contains approximately 30 burial pits, into which has been placed a varialyoof disposable wastes, contaminated with low levels of thorium and uranium, with enrichments ranging from depleted to 97L Measurements were made to determine if these materials are migrating from the burial pits into the environment, and if so, at what levels.
Survey results indicate that contamination is not moving from the pits, and that previous burials of radioactive waste on the site have had little or no effect on the environment.
I NUREG/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADIDACTIVE WASTE DISPOSAL FACILITY.
Southern States Energy Doord.
September 1983.
123pp.
8310070320.
20679:113.
The attributes of the Southern States Energy Board (SSEB) enable i
it to view federal / state interface problem ares from a perspective that can be uniquely constructive.
The board is sensitive to the interests of both federal and state levels of government since it is composed of. member states with common regional interests and confirmed by federal legislative action.
It has been most effective when exercising a leadership role in finding procedures and practices that use the resources of both levels of government that are mutually supportive and nonduplicative.
SSEB began an NRC-funded-effort in that direction related'to nuclear power p; ant siting in June 1975, entitled " Regional-Interstate Nuclear Facility Siting Procedure Demonstration Project."
SSEB approached the problem by working with interested states to analyze various elements of the licensing process, in particular with NEPA review procedures f or interstate coordination where potential impacts extend beyond a single state and where the facility serves an i
interstate or regional need.
SSEB also served as a catalyst in the development of a region-wide nuclear facility siting procedurc that could improve the effectiveness and timeliness of the regulatory process.
68
Contractor Report Number Index This index lists, in alphabetical order, the contractor-issued report codes for the NRC contractor reports in this compliation. Each contractor code is cross-referenced to the NUREGICR for the report and to the 10-digit NRC Document Control System accession number.
SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUNDER NUMBER ANL-82-51 NUREQ/CR-3238 EGG-2269 NUREG/CR-3419 ANL-83-11 VO1 NUREG/CR-3359 V01 EGG-2271 NUREG/CR-3437 ANL-83-38 NUREG/CR-3388 EPRI NP-2313 NUREG/CR-2575 ANL-83-44 NUREG/CR-3339 EPRI NP-2375 NUREG/CR-2573 ANL-83-61 NUREG/CR-3420 EPRI NP-2376 NUREC/CR-2574 BHARC-400/83/OO NUREQ/CR-3196 EPRI NP-3093 NUREG/CR-3223 BHARC-400/83/01 NUREQ/CR-3215 VO1 GEAP-22051 NUREG/CR-2573 BHARC-400/83/02 NUREQ/CR-3215 VO2 GEAP-22052 NUREG/CR-2574 BMI-2103 NUREG/CR-3365 GEAP-22053 NUREG/CR-2575 BMI-2105 NUREC/CR-3405 V01 GEAP-30157 NUREG/CR-3223 EMI-2106 NUREQ/CR-3447 V01 CP-R-221020 NUREG/CR-3371 VOi BNL-NUREG-51454 NUREC/CR-2331 V02 N4 GP-R-221020 NUREQ/CR-3371 V02 BNL-NUREG-51492 NUREQ/CR-2475 HEDL-TME-82-21 NUREC/CR-2805 V04 BNL-NUREG-51494 NUREQ/CR-2482 V04 HEDL-TME-83-8 NUREG/CR-2146 V02 BNL-NUREG-51630 NUREC/CR-3091 V02 IEB-80-21 NUREG/CR-3048 BNL-NUREG-51645 NUREC/CR-3148 IEB-81-02 NUREG/CR-3050 BNL-NUREG-51654 NUREC/CR-3188 LA-9596-MS NUREG/CR-3355 BNL-NUREG-51658 NUREG/CR-3219 V01 LA-9704-MS NUREG/CR-3280 BNL-NUREG-51664 NUREQ/CR-3241 LA-9708-MS NUREG/CR-3221 BNL-NUREG-51667 NUREG/CR-3266 LA-9749-MS NUREG/CR-3278 BNL-NUREG-51671 NUREQ/CR-3282 LBL-16363 NUREG/CR-3475 BNL-NUREC-51672 NUREQ/CR-3283 NEA-2022 NUREG/CR-3402 BNL-NUREQ-51673 NUREQ/CR-3284 ORNL-5838 NUREG/CR-2506 BNL-NUREG-51678 NUREG/CR-3315 ORNL/NSIC-2OO NUREG/CR-2000 VO2 N6 BNL-NUREQ-51681 NUREG/CR-3341 ORNL/NSIC-200 NUREG/CR-2000 VO2 N7 BNL-NUREG-51689 NUREG/CR-3381 ORNL/NSIC-200 NUREG/CR-2000 V02 N8 EGO-2104 NUREG/CR-2148 ADD 01 ORNL/NSIC-213 NUREG/CR-3122 EGG-2230 NUREQ/CR-3003 VO2 ORNL/TM-8204 NUREG/CR-3138 EGG-2248 NUREQ/CR-3214 ORNL/TM-83OO NUREG/CR-2669 EGG-2258 NUREG/CR-3289 ORNL/TM-8443/V4 NUREG/CR-2874 V04 4
U SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER ORNL/TM-85 5 NUREG/CR-2989 SAND 51-2409 NUREG/CR-2402 ORNL/TM-8649 NUREG/CR-3155 SAND 82-1137 NUREC/CR-2726 ORNL/TM-8661 NUREG/CR-3265 SANDB2-1197 NUREG/CR-2739 ORNL/TM-8741 NUREG/CR-3161 SANDO2-1980 NUREG/CR-3269 ORNL/TM-8781 NUREG/CR-3331 SANDB2-2726 NUREC/CR-2631 ORNL/TM-8787/V1 H?fREQ/CR-3334 VO1 SAND 82-2969 NUREG/CR-3111 VO1 ORNL/TM-8769 NUREC/CR-3333 SANDB2-7205 NUREG/CR-2982 RO1 2
ORNL/TM-8796/V1 NUREG/CR-3200 VO1 SANDB2-7212/2 NUREG/CR-3085 VO2 ORNL/TM-8809 NUREG/CR-3364 SANDB2-7212/3 NUREG/CR-3085 V03 PARAMETER IE-12 NUREG/CR-3048 SANDB2-7212/4 NUREG/CR-3085 V04 PARAMETER IE-12 NUREG/CR-3050 SAND 83-0538 NUREG/CR-3234 PNL-4174 NUREG/CR-2658 SAND 83-0832 NUREG/CR-3257 PNL-4221 NUREG/CR-2850 V02 SAND 83-0833 NUREG/CR-3258 PNL-4294 NUREG/CR-2835 SAND 83-1038 NUREG/CR-3277 PNL-4453 NUREG/CR-2938 SAND 83-1171 NUREG/CR-3329 V01 PNL-4559 NUREG/CR-3083 SAND 83-1502 NUREG/CR-3401 PNL-4566 NUREG/CR-3093 SAND 83-7116 NUREG/CR-3394 VO1 PNL-4641 NUREG/CR-3175 SAND 83-7116 NUREG/CR-3394 V02 PNL-4655 NUREG/CR-3215 V01 SAND 83-7123 NUREG/CR-3423 PNL-4655 NUREG/CR-3215 V02 UCID-19465 NUREG/CR-3017 PNL-4669 NUREG/CR-3272 UCID-19703 NUREG/CR-3270 PNL-4679 NUREG/CR-3196 UCID-19719 NUREG/CR-3345 PNL-4694 NUREG/CR-3259 UCRL-53047 NUREG/CR-3132 PNL-4705-1 NUREG/CR-3307 VO1 UCRL-53948 NUREC/CR-3133 PNL-4774 NUREG/CR-3384 PNL-4782 NUREG/CR-3304 SAIO1583-368LJ NUREC/CR-3264 SAND 79-2247/4 NUREG/CR-1246 V04 SAND 81-1529 NUREC/CR-2238 V04 SAND 81-1655 NUREG/CR-2254 l
l l
70
I l
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.
ADAMS,J.W.
NUREC/CR-3381: EVALUATION OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
ADISOMA,G.S.
NUREC/CR-3473: ROCK NASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
AW, T. M.
NUREC/CR-3282: INTERNAL HYDROGEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
AMICO,P.J.
NUREG/CR-3085 V02: INTEHIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT. Vol 11 - Appendix A And B.1 Thru B.4. Docket No. 50-245.(Northeast Nuclear Energy Compang)
NUREG/CR-308b V03: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT. Vol III Appendix B.5 Thru B.8. Docket No. 50-245.(Northeast Nu: lear Energy)
NUREC/CR-3085 V04: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol IV Appendix B.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
AMOS,C.N.
NUREG/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER THROUGH SLITS.
ANDERSEN,J.G.
NUR EC/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
ANDERSON,P.D.
NUREC/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1983.
ATWOOD,C.L.
NUR EG/CR-3289: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES. Estimates Based on Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, 1976-1981.
NUREG/CR-3437: DATA ANALYSIS USING BIONOMIAL FAILURE RATE COMMON CAUSE MODEL.
BAKER,D.A.
NUREQ/CR-2850 V02: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
BALDWIN,C.A.
NUREC/CR-3333: NEUTRON SPECT7AL CHARACTERIZATION OF THE FOURTH NUCLEAR REGULATORY COMMISSION HEAVY SECTION STEEL TECHNOLOGY IT-CT 71
IRRADIATION EXPERIMENTS: Dosimetry And Uncertainty Analysis.
BALL,S.J.
NUREG/CR-2844 V04: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMDER 31,1982.
BANKOFF,S.G.
NUREG/CR-3060: A CRITICAL REVIEW OF FLOODING LITERATURE.
BANKS.W.W.
NUREG/CR-3003 VO2: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE R AY TUBE-GENERATED DISPLAYS.
BARD,C.S.
NUREG/CR-2506: UNCERTAINTIES IN GEDLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEGUENCES.
BARNES,V.
NUREG/CR-3196: DRUG AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
B ARTTER, W. D.
NUREG/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE 3
PERSONNEL RELIABILITY MODEL.
BATTLE,R.E.
NUR EG/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
BEBAHANI,A.
NUR EG/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
BECKER,M.
NUREG/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
NUREG/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABIL'ITY.
BELL B.J.
NUREG/CR-2254: A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS.
BERLAD,A.L.
NUREG/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS.
B ER NA, 0. A.
NUREG/CR-2148 ADD 01: FRAP-T6:A COMPUTER CODE FOR THE TRANSIENT ANALYSIS OF OXIDE FUEL RODS.
BERNARD,H.
NUREG-0984: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE CORNELL UNIVERSITY TRIGA RESEARCH REACTOR. Docket No.50-157.
BERNATOWICZ,H.
NUR EG/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
BERTUCCA,J.
NUREG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
BINDMAN.A.
NUR EG/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition.
BISHOP.D.J.
(
NUREG/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS l
FROM THE H.B.
ROBINSON STEAM ELECTRIC PLANT.
l BLACKMAN H.S.
NUREG/CR-3003 V02: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
i BLAKELY,K.D.
l NUR EG/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM j
72
-SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
BOGARRA,J.
NURE0/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
500 DAN,E.J.
NURE0/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)-
LEVELS IN CDPflERCIAL NUCLEAR POWER PLANTS.
L
-B0HN,M.P.
NUREO/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO L
DEGRADATION OF STEAM OENERATOR AND REACTOR COOLANT PUMP h
SUPPORTS. Seismic Safety Margins Research Program.
BOOTH,L.F.
NUPEG/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL SITE HEMATITE, MISSOURI.
NUREO/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
BORGONOVI,G.
NUREO/CR-3264: PARAMETRIC EFFECTS ON NONI,ESTRUCTIVE ASSAY TECHNIGUES.
- BOYACK, B. E.
{
NUREQ/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
BROCARD,D.N.
NUREQ/CR-2982 RO1: BUOYANCY, TRANSPORT,AND HEAD LOSS OF FIBROUS REACTOR INSULATION.
BRONSON,F.L.
NUREQ/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL SITE HEMATITE, MISSOURI.
BROWNSON,F.L.
NUREQ/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
BURQY,D.
NUREG/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS.-Volume 1: Project Approach And Methodology.
NUREG/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
CAIN,G.R.
NUREG/CR-3362: THE INFLUENCE OF LYMPH 0 HEMATOPOIETIC DISEASE IN ALTERING RADIATION INDUCED CANCER SENSITIVITY IN MICE.
CAMP,A.L.
NUREC/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
I CAPFBELL,D.J.
NUREG/CR-2989: RELIABILITY OF EMERCENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
CAMPBELL,J.E.
NUREQ/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTEG IN GEOLOGIC REPOSITORIES.
CAPLAN,J.S.
i NUREQ/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
i CARTWRIGHT,K.
NURE0/CR-2478 VO2: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
l CASTLE,J.N.
l NUR EG/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1983.
CATTON, I.'
NUREQ/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1983.
CHAN,M.K.W.
NUREQ/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A REVIEW OF THE LITERATURE.
73
-w-w
-_y
U CHANG,F.K.
NUREG/CR-3327: ANALYSIS OF STRONG-MOTION DATA FROM THE NEW HAMPSHIRE EARTHOUAKE OF 18 JANUARY 1982.
CHANG,T.Y.
NUREG-1018: SEISMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS.
Status Report, Unresolved Safety Issue A-46.
NUR EG/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS CHAO,B.T.
NUREG/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EGUATIONS OF MULTIPHASE FLOW.
CHENG,L.Y.
NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION, SONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
CHEUNG,Y.K.
NUREG/CR-2574; BWR REFILL-REFLOOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT
. TRAC-BWR COMPONENT MODELS.
C EVERTON, R. D.
NUREG/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING ON THE INNER SURFACE.
CHILTON,A.B.
NUREG/CR-3425: LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS.
CHITTY,D.E.
NUREG/CR-31CO: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
CHU,K.H.
NUREC/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
CHU,M.S.Y.
NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
NUREG/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
CLAPP,N.E.
i NUREG/CR-2874 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION CF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
CLARK,D.E.
NUREG/CR-3472 VO1: SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS.
CLEGG,L.B.
NUREG/CR-3419: EXPERIMENT DATA REPGRT FOR SEMISCALE MOD-2B POWER LOSS TEST SERIES (TEST S-PL-1,-2,AND-3).
CLEVELAND,J.C.
NUREG/CR-2844 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES l
FOR THE DIVISION OF ACCIDENT EVALUATION, QUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
COHEN,L.
NUREG-0837 VO2 NO4: NRC TLD DIRECT RADIATION MONITORING
-NETWORK. Progress Report, October - December 1982.
COLBURN,N.I.
NUREG/CR-3473: ROCK MASS SEALING-EXPERIPENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1903.
COLBURN,T.G.
NUREG-1011: DPAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERAT(Nt REPAIR AT THE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
~50-266.(Wisconsin Electric Power Company)
NUREG-1011: FINAL ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR REPAIR AT THE POINT BE/;CH NUCLEAR PLANT, UNIT 1. Doc ket No.
74 :
~ --
50-266.(Wisconsin Electric Power Company)
COLEMAN,D.R.
NURE0/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
NURE0/CR-3416 V02: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
i COMBS,S.K.
NURE0/CR-3138: MEASUREMENT OF TWD-PHASE FLOW AT THE CORE / UPPER PLEMJM INTERFACE FOR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIDNS.
I CONKLIN,J.C.
i NUREG/CR-2874 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, l
OCTOBER 1 - DECEMBER 31,1982.
COpfdOR, P. E.
l NURE0/CR-3215 V01: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES i
WITH NUCLEAR POWER PLANTS.An Organizational Overview.
NURE0/CR-3213 V02: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
i COSTANTINO,C.
NURE0/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN DPERATING NUCLEAR POWER PLANTS.
COSTELLO,F.
NUREG-0837 V02 N04: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, October - December 1982.
COVER,L.E.
NURE0/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO i
DEGRADATION OF STEAM OENERATOR AND REACTOR COOLANT PUMP l
SUPPORTS. Seismic Safety Margins Research Program.
CRANWELL,R.M.
NUREQ/CR-2739: DATA BASE FOR BASALT METHODOLOGY.
CUMMINGS,J.C.
NUREQ/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
CURRERI,J.
NURE0/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
DAEMAN,J.J.K.
NUREO/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE l
PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
l DAVIS,M.S.
NUREC/CR-3219 VO1: DRAFT TECHNICAL POSITION SUBTASK 1.1: WASTE PACKAGE PERFORMANCE AFTER REPOSITORY CLOSURE.
DAVIS T.C.
NURE0/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIP APPLICATIONS.
DE JARLAIS,G.
NURE0/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC SIMULATION.
DEEDS.W.E.
NUREG/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM OENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
NURE0/CH-32OO V01: EDDY-CURRENT INSPECTION FOR STEAM OENERATOR TUBING PRDORAM GUARTERLY' PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1993.
DENNING,R.S.
NURE0/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
DICKSON,J.H.
NUREG-0998:
U. S.
NUCLEAR REQULATORY COMMISSION 1982 ANNUAL REPORT.
DISCHLER,S.A.
NUREO/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BDREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
75
DOOD,C.V.
NUREG/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
NUREG/CR-3200 V01: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
DONG,R.Q.
t NUREG/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAR i
FACILITIES WITH SEISMIC EQUIPMENT QUALIFICATION IN NUCLEAR PLANTS (A-46).
DOUGHERTY,D.
NUREG/CR-3381: EVALUATION OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
DREW,D.A.
NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION, SONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
EDLZR,S.K.
4 l
NUREG/CR-3307 V01: REACTOR SAFETY RESEARCH PROGRAMS. Quarterly Report, January -March 1983.
EIBENHAVER,C.M.
NUREG/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
NUREG/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOS! METERS AND PORTABLE INSTRUMENTS Fort DETERMINING NEUTRON DOSE EGUIVALENT AT i
NUCLEAR POWER PLANTS.
ELLINGSWOOD,R.
NUREG/CR-3341: PROBABILISTIC DESCRIPTIONS OF RESISTANCE OF SAFETY-RELATED NUCLEAR STRUCTURES.
ENGEL,D.W.
NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
I EPPERSON,D.
NUREG/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
FAKOR Y, M. R.
NUREG/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A BWR/4.
FELMY,A.R.
NUREG/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH SYSTEM.Model Formulations And Preliminary Results.
FIELDS,S.R.
NUREG/CR-2146 V02: DYNAMIC ANALYSIS TO ESTABLISH NORMAL SHOCK AND VIBRATION OF RADIDACTIVE MATERIAL SHIPPING PACKAGES: Guarterly Progress Report. April 1,1981 - June 30,1981.
FOLEY,W.J.
NUREG/CR-3048: CLOSE0VT OF IE BULLETIN B0-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
NUREG/CR-3050: CLCSEQUT OF IE BULLETIN 81-02 AND SUPPLEMENT: Failure Of Gate-Type Valves To Close Against Differential Pressure.
FRESHLEY NUREG/CR-3272: LOCA SIMULATION IN NRU FROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
FUENKAJORN,K.
NUREG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
GALLUP,D.R.
NUREG/CR-2631: A LGGICAL FRAMEWORK FOR IDENTIFYING EGUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
GEE,G.W NUREG/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH I
76 1
SYSTEM.Model Formulations And Preliminary Results.
GERTMAN,D.I.
NURE0/CR-3003 V02: HUMAN ENGINEERING DESIGN CONSIDE3ATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
OMi:RSON, P.
NURE0/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF BEVERE DYNAMICS.
GIDA,R.G.
NURE0/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
GILMORE,W.E.
NURE0/CR-3003 V02: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DIS
- LAYS.
OLANCY,J.
NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
OLISSMEYER,J.A.
NURE0/CR-2835: REVIEW OF MODELS APPLICABLE TO ACCIDENT AEROSOLS.
GOLDMAN,A.S.
NURE0/CR-3221: MC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION.
00ZANI,T.
NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
6RADY,L.M.
NURE0/CR-1246 V04: SAFE USERS MANUAL. Volume 4: Computer Programs.
GRAY,I.
{
NURE0/CR-3447 V01: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methads.
GREER,W.B.
NURE0/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
GROFF,D.W.
NURE0/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL.
SITE HEMATITE, MISSOURI.
2 NURE0/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
GRUNDL,J.A.
NURE0/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
GUTKNECHT,P.J.
NURE0/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
GUTHACHER,R.G.
NURE0/CR-3221: MC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION.
GUZOWSKI,R.V.
NURE0/CR-2739: DATA BASE FOR BASALT METHODOLOGY.
HAAS,P.M.
NURE0/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PL ANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
HAGAN,W.
NURE0/CR-3264: PARAMETRIC EFFECTS ON NONTESTRUCTIVE ASSAY TECHNIGUES.
HARLAN,R.
NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIQUES.
HARRINGTON,R.M.
NUREQ/CR-2844 V04: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES FCR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
HARRIis0N, D. L.
i NURE0/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
HARRISON,F.L.
77
4 NURE0/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM THE H.B.
ROBINSON STEAM ELECTRIC PLANT.
NURE0/CR-3133: THE SENSITIVITY OF ADULT, EMBRYONIC AND LARVAL CRAYFISH q
PROCAMBARUS CLARKII TO COPPER.
MALY, R. J.
NURE0/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
HEBDON,F.J.
NUREG-1022: LICENSEE EVENT REPORT SYSTEM. Description 0F System Ared Guidelinks For Reporting.
i HENCH,L.L.
NURE0/CR-3472 Vol: SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS.
i HENNICK,A.
NURE0/CR-3048: CLOSEOUT OF IE BULLETIN 90-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
NUREO/CR-3050: CLOSEQUT OF IE BULLETIN 81-02 AND SUPPLEMENT: Failure Of Gate-Type Valves To Close Against Differential Pressure.
i HERZOG,B.L.
NURE0/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT l
WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
ESSON,G.M.
1 NUREO/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
HISER,A.L.
NURE0/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
HOBSON,D.D.
NURE0/CR-3265: EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS.
H0 HORST,J.K.
NUREO/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS
+
OF OXIDE FUEL RODS.
HOLTER,G.M.
NURE0/CR-0130 ADD 02: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION.Effacts On Decommissioning Of Interim Inability To Dispose Of Wastes Offsite.
l NURE0/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR POWER STATION. Effects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
HOOPER,R.L.
NURE0/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY GOALS FOR NUCLEAR POWER PLANTS.
HOPE,A.M.
NURE0/CR-3196: DRUS AND ALCOHOL ABUSE:THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
HOPE,E.
NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
HORTON,W.
NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
HOWAR D, G. E.
NURE0/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
l H0Y, H. C.
NURE0/CR-3122: POTENTIAL DAMAGING FAILURE MODES OF HIGH-AND MEDIUM-VOLTAGE ELECTRICAL EGUIPMENT.
i
.HSIEH,B.J.
NURE0/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION METHODOLD0Y AND APPLICATION.
HSIEH,P.A.
NURE0/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY 78
-. - =..
l EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
HU,K.
NUREG/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF SEVER DYNAMICS.
HUDSON,H.G.
NUREG/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMPERCIAL NUCLEAR POWER PLANTS.
HWANG,H.
NUREG/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT GTRUCTURAL LDADS.
ISHII,M.
NUREG/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESION.
ISKANDER,S.K.
NUREG/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING OH THE INNER SURFACE.
IYER,K.
NUREG/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF SEVER DYNAMICS.
JACKSON,M.S.
NUREG/CR-3215 VO1: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
NUREC/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
JAMALI,K.M.A.
i NUREG/CR-3241: PROBABILISTIC METHODS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SEGUENCES IN LOOP-TYPE LMFBRS.
JO, J. H.
NUREG/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CO AT BNL IN FY 1981.
JOHNSON,K.I.
NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBADILITY OF REACTOR PRESSURE VESSEL FAILURE.
i i
JOHNSON,T.M.
NUREC/CR-2478 VO2: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION A WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
1 JOHNSTON,B.A.
NUREG/CR-31SO: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FR SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
i KAKO, T.
NUREG/CR-3283: FIRST EXCURSION PROBLEMS FOR CAUSSIAN VECTOR PROCESSE L
KAM,F.B.K.
NUREG/CR-3333: NEUTRON SPECTRAL CHARACTERIZATION OF THE FOURTH NUCLE REGULATORY COMMISSION HEAVY SECTION STEEL TECHNOLOGY 1T-CT IRRADIATION EXPERIMENTS: Dosimetry And Uncertainty Analysis.
KEHLER,P.
NUREG/CR-3238: PNABIM: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIQUE OF MASS FLOW MEASUREMENT.
KENNEDY,J.K.
NUREC/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
KERWIN,N.
NUREC/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
KISNER,R.A.
NUREC/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT' CON
' FUNCTIONS TO HUMAN OR AUTOMATIC CONTHOL.
KLECKER,R.W.
79
i NUREQ/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
KLEIN,S.J.
NUREG/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And l
Computer Modeling of Trench Cover Design.
MMETYK,L.N.
NUREG/CR-3257: RELAP5 ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
NUREG/CR-3258: RELAP5 ASSESSMENT: SEMISCALE NATURAL CIRCULATION TESTS S-NC-2 AND S-NC-7.
KlEE, H. E.
NUREG/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
KNEZOVICH,J.P.
NUREG/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM THE H. B.
ROBINSON STEAM ELECTRIC' PLANT.
KOCAMUSTAFAOGUL 4
NUREC/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURA'. AND FORCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESIGN.
KOCHER,D.C.
NUREG/CR-2506: UNCERTAINTIES IN GEOLOMJ DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEGUENCES.
KDESTEL,A.
NUREG/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
MONZEK,G.J.
NUREG/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Rediation Exposure And Costs To Selected Parameters.
KOT, C. A.
NUREG/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION METHODOLOGY AND j
APPLICATION.
KDUSARI,B.
NUREG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
KUPIEC,C.F.
NUREG/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
LAHEY,R.T.
NUREG/CP.-3372: AN ANALYSIS OF WAVE DISPERSION, SONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
NUREG/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
NUREG/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITOMETER AND ITS APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
j NUREG/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
NUREG/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A BWR/4.
LARBON,T.H.
NUREG/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
LEDBETTER,R.
NUREC/CR-33SO: CURRENT METHODOLOGIES FOR ASSESSING SEISMICALLY INDUCED SETTLEMENTS IN BOIL.
LEE,S.C.
NUR EG/CR-3060: A CRITICAL REVIEW OF FLOODING LITERATURE.
80 i
LEMPCES,C.
'NUREG/CR-3371 V01: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROO CREWS.
Volume 1: Project Approach And Methodology.
NURE0/CR-3371 V02:
TASK ANALYSIS OF NUCLEAR POWER PLANT CONT.OL ROOM CREWS. Volume 2: Data Results.
LINDEMER,T.B.
NURE0/CR-2844 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUD FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
LINK,B.W.
NURE0/CR-3336:
EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTE. Summary Report.Ames Laboratory Research Reactor. Docket No.
4 50-116.
NURE9/CR-3370: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS SUMMA REPORT NORTH CAROLINA STATE UNIVERSITY RESEARCH AND TRAININ2 RE LIPPINCOTT,E.P.
NUREG/CR-2805 V04: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM. Guarterly Progress Report,0ctober 1982 - December 1982.
LOMMERS,L.
NUREC/CR-3306: LWR AND HTOR COOLANT DYNAMICS:1HE CONTAINMENT OF DYNAMICS.
LOSS,F.J.
NUREC/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
MAf#i, B.
NUREQ/CR-3451:
BENCHMARK PROBLEMS FOR RADIOLOGICAL dSSESSMENT CODES.
MARSH,G.M.
NURE0/CR-3118 V01:
HEALTH EFFECTS OF OCCUPr".ONAL EXPOSURE TO HAZARDOUS i
CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING R ADIATION Volume 1: Executive Summary.
NUREG/CR-3118 V02: HEALTH EFFECTS OF GCCUPATIONAL EXPOSURE TO H
}
CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING i
RADIATION. Volume 2: Supporting Documentation And Review Of Epidemiologic Studies Of Specific Chemicals.
MCCLUNG,R.W.
NUREQ/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING j
- PROGRAM, ANNUAL PROGRESS REPORT FON PERIOD ENDING DECEMBER 31,1983.
MCDANIEL,T.
NUREQ/CR-3264:
PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
MCDOWELL,C.S.
NUREC/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL SITE HEMATITE, MISSOURI.
MCELROY, W. N.
NUREQ/CR-2805 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROG ~ TAM.Guarterly Progress Report,0ctober 1982 - December l
1982.
MCOLAUN,J.M.
NUREC/CR-3258: RELAP5 ASSESSMENT: SEMISCALE NATURAL CIRCULATION TESTS S-NC-2 AND S-NC-7.
MCOUIRE,M.V.
NUREQ/CR-3196: DRUG AND ALCOHOL ABUSE: THE BAEES FOR EMPLOYEE ASSISTANC PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
MCLAUGHLIN,S.D.
NUREQ/CR-3215 VO1:
ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
NUREC/CR-3215 VO2:
ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
MCLUNG,R.W.
NUREC/CR-32OO V01:
EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
MEACHUM,T.R.
81
l NUREG/CR-3299: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL A~tSEMBLIES. Estimates Based on Licensee Event Reports At U.S.
l Commercial Nuclear Power Plants, 1976-1981.
l MILES,D.E.
NUR EG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
MILLER,A.
i NUREG/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL R00M CREWS. Volume 1: Project Approach And Methodology.
NUREG/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
MILLER,D.W.
NUREG/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON l
SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
MILLER,N.E.
NUREC/CR-3405 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR l
HIGH-LEVEL WASTE PACKAGING. Annual Report, March 1992 - April 1983.
NUREG/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
MILLER,R.L.
NUREG/CR-3336: EVALUATION CF NUCLEAR DECOMMISSIONING PROJECTS. Summary Report,Ames Laboratory Research Reactor. Docket No. 50-116.
NUREG/CR-3370: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS
SUMMARY
REPORT NORTH CAROLINA STATE UNIVERSITY RESEARCH AND TRAINING REACTOR.
MILLS,M.
NUREC/CR-3451: BENCHMARK PROBLEMS FOR RADIOLOGICAL ASSESSMENT CODES.
i MISHIMA.J.
I NUREG/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A REVIEW OF THE LITERATURE.
i MITCHELL,D.H.
NUREG/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
MOHR,C.L.
NUREG/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
MORRIS,F.A.
NUREG/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY GOALS FOR NUCLEAR POWER PLANTS.
NOTLEY,F.
NUREC/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 38).
i MURPHY,E.S.
NUREG/CR-0130 ADDO2: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIGNING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Effects On Decommissioning Of Interim Inability To Disposa Of Westes Offsite.
NUREG/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR POWER GTATION. Effects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
MYERS D.A.
L NUREG/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
i MYERS,L.L.
NUREG/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
NADEL,M.V.
NUREG/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
NALEZNY,G.L.
NUREG/CR-3214:
SUMMARY
OF NUCLEAR REGULATORY COMMISSION'S LOFT PROGRAM i
l EXPERIMENTS.
I NANSTAD,R.K.
82
NUREO/CR-3265: EFFECTS OF DFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS.
ELSON,R.F.
NUREG/CR-3425: LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS.
EUMAN,S.P.
l NUREG/CR-3213: PRESSURE TESTING OF FRACTURCO ROCKS - A METHODOLOGY t
EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
NEYMOTIN,L.
NUREG/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1991.
NOURBAKHSH,H.P.
NURE9/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF SEVERE l
DYNAMICS.
NYERGES,P.L.
NUREO/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
OMAWA,K.
NUREG/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITOMETER AND ITS APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
OLSON,J.
NUREG/CR-3215 VO1: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
NUREG/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
ORLOV,9.
NURE0/CR-3264: PARAMETRIC EFFECTS DN NONDESTRUCTIVE ASSAY TECHNIQUES.
ORTIZ,N.R.
NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
NUREC/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
OSBORN, R. N.
NUREC/CR-3215 Vol: DRGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
2 NUREQ/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR PDWER PLANT. Perspectives For Organizational Assessment.
PARAMESWARAN,V.
NUREQ/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT TRAC-BWR COMPONENT MODELS.
PARAMORE,B.
I NUREG/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NUREC/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL RDOM CREWS. Volume 2: Data Results.
PARCHEN,L.J.
NUREG/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
PARIS,P.C.
NUREG/CR-3464: THE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTIAL THROUGH WALL CRACKS.
PARK,O.C.
NUREQ/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
NUR EG/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
PARK,J.K.
NUR EQ/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
83
PECK,S.I.
NUREQ/CR-3387: RADIDLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL SITE HEMATITE, MISSOURI.
NUREQ/CR-3406: RADIDLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
PELDGUIN,R.A.
NUREQ/CR-2050 VO2: POPULATION DOSE CDMMITNENTS DUE TO RADIDACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
PEPPING,R.E.
NUREQ/CR-2402: RISK ANALYSIS METHODOLDQY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
NUREQ/CR-3111 VO1: AN ASSESSMENT DF THE PROPOSED RULE (10CFR60) FOR i
DISPOSAL DF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
j PETERSON,S.R.
NUREQ/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLDW-THROUGH SYSTEM.Modct Formulations And Preliminary Results.
PODOWSKI,M.
NUREQ/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
4 POLITO,J.
j NUREQ/CR-3423: THE SAFEQUARDS NETWORK ANALYSIS PRDCEDURE SNAP: A User's Manual.
PRICE,H.E.
NUREQ/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTRDL.
PUGH C.E.
NUREQ/CR-3334 V01: HEAVY SECTION STEEL TECHNOLDQY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983.
PULLIAM,R.
NUR EQ/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTRDL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
RADFORD.L.R.
NUREG/CR-3196: DRUC AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
RANKIN,W.L.
NUREQ/CR-3196: DRUQ AND ALCDHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PRDORAMS IN THE NUCLEAR UTILITY INDUSTRY.
RASNUSON,D.M.
NUREQ/CR-3447 V01: RESEARCH PRIORITIZATIDN USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methods.
RAZVI,J.
NUREQ/CR-3264: PARAMETRIC EFFECTS DN NONDESTRUCTIVE ASSAY TECHNIGUES.
REICH,M.
NUREQ/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED l
ELECTRICAL AND MECHANICAL EGUIPMENT IN OPERATING NUCLEAR POWER l
PLANTS.
NUREQ/CR-3284: REVIEW DF C"ARENT ANALYSIS METHODOLOGY FOR REINFORCED i
CONCRETE STRUCTURAL EVALUATIONS NUREQ/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTURAL LOADS.
RICE D.W.
NUREQ/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FRDM THE H.B.
ROBINSON STEAM ELECTRIC PLANT.
i NUREQ/CR-3133: THE SENSITIVITY DF ADULT,EMBRYDNIC AND LARVAL CRAYFISH l
PROCAMBARUS CLARKII TO COPPER.
l ROHATGI,U.S.
NUREQ/CR-3148: INDEPENDENT-ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1981.
j 84
RUSSCHER,O.E.
NUREQ/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
NUREC/CR-3272: LDCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
l SACKETT,K.E.
l NUREC/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2B POWER LOSS TEST SERIES (TEST S-PL-1,-2,AND-3).
SAHA, P.
NUREQ/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1981.
SALLACH,R.A.
NUREQ/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH,CD32-)
RECIPROCAL SYSTEM l
SALTOS,N.
I NUREQ/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
SANDHOP,J.R.
NUREQ/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
SAUNDERS,J.H.
NUREQ/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
SAUTER,A.
NUREQ/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING ON THE INNER SURFACE.
SAWYER,C.R.
NUREQ/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
SCHROCK,V.E.
NUREQ/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER THROUGH SLITS.
SCHROEDER,L.
NUREQ/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NUREQ/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
SCHWARTZ,R.B.
NUREQ/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
NUREQ/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EQUIVALENT AT NUCLEAR POWER PLANTS.
SCHWEIT2ER,D.E.
NUREQ/CR-3219 VO1: DRAFT TECHNICAL POSITION SUBTASK 1.1: WASTE PACKAGE PERFORMANCE AFTER REPOSITORY CLOSURE.
SCOTT,W.G.
NUREQ/CR-3215 VO1: ORGANIZATIONAL ANALYSIS AND SAFETY FDR UTILITIES WITH NUCLEAR PCWER PLANTS.An Organizational Overview.
NUREQ/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives for Organizational Assessment.
SEBRELL,W.
NUR EQ/CR-3234: THE POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH ELECTRICAL PENETRATION ASSEMBLIES UNDER SEVERE ACCIDENT CONDITIONS.
SERNE,R.J.
NUREQ/CR-2938: TAILINGS TREATMENT TECHN!GUES FOR URANIUM MILL WASTE: A Review Of Existing Information.
NUREQ/CR-3404: PREDICTIVE GEDCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH SYSTEM.Model Formulations And Preliminary Results.
SHA,W.T.
NUREQ/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EQUATIONS OF MULTIPHASE FLOW.
85
l
.HAFAGHI A.
NUREG/CR-3447 VO1: RESEARCH PRIDRITIZATION USING THE ANALYTIC HIERARCHY
. PROCESS. Volume 1: Basic Methods.
SHAH,V.N.
NUREG/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS OF OXIDE FUEL RODS.
SHARMA.S.
NUREG/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS SHAUG,J.C.
NUREG/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
NUREC/CR-2574:.BWR REFILL-REFLDOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT TRAC-BWR COMPONENT MODELS.
SHERMAN,M.P.
l NUREG/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
SHERWOOD.D.R.
i NUREC/CR-2938: TAILINGS TREATMENT TECHNIGUES FOR URANIUM MILL WASTE: A Review Of Existing Information.
SHIEH,L.C.
NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION DF STEAM QENERATOR AND REACTOR COOLANT PUMP l
SUPPORTS. Seismic Safety Margins Research Program.
SHINOZUKA M.
l NUREG/CR-3283: FIRST EXCURSION PROBLEMS FOR QAUSSIAN VECTOR PROCESSES.
I SHOOMAN,M.
NUREG/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR PDWER PLANT STRUCTURAL LDADS.
SIBULKIN,M.
NUREC/CR-7475: HYDRDGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS.
SIEFKEN.L.J.
i NUREG/CR-2148 ADD 01: FRAP-T6:A COMPUTER CODE FOR THE TRANSIENT ANALYSIB l
OF OXIDE FUEL RODS.
l SIEGEL A.I.
NUREG/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE
[
PERSONNEL RELIABILITY MODEL.
SIMAN-TOV,I.
NUREG/CR-2844 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1983.
SIMONEN.F.A.
NUR EC/CR-3304: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
SIMPSON,E.S.
NUREG/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
SJOREEN,A.L.
NUREG/CR-2506: UNCERTAINTIES IN GEDLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL C ONSEGUENCES.
PLOVIK,C.
NUREC/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1981.
SMITH,J.H.
NUREC/CR-32DO VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING l
PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
l SMITH.P.D.
NUREG/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAR FACILITIES WITH SEISMIC EQUIPMENT GUALIFICATIDN IN NUCLEAR PLANTS L
86
(A-46).
SOMER S. W. M.
NUREC/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BUR SITE HEMATITE MISSOURI.
SOMMERS,P.E.
NUREC/CR-3215 VO1: DRGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
NUREC/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For OrDanizational Assessment.
l SOO, P.
NUREG/CR-2482 VO4: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1982 - March 1983.
NUREC/CR-3091 VO2: REVIEW OF WASTE PACKAGE VERIFICATION TESTS. Semiannual Report Covering The Period October 1982 - March 1983.
NUREG/CR-3282: INTERNAL HYDROCEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
SDO,S.L.
NUR EG/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EQUATIONS OF MULTIPHASE FLOW.
SOUTH,D.L.
NUR EC/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOL PLUG PERFORMANCE. Annual Report. June 1982 - May 1983.
SRINIVASAN M.C.
NUREC/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION NETH3DOLOGY AND APPLICATION.
STAHL.D.
NUREG/CR-3405 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report March 1982 - April 1983.
NUREG/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
STALLMAN,F.W.
NUR EG/CR-3333: NEUTRON SPECTRAL CHARACTERIZATION OF THE FOURTH NUCL REGULATORY COMMISSION HEAVY SECTION STEEL TECHNOLOGY 1T-CT i
IRRADIATION EXPERIMENTS: Dosimetry And Uncertainty Analysis.
STERNER R.W.
NUR EG/CR-3373: AIR / WATER SUBCHANNEL NEASUREMENTS OF THE EQUILIBRIUM GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
STEVENS.D.L.
NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
STIRPE,D.
NUREQ/CR-3221: HC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALVATION.
STOCKMAN.H.W.
NUREG/CR-3401: LABORATORY-SCALE SODIUM-CARBONATE ACCREGATE CONCRETE INTERACTIONS.
STOHR.C.
NUREG/CR-2478 VO2: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
STORMANT,J.C.
NUREQ/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June 1982 - May 1983.
STREIT,R.L.
NUR EG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS. Seismic Safety Margins Research Program.
STROSNIDER,J.
NUR EG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBADILITY OF 87
=
REACTOR PRESSURE VESSEL FAILURE.
SUBUDHI,M.
NURE9/CR-3266: SEISMIC 8: DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
SUGARMAN,A.
NUREQ/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
SULLIVAN', P.
NURE9/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATIDN.Second Edition.
SUMMERS,R.M.
NURE0/CR-3277: RELAP5 ASSESSMENT: SEMISCALE MOD-3 SMALL BREAK TEST.
SUO-ANTTILA,A.
. NUR EG/CR-3401: LABORATDRY-SCALE SODIUM-CARBONATE AGOREGATE CONCRETE INTERACTIONS.
SUTTER,S.L.
NURE0/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC AIR.
SWAIN,A.D.
NURE0/CR-2254:.A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS.
SWANSON,D.G.
NUREC/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1983.
SWARTZMAN,9.
NUREC/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition.
TADA,H.
NUREC/CR-3464: THE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTIAL THROUGH WALL CRACKS.
TALNAGI,J.W.
NUREG/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTIDft SYSTEMS.
THEOFANOUS,' T. G.
NUREQ/CR-3306: LWR AND HTOR COOLANT DYNAMICS:THE CONTAINMENT OF SEVERE 3
DYNAMICS.
THOMAS,D.9.
NUREG/CR-3138: MEAIPUREM8ENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FDR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
THOMPSON, J. E.
NUREG/CR-2575: BWR FULL INTEGRAL S!t1ULATION TEST (FIST) PROGRAM TEST PLAN.
THOMPSON, S. L.
NUREG/CR-3257: RELAP5 ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
NUREO/CR-3329 V01: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
THOMPSON,T.
NUREG-OS37 VO2 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, October - December 1982.
TSURI,A.
NURE0/CR-3283: FIRST EXCURSION PROBLEMS FOR GAUSSIAN VECTOR PROCESSES.
TYLER,S.W.
NURE0/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
TYNDALL,R.L.
NURE0/CR-3364: PRESENCE OF PATHOGENIC MICRODRGANISMS IN POWER PLANT COOLING WATERS.
VAN COTT,H.
NURE0/CR-3371 V01: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM 88
CREWS. Voluso 1:ProJoct Approach And Methodology.
NUREC/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
VAN TUYLE,G.J.
NUREC/CR-31BG: THE STEADY STATE POROUS BODY ASSEMBLY CODE (SPAC): Users 1
l Manual.
VANNONI,M.G.
NUREC/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EQUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
VESELY,W.E.
NUREC/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
NUREC/CR-3447 VO1: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methods.
VISKANTA,R.
NUREC/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF SEVER DYNAMICS.
VOCT, D.
NUREC/CR-3451: BENCHMARK PROBLEMS FDR RADIOLOGICAL ASSESSMENT CODES.
WADHAM,R.
NUR EC/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES WAHI,K.K.
NUREC/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES BEDDED SALT. Final Report.
NUREG/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
WALTON,W.B.
NUREC/CR-31BO: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FRO SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
WANC,P.C.
NUREC/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTURAL LOADS.
WARLING,J.C.
NUREG/CR-3119 VO1: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARD CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADI ATION. Volume 1: Executive Summary.
NUREG/CR-3118 VO2: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARD CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING R ADIATION. Volume 2: Supporting Documentation And Review Of Epidemiologic Studies Of Specific Chemicals.
NUREC/CR-3119 VO3:
HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING R ADIATION. Volume 3: Abstracts of NIOSH Criteria Documents.
WEBB,B.J.
NUREC/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
WEED,J.E.
NUREG/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS.
WEISS,A.J.
NUREG/CR-2331 VO2 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE O NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report,0ctober
-December 31,1982.
WELLS,J.E.
NUR EC/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS. Seismic Safety Margins Research Program.
WESTRICH,H.R.
NUR EG/CR-3401: LABORATORY-SCALE SODIUM-CARBONATE AGGREGATE CONCRETE INTERACTIONS.
WILDUNG,N.J.
89
NUREG/CR-3272: LOCA 91MULATION IN NRU PROGRAM. Data Report For The Fourth Materials Exoeriment (MT-4).
WILSON,C.L.
NUREG/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
NUREG/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
WOLF,J.J.
NUREG/CR-2649: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
WYSDC KI. J. J.
NUREG/CR-3394 V01: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LDSS OF CDOLANT ACCIDENTS. Executive Summary And Significant Findings.
NUREC/CR-3394 V02: PROBA3ILISTIC ASSLSSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF CDOLANT ACCIDENTS. Compendium Of Calculated Break Probabilities, Sump Blockage Frequencies,And Calculational Program Listing.
WYTHE,D.M.
NUR EG/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI) i LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS.
YANG,C.H.
NUREG/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTDR ACCIDENTS.
ZANETOS.M.A.
NUREQ/CR-3118 Vol: HEALTH EFFECTS OF DCCUPATIONAL EXPOSURE TO HAZARDGUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH N01ES ON IONIZING RADIATION. Volume 1: Ex ecutive Summary.
NUREG/CR-3118 V02: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS
' CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING R ADI ATIDN. Volume 2: Supporting Documentation And Review Of Epidamiologic Studies Of Specific Chemicals.
NUREG/CR-3118 V03: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 3: Abstracts of NIOSH Criteria Documents.
ZUCKER.M.
NUREG/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
W
..-r.-,.
l l'
Subject index This Index was developed from keywords and word strings in titles and ab-stracts. During this development period, there will be some redundancy, which will be removed later when a reasonable thesaurus hst been developed through experience. Suggestions for improvements are welcome.
AC Power System NUREQ/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
ATWS NUREG-1000 V02: GENERIC IMPLICATIONS OF ATWS EVENTS AT THE SALEM NUCLEAR POWER PLANT. Licensee And Staff Actions.
Abnormal Occurrence NUREG-0090 V06 N01: REPORT TO CONGRESG ON ABNORMAL OCCURRENCES. Jar.uary-March 1983.
Accelerograms NUREQ/CR-3327: ANALYSIS OF STRONO-MOTION DATA FROM THE NEW HAMPSHIRE EARTHGUAKE OF 18 JANUARY 1982.
Accident Sequence NUR EQ/CR-3234: THE POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH ELECTRICAL PENETRATION ASSEMBLIES UNDER SEVERE ACCIDENT CONDITIONS.
NUREQ/CR-2238 V04: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEMBER 1981. Volume 20.
NUREQ/CR-3241: PROBABILISTIC METHODS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SE3UENCES IN LOOP-TYPE LMFBRS.
Accident NUREQ/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS.
NURE0/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
NURE0/CR-32)4:
SUMMARY
OF NUCLEAR REQULATORY COMMISSION'S LOFT PROGRAM E%PERINEN7S.
NUREQ/CR-3257: RELAP5 ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
NURE0/CR-3258: RELAP5 ASSESSMENT:8EMISCALE NATURAL CIRCULATION TESTS 8-NC-2 AND S-NC-7.
r l
NUR EG/CR-3277: RELAP5 ASSESSMENT:SEMISCALE MOD-3 SMALL BMEAK TEST.
NURE0/CR-3329 VO1: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
NUREQ/CR-2874 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
NUR EQ/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC AIR.
NUREQ/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
Aeroso!
NUREQ/CR-2835: REVIEW OF MODELS APPLICABLE TO ACCIDENT AEROSOLS.
NUREQ/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS 91
AND SOLUTIONS IN STATIC AIR.
Aggregates NUREG/CR-3301: LABORATORY-SCALE SODIUM-CARBONATE AGGREGATE CONCRETE INTERACTIONS.
Air / Water Flow NUR EG/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM QUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
Airborne Particle Release NUREC/CR-2835: REVIEW OF MODELS APPLICABLE TO ACCIDENT AEROSOLS.
Airborne Releases NUREG/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC AIR.
l Analytic Hierarchy Process l
NUREG/CR-3447 V01: RESEARCH PRIORITIZATION USIfJG THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methods.
Anchor / Darling NUR EG/CR-3048: CLOSEOUT OF IE BULLETIN 80-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
l Anticipated Transients Without Scram NUREG-1000 V02: GENERIC IMPLICATIONS OF ATWS EVENTS AT THE SALEM NUCLEAR POWER PLANT. Licensee And Staff Actions.
Automatic Reactor Trip System NUREG-0090 V06 N01: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. January-March 1983.
Automation NUR EG/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
4 BLAST NUREG/CR-2874 V04: HIGH-TEMPERATURE CAS-COOLED REACTOR SAFETY STUDIES l
FOR THE DIVISION GF ACCIDENT EVALUATION, QUARTERLY PROGRESS REPORT, l
OCTOBER 1 - DECEMBER 31,1982.
BWR Refill-Reflood Program NUR EG/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
Basalt NUREG/CR-2739: DATA BASE FOR BASALT METHODOLOGY.
Basin Water NUREG/CR-3364: PRESENCE OF PATHOGENIC MICROORGANISMS IN POWER PLANT COOLING WATERS.
l Benchmark Problems NUREG/CR-3451: BENCHMARK PROBLEMS FOR RADIOLOGICAL ASSESSMENT CODES.
l Binomial Failure Rate i
NUR EG/CR-3437: DATA ANALYSIS USING BIONOMIAL FAILURE RATE COMMON CAUSE MODEL.
Bioassay NUREC/CR-3133: THE SENSITIVITY OF ADULT EMBRYONIC AND LARVAL CRAYFISH PROCAMBARUS CLARKII TO COPPER.
Biosphere NUREG/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEQUENCES.
Block Valve l
NUREG/CR-3050: CLOSEOUT OF IE BULLETIN 81-02 AND SUPPLEMENT: Failure Of Gate-Type Valves To Close Against Differential Pressure.
Borshole Plugging NUREC/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
Burial Site NUREG/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUGTION ENGINEERING BURIAL SITE HEMATITE, MISSOURI.
NUR EG/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
92
L i
ERWIN PLANT BURIAL SITE.
Burnup NUREG/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL RDD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
CARDS NUREG/CR-2146 V02: DYNAMIC ANALYSIS TO ESTABLISH NORMAL SHOCK AND VIBRATION OF RADIDACTJVE MATERIAL SHIPPING PACKAGES: Guarterly Progress Report. April 1,1981 - June 30,1981.
CCFL
~
NUREG/CR-3060: A CRITICAL REVIEW DF FLOODING LITERATURE.
COMPARE NUREG/CR-3278: HYDRDGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
t Calibration NUREG/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM i
THE NBS D(2)D-MODERATED (252)CF SOURCE.
Cancer NUREC/CR-3118 V01: HEALTH EFFECTS OF DCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING I
RADIATION. Volume 1: Executive Summary.
NUREG/CR-3118 V02: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING R ADIATIDN. Volume 2: Supporting Documentation And Review Of Epidemiologic Studies Of Specific Chemicals.
NUREG/CR-3118 V03: HEALTH EFFECTS OF DCCUPATIONAL EXPOSURE TO HAZARDOUS 4
CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATIDN. Volume 3: Abstracts of NIOSH Criteria Documents.
l Cask-Rail NUREG/CR-2146 VO2: DYNAMIC ANALYSIS TO ESTABLISH NORMAL SHOCK AND 1
VIBRATION DF RADIDACTIVE MATERIAL BHIPPING PACKAGES: Guarterly l
Progress Report. April 1,1981 - June 30,1981.
l Certificates Of Compliance NUREG-0303 V01 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
l NUREG-0383 VO2 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR R ADIDACTIVE MATERIALS PACKAGES. Certificates Of Compliance.
NUREG-0383 VO3 RO3: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Guality Assurance Programs For Radioactive Material Packages.
Chemical Reactions NUREG/CR-3401: LABORATORY-SCALE SODIUM-CARBONATE AGGREGATE CONCRETE INTERACTIONS.
Cladding Deformation NUREG/CR-3416 VO2: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIG CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
Cladding Materials Deformation 3
NUREG/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACIDR.
Cladding NUREG/CR-3155: MODIFICATION DF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING DN THE INNER SURFACE.
Cleanup NUREG/CR-3381: EVALUATIDN OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
Codes NUREG/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1981.
Combustion Products NUREG/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A REVIEW OF THE LITERATURE.
93 N.
Cschustien NURE0/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS.
NURE0/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
Common Cause Failure Rates NURE0/CR-3437: DATA ANALYSIS USING BIONOMIAL FAILURE RATE COMMON CAUSE MODEL:.
Compact Toughness NURE0/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
d Compensated Ion Chamber NURE0/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SEN8 ORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
4 Computer Code l
NURE0/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS OF OXIDE FUEL RODS.
NURE0/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODEL8 FOR THE BWR VERSION OF TRAC.
NURE0/CR-2574: BWR REFILL-REFLOOD PROGRAM TA8K 4.7 - MODEL DEVELOPMENT TRAC-BWR COMPONENT MODEL8.
NURE0/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRES 8URE VESSEL WITH CLADDING ON THE INNeER SURFACE.
NURE0/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRES 8URE VESSEL FAILURE.
NURE0/CR-3451: BENCHMARK PROBLEMS FOR RADIOLOGICAL ASSESSMENT CODES.
NURE0/CR-3425: LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND 8PHERICAL PHANTOMS.
Computer Program NURE0/CR-3238: PNASIM: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIQUE OF MASS FLOW MEASUREMENT.
NUREQ/CR-1246 V04: SAFE USERS MANUAL. Volume 4: Computer Programs.
Concrete-Steel Interaction NURE0/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS Concrete NURE0/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH,C032-)
RECIPROCAL SYSTEM NURE0/CR-3284i REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALIJATIONS NVRE0/CR-3401: LABORATORY-SCALE SODIUM-CARBONATE A00REGATE CONCRETE INTERACTIONS.
Conservation NURE0/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EGOATIONS OF MULTIPHASE FLOW.
Container NURE0/CR-2482 V04: REVIEW OF DOE WASTE PACKAGE PROGRAM. But, task 1. 1 -
National Waste Package Program,0ctober 1982 - March 1983.
Containment Emergency Sump Performance NURE0/CR-3394 V01: PROBABILISTIC A88E88 MENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Executive Summary And Significant Findings.
NURE0/CR-3394 V02: PROBABILISTIC A88EB8 MENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Compendium Of Calculated Break Probabilities, sump Blockage Frequencies,And Calculational Program Listing.
Containment NURE0/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS.
NURE0/CR-2492 V04: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1982 - March 1983.
NURE0/CR-3091 V02: REVIEW OF WASTE PACKAGE VERIFICATION 94 i
L __
TESTS. Semiannual Report Covering The Period October 1982 - March 1983.
NURE0/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
NURE0/CR-3385: PEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
. NURE0/CR-2938: TAILINGS TREATMENT TECHNIQUES FOR URANIUM MILL WASTE: A L
Review Of Existing Information.
Control Functions NURE0/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
Control Room Crows NURE0/CR-3371 V01: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NURE0/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
4 Convection Circulation Loop NURE0/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVECTION LOOP AND THE1R APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESION.
Cooling-Tower NURE0/CR-3364: PRESENCE OF PATHOGENIC MICROORGANISMS IN POWER PLANT COOLING WATERS.
Copper Sensitivity NURE0/CR-3133: THE SENSITIVITY OF ADULT, EMBRYONIC AND LARVAL CRAYFISH PROCAMSARUS CLARKII TO COPPER.
Copper NURE0/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM THE H.B.
ROBINSON STEAM ELECTRIC /LANT.
- Core Behavior NURE0/CR-3214:
SUMMARY
OF NUCLEAR REQULATORY COMMISSION'S LOFT PROGRAM EXPERIMENTS.
Core Cooling NURE0/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A SWR /4.
Core Jet Length NURE0/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC BIMULATION.
Core Melt NURE3/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF BEVERE DYNAMICS.
Core Meltdown NORE0/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report. April 1980 - April 1983.
Core Retention System NURE0/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1983.
Cost NURE0/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE j
NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameters.
Countercurrent Flow Limitation NURE0/CR-3060: A CRITICAL REVIEW OF FLOODING LITERATURE.
l Crack Mouth 1
NUREQ/CR-3402: ALTERNATVE PROCEDURES FOR U-R CURVE DETERMINATION.
Cracking L
NURE0/CR-3048: CLOSEOUT OF IE BULLETIN 80-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
Cracks NURE0/CR-3464: THE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTI AL THROUGH WALL CRACKS.
96 er
-=r='*
r
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NUREG/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER THROUGH SLITS.
Crayfish NUREQ/CR-3133: THE SFNSITIVITY OF ADULT, EMBRYONIC AND LARVAL CRAYFISH PROCAMBARUS CLARKII TO COPPER.
Critical Flow NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION, SONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
' Cylindrical Ccre Test Facility Evaluation NUREG/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 39).
Decommissioning NUREC/CR-0130 ADDO2: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Effects On Decommissioning Of Interim Inability To Dispose Of Wastes Offsite.
NUREQ/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DECONHISSIONING A REFERENCE BOILING WATER REACTDR POWER STATION. Ef fects On Decommissioning of Interim Inability Tc Dispose of Wastes Offsite.
NUREG/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameters.
NUREG/CR-3336: EVALUATIDN OF NUCLEAR DECOMMISSIONING PROJECTS. Summary Report,Ames Laboratory Research Reactor. Docket No. 50-116.
NUREQ/CR-3370: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS
SUMMARY
REPORT NORTH CAROLINA STATE UNIVERSITY RESEARCH AND TRAINING REACTOR.
Decontamination NUR EG/CR-3381: EVALUATION OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECGNTAMINATION PROCESS.
Degraded Core Accident 5
NUREG/CR-3278: HYDRDGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
l Densitometer NUREG/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITOMETER AND ITS. APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
Detection NUR EG/CR-3221: MC8 A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION.
Diesel Generator Failure NUREG/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
Discharge Waters NUREC/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM THE H B.
ROBINSON STEAM ELECTRIC PLANT.
Displacement Technique NUREG/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
Display Screens r
NUREG/CR-30C3 VO2: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
Disposal Charge NUREG/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameters.
Disposal NUREG/CR-3111 VO1: -AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEDLOGIC REPOSITORIES.
Dose-To-Man I
NUREQ/CR-3451: BENCHMARK PROBLEMS FOR RADIDLOGICAL ASSESGMENT CODES.
l
- Dose l
NUREG/CR-2850 VO2: POPULATION DOSE COMMITMENTS DUE TO RADIDACTIVE l
RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
l 96 I
V Dosimeters NUREG/CR-3400: ANALYSIS OF PEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EOUIVALENT AT e
. NUCLEAR POWER PLANTS.
Dosimetry Improvement Program NUREG/CR-2805 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY j
IPPROVEPENT PROGRAM. GuarterIg Progress Report,0ctober 1982 - Decemb er L
1982.
Draft Environmental Statement t
NUREG-1011: DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM QENERATOR REPAIR AT THE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
50-266.(Wisconsin Electric Power Company)
Drug And. Alcohol Abuse NUREG/CR-3196: DRUG AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN TE NUCLEAR UTILITY INDUSTRY.
4 i
Earthquake NUREG/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
i NUREG/CR-3327: ANALYSIS OF STRONG-MOTION DATA FROM THE NEW HAMPSHIRE EARTHGUAKE OF 18 JANUARY 1982.
NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS. Seismic Safety Margins Research Program.
NUREQ/CR-33SO: CURRENT METHODOLOGIES FOR ASSESSING SEIBMICALLY INDUCED SETTLEMENTS IN SOIL.
NUREQ/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAR FACILITIES WITH SEISMIC EQUIPMENT QUALIFICATION IN NUCLEAR PLANTS (A-46).
Eddy-Current Inspection MUREQ/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
NUREG/CR-32OO V01: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
t' Effluent Solutions NUREG/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMFNTS IN A FLOW-THRDUGH SYSTEM.Model Formulations And Preliminary Results.
Electrical Equipment Failures NUREQ/CR-3122: PDTENTIAL DAMAGING FAILURE MODES OF HIGH-AND MEDIUM-VOLTAGE ELECTRICAL EGUIPMENT.
Electrical Penetration Assemblies NUREQ/CR-3234: THE POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH ELECTRICAL PENETRATION ASSEMELIES UNDER SEVERE ACCIDENT CONDITIONS.
Electrical
- NUREG/CR-3266: SEISMIC & DYNAMIC QUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
Electromagnetic Interference 4
NUREG/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS.
Emergency Core Cooling System NUREQ/CR-3394 V01: PROBABILISTIC ASSESSMENT G! RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Executive Summary And i
Significant Findings.
NUREG/CR-3394 VO2: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP
[
BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Compendium 0F Calculated Break Probabilities, Sump Blockage Frequencies,And Calculational Program Listing.
Emergency Onsite AC Power Systems 97 b
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E V
NUREG/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
Emergency Operating Procedures NURE0/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM C REWS. Volume 2: Data Results.
Employee Assistance Program NURES/IR-3196: DRUS AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PRD0 RAMS IN THE NUCLEAR UTILITY INDUSTRY.
Energetic Fuel / Coolant Interactions 4
NURE0/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF SEVERE 1
DYNAMICS.
Energetics NUREG/CR-3392: PMOCE88 NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition.
Enforcement Actions NUREG-0940 V02 NO2: ENFORCEMENT ACTIONS: SIONIFICANT ACTIONS RESOLVED.Guarterly Progress Report, April-June 1983.
Epicor-II NUREG/CR-3381: EVALUATION OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
Equilibrium Guality NURE0/CR-3373: AIR / WATER BUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM GUALITY AND MA88 FLUX DISTRIBUTION IN A ROD BUNDLE.
Equipment Qualification NUREG-1010: BEISMIC GUALIFICATION OF EGUIPMENT IN OPERATING PLANTS.
Status Report, Unresolved Safety Issue A-46.
Equipment NURE0/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EQUIPMENT IMPORTANT TO BAFETY IN NUCLEAR PDWER PLANTS.
NUREC/CR-3017: CORRELATION OF BEIBMIC EXPERIENCE DATA IN NON-NUCLEAR FACILITIES WITH SEISMIC EQUIPMENT GUALIFICATIDN IN NUCLEAR PLANTS (A-46).
Exposure NURE0/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EQUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
FIST NURE0/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TEST PLAN.
FRAP-T6 NURE0/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FDR THE TRANSIENT ANALYSIS OF OXIDE FUEL RODS.
FRAPCON-2 NURE0/CR-3416 V02: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS C APABILITIES. Phaos 1 Topical Rept Vol 2: Code Evaluation.
FRAPCON NURE0/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
Fault Retes NURE0/CR-3289: COMMON CAUBE FAULT RATES FOR INSTRUMENTATION AND CONTROL A88EMBLIES. Estimates Based on Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, 1976-1981.
Final Environmental Impact Statement NUREG-0925: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF THE TETON URANIUM ISL PROJECT. Docket No. 40-8781. (Teton Exploration Drilling,Inc.)
Final Environmental Statement NUREG-1011: DRAFT ENVIRONMENTAL STATEMENT RELATED TO BTEAM QENERATOR REPAIR AT THE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
50-266.(Wisconsin Electric Power Company)
Fires 98
.y NURE0/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A REVIEW OF THE i
LITERATURE.
Fission Gas Release NURE0/CR-3414 V02: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
Flaw Behavior NURE0/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING ON THE INNER SURFACE.
Flaws NURE0/CR-3334 V01: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983.
Flooding NURE0/CR-3040: A CRITICAL REVIEW OF FLOODING LITERATURE.
Floor Response Spectra NURE0/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
(
Flow-Measurements NURE0/CR-3238: PNASIN: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIGUE OF MASS FLOW MEASUREMENT.
Flowpaths NURE0/CR-3473: ROCK MAS 8 SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFDRMANCE. Annual Report, June 1982 - May 1983.
Fracture Toughness NURE0/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VE8SEL FAILURE.
Fractured Rocks i
NUME0/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
Fuel Assemblies i
NURE0/CR-1754: TEClelOLD0Y, SAFETY AND COST 8 0F DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning i
Radiation Exposure And Costs To Selected Parameters.
4~
Fuel Bundle NURE0/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The i
Fourth Materials Experiment (MT-4).
NURE0/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
Fuel Channel Analysis NURE0/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT TRAC-BWR COMPONENT MODELS.
Fuel Rod NUREQ/CR-2148 ADD 01: FRAP-T4: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS OF DXIDE FUEL RODS.
Fuel NURE0/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
NURE0/CR-3416 V02: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
Full Integral Simulation Test NURE0/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TEST PLAN.
Gamma-Ray Scattering NURE0/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITOMETER AND ITS APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
Cas-Cooled Reactor NURE0/CR-2874 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
99 i
Cas-Liquid Flow NUREG/CR-3060: A CRITICAL REVIEW OF FLOODING LITERATURE.
Gaussian NUREG/CR-3283: FIRST EXCURSION PROBLEMS FOR CAUSSIAN VECTOR PROCESSES.
Geologic Disposal NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
NUREG/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEGUENCES.
Geologic Repositories NUREG/CR-3111 V01: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
NUREQ/CR-3405 V01: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, March 1982 - April 1983.
Geological NUREG/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
Graphic Displays NUREQ/CR-3003 YO2: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
Groundwater Contamination NUREG/CR-2938: TAILINGS TREATMENT TECHNIGUES FOR URANIUM MILL WASTE: A Review Of Existing Information.
Groundwater Corrosion NUREC/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERI ALS USED FOR HIGH-LEVEL ~ WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
Groundwater Flow NUREG/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL C ONSEGUENCES.
Groundwater NUREG/CR-2739: DATA BASE FOR BASALT METHODOLOGY.
NUREG/CR-3259: LEAh DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
HLW-Borosilicate Glasses NUREG/CR-3472 V01: SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS.
Half-Bead Weld Repair NUREQ/CR-3265: EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS.
Hazardous Substances NUREG/CR-3118 V01: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IvNIZING RADIATION. Volume 1: Enecutive Summary.
NUREG/CR-3118 V02: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 2: Supporting Documentation And Review Of Epidemiologic Studies Of Specific Chemicals.
NUREG/CR-3118 V03: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING R ADIATION. Volume 3: Abstracts of NIOSH Criteria Documents.
Health Risk NUREG/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RAD 10NUCLIDES AND RAD 10 LOGICAL CONSEGUENCES.
Heat Transfer NUREG/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
Heavy-section Steel Technology 100
NUREC/CR-3334 V01: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983.
Heisadampfreaktor NUREQ/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
High Level Nuclear Waste NUREQ/CR-3219 VO1: DRAFT TECHNICAL POSITION SUBTASK 1.1: WASTE PACKAGE l
PERFORMANCE AFTER REPOSITORY CLOSURE.
High Level Radioactive Waste NUREG/CR-3091 V02: REVIEW OF WASTE PACKAGE VERIFICATION TESTS. Semiannual Report Covering The Period October 1982 - March 1983.
High Level Waste Package NUREQ/CR-2482 V04: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1982 - March 1983.
High Level Waste Repository NUREC/CR-3451: BENCHMARK PROBLEMS FOR RADIOLOGICAL As3ESSMENT CODES.
High-Level Radioactive Waste NUREG/CR-3405 V01: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, March 1982 - April 1983.
NUREG/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES.
High-Level Waste NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
High-Temperature NUREG/CR-2874 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
Hole Sealing NUREG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - Nag 1983.
Human Behavioral Methodologies NURE0/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
Human Engineering Design NUREC/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NUREG/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
Human Factors Program Plan NUREG-09BS VO1:
U. S.
NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.
Human Reliability Analysis NUREG/CR-2254: A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS.
l Humans i
NUREG/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
Hydraulic Testing NUREG/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY l
EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
i Hydrodynamics NUREG/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC SIMULAYION.
Hydrogen Embrittlement NUREG/CR-3282: INTERNAL HYDROCEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
Hydrogen NUREG/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR 101 i
ACCIDENTS.
NUREG/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
NURE0/CR-3275: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
IE Bulletin 81-02 NURE0/CR-30*0: CLOSEQUT OF IE BULLETIN 91-02 AND SUPPLEMENT: Failure 07 Gate-Type Valves To Close-Against Differential Pressure.
IPSAR-
-NUREG-OS$1 801: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM -
R.-
E.
GINNA NUCLEAR POWER PLANT. Docket No.
50-244.(Rochester Gas And Electric Corporation)
IREP NURE0/CR-3085 V02: INTERIM RELIABILITY EVALUATION PRDORAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol II - Appendix A And 3.1 Thru B.4. Docket No. 50-245.(Northeast Nuclear Energy Company)
NUREG/CR-3055 V03: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol III Appendix B.5 Thru B.S. Docket No. 50-245.(Northeast Nuclear Energy)
NURE0/CR-3085 VO4: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE HILLSTONE POINT UNIT 1 NUCLEAR POWdR PLANT.Vol IV Appendix B.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
Ice Condensor Containment NURE0/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
Inspections NUREG-0040 V07 NO2: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report, April 1983 - June 1983.(White Book)
Institutional Problems NURE0/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY 00ALS FOR NUCLEAR POWER PLANTS.
I Instrument Development Loop i
NURE0/CR-3139: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FOR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
Instrumentation And Control Assemblies NURE0/CR-3289: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL
-ASSEMBLIES. Estimates Based on Licensee Event Reports At U.S.
Commercial Nuclear Power Plants, 1976-1921, i
Insulation NURE0/CR-2982 RO1: BUDYANCY, TRANSPORT,AND HEAD LOSS OF FIBROUS REACTOR INSULATION.
NURE0/CR-3394 V01: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Executive Summary And Significant Findings.
NURE0/CR-3394 V02: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Compendium Of Calculated Break Probabilities, Sump Blockage Frequencies,And Calculational Program Listing.
Intake Waters NURE0/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM-THE H.B.
ROBINSON STEAM ELECTRIC PLANT.
L Integrated Plant Safety Assessment Report NUREG-0821 801: INTEGRAYED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - R.
E.
GINNA NUCLEAR POWER PLANT. Docket No.
50-244.(Rochester Gas And Electric Corporation)
NUREG-0525 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - BIG ROCK POINT PLANT. Docket No. 50-155.
(Consumers Power Company)
Interim Reliability Evaluation Program NURE0/CR-3085 V02: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol II - Appendix A And B.1 Thru B.4. Docket No. 50-245.(Northeast Nuclear Energy Company)
NURE0/CR-3095 V03: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF 102
THE MILL 8 TONE POINT UNIT 1 NUCLEAR POWER PLANT. Val III Appendix B.5 Thru B.8. Docket No. 50-245.(Northeast Nuclear Energy)
NURE0/CR-3085 V04: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILL 8 TONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol IV Appendix B.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
Inventory Difference NUREG-0430 V03 NO2: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Date. July - December 1982.
Inverted Annular Flow NURE9/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC SIMULATION.
Irradiated Reactor Fuel NUREG-0725 R03: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
Irradiation NURE9/CR-3282: INTERNAL HYDROGEN EM8RITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
Jet Pumps NURE9/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A BWR/4.
Kinetics Parameters NURE0/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
LER NUREG-1022: LICENBEE EVENT REPORT SYSTEM. Description Of System And Guidelines For Reporting.
NUREO/CR-2000 V02 N6: LICENBEE EVENT REPORT (LER) COMPILATION:For Month of June 1983.
NUREG/CR-2000 V02 N7: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of July 1983.
NUREG/CR-2000 V02 N8: LICENSEE EVENT REPORT (LER) CDNPILATION:For Month Of August 1983.
LOCA NUREG/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TEST PLAN.
NUREG/CR-2982 RO1: BUOYANCY, TRANSPORT,AND HEAD LDSS OF FIBROUS REACTOR INSULATION.
NUREQ/CR-3083: EXPERIMENT OPERATION 3 PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
NUREO/CR-3138: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FOR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
NURE0/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
NURE0/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
NUREG/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A BWR/4.
LOFT NUREG/CR-3214:
SUMMARY
OF NUCLEAR REGULATORY COMMISSION'S LOFT PROGRAM EXPERIMENTS.
LWR-P V-SDIP NUREG/CR-2805 V04: LWR PRES 8URE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1982 - December 1982.
Lawsuit NUREG-1020 V01: OPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al V.The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket No. 50-289.
1 NUREG-1020 V02: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT DN TMI-1. General Public Utilities Corporation, et al V.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket 103
g r
No.50-289.
Leaching Mechanisms NUREG/CR-3472 VO1: SURFACE PRDPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS.
Leek' Detection Systems 4
NUREG/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS I
IMPDUNDMENTS WITH SYNTHETIC LINERS.
. Leak-Flow Rates NUREG/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCDOLED WATER THROUGH SLITS.
Leak NUREG/CR-3234: THE POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH i
ELECTRICAL PENETRATION ASSEMBLIES UNDER SEVERE ACCIDENT CONDITIONS.
' Legal Challenges NUREC/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY GDALS FOR NUCLEAR POWER PLANTS.
Licensed Fuel Facilities NUREG-0430 VO3 NO2: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. July - December 1982.
Licensed Operating Reactors NUREG-OO2O VO7 NO3: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPDRT. Data As Of February 28,1983.(Greg Book)
NUREG-OO2O YO7 NO4: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of March 31,1983. (Greg Book)
NUREG-OO2O VO7 N06: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of May 31,1983.(Greg Book)
Licensee Event Report i
NUREG-1022: LICENSEE EVENT REPORT SYSTEM. Description Of System And Guidelines For Reporting.
NUREC/CR-2OOO VO2 N6: LICENSEE EVENT REPORT (LER) COMPILATION:For Month of June 1983.
NUREC/CR-2OOO YO2 N7: LICENSEE EVENT REPORT (LER) COMPILATION:For Month of July 1983.
NUREG/CR-2OOO YO2 NS: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of August 1983.
NUREC/CR-3289: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES. Estimates Based on Licensee Event Reports At U. S.
Commercia! Nuclear Power Plants, 1976-1981.
Licensing Actions NUREG-0748 VO3 N06: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
.Dcta As Of June 30,1983.(Orange Book)
NUREG-0748 VO3 NO7: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of July 31,1983.(Orange Book)
Licensing Examinations NUREG-0985 YO1:
U. S.
NUCLEAR REGULATORY CDMMISSION HUMAN FACTORS PROGRAM PLAN.
Licensing Reviews NUREG-0500 V12 N06: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of June 30,1983.(Blue Book)
NUREG-0500 V12 NO7: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As l
Of July 31,1983 (Blue Book)
NUREG-0580 V12 NOS: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As of August 31,1983 (Blue Book)
Licensing NUREG/CR-2238 VO4: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEEBER 1981. Volume 20.
Load Line.
NUREG/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
Loading Conditions NUREG/CR-3283: FIRST EXCURSION PROBLEMS FOR CAUSSIAN VECTOR PROCESSEE.
104
Lccdo NUREC/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTURAL LOADS.
Loss Of Forced Circulation i
NUREG/CR-3306: LWR AND HTGR COOLANT DYNAMICS: THE CONTAINMENT OF SEVERE DYNAMICS.
Loss-Detection NUREG/CR-3221: MCt:A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE
\\
ESTIMATES AND PERFORMANCE EVALUATION.
Loss-Of-Coolant Accident NUREC/CR-2982 RO1: BUOYANCY, TRANSPORT,AND HEAD LDSS OF FIBROUS REACTOR INSULATIDN.
NUREG/CR-3083: EXPERIMENT OPERATIDNS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTDR.
NUREG/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
NUREG/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The i
Fourth Materials Experiment (MT-4).
Loss-Of-Fluid Test NUREG/CR-3214:
SUMMARY
OF NUCLEAR REGULATORY COMMISSION'S LOFT PROGRAM i
EXPERIMENTS.
Low Fracture Toughness NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPDRTS. Seismic Safety Margins Research Program.
Low Level Radioactive Waste NUREQ/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADIDACTIVE WASTE DISPOSAL FACILITY.
MC&A NUREG/CR-3221: MC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION.
Magnetic Field NUREG/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS.
Main Feedwater Line Break NUREG-0090 VO6 N01: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. January-March 1983.
Maintenance Personnel NUREG/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
(
Malcolm Foundry NUREG/CR-3048: CLOSEOUT OF IE BULLETIN 80-21: VALVE PARTS SUPPLIED BY MALCOLM FDUNDRY.
Management NUREG-0985 VO1:
U. S.
NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.
NUREG/CR-3215 VO2: DRGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
Manual NUREC/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
Mass. Flow Measurement NUREQ/CR-3138: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FOR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
Mass Flux Distribution NUREG/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM l
. GUALITY AND. MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
l Materials Centrol And. Accounting NUREG/CR-3221: MC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION.
Materials Deformation / Rupture NUR EG/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The 105
~
Fourth Materials Experit-ent (MT-4).
Mechnical Eguipment NURE0/CR-3266: SEISMIC 4: EYNAMIC QUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
Migration NURE0/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS 80!.UTIONS AND SEDIMENTS IN A FLOW-THROUGH SYSTEM.Model Formulationv And Preliminary Results.
Model Test NURE0/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 38).
Monte Carlo NURE0/CR-3425: LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS.
Multiphase System NURE0/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EGUATIONS OF MULTIPHASE FLOW.
NDA NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIQUES.
NUFREG-N NURE0/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR CDUPLED DENSITY-WAVE OSCILLATIONS IN j
BOILING WATER NUCLEAR REACTORS.
Naegleria NURE0/CR-3364: PRESENCE OF PATHOGENIC MICRDORGANISMS IN POWER PLANT COOLING WATERS.
Neutron Dose NURE0/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EQUIVALENT AT NUCLEAR POWER PLANTS.
Neutron Exposure NURES/CR-2 SOS V04: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1992 - December 1982.
Neutron Sensor NURE0/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
Neutron Sources NURE0/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
Nodal Analysis NUREQ/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
Nondestructive Examination NURE0/CR-3307 V01: REACTOR SAFETY RESEARCH PRDORANS.Guarterly i
' Report, January -March 1983.
Nuclear-Coupled Density Wave Oscillation NURE0/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR CDUPLTD DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
DCA-I NURE0/CR-3155: MODIFICATION DF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING DN THE INNER SURFACE.
l ORECA NORE0/CR-2874 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,0UARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
Off-Specification NURE0/CR-3265: ' EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPA!RS.
i 106
i Offsite AC Power Loss NUREG/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-28 POWER LOSS TEST SERIES (TEST S-PL-1,-2 AND-3).
Onsite Waste Storage NUREG/CR-0130 ADDO2: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Ef f ects On Decommissioning Of Interim Inability To Dispose Of Wastes Offsite.
NUREG/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND CJSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR POWER STATION. Effects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
7 Operating Reactors NUREC-0748 VO3 N06: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of June 30,1983.(Orange Book)
NUREG-0748 VO3 NO7: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of July 31,1983.(Orange Book)
Operating Units NUREG-OO2O VO7 NO3: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of February 28,1983.(Greg Book)
NUREG-OO2O VO7 NO4: LICENSED DPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of March 31,1983. (Greg Book)
NUREG-OO2O VO7 N05: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of April 30,1983.(Greg Book)
NUREG-OO2O VO7 N06: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of May 31,1983.(Greg Book)
Operator Performance NUREC/CR-3OO3 VO2: HUMAN ENGINEERING DESIGN CONSIDERATIDNS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
Operators NUREG/CR-3371 VO1: TASK ANALYSIS DF NUCLEAR POWER PLANT CONTROL RDOM CREWS. Volume 1: Project Approach And Methodology.
NUREG/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL RODM CREWS. Volume 2: Data Results.
Organizational Factors NUREC/CR-3215 VO1: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
Organizational NUREC/CR-3215 VO2: DRGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
PAPA i
NUREC/CR-3241: PROBABILISTIC METHOCS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SEQUENCES IN LOOP-TYPE LMFBRS.
PNASIM l
NUR EQ/CR-3238: PNASIM: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIGUE OF MASS FLOW MEASUREMENT.
PORV NUREG/CR-3050: CLOSEOUT OF IE BULLETIN 81-02 AND SUPPLEMENT: Failure Of Oate-Type Valves To Close Against Differential Pressure.
l Packages NUREG-0383 VO2 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR j
RADIOACTIVE MATERIALS PACKAGES. Certificates Of Compliance.
NUREC-0383 VO3 RO3: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Summary Report OF NRC Approved Guality l
Assurance Programs For Radioactive Material Packages.
Packing Material j
NUREQ/CR-2482 VO4: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1982 - March 1983.
I Parallel Channel Effects NUREC/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING l
EMERGENCY CORE COOLING IN A BWR/4.
Personnel Dosimeters 107
NUREC/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
Personnel NUREG/CR-3196: DRUG AND ALCOHOL ABUSE: THE BASES FDR EMPLOYEE ASS 7 STANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
Piping Response NUREC/CR-31BO: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FRDM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
Piping Systems NUREG/CR-3394 VO1: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LDSS OF COOLANT ACCIDENTS. Executive Summary And Significant Findings.
NUREG/CR-3394 V02: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LDSS OF CDOLANT ACCIDENTS. Compendium OF Calculated Break Probabilities, Sump Blockage Frequencies,And Calculational Program Listing.
Piping NUR EG/CR-3464: THE APPLICATION DF FRACTURE PRDOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTIAL THROUGH WALL CRACKS.
NUREC/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER THROUGH SLITS.
Plugs NUR EG/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BDREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
Population Dynamics NUREC/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition.
Population Radiation Dose Commitments NUREG/CR-2850 V02: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
Porous Body Models i
NUREG/CR-3188: THE STEADY STATE POROUS BODY ASSEMBLY CODE (SPAC): Users Manual.
Power Distribution NUREG/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
Power Loss NUREC/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-28 POWER LOSS TEST SERIES (TEST S-PL-1,-2,AND-3).
Pressure Vessel Surveillance NUREC/CR-2805 V04: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1982 - December 1982.
Pressure Vessel NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
Primary-Coolant Containment Systems NUREC/CR-3334 VO1: HEAVY SECTIDN STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983.
Probabilistic Accident Progression' Analysis NOREG/CR-3241: PROBABILISTIC METHODS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SEGUENCES IN LOOP-TYPE LMFBRS.
Probabilistic Risk Analyses
)
NUREC/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
j NUREC/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
]
Probabilistic Risk Assessment NUR EC/CR-2254: A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS.
Proportional Counters NUREG/CR-3400: 'NALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND 108
PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EQUIVALENT AT NUCLEAR POWER PLANTS.
Public Health NUREG/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EQUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
4 Pulsed-Neutron Activation NUREC/CR-3238: PNASIM: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIGUE OF MASS FLOW MEASUREMENT.
Guality Assurance Programs NUREG-0383 VO1 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
NUREG-0383 VO2 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Certificates Of Compliance.
NUREG-0383 VO3 RO3: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Guality Assurance Programs For Radioactive Material Packages.
RELAP5 NUREC/CR-3257: RELAP5 ASSESSMENT:LDFT TURBINE TRIP L6-7/L9-2.
NUREC/CR-3258: RELAP5 ASSESSMENT: SEMISCALE NATURAL CIRCULATION TESTS S-NC-2 AND S-NC-7.
l NUREC/CR-3277: RELAP5 ASSESSMENT: SEMISCALE MOD-3 SMALL BREAK TEST.
i NUREC/CR-3329 VO1: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
l RELAP5/ MOD 1 NUREG/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAPS/ MOD 1 CODES l
AT BNL IN FY 1981.
NUREG/CR-3257: RELAPS ASSESSMENT: LDFT TURBINE TRIP L6-7/L9-2.
Radiation Exposure NUREG/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameters.
NUREG/CR-3118 VO1: HEALTH EFFECTS DF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 1: Executive Summary.
NUREG/CR-3118 VO2: HEALTH EFFECTS DF DCCUPATIDNAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 2: Supporting Documentation And Review Of Epidemiologic Studies Of Specific Chemicals.
NUREG/CR-3118 VO3: HEALTH EFFECTS OF DCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IDNIZING RADIATION. Volume 3: Abstracts of NIOSH Criteria Documents.
Radiation Sensitivity NUREG/CR-3362: THF INFLUENCE OF LYMPHOHEMATOPOIETIC DISEASE IN AL1ERING RADIATION INDUCED CANCER SENSITIVITY IN MICE.
Redioactive Material Packages NUREG-0383 VO1 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
NUREG-0383 VO2 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of Compliance.
NUREG-0383 VO3 RO3: DIRECTORY DF CERTIFICATES OF COMPLIANCE FOR RADIDACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Guality Assurance Programs For Radioactive Material Packages.
Radioactive Materials Shipping Package NUREC/CR-2146 VO2: DYNAMIC ANALYSIS TO ESTABLISH NORMAL SHOCK AND VIBRATION OF RADIOACTIVE MATERIAL SHIPPING PACKAGES: Guarterly Progress Report. April'1,1981 - June 30,1981.
Rcdioactive Waste NUREQ/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
109
ERWIN PLANT BURIAL SITE.
Radiological Survey NUREG/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL SITE HEMATITE, MISSOURI.
NUREQ/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
Radionuclide Releases NUREG/CR-2850 VO2: POPULATIDN DDSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
Radionuclide Transport NUREG/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEGUENCES.
NUREG/CR-3451: BENCHMARK PROBLEMS FOR RADIDLOGICAL ASSESSMENT CODES.
Reactor Coolant Pump Supports NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS.-Seismic Safety Margins Research Program.
. NUREG/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING ON THE INNER SURFACE.
Reactor Safety NUREG/CP-OO47: TRANSACTIONS OF THE ELEVENTH WATER REACTOR SAFETY RESEARCH INFDRMATION MEETING.To Be Held At National Bureau of Standards,Gaithersburg, Maryland,0ctober 24-28,1983.
NUREG/CR-3307 VO1: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report, January -March 1983.
NUREG/CR-3359 Vol: PHYSICS OF REACTOR SAFETY.Guarterly Report. January -
March 1983.
Reciprocal System t
NUREC/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH,CO32-)
j RECIPROCAL SYSTEM j
Recirculating Piping Loop NUREG/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM l
SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
i Recirculation Sump NUREC/CR-3394 VO1: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF CDOLANT ACCIDENTS. Executive Summary And Significant Findings.
NUREG/CR-3394 VO2: PROBABILISTIC ASSESSMENT DF RECIRCULATION SUMP i
BLOCKAGE DUE TO LDSS OF COOLANT ACCIDENTS. Compendium DF Calculated Break Probabilities. Sump Blockage Frequencies,And Calculational l
Program Listing.
l
-Refill and Reflood l
NUREC/CR-3138: MEASUREMENT DF TWO-PHASE FLOW AT THE CDRE/ UPPER PLENUM INTERFACE FOR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
Reflood NUREQ/CR-3138: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM f
INTERFACE FOR A PWR GEDMETRY UNDER SIMULATED REFLDOD CONDITIONS.
Regulatory Agenda NUREG-0936 VO2 NO2: NRC REQULATORY AGENDA.Guarterly Report, March - May l
1983.
Regulatory And Technical Reports NUREG-0304 VOG NO2: REQULATORY AND TECHNICAL REPORTS. Compilation For Second Guarter 1983.
Regulatorg' Licensing NUREG-0580 V12 N06: REGULATORY LICENSING STATUS
SUMMARY
REPORT.Date As Of June 30,1983.(Blue Book)
NUREG-0580 V12 NOG: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As
{
of August 31,1983 (Blue Book) l 110
Reinforced Concrete NUREG/CR-3003 V02: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE R AY TUBE-GENERATED DISPLAYS.
NUR EG/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS Release NUREG/CR-2482 VO4: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1902 - March 1983.
NUREG/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEGUENCES.
NUREG/CR-3111 VO1: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES.
NUREG/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
Remmeters NUREC/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EQUIVALENT AT NUCLEAR POWER PLANTS.
Repositories NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
Repository NUREG/CR-2482 VO4: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1982 - March 1983.
NUREC/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-CROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEQUENCES.
NUREC/CR-3219 VO1: DRAFT TECHNICAL POSITION SUBTASK 1.1: WASTE PACKAGE PERFORMANCE AFTER REPOSITORY CLOSURE.
Research Programs NUREG/CR-3447 Vol: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methods.
Research Safety NUREG/CR-2238 VO4: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEMBER 1981. Volume 20.
Restart Hearing NUREG-1020 VO1: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al V.The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket No. 50-289.
NUREG-1020 V02: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al V.
The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket No.50-289.
Risk Assessment Methodology NUR EG/CR-2402: RISK (NALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
Risk Importance Measures NUREG/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
Risk NUREC/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
Rod Bundle NUREC/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
Rod Deformation NUREG/CR-3416 Vol: EVALUATION OF POWER REACTOR FUEL POD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
Rod NUREG/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
111
b
. NUREC/CR-3416 VO2: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
SAFE NUREC/CR-1246 VO4: SAFE USERS MANUAL. Volume 4: Computer Programs.
SCORE NUREG/CR-2874 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
SEP NUREG-0821 SO1: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - R.
E.
GINNA NUCLEAR POWER PLANT. Docket No. 244.(Rochester Gas And Electric Corporation)
SIMULATE NUREG/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
SNAP NUREQ/CR-3423: THE SAFEGUARDS NETWORK ANALYSIS PROCEDURE SNAP: A User's Manual.
SPAC NUREC/CR-3188: THE STEADY STATE POROUS BODY ASSEMBLY CODE (SPAC): Users Manual.
SSEL NUREG-0525 RO7: SAFEQUARDS
SUMMARY
EVENT LIST (SSEL), REVISION 7.
SURF NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
Embotage NUREG-0525 RO7: SAFEQUARDS
SUMMARY
EVENT LIST (SSEL),REVISIDN 7.
Safeguards Automated Facility Evaluation NUREG/CR-1246 VO4: SAFE USERS MANUAL. Volume 4: Computer Programs.
Safeguards Network Analysis NUREG/CR-3423: THE SAFEGUARDS NETWORK ANALYSIS PROCEDURE SNAP: A User 's
. Manual.
Safeguards Summary Event List NUREG-0525 RO7: SAFEGUARDS
SUMMARY
EVENT LIST (SSEL), REVISION 7.
Safety Assassments NUREC/CP.-32IS VO1: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
Safety Code Development NUREC/CR-2874 VO4: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION,GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
Safety Evaluation Report NUREG-0420 SO4: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF THE SHOREHAM NUCLEAR POWER STATION, UNIT NO.
- 1. DOCKET NO. 50-322.
l (Long Island Lighting Company)
NUREG-0519 SO5: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF LA SALLE COUNTY STATION, UNITS 1 AND 2. Docket Nos. 50-373 And 50-374.(Commonwealth Edison Company) i NUREG-0675 S16: SAFETY EVALUATION REPORT RELATED TO THE DPERATION OF DIABLO CANYDN NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)
NUREG-0675 S18: SAFETY EVALUATIDN REPORT RELATED TO THE OPERATION OF l
DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)
J NUREG-0852 802: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN OF 4
THE STANDARD NUCLEAR STEAM SUPPLY REFERENCE SYSTEM,CESSAR SYSTEM
- 80. Docket No. 50-470.(Combustion Engineering, Incorporated)
NUREG-0881 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF f
WDLF CREEK GENERATING STATION, UNIT 1. Docket No. 50-482.(Kansas Gas And Electric Compangeet al.)
S 112
NUREG-0979 B01: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN APPROVAL OF THE GESSAR II BWR/6 NUCLEAR ISLAND DESIGN. Dscket No.
50-447.(General Electric Company)
NUREG-0984: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE CORNELL UNIVERSITY TRIGA RESEARCH REACTOR. Doc k et No. 50-157.
NUREG-0984: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE CORNELL UNIVERSITY TRIGA RESEARCH REACTOR. Docket No.50-157.
NUREG-0988: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE DMAHA VETERANS ADMINISTRATION MEDICAL CENTER. Docket No. 50-131.
NUREG-0991: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LIMERICK GERNERATING STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-353.(Philadelphia Electric Company)
NUREG-0996: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE UNIVERSITY OF DKLAHOMA RESEARCH REACTOR. Doc ket No. 50-112.
NUREG-1007: SAFETY EVALUATION REPORT RELATED TO THE LICENSE RENEWAL AND POWER INCREASE FOR THE NATIONAL BUREAU OF STANDARDS REACTOR. Docket No. 50-184.
NUREG-1010: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE ZERO-POWER REACTOR AT CORNELL UNIVERSITY. Docket No. 50-97. (Cornell University)
NUR EG-1016: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE IDWA STATE UNIVERSITY. Docket No. 50-116.
Safety Functions NUREC/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
Safety Goals NUR EC/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY COALS FOR NUCLEAR POWER PLANTS.
Safety Research NUR EG/CP-OO47: TRANSACTIONS OF THE ELEVENTH WATER REACTOR SAFETY RESEARCH INFDRMATION MEETING.To Be Held At National Bureau of Standards.Gaithersburg, Maryland,0ctober 24-28,1983.
NUREG/CR-2331 VO2 N4: SAFETY RESEARCH PROGRAMS SPONSDRED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Quarterly Progress Report,0ctober
-December 31,1982.
Safety-Related Nuclear Structures NUREG/CR-3OO3 VO2: HUMAN ENGINEERING DESIGN CONSIDERATIDNS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
Safety NUREG-0485 VO5 NOB: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of August 31,1983.(Buff Book)
NUREG-0828 DRFT: ' INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - BIG ROCK POINT PLANT. Docket No. 50-155.
(Consumers Power Company)
NUREC/CR-2238 VO4: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEMBER 1981. Volume 20.
I NUR EG/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EGUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
NUREG/CR-3215 VO2: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
Salt NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
Scaling Criteria NUREC/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION 113
7 LOOP DESIGN.
Seismic Analysis NUREG/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM i-SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
-Seismic Category I Nuclear Structures NUREG/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTURAL LOADS.
Seismic Events NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK'DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT. PUMP i
SUPPORTS. Seismic Safety Margins Research Program.
Seismic Qualification NUREG/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAft i
FACILITIES WITH SEISMIC EQUIPMENT GUALIFICATION IN NUCLEAR PLANTS (A-46).
Seismic Settlement NUREG/CR-33SO: CURRENT METHODOLOGIES FOR ASSESSING SEISMICALLY INDUCED SETTLEMENTS IN SOIL.
Seismic NUREG/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EGUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
Semiscale Mod-2b NUREG/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2B POWER LOSS TEST SERIES (TEST S-PL-1,-2,AND-3).
Shipment NUREG-0725 RO3: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
Single-Phase Flow NUREG/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESIGN.
Site NUREG/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADIDACTIVE WASTE DISPOSAL FACILITY.
Slab NUREG/CR-3425: LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS.
Slits NUREG/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBC00 LED WATER THROUGH SLITS.
Small Break Transients NUREG/CR-3277: RELAP5 ASSESSMENT: SEMISCALE MOD-3 SMALL BREAK TEST.
Sodium NUREG/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH,CO32-)
RECIPROCAL SYSTEM Sonic Velocity NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION, SONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
Source NUREC/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE, Southern States Energy Board NUREG/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITY.
Spent Fuel Shipment NUREG-0725 RO3: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
Spent Unreprocessed Fuel l
NUREG/CR-2402: RISK ANALYSIG METHODOLOGY FOR SPENT FUEL REPOSITORIES IN 114
p:y T
6 BEDDED SALT. Final'Roport.
Spherical Phantoms-NUREC/CR-3425: - LOW-ENERGY PHOTON DOSE DEPOSITION IN TISSUE SLAB AN
. SPHERICAL PHANTOMS.
~ Staffing-NUREG-0985 V01: U. S.
NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.
States-NUREG/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADIDACTIVE WASTE DISPOSAL' FACILITY.
Station Blackout NUREG/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2B POWE TEST SERIES-(TEST S-PL-1,-2,AND-3).
[
~
Steady State Porous Body Assembly Code i
NUREC/CR-3188:- THE STEADY STATE POROUS BODY ASSEMBLY CODE (SPAC Manual.
Steam Explosion NUREG/CR-3306: LWR AND HTGR COOLANT DYNAMICS: THE CONTAINMENT OF SEVERE DYNAMICS.
' Steam Generator ~ Repair NUREG-1003:
DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR
{
~ REPAIR AT H.B.
ROBINSON STEAM ELECTRIC PLANT UNIT NO.
2.
Docket No.
50-261.-(Carolina Power And Light Company)
NUREC-1011:
DRAFT' ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR F-REPAIR AT THE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
[:
50-266.(Wisconsin Electric Power Company)
NUREG-1011:
DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM GENERATOR REPAIR AT THE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
50-266.(Wisconsin Electric Power Company)
Steam Generator Tubing NUREC/CR-3161:
I-EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
NUREG/CR-3200 V01: EDDY-CURRENT INSPECTIGN FOR STEAM GENERATOR TUB PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
(
Strategic Special Nuclear Material NUREG/CR-3221: MC4 A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORP'ANCE EVALUATION..
Structural Loads NUREG/CR-3315: A CONSENSUS EST1HATION STUDY OF SUCLEAR POWER PLANT 3
STRUCTURAL LOADS.
- Struetural Response NUREC/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT i
STRUCTURAL LCADS.
- Structures NUREG/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION METHODOLOGY AND
[
APPLICATION.
Subchannel Measurement NUREG/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM l
GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
l Subcooled Water i
NUREG/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER THROUG i
. SLITS.-
~
(<
NUREG/CR-2982 RO1: BUDYANCY, TRANSPORT, AND HEAD LOSS OF FIBROUS REACTOR P
INSULATION.
Support Structure
~
NUREG/CR-2005 VO4: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY
-IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1982 - December 1982.
System Codes-115 l
NUREG/CR-3329 VO1: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
System Identification Methodologies NUR EG/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION METHODOLOGY AND APPLICATION.
System Refill NUREG/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
Systematic Evaluation Program NUREG-0485 VOS NO6: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of June 30,1983.(Buff Book)
NUREG-0485 VO5 NO7: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of July 31,1983.(Buff Book).
NUREG-0485 VO5 NOB: SYSTEMATIC EVALUATIGN PROGRAM STATUS
SUMMARY
REPORT. Data As Of August 31,1983.(Buff Book)
NUREG-OS21 SO1: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - R.
E.
GINNA NUCLEAR POWER PLANT. Docket No.
50-244. (Rochester Gas And Electric Corporation)
NUREG-0828 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT REPORT SYSTEMATIC EVALUATION PROGRAM - BIG ROCK. POINT PLANT. Doc ket No. 50-155.
(Consumers Power Company)
Systems Codes NUREG/ CR-3257: RELAPS ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
TLD NUREG-0837 VO2 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, October - December 1982.
NUREG/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EQUIVALENT AT NUCLEAR POWER PLANTS.
TRAC-PD2 NUREG/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1981.
TRAC NUREG/CR-2573: ' BWR REFILL-RELOAD PRDORAM TASK 4. 7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
NUREC/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT 4
TRAC-BWR COMPONENT MODELS.
NUREG/CR-32SO: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
NUREQ/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 38).
Task Analysis i
NUREQ/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NUREG/CR-3371 VO2: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM l
C REWS.' Volume 2: Data Results.
Tearing Instability NOREG/CR-3464: THE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING l
CIRCUMFERENTIAL THROUGH WALL CRACKS.
Thermal-Hydraulic l
NUREG/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
NUREG/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4. 7 - MODEL DEVELOPPENT TRAC-BWR COMPONENT MODELS.
I NUREG/CR-2575: BWR FULL INTEGRAL SIMULATIDN TEST (FIST) PROGRAM TEST PLAN.
NUREG/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
NUREG/CR-3214:
SUMMARY
OF NUCLEAR REGULATORY COMMISSION'S LDFT PROGRAM l
EXPERIMENTS.
NUREG/CR-3257: RELAP5 ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
116
7 NURED/CR-3258: RELAP5 ASSESSMENT: SEMISCALE NATURAL CIRCULATION TESTS S-NC-2 AND S-NC-7.
NUREQ/CR-3277: RELAPS ASSESSMENT: SEMISCALE MOD-3 SMALL BREAK TEST.
NUREG/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
NUREQ/CR-3329 V01: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
NOREQ/CR-3375: THE DEVELOPMENT OF NUFREG-N AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
NUREQ/CR-3359 VO1: PHYSICS OF REACTOR SAFETY.Guarterly Report, January -
March 1983.
Thermoluminescent Dosimeter NUREG-OB37 V02 NO4: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, October-- December 1982.
TiCode-12 NUREQ/CR-3282: INTERNAL HYDROGEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
Title List i
NUREG-0540 V05 NO4: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. April-I-30, 1983.
NUREG-0540 VOS NO7: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
July 1-31,1983.
Topical Report Reviews NUREG-0390 V07 N01: TOPICAL REPORT REVIEW STATUS.(Blue Book)
Toxicity NUREG/CR-3173: THE SENSITIVITY OF ADULT, EMBRYONIC AND LARVAL CR' ' FISH PROCAMBAROJ CLARKII TO COPPER.
Training NUREG-0985 VO1:
U. S.
NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.
NUREQ/CR-3371 V01: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NUREQ/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
Transient Reactor Analysis Code NUREQ/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4.7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF' TRAC.
NUREC/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT TRAC-BWR COMPONENT MODELS.
NUREG/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
Transient Reactor Analysis NUREG/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATIDA MODEL TEST C1-19 (RUN 38).
Transient NUREG/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS OF OXIDE FUEL RODS.
NUREG/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TEST PLAN.
NUREQ/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
NUREG/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
NUR EQ/CR-3241: PROBABILISTIC METHODS FOR IEENTIFICATION OF SIGNIFICANT ACCIDENT SEGUENCES IN LOOP-TYPE LMFBRS.
l Transportation NUREG-0725 R03: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
Trench Covers
- NUREQ/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
117
Tube Support NUR EG/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING
. PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
Turbine Trip NUREG/CR-3257: RELAPS ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
Two-Phase Flow NUREQ/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVEPTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESIGN.
Two-Phase NUREG/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITOMETER AND ITS APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
UPL NUR EC/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
USI A-46 NUREG-1018: SEISMIC QUALIFICATION OF EGUIPMENT IN OPERATING PLANTS.
Status Re' port, Unresolved Safety Issue A-46.
Uranium Mill Tailings Ponds NUREG/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
Uranium Mill Tailings NUREC/CR-2938: TAILINGS TREATMENT TECHNIGUES FOR URANIUM MILL WASTE: A Review Of Existing Information.
NUREG/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH SYGTEM.Model Formulations And Preliminary Results.
VISA NUR EC/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
Valve NUR EG/CR-3048: CLOSEOUT OF IE BULLETIN 80-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
NUREG/CR-3050: CLOSEOUT OF IE BULLETIN 81-02 AND SUPPLEMENT: Failure Of Gate-Type Valves To Close Against Differential Pressure.
Vendor NUREG-0040 VO7'NO2: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS R EPORT. Guarterly Report, April 1983 - June 1983.(White Book)
Vessel Integrity NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
Waste Disposal Sites NUREG/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
Waste Disposal NUREG/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEA7CH AND TEST REACTORE, Sensitivity Of Decommissioning Radiation Exposcre And Costs To S,jlected Parameters.
NUREG/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITY.
Waste Form Degradation NUREG/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
Waste Package NUREC/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
Waste Package / Repository NUREC/CR-3091 VO2: REVIEW OF WASTE PACKAGE VERIFICATION 118
r j;
TESTS. Semiannual Report Covering The Period October 1982 - March 1983.
Waste Packages l
NUREG/CR-3405 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, March 1982 - April 1983.
Waste NUREG/CR-2482 VO4: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask
- 1. 1 -
National Waste Package Program,0ctober 1982 - March 1983.
NUREG/CR-3381: EVALUATIDN OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
NUREC/CR-0130 ADDO2: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Ef fects On Decommissioning Of Interim Inability To Dispose Of Wastes Offsite.
NUREC/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR POWER STATION. Effects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
NUREG/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
Water Hammer NUREG-OO90 VO6 NO1: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. January-March 1983.
Wave Dispersion NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION, SONIC VELDCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
Weld NUREG/CR-3265: EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS.
Whole Body Irradiation NUREG/CR-3362: THE INFLUENCE OF LYMPHOHEMATOPOIETIC DISEASE IN ALTERING RADIATION INDUCED CANCER SENSITIVITY IN MICE.
korkers NUREG/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EGUIVALENT AT NUCLEAR POWER PLANTS.
Yokes NUREG/CR-3048: CLOSEOUT OF IE BULLETIN 80-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
l 6
t 119 i
i l
NRC Originating Organization Index (Staff Reports)
This index l!sts 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 NUREC-0837 VO2 N04: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, October - December 1982.
REGION 4, OFFICE OF DIRECTOR NUREG-0040 V07 NO2: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report, April 1983 - June 1983.(White Book)
EDO - OFFICE OF ADMINISTRATION DIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL NUREG-0304 V0O NO2: REGULATORY AND TECHNICAL REPORTS. Compilation Fcr
~
Second Guarter 1983.
NUREG-0540 VOS N04: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. April 1-30, 1983.
NUREG-0540 V05 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
May 1-31,1983.
NUREG-0540 VOS N06: TITLE LIST CF DOCUMENTS MADE PUBLICLY AVAILABLE.
June 1-30,1983.
NUREG-0540 VOS N07: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.
July 1-31,1983.
DIVISION OF RULES AND RECORDS NUREG-0936 V02 NO2: NRC REGULATORY AGENDA.Guarterly Report, March -
May 1983.
EDO - OFFICE FOR ANALYSIS & EVALUATION OF GPERATIONAL DATA DIRECTOR'S OFFICE NUREG-0090 V06 N01: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. January-March 1983.
NUREG-1022: LICENSEE EVENT REPORT SYSTEM. Description Of System And Guidelines For Reporting.
I l-l OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) 1
\\
l l
121
DIRECTOR'S OFFICE. OFFICE OF INSPECTION AND ENFORCEMENT NUREG-0430 VO3 NO2: LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. July - December 1982.
ENFORCEMENT STAFF NUREQ-0940 VO2 NO2: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED.Ouartarig Progrest Report April-June 1983.
OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEQUARDS OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDM. DIRECTOR NUREG-0725 RO3: PUBLIC INFORMATION CIRCULAR FOR SHIPMENTS OF IRRADIATED REACTOR FUEL.
NUREC-0925: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF THE TETON URANIUM ISL PROJECT. Docket No. 40-8781. (Teton Explcration Drilling,Inc.)
DIVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG-C383 VO1 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Summary Report Of NRC Approved Packages.
NUREG-0383 VO2 R06: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of Compliance.
NUREO-0383 VO3 RO3: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKACES. Summary Report Of NRC Approved Ouality Assurance Programs For Radioactive Material Packages.
DIVISION OF SAFEQUARDS NUREO-0525 RO7: SAFECUARDS
SUMMARY
EVENT LIST (SSEL), REVISION 7.
U. S.
NUCLEAR REQULATORY COMMISSION NRC - NO DETAILED AFFILIATION CIVEN NUREQ/CR-3447 VO1: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Dasic Methods.
OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)
OFFICE OF NUCLEAR REQULATORY RESEARCH, DIRECTOR NUREQ/CP-OO47: TRANSACTIONS OF THE ELEVENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.To Be Held At National Bureau of Standards,Gaithersburg, Maryland,0ctober 24-28,1983.
EDO-RESOURCE MANAGEMENT OFFICE OF RESOURCE MANAGEMENT, DIRECTOR NUREO-0485 VOS N06: SYSTEMATIC EVALUATION PROGRAM UTATUS
SUMMARY
REPORT. Data As Of June 30,1983.(Buff Book)
NUREG-0485 VOS NO7: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As 09 July 31,1983. (Buf f Book).
NUREG-0485 VO5 NOB: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of August 31,1983.(BuPf Book)
NUREC-0998:
U. S.
NUCLEAR REGULATORY COMMISSION 1982 ANNUAL REPORT.
MANAGEMENT INFORMATION BRANCH NUREG-OO2O VO7 NO3: LICENSED OPERATINQ REACTORS STATUS
SUMMARY
122
l REPORT. Data As Of February 28,1983.(Greg Book)
NUREG-OO20 V07 NO4: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of March 31,1983. (Greg Book)
NUREG-0020 V07 N05: LICEN3ED OPERATING REACTORB STATUS
SUMMARY
REPORT. Data As Of April 30,1983.(Greg Book)
NUREG-0020 V07 N06: LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Data As Of May 31,1983.(Greg Book)
NUREG-0390 V07 N01: TOPICAL REPORT REVIEW STATUS.(Blue Book)
NUREG-0500 V12 N06: REOULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of June 30,1983.(Blue Book)
NUREG-0580 V12 N07: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As Of July 31,1983 (Blue Book)
NUREG-0580 V12 N08: REGULATORY LICENSING STATUS
SUMMARY
REPORT. Data As of August 31,1993 (Blue Book)
NUREG-0606 V05 NO3: UNRESOLVED SAFETY ISSUES
SUMMARY
. Data As Of August 19,1983. (Aqua Book).
NUREG-0748 V03 N06: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of Juno 30,1983.(Orange Book)
NUREG-0748 V03 N07: OPERATING REACTORS LICENSING ACTIONS
SUMMARY
. Data As Of July 31.1983.(Orange Book)
NUREG-0871 V02 NO3:
SUMMARY
INFORMATION REPORT. Data As Of June 30,1983.(Brown Book)
DFFICE OF NUCLEAR REACTOR REGULATION (PDST 4/28/90)
OFFICE OF NUCLEAR REACTOR REOULATION, DIRECTOR NUREG-0485 V05 N06: SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT. Data As Of June 30,1983.(Buff Book)
NUREG-0852 802: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN 4
OF THE STANDARD NUCLEAR STEAM SUPPLY REFERENCE SYSTEM,CESSAR SYSTEM
- 90. Docket No. 50-470.(Combustion Engineering, Incorporated)
NUREG-0988: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF TE OPERATING LICENSE FOR THE RESEARCH REACTOR AT THE OMAHA VETERANS ADMINISTRATION PEEDICAL CENTER. Docket No. 50-131.
NUREG-1000 V02: GENERIC IMPLICATIONS OF ATWS EVENTS AT THE SALEM i
NUCLEAR POWER PLANT. Licensee And Staff Actions.
NUREG-1020 V01: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFliCT ON TMI-1. General Public Utilities Corporation,et al V.The Babcock &
Wilcox Company,et al.Three Mile Island Nuclear Station, Unit 1.
Docket No. 50-289.
NUREG-1020 V02: GPU V.
B&W LAWSUIT REVIEW AND ITS EFFECT ON TMI-1. General Public Utilities Corporation,et al V.
The Babcock &
Wilcor Company,et al.Three Mile Island Nuclear Utation, Unit 1.
Docket No.50-289.
DIVISION OF HUMAN FACTORS SAFETY NUREG-0985 V01:
U. S.
NUCLEAR REOULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.
DIVISION OF LICENGING NUREG-0420 SO4: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF THE SHOREHAM NUCLEAR POWER STATION, UNIT NO.
- 1. DOCKET NO. 50-322.
(Long Island Ligh ting Company's i
NUREG-0519 805: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LA SALLE COUNTY STATION, UNITS 1 AND 2. Docket Nos. 50-373 And 50-374.(Commonwealth Edison Company)
NUREG-0675 S16: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos.'50-275 And 50-323.(Pacific Gas And Electric Company)
NUREG-0675 S18: BAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 123
m And 50-323.(Pacific Gas And Electric Company)
NUREG-0821 801: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - R.
E.
GINNA NUCLEAR POWER PLANT.Doeket No.
50-244.(Rochester Oss And Electric Corporation)
NUREG-OSS1 803: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WOLF CREEK GENERATING STATION UNIT 1. Docket No. 50-482.(Kansas Gas And Electric Company,et al.)
NUREG-0979 801: SAFETY EVALUATION REPORT ~ RELATED TO THE FINAL DESIGN APPROVAL OF TE GESSAR II BWR/6 NUCLEAR ISLAND DESION. Docket No.
50-447.(General Electric Company)
NUREG-0984: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE CORNELL UNIVERSITY TRIGA RESEARCH REACTOR. Docket No.50-157.
NUREG-0991: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF LIPERICK GERERATING STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-353.(Philadelphia Electric Company)
NUREG-0996: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE UNIVERSITY OF OKLAHOMA RESEARCH REACTOR. Docket No. 50-112.
NUREG-1003: DRAFT ENVIRONMENTAL STATEMENT RELATED TO STEAM OENERATOR REPAIR AT H.B.
ROBINSON STEAM ELECTRIC PLANT UNIT NO.
2.
Docket No.
50-261.(Carolina Power And Light Company)
NUREG-1010: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF TE OPERATING LICENSE FOR THE ZERG-POWER REACTOR AT CORNELL UNIVERSITY. Docket No. 50-97. (Cornell University)
NUREG-1011: DRAFT ENVIRONMENTAL STATEMENT RELATED TO OTEAM OENERATOR REPAIR AT TE POINT BEACH NUCLEAR PLANT UNIT NO.
1.
Docket No.
50-266.(Wisconsin Electric Power Company)
NUREG-1011: FINAL ENVIRONMENTAL STATEMENT RELATE') TO STEAM OENERATOR REPAIR AT THE POINT BEACH NUCLEAR PLANT, UNIT 1. Docket No.
50-266.(Wisconsin Electric Power Company)
NUREG-1016: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LIC2NSE FOR THE RESEARCH REACTOR AT THE IOWA STATE UNIVERSITY.Dacket No. 50-116.
DIVISION OF SAFETY TECHNOLOGY NUREG-1018: BEIBMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS.
Status Report, Unresolved Safety Issue A-46.
1 l
t 124
i 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 than 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 organir.ation, if furtner information is needed, refer to the main citation by the NUREG/CR number.
EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA DIRECTOR'S OFFICE NUREG/CR-2000 V02 N6: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of June 1983.
NUREG/CR -2000 V02 N7: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of July 1983.
NUREG/CR-2000 V02 NB: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of August 1983.
NUREG/CR-3122: PO1ENTIAL DAMAGING FAILURE MODES OF HIGH-AND MEDIUM-VOLTAGE ELECTRICAL EGUIPMENT.
OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)
DIVISION OF EMERGENCY PREPAREDNESS & ENGINEERING RESPONSE (POST 830103)
NUREG/CR-3048: CLOSEOUT OF IE BULLETIN 80-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
NUREG/CR-3050: CLOSEOUT OF IE BULLETIN 81-02 AND SUPPLEMENT: Failure Of Gate-Type Valves To close Against Differential Pressure.
DIVISION OF GA, SAFEGUARDS & INSPECTION PROGRAMS (POST 830103) i NUREG/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EQUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS DIVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGINEERING BURIAL SITE HEMATITE MISSOURI.
NUREO/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT DURIAL SITE.
DIVISION OF WASTE MANAGEMENT NUREG/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratory Evaluation And Computer Modeling of Trench Cover Design.
NUREG/CR-2482 V04: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Waste Package Program,0ctober 1982 - March 1983.
125
v i
l NUREG/CR-3091 VO2: REVIEW OF WASTE PACKAGE VERIFICATION TESTS. Semiannual Report Covering The Period October 1982 - March 1983.
NUREQ/CR-3219 VO1: DRAFT TECHNICAL POSITION SUBTASK 1.1: WASTE PACKAGE PERFORMANCE AFTER REPOSITORY CLOSURE.
NUREC/CR-3259: LEAK DETECTION SYSTJMS FOR URANIUM MILL TAILINGS IMPOUNDMENTS WITH SYNTHETIC LINERS.
NUREC/CR-3381: EVALUATION QF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
NUREQ/CR-3451: BENCHMARK PROBLEMS FOR RADIOLOGICAL ASSESSMENT CODES.
OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)
DIVISION OF ACCIDENT EVALUATION NUREG/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TRANSIENT ANALYSIS OF OXIDE FUEL RODS.
NUREQ/CR-2238 V04: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEMBER 1981. Volume 20.
NURE0/CR-2331 V02 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR RECULATORY RESEARCH.Guarterly Progress Report,0ctober
-December 31.1982.
NUREQ/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4. 7 - MODEL DEVELOPMENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
NUREC/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4.7 - MODEL DEVELOPMENT TRAC-BWR COMPONENT MODELS.
NUREQ/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TEST PLAN.
NUREQ/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
NUREG/CR-2844 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVIS1GN OF ACCIDENT EVALUATION,GUARTERLY PROGRES5 REPORT, OCTOBER 1 - DECFMBER 31,1982.
NUREG/CR-3060: A CaITICAL REVIEW OF FLOODING LITERATURE.
NUREG/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTO.1.
NUREG/CR-3138: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FOR A PWR GEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
NUREQ/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT BNL IN FY 1981.
NUREG/CR-3188: THE STEADY STATE POROUS BODY ASSEMBLY CODE (SPAC): Users Manual.
NUREQ/CR-3214:
SUMMARY
OF NUCLEAR REQULATORY COMMISSION'S LOFT PROGR AM EXPERIMENTS.
NUREQ/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
.I NUREQ/CR-3238: PNASIM: A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIGUE OF MASS FLOW MEASUREMENT.
NUREC/CR-3241: PROBABILISTIC METHODS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SEGUENCES IN LOOP-TYPE LMFBRS.
NUREQ/CR-3257: RELAPS ASSESSMENT: LOFT TURBINE TRIP L6-7/L9-2.
NURC3/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH,CO32-)
RECIPROCAL SYSTEM NUREQ/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The Fourth Materials Experiment (MT-4).
NUREG/CR-3277: RELAPS ASSESSMENT: SEMISCALE MOD-3 SMALL BREAK TEST.
NUREG/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
NUREG/CR-3307 V01: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report, January -March 1983.
NUREO/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC SIMULATION.
NUREQ/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 38).
NUREQ/CR-3359 V01: PHYSICS OF REACTOR SAFETY Guarterly Report. January
{
126
9
- March 1983.
NUREG/CR-3372: AN ANALYSIS OF WAVE DISPERSION, SON 1s VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
NUREG/CR-3373: AIR / WATER SUBCHANNEL MEASUREMENTS OF THE EQUILIBRIUM GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
NUREG/CR-3374: THE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITONETER AND ITS APPLICATION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
NUREG/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR TIE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN BOILING WATER NUCLEAR REACTORS.
NUREG/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EMERGENCY CORE COOLING IN A BWR/4.
NUREG/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
NUREG/CR-3401: LABORATORY-SCALE SODIUM-CARBONATE AGOREGATE CONCRETE INTER ACTIONS.
NUREG/CR-3419: EXPERIMENT DATA REPORT FOR SEMISCALE MOD-2B POWER LOGS TEST SERIES (TEST S-PL-1,-2,AND-3).
NUREG/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESIGN.
NUREG/CR-3434: TIME AVERAGING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EQUATIONS OF MULTIPHASE FLOW.
NUREG/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER THROUGH SLITS.
ANALYTICAL MODELS BRANCH NUREG/CR-3258: RELAP5 ASSESSMENT: SEMISCALE NATURAL CIRCULATION TESTS S-NC-2 AND S-NC-7.
DIVISION OF FACILITY OPERATIONS NUREG/CR-2254: A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS.
NUREG/CR-2331 VO2 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Guarterly Progress Report, Octaber
-December 31,1982.
NUREG/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
NUREG/CR-3003 V02: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GENERATED DISPLAYS.
NUREG/CR-3118 V01: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 1: Executive Summary.
NUREG/CR-3118 V02: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO
' HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 2: Supporting Documentation And Review Of Epidemiologic Studies Of Specific Chemicals.
NUREG/CR-3118 V03: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO
' HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume '3: Abstracts of NIOSH Criteria Documents.
NUREG/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
NUREG/CR-3196: DRUG AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
NUREG/CR-3215 VO1: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
NUREG/CR-3215 V02: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
NUREG/CR-3221: MC8 A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVALUATION.
NUREG/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIQUES.
127
.V NUREQ/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
NUREQ/CR-3371 VO1: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
NUREQ/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
NUREQ/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
NUREQ/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EGUIVALENT AT NUCLEAR POWER PLANTS.
NUREQ/CR-3423: THE SAFEQUARDS NETWORK ANALYSIS PROCEDURE SNAP: A User's Manual.
NUREQ/CR-3425: LOW-ENERQY PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS.
DIVISION OF HEALTH, SITING & WASTE MANAGEMENT-HUREQ/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES -GROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEQUENCES.
NUREQ/CR-2938: TAILINGS TREATMENT TECHNIGUES FOR URANIUM MILL WASTE:
A Review Of Existing Information.
NUREQ/CR-3111 Vol: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
NUREQ/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM THE H.B.
ROBINSON STEAM ELECTRIC PLANT.
NUREQ/CR-3133: THE SENSITIVITY OF ADULT, EMBRYONIC AND LARVAL CRAYFISH PROCAMBARUS CLARKII TO COPPER.
NUREQ/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
NUREQ/CR-3282: INTERNAL HYDROGEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
NUREQ/CR-3327: ANALYSIS OF STRONG-MOTION DATA FROM THE NEW HAMPSHIRE EARTHGUAKE OF 18 JANUARY 1982.
NUREQ/CR-3362: THE INFLUENCE OF LYMPH 0 HEMATOPOIETIC DISEASE IN ALTERING RADIATION INDUCED CANCER SENSITIVITY IN MICE.
NUREQ/CR-3364: PRESENCE OF PATHOGENIC MICRODRGANISMS IN POWER PLANT COOLING WATERS.
NUREQ/CR-3380: CURRENT METHQDOLOGIES FOR ASSESSING SEISMICALLY INDUCED SETTLEMENTS IN SOIL.
NUREQ/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition.
NUREQ/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH SYSTEM.Model Formulations And Preliminary Results.
^
NUREQ/CR-3405 V01: 'LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, March 1982 - April 1983.
NUREQ/CR-3427 VO1: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
NUREQ/CR-3472 V01: SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS.
NUREQ/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
NUREQ/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVEL RADI0 ACTIVE WASTE DISPOSAL FACILITY.
DIVISION OF RISK ANALYSIS NUREQ/CR-1246 V04: SAFE USERS MANUAL. Volume 4: Computer Programs.
NUREQ/CR-2146 V02: DYNAMIC ANALYSIS TO ESTADLISH NORMAL SHOCK AND VIBRATION OF RADIDACTIVE MATERI AL SHIPPING PACKAGES: Guarterly Progress Report.Apr:1 1,1981 - June 30,1981.
128
l NURE0/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN BEDDED SALT. Final Report.
NUREG/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A-REVIEW OF THE LITERATURE.
NURE9/CR-2739: DATA BASE FOR BASALT METHODOLOGY.
NURE0/CR-2SL5: REVIEW OF MODELS APPLICABLE TO ACCIDENT AEROSOLS.
NURE0/CR-3085 V02: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF TE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol II - Appendix A And B.1 Thru B.4. Docket No. 50-245.(Northeast Nuclear Energy Company)
NUREG/CR-3085 V04: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE' POINT UNIT 1 NUCLEAR POWER PLANT.Vol IV Appendix B.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
NUREG/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN STATIC AIR.
NURE9/CR-3175: INSTITUTIONAL I:1PLICATIONS OF EFTABLISHING SAFETY GOALS FOR NUCLEAR POWER PLANTS.
NUREG/CR-3289: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES. Estimates Based on Licensee Event Reports At U. S.
Commercial Nuclear Power Plants, 1976-1991.
NUREO/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
NURE9/CR-3437: DATA ANALYSIS USING BIONOMIAL FAILURE RATE COMMON CAUSE MODEL.
NUREG/CR-3447 V01: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Methods.
DIVISION OF ENGIEERING TECHNOLOGY NURE9/CR-0130 ADD 02: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Effects On Decommissioning Of Interim Inability To Dispose Of Wastes Offsite.
NURE9/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE BOILING WATER REACTOR POWER STATION. Effects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
NUREG/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameters.
NURE0/CR-2331 V02 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report,0ctober
-Deeember 31,1982.
NUREO/CR-2805 V04: LWR PRESSURE VEBSEL SURVEILLANCE DOSIMETRY IMPROVEMENT PROGRAM.Guarterly Progress Report,0ctober 1992 -
December 1982.
NUREO/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE YESSEL WITH CLADDING ON THE INNER SURFACE.
NURE0/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
NUREG/CR-31SO: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
NUREG/CR-3200 V01: EDDY-CURRENT INSPECTION FOR STEAM OENERATOR TUBINA PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1993.
NUREG/CR-3234: THE POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH ELECTRICAL PENETRATION ASSEMBLIES UNDER SEVERE ACCIDENT CONDITIONS.
NURE0/CR-3265: EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS.
NURE0/CR-3283: FIRST EXCURSION PROBLEMS FDR QAUSSIAN VECTOR PROCESSES.
NURE9/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS NUREG/CR-3307 V01: REACTOR SAFETY RESEARCH PROGRAMS. Quarter 1g Report, January -March 1983.
129
m i
O NUREG/CR-3315: A CONSENSUB ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTURAL LDADS.
j NURE0/CR-3333: NEUTRON SPECTRAL CHARACTERIZATION OF THE FOURTH NUCLEAR REGULATORY COMMISSION HEAVY SECTION STEEL TECHNOLOGY 1T-CT i
IRRADIATION EXPERIMENTS: Dosimetry And Uncertainty Analysis.
j NURE0/CR-3334 VO1: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983.
NURE0/CR-3336: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Summary Repor4,Ames-Laboratorg Research Reactor. Docket No. 50-116.
NURE0/CR-3341: PROBABILISTIC DESCRIPTIONS OF RESISTANCE OF SAFETY-RELATED NUCLEAR STRUCTURES.
NUREG/CR-3370: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS
SUMMARY
REPORT NORTH CAROLINA STATE UNIVERSITY RESEARCH AND TRAINING i
REACTOR.
i NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY 0F REACTOR PRESSURE VESSEL FAILURE.
NUREO/CR-3388: EVALUATION OF SYSTEM IDENTIFICATION METHODOLOGY AND APPLICATION.
NUREG/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
EDO-RESOURCE MANAGEMENT DIVISION OF DATA AUTOMATION h MANAGEMENT INFORMATION NURE0/CR-2850 VO2: POPULATION DOSE COMMITMENTS DUE TO RADIDACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
OFFICE OF NUCLEAR REACTOR REQULATION (POST 4/28/80)
DIVISION OF ENGINEERING i
NUREO/CR-3266: SEISMIC h DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
NURE9/CR-3464: THE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUPFERENTIAL THROUGH WALL CRACKS.
DIVISION OF SYSTEMS INTEGRATION (POST 811005)
NUREG/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR l'
ACCIDENTS.
NUREO/CR-3270: INVESTIGATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS.
NURE0/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
j NUREG/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Report, April 1980 - April 1983.
NUREG/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAINMENT OF j
SEVERE DYNAMICS.
ACCIDENT EVALUATION BRANCH t
(
NUREO/CR-3329 VO1: THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
j NUREG/CR-3416 VO1: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS i
CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
NUREO/CR-3416 VO2: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
DIVISION OF SAFETY TECHNOLOGY NURE0/CR-2982 RO1: BUOYANCY, TRANSPORT,AND HEAD LOSS OF FIBROUS REACTOR INSULATION.
NUREG/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
130
NUREO/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAR FACILITIES WITH SEIBMIC EQUIPMENT GUALIFICATION IN NUCLEAR PLANTS (A-46).
NUREG/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEGRADATION OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS. Seismic Safety Margins Research Program.
NUREG/CR-3394 V01: PROBABILISTIC AESESSMENT OF RECIRCULATION SUMP BLOCMAGE DUE TO LOSS OF COOLANT ACCIDENTS. Executive Summary And Significant Findings.
NUREO/CR-3394 V02: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCMAGE DUE TO LOSS OF COOLANT ACCIDENTS. Compendium Of Calculated Break Probabilities, Sump Blockage Frequencies,And Calculational Program Listing.
1; i
l f
131
Contractor index l
l l
This index lists, in alphabetical order, the contractors that prepared the l
NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports, if further information is needed, refer to the main citation by the NUREG/CR number.
ALDEN RESEARCH LABORATORY NUREQ/CR-2982 RO1: BUOVAP!CY, TRANSPORT, AND HEAD LOSS OF FIBROUS REACTOR INSULATION.
ANCO TESTING LABORATORY, INC.
NUREQ/CR-3180: PIPE DAMPING STUDIES AND NONLINEAR PIPE BENCHMARKS FROM SNAPBACK TESTS AT HEISSDAMPFREAKTOR.
APPLIED SCIENCE ASSOCIATES, INC.
NUREC/CR-3299: CORE MELT MATERIALS INTERACTIONS EVALUATIONS. Final Repert. April 1980 - April 1983.
ARGONNE NATIONAL LABORATORY NUR EQ/CR-3238: PNASIN:A PROGRAM FOR SIMULATION OF THE PULSED-NEUTRON ACTIVATION TECHNIGUE OF MASS FLOW MEASUREMENT.
NUREQ/CR-3339: AN EXPERIMENTAL STUDY OF INVERTED ANNULAR FLOW HYDRODYNAMICS UTILIZING AN ADIABATIC SIMULATION.
NUREQ/CR-3359 V01: PHYSICS OF REACTOR SAFETY.Guarterly Report, January -
March 1983.
NUREQ/CR-3308: EVALUATION OF SYSTEM IDENTIFICATION METHODOLOGY AND APPLICATION.
NUREQ/CR-3420: SCALING CRITERIA FOR TWO-PHASE FLOW NATURAL AND FORCED CONVECTION LOOP AND THEIR APPLICATION TO CONCEPTUAL 2X4 SIMULATION LOOP DESION.
NUR EQ/CR-3434: TIME AVERAQING OF LOCAL VOCAL VOLUME-AVERAGED CONSERVATION EGUATIONS OF MULTIPHASE FLOW.
ARIZONA, UNIV. OF, TUCSON, AZ NUR EQ/CR-3213: PRESSURE TESTING OF FRACTURED ROCKS - A METHODOLOGY EMPLOYING THREE-DIMENSIONAL CROSS-HOLE TESTS.
l NUR EQ/CR-3473: ROCK MASS SEALING-EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1982 - May 1983.
ARMY, DEPT. OF, ARMY ENGINEER WATERWAYS EXPERIMENT STATION NUREQ/CR-3327: ANALYSIS OF STRONC-M0'(ION DATA FROM THE NEW HAMPSHIRE EARTHOUAKE OF 18 JANUARY 1982.
NUREQ/CR-3380: CURRENT NETHODOLOGIES FOR ASSESSING SEISMICALLY INDUCED SETTLEMENTS IN SOIL.
BATTELLE HUMAN AFFAIRS RESEARCH CENTERS I
NUREQ/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY GOALS FOR NUCLEAR POWER PLANTS.
NUREQ/CR-3196: DRUQ AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES NUREQ/CR-3118 VO1: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS 133
CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 1: Executive Summary.
NUREG/CR-3118 V02: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 2: Supporting Documentetion And Review Of Epidemiologic Studies Of Specific Chemicals.
NUREQ/CR-3118 V03: HEALTH EFFECTS OF OCCUPATIONAL EXPOSURE TO HAZARDOUS CHEMICALS: A COMPARATIVE ASSESSMENT WITH NOTES ON IONIZING RADIATION. Volume 3: Abstracts of NIOSH Criteria Documents.
NUR EQ/CR-3385: MEASURES OF RISK IMPORTANCE AND THEIR APPLICATIONS.
NUREQ/CR-3405 V01: LONG-TERM PERFORMANCE OF MATERI ALS USED FOR HIGH-LEVdL WASTE PACKACING. Annual Report, March 1982 - April 1983.
NUREG/CR-3427 V01: LONQ-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING GUARTERLY REPORT APRIL - JUNE 1983.
NUREQ/CR-3447 V01: RESEARCH PRIORITIZATION USING THE ANALYTIC HIERARCHY PROCESS. Volume 1: Basic Nethods.
BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORY
=
NUREQ/CR-0130 ADD 02: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE PRESSURIZED WATER REACTOR POWER STATION. Effects On Decommissiening Of Interim Inability To Dispose Of Wastes Offsite.
NUREQ/CR-0672 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING A REFERENCE DOILING WATER REACTOR POWER STATION. Effects On Decommissioning of Interim Inability To Dispose Of Wastes Offsite.
NUREQ/CR-1756: TECHNOLOGY, SAFETY AND COSTS OF DECOMMISSIONING REFERENCE NUCLEAR RESEARCH AND TEST REACTORS. Sensitivity Of Decommissioning Radiation Exposure And Costs To Selected Parameterr.
NUREQ/CR-2658: CHARACTERISTICS OF COMBUSTION PRODUCTS: A REVIEW OF THE LITERATURE.
NUREC/CR-2835: REVIEW OF MODELS APPLICABLE TO ACCIDENT AEROSOLS.
NUREC/CR-2850 V02: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1980.
NUREC/CR-2938: TAILINGS TREATMENT TECHNIGUES FOR URANIUM MILL WASTE: A Review Of Existing Information.
=
NUR EQ/CR-3083: EXPERIMENT OPERATIONS PLAN FOR THE TH-2 EXPERIMENT IN THE NRU REACTOR.
]
NUREC/CR-3093: AEROSOLS GENERATED BY RELEASES OF PRESSURIZED POWDERS AND SOLUTIONS IN GTATIC AIR.
NUREC/CR-3175: INSTITUTIONAL IMPLICATIONS OF ESTABLISHING SAFETY COALS
=
FOR NUCLEAR POWER PLANTS.
=
NUR EO/CR-3196: DRUG AND ALCOHOL ABUSE: THE BASES FOR EMPLOYEE ASSISTANCE
]
PROGRAMS IN THE NUCLEAR UTILITY INDUSTRY.
NUREQ/CR-3215 V01: ORGANIZATIONAL ANALYSIS AND SAFETY FOR UTILITIES WITH NUCLEAR POWER PLANTS.An Organizational Overview.
2 NUREQ/CR-3215 V02: ORGANIZATIONAL ANALYSIS AND SAFETY UTILITIES WITH NUCLEAR POWER PLANT. Perspectives For Organizational Assessment.
j NUREC/CR-3259: LEAK DETECTION SYSTEMS FOR URANIUM MILL TAILINGS 5
IMPOUNDMENTS WITH SYNTHETIC LINERS.
NUREQ/CR-3272: LOCA SIMULATION IN NRU PROGRAM. Data Report For The B
NUREO/CR-3307 V01:
Fourth Materials Experiment (MT-4).
REACTOR SAFETY RESEARCH PROGRAMS.Ouarterly Report. January -March 1983.
NUREG/CR-3384: VISA - A COMPUTER CODE FOR PREDICTING THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.
NURE0/CR-3404: PREDICTIVE GEOCHEMICAL MODELING OF INTERACTIONS BETWEEN j
URANIUM MILL TAILINGS SOLUTIONS AND SEDIMENTS IN A FLOW-THROUGH a
SYSTEM.Model Formulations And Preliminary Results.
BIOTECHNOLOGY, INC.
d NUREG/CR-3371 V01: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1 Project Approach And Methodology.
2 NUREG/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM E
_?
[
134
=
I CREWB. Volume 2: Data Results.
SROOKHAVEN NATIONAL LABORATORY NURE0/CR-2331 V02 N4: SAFETY RESEARCH PROGRANS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report,0ctober
-December 31,1982.
NURE9/CR-2475: HYDROGEN COMBUSTION CHARACTERISTICS RELATED TO REACTOR ACCIDENTS.
NUREG/CR-2482 V04: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 -
National Weste Package Program,0ctober 1982 - March 1983.
NUREG/CR-3091 V02: REVIEW OF WASTE PACKAGE VERIFICATION TESTS. Semiannual Report Covering The Period Ocsober 1982 - March 1983.
NUREG/CR-3148: INDEPENDENT ASSESSMENT OF TRAC-PD2 AND RELAP5/ MOD 1 CODES AT Bft. IN FY 1981.
NURE0/CR-31BS: THE STEADY STATE POROUS BODY ASSEMBLY CODE (SPAC): Users Manual.
1 NUREO/CR-3219 V01: DRAFT TECWICAL POSITION SUBTASK 1.1: WASTE PACKAGE PERFORMANCE AFTER REPOSITORY CLOSURE.
l' NUREG/CR-3241: PROBABILISTIC METHODS FOR IDENTIFICATION OF SIGNIFICANT ACCIDENT SEGUENCES IN LOOP-TYPE LMFBRS.
NUREG/CR-3266: SEISMIC & DYNAMIC GUALIFICATION OF SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT IN OPERATING NUCLEAR POWER PLANTS.
NUREG/CR-3282: INTERNAL HYDROGEN EMBRITTLEMENT OF TITANIUM ALLOY TiCODE-12 AT ROOM TEMPERATURE.
NUREG/CR-3283: FIRST EXCURSION PROBLEMS FOR GAUSSIAN VECTOR PROCESSES.
NUREQ/CR-3284: REVIEW OF CURRENT ANALYSIS METHODOLOGY FOR REINFORCED CONCRETE STRUCTURAL EVALUATIONS NURE0/CR-3315: A CONSENSUS ESTIMATION STUDY OF NUCLEAR POWER PLANT STRUCTURAL LDADS.
NUREQ/CR-3381: EVALUATION OF THREE MILE ISLAND UNIT 2 REACTOR BUILDING DECONTAMINATION PROCESS.
BURNS & ROE CD.
NUREO/CR-3394 YO1: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS.Frecutive Summary And Significant Findings.
NUREG/CR-3394 V02: PROBABILISTIC ASSESSMENT OF RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Compendium Of Calculated Break Probabilities Sump Blockage Frequencies And Calculational f
Program Listing.
CALIFORNIA, UNIV. OF, DAVIS NURE0/CR-3362: THE INFLUENCE OF LYPFH0HEMATOPO! ETIC DISEASE IN ALTERING RADIATION INDUCED CANCER SENSITIVITY IN MICE.
COMMERCE, DEPT. OF, NATIONAL BUREAU OF STANDARDS NUREQ/CR-3341: PROBABILISTIC DESCRIPTIONS OF RESISTANCE OF SAFETY-RELATED NUCLEAR STRUCTURES.
NURE0/CR-3399: EXPERIMENTAL VERIFICATION OF THE NEUTRON SPECTRUM FROM THE NBS D(2)D-MODERATED (252)CF SOURCE.
NUREO/CR-3400: ANALYSIS OF MEASUREMENTS WITH PERSONNEL DOSIMETERS AND PORTABLE INSTRUMENTS FOR DETERMINING NEUTRON DOSE EGUIVALENT AT NUCLEAR POWER PLANTS.
CONTROL DATA CORP.
NURE9/CR-3416 V01: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 1: Data Evaluation.
NUREG/CR-34M VO2: EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABILITIES. Phase 1 Topical Rept Vol 2: Code Evaluation.
DEL RESEARCH CORP.
NUR E9/CR-3464: TIE APPLICATION OF FRACTURE PROOF DESIGN METHODS USING TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTIAL THROUGH WALL CRACKS.
i l-135 L
E060, INC.
NURE0/CR-2148 ADD 01: FRAP-T6: A COMPUTER CODE FOR THE TP.ANSIENT ANALYSIS
-0F OXIDE-FUEL RODS.
NUREG/CR-3003 V02: HUMAN ENGINEERING DESIGN CONSIDERATIONS FOR CATHODE RAY TUBE-GEERATED DISPLAYS.
NURE0/CR-3214:
SUMMARY
OF NUCLEAR REGULATORY COMMISSION'S LOFT PROGRAM 4
EXPERIMENTS.
NURE0/CR-3289: COMMON CAUSE FAULT RATES FOR INSTRUMENTATION AND CONTROL ASSEMBLIES. Estimates Based on Licensee Event Reports At U.S.
, Commercial Nuclear Power Plants, 1976-1981.
NURE0/CR-3419: EXPERIMENT DATA REPORT FOR BEMISCALE MOD-2B POWER LOSS l
TEST SERIES (TEST S-PL-1,-2,AND-3).
i NURE0/CR-3437: DATA ANALYSIS USING BIONOMIAL FAILURE RATE' COMMON CAUSE MODEL.
NURE0/CR-3464: TM APPLICATION OF. FRACTURE PROOF DESIGN METHODS USING
' TEARING INSTABILITY THEORY TO NUCLEAR PIPING POSTULATING CIRCUMFERENTIAL THROUGH WALL CRACKS.
ELECTRIC POWER RESEARCH INSTITUTE NUREG/CR-2575; BWR FULL INTEGRAL SIMULAT* ION TEST (FIST) PROGRAM TEST PLAN.
4 NURE0/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
1 ENSA. INC.
NURE0/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
i FLORIDA STATE UNIV.
NUREG/CR-3472 V01: SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF t
ALTERNATIVE WASTE FORMS.
GENERAL ELECTRIC CD.
NURE0/CR-2573: BWR REFILL-RELOAD PROGRAM TASK 4. 7 - MCDEL DEVELOPENT BASIC MODELS FOR THE BWR VERSION OF TRAC.
NURE0/CR-2574: BWR REFILL-REFLOOD PROGRAM TASK 4. 7 - NODEL DEVELOPENT TRAC-BWR COMPONENT MODELS.
NURE0/CR-2575: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TEST PLAN.
NURE0/CR-3223: BWR REFILL REFLOOD PROGRAM FINAL REPORT.
GENERAL PHYSICS CORP.
NURE0/CR-2726: LIGHT WATER REACTOR HYDROGEN MANUAL.
NUREO/CR-3371 V01: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 1: Project Approach And Methodology.
' NUREG/CR-3371 V02: TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS. Volume 2: Data Results.
HAWORD ENGIMERING DEVELOPENT LABORATORY NUREG/CR-2146 V02: DYNAMIC ANALYSIS TO ESTABLISH NORMAL SHOCK AND 2
VIBRATION OF RADIDACTIVE MATERIAL SHIPPING PACKAGES:Guarterly Progress Report. April 1,1981 - June 30,1981.
NURE0/CR-2805. V04: LWR PRESSURE VESSEL SURVEILLANCE DOSIMETRY IlfROVEMNT PROGRAM. Guarterly Progress Report, Octaber 1992 - December 1982.
.ILLINDIS, STATE OF-NUREG/CR-2478 V02: A STUDY OF TRENCH COVERS TO MINIMIZE INFILTRATION AT WASTE DISPOSAL SITES. Task II Report - Laboratcry Evaluation And Computer Modeling of Trench Cover Design.
ILLINDIS, UNIV. OF, URBANA-CHAMPAIGN NURCO/CR-3425: LOW-ENER0Y PHOTON DOSE DEPOSITION IN TISSUE SLAB AND SPHERICAL PHANTOMS.
],
. INTER A ENVIRONMENTAL CONSULTANTS, INC.
NUREG/CR-3111 V01: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIDACTIVE WASTES IN GEOLOGIC REPOSITORIES.
LAWRENCE BERKELEY LABORATORY NURE0/CR-3475: CRITICAL DISCHARGE OF INITIALLY SUBCOOLED WATER TFNt0 UGH SLITS.
136
LAWRENCE LIVERMORE LADORATORY NUREQ/CR-3017: CORRELATION OF SEISMIC EXPERIENCE DATA IN NON-NUCLEAR FACILITIES WITH SEISMIC EGUIPMENT GUALIFICATION IN NUCLEAR PLANTS (A-46).
NUREQ/CR-3132: CONCENTRATION AND DISTRIBUTION OF COPPER IN EFFLUENTS FROM THE H.B.
ROBINSON STEAM ELECTRIC PLANT.
NUREQ/CR-3133: THE SENSITIVITY OF ADULT, EMBRYONIC AND LARVAL CRAYFISH PROCAMBARUS CLARKII TO COPPER.
NUREQ/CR-3270: INVESTICATION OF ELECTROMAGNETIC INTERFERENCE (EMI)
LEVELS IN COMMERCIAL NUCLEAR POWER PLANTS.
NUREQ/CR-3345: AN ASSESSMENT OF POTENTIAL INCREASES IN RISK DUE TO DEORADATION OF STEAM CENERATOR AND REACTOR COOLANT PUMP SUPPORTS. Seismic Safety Margins Research Program.
LOS ALAMOS SCIENTIFIC LABORATORY NUR EQ/CR-3221: MC&A REFORM AMENDMENT: EXAMPLE SYSTEM BETA VARIANCE ESTIMATES AND PERFORMANCE EVAL 8)ATION.
NUREQ/CR-3278: HYDROGEN BURN ANALYSES OF ICE CONDENSER CONTAINMENTS.
NUREQ/CR-3280: TRAC-PF1 DEVELOPMENTAL ASSESSMENT.
NUREQ/CR-3355: TRAC-PD2 POSTTEST ANALYSIS OF THE CCTF EVALUATION MODEL TEST C1-19 (RUN 38).
MATERIALS ENGINEERING ASSOCIATES, INC.
NUREQ/CR-3402: ALTERNATVE PROCEDURES FOR J-R CURVE DETERMINATION.
NORTHWESTERN UNIV.
NUREQ/CR-3060: A CRITICAL REVIEW OF FLOODING LITERATURE.
DAK RIDGE NATIONAL LABORATORY NUREQ/CR-2000 V02 N6: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of June 1983.
NUREQ/CR-2000 V02 N7: LICENSEE EVENT REPORT (LER) COMPILATION: For Month of July 1983.
NUREQ/CR-2000 V02 N8: LICENSEE EVENT REPORT (LER) COMPILATION: For Month DF August 1983.
NUREQ/CR-2506: UNCERTAINTIES IN GEOLOGIC DISPOSAL OF HIGH-LEVEL WASTES
-OROUNDWATER TRANSPORT OF RADIONUCLIDES AND RADIOLOGICAL CONSEQUENCES.
NUREQ/CR-2669: FRONT-END ANALYSIS FOR NUCLEAR POWER PLANT MAINTENANCE PERSONNEL RELIABILITY MODEL.
NUREQ/CR-2844 V04: HIGH-TE.MPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.GUARTERLY PROGRESS REPORT, OCTOBER 1 - DECEMBER 31,1982.
NUR EQ/CR-2989: RELIABILITY OF EMERGENCY AC POWER SYSTEM AT NUCLEAR POWER PLANTS.
NUREQ/CR-3122: POTENTIAL DAMAGING FAILURE MODES OF HIGH-AND MEDIUM-VOLTAGE ELECTRICAL EQUIPMENT.
NUREQ/CR-3138: MEASUREMENT OF TWO-PHASE FLOW AT THE CORE / UPPER PLENUM INTERFACE FOR A PWR CEOMETRY UNDER SIMULATED REFLOOD CONDITIONS.
NUREQ/CR-3155: MODIFICATION OF OCA-I FOR APPLICATION TO REACTOR PRESSURE VESSEL WITH CLADDING ON THE INNER SURFACE.
NUR EQ/CR-3161: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM, ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31,1983.
NUREQ/CR-3200 V01: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM GUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 31,1983.
NUREQ/CR-3265: EFFECTS OF OFF-SPECIFICATION PROCEDURES ON THE MECHANICAL PROPERTIES OF HALF-BEAD WELD REPAIRS.
NUREQ/CR-3331: A METHODOLOGY FOR ALLOCATING NUCLEAR POWER PLANT CONTROL FUNCTIONS TO HUMAN OR AUTOMATIC CONTROL.
NUREQ/CR-3333: NEUTRON SPECTRAL CHARACTERIZATION OF THE FOURTH NUCLEAR REGULATORY COMMISSION HEAVY SECTION STEEL TECHNOLOGY iT-CT IRRADIATION EXPERIMENTS: Dosimetry And Uncertainty Analysis.
NUREG/CR-3334 V01: HEAVY SECTION STEEL TECHNOLOGY PROGRAM GUARTERLY PROGRESS REPORT FOR JANUARY - MARCH 1983.
137
- - - - ~
MURE0/CR-3364: PRESENCE OF PATHOGENIC MICRODRGANISMS IN POWER PLANT COOLING WATERS.
OHIO STATE UNIV.
NURE0/CR-3182: DEVELOPMENT OF AN IN-SITU METHOD FOR TESTING NEUTRON SENSORS USED IN NUCLEAR POWER PLANT REACTOR PROTECTION SYSTEMS.
PARAMETER, INC.
I NURE0/CR-3048: CLOSEQUT OF IE BULLETIN B0-21: VALVE PARTS SUPPLIED BY MALCOLM FOUNDRY.
NURE0/CR-3050: CLOSEQUT OF IE BULLETIN 91-02 AND SUPPLEMENT: Failure Of Gate-Type Valves To Close Against Differential Pressure.
PRITSKER ts ASSOCIATES,INC.
NUREG/CR-3423: TE SAFE 00ARDS NETWORK ANALYSIS PROCEDURE SNAP: A User's Manual.
PURDUE UNIV., WEST LAFAYETTE, IN NURE0/CR-3306: LWR AND HTOR COOLANT DYNAMICS: THE CONTAIFBENT OF SEVERE DYNAMICS.
RADIATION MANAGEENT CORP.
NURE9/CR-3387: RADIOLOGICAL SURVEY OF THE COMBUSTION ENGIEERING BURIAL SITE E MATITE, MISSOURI.
NURE9/CR-3486: RADIOLOGICAL SURVEY OF THE NUCLEAR FUEL SERVICES,INC.
ERWIN PLANT BURIAL SITE.
RENSSELAER POLYTECMIC INST., TROY, NY NURE9/CR-3372: AN ANALYSIS OF WAVE DISPERSION,BONIC VELOCITY AND CRITICAL FLOW IN TWO-PHASE MIXTURES.
NURE0/CR-3373: AIR / WATER SUBCliANNEL MEASUREMENTS OF THE EQUILIBRIUM GUALITY AND MASS FLUX DISTRIBUTION IN A ROD BUNDLE.
NURE0/CR-3374: TE DEVELOPMENT OF A GAMMA-RAY SCATTERING DENSITOMETER AND ITS APPLICAT' ION TO THE MEASUREMENT OF TWO-PHASE DENSITY DISTRIBUTION IN AN ANNULAR TEST SECTION.
NURE9/CR-3375: THE DEVELOPMENT OF NUFREG-N,AN ANALYTICAL MODEL FOR THE STABILITY ANALYSES OF NUCLEAR COUPLED DENSITY-WAVE OSCILLATIONS IN l
BOILING WATER NUCLEAR REACTORS.
NURE0/CR-3376: PARALLEL CHANNEL EFFECTS AND LONG-TERM COOLING DURING EER0ENCY CORE COOLING IN A BWR/4.
NUREG/CR-3377: MODEL ANALYSIS FOR REACTOR KINETICS AND STABILITY.
SANDIA LABORATORIES NURE0/CR-1246 V04: SAFE USERS MANUAL. Volume 4: Computer Programs.
NUREG/CR-2238 V04: ADVANCED REACTOR SAFETY RESEARCH GUARTERLY REPORT OCTOBER - DECEMBER 1991. Volume 20.
NURE0/CR-2254: A PROCEDURE FOR CONDUCTING A HUMAN RELIABILITY ANALYSIS FOR NUCLEAR POWER PLANTS.
l NUREG/CR-2402: RISK ANALYSIS METHODOLOGY FOR SPENT FUEL REPOSITORIES IN I
BEDDED BALT. Final Report.
NUREG/CR-2631: A LOGICAL FRAMEWORK FOR IDENTIFYING EQUIPMENT IMPORTANT TO SAFETY IN NUCLEAR POWER PLANTS.
NURE9/CR-2726: LIGHT WATER REACTOR HYDR 00EN MANUAL.
NUREG/CR-2739: DATA BASE FOR BASALT ETHODOLOGY.
NUREO/CR-2982 RO1: BUDYANCY, TRANSPORT,AND HEAD LOSS OF FIBROUS REACTOR INSULATION.
NUREG/CR-3085 V02: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF T E MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol II - Appendia A And 5.1 Thru B.4. Docket No. 50-245.(Northeast Nuclear Energy Company)
NUREG/CR-3085 V03: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol III Appendix B.5 Thru B.S. Docket No. 50-245.(Northeast Nuclear Energy)
NUREG/CR-3085 V04: INTERIM RELIABILITY EVA! UAiION PROGRAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWR PLANT. Vol IV Appendix B.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
NURE0/CR-3111 V01: AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES.
138
r-l NURE0/CR-3234: TM POTENTIAL FOR CONTAINMENT LEAK PATHS THROUGH ELECTRICAL PENETRATION ASSEMBLIES UNDER SEVERE ACCIDENT CON 1
NUREG/CR-3257:
l RELAP5 ASSESSENT: LOFT TURBINE TRIP L6-7/L9-2.
NUREG/CR-3258: RELAPS ASSESSMENT:SEMISCALE NATURAL CIRCULATION TESTS I
S-NC-2 AND S-NC-7.
NUREG/CR-3269: THE NADH-RICH CORNER OF THE (NA+,CA+) (OH.CO32-)
RECIPROCAL SYSTEM 1
NUREO/CR-3277: RELAP5 ASSESSENT: GEMISCALE MOD-3 SMALL BREAK TEST.
i NURE9/CR-3329 VO1:
l THERMAL HYDRAULIC ANALYSIS RESEARCH PROGRAM GUARTERLY REPORT JANUARY-MARCH 1983.
NUREG/CR-3394 Vol: PROBABILISTIC ASSESSMENT OF, RECIRCULATION SUMP BLOCKAGE DUE TO LOSS OF COOLANT ACCIDENTS. Executive Sueciary And Significant Findings.
NUREG/CR-3401: LABORATORY-SCALE SODIUM-CARBONATE A00REGATE CONCRETE INTERACTIONS.
SCIENCE APPLICATIONS. INC.
NUREO/CR-3085 V02: INTERIM RELIABILITY EVALUATION PROGRAM: ANAL THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT. Vol'.1 - Appendix A And B.1 Thru B.4. Docket No.
50-245.(Northeast Nue?
'e Energy Company)
NUREG/CR-3085 V03:
INTERIM RELIABILITY EVALUATION PRCAAM: ANALYSIS OF THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol III Appendix B.5 Thru B.8. Docket No.
50-245.(Northeast Nuclear Energy)
NUREG/CR-3085 VO4: INTERIM RELIABILITY EVALUATION PROGRAM: ANALYS THE MILLSTONE POINT UNIT 1 NUCLEAR POWER PLANT.Vol IV Appendix B.9 Thru B.19 And C. Docket No.50-245.(Northeast Nuclear Energy)
NUREQ/CR-3111 Vol:
AN ASSESSMENT OF THE PROPOSED RULE (10CFR60) FOR DISPOSAL OF HIGH-LEVEL RADI0 ACTIVE WASTES IN GEGLOGIC REPDGITO NURE0/CR-3264: PARAMETRIC EFFECTS ON NONDESTRUCTIVE ASSAY TECHNIGUES.
SOUTHERN STATES ENERGY BOARD NUREG/CR-3487: REGIONAL-INTERSTATE SITE REVIEW PROCEDURE: LOW-LEVE RADIDACTIVE WASTE DISPCWL FACILITY.
TEKEKRON, INC.
NUREG/CR-3451:
BENCHMARK PROBLEMS FOR RADIOLOGICAL ASSESSMENT CODES.
UNITED NUCLEAR CORP.
NUREQ/CR-3336:
EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS. Summary Report,Ames Laboratory Research Reactor. Docket No.
50-116.
NUKE 9/CR-3370: EVALUATION OF NUCLEAR DECOMMISSIONING PROJECTS SU REPORT NORTH CAROLINA STATE UNIVERSITY RESEARCH AND TRAINING WASHINGTON, UNIV. OF, SEATTLE, WA NUREG/CR-3392: PROCESS NOTEBOOK FOR AGUATIC ECOSYSTEM SIMULATION.Second Edition.
r 139 i
Licensed Facility index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number, if fur-ther information is needed, refer to the main c!tation by the NUREG number.
50-155 Big Rock Point Nuclear Plant, Consumers Power Co NURFG-0828 DRFT STN-50-470 CESSAR Sy s t em 80, Combustion Engineering NUREG-0852 S02 50-537 Citnch River Breeder Reactor, Project Management Corp.
NUREG-Ov68 VO! ERR 50-157 Cornell Univ. Research Reactor NUREC-0984 50-97 Cornell Univ. Zero Power Reactor NUREG-1010 50-275 Diablo Canyon Nuclear Power Plant. Unit 1. Pacific Gas & Electric Co NUREG-0675 S16 50-275 Diablo Canyon Nuclear Power Plant, Unit 1. Pacific Cas & Electric Co NUREG-0675 SIR 50-323 Diablo Canyon Nuclear Power Plant, Unit 2. Pacific Gas & Electrac Co NUREG-0675 Sie 50-323 Diablo Canyon Nucleae Power Plant. Unit 2. Pacific Ces & Electric Co NUREG-0675 S18 STN-50-447 CESSAR-238, General Electric Co NUREG-0979 SO!
50-261 N.
B.
Robinson Plant. Unit 2, Carolina Power and Light Co NUREG-iOO3 50-261 t i.
B.
Robinson Plant. Unit 2, Carolina Power and Ltght Co.
NUREC/CR-3132 50-116 lowa State Univ. 'te sear ch Reac tor NUREG-1016 50-11b lowa State Univ. Research Reactor NUREG/CR-3336 50-116 lowa State Univ. Research Reactor NUREG/CR-3336 ERR 40-2061 Kerr-McGee Chemical Corp., Oklahoma City, OK, NUREG-0904 ERR 50-373 LaSalle County Station, Li s t 1. Commonwealth Edison Co.
NUREG-0519 SOS50-37A LaDalle County Station. Unit 2, Commonwealth Edison Co.
NUREG-0519 S05 50-332 Limerich Generating Station, Unit 1. Philadelphia Electric Co NUREG-0991 50-333 Limerick Generating Station, Unit 2, Philadelphia Electric Co NUREG-0991 50-245 Millstone Nuclear Power Station. Unit
- 1. Northeast Nuclear Energy Co NVREG/CR-3085 VO2 50-245 Millstone Nuclear Power Station, Unit le Northeast Nuclear Energy Co NVREG/CR-3085 VO3 50-245 Millstone Nuclear Power Station, Unit 1. Northeast Nuclear Energy Co NUREC/CR-3085 VO4 50-194 National Bureau of Standards Reactor. National Bureau of Standards NUREC-1007 50-297 North Carolina State Univ. PULSTAR Reactor NVREC/CR-3370 70-0143 Nuclear Fuel Services. Inc.. Erwin, TN, NVREG/CR-3486 50-131 Omaha Vet. Admin. Hosp. Research Reactor NUREG-0989 50-266 Point Beach Nuclear Plant. Unit 1, Wisconsin Electric Power Co NUREG-1011 50-266 Point Beach Nuclear Plant. Unit
- 1. Wisconsin Electric Power Co NUREG-1011 50-244 Robert Emmet Oinna Nuclear Plant, Unit 1. Rochester Can & Electric C NUREG-0821 601 50-272 Salen Nuclear Generating Station. Unit 1.Public Service Electric & G NUREG-1000 VO2 50-322 Shoreham Nuclear Power Station. Long Island Lighting Co NUREG-0420 SO4 40-8781 Teton Esploration Drilling Co..
Inc.. Casper. WY, NVREG-0925 50-289 Three Mile Island Nuclear Station, Unit 1.
Metropolitan Edison Co.
NUREG-1020 VO!
50-289 Three Mile Island Nuclear Station, Unit 1.
Metropolitan Edtson Co NVREG-1020 v02 50-320 Three Mile Island Nucle'er Station, Unit 2.
Metropolitan Edison Co NUREG/CR-3381 50-112 Univ. of Oklahoma Research Reactor NUREG-0996 STN-50-402 Wolf Creek Generating Station, Hansas Gas & Electric NUREG-0881 SO3 141
L h,
Fc m 335 U.S. NUCLE AR RE"ULATCOY COMMISSION BIBLIOGRAPHIC DATA SHEET NUREG-0304fol.8,No.3
- 4. TlTLE AND SUBTITLE (Add Vo/ume No.,if apprepnatel R gulatory and Technical Reports
- 2. (Leave bleanJ [
E Compi on for Third Quarter 1983
- 3. RECIPIEN ACCESSION NO.
- 7. AUTHOR (S)
- 5. DATE [ ORT COMPLE TED M ON jYEAM
- 9. FERFORMING ORGAN TION NAME AND MAILING ADDRESS //nclude leo Code /
D REPORT ISSUEO Division of Tec ical Information and Document Control
~m IvE^a Office of Admini ation vember 1983 U.S. Nuclear Regul ory Commission
[ft '"* * *"'" >
Washington, DC 205 g,,,,,,,,,
12, SPONSORING ORGANIZATION NA ' AND M AILING ADDRE SS (/nclude l<a Coael
/-
p Same as 9, above.
It FIN NO.
- 13. TYPE OF REPORT P
- OD COVE RE D //nclus,ve dates /
Quarterly f
July - September 1983
- 15. SUPPLEMENTARY NOTES 14 (trave o/ anal
- 16. ABSTR ACT 200 words or / css /
This compilation lists all NRC regulato and technical reports publi.shed under the NUREG series during the third quar f 1983.
l
/
/
l r-
- 17. KEY WORDS AND DOCUMENT A-LYSIS 17a DESCRIPTORS s
17b. IDENTIFIE RS O ENDED TERMS h
- 18. AVAILABILI ST A TE ME NT 19 SE CURITY CLASS ITh,s reoorrl 21 NO OF PAGES Unlimited Unclassifind 20 SeCua TY Ct ASS /ra,s o,ri 22 Price Unclassified s
NRC FORM 335 41181 m_
f f
Main Citations I
and Abstracts Contractor Report 2
Number index h
Personal Author Index 4
Subject index 120555078877 1
99999 i
US NRC l
ADM-DIV OF TIDC POLICY C PUB MGT BR-POR NUREG i
k-501 WASHINGTON DC 20555 5
NRC Originating Organization Index NRC Contractor 6
sponsor index 7
Contractor index
- '8 Licensed Facility Index