ML20126H142
| ML20126H142 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1981 |
| From: | Mckenzie L, Mckinney U, Oliu W NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V05-N03, NUREG-304, NUREG-304-V5-N3, NUDOCS 8104070006 | |
| Download: ML20126H142 (150) | |
Text
{{#Wiki_filter:- - - - - - - - - - - - - - - - - - - - - - - - NUREG-0304 Vol. 5, No. 3 Regulatory and Technical Reports Compilation for Third Quarter 1980 July - September b i f[g~i=,9cR
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NUREG-0304 Vol. 5, No. 3 1
- T~::~ ~ ~. - _ - . _ . . - . . . - . . . _ _ . - _ . .
Regulatory and Technical Reports l Com allation for Thirc Quarter 1980 July - September
. _ _ . _ .__.___T___ _ _ _ _ _ _
Date Published: January 1981 Compiled by Walter E. Oliu and Linda McKenzie Indexed by Ursula Mckinney Division of Technical Information and Document Control Office of Administration ! U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i 4% (4 t t l l l l l l l
CONTENTS v Preface................................................................... Thumb Tab 1 Sequential List of Staf f-Generated Reports Designated NUREG . . . . . . . . . . . . . . Sequential List of NRC Contractor-Generated Reports Designated 2 NUREG/CR ............................................................... 3 Keyword Index to Reports.................................................. Cross-Reference to Contractor Report Numbers.............................. 4 iii
Preface This compilation lists formal regulatory and technical reports issued from July through September 1980 by the U.S. Nuclear Regulatory Commission (NRC) staff and by NRC contractors. The compilation is divided into two major sections. l The first major section consists of a sequential listing of all NRC reports in report-number order. The first portion of this sequential section lists staff I reports, and the second portion lists contractor reports. Each report citation in the sequential section contains full bibliographic information, including: l NRC report number Report title Month and year of issuance Contractor (if appropriate) Contractor report number (if appropriate) NRC originating or sponsoring office Availability (where report can be obtained) Abstract The second major section of this compilation consists of a key-word index to report titles. Each key word is cross-referenced to the report or reports in the sequential listing that contains that word. The third major section contains an alphabetically arranged listing of contractor report numbers cross-referenced to their corresponding NRC report numbers. How to Read Citations ,
@ ^ h f i NUREG-0678 The Effects of Temperature, Moisture, Concentration, l Pressure, and Mass Transfer on the Adsorption of l Krypton and Xenon on Activted Carbon. I
@ August 1980.
@ - ACRS GP0 NTIS: @
5 This report is a critical review of the published literature on the adsorption of radioactive krypton and xenon on activated charcoal. The report includes a tabulation and evaluation of the adsorption coefficients for these two gases as related to temperature, pressure, moisture, mass transfer effects, and the nature of the carrier gas. Wherever possible, the resulting data have been used to develop simple correlations for quantitatively evaulating the @ eifects of these parameters on noble gas adsorption. Important conclusions of the study include the observations that (a) individual charcoals have a wide range of adsorption coefficients and therefore the performance of a given bed is heavily dependent on the quality of the charcoal it contains; (b) because ; of the detrimental effects of mass transfer on noble gas adsorption, consideration I should be given to including this factor in developing technical specifications for adsorption beds; and (c) additional research is needed on the determination of the interelationship of moisture and temperature and their effects on adsorption bed performance. v
r l
.@ 'NRC report r. umber
@ Title
[
.@ Month and year issued' f Y
@. Originating office-
.@. GP0/NRC Availability . l i
.@ National Technical Information Service availabilityL
@ Abstract i'
The key to abbreviations for.NRC Offices appears at the end of this preface. { Contractor Report '
@ ^ ,
~ f t NUREG/CR-0508- Security Communication Systems for Nuclear Fixed hSite Facilities.
.@ : July 1980,
@ : Y-12 Plant,'0ak Ridge '
Y/0W-128
@ : OSD GPO NTIS d 7 '
e This report presents a basic discussion of communication tecnniques and factors-relevant to designing communication systems for nuclear fixed-site facility security systems. The reader is provided communication fundamentals, design considerations, and. specification techniques. Copious references and an annotated ; bibliography are provided for individuals'who desire to delve deeper than the ' limits and areas of study of this report. Ease of reading and use of this rep' ort are enhanced by relegating detailed communication design treatise to
+
the Appendices. Sample procurement specifications are provided-throughout the report for various communication system components and are distinguished from the regular text by using a smaller type. *
@ NRC report number
@ Title ;
@ Month and year issued "
@ Contractor (
i f vi i i i f 7
f
@ Cont.ractor report number @ NRC sponsoring office @ GP0/NRC availability- @ National Technical Information Service availability @ Abstract Availability of NRC Publications Copies of Nim staff and contractor reports may be purchased either from the NRC-GP0 Sales Office or from the National Technical Information Service, Springfield, Virginia 22161. To. purchase documents from the NRC-GPO Sales Office, send a check or money order, payable to the Superintendent of Documents, '
to the following address: U.S. Nuclear Regulatory Commission ATTN: Sales Manager Washington, D.C. 20555 You may charge any purchara to your GP0 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 Inter national 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. q l NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff generated report. Contractor prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established,' codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as
.various other numbers that could not be correlated with NRC sponsorship of the work being reported.
In addition'to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings. All :.hese report codes are controlled and assigned by the NRC Division of Technical Information and Document Control. vii
Abbreviations for NRC Offices ACRS - Advisory Committee for Reactor Safeguards ADM - Office of-Administration ELD - Office of the Executive Legal Director IE - Office of Inspection and Enforcement IP - Office of International Programs MPA - Office of Management and Program Analysis NMSS - Office of Nuclear Material Safety and Safeguards NRR - Office of Nuclear Reactor Regulation CON - Office of the Controller PE - Office of Policy Evaluation RES - Office of Nuclear Regulatory Research SD - Office of Standards Development SP - Office of State Programs viii
E Sequential List of Staff-Generated Reports Designated NUREG i 1 l l l l l f
Report No. Bibliographic Dtta NUREG-0011, Supp. 2 Safety Evaluation Report Related to Operation of Sequoyah Nuclear Plant, Units 1 and 2, Docket Nos. 50 327 and 50-328.. Tennessee Valley Authority, Supp. No. 2. August 1980. ONRR -GPO. NTIS. Supplement No. 2 to the Safety Evaluation Report of Tennessee Valley Authority's application for licenses to operate its Sequoysh Nuclear Plant Units 1 and 2, located in Hamilton County, Tennessee, has been prepared by the Of fice of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been completely resolved at the nine of publica-tion of the Safety Evaluation Report, and defines the requirements tha *. must be met for full-power operation.
-NUREC-0011, Supp. 3 Safety Evaluation Report Related to Operation of Sequoyah Nuclear Plant, Units 1 and 2, Docket Nos. 50-327/328. Tennessee Valley Authority, Supp, 3.
September 1980. ONRR GPO. NTIS, Supplement No. 3 to the Safety Evaluation Report of Tennessee Valley Authority's application for licenses to operate its Sequoyah Nuclear Plant Units 1 and 2, located in llamilton County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement provides furthe. information on the hydrogen control measures for Unit 1. KUREG-0053, Supp. 11 Safety Evaluation Report Related to the Operation of North Anna Power Station, Unit 2, Virginia Electric and Power Company, Docket No. 50-339. Supplement No. 11. August 1980. ONRR GPO. NTIS. On June 4, 1976 the Nuclear Regulatory Commission issued its Safety Evaluation regard-ing the application for licenses to operate the North Anna Power Station Unita 1 and
- 2. The application was filed by Virginia Electric and Power Company. Supplement No. I to the Safety Evaluation Report was issued on June 30, 1976; Supplement No. 2 was issued on August 2 1976; Supplement No. 3 was issued on September 15, 1976; Supplement No. 4 was issued on December 8, 1976; Supplement No. 5 was issued on December 29, 1976; Supplement No. 6 was issued on February 2,1977; Supplement No. 7 was issued on August 18, 1977; Supplement No. 8 was issued on December 14, 1977; Supplement No. 9 was issued on March 31, 1978, and Supplement No.10 was issued on April 10, 1980. Supplements 1 through 9 documented the resolution of several out-standing items. Supplement No. 10 addresses the requirements for fuel loading and conducting low power testing of North Anna Unit 2. This supplement, No. 11, addresses the requirements which must be completed prior to the issuance of a full-power operating license for Unit 2.
NUREG-0053, Supp. 12 Safety Evaluation Report Related to the Operation of North Anna Power Station, Unit 2, Docket No. 50-339. Supplement No. 12. August 1980. OKRR G PO . NTIS. On June 4,1976 the Nuclear Regulatory Commission issued ita Safety Evaluation regard-ing the application for licenses to operate the North Anna Power Station Units 1 and
- 2. The application was filed by Virginia Electric and Power Company. Supplement No. 1 to the Safety Evaluation Report was issued on June 30, 1976; Supplement No. 2 was issued on August 2 1976; Supplement No. 3 was issued on September 15, 1976; Supplement No. 4 was issued on December 8, 1976; Supplement No. 5 was issued on December 29, 1976; Supplement No. 6 was issued on February 2, 1977; Supplement No. 7 was issued on August 18, 1977; Supplement No. 8 was issued on December 14, 1977; Supplement No. 9 was issued on March 31, 1978, and Supplement No. 10 was issued on April 10, 1980. Supplements 1 through 9 documented the resolution of several out-standing items. Supplement No 10 addresses the requirements for fuel loading and conducting low-power testing of North Anna Unit 2. Supplement No. 11 addresses the requirements which must be completed prior to the issuance of a full-power operating license int Unit 2. This Supplement, No. 12, addresses emergency preparedness.
2 l l
4 i 1 . Report No. .Bibliogrcphic Data [ FUREC-0090, Vol 3 No,'l Report to Congress on Abnormal Occurrences, January-March 1980. I September 1980. OMPA CPO. kTIS. Section 208 of the Energy Reorganization Act of 1974 Identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines [ to be significant from the standpoint of public health or safety and requires a ; quarterly report of such events to be made to Congress.' This report, the twentieth in the series, covers the period January I to March 31, 1980. During the period, there were two abnormal occurrences at the nuclear power plants licensed to operate: one involved exposures to beta radiation in excess of regulatory limits and the i second (a generic concern) involved a transient initiated by partial loss of power. ' There was one abnormal occurrence at the fuel cycle facilities (other than nuclear power plants); the incident involved a loss of confinement system which resulted in inhalation of plutonium by an employee. There was one abnormal occurrence at other - licensee facilities; the incident involved overexposure to individuals in unrestricted areas. There were no ataormal orrurrences reported by the Agreement States. This report also contains intornation updating previously reported abnormal occurrences. NUREG-Oll7, Supp. 4 Safety Evaluation Report Related to the Operation of Joseph M. Farley Nuclear Plant, Unit 2. Docket No. 50-564, Alabama Power Company. Supplement 4 to NUREG-75/034. September 1980. ONRR GPO. NTIS. Supplement No. 4 to the Safety Evaluation Report of Alabama Power Company's applica-tion for licenses to operate its Joseph M. Farley Nuclear Plant Unita 1 and 2, located in Houston County, Alabama, has been prepared by the Office of Nuclear Reactor , Regulation of the U.S. Nuclear Regulatory Commission. This supplement providea the NRC staf f's evaluation of Alabama Power Company's FSAR Amendment Nos. 67 through 74 for the Farley Nucicar Plant, and its response to other safety issues, including the TMI-2 Action Plan, that have arisen since Supplement No. 3 was issued in June 1977. The staff concluded that the Farley Nuclear Plant Unit 2 may be issued a license for fuel loading and low-power testing. NUREG-0134, Add, 2 Final Environmental Statement Related to the Operation of North Anna Power Station, Unit 1 and 2, Docket No. 50-338 and 50-339. Virginia Electric and Power Company. August 1980. OKRR GPO. NTIS. A Final Environmental Statement for the North Anna Power Station Units I and 2, proposed for operation by Virginia Electric and Power Company, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear fegulatory Commission. Addendum 2 to the Final Environmental Statement clarifies or amplifies information with regard to the Table S-3 and does not affect the cost-benefit conclusion already made in the Final Environmental Statement and Addendum. NUREG-0304, Vol. 4 Regulatory and Technical Reports Compilation for 1979. July 1980. OADM GPO. NTIS. This report contains a compliation of all NRC Regulatory and Technical reports published under the NUREG series during 1979. l l NUREG-0313, Rev. 1 Technical Report on Material Selection and Processing Guidelines for BWR Coolant j Pressurr Boundary Piping. ) July 1960. ONRR GPO. NTIS. { 1 This report updates and supersedes the NRC technical positions established in NUREG-0313 '
" Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," published in July 1977. This report sets forth the NRC staf f's revised acceptable methods to reduce the intergranular stress corrosion cracking susceptibility of BVR ASME Code Class 1, 2, & 3 pressure boundary piping and safe ends. For plants that cannot fully comply with the material selection, testing, and processing guidelines of this document, varying degrees of augmented inservice inspection and leak detection requirements are presented.
i I [; port No. Bibliographic Data NUREG-0386, Supp. 2 United States Nuclear Regulatory Commission Staff Practice & Procedure Digest. Supplement 2 to Digest No. 2. September 1980. OELD GPO. NTIS. This second supplement to the second edition of the NRC Staff Practice and Procedure Digest contains a digest of a number of Commission, Atomic Safety and Licensing Appeal Board and Atomic Safety and Licensing Board decisions issued during the period from 10/1/78 to 12/31/78 interpreting the NRC's Rules of Practice in 10 CFR Part 2. The supplement also includes, to a very limited degree, material from adjudicatory decisions and regulation changes after December 31, 1978. The supplement, which is intended to be used as a " pocket-part" supplement to the DJgest itself, includes a number of new subsections and topics not covered in the Digest. The new subsections are noted in the index for the supplement. The Practice and Procedure Digest and the supplements thereto were prepared by attorneys in the NRC's Office of the Executive Legal Director as an internal research tool. Because of the Digest's usefulness to these attorneys, it was decided that it might also prove useful to members of the public. Accordingly, the decision was made to publish the Digest and subsequent editions thereof and supplements thereto periodically. [ NUREG-0436, Rev 1, Supp 1 Plan for Reevaluation of NRC Policy on Decommissioning of Nuclear Facilities. December 1978 to July 1980. August 1980. OSD G PO . NTIS. This report supplements and updates the information presented in EUREG-0436, Rev. 1, of the same title and dated December 1978. Supplement I defines new terminology for the decommissioning alternatives. It updates the status and schedules for developing the information base, the draft generic environmental impact statement, and the rulemaking. In addition, schedules for regulatory guides to support the rules are presented. NUREG-0452, Rev 3 Standard Technical Specifications for Westinghouse Pressurized Water Reactors, Revision 3 September 1980. ONRR GPO. NTIS. The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the USNRC for use in the licensing process of current Westinghouse pressurized water reactors. The W-STS sets forth the limits, operating conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements. NUREG-0487, Supp 1 MARK 11 Containment Lead Plant Program Load Evaluation and Acceptance Criteria - Generic Technical Activities A8 and A39. I September 1980. ! ONRR GPO. NTIS. f The staff issued a report, NUREG-0487, in October 1978 that provided acceptance criteria for the suppression pool dynamic loads associated with safety relief valve discharges and loss-of-coolant accidents for the lead MARK 11 plants. This report is a supplement to NUREG-0487. Issuance of this supplement concludes the MARK 11 Lead Plant Program except for the condensstion oscillation and the chugging load specifica-tions. It contains an evaluation of the proposed alternatives to the lead plant acceptance criteria and an update of the ongoing MARK II Long Term Program. This evaluation was conducted as a part of the NRC's Generic Technical Activities A-8 and A-39. NUREG-0525, Rev 2 Safeguards Summary Event List (SSEL). September 1980. ONMSS GPO. NTIS. The Safeguards Summary Event List (SSEL) provides data on nine categories of Safeguards-related events involving NRC licensed material or licensees. It is deliberately broad in scope for two main reasons. Firat, the list is designed to serve as a reference document. It is as complete and accurate as possible. If additional
Report No. Bibliogrcphic Data information is obtained on an event, it will be incorporated into future revisions of the list. Second, the list is intended to provide 66 broad a perspective of the nature of licensee-related events as possible. NUREG-0535 Review and Assessment of Package Requirements (Yellowcake) and Einergency Response to Transportation Accidents. July 1980. ONMSS GPO. NTIS. Following the accidental apill of yellowcake in Colorado in September 1977, a U.S. Nuclear Regulatory Commissica/U.S. Department of Transportation study group met to consider topics on yellowcake packaging and response to transportation accidents involving any radioactive material. The group concluded that in an accident the state and local government agencies are responsible for controlling the scene, that carriers are responsible for notifying authorities and isolating and cleaning up any spilled ru/(cactive material, and that shippers are responsible for infonning others of hazards af the cargo. The group recommended that these parties prepare plans for carrying out these responsibilities. The group also recommended transportation data-gathering programs but, based on cost effectiveness arguments, did not recommend additional package requirements for yellowcake shipments. NUREG-0559 A Comparative Analysis of LWR Fuel Designs. July 1980. ONRR CPO. NTIS. The computer code GAPCON-THERMAL-2 was used to generate thermal performance predictions f or the spectrum of commercial light-water-reactor fuel designs at four dif ferent steady-state power levels. The input parameters for the code were obtained from design data that are nonproprietan' and are tabulated in this report. Calculated values of maximum fuel temperature, average fuel temperature, stored energy, gap conductance, fission gas release and rod internal pressure are plotted as a function of burnup. Radial fuel pellet temperatures are also plotted at one burnup level. NUREG-0612 Control of Ileavy Loads at Nuclear Power Plants Resolution of Generic Technical Activity A-36. July 1980. ONRR GPO. NTIS. i This report summartres work performed by the NRC staff in the resolution of Generic ! Technical Activity A-36, " Control of Heavy Loads Near Spent Fuel." Generic Technical Activity A-36 is one of the generic technical subjects designated as " unresolved safety issues" pursuant to Section 210 of the Energy Reorganization Act of 1974. The report describes the technical studies and evaluations performed by the NRC staff, the staff's guidelines based on these studies, and the staff's plans for implementa-tion of its technical guidelines. NUREG-0653 1 Report on Nuclear Industry Quality Assurance Procedures for Safety Analysis Computer Code Development and Use. I August 1980. ONRR GPO. NTIS. As a result of a request from Commissioner V. Gilinsky to investigate in detail the l causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March 1978, the staff undertook an extensive investigation of vendor quality ) control practices as applied to safety analysis computer code development and use, i This investigation included conducting inspections of code development and use practices I of the four major light-water-reactor nuclear steam supply system vendors and .s major reload fuel supplier. The conclusion reached by the staff as a result of the investi-gation is that vendor practices for code development and use are basically sound. A i i number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality i assurance (QA) review and inspection process for computer codes and identified areas for improvement. l l l l l
Report No. Bibliographic Data. NUREG-0658, Rev 1 Technical Specifications, Sequoyah Nuclear Plant, Unit No. 1, Docket No. 50-327, Appendix "A" to License No. DPR-77. September 1980. OKRR GPO. NTIS. The Sequoyah Unit 1 Technical-Specifications were prepared by the U.S. Nuclear Regulatory Commission. The Sequoyah Unit 1 Technical specifications set forth the limits, operating conditions and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. NUREG-0660 Vol 1 NRC Action Plan Developed as a Result of the THI-2 Accident, Revision 1, Vol 1. July 1980. OEDO GPO. NTIS. The Action Plan provides a comprehensive and integrated plan for all actions judged necessary by the Nuclear Regulatory Commission to correct or improve the regulation and operation of nuclear f acilities based on the experience from the accident at the Three Mile Island Unit 2 nuclear facility and the official studies and investigations of the accident. The major portion of Revision 1 is a revised version of Chapter V, - which delineates the intentions of the Commission, in recognition of interrelation-ships that call for correlated planning and action, the items in the chapter have been grouped into seven subject areas. NUREG-0660, Vol 2 NRC Action Plan Development as a Result of the THI-2 Accident, Revision 1, Vol. 2. July 1980. OEDO GPO. NTIS. The Action Plan provides a comprehensive and integrated plan for all actiona judged necessary by the Nuclear Regulatory Commission to correct or improve the regulation and operation of nuclear f acilities based on the experience from the accident at the " Three Mile Island Unit 2 nuclear facility and the official studies and investiga-tions of the accident. The major portion of Revision 1 is a revised version of Chapter V, which delineates the intentions of the Commission. In recognition of interrelationships that call for correlated planning and action, the items in the chapter have been grouped into seven subject areas. NUREG-0661 MARKIContainmentLongTermProgramSafetyEvaluationReport,NesolutionofGeneric Technical Activity A-7. February 1977 to December 1979. July 1980. ONRR GPO. NTIS. During testing for an advanced Boiling Water Reactor (BWR) containment system design (MARK-III), suppression pool hydrodynamic loads were identified which had not been considered in the original design of the MARK I containment system. To address this issue, a MARK I Owners Group was formed and the assessment was divided into a short-term and long-term program. The results of the NRC staf f's review of the MARK I Containment Short-Term Program are described in NUREG-0408. This report describes the results of the NRC staf f's review of the generic MARK I Containment Long-Term Program (LTP). The LTP was conducted te provide a generic basis to define suppression pool hydrodynamic loads and the related structural acceptance criteria, such that a comprehensive reassessment of each MARK I containment system would be performed. A series of experimental and analytical programs were conducted by the MARK I Dwners Group to provide the necessary bases for the generic load definition and structural assessment techniques. The generic methods proposed by the MARK I Owners Group, as modified by the NRC sta f f's requirements, will be used to perform plant-unique analyses, which will identify the plant modifications, if any, that will be needed to restore the originally intended margin of safety in the MARK I containment designs. North Anna Power Station Unit 2 Technical Specifications Appendix "A" to License KUREG-0664, Rev 1 No. NPF-7. August 1980. ONRR GPO. NTJS. The North Anna Unit 2 Technical Specifications, which were prepared by the U.S. . Nuclear Regulatory Commission, set forth the limits, operating conditions and other { requirements applicable to nuclear reactor f acility operation as set forth in 10 CFR 50.36 for the protection of the health and safety of the public.
Report No. Bibliographic Data NUkEG 0675, Supp 10 Safety Evaluation Report Related to Operation of Diablo Canyon Nuclear Power Station, Units 1 and 2, Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Supplement No. 10. August 1980. ONRR GPO. NTIS. Supplement No. 10 to the Safety Evaluation Report for Facific Gas and Electric Company's application for licenses to operate the Diablo Canyon Nuclear Power Station (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 Nuclear Regulatory Commission. The purpose of this Supplement is to discuss TM1-2-related requirements that must be met prior to fuel load of the Diablo Canyon facilities. NUREG-0678 The Effects of Temperature, Moisture, Concentration, Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon. August 1980. ACRS G PO . NTIS. This report is a critical review of the published literature on the adsorption of radioactive krypton and menon on activated charcoal, The report includes a tabulation and evaluation of the adsorption coefficients for these two gases as related to temperature, pressure, moisture, mass transfer effects, and the nature of the carrier gas. Whenever possible, the resulting data have been used to develop simple corre-lations for quantitatively evaluating the effects of these parameters on noble gas adso rpt ion , important conclusions of the study include the observations that (a) individual charcoals have a wide tange of adsorption coefficients and therefore the performance of a given bed is heavily dependent on the quality of the charcoal it contains; (b) because of the detrimental effects of mass transfer on noble gas adsorp-tion, consideration should be given to including this factor in developing technical specifications for adsorption beds; and (c) additional research is needed on the determination of the interrelationship of moisture and temperature and their ef fects on adsorption bed performance. NUREG-0679 Pipe Cracking Experience in Light-Water Reactors, 1967 through 1979. August 1980. OKRR GPO. NTIS. Commercial light-water reactors have experienced pipe cracking since 1965. This report summarizes pipe ersching experience in light-water reactors as reported in Licensee Event Reports from 1967 through 1979, other licensee and vendor reports, and Office of Inspection and Enforcement Bulletins. Pipe cracks which were environmentally induced, such as stress corrosion cracking of metal sensitized by welding and heat treatment, were most prevalent. Feedwater pipes experienced fatigue cracking from thermal stress and many small lines developed leaks sa a result of fatigue caused by vibration. Cracking incidents are separated into generic categories and listed by reactor type, pipe sise, and systems affected. NUREG-0684 Summary of Public Commenta and NRC Staff Analysis Relating to Rulemaking on Emergency Planning for Nuclear Power Plants. September 1980. OSD GPG. NTIS. This NUREG provides e summary and discussions of public comments received during the expedited rulemaking to upgrade emergency preparedness around nuclear power reactor sites. The final rule was published in the Federal Register (45 FR 55402) on August 19, 1980. The informatian in NUREG-0684 was excerpted in the main f rom internal paper SECY-80-275 (June 3, 1980) which forwarded the final rule to the Commission for consideration. This document, along with NUREG-0628, NUREG/CP-Doll, and the materials cited in the Final Rules, should be considered a compendium of the major issues raised in this proceeding and acted upon by the Commission. NUREG-0605 Environmental Assessment for Effective Changes to 10 CFR Part 50 and Appendix E to 10 CTR Part 50 Emergency Planning Requirements for Nuclear Power Plants. August 1980. OSD GPO. NTIS. The staf f of the U.S. Nuclear Regulatory Commission has prepared an Environmental Assessment for changes to the regulations governing emergency planning requirements. Based on this essessment, the Director, Of fice of Standards Development, determined
Report No. Bibliographic Data that an Environmental Impact Statement would not be prepared for the rule changes and directed that a " Negative Declaration; Finding of No Significant Impact" be prepared and published in the Federal Register. The Environmental Assessment is presented and ; the Federal Register Notice is attached as Appendix B. (Included in Appendix B is an analysis of comments received on an earlier draft version of this Assessment (45 FR 3913, January 21, 1980).) The effective rule changes are included as Appendix C for completeness. KUREG-0686 Final Environmental Statement Related to Primary Cooling System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Docket No. 50-010. October 1980. ONRR GPO. NTIS. The staff has considered the environmental impact and economic costs of the proposed primary cooling system chemical decontamination at Dresden Nuclear Power Station Unit I located in Grundy County, Illinois. This statement focuses on the occupational radiation exposure associated with the proposed Unit I decontamination program, on alternatives to chemical decontamination, and on the environmental impact of the dispesal of the solid radioactive waste generated by this decontamination. The staff ha' concluded that the proposed decontamination will not significantly affect the goality of the human environment. Furthermore, any impacts from the decontamination program are outweighed by its benefits. NUREG-0691 Investigation and Evaluation of Cracking lucidents in Piping in Pressurized Water Reactors. September 1980. GPO. NTIS. This report summarizes an investigation of known cracking incidents in pressurized-water-reactor plants. Several instances of cracking in feedwater piping in 1979, together with reported cases of stress corrosion cracking at Three Mile Island Unit 1, led to the establishment of the third Pipe Crack Study Group. Major differences between the scope of the third PCSG and the previous two are: (1) the emphasis given to systems safety implications of cracking and (2) the consideration given all cracking mechanisms known to affect PWR piping, including the failure of small lines in secon-dary safety systems. The present PCSG reviewed existing information on cracking of PWR pipe systems, either contained in written records or collected f rom meetings in the United States, and made recommendations in response to the PCSG charter questions and to other major items that may be considered to either reduce the potential for cracking or to improve licensing bases. KUREG-0692 Final Environmental Statement Related to Steam Generator Repair at Surry Power Station, Unit No. 1. Virginia Electric and Power Cmpany, Docket No. 50-280. July 1980. DNRR GPO. NTIS. l A Final Environmental Statement for the Surry Power Station Unit 1 Steam Generator Repair Program has been prepared by the Of fice of Nuclear Reactor Regulation. This Statement considers the environmental impacts and economic costs of the proposed steam generator repair at Surry Power Station Unit 1. The Statement focuses on the occupational radiation exposure associated with the proposed Unit I repair program t and on alternatives to reduce this exposure. It concludes that the proposed repair will not significantly affect the quality of the human environment. Furthermore, any impacts from the repair program are outweighed by its benefits. Also included are comments of Federal, State and local governmental sgencies and certain nongovernmental , organizations and individuals. KURE G-Ot96 Functional Criteria for Emergency Response Facilities. July 1980. ONRR GPO. NTIS. There are four facilities related to the function of mitigating nuclear reactor . abnormal situations: (1) The Safety Parameter Display System (SPDS), (2) the onsite Technical Support Center (TSC), (3) the Emergency Operations Facility (EOF), and (4) the Nuclear Data Link (NDL) providing information to the KRC Operations Center. This report delineates the f unctional criteria for these f acilities in sufficient detail and scope for licensee and other emergency preparedness planners to design an integrated emergency resource capability for inclusion in emergency plans. More i detailed acceptance criteria are being developed separately and will be issued in the future.
Report No. Bibliographic Data NUREG-0698 NRC Plans for Cleanup Operations at Three Mlle Island Unit 2. July 1980. ONRR G PO . NTIS. The objective of this NRC master plan is to define the functional role of the NRC in cleanup operations at Three Mile Island Unit 2 to assure that agency regulatory responsibilities and objectives will be fulfilled. The plan outlines NRC functions in THI-2 cleanup operations in the following areas: (1) the functional structure of NRC in i*e coordination with other government agencies, the public, and the licensee. (2) the functional roles of these organizations in cleanup operations, (3) the NRC review and decision-making procedure for the licensee's proposed cleanup operation, (4) the NRC/ licensee schedule of major actions, and (5) NRC's functional role in overseeing implementation of approved licensee activities. NUREG-0699 Comments on the NRC Safety Research Program Budget for Fiscal Year 1982. , July 1980. ! ACRS GPO. NTIS. Recommendations of the Advisory Committee on Reactor Safeguards are presented to the Commissioners for their consideration for IY82 budget for the NRC safety research program. NUREG-0702 Final Environmental Statement Related to the Operation of Gas Hills Uranium Project, j Docket No. 40-299, Union Carbide Corporation. I August 1980. ONMSS GPO. NTIS. A Final Environmental Statement for Union Carbide Corporation related to the renewal of Source Material License SUA-648 for the Gas Hills Uranium Project located in { Natrona County, Wyoming (Docket No. 40-299) has been prepared by the Office of Nuclear Material Safet) and Safeguards. This statement provides (1) a summary of environmental impacts and adverse ef fects of the proposed action and (2) a consideration of principal alternatives. Also included are comments of governmental agencies and other organiza-tions on the Draft Environmental Statement for this project, and staf f responses to these comments. The NRC has concluded that, af ter weighing the environmental, economic, technical, and other benefits of the Gas Hills Uranium Project against environmental and other costs and considering available alternatives, the action called for is renewal of the source material license, subject to stipulated conditions. NUREG-0703 Potential Threat to Licensed Nuclear Activities from Insiders (Insider Study). July 1980. ONMSS GPO. NTIS. Ine Insider Study was undertaken by NRC staf f at the request of the Commission. Its j objectives were to: (1) determine the characteristics of potential insider adversaries I to licensed nuclear activities; (2) examine security system vulnerabilities to insider adversaries; and (3) assess the ef fectiveness of techniques used to detect or prevent insider malevolence. The study analyzes insider characteristics as revealed in incidents cf thef t or sabotage that occurred in the nuclear industry, analogous industries, government agencies, and the military. Adversa ry chracteristics are grouped into four categories: position-related, behavioral, resource and operational. It also analyzes (1) the five security vulnerabilities that most frequently accounted for the success of the insider crimes in the data base; (2) the 11 means by which insider crimes were most of ten detected; and (3) four major and six lesser methods aimed at preventing insider malevolence. In addition to case history information, the study contains data derived f rom non-NRC studies and from interviews with over 100 security experts in industry, government (federal and state) and law enforcement. KUREG-0706, Vol 1 Final Generic Environmental Impact Statement on Uranium Hilling Project M-25: Volume 1 - Summary and Text. September 1980. ONMSS GPO. NTIS. The Final Generic Environmental Impact Statement (CEIS) on Uranium Milling focuses primarily upon the matter of mill tallings disposal. It evaluates both the costs and benefits of alternative tailings disposal modes and draws conclusions about criteria which should be incorporated into regulations. Both institutional and technical controls are evaluated. Health impacts considered were both short- and long-term. Restatement and resolution of all public comments received on the draft (GEIS) are presented. There are three volumes: Volume I is the main text and Volumes II and III are supporting appendices.
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Report No. Bibliogrrphic Data NUREG-0706, Vol 2 Final Generic Environmental Impact Statement on Uranium Milling Project M-25: Volume II - Appendices A-F. September 1980. ONMSS GPO. NTIS. The Final Generic Environmental Impact Statement (GEIS) on Uranium Milling focuses. primarily upon the matter of mill tailings disposal. It evaluates both the costs and benefits of alternative tailings disposal modes and draws conclusions about criteria which should be incorporated into regulations, Both institutional and technical controls are evaluated, Health impacts considered were both short- and long-term. Restatement and resolution of all public comments received on the draft (GEIS) are presented. There are three volumes: Volume I is the main text and Volumes 11 end III are supporting appendices.
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NUREG-0706, Vol 3 Final Ceneric Environmental Impact Statement on Uranium Milling Project'H-25: Volume III - Appendices G-V. , September 1980, i DNMSS GPO. NTIS, The Final Generic Environmental Impact Statement (GEIS) on Uranium Hilling focuses- . primarily upon the matter of mill tailings disposal. It evaluates both the costs and L benefits of alternative tailings disposal modes and draws conclusions about criteria which should be incorporated into regulations. Both institutional and technical i controls are evaluated. Health impacis considered were both short- and long-term. Restatement and resolution of all public comments received on the draf t (GEIS) are presented. There are three volumes: Volume I is the main text and Volumes II and j III are supporting appendices. NUREG-0715 Task Force Report on Interim Operation of Indian Point. August 1980. ORES GPO. NTIS. On May 30, 1980 the Commission issued an order establishing a four pronged approach-for resolving the issues raised by the Union of Concerned Scientists' petition regarding the Indian Point nuclear facilities. Among other things a Task Force on Interim Operation was established to address the question of whether Indian Point Units 2 and - 3 should or should not be allowed to operate during the pendency of a planned adjudi-cation. Specifically, the Task Force report deals with two major issues. The first issue relates to accident risk as a function of population density and distribution around the plant. New York City is less than 50 miles to the south of the Indian Point site. The Task Force compared Indian Point risks (e.g., health impacts and property damage) with those of other reactor sites and designs, distinguishing between the effects of population densities and of design and other factors. Secondly, the Task Force examined the economic, social and other "nonsafety" ef fects of shutting down or reducing the power levels of either or both reactors. In particular, the Task Force compared projected peak demands for energy with projected available capacity to determine if reducing power levels at Indian Point would affect system reliability l' in the summer of 1980 l NUREG-0721 Acceptance Criteria for the Physical Protection Upgrade Rule Requirements for Fixed Sites. September 1980. ONMSS GPO. NTIS. I This document has been developed as a tool to assist in providing consistent evaluation l of upgraded physical security plans submitted in response to the Physical Protection Upgrade Rule, effective March 25, 1980. It presents a means for assuring licensee , t compliance with every regulatory requirement of particular significance to the protec-Lion of the public health and safety. Acceptance criteria are included to determine the extent to which each licensee meets the regulatory requirements. The format parallels Regulatory Guide 5.52, " Standard Format and Content of a Licensee Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites (Other than Nuclear Power Plants)." l t l
i Report No. Bibliogrrphic Dsta NUREG-0722 The Effects of Natural Phenomena on the Exxon Nuclear Company Mixed Oxide Fabrication Plant at Richland, Washington. September 1980. ONMSS GPO. NTIS. An Analysis of the Effects of Natural Phenomena on the Exxon Nuclear Company Mixed i j Oxide Fabrication Plant at Richland, Washington has been prepared by the Office of 1 l Nuclear Material Safety and Safeguards. The analysis is in support of the Special Nuclear Materials I.icense held by the subject company. It addresses the probable effects of damage to the Exxon Nuclear Company Mixed Oxide Fabrication Plant by severe weather and earthquake and expresses the consequence of damage as dose to several human receptors. The doses that result from facility damage are multiplied by the occurrence rate for the initiating event yielding the yearly risk. NUREG-0727, Add. Final Environmental Statement Related to the Operation of the Joseph M. Farley Nuclear Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364. September 1980. ONRR GPO. NTIS. This addendum to the Final Environmental Statement addresses the environmental dose commitments and health effects from fuel cycle releases, fuel cycle socioeconomic impacts, and possible cumulative impacts pending further treatment by rulemaking. NUREG-0728 Report to Congress: NRC Incident Response Plan. September 1980. l OIE GPO. NTIS. l 1 The Nuclear Regulatory Conraission (NRC) regulates civilian nuclear activities to protect the public health and safety and to preserve environmental quality. An Incident Response Plan has been developed which assigns responsibilities for responding j to any potentially threatening incident involving NRC licensed activities and for l assuring that the NRC will fulfill its statutory mission. l NUREG-0729 Report to Congress on NRC Emergency Communications. l September 1980. j OIE GPO. NTIS. I The accident at Three Mile Island highlighted the need for improved communications among the NRC and other organizations which respond to such emergencies. ibis report summarizes the communications problems identified by several major review groups after the accident, the status of corrective actions, and NRC plans to improve communications still further. NUREG-0730 Report to Congress on the Acquisition of Reactor Data for the NRC Operations Center. September 1980. OIE GPO. NTIS. I l This report considers alternative methods for transmission of data from operating l , nuclear reactors to the NRC Operations Center in order for NRC to carry out its ! l responsibilities in a nuclear emergency. The report considers the spectrum of roles NRC will play, discusses the various alternatives and describes in detail one data link concept which could meet the NRC data requirements. In addition, the report 4 considers the data link in relation to other required emergency facilities and presents an implementation plan and schedule for an automatic system, if a decision is made to : proceed. NUREG-0732 Answers to Frequently Asked Questions about Cleanup Activities at Three Mile Island, Unit 2. September 1980. I ONRk GPO. NTIS. J This document presents answers to frequently asked questions about plans for cleanup and decontamination activities at Three Mile Island Unit 2. Answers to the questions asked ere based co information in the NRC " Draft Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting ! from March 28, 1979 Accident, Three Mile Island Nuclear Station, Unit 2," NUREG-0683. I J
l l l Sequential List of NRC Contractor-Generated Reports Designated NUREG/CR l i l l l l I l 1 l
Report No. fibliogrephic Data FUREG/CR-0055, Vol 2 Verification of Subsurface Conditions at Selected " Rock" Accelerograph Stations in California. September 1980. Shannon & Wilson; Agbabian Assoc. ORES GPO. NTIS. Second in a series of reports to investigate the subsurface soil conditions at accelerograph stations which have been categorized by other researchers as " rock"~ sites. Contained in this volume are our findings of the site conditions at 29 accelerograph stations in California. Subsurface conditions at the sites were investi-At three. gated with a geologic reconnaissance and a review of available boring data. sites, where boring data was not available, a test hole was drilled to better define the depth to rock. Of the 29 " rock" sites that were investigated, less than half could be verified as being founded on or within about 20 feet of rock. This would imply that over half of the stations are really soil sites. KUREG/CR-0055, V. 2 App Verification of Subsurf ace Conditions at Selected " Rock" Accelerograph Stations in California. j September 1980. Shannon & Wilson; Agbabian Assoc. ORES GPO. NTIS. Appendix to the second in a series of reports to investigate the subsurface soll conditions at accelerograph stations which have been categorized by other researchers as " rock" sites in California. Appendix to Volume 2 contains the accelerograms which have been recorded at the sites investigated. Subsurface conditions at the sites were investigated with a geologic reconnaissance and a review of available boring data. At three sites, where boring data was not available, a test hole was drilled to better define the depth to rock. Of the 29 " rock", sites that were investigated, less than half could be verified as being founded on or within about 20 feet of rock. This would imply that over half of the stations are really soil sites. NUREG/CR-0055, Vol 3 Verification of Subsurface Conditions at Selected " Rock" Accelerograph Stations in California. September 1980. Shannon & Wilson and Agbabian Assoc. ORES GPO. NTIS, This report is the third in a series presenting geotechnical information for accelero-graph stations in California that have been classified as " rock" sites by one or more investigators. This volume discusses the findings at five locations which are in the vicinity of nine accelerograph stations. Although each of these stations was originally discussed in one of the earlier reports in this series, the available information on . the subsurface conditions at each site was insufficient for determining the dynamic ~ properties of the subsurface soils or the depth to rock. Consequently, this study was performed to supplement these previous reports with more detailed subsurf ace information based on deep borings, downhole geophysical measurements and laboratory tests. NUREC/CR-0200 SCALE: A Modular Core System for Performing Standardized Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSASI and CSAS2. August 1980. Oak Ridge National. Lab ORKL/NUREC/CSD-2, Vol 1 ORES GPO! KTIS. I Under contract with the Nuclear Regulatory Commission, the Computer Sciences Division at Oak Ridge National Laboratory has developed the SCALE system for performing Standardized Computer Analyses for Licensing Evaluation of nuclear systems. The SCALE systen includes a number of selected data libraries as well as various calcula-tional modules for performing criticality, shielding and heat transfer analys,es. This document describes the CSASI and CSAS2 control modules which shield the group-averaged cross-section data for the given situation and perform a oce-dimensional discrete nrdinates or multidimensional Monte Carlo calculation to obtain the effective neutron multiplication factor (k-eff) for the configuration described by the user. l
Report No. Bibliographic Data NUREG/CR-0453 Rev i SIMMER-II: A Computer Program for LMFBR Disrupted Core Analysis. July 1980. Los Alamos Scientific Lab LA-7515-M, Rev ORES GPO. NTIS. Physical models, numerical methods, and program description are presented for SIMMER-II, e computer program to predict the neutronte and fluid dynamic behavior of an LMFBR during a hypothetical core-disruptive accident. Either the time-dependent multigroup neutron diffusion or neutron transport equation is solved by the quasistatic method to predict reactor neutronic behavior during the accident. Cross sections depend on temperature and background cross sections.. The structure, liquid, and vapor fields are modeled to predict the fluid-dynamic behavior of the reactor. Each field consists of density components to follow the material motion and energy components to predict the material temperatures. For typical accident calculations, the materials are fertile fuel, fissile fuel, stainless steel, sodium, control material, and fission gas. Heat, mass, and momentum transfer are calculated among the three fields and theilt components. NUREG/CR-0508 Security Communication Systems for Nuclear Fixed-Site facilities. July 1989. Y-12 Plant, Oak Ridge Y/DW+128 OSD GPO. NTIS. l 1 l' This report presents a baste discussion of communication techniques and factors relevant to designing communication systems for nuclear fixed site facility security systems. The reader is provided cosmiunication fundamentals, design considerations, , and specification techniques. Copious references and an annotated bibliography are l provided for individuals who desire to delve deeper than the limits and areas of I study of this report. Ease of reading and use of this report are enhanced by rele-gating detailed communication design treatise to the Appendices. Sample procurement specifications are provided throughout the report for various communication system components and are distinguished f rom the regular text by using a smaller type. NUREC/CR-0603 A Risk Assessment of a Pressurized Water Reactor for Class 3-8 Accidents. September 1980. Brookhaven National Lab BNL-NUREG-50950 ORES GPO. NTIS. An assessment has been made of the impact on societal risk of Class 3-8 accident sequences as defined by Appendix D to 10 CFR 50. The present analysis concentrates on a pressurized water reactor and utilizes realistic assumptions when practical. Conclusions are drawn as to the relative importance of the analyzed accidents and their impact on the development of a complete societal risk curve. NUREC/CR-0720 LWR Pressure Vessel Irradiation Surveillance Dosimetry Quarterly Progress Report, October-December 1978. July 1980. Hanford Engineering Develop. Lab HEDL-TME-79-IS ORES GPO. NTIS. This report describes progress made in the Light Water Reactor Pressure Vessel . Irradiation Surveillance Dosimetry Program during October-December 1978. The primar) I objective of the program is to prepare an updated and improved set of dosimetry, damage correlation, and associated reactor analysis ASTM Standards for LWR-PV irradia - tion surveillance programs. Supporting this objective are a series of analytical and l experimental validation and calituation studies in " Standard, Reference, and Control *4ed j Environment Benchmark Fields," reactor " Test Regions," and operating power reactor (
, " Surveillance Positions." I i
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l I Report No. Bibliographic Data NUREG/CR-0723 Calculations of the Skyshine Gamma-Ray Dose Rates from Independent Spent fuel Storage Installations (ISFS1) Under Worst Case Accident Conditions. September 1980. , Oak Ridge National Lab ORKL/KUREG/TM-316 ONMSS GPO. NTIS. Calculations of the skyshine gamma-ray dose rates from three spent fuel storage pools under worst case accident conditions have been made using the discrete ordinates code DOT-IV and the Monte Carlo code NORSE and have been compared to those of two previous methods. The DNA 37N-210 group cross section library was utilized in the calculations, together with the Claiborne-Trubey gamma-ray dose factors taken f rom the library. 5 Plots of all results are presented. It was found that the dose.was a strong function of the iron thickness over the spent fuel assemblies, the initial angular distribution of the emitted radiation, and the photon source near the top of the assemblies. NUREC/CR-0742 Review and Integration of Existing Literature Concerning Potential Social Impacts of > Transportation of Radioactive Materials in Urban Areas. July 1980. Rice Univ; Univ of Texas; Sandia Natl. Lab SAND 78-7017 OSD GPO. NTIS. The symbolic interactionist/ collective behavior approach within sociology is applied to the transport of radioactive materials through urban environs, indicating that social impacts of such transport would extend far beyond objectively measurable radiological impacts of normal (incident free) transport, accidents during transport (with or without radiation release) or diversion by terrorists. This approach is t used to delineate the major cultural frames of reference that interested publics and special groups might use in interpreting events surrounding radioactive material ! transport, and to specify probable social impacts of seven scenarios. These impacts include: (1) uncertainty, fear and mistrust, (2) processes, (3) initial agency responses, (4) subsequent collective behavior responses, and (5) a wide range of more general impacts on U.S. culture and special structure. NUREG/CR-0744 Identification and Assessment of the Social Impacts of Transportation of Radioactive Materials in Urban Environments. July 1980. Battelle Human Affairs; Sandia National Lab SAND 79-7032,. B-HARC-411-049 OSD G PO . NTIS. This study provides an assessment of the range of social impacts that have occurred or may result from the transportation of radioactive materials through urban environ-ments. These impacts are identified by retrospectively examining representative cases and reviewing selections from the theoretical and the technical literatures on social impacts. The likelihood of social impacts occurring in the future is assessed on the basis of observed impacts to date. The study analyzes five categories of l social impacts (psychological, sociological, political, legal, and organizational) l resulting f rom four causative transportation events (incident f ree, vehicular accidents, [ human errors and deviations from accepted quality assurance practices, and malevolent l acts). The methods of analysis include evaluations of public opinion surveys, selected literature reviews, analysis of selected case studies, interviews with knowledgeable officials and citizens, legal analyses, review of relevant informed professional judgement. This study concludes that political and legal impacts have been more substantial than psychological and sociological impacts, and for a given magnitude of physical consequences, radioactive materials transportation has greater social impacts l than hazardous materials transportation generally. Malevolent acts, among the four causative event categories, appear to hold the greatest potential for severe social impacts. I l NUREG/CR-0761 Extended Analysis of Data from the 1/5-Scale MARK I Boiling Water Reactor Pressure Suppression Experiment. July 1980. Lawrence Livermore Lab UCRL-52707 ORES GPO. NTIS. An extensive analysis of data from the 1/5-scale MARK I BWR Pressure Suppression Experiment (PSE) air transient tests has been completed. Primary focus was placed on
Report No. Bibliographic Data computing a best estimate of the hydrodynamic vertical load function (HVLF) and determining the associated peak forces and their standard error. These results were then applied to develop the sensitivity of the HVLF to various major parameters (for example, drywell pressurization rate), to evaluate the impulse of the HVLF, and to analytically model the response vertical load function (RVLF). In addition, a complete = evaluation of the enthalpy flux distribution in the vent system was provided for each test Finally, pool swell dynamics were quantified for a subset of the test series and correlated to the observed ringheader strut loads. NUREG/CR-0893 Acute Toxicity and Bioaccumulation of Chloroform to Four Species of Freshwater Fish. August 1980. Battelle pacific Northwest Lab PNL-3046 ORES GPO. NTIS. Acute toxicity of chloroform to four species of freshwater fish was studied in flow-through 96-hr toxicity tests. Chloroform is toxic to fish in the tens of parts per million, a concentration well above that which would be expected to be produced under normal power plant chlorination conditions. Investigations of acute toxicity of chloroform and the bloaccumulatinn of chlorinated compounds in tissues of fish revealed differences in tolerance levels and tissue accumulations. Mean 96-hr LCKO' for chloroform were 18 ppm for rainbow trout and bluegill, 51 ppm for largmaouth 5 ass and 75 ppm for channel catfish. Mortalities of bluegill and largemouth bass occurred during the first 4 hr of exposure while rainbow trout and channel catfish showed initial tolerance and mortalities occurred during the latter half of the 96-hr exposure. Rainbow trout had the highest level of chloroform tissue accumulation (7 mg/g tissue) and catfish the second highest (4 mg/g tissue) followed by bluegill and largemouth bass which each accumulated about 3 mg/g tissue. Accumulation of chloroform was less than one order of magnitude above water concentrations for all species. NUREG/CR-0961 Qualification Test Results on 1550*C and 2200*C I/16-Inch 0.D. Fuel Centerline Thermocouples for the LOFT program, September 1980. Hanford Engineering Develop. Lab KEDL-TME-79-50 ORES CPO. NTIS. The technology and commercial vendors have been developed for fabrication of thermo-couples to measure fuel centerline temperatures to 2200'C in the LOFT reactor. Two model A and one model B qualification thermocouples satisfied all test requirements during life tests at 2200'C and 1550*C. The est output drifted less than 2% during 400-hour tests at the maximum test temperatures of 2200'C and 1550*C, Measurement performance remained unimpaired af ter 145'C/s transient survival tests. The thermo-couples did not meet the time response requirement of one second. Time responses of 4\ seconds at 1550'C and 24 seconda at 2200*C were measured. However, this result was not considered too negative to preclude useful temperature measurement of fuel centerline temperatures in the LOFT reactor. The first qualification thermocouples satisfied all other test requirements. NUREG/CR-0985, Vol 3 Geotechnical and Strong Motion Earthquake Data from U.S. Accelerograph Stations. September 1980. Shannon & Wilson; Agbabian Assoc. ORES GPO. NT!S. This is the third in a series of reports to investigate the subsurface soil conditions j at selected accelerograph stations. Contained in this volume are the findings of ; site conditions at six accelerograph stations in the western U.S. These stations are I located at Gilroy, California; Logan, Utah; Bozeman, Montana; Tacoma, Washington; and two sites in Helena, Montana. Subsurface conditions at the first four sites were ! investigated by means of deep borings, field geophysical testing, and laboratory l testing of soil samples retrieved from the borings. Investigation of the Helena sites consisted of only a geologic reconnaissance since both accelerograph stations are founded on rock, i f i I l I
l Report No. Bibliogrephic Dsta NUREG/CR-0985, Vol 4 Geotechnical and Strong Motion Earthquake Data from U.S. Accelerograph Stations. Septemb'er 1980. Shannon & Wilson; Agbabian Assoc. ORES GPO. NTIS. This is the fourth in a series of reports presenting geotechnical and seismic data for selected accelerograph stations. This volume discusses the findings at five stations, one each in the cities of Anchorage, Alaska; Seattle, Washington; Olympia, Washington; and two in Portland, Oregon. This report contains information for each site describ-ing the station building and instrumentation, geology and seismology of the area, and site conditions. Deep borings, downhole geophysical measurements, and laboratory tests were conducted at each station, except Olympia, to evaluate the subsurface conditions. Since subsurface data was already available for Olympia, field and laboratory testing was not conducted for this study. NUREG/CR-0985, Vol 5 Geotechnical and Strong Motion Earthquake Data from U.S. Accelerograph Stations. September 1980. Shannon & Wilson and Agbabian Assoc. ORES GPO. NTIS. This is the fifth in a series of reports presenting geotechnical and seismic data for selected accelerograph stations. This volume discusses the findings at five stations, one each in the following locations: Fairbanks, Alaska; Petrolfa, California; Hollister, California; Los Angeles, California; and New Madrid, Missouri. This report contains l information for each site describing the station building and instrumentation, geology l and seismology of the area, and site conditions. Deep borings, downhole geophysical measurements, and laboratory tests were conducted at each station to evaluate the subsurface conditions. Reactor Safety Research Programs. Quarterly Report - July-September 1979. NUREG/CR-1009 July 1980. Battelle Pacific Northwest Lab PNL-3040-3 ORES GPO. NTIS. This document summarizes the work performed by Pacific Northwest Laboratory from July through September 1979 for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Each program is considered separately and is discussed according to major tasks or topics, depending on the nature of the project. NUREG/CR-1038 Dynamic, Inelastic Buckling Analysis of Mark I Torus Support Columns. August 1980. Lawrence Livermore Lab URCL-52723 ONMSS GPO. NTIS. Columns that support the MARK I BWR containment tori are subjected to short-duration To accurately predict the actual dynamic loads during some accident conditions. response under these conditions, an analysis must incorporate the dynamic behavior of the columns. Two different analysis methods were used in an effort to solve this dynamic, inelastic buckling problem. Neither method adequately solved the problem. The finite element approach gave unstable results for short-duration dynamic luad pulses. The finite difference approach did notResults incorporate the plasticity effects that were obtained, however, needed to accurately predict column response. indicated that little additional load capaetty could be realized for the combinations The of dynamic and static loads, and for the column configurations considered. results presented in this report indicate that design modifications will be required either to reduce the magnitude of the dynamic load or to increase the strength of the torus support columns. Acute Ef fects of Inhalation Exposure to Uranium Hexafluoride and Patterns of Deposition. NUREG/CR-1045 UF6/UO 2 F, Studies in Experimental Animals. August 1980. University of Rochester , ORES GPO. NTIS. ! and KF toxicity I This interim report describes the animal research completed on UO,F 2 following inhalation, intratracheal instillation and intravenous Injection during the l first year of a continuing study. The principal impetus to the study was a needed l
Report No. Bibliographic Data evaluation of exposure or intake parameters with respect to urinary excretion levels and renal retention values for uranium inasmuch as these interrelationships are critical to the currently advocated bioassay procedures for uranium workers. Results to date support the relation between absorbed dose and urinary elimination rate pronosed by the ICRP for +6U compounds, indicate pulmonary retention of UO F, is extremely short (half-time hours), and suggest the threshold absorbed dose2 f6r producing renal injury is of the order of 10 mg kg 1 body weight. In dogs and rats, glucosuria seems to be a more sensitive indicator of renal dysfunction than albuminuria or plasma urea nitrogen. levels. but present results are equivocal.Possible syncrgism between UO 2 F, and HF is under study Retention functions for kidney and bone await further experiments in both species. KUREG/CR-1048 Technical Safeguards lasues for Alternative Fuel Cycles. September 1980. Brookhaven National Lab DNL-NUREG-51182 ONMSS GPO. NTIS. This study involved a preliminary analysis and assessment of the safeguardability of 21 alternative Assessment fuel cycles proposed under the Nonproliferation Alternative Systems Program (NASAP) of the Department of Energy for technical safeguards issues and problems that might affect regulation or licensing. The approach adopted was to identify generic features, common to two or more fuel cycles, and assess these independently of the fuel cycle in which they are involved. Then the individual fuel cycles were reviewed in order to identify additional unique features -- i.e. , those associated with a single fuel cycle only. NUREG/CR-1071 Critical Experiments with Interstitially-Moderated Arrays of Low-Enriched Uranium Oxide. Topical Report on Reference Critical Experiments. September 1980. Rockwell Internat. Rocky Flats Plant RFP-3008 ORES GPO, NTIS. The critical separation between two tables supporting arrays and cans containing low-enriched uranium oxide has been measured for twenty-one (21) reflected configura-tions having interstitial layers of moderating material between cans. The critical separation varied between 0.23 and 1.84 cm. *" '* 4.46% U-235, compacted to a density of 4.7 g/cm3, The uranium oxide (U3 and adjusted to an8)11/U atomic ratio of 0.77 by the addition of water. Each can weighs 16 kg and is a 15.3 cm cube. Interstitial plastic moderator 1.0, 1.3, or 2.5 cm thick separates cans of the three-dimensional array. Some experiments include thin sheets of neutron absorbing materials, such as mild steel or polyvinyl chloride, surrounding each can. Arrays are closely reflected by thich cuboidal shells of plastic or concrete. The parameter varied to achieve criticality is the number of cans in the array. The smallest number of cans (40) occurs with 2.5-cm-thick moderator, no absorber, and concrete reflector. The largest reflectors.(100) occurs for several combinations of absorber and moderator in both For otherwise similar configurations, concrete is the better reflector in all cases. KUREG/CR-1109 Measurement of Radon Diffusion from Uranium Mill Tailing Piles. March 1980. Battelle Pacific Northwest Lab PNL-3187 ORES GPO. NTIS. The concentrations of Ra-226 and Rn-222 (Pb-214) were measured as a function of depth within a uranium mill tailings pile by in-situ gamma-ray spectrometry. Radon diffusion and exhalation rates were determined from the concentration gradients by employing an integral solution distribution of parent of the the dif fusion equation that accommodates a nonuniform depth radium. Radon diffusion coefficients of 0.0002 and 0.0017 cm2/sec, and exhalation rates of 60 and 275 atoms /cm2/see were determined for two locations with differing soil moisture content. i l i
Bibliographic Data Report No. A Preliminary Assessment NUREG/CR-lll3 Licensability of CANDU-Type Reactors in the United States. of the R and D Requirements. August 1980. University of California UCLA-ENG-7853 ONRR GPO. NTIS. An assessment is provided of the R&D required to establish the licensability of a CANDU-type reactor in the U.S. It is shown that the bulk'of the R&D effort shoulf establish the integrity of the pressure tubes and the effects of the pressure tubs f ailure on the remainder of the system. Three possible R&D program options are defined and discussed; it is concluded that one of these options is likely to require less R&D than the other two. The principle underlying this option is that the pressure tubes would be shown to have a moderately low probability of sudden, gross failure and that the effects of a single failure would not lead to unacceptable consequences. In other areas where R&D work would be necessary, more of the problems would be similar to those encountered in LWRs; however, two novel problems are identsfied, viz: (a) investigation of the effectiveness of the moderator as an alternative emergency cooling system and (b) the ef fect of the difference in reactor configuration (horizontal heat source) on natural circulation. Dverall, it is concluded that a relatively small amount of additional R&D should be sufficient to support a license application to build a CANDU-type reactor in the U.S. NUREG/CR-lll9 Piping Inelastic Fracture Mechanics Analysis. July 1980. Naval Research Lab NRL Memo Rpt 4259 ONRR GPO. NTIS. This report summarizes the results and conclusions of Tasks 1 and 2 of the study on
" Piping Inelastic Fracture Mechanics Analysis." In these tasks, available experimental data and the analytical methods for predicting rupture of LWR piping are assembled and assessed. The analytical techniques investigated can be catalogued into three major grouper structural response, semiempirical methods, and the J-controlled growth approach. A lenk-before-break condition is also investigated.
Seismic Safety Margins Research Program (Phase 1). Progress Report No. 7. NUREG/CH-il20, Vol 3 August 1980. Lawrence Livermore National Lab ORES GPO. NTIS. This document is a progress report on the Seismic Safety Margins Research Program (SSHRP) covering the period April 1, 1980 through June 30, 1980. .The report gives a general description of the program, together with financial summaries and individual project details. Each project is summarized to show accomplishments, schedules, milestones and completion dates, budget and expenditures, and any concerns that may affect the project. NUREG/CR-Il66 COPS Model Estimates of LLEA Availability Near Selected Reactor Sites. July 1980. Sandia Lab, Livermore ORES GPO. NTIS. The COPS computer model has been used to estimate local law enforcement agencyThe (LLEA) officer availability in the neighborhood of selected nuclear reactor sites. results of these analyses are presented both in graphic and tabular form in this report. NUREG/CR-1184 Evaluation of Simulator Aiequacy for the Radiation Qualification of Safety Related Equipment. July 1980. Sandia Lab, Albuquerque SAND 79-1787 ORES GPO. NTIS. Radiation qualification will be addressed in this report in order to evaluate the adequacy of radiation simulators typically used in qualification test simulations. " Adequacy" Where possible, discussion of combined environment effects will be made. need not be based on one-to-one correspondence of the actual radiation signature with i
Report No. Bibliographic Data a simulator signature, although that would be sufficient to assure adequacy. Instead, adequacy is to be judged on the basis of equivalence of equipment " damage" as a result of the exposure; under that definition of adequacy, the radiation signatures may not be identical but the damage (and damage mechanisms) must be quite similar.
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NUREG/CR-1185 COMPARE-MOD 1 Code, Addendum 1. August 1980. Los Alamos Scientific Lab IA-7199-MS, Add 1 ONRR GPO. NTIS. The COMPARE-MOD 1 code has been extended to incorporate an accounting for loss coefficient detail, calculation of forces and moments, and plotting of calculated results. The loss codificient detail feature includes a complete breakdown of the loss coefficient components, which facilitates checking the input, the calculation of the f riction component, and the summation of the components to provide the total loss , coefficient. The force-moment capability is based on a general orthogonal cartesian coordinate system that allows pressure-bearing surfaces of arbitrary orientation and location. Plotting is based on the DISSPLA system and features the convenient plotting of key parameters such as pressures, forces and moments, and plotting of any code variable by means of special plotting procedures. NUREG/CR-il92 Material Accounting as Required by the United States Nuclear Regulatory Commission Capabilities and Vulnerabilities. September 1980. . Lawrence Livermore National Lab UCRL-52734 > ORES GPO. NTIS. The Lawrence Livermore National Laboratory has undertaken to assist the U.S. Nuclear ' Regulatory Commission in upgrading its material accounting (MA) regulations. As part of this effort, this report evaluates and critiques the current status of material accounting as required by Parts 70.51 through 70.59 of the Code of Federal Regulations. Through the development and assessment of a generic, minimal MA system, the capabilities of the MA system have been delineated, and the vulnerabilities of the MA system to deliberately induced system failures (i.e., tampering) have been determined. NUREG/CR-Il95 Measurement of XE-133 C-14 and Tritium in Air and I-!31 in Vegetation and Milk Around 1 the Quad Cities Nuclear Power Station. July 1980. Exxon Nuclear; Idaho National Engineering Lab ENICO-1023 , ORES GPO. NTIS. [ The objective of this program was to evaluate the air-grass-cow-milk model for predicting the 1-131 dose from normal operations at an actual LWR. The Quad Cities Nuclear Station was selected as the candidate site and measurements were made for a three- ' month period. The station gaseous ef fluent release rate was determined for 1-131, Xc-133 Kr-85, C-14, and 11-3. The chemics! species of the released lodine and the oxidization state of the released C-14 and H-3 were determined. In the environment, measurements were made of the 1-131 concentration in air at five sites. At two of these sites, determinations were made of the 1-131 concentration in vegetation, milk, and precipitation and the Xe-133, C-14, and X-3 concentrations in air. NUREG/CR-1198, Vol 1 Design Guidance and Evaluation Methodology for Fixed Site Physical Protection Systems, Volume 1. July 1980. Sandia National Lab, Albuquerque SAND 79-2378/l ONMSS GPO. NTIS. Design guidance products and a system performance evaluativi methodology bave been developed to aid the Nuclear Regulatory Commission in the implementation of new regulations designed to upgrade the physical protection of nuclear fuel cycle facilities. The evaluation methodology which incorporates the design guidance products, provides a means of arriving at an overall measure of performance for each capablFity required ! in the regulations. To arrive at this measure of performance, first the scores associated with responses to a series of equipment and procedure quessis marsres are ) aggregated. The aggregation of scores then proceeds through successive levels of a I hierarchical structure developed for each capability.
_ . _ . . . - _ ._. ~ _ . . _ . _ _ Heport No. Bibliographic Data , NUREG/CR-Il98, Vol 2 Design Guidance and Evaluation Methodology for Fixed Site Physical Protection Systems, Volume 2 July 1980. Sandia National Lab, Albuquerque SAND 79-2378/2 ONMSS Gp0. NTIS. Design guidance products and a system performance evaluation methodology have been developed to aid the Nuclear Regulatory Commission in the implementation of new regulations designed to upgrade the physical protection of nuclear fuel cycle facilities.
- The evaluation methodology which incorporates the design guidance products, provides [
a seans of arriving at an overall measure of performance for each capability required in the regulations. To arrive at this measure of performance, first the scores associated with responses to a series of equipment and procedure questionnaires are aggregated. The aggregation of scores then proceeds through successive levels of a hierarchical structure developed for each capability. NUkEG/CR 1262 Risk Methodology for Geologic Disposal of Radioactive Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. July 1980. Sandia Lab, Albuquerque (
~
SAND 80-0157 ORES GPO, NTIS. A method for is.jucing a desired rank correlation matrix on a multivariate input ran-dom variable is introduced in this paper. This method is simple to use, is , dist ribution-free, preserves the exact- form of the marginal distributions on the input variables, and may be used with any type of sampling scheme for which correlation of input variables is a meaningful concept. A small simulation study provides an estimate of the bias and variability involved in the method. Input variables used in a model j for study of geologic disposal of radioactive waste provide an exangle of the usefulness of this procedure. NUREG/CH-1277 Shock Environments for Large Shipping Containers During Rail Coupling Operations. August 1980. Sandia National Lab, Albuquerque SAND 79-2168 ORES GPO. NTIS. Sandia National Laboratories participated in a study to define the shock environments to which large, fissile material shipping containers may be exposed during rail-coupling operations. Tests were conducted using impact velocities up to 17.98 km/h (11.17 mph). The cargo on the rail cars consisted of a 36-tonne (40-ton) cask mounted on a skid or a 64-tonne (70-ton) cask. The rsil cars were equipped with either standard draf t gear, hydraulic end-of-car draf t gear, or a sliding center sill cushion underf rame, The maximum peak acceleration and its pulse duration were determined for the longitudinal, transverse, and vertical axes of the two casks. NUREG/CR-1282 Statistical Analysis of Steam Generator Inspection Plans and Eddy Current Testing. August 1980. Sandia National Lab, Albuquerque SAND 79 2300 ORRR GPO. NTIS. r Methods are given for choosing steam generator inspection plans which have specified statistical properties. The methods include adjustments for the possible effect of eddy current testing measurement error. Available measurement error data are analyzed and statistical methods are given for analyzing the results of in-service inspection l of steam generators. To evaluate candidate inspection plans, a computer simulation is proposed and illustrated. Additionally, an exponential / double exponential probabil-ity model for the distributions of degradation and measurement error is developed and ; used in a parametric study of some statistical properties of inspection plans.
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e geport No. Bibliographic Data NUREG/CR-1289 - Evaluation of Isotope Nigration-Land Burial Water Chemistry at Commercially Operated ' Low Level Radioactive Waste Disposal Sites.
-July 1980.
Brookhaven National Lab BNL-FUREG-51143 ORES GPO. NTIS. In cooperation with the U.S. Ocological Survey, a field and laboratory program was initiated to study the existing commercial low-level radioactive waste disposal-sites. This investigation will provide source term data for radionuclides and other solutes in trench waters at the sites and will describe the physical, chemical, and biological properties of the geochemical system that control the movement of radio-nuclides. This study was also initiated to provide experimental research support to i the U.S. Nuclear Regulatory Commission for development of criteria for the selection and licensing of solid low-level radioactive waste disposal altes. The disposal' sites sampled to date are located at West Valley, New York! Maxey Flats, Kentucky; Barnwell, South Carolina; and Sheffield, 1111ools. Procedures for the collection, preparation and analysis (particularly under anaerobic conditions) of trench waters were developed, when necessary, to supplement standard procedures. Inorganic, organic, and radiochemical constituents in trench waters are measured and their relevance to movement of radionuclides is being evaluated. NUREG/CR-1295 The ORNL State-Level Electricity Demand Torecasting Model. ' July 1980. Oak Ridge National Lab ORNL/KUREG-63 ORES GPO. NTIS. This report presents further results of validating Version 1 of the ORNL-SLED Nodel, an investigation of structural changes in electricity demand, an update of the Version I model as Version II, the electricity cost forecasting model, and the forecasts j of electricity demand and prices by sector and by state for 1977-2000. A new set of the assumptions of exogenous variables was developed and used to forecast electricity demand and prices. The forecast rates of growth in total electricity demand vary considerably from state to state: Arizona has the highest rate (8.2%) and Massachusetts and lilinois have the lowest rate (2.9%) for the 1976-90 period. For the United States as a whole, the forecast annual growth rates of total electricity demand are 4.5%, 4.1% and 5.5% for the base, high-price, and low price cases, respectively, for the same period. NUREC/CR-1298 Growth and Histological Effects to Protothaca Staminea (Littleneck Clam) of Long-Tern Exposure to Chlorinated Sea Water. August 1980. i Battelle Pacific Northwest Lab PNL-3158 ORES GPO. NTIS. There has been considerable concern about the potential for long-term ef fects to ' marine organisms f rom chlorinated seawater. As part of a larger study to investigate the effects of materials resulting from seawater chlorination on marine organisma, ' groups of littleneck clama, Protothaca staminca, were exposed to seawater that had been chlorinated. Two experiments were conducted. In one test, groups of littleneck clams were exposed to dilutions of chlorinated seawater tbat had average chlorine produced oxidant (CPO) concentrations of 16 mg/P or less. In the second test, groups of clams were exposed to chlorinated seawater-unchlorfuated seawater mixtures that had target CPO concentrations of 0, 6, 12, 25, 50 and 100 mg/P. In the first experiment, length measurements were made on all class at approximately one-month intervals for three months. In the second test, lengt.h, weight, depth, width and edge etching were used to measure growth, and subsamples were harvested and measured at one-month intervals. In addition, clams were preserved for histological examination. ' NUREC/CR-1302 Study of Plutonium Oxide Powder Emissions from Simulated Shipping Container Leaks. August 1980. Battelle Columbus Lab; Battelle Pacific Northwest Lab PNL-3278 ORES GPG. NTIS. l ' To provide data to facilitate the predictions of Puo, caissions through leaks in Puo shipping containers under accident conditions, a series of experiments was conducted 2 ' using Puo2 Powder and an esperimental system designed to simulate a shipping container
Report No. Bibliographic Data leak. Over two hundred experiments were completed. The experimental parameters investigated were the leak size / type, internal system pressure, agitation of the 4 apparatus, leak orientation with respect to the powder location and the run time. No -l single parameter appeared to have any observable ef fect on the quantities of Pu0 2 emitted. However, there was an apparent dependency on the interaction between the orifice area and the internal pressure. The dependency took the form of a function ofA/P. Although this functional form was suggested by the data, the data were not sufficient to allow a more detailed function to be determined. The results of experiments in which the run time was variable produced the observation that changes j in the run time did not result in changes in the quantitles of Puos , emitted. This observation led to the conclusion that the majority of Pu0g observe 3 is emitted during the initial pressurization of the leak tube. KUREG/CR-1308, Vol 2 fixed Site Neutralization Model Programmer's Model. September 1980. Sandia Lab, Albuquerque SAND 79-2242 ORES GPO. NTIS. The Fixed Site Neutralization Model (FSNM) is a stochastic, time-stepped simulation of an engagement process whereby an adversary force attempts to steal or sabotage sensitive (e.g., nuclear) materials being guarded by a security force on a fixed site and a response force that is offsite. It is anticipated that the FSNM will assist regulatory bodies of the U.S. Government in evaluating fixed site physical protection systems at various installations in a variety of scenarios. In resolution, the model has representations of individual activities, plans, perceptions, psychological profiles, skills, and equipment. The forces simulated may involve as many as 50 individuals. For purposes of efficiency, most data input to the Fixed Site Neutraliza-tion Model are in binary form. Both preprocessors and the FSNM itself are written in FORTRAN. NUREC/CR-1333 Flow Topography Instrumentation and Analysis System. August 1980. Creare, Hanover, NH TN-314 ORES GPO. NTIS. An instrumentation system has been developed and used to record, display and analyze t.wo-phase flow topographies. This report describes the devices designed; the computer ' analysis techniques used to derive two-phase flow distributions, velocities, and interfaces; and the application and demonstration of these techniques during experiments in a model reactor vessel. These techniques have broad instrumentation applicability. KUREC/CR-1334 Validation of a Monte Ccrlo Code for Radiation Streaming Analyses. September 1980. Mathematical Applications Group, Elmsford, NY KR-7068 ORES GPO. NTIS. Calculations have been performed with the SAM-CE computer code to demonstrate that an-off-the-shelf Monte Carlo radiation transport code can, with affordable computer running times, adequately predict dose levels, due to streaming, in the vicinity of an operating nuclear power plant. The configuration of the Millstone-2 plant was simulated and neutron and secondary gamma ray cases were calculated on the operating floor, and elsewhere both with and without a water tank shield at the top of the cavity. Agreement between calculations and actual measurements was excellent; usually considerably better than a factor of two. NUREG/CR-1340 State-of-the-Art Study Concerning Near-Field Earthquake Ground Motion. August 1980. Systems, Science and Software, La Jolla, CA SSS-R-80-4217 ORES CPO. NTIS. This report presents a status summary of a continuing investigation into the applica-bility of theoretical earthquake source modeling to the definition of design ground motion environments for nuclear power plants located in the near-field'of potentially active faults. A wide variety of proposed near-field ground motion prediction procedures are described and evaluated, it is concluded that existing empirical procedures for
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l Report No. Bibliographic Da_ta_ I predicting near-field ground motion characteristics are not adequately constrained by the available strong motion data, leading to order of magnitude uncertainty in the prediction of some parameters. .On the other hand, the review of the proposed theoretical source descriptions has identified a number of model parameters and assumptions which are also not well constrained either by data or theory and which may affect the .j near-field ground motion estimates predicted by these models. Preliminary parametric study results are presented, for example, which demonstrate that different, commonly . employed assumptions concerning the initiation and stopping of earthquake faulting. I can have a significant effect on the computed near-field ground motion characteristics. l NUREG/CR-1347 Thermocouple Signal Sensitivity to the Sheath Thickness of Thermal-Hydraulic Test 1 i Facility Indirectly Heated Electric Fuel Pin Simulators. 1 September 1980. Oak Ridge National Lab ORNL/NUREG-62 I ORES GPO. NTIS. I Analysis of data in large loss-of-coolant accident experimental facilities often requires extensive use of signala recorded from thermocouples embedded in indirectly heated electric fuel pin simulators (EFPS). These signals, conv.erted to temperature, are used in the numerical determination of EFPS experimental conditions, including transient surface temperature, transient autFace heat flux, and transient internal radial temperature distribution. Important points that arise in using the recorded thermocouple signals as a basis for subsequent analysis include (1) the effect of the distance that the thermocouple bead is located from the EFPS surface on the ability of the thermocouple to resolve rapidly changing boundary phenomena and (2) the extent that this depth influences the time response associated with the thermocouple. Several numerically solved ETPS transients (where boundary conditions were specified, and surface temperature, surface heat flux, and internal radial temperature histories were subsequently calculated) are presented in an effort to establish these two relationships and to form a foundation for recommending designs which will minimize the adverse ef fects of these relationships by specifying optimal thermocouple radial positions in future EFPSs. NUREG/CR-1349 Reactor Safety Research Programs. Quarterly Report, October-December 1979. August 1980. Battelle Pacific Northwest Lab TNL-3040-4 ORES GPO. NTIS. This document summarizes the work performed by Pacific Northwest Laboratory from Oct ober 1 through December 31,19M, for the Division of Reactor Safety Research l within the Nuclear Regulatory Comm.ssion. Evaluation of nondestructive examination ' (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the fessib1llty of detemining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systema, examir.ing NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program ' at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho l National Engineering Laboratory. These programs will provide data for comput.cr I modeling of reactor system and fuel performance during various abnormal operating conditions. NUREG/CR-1355 CONAN: An UfFBR Containment Response Computer Code. July 1980. l Brookhaven National Lab ' BNL-NUREG-51151 ONRR GPL . NTIS. A descriptica is given for the mathematical models used in the CONAN containment analysir co e. The code was designed to study the particular phenomena which are j important nd limiting in the response of an DiFBR containment system to a hypothetical i core melt .hrough accident. Results obtained using the CONAN code are compared to ! the omet solution for an ediabatic system and to results obtained from the CACECO
Report No. Bibliographic Data containment code. The conclusion drawn from results obtained with CONAN is that the processes of evaporation and condensation, which are not treated mechanistically in the CACECO code, do have a significant effect on the transient and tend to strongly limit the severity of the accident in terms of containment pressurization. KUREG/Ck-1356 High Cycle Fatigue Behavior of Incoloy 800H in a Simulated High-Temperature Gas-Cooled Reactor Helium Environment. July 1980. Brookhaven National Lab BNL-NUREG-51156 ORES GPO. NTIS. The current study was an attempt to evaluate the high cycle fatigue strength of Incoloy 800H in a High-Temperature Gas-Cooled Reactor helium environment containing significant quantities of moisture. As-heat-treated and thermally-aged materials were tested to determine the ef fects of long-term corrosion in the helium test gas. Results from in-helium tests were compared to those from a standard air environment. KUREG/CR-1357 Heat Removal Characteristics of Volume Heated Boiling Pools with Inclined Boundaries. July 1980. Brookhaven National Lab BKL-KUREG-51157 ORES GPO. NTIS. The heat transfer characteristics of volume-heated boiling pools are of importance in j the safety analysis of hypothetical core disruptive accidents (HCDA) in liquid metal fast breeder reactors (UTFBRs). In general, these pools would be composed of molten fuel and steel and would generate heat as a result of fission product decay. The fluid dynamic characteristics, as well as the containability of such boiling systems, would depend intimately on the heat loads applied to the surrounding boundaries. In s addition, the thermodynamic and hydrodynamic states of the boiling mixture might i determine the initial or boundary conditions for separate but related phenomena, such as nuclear recriticality, structural integrity, flow and freezing of multiphase fluids, ete-. This report presents new experimental data for local boundary heat transfer coefficients and average void fraction in volume-boiling pools and comparea [ these results to previous experimental data and to existing empirical models. ; NUREG/CR-1358 Vegetational Cover in Monitoring and Stabilization of Shallow Land Burial Sites. August 1980. P University of California UCLA 12-1235 ORES GPO. KilS. The. year FY79 was a transition year between start up of work at the low-level ! waste burial site at Maxey Flats, Kentucky and completion of previous work involving r laboratory studies with radionuclides. All of our studies are designed to solve problems or verify situations that exist in the field. The thrust at Maxey Flats by this group involves soil moisture and radionuclide movement at that burial site in a , i humid region. Vegetation cover is being manipulated, rooting depth is being studied, ' water penetration and flow are being measured, radionuclide uptake by plants and concentration in components of soil moisture are being measured. Goals are to determine how water is penetrating trenches and how to minimize such penetration, Laboratory
- studies involve fission and transuranic radionuclides with a future focus placed primarily upon field problems related to low-level waste burial problems and soils. !
Some past studies being completed involved transuranic elements and a cross section + of USA soils. Different sized containers have been involved in the studies so that ; results can be extrapolated to field conditions. Analytical work is almost completed ' and the data are being synthesized. Some preliminary organization of the data is included in this annual report. Concentration ratios, plant part discrimination ,a ratios and radionuclide ratios are included in the initial evaluation. The laboratory phase of this study is to be ccmpleted in the next fiscal year. j
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Report No. Bibliographic Data
- < NUREG/CR-1364 Uncertainty Analysis for a PWR Loss-of-Coolant Ac'cident
- 11. Alternative Core Damage Es tima t o rs .
August 1980. Sandia National Lab, Albuquerque SAND 80-0629 ORES GPO. NTIS. In reinvestigating the data base used to analyse the statistical behavior of peak ' clad temperature (PCT), we have defined several new core damage indices or estimators, and have introduced and evaluated the concepts of local and global core damage. Methods of enhancing the importance of thermal-hydraulic variables compared to fuel variables have also been success fully tested. This trade-off is accomplished by an ~ l Increase in the number of calculations required to create a data base. NUREG/CE-1375 User's Manual for USINT: A Program for Calculating Heat and Mass Transfer in Cancrete Subjected to High Heat Fluxes, t August 1980. Sandia National Lab, Albuquerque ' SAKD79-1694 ORES GPO. .NTIS. This report is a user's manual for the intelligent' application of PROGRAM USINT which ; is for calculation of heat and mass transfer in concrete subjected to high heating rates. The describing differential equations for energy, mass transfer of water and CO2 are provided along with appropriate boundary and initial conditions. . The concrete is considered to contain two basic regions: wet and dry. In the wet region . steam, r CO 2 and liquid water may co-exist bat in the dry region there is no liquid water There is also the possibility of a third region in which there is only liquid water and no gas ~u. Decomposition of the concrete is treated by 6tilizing first order chemical kinetic equations; three reactions are assumed that. treat evaporable mater, chemically bound water and C0 2 . A modified Clausius-Clapeyron equation is used as the equation of state in the wet region. , NUREG/CR-1377 Risk Methodology for Geologic Disposal of Radioactive Waste: Transport flodel Sensitivity Analysis. August 1980. Sandia National Lab, Albuquerque SAND 80-0644 " ORES GPO. NTIS. ' In this report, a sensitivity analysis methodology is demonstrated for geosphere transport. The sensitivity analysis uses two transport simulators. One simulator is ; the general, multi-dimensional numerical model SWIFT. The other is the simplified it network flow model NWFT, which contains a one-dimensional radionuclide transport f simulator. Statistical techniques used in the sensitivity analysis' include Latin , hypercube sampling and stepwise regression on ranks. The demonstration problems are , based on a reference site geology and hydrology which, although hypothetical, con-tains properties of real sites.' Three different waste release scenarios are examined. . NUREG/CR-1380. Vol 1, ES Assessment of Current Onsite Inspection Techniques for Light Water Reactor Fuel j Systems - Volume 1 - Executive Summary. July 1980, , 1 Battelle Pacific Northwest Lab PNL-3325 OKRR GPO. NTIS. i t Onsite inspection techniques currently used on fuel systems at domestic commercial light-water reactors were assessed. Those techniques are visual, gumma scanning, ; sipping, mensural, eddy current, and ultrasonic inspections. The assessment involved < a literature survey, meetings with all five reactor fuel suppliers, and visits to three reactor sites.
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Report No. Bibliographic Data NUREG/CR-1394 Diffusion Near Buildings as Determined from Atmospheric Tracer Experiments. ' September 1980. > NOAA, Air Resources Lab NOAA Tech Memo ERL ARL-84 , ORES GPO. NTIS. Data from the innermost arca and roof top samplers of the Rancho Seco and EOCR field studies were used to examine diffusion close to a building. The minimum length plume 5 paths were determined from each release location to each sampler position at these two test sites. Measured concentrations, normalized by source strength (C/Q), were plotted versus plume path length and an envelope containing 95% of the measured i values of C/Q was determined. The curves f rom the two sites were similar in shape and implied three zones of diffusion. Comparisons were also made with current NRC ' methods for predicting maximum expected concentrations close to a building. The NRC model overestimated concentrations in all but one case. The model was generally within an order of magnitude at EOCR, and within two orders of magnitude at Rancho Seco. NUKEG/CR-1396 Prompt Burst Energetics Experiments: Fresh Uranium Carbide / Sodium Series. July 1980. Sandia National Lab, Albuquerque , DAND80-0820 I ORES GPO. NTIS. The current program in Prompt Burst Energetics (PBE) at Sandia Laboratories involves an in-pile experimental and complementary analytical investigation of the energetics of fuel-clad-coolant systems subjected to energy deposition conditions associated with super-prompt critical excursions. In particular, the emphasis to date has been on autoclave tests of single intact fuel pins in the presence of stagnant sodium irradiated in the experiment cavity of the Annular Core Pulse Reactor (ACPR) and on , the supportive analysis of those tests. This report describes the experiment per-formed in the Annalar Core Pulse Reactor. NUREG/CR-1400 Quarterly Technical Progress Report on Water Reactor Safety Programs Sponsored by the Nuclear Regulatory Commission's Division of Reactor Saf ety Research - April-June 1980.
- August 1980.
EG&G Idaho EGG-2048 ORES GPO. NTIS. , EG&G Idaho, Inc., performs water reactor safety research at the Idaho National Engineering Laboratory under the sponsorship of the U.S. Nuclear Regulatory Commission's (NRC) Division of Reactor Safety Research. The current water reactor research activities are acecmplished in the Semiscale Program, the Loss-of-Fluid Test (LOFT) Experimental Program, the Thermal Fuels Behavior Pry;-*m, the Code Development and Analysis Program, the Code Assessment and Applications Program, and the 2D/3D Program. I NUREG/CR-1402 Advanced Reactor Safety Research Division Quarterly Progress Report. October 1-December 31, 1979. September 1980. Brookhaven National Lab BNL-NUREG-51177 j ORES GPO. NTIS. i This quarterly report describes current activities and technical progress during , Octcber-December 1979 in t.* Advanced Reactor Safety Research Program. The projects y repossed this quarter are in N Safety Evaluation, SSC code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.
l l Report No, Bibliographic Data NUREC/CR-1403 Water Reactor Safety Research Division Quarterly Progress Report. October-December 1979. July 1980. Brookhaven National Lab BNL-KUREG-51178 ORES GPO. KTIS. The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USKRC Division of Reactor Safety Research. The projects reported. each quarter are the following: LWR Thermal liydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing. NUREC/CR-1404 Transient Analysis of Coolant Flow and lieat Transfer in LMFBR Piping Systems. September 1980. Brookhaven National Lab BNL-NUDEG-51179 ORES GPO. NTIS. A one-dimensional model for transient analysis of coolant flow and heat transfer in LMFBR piping systems is presented, in which energy equations are formulated for the coolant and pipe wall using the nodal heat balance approach. An implicit integration scheme is applied to the coolant equation with explicit wall heat flux, allowing the solution to march in the flow direction. Uncertaihties in Nusselt number correlations are shown to have the greatest impact on overall heat transfer coefficients at low flow conditions. For coolant dynamics, each pipe run between components in treated as a lumped control volume. For the transient cases studied: (a) the predicted response using the coolant-wall model is in excellent agreement with a more detailed model that includes insulation heat losses, while the transport delay and coolant .. mixing models appear to be inadequate. (b) The degree of axial nodalization required for a converged solution is indeed bounded. (c) Timcatep control is found to be most efficiently achieved using the characteristic time approach. (d) The predicted flow decay is found to be only marginally affected by the Reynolds number dependence of friction factor in the pipings and IHX. NUREG/CR-1405 The NACOM Code for Analytti of Postulated Sodium Spray Fires in LMFBRs. September 1980. Brookhaven National Lab BNL-NUREG-51180 ORES GPO. N11S. An analysis of potential sodium spills and fires in liquid metal fast breeder reactors has been made to assess the maximum equipment cell loading conditions. A computer code called NACOM (sodium combustion) has been developed at Breokhaven National Laboratory (BNL) to analyze sodium spray fires. This report contains a detailed description of physical models used in this code as well as programming aspects. The single droplet combustion model and the model describing the droplet's motion are verified. Comparisons between NACOM predictions and SPRAY-3A predictions of the Atomics International (AI) LTV Jet Tests are made. RUREC/CR-1408 Instrumented Impact Properties of Zircaloy-Oxygen and Zircaloy-Ilydrogen Alloys. July 1980. Argonne National Lab ANL-80-14 ORES GPO. NTIS. This report summarizes results of instrumented impact tests on subsize Charpy speci-mens of Zircaloy-oxygen and Zircaloy-hydrogen alloys. These results can be used with other property data to determine the response of reactor fuel assemblies to mechanical loads during the latter stages of hypothetical LOCA transients and seismic conditions. Mathen.atical analyses are used to evaluate the dynamic response of reactor fuel assemblies subjected to LOCA and seismic loads. Failure boundaries have been estab-lished for thermal shock as well as for impact and compressive loads at low temperatures.
Report No. Bibliographic Data NUREG/CR-1410 Report ,of the Zion / Indian Point Study: Volume 1. Augast 1980. Sandia Natienal Lab, Albuquerque SAND 80-0617/1 ORES GPO. NTIS. This report contains detailed results of a study for the identification of reactor i core-melt accident mitigation measures at the Zion and Indian Point plants. Miti-gation strategies have been identified that show promise of providing large reduction , in consequences for specific accident sequences. However, without an overall risk analysis, it is not clear to what extent a given mitigation scheme reduces overall risk. The study evaluated filtered-vented containment systems, steam explosions, hydrogen burning, hydrogen control measures, melt / concrete and melt /MgG interactions, and meltdown phenomenology. NUREG/CR-1411, Vol 1 Report of the Zion / Indian Point Study. September 1980. Los Alamos Scientific Lab LA-8306-MS ORES GPO. NTIS. This report presents analyses performed at the Los Alamos Scientific Laboratory in support of the U.S. Nuclear Regulatory Commission Zion / Indian Point Project. The analyses were performed in cooperation with Sandia Laboratories-Albuquerque. Three, areas of work are discussed: (1) Steam explosion energetics; (2) Fluid slug-vessel head impact; and (3) Structural dynamics. NUREG/CR-1411, Vol.2 Report of the Zion / Indian Point Study. July 1980. Los Alamos Scientific Lab LA-8306-MS ORES GPO. NTIS. This repurc presente analyses performed at the Los Alamos Scientific Laboratory (LASL) in support of the U.S. Nuclear Regulatory Commission (NRC) Zion / Indian Point Project (Z/IP). The analyses were performed in cooperation with Sandia Laboratories-Albuquerque. Three areens of work are discussed: (1) Steam expression energetics; (2) Fluid slug-vessel impact; and (3) Structural dynamics. Sandia analyses and the overal study summary and conclusions are provided in two companion Sandia reports, NUREG/CR-1409 (SAND 80-0617) and NUREC/CR-1410 (SAND 80-0617/1). NUREG/CR-1420 Health Status and Body Radioactivity of Former Thorium Workers. September 1980. Argonne National Lab ANL-80-37 ORES GPO. NTIS. l This is a progress report of a study of the health effects of industrial exposure to thorium. The objectives of the study are: (1) to assess possible health effects of employment in the thorium milling industry by comparison of mortality and morbidity characteristics of former thorium workers with those of suitable general populations; (2) to examine dt. ease outcomes by estimated exposure levels of thorium and thoron , I daughter products for possibic radiation-related effects; and (3) to determine the ! I body distribution of inhaled ttorium (and daughters) and rare earths in humans by radioactivity measurements in v2vo and by analysis of autopsy samples. The principal endpoints for investigation are respiratory diseases and cancers of lung, liver, bone, and bone marrow. I NUREG/CR-1428 Solubility Classification of Airborne Uranium Products from LWR-Fuel Plants. August 1980. i Battelle Pacific Northwest Lab PNL-3411 ' OSD GPO. NTIS. l l Airborne dust samples were obtained from various locations within plants manufacturing fuel elemt ats for light-water reactors, and the dissolution rates of uranium f rom ! tbase samples into simulated lung fluid at 37'C were measured. These measurements were used to classify the solubilities of the samples in terms of the lung clearance ( model proposed by the International Commission on Radiological Protection. Similar I
Report No. Biblingraphic Data i l evaluations were performed for samples of pure uranium compounds expected as components i in plant dust. The variation in solubility classifications of dust encountered along I the fuel production lines is described and correlated with the process. chemistry and
'the solubility classifications of the pure uranium compounds.
NUREG/CR-1438 Steam Line Dynamics. September 1980. Brookhaven National Lab BNL-NUREG-51186 ORES- GPO. NTIS. A computer program has been developed for the prediction of transients in the main I steam supply lines of power plants. The program is specifically suited for acoustic transients, induced by sudden value actions. The program has been. assessed as a stand-alone program by comparing computer results with both analytical and experimental results. The agreement is good. The program is designed to serve as part of a , systems analysis code and has been implemented in the it. version of RAMONA-Ill, a ! boiling water reactor systems code. NUREG/CH-1445 Preparation of Working Reference Materials: Calcined Waste Recovery _ Products Containing Uranium or Plutonium. September 1980. I
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Los Alamos Scientific Lab LA *346 OSD GPO. NTIS. Procedures are presented for preparing calcined waste recovery products that have ; assigned values of uranium and plutonims contents and isotopic distributions. These working reference materials are used to calibrate and maintain measurement control surveillance of chemical methods for analyzing plant process materials. Statistical treatments are discussed that preside a measure of the reliability of working reference , materials in applications to nu icar material accountability and safeguards. NUREG/CR-1448 Physical Protection of Nuclear Facilities. Quarterly Progress Report, January-March 1980. August 1980. Sandia National Lab, Albuquerque SAND 80-1006 ORES GPO. NTIS. This report presents evaluation rethodology efforts concerned primarily with the Safeguards Automated Facility Evaluation (SAFE) methodology and the Brief Adver-sary Threat Loss Estimator (BATLE) model. In support of a study on design concepts for sabotage protection, the SAFE methodology was applied to the Standardized Nuclear Unit Power Plant System (SNUPPS) facility. Alternative SNUPPS facility designs were also analyzed using SAFE. The activities this quarter were principally related to facility characterization or evaluation methodology tasks. Facility characterization activities concentrated on the vital area analysea of operating reactor facilities. In addition, existing computer codes for rank ordering of vital areas were extended by the addition of subroutines to allow calculations of approximations to the importance measures. Several new approximation methods applicable to the vital area ranking techniques were also examined. NUREG/CR-1450 Multirod Burst Test Program Progress Report for July-December 1979. September 1980. Oak Ridge National Lab ORNL/NUREG/TM-392 ' ORES GPO. NTIS. A series of scoping tests designed to explore the effect of shroud heating on zircaloy cladding deformation was conducted in the single-rod test f acility, which was recently modified to permit independent beating of the shroud under specified conditions. To i facilitate comparison of the test results, the series included tests under specified + conditions used previously. Significantly greater deformation was observed in heated shroud tests than would be expected from unheated shroud tests. Fabrication of fuel pin simulators for the B-5 (8 x 8) bundle test continued with N 90% of the required number being completed. Five fuel pin simulators, identical to the simulators used ! in the Japanese Atomic Energy Research Institute multirod bundle burst tests, were i delivered by the Japanese manufacturer. The surface temperature distribution of the - f 1
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t Report No. Bibliographic Data simulators was characterized for several heating rates by infrared scanning and was compared to similar characterizations of Oak Ridge National Laboratory simulators. Plans are under way for conducting burst tests on the Japanese simulators in the single-rod test facility. NUREG/CR-1457 Regional Relationships Among Earthquake Magnitude Scales. 3 September 1980. Lawrence Livermore National Lab 5 UCRL-52745 h ORES CPO. NTIS. i The seismic body-wave magnitudeb a of an earthquake is strongly affected by regional variations in the Q structure, composition, and pr.ysical state within the earth. Therefore, because of differences in attenuation of P-waves between the western and eastern United States, a problem arises when comparingb a 's f r the two regions. A regional magnitude bias exists which, depending on where the earthquake occurs and where the(P-waves are recorded, can lead to magnitude errors as large as one-third i unit. There is also a significant difference between a and M gvalues for earthquakes-b in the western United States. An empirical link between the g of an castern U.S. earthquake and the M g of an equivalent western earthquake is gIven by Mg = 0.57 + s 0.92 (g ) East. This result is important when comparing ground motion between the two regTons and for choosing a set of real western U.S. earthquake records to repre-sent eastern earthquakes. NUREG/CR-1466 Predicting Life Expectancy and Simulating Age of Complex Equipment Using Accelerated- r Aging Techniques. July 1980. Sandia Lab, Albuquerque SAND 79-1561 ORES GPO. NTIS. This document outlines some of the types of experiments which can be used to improve reliability, simulate age, and predict life expectancy of complex equipment. Brief discussion in given of failure ac3e tests and compatibility tests, which often are useful qualitative aging information. A detailed discussion is presented on accelerated aging methods, emphasizing an approach based on kinetic rate expression. It is concluded that, when properly conceived and carried out, accelerated aging i studies of materials and simple componen*s offer the best opportunity for making , quantitative age simulations and lifetime predictions of equipment. NUREG/CR-1471 An Assessment of LWR Fuel-Failurt Propagatra Potential: Literature Survey. August 1980. Los Alamos Scientific Lab LA-UR-80-45-15 ' ONRR GPO. RTIS. In this report a preliminary investigation is made to identify the consequences of small-scale, off-normal, localized events that induce single or few-element fuel rod failure on the potential to cause adverse effects or failure propagation to adjacent fuel rods, for commercial light-water-reactor (LWR) design. Since previous assessments of the potential for fule-f ailure propagation have centered primarily on liquid metal fast breeder reactor (LHFBR) designs, a review of such work is presented for the t purpose of identifying generic concerns that might also be applicable to LWR systems. LWR fuel-f ailure behavior under both normal and mild of f-normal conditions are also reviewed and phenomena identified that could be of concern in fuel-failure propagation. , The preliminary results of this review indicate that either rapid fission gas release ; (from the plenum region of a failed fuel element) or release of molten or finely-fragmented solid fuel (from the core region of a failed fuel element) to the coolant , stream, inducing subchannel coolant expulsion and attendant vapor or gas blanketing l and undercooling of adjacent intact fuel rods, appear to be the.most probable initi-ating mechanisms by which fuel-failure propagation could occur. g h [ t t T
I l l Report No. Bibliographic Data j NUREG/CR-1472 Structural Integrity of Water Reactor Pressure Boundary Components. ' Quarterly Progress Report, January-March 1980. August 1980. Naval Research Lab NRL Memo Rpt 4254 ORES GPO. NTIS. This report describes progress in a continuing program to characterize material ' properties performance with respect to structural integrity of light-water-reactor pressure boundary components. Progress under f racture mechanics describes J-R curve trends from a low shelf A302-B steel and includes a comparison of R curves by the multispecimen and single specimen compliance procedures. Fatigue crack growth rates are being determined for a variety of pressure vessel and piping steels in simulated . nuclear coolant environments. Static load cracking in this environment has been observed in bolt-loaded specimens taken from weld heat-affected zones. Work in radiation sensitivity and postirradiation properties recovery has defined tensile property changes under cyclic annealing and reirradiation treatment. Recent progress is described in radiation studies involving reactor vessel steels in a coordinated
'1AEA program. Also reported are notch ductility tests of reference steels of the NRC light water reactor, pressure vessel' irradiation dosimetry program.
NUREG/CR-1474 An Algorithm to Estimate Field Concentrations Under Nonsteady Meteorological Conditions from Wind Tunnel Experiments. September 1980. Colorado State University ORES GPO. NTIS. Highest concentrations at ground level are often produced from surface sources with stable atmospheric conditions and near calm winds. This report describes a weighted data methodology developed to predict surface concentrations from staticnary wind-tunnel measurements and actual meteorological' wind fields. Field measurements made downwind of the Rancho Seco Nuclear Power Station in 1975 have been compared against a set of wind-tunnel measurements around a 1:500 scale model of the same facilities. The weighted data algorithm was realistic in predicting centerline concentration values as well as the horizontal spread of the plume. KUREG/CR-1475 Wind-Tunnel Measurements of Dispersion and Turbulence in the Wakes of Nuclear Reactor Plants. September 1980. Colorado State University 3' ORES GPO. NTIS. Between 1975 and 1979 via contracts between the !?C and Colorado State University, a sequence of laboratory experiments have been performed to evaluate the influence of nuclear reactor. building complexes on dispersion of efflu(nts released into their wakes. This study involved research directed toward quantifying the wake-dispersion interaction as well as a validation exercise to compare laboratory and field measure-ments about specific altes. This report presents the program objectives and summarizes
- t.he results of two model/ field building dispersion experiments; a comparison of perturbation model predictions to model measurements of velocity deficit, turbulence excess, and temperature or concentration perturbations; an examination of the ef ficacy of a new algorithm used to predict full-scale concentrations downwind of buildings in nonstationary wind fields from wind tunnel measurements; preliminary measurements of close-in dispersion near obstacles; and behavior of a stratification wind tunnel designed to study coastal atmospheric boundary layer behavior,.
NUREC/CR-1477 Heavy-Section Steel Techno^. s Program Quarterly Progress Report for January-March 1980. July 1980. Oak Ridge National Lab ; ORNL/NUREC/TM-393 1 ORES GPO. NTIS. ! Comprises studies related to all areas of the technology of materials f abricated < into thick-section primary-coolant containment systems of light-water-cooled nuclear I power reactors. The investigation focuses on the behavior and structural integrity l of steel pressure vessels containing cracklike flaws. Current work is organized into five tasks: (1) program administration and procurement, (2) fracture mechanics , analyses and investigations, (3) investigations of irradiated materials, (4) thermal ) shock investigations, and (5) pressure vessel investigations. 1 l l 1
l Report No. Bibliographic Data NUREG/CR-1480 Summary of Thermal Hydraulic Calculations for a Pressurized Water Reactor. July 1980. Los Alamos Scientific Lab LA-8361-MS OMRR GPO. NTIS. The results of two transients involving the loss of a steam generator in a single-pass, ' steam generator, pressurized-water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steau bubble in the core while the pressurizer is solid. Calculations show that continued injection of high-pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island. NUREG/CR-1481 Financing Strategies for Nuclear Power Plant Decommissioning. July 1980. Temple, Barker & Sloan, Lexington, KA OSD GPO. NTIS. The report analyzes several alternatives for financing the decommissioning of nuclear ; power plants from the point of view of assurance, cost, equity, and other criteria. Sensitivity analyses are performed on several important variables, and possible impacts on representative companies' rates are discussed and illustrated, t KUREG/CR-1482 Nuclear Power Plant Simulators: Their Use in Operator Training and Requalification. August 1980. Ook Ridge National Lab ORNL/NUREG/TM-395 ORES GPO. NTIS. The report presents the resulta of a study performed for the Nuclear Regulatory Commission to evaluate the capabilities and use of nuclear power plants simulators either built or being built by the U.S. nuclear power industry; to determine the adequacy of existing standards for simulator design and for the training of power l plant operators on simulators; and to assess the issues about simulator training programs raised by the March 28, 1979 acciden* at Three Mile Island Unit 2. It is the conclusion of this study that both ANSI /ANS standards should be expanded, strengthened, and endorsed by the NRC; that simulator training should be required; and that a well-defined regulatory structure for simulators and simulator training , programs should be developed. The most obvious deficiency in the present standards is the absence of a methodology and data base to comprehensively and objectively assess the relative importance of specific malfunctions and thereby define the most important exercises to be included in a training program. This deficiency was considered to be sufficiently serious that the present study was expanded to include the development of a preliminary methodology and to il19 strate its application. NUREG/CR-1484 Dynamic Analysis to Establish Normal Shock and Vibration of Radioactive Material Shipping Packages - Quarterly Progress Report October 1-December 31, 1979. August 1980. Hanford Engineering Develop. Lab KLDL-TME-80-24 ORES GPO. NTIS. A computer program MARCS (Modal Analysis of a Rail Car-Cask System) was written to perform a modal analysis on the systems represented by the CARDT and CARDS (Cask-Rail Car Dynamic Simulator) models. Parameters generated by MARCS will be used to generate frequency response spectra. A preliminary evaluation of the performance of CARDS was made by comparing calculated reruits with response variables measured during Test 3 of the series of testa conducted at the Savannah River Laboratories, Aiken, SC. l 1 NUREG/CR-1485, Vol 1, No. 1 Safeguards Material Control and Accounting Program: Quarterly Report, October-December 1979. , September 1980. Lawrence Livermore National Lab UCRL-52715-80-1 ORES GPO. NTIS. Activity for the quarter October-December 1979 in the Material Control Saf eguards Evaluatioe Program, conducted for the U.S. Nuclear Reguatory Commission (KRC) at Lawrence Lavermore National Laboratory, is summarized. Progress was made in developing j l
Report No. Bibliographic Data the automated safeguards assessment tool called the Structured Assessment Approach (SAA) Program, giving particular attention to enhanced collusion analysis. Work has-continued on the development of the Aggregated Systems Model (ASM) in support of the ; NRC development of MC&A upgrade segulations, and we include value-impact analyses of alternative safeguard rules, a first-cut safeguards cost model, and a study of the impacts of MC&A regulations on licensees. The report concludes with a description of more work in support of the MC&A upgrade rule development, which is our ev.sluation ' and critique of the current NRC material accounting regulations, an attempt to identify inherent vulnerabilities. NUREC/CR-1486 Seasonal Vibration of 10-Square Mile Probable Maximum Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. $3). 6 July 1980. National Weather Service, NOAA OSD GPO. NTIS. Estimates of the probable maximum precipitation (i.e., the theoretically greatest depth of precipitation for a given duration that is physically possible over a particular drainage basin at a certain time of year) are given in this study for durations from 6 to 72 hours for each month of the year for 10 m12 areas. The results > are in a generalized form; that is, on maps allowing use for planning and design of any present or proposed structure for the United States east of the 105th meridian. While smaller sized areas have greater values, especially for the warm season, they are not defined in the study. For the winter season, values for smaller areas are not appreciably different from the 10 mi2 estimates in this study. The probable maximum precipitation estimates show a smooth variation with duration, season, and location. NJkEG/CR-1487 Vital Area Analysis Using SETS. August 1980. Sandia National Lab, Albuquerque SAND 80-1095 ORES CPO. NTIS. This report describes the use of the Set Equation Transformation System (SETS) for vital area analysis. Several concepts are introduced which enable the analyst to ' construct more efficient SETS user programs to perform vital area analysis. The advantages of performing the transformation of variables without first determining the minimal cut sets of the fault tree are discussed. A " bottom-up" approach to solving a fault tree is presented. The techniques described for vital area analysis are also suitable and efficient fan many kinds of common cause analyses. NUREC/CR-1488 An Ultrasonic Thermometry System for Measaring Very High Temperatures in Reactor Safety Experiments. July 1980. Sandia Lab, Albuquerque SAND 79-0621 OkES CPO. NTIS. Ultrasonic thermometry has many potential applications in reactor safety experiments, where extremely high temperatures and lack of visual access may preclude the use of conventional Jiagnostics. This report details ultrasonic thermometry requirements for one auch experiment, the molten fuel pool experiment. Sensors, transducers, and signal processirs electronies are described in detail. Axial beat transfer in the sensors is modelled and found acceptably small. Measurement errors, calculations of their ef fect, and ways to minimize them are given. A rotating sensor concept is discussed which holds promise of alleviating sticking problems at high temperature. Applicatious of ultrasonic thermometry to three in-core experiments are described. KUREC/CR-1489 Best Estimate Method vs Evaluatioa Method: A Comparison of Two Techniques in Evaluating Seismic Analysis and Desie: July 1980, Lawrence Livermore National Lab UCRL-$2746 ORES GPO. NTIS. The concept of how two techniques, Best Estimate Method and Evaluation Method, may be applied to the traditional seismic analysis and design of a nuclear power plant is introduced. Only the four links of the seismic analysis and design methodology chain
Report No.. Bibliographic Data (SMC)--seismic input, soil-structure interaction, major structural response, and ' subsystem response--are considered. Thc objective is to evaluate the compounding of conservatisms i: the seismic analysis and design of nuclear power plants, to provide guidance for judgmencs in the SMC, and to concentrate the evaluation on that part of , the seismic analysis and design which is familiar to the engineering community. An example applies the ef fects of three-dimensional excitations on a model of a nuclear power plant structure. The example demonstrates how conservatisms accrue by coupling two links in the FMC and comparing those results to the effects of one link alone. l The utility of employing the Best Estimate Method vs. the Evaluation Method is also demonstrated. NUREG/CR-1492 Qualification Testing Evaluation Program Light-Water Reactor Safety Research Quarterly Report, July-September 1979. September 1980. Sandia National Lab, Albuquerque SAND 80-Ill7 ORES GPO. NTIS. The July-September 1979 quarter can be characterized as a period of formal reporting and continuing effort in the Qualification Testing Evaluation (QTE) Program. Under Task 1, the principal effort was devoted to the preparation for continuation of verification tests of the qualification of Browns Ferry Unit 3 connector assemblies. Radiation and thermal-aging experiments were completed last quarter; checkout of the superheat test capability was continued in preparation for the accident-simulation test phase. Under Task 3, effort was directed towards the effects of the sorption characteristics of a material with respect to quantitatively accelerating the degrada-tion. When experiments are properly run and analyzed, Arrhenius behavior is often found. The balance of the quarterly effort centered on continuation of ongoing projects. NUREG/CR-1501 Monoclinal Structure and Shallow Faulting of the Reelfoot Scarp as Estimated from , Drill Holes with Variable Spacings. July 1980.
- Vanderbilt University ORES GPO. NTIS.
The Reelfoot scarp is an east-facing slope on the Mississippi River alluvial plain. It descends eastward about 20 feet across a distance of about 600 feet. The scarp is mainly a monocline. However, a small fault occurs at the foot of the slope, along t which there was about 3 feet of graben collapse which subsequently (maybe soon after 1812) filled in which soil washed down tha adjacent monoclinal slope. The monoclinal structure was clearly shown by six drill holes spaced about 100 feet apart. A fault was indicated by these same holes, but a fault was not demonstrated by drilling until holes were drilled closer together than 10 feet. NUREG/CR 1504 A User's Guide to EPIC, a Computer Program to Calculate the Motion of Fuel and
- Coolant Subsequent to Pin Failure in a LMFBR.
July 1980. Argonne National Lab ANL-80-41 ORES GPO. NTIS. The computer code EPIC models fuel and coolant motion which results from internal i fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of a sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions whereby t fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel f rom the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is'used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a FORTRAN-IV program requiring 400K bytes of storage oJ the IBM 370/195 computer.
Report Fo. . Bibliographic Data NUREG/CR-1505 Advanced Reactor Safety Research Division Quarterly Progress Report, January 1-March 31, 1980. September 1980. Brookhaven National Lab BNL-NUREG-51217 ORES ' GPO. NTIS. This quarterly report describes current activities and technical progress during January-March 1980 in the Advanced Reactor Safety Research Program. The projects y reported are HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation. NUREC/CR-1509 Light Water Reactor Safety Research Program Quarterly Report, January-March 1980. September 1980. Sandia National Lab, Albuquerque SAND 80-1304/1 of 4 ORES GPO. NTIS. The Molten Fuel Concrete Interactions (MFCI) study is comprised of experimental and analytical investigations of the chemical and physical phenomena associated with interactions between molten core materials and concrete. Such interactions are possible during hypothetical' fuel-melt accidents in light water reactors (LWRs) when molten fuel and steel from the reactor core penetrate the pressure vessel and cascade 1 onto the concrete substructure. The purpose of the MFCI study is to develop an ' understanding of these interactions suitable for risk assessment. Emphasis is placed on identifying and investigating the dominant interaction phenomena occurring between prototypic materials in order to evaluate: (1) The generation rate and nature of evolved noncondensable gases; (2) The effects of gas generation on fission products release; and (3) The mechanism, rate, and directional nature of concrete erosion by the melt. NUREC/CR-1513 Evaluation of Isotope Migration-Land Burial Water Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report-, Janua ry-March 1980. July 1980. Brookhaven National Lab BNL-NUREG-51219 ORES GPO. NTIS. This report presents results for tritium analyses for soil cores takes, at West- Valley, NY, and Barnwell, SC. Tritium movement at West Valley appears to be ditiusion controlled. . The Barnwell core data suggests that coring has intersected a water flow path below the trench. An apparatus has been designed for flow through column K determinations l d and is described. Gel filtration experiments using spiked trench water from West l Valley have been continued using a longer column than used in previous work. Increased ) resolution of DOC components has been observed. { I KUREG/CR-1514 Properties of Radioactive Wastes and Waste Containers. Quarterly Progress Report, January-March 1980. July 1980. Brookhaven National Lab BNL-KUREG-51220 ORES GPO. NTIS. This report presents first quarter 1980 progress in research on properties of radio-active wastes and waste containers at Brookhaven National Laboratory. Solidification creeriments were performed with organic ion-exchange resins using Portland Type II' ser%t to intritigate waste to binder ration which results in monolithic waste forms, periments le conducted to establish appropriate waste / binder ratios within which
' simulated boric acid reactor waste may be incorporated into Portland Type KKK cement, to produce acceptable waste forms. A "two-part" urea-formaldehyde process was used to solidify four simulated LWR waste streams, viz., lon-exchange bead resins, diatomaceous earth, sodium sulfate, and boric acid wastes.
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l l 1 Report No. Bibliographic Data NUREG/CR-1516 Nuclear Reactor Safety Quarterly Progress Report, October 1-December 31, 1979. August.1980. Los Alamos Scientific Lab LA-8299-PR ; ORES GPO. NTIS. l l This guarterly repcrt summarizes technical progress from a continuing nuclear reactor safety research program. The reporting period is from October 1 to December 31, 1979. This research effort concentrates on providing an accurate and detailed understanding of the response of nuclear reactor systems to a broad range of postulated accident conditions. The report is mainly organized according to reactor type. Major sections deal with light-water reactors (LWRs), liquid metal fast breeder reactors (LMFBRs), high-temperature gas-cooled reactors (RTGRs), and gas-cooled fast reactors (GCFRs). NUREG/CR-1518 Assessment of Core Penetration of a PWR Reactor Vessel and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents. August 1980. Sandia National Lab, Albuquerque SAND 80-0701 ORES GPO. NTIS. A brief analysis of events surrounding a PWR reactor vessel f ailure following a core meltdown was performed. The purpose of the analysis was to assess the impact of such events on a containment building filtered vent. Specific accidents considered included , a loss of AC power and auxiliary feedwater (TMLB'), a small-break LOCA with ECCS failure (S2D) and large-break LOCA with failure of the containment heat removal system (ABG). The MARCH computer code analysis of these accidents (Indian Point 3 and Zion reactors) was used as a basis of comparison. The major findings are ' (1) location and size of vessel rupture in the TMLB' accident could significantly affect the pressure history in containment and the subsequent loading on the filtered vent; (2) high internal reactor pressure (f rom rapid debris slumping into the lower head water) could cause steam generator tube failure and thus failure of secondary containment; (3) significant contaissent building pressure rise could occur from molten debris dropping into the reactor cavity if there is adequate water in the cavity for complete quenching; and (4) the coolability of total-core in-vessel or ex-vessel particle beds by natural circulation (assuming an adequate coolant supply) can neither be assured nor excluded at this time. Suggested research to resolve-uncertainties in the above items is discussed. NUREG/CR-1520 Experiment Data Report for LOFT Anticipated Transient Experiment L6-5. July 1980. EG&G Idaho EGG-2045 ORES GPO. NTIS. This report presents experimental data from the first anticipated transient experiment (Experiment L6-5) cenducted in the Loss-of-Fluid Test (LOFT) facility. The data are uninterpreted but readily usable for the nuclear community in advance of detailed analysis and interpretation. Experiment L6-5 was a loss of secondary feedwater anticipated transient performed on May 29, 1980, and was part of the LOFT Eraerimental Program conducted by EG&G Idaho, Inc. , for the U.S. Nuclear Regulatory Commission. This experiment is part of the LOFT Non-LOCE Test Series L6 which was designed to provide data for investigating the thermal-hydraulic response of the LOFT reactor system for transient initiation to plant restabilization af ter reactor scram. NUREJ JR-1521 High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety > Research Quarterly Progress Report, January 1-March 31,1980. ; September 1980. Oak Ridge National Lab ORNL/NUREG/TM-397 ORES GPO. NTIS. 4 Work continued on development of the ORTAP, ORECA, and BLAST codes; verification studies were continued on the ORECA, CORTAP, and BLAST codes. An improved steam turbine plant model (ORTURB) for use in ORTAP was developed and checked. Predictions from BLAST, CORTAP, and ORECA were compared with various transient test data from the Fort St. Vrain reactor.
I i l Report No. Bibliographic Data KUREC/CR 1526, Vol 1 Physics of Reactor Safety. Quarterly Repurt for January-March 1980. July 1980. Argonne National Lab ANL-80-54, Vol 1 ORES GPO. NTIS. . This quarterly progress report summarizes work done during the months of January-March 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research. - The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-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-1527 Enhancement of the Nuclear Materials Management and Safeguards System. August 1980. Boeing Computer Services, Vienna, VA ONMSS GPO. NTIS. The Nuclear Regulatory Commission (NRC) awarded a competitive contract to Boeing Computer Services Company to implement specific recommendations developed under the Enhancement of the Nuclear Materials Management and Safeguards System (EKRAS) Contact, NRC-02-78-08;, and to perform analysis in other specified areas of safeguards concern. The results of the activities of this contract were the production of program specifi-cations for enhancements to the Nuclear Materials Manage 6aent and Refcguards System (KMMSS) in the areas of inventory difference and authori?ci possession limit data; the production of acceptance test procedures for testin6 nne implemented capability; an analysis of the KMMSS Safeguards Monitor (SM-1) report and recommendations for its improvement; the production of program specifications for enhancements to the NKMSS SM-1 report based on selected recommendations; an analysis of NMMSS shipper-receiver difference data processing; and the production of a safeguarda user's manual containing KMMSS reports related to analysis performed under this contract. This activity is pcrt of KRC's ef fort to continuously enhance their nuclear materials accounting system. NUREG/CR-1528 Safeguards User's Manual for Nuclear Materials Management and Safeguards System. August 1980. Boeing Computer Services, Vienna, VA ONMSS GPO. NTIS. The computerized system used by KRC to receive, store, analyze, and report information on the nuclear material possessed by each licensee is called the Nuclear Materiai Management and Safeguards System (NMMSS). It is located at the DOE computer facility, Oak Ridge, Tennessee. In September 1978, a contract (NRC-02-78-083) was awarded to determine and document inventory difference (ID) inconsistencies between NMMSS and the Safeguards Status Reporting System (SSRS) maintained by the Office of Inspection and Enforcement (I&E). This work resulted in an extensive analysis of NMMSS. This report documents the results of this study in the form of a User's Manual that would be useful to Safeguards Analysis. Twenty NMMSS reports were identified that contain licensee data related to inventory differences, authorized possession limits, and shipper-receiver differences. Each NMMSS report is described with a general descrip-tion, processing to produce the report, and a sample report page with the data elements identified. Also included is a summa ry of the data input to KMMSS, its processing, and a description of the files and records in the NMMSS data base. NUREG/CR-1529 Two-Phase flow Measurements with Advanced Instrumented Spool Pieces. September 1980. Oak Ridge National Lab ORNL/NUREG-72 ORES GPO. NTIS. A series of two phase, air-water and steam-water tests performed wita instrumented piping spool pieces is described. The behavior of the three-beam densitometer, turbine meter, and drag flowmeter is discussed in terms of two-phase models. Results ; from application of some two-phase mass flow models to the recorded spool piece data are shown. Results of the study are used to make recommendatitas regarding spool l piece design, instrument selection, and data reduction methods to obtain more accurate l measurements of two-phase flow parameter ^. l l
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Report No. Bibliogrcphic Date NUREG/CR-1536 PRESBC: Pressure Boundary Conditions for the K-FIX Code. August 1960. Los Alamos Scientific Lab LA-NUREG-6623, Supp 3 ORES GPO. NTIS. Recommended pressure boundary condition modifications are described for the computer code K-FIX, which has been published in the report LA-NUREC-6623 and released to the National Energy Sof tware Center in April 1977. NUREG/CR-1537 Gap Conductance Test Series fuel Characterization Data Report. September 1980. EG&G Idaho ' EGG-2046 ORES GPO. NTIS. The physical, chemical, mechanical, and metallurgical properties of the UO2 fuel tased in the Power Burst Facility Cap Conductance Test Series are presented. These data were obtained from nondestructive and destructive examinations of representative fuel pellets performed by EC&G Idaho, Inc., and by Battelle Pacific Northwest Laboratories. These data characterize the initial fuel condition, and are necessary to understand and evaluate fuel rod behavior during irradiation testing in the Gap Conductance Test > Series. NUkEC/CR-1538 PBF/ LOFT Lead Rod Test Series Test Results Report. July 1980. EG&G Idaho EGG-2047 ORES GPO. NTIS. Results of the Power Burst Facility / Loss-of-Fluid Test (PBF/ LOFT) Lead Rod sequential blowdown test series conducted in the PBF are presented. The tests were performed to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa) light-water-reactor design fuel rods when subjected to a series of large, double-ended cold leg break luss-of-coolant accident (LOCA) tests, and to determine whether subjecting these deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature and pressure measurements with out-of-pile experiment data, by comparison of steady-state fuel centerline temperature response, and by posttest visual examinations and cladding j diametral measurements. NUREC/CR-1539 A Methodology and a Preliminary Data Base for Examining the Health Risks of Electricity Generation from Uranium and Coal fuels. September 1980. Oak Ridge National Lab; Science Applications, La Jolla, CA " ORNL/Sub-7615/1 ORES GPO. NTIS. An analytical model was developed to assess and examine the health effects associated with the production of electricity f rom uranium coal fuels. The model is based on a systematic methodology that is both simple and easy to check, and provides details about the various components of health risk. A preliminary set of data that is needed to calculate the health risks was gathered, normalized to the model facilities, and presented in a concise manner. Additional data will become available as a result of other evaluations of both fuel cycles, and they should be included in the data base. 1 NUREG/CR-1547 triticality Experiments with Subcritical Clusters of 2.35 Wt% and 4.31 Wt% U-235 Enriched UO Rods in Water at a Water-to-Fuel Volume Ratio of 1.6. July 1980. 2 , Battelle Pacific Northwest Lab PNL-3314 * ( ORES GPO. NTIS. i The results from the fourth in a series of criticality experiments are presented in i this paper. This fourth set of experiments involves clusters of either 2.35 wt% or 4.31 wt1 U-235 enriched UO 2 fuel r ds immersed in water as in previous experiments but a neutron moderation approximating that found typically in boiling-water-reactor and i I ! l
. ~ - - I l Report No. Bibliographic Data pressurized-water-reactor type fuel assemblies. (Previous experiments were performed' near optimum neutron moderation.) The critical separation between multiple suberitical fuel assemblies, both with and without fixed neutron poisons, was determined. NUREG/CR-1548 The NDT-COMP 9 Microcomputer. September 1980. Oak Hidge National Lab ORNL/NUREC/TM 390 ORES GPO. HTIS. An 8080-based microcomputer system, the NDT-COMP 9, has been designed for instrumen - tation control and data analysis in eddy-current tests. The NDT-COMP 9 represents a significantly more powerful computer system than the NDT-COMP 8 microcomputer f rom which it was developed. The NDT-COMP 9 system is contained on a 240- by 120-mm (9.5-by 4.8 in.) circuit board and will it in a four-wide Nuclear lastrumentation Module (NIM) bin with 26-pin edge connectors. In addition to the 8080-compatible central l processing unit (CPU), an arithmetic processing unit (APU) is available to provide up ) to 32-bit fixed- or floating point, basic or transcendental math functions. The 16K of read only memory (ROM) and random access memory (RAM), one serial input-output (I/0) port (RS-232-C at a maximum speed of 9600 baud), and 72 parallel 1/0 ports are available. The baud rate is under software control. A system monitor and math package are available for use with the microcomputer. 7 NUREG/CH-1549 Estimates of Uranium Content and Radon Flux for Uranium Hine Dumps Based on Borchold Radioactivity Logs. ! July 1980. Kilborn/NUS, Denver, CO ONMSS GPO. NTIS. In exploratory drilling to locate uranium deposits, borehole logs of gamma radiation from aturally radioactive elements are utilized to indicate the presence of uranium and the concentrations in which it is found at various depths. This report describes a method of otilizing borehole log data to estimate uranium concentrations in the rock surroundtL. 6. 'verlying uranium deposits and to predict radon releases from waste rock brought to tne ground surf ace in mining operations. The method can be used to , predict radon releases before L2ning operations.are started so that potential environ-mental impacts can be evaluated. The estimates of uranium concentration are generally within 20 percent of true values af ter correcting for concentrations of naturally radioactive thorium and potassium in the normal range; variations in emanation coef-ficients and diffesion rates for radon can introduce errors in radon flux estimates, but the estimates should be the correct order of magnitude in most mining regions.
- NURE0/CR-1552 Development and Verification of Fire Tests for Cable Systems and System Components.
September 1980. Underwriters Lab UL-USNC 75 Q ORES GPO. NTIS. Experiments were conducted to study the effects of a forced ventilation on the results of the IEEE 383 flame test for tray cables. Three sets of experiments were conducted on three types of cables. The first set was a control in which cable samples were tested in a f ree-convection environment within an 8 x 8 x 8 f t (2.44 x 2.44 x 2.44 m) enclosure. In the second set, the cable samples were tested within an enclosure with a 1500 CFM (708 1/s) forced ventilation. In the third set, the cable samples were tested within an enclosure with the ventilation rate of either 1200 CFM (566 1/s) or 1800 CFM (849 1/s). The results showed that the rate of flame propagation and the i maximum cable damage were not affected when the enclosure and forced ventilation were used, it is recommended that the test method be revised to specify the use of an enclosure and forced ventilation. NUREC/CR-1554 On the Motions of Particles in Turbulent Flows. July 1980. State University of NY, Stony Brook ORES C PO . NTIS. The present paper describes theoretical and experimental studies of the behavior of turbulant particle dispersion flows. In particular, dispersions of low particle concentration are considered in a turbulent pipe flow and the particle deposition at the wall is studied. The theoretical treatment is basd on an analysis of the particles'
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i keport No. Biblio2ragh: c Data with safe-shutdown-earthquake (SSE) mechaaical loadings and employing conventional stress analysis and fracture-mechanics techniques. KUREC/CR-1569 Seismicity and Tectonic Relationships for Upper Great Lakes Precambrian Shield Province. July 1980. Minnesota Geological Survey ORES GPO. NTIS. This is a final report comprising a three-year study of the seismicity of Minnesota including the procurement and installation of a six-station seismograph system. This system was deployed in a microcarthquake monitoring array, An earth model was developed based on signals from mine blasts and regular earthquake bulletins were published. Descriptions of the model, methodologies, and three significant earth-quakes are given. NUREG/CR-1575 Hydrogen-Mixing in a Closed Containment Compartment Based on a One-Dimensional Model with Convective Effects. September 1980. Los Alamos Scientific Lab LA-8429-MS ONRR GPO. NTIS. A transient, 1-dimensional finite difference model was developed for determining the hydrogen concentration variation with position in a closed containment compartment caused by radiolysis following a LOCA. The model includes mixing due to molecular and eddy diffusion and natural convection. For representative compartments, the maximum hydrogen concentration difference between the bottom and top of the compartment never exceeds 0.25 volume percent when all 3 mixing mechanisms are considered for a range of parameters. NUREG/CR-1579 Considerations on Nuclear Data Link Implementation in Relation to the Technical Support Center, Emergency Operations Cester and Safety Parameter Display System. September 1980. Sandia National Lab, Albuquerque SAND 80-1618 ORES GPO. NTIS. Upon the occurrence of a significant event at a nuclear plant, both the Nuclear Regulatory Commission and the licensee must carry out certain roles to mitigate possible consequences. To accomplish their respective roles effectively, both require timely data from the plant instrumentation systems. The relationship of the NRC-oriented Nuclear Data Link to the individual licensee-oriented Technical Support Centers, Emergency Operations Facilities, and Safety Paramter Display Systems has been examined with regard to implementation of data acquisition, communication, and display require-ments. The possible use of a common data acquisition processor for all four systems is discussed, along with technical considerations important in the implementation of such an approach. A common data acquisition system is not recommended but could be l successfully implemented if tight control is exercised. Some of the anticipated difficulties in developing standardized data displays are outlined. Duplication at the NRC Operations Center of those displays available at the reactor sites will be extremely dif ficult unless industry-wide standa::dization of displays is implemented. KUREG/CR-1581, Vol 1 Evaluation of Mathematical Models for Characterizing Plume Behavior from Cooling Towers: Vol. 1--Dispersion from Single 6t; Multiple Source Natural Draft Cooling Towers. September 1980. Argonce National Lab ORES GPO. NTlS. Fif teen mathematical models for visible plume prediction from natural draft cooling towers are evaluated theoretically and tested with 39 sets of single-tower visible plume field data from three sites. Seven of these models with the capability of treating plumes from multiple towers are further tested with 26 sets of multiple tower data from two sites. The visible plume outlines provided by these data give , information on the trajectory of the plume as well as dilution. The model/ data comparisons prepared in this study revealed systematic behaviors in the predictions of most models which were able to be traced back to model assumptions. A wide range of predictions was found to occur among the models. No one model performed consis-tently well for all data sets. Theoretical analysis of the model formulations revealed
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Report No. Bibliogrephic Data frequency response to the surrouding eddy motion. A scheme is developed to determine the location of the joining boundary between two distinguishable regions of the particle transport; a turbulent dif fusion-controlled region and a region in which the particle motion is controlled by the mean velocity field of the fluid as in laminar flows. The paper also gives first results of experimental investigations carried out ) in an upward turbulent air flow in a vertical pipe-test-rig. Measurements were. j carried out-for various sizes and it is shown that'the resulta qualitatively support l the theoretical findings. KUREG/CR-1557' Steam-Water Mixing and System Hydrodynamics Program - Task 4 - Quarterly Progress Report, April 1-June 30, 1979. July 1980. Battelle Columbus Lab BM1-2038 ORES GPO. NTIS. r During this quarter, analysis included additional development of the 1* scaling param-eter for tubes, review of results from air-water tests in distorted geometries, and correlation of results from countercurrent flow condensation tests in a rectangular test section. Experimental work this quarter included completion of condensation tests and heat partitioning studies in the rectangular test section, and air-water and steam- 5 water tests in the 2/15-scale model with shortened and extended annulus lengths. The-instrumented break leg spool piece was received from IKEL and installed in the 2/15-scale facility. NUREG/CR-1563 Eddy-Current inspection for Steam Generator Tubing Program Annual Progress Report for Period Ending December 31, 1979. t August 1980 Oak Ridga National Lab . OPAL /KUREG/TM-398 ORES GPO. NTIS. . This report presents the ORNL program to improve the eddy-current inspection capabilities for in-service inspection of steam ~ generators presently undertaken for the Nuclear Regulatory Commission. Eddy-current methods provide the best in-service inspection of steam generator tubing, but present techniques can produce ambiguity because of the many independent variables that af fect the signals. The current development program has used mathematical models and developed or modified computer programs to design optimum probes, instrumentation, and techniques for multifrequency, multiproperty examinations. Interactive calculations and experimental measurements have been made with the use of modular eddy-current instrumentation and a minicomputer. More testing is needed for all the different combinations of cases and different types of defects, r
.KUREG/CR-1564 Comparison of CONTEMPT-LT Containment ' Code Calculations with Marviken, LorT, and Battelle-Frankfurt Blowdown Tests.
July 1980. Los Alamos Scientific Lab LA-8423-MS OKRR GPO. NTIS. This study compared the CONTEMPT-LT/026 containment analysis code calculations with large-scale test results. LASL reviewed 7 large-scale experimental test programs and selected 5 of the 16 Marviken tests for pressure-suppression containment analysis comparisons and i LOFT test as a secondary investigation. In addition, 1 Marviken test was used to investigate the effects of 18 code parameter variations. A single , Battelle-Frankfurt test was used for a dry containment comparison. NUREG/CR-1568 LWR Fuel Rod Post-Subcooled Blowdown Scoping Analysis. August 1980. ; EG&G Idaho _ ONRR GPO. KTIS. s Thermal transients which occur during the post-subcooled blowdown regime of a postulated loss of-coolant accident (LOCA) can caust significant changes in light water reactor (LWR) f uel rod material properties and geomet ry. ~ The ef fects of these st ructural changes must be assessed to insure that a coolable geometry of the fuel system is maintained. An overall assessment of the fuel rod structural integrity has been made . by considering the degraded structural properties of the fuel rod in conjunction
7 . Report No Bibliographic Data that models which correctly predict the plume trajectory due to the entrainment mechanism alone will overpredict dilution. The more successful models employ an additional mechanism to provide additional bending without additional mixing. The correctness of any of the additional bending mechanisms remains to be determined. The model/ data discrepancies are partly due to model errors and partly due to data measurement errors. The accuracy of the data makes it unlikely for a model to predict - better than a factor of l\-2 in most, perhaps 90%, of all data cases. NUREG/CR-1581, Vol 2 Evaluation of Mathematical Models for Characterizing Plume Behavior from Cooling Towers: Vol.2--Salt Drift Deposition from Natural Draft Cooling Towers. September 1980. Argonne National Lab ORES GPO. NTIS. Twelve mathematical models for salt drift deposition from natural draft cooling towers are evaluated in terms of performance with prototype data and validity of theoretical assumptions. Model predictions are compared with field data acquired at the Chalk Point Power Plant during 1975, 1976, and 1977. Large, often several orders of magnitude differences among model predictions existed for runs with the 1975-1977 data. Since the field data are limited, extrapolation of the model's performance to significantly different types of environmental conditions and farther from the tower than 1 km should be done with caution. Each model reviewed was shown to have limita-tions in two categories. There were asumpticas that were either not correct physically or not state-of-the-art. Secondly, there were assumptions whose correctness is presently unknown. Sensitivity and comparative studies were also conducted and are described. NUREG/CR-1581, Vol 3 Evaluation of Mathematical Models for Characterizing Plume Behavior from Cooling Towers: Vol. 3--Plume Rise f rom Mechanical braf t Cooling Towers. September 1980. University of Illinois ORES GPO. NTIS. Various methods commonly used to predict the length and height of the visible piame produced by an array of mechanical-draft cooling towers are evaluated by comparing predictions with observational data from the Benning Road Power Station and from a small array of towers at the Purdue University Power Plant. Four different approaches-- empirical, integral, cloud-physics, and finite-difference--are examined. Statistical estimates of predictive capability are given. Problems inherent in the application of these approaches are discussed. Observations concerning areas of weakness and thus areas of potential improvement are made. NUREG/CR-1582, Vol 2 Seismic Hazard Analysis. A Methodology for the Eastern United States. August 1980. Lawrence Livermore Lab ONRR GPO. NTIS. This report presents a probabilistic approach for estimating the seismic hazard in l the Central and Eastern United States. The probabilistic model (Uniform Hazard Methodology) systematically incorporates the subjective opinion of several experts in the evaluation of seismic hazard. Subjective input, assumptions, and associated hazard are kept separate for each expert so as to allow review and preserve diversity of opinion The report is organized into five sections: Introduction, Methodology Comparison, Subjective Input, Uniform Hazard Methodology (UHM), and Uniform Hazard Spectrum. Section 2, Methodology Comparisn, briefly describes the present approach and compares it with other available procedures. The remainder of the report focuses on the UHM. Specifically, Section 3 describes the clicitation of subjective input; Section 4 gives details of various mathematical models (earthquake source geometry, magnitude distribution, attenuation relationship) and how these models are combined to calculate seismic hazard. The last section, Uniform Hazard Spectrum, highlights the main features of typical results. Specific results and sensitivity analyses are not presented in this report. I I
Report No. Bibliographic Data KUREG/CR-1582, Vol 3 Seismic Hazard Analysis. Solicitation of Expert Opinion. s August 1980. Lawrence Livermore Lab ONRR GPO. NTIS. This report presents a detailed tabulation of ten experts' answers to a questionnaire on seismicity and ground motion characteristics of the Central and Eastern United States. The goal in eliciting such information was to obtain a subjective representation, of parameters that affect seismic hazard in order to supplement the very limited historical data that are available in these regions. Not only was the "most probable value" sought in each case, but also, whenever possibles .the entire probability distribution to be used in a probabilistic hazard analysis. The questionnaire was divided into five sections: Source Zone Configuration, Maximum Earthquakes, Earthquake Occurrence, Ground Motion Models, and Overall Level of Confidence. The last section was designed to develop a synthesis of opinion, if need be. The questionnaire.was designed to contain redundancy to provide cross-checking and establish consistency in the results. NUPfn/CR-1583 Radon Release and Dispersion from an Open Pit Uranium Mine. July 1980. Argonne National Lab AKL/ES-97 ORES GPO. NTIS. Radon-222 flux from representative sections of the United Nuclear St. Anthony open-pit mine complex was' measured. In January 1979, permission was obtained from United Nuclear Corporation to install equipment for these measurements at the St. Anthony Mine in the Grants, New Mexico, mineral belt. This report describes measurements and details of the studies performed from March through September 1979. Results of the following tasks undertaken in this study are presented: (a) Measurement of' radon flux-from the ground, (b) Measurement of working level and airborne radon concentrations, (c) Measurement of. meteorological parameters, (d) Development of a' theoretical model to describe the release of radon from open pit mines. NUREG/CR-1584 Psychological Stress for Alternatives of Decontamination of TMI-2 Reactor Building Atmosphere. August 1980. Human Design Group, Olney, MD ONRR CPO. NTIS. The purpose of this report is to consider the nature and level of psychological stress that may be associated with each of several alternatives for decontamination. The report briefly reviews some of the literature on stress, response to major disaster or life stressors, provides opinion on each decontamination alternative, and considers possible mitigative actions to reduce psychological stress. The report concludes that any procedure that is adapted for the decontamination of.the reactor building atmosphere will result in some psychological stress. The stress, however, should abate as contamination is reduced and uncertainty is diminished. The advantages of the purge alternative are the rapid completion of the decontamination and the conse-quent elimination of future uncontrolled release. Severe stress effects are less likely if the duration of stressor exposure is reduced, if the feeling of public control is increased, and if the degree of perceived safety is increased. The long delays, continued uncertainty, and possibility of uncontrolled release that charac-terize the other alternatives may offset the perception that they are safer. In addition, chronic stress could be a consequence of long delays and continued uncertainty. NUREG/CR-1585 Modeling Tornado Dynamics. September 1980. Aeronautical Research Assoc, Princeton ARAP Rpt 421 ORES GPO. NTIS. This report details the results of a research program aimed at providing NRC with a definitive theoretical model of the tornado, low-level flowfield. It includes a critical review of existing theoretical models and detailed ae ounting of an i axisymmetric numerical model based on turbulent transport theory. Model verification tests show reasonable comparsions with laboratory and limited field results. An extensive model sensitivity analysis shows the flowfield is most dependent on the ambient vertical vorticity and horizontal convergence occurring in the parent thunderstorm. It is also quite sensitive to surface roughness. Dominant features of
Report No Bibliographic Data the model results are the low altitude at which the maximum windspeed occurs and.the
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large magnitude of velocity fluctuations which occur close to the surface. When available dual doppler observations are used to impose boundary conditions on a model domain with a 1-Km radius and 1-Km height, estimates of the maximum windspeed suggest ' that speeds in excess of 125 m/see should be' exceedingly rare. An analytic fit to the complex model results is provided so that the model wind distribution can be used in further engineering design studies. Five previously published papers reporting partial results of the research program are included as appendices. NUREG/CR-1586 Respirator Studies for the Nuclear Regulatory Commission, Evaluation and Performance , of Escape-Type Self-Contained Breatbing Apparatus, October 1, 1978-September 30, 1979. S September 1980. Los Alamos Scientific Lab LA-8432-MS ORES GPO. NTIS. The performance of escape-type breathing apparatus was evaluated for weight, comfort, , ease of use, and protection factor (calculated from facepiece leakage). All of the devices tested provided a self-contained air supply of 5-to 15 min duration.. Five of l them have the provision to connect.an air line but allow the use of the self-contained supply for safe egress. The air supply was stored in cylinders, tubing, or disposable containers. Respiratory inlet coverings were half masks, full facepieces, hoods, and mouthpieces. An estimate is given for the case of quick donning, Recommendations for conditions for use of the equipment are given. NUREG/CR 1590 Evaluation Methodology for Fixed-Site Physical Protection Systems. September 1980. Sandia National Lab, Albuquerque SAND 80-0505 ONMSS GPO. NTIS. A system performance evaluation methodology has been developed to aid the Nuclear. Regulatory Commission (NRC) in the implementation of new regulations designed to upgrade the physical protection of nuclear fuel cycle f acilities. The evaluation methodology, called Safeguards Upgrade Rule Evaluation (SURE), provides a means of explicitly incorporating measures for highly important and of ten dif ficult to quantify performance factors, e.g., installation, maintenance,. training and proficiency levels, compatibility of components in subsystems, etc. This is achieved by aggregating resposes to component and system questionnaires through several successive levels of a functional hierarchy developed for each primary performance capability specified in the regulations, 10 CFR 73.45. An overall measure of performance for each capability is the result of this aggregation process. This paper provides a description of SURE. NUREG/CR-1592 Compressible Analysis of Inlet Plenum Pressure Rise due to Sodium Boiling in Fuel r Subassemblies During Pump Coast-Down of an LMFBR. August 1980. Argonne National Lab ANL-80-48 ORES GPO. NTIS. The effect of sodium compressibility and steel elasticity on the rise in inlet plenum pressure occurring during boiling in a loss-of-flow accident in an LMFBR has been investigated using the PTA-2 code. These effects do not seem large enough to require conbideration in accident analysis. The pressure rise is less for pool than for loop designs. KUREG/CR-1593 Performance Testing of Personnel Dosimetry Services: Alternatives and Recommendations ' for a Personnel Dosimetry Testing Program. August 1980. , University of Michigan -t OSD GPO. NTIS. The Nuclear Regulatory Commission (KRC) is considering an amendment to 10 CFR P&rt 20 = that would require their licensees to use only processors of personnel dosimetry , devices (e.g. , film badges and thermoluminescent dosimeters) that have been' certified. m- [ Although this action would have a direct effect only on those processors that service NRC licensees, it would most likely lead indirectly to a nationally recognized certi-fication program for all dosimetry processors. The objectives of this Report are to , l
i Reyort No. Bibliographic Data consider a variety of alternatives that would influence a certification program, to consider the advantages and disadvantages, values and impacts, of each alternative, and to make a recommendation for each alternative. Among the considerations discussed are: (1) is a certification program necessary? . (2) What standard should be used for a testing prgoram? (3) What type of organization should test dosimetry processors? (4) How often should a processor be . retested? (5) What appeals procedures should be available to a processor 7 (6) What are realistic estimates of the costs of a testing program? i NUREG/CR-1601 Critical Experiments, Measurements, and Analyses to Establish a Crack Arrest Methodology for Nuclear Pressure Vessel Steels. [ July 1980. ' Battelle Columbus I.ab BMI-2055 ORES CPO. NTIS. Analysea have been performed on ORNL Thermal-Shock Experiment TSE-5 using a modified plastic dynamic finite-difference-solution procedure. These used two different postulated dynamic-fracture-toughness relations. In both cases, the dynamic analysis predicted that crack arrest would occur well beyond the point suggested by a quasistatic analysis. Crack-initiation studies on the steel used in TSE-5 revealed a large degree of scatter in K A statistical method was used to estimate lower-bound I.Ehree toughness based on the parameter Weibull distribution. Preliminary fractographic examination indicates that the fracture origin can be located and that this technique-may be used to elucidate the source of the scatter. , NUREG/CR-1602 Strength and Stif fness of Tensioned Heinforced Concrete Panels Subjected to Nembrane Shear, Two-Way Reinforcing. , July 1980. Cornell University ORES GPO. NTIS. This report presents results of an experimental program to investigate seismic shear transfer in a cracked reinforced concrete containment vessel without diagonal rein-forcement. The test specimen was designed and constructed to represent the stress conditions in the wall of a pressurized containment subjected to tangential shear stresses such as those induced by an earthquake. Four monotonic and twelve reversing shear load tests were done on 4 ft square by 6 in thick flat specimens reinforced with steel bars in two orthogonal directions. Test parameters included the level of biaxial tension applied to the bars (from 0 to 0.9f and the loading history. Results are given for strength, stiffness, developmEn)t of cracking, and degradation effects produced by cyclic shear. Engineering models for predicting strength and 6 stif fness are given, along with preliminary design implications. A comprehensive review of pertinent literature is also included. NUREC/CR-1606 An Evaluation of Condensation-Induced Water Hammer in Preheat Steam Generators. September 1980. Brookhaven National Lab ,t BNL-KUREG-51248 ONRR GPO. NTIS. ! At the request of the Division of Systems Safety of USNRC, BNL evaluated the potential of condensation-induced water hammer in preheat-type steam generators. Westinghouse 1/8-scale water hammer tests and data analysis were reviewed. BNL has concluded that water bammers occurred in the feedwater line during many of the 1/8-scale tests and i were probably caused by steam bubble entrapment and collapse in the partially filled , feedwater line. The Westinghouse scaling laws were also independently reviewed. The ' present state-of-the-art on the condensation heat transfer and the mechanism of vapor cavity formation precludes us f rom. deriving any credible scaling criteria. However, , under certain operating conditions the condensation-induced void collapse could be an oscillatory process. This may partially explain the' apparent randomness of the water hammer phenomenon seen in most experimental studies. The full-scale preheat-type steam generators of both the Westinghouse and the Combustion Engineering design have been reviewed from the viewpoint of condensation-induced water hammer, it is recom-mended that each plant should be reviewed separately to identify the worst situation (s) for the condensation-inducti water hammer, and the appropriate verification test (s) should be performed in planta. In addition, basic research should be sponsored in order to enhance our understanding in this area.
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Report No. Bibliographic Data i NUREG/CR-1607 Drop-Size Estimates for a Loss-of-Coolant Accident. August 1980. Los Alamos Scientific Lab LA-8449-MS ONRR CPO. NTIS. Drop sizes ranging between 16 and 76 pm are estimated for loss-of-coolant-accident (LOCA) conditions. A break-size diameter of 0.3 m (1 ft.) and liquid temperature of i 590 K (600'F) are assumed. The best estimate is that the drop size will be less than , 16 pm due to the combined effects of heterogeneous and homogeneous nucleation and of , aerodynamic atomization. The calculations are based on an extrapolation of available i lowtemperature fragmentation data to typical LOCA conditions. The extrapolation is suported by a semtempirical fragmentation model that is consistent with low-temperature , measurements reported in the literature. If drops are formed by some unknown process ! upstream of the break, the largest drop that can escape fragmentation when passing ! through the break opening is estimated to be 46 pm. NUREG/CR-1608 Scenario Development and Evaluation Related to the Risk Assessment of High Level Radioactive Waste Repositories. August 1980. GGS, Urbana, IL CGS /NR85F060 ORES GPO. NTIS. This study elamines the consequences of disruptive features within a geologic repository system intended for the isolation of high-level nuclear. waste. The presence of fault zones having high hydraulic conductivity does not change significantly the confinement ' capabilities of a reference repository system (RRS) in bedded salt. Fault zones of low hydraulic conductivity can reduce or enhance the confinement capabilities.
- depending upon conductivity magnitude and location of the depository in the system.
The analysis. indicates that with accuraie characterization and proper understanding, ' an RRS with a fault zone might be utilized to provide confinement characteristics
- superior to those of the unfaulted system. The presence of fractures distributed uniformly throughout a typical granite mass could result in unacceptable confinement capability for a repository system in granite. Preliminary resulta emphasize the- 6 need to evaluate critically the assumptions involved in modeling groundwater flow in ,
fractured media and to accurately characterize the repository system. i NUREC/CR-1609 A Deterministic-Probabilistic Model for Contaminant Transport. August 1980. ! CCS, Urbana, IL
- CGS /NR85UO60 '
ORES, GPO. NTIS. This manual describes a deterministic-probabilistic contaminant transport (DPCT) > computer model designed to simulate mass transfer by ground-water movement in a } vertical section of the earth's crust. The model can account for convection, disper- , sion, radioactive decay, and cation exchange for a single component. A velocity is a calculated from the convective transport of the ground water for each reference particle in the modeled region; dispersion is accounted for in the particle motion by adding a random component to the deterministic motion. The model is sufficiently general to enable the user to specify virtually any type of water table or geologic configuration, and a variety of boundary conditions. A major emphasis in the model -; development has been placed on making the model simple to use, and information pro- ! vided in the User Manual will permit changes to the computer code to be made relatively ; easily for those that might be required for specific applications, i i NUREG/CR-1610, Vol 1, No. 1 Inspection Methods for Physical Protection Project: Quarterly Report, March-May 1980. September 1980. Lawrence Livermore National Lab UCID-18123-80-1 ' ORES GPO. NTIS. h This is the fif th quarterly report to the U.S. Nuclear Regulatory Commission (NRC) on the progress at Lawrence Livermore National Laboratory (LLNL) in the Inspection Methods for Physical Protection (IMPP) project. Besides reporting on trips for field tests and data acquisition, the feasibility studies for field evaluation of procedures, and the progress of the E-field intrusion detector training film, the report details the production status of the 23 procedures in the draft module 81100 replacement 5 series already delivered to NRC and the status of 28 procedures now being written for .
h Report No. Bibliographic Data transportation of irradiated fuel and for possession and use of fesmula quantities of strategic special nuclear materials (SSKM). , t KUREC/CR-1620 Survey of Current State Radiological Emergency Response Capxbilities for Transportation ! Related Incidents. ' September 1980. Indiana University OSD GPO. NTIS. This volume is the final report of a project to survey current state radiological emergency response capabilities for transportation-related incidents. The survey was performed to provide the KRC with information useful in the development of guidelines for state organizations and planning for emergency response. The report includes the results of a mail and telephone survey of state emergency response of ficials; informa-tion gleamed from radiological emergency response plans and related official document; and some general conclusions and recommendations drawn in part from interviews conducted and site visits'to selected states. NUREG/CR-1621 A Characterization of Faults in the Appalachian Foldbelt. September 1980. Florida State University OSD GPO. NTIS. The characterization is a synthesis of available data on geologic faults in the Appalachian foldbelt regarding their description, generic implications, rate of movement, and potential as geologic-seismic hazards. It is intended to assist applicants and reviewers in evaluating faults at sites for nuclear facilities. Appalachian faults were found to fall into 13 groups which can be defined on either their temporal, generic, or descriptive properties. They are as follows: Group 1 Faults with demonstrable Cenozoic movement; Group 2, Wildflysch type thrust sheets; Group 3, Bedding plane thrusts - decollements; Group 4, Pre- to synmetamorphic thrusts in medium- to high grade terranes; Group 5, Post metamorphic thrusts in medium- to , high-grade terranes;; Group 6 Thrusts rooted in low crystalline basement; Group 7, ' High angle reverse faults; Group 8, Strike slip faults; Group 9, Normal (block) faults; Group 10, Compound f aults; Group 11, Structural lineaments; Group 12, Faults associated with local centers; and Group 13, Faults related to geomorphic phenomena. Unhealed faults (groups 1, 6, 8, 9, and 12) must be considered candidates for reac-tivation. Healed brittle or ductile faults (groups 4, 5, and 10) are not places of mechanical discontinuity and are unlikely candidates for reactivation. The remaining , groups (2, 3, 7, 11, and 13) should be individually assessed as to their potential for reactivation. KUREC/CR-1624 Load Combination Program. Progress Report No. 5. September 1980. Lawrence Livermore National Lab
'UCID-18674 ORES GPO. NTIS.
This cocument is a progress report on the Load Combination Program (LCP) covering the geriod April 1,1980 through June 30, 1980. The report gives a general description of the program by project and tasks, together with financial summaries, technical reports generated, and meeting attendance. Two appendixes which discuss technical subjects are also included. NUREG/CR-1625 Steam-Water Mixing and System Hydrodynamics Program - Task 4 - Quarterly Progress Report, July 1-September 30, 1979. August 1980. Battelle Columbus Lab ORES GPO. NTIS. During this quarter we analyzed results from high bypass air-water tests and f rom low subcooling steam-water tests in the 2/15-scale model, continued development of a mechanistic model for ECC penetration, and analyzed results from steam-water tests in the simple tube facility. The experimental efforts during this quarter were directed to completion of the installation of the annulus void measurement system and the instrumented spool piece for the break leg, and subsequent checkout and acquisition of initial data.
Report No. Bibliographic Data NUREG/CR-1629 In-Plant Source Term Measurements at Turkey Point Station - Units 3 and 4. September 1980. EG&G Idaho; Allied Chemical Corp ORES GPO. NTIS. This report presents data obtained at Turkey Point Units #3 and #4 as a part of the inplant source term measurement program in operating light water reactors (LWRs). The primary objective of this program is to provide the Nuclear Regulatory Commission (NRC) with operational data that can be used in evaluation of plant designs for 11guld and gaseous waste treatment systems. Data presented were obtained at the Turkey Point Power Station operated by FJorida Power and Light. 'This plant is the t,hird in a planned series of six operating LWRs to be studied. Data from all plants will be combined and interpreted to provide a data base for radioisotupe inventory in plant systems, radioactive waste treatment system performance, and source terms for both liquid and gaseous systems. One of the primary objectives in performing measure-ments at Turkey Point was to study primary-to-secondary leaks if they occurred and to determine partition f actors in steam generators. The opportunity to study primary-to-secondary leaks occurred twice during the inplant measurement period. Results of these studies together with measurements performed on the liquid and gaseous systems at Turkey Point are presented. NUREG/CR-1631 Lateral Loads on Vent Pipe in Steam Chugging. August 1980. University of California, LA ORES GPO. NTIS. The quasisteady injection of steam into a pool of subcooled water was investigated. The resulting phenomenon was studied with emphasis on structural loading on the steam downcomer. From experimental data at a single steam mass flux it was found that pressure pulsations in the pool were temperature dependent. At low pool temperatures pressure pulsations were found to have high magnitudes and occur at low frequencies. At higher pool temperatures amaller, high frequency pressure pulses were observed. Three types of pressure pulsations were observed to occur within the pool. Pressure pulsations from (1) bubble growth and bubble shape changes, (2) bubble collapse, and (3) water slug movement within the downtomer were observed and recorded. Loadings on the steam downcomer were seen to be influenced only by bubble collapse and the magnitudes were independent of pool temperature. NUREG/CR-1632 Hydrodynamics of a Vapor Jet in Subcooled Liquid. September 1980. University of California, IA ORES GPO. NTIS. The report addresses the two-phase flow of a steam jet, injected vertically downward from a submerged pipe into a stagnant pool of subcooled water. Direct contact heat and mass exchange rates and hydrodynamic pressure pulsations are investigated. The emphasis of the present work pertains to dynamics of the subsonic (unchoked) regime of the exit steam injection rate. Experimental results indicate a large influence by ' the pool subcooling and jet diameter on the frequency and amplitude measurements of l steam jet pulsations. In contrast, the exit steam velocity has a minor effect on the i Jet pulsation characteristics. A numerical model for steam jet pulsat. ions is proposed, based on the experimental observation. Reasonable agreement between numerical and experimental results has been achieved. I NUREG/CR-1634 Film Entrainment and Drop Deposition for Two-Phase Flow. l August 1980. Los Alamos Scientific I.ab LA-8475-MS ONRR GPO. NTIS. A model for estimating the rete of film mass entrainment for drop-annular flow, based on film disturbance wave stability, was developed. The model was verified by applica-tion to tests involving deposition and entrainment. To account for the deposition, a l recent particle mass diffusion correlation was used. Application of the entrainment and deposition models confirmed that a flow passage length (L/D) of about 100 or more l- is required to achieve near equilibrium, for a zero initial entrainment flow. The assumption of an initially fully entrained flow remaining approximately so, as used in nuclear power plant subcompartment analysis, is shown to be appropriate, l l
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I 1 l Report No. Bibliographic Data 1 KUREC/CR-1635 Nuclear Plant Reliability Data System 1979 Annual Reports of Cumulative System and Component Reliability. September 1980. Southwest Research Institute OMPA GPO. NTIS. / This NpRDS document includes two annual reports: (1) Annual Report of Cumulative System Reliability (A02), (2) Annual Report of Cumulative Component Reliability (A03).
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Both annual reports provide generic reliability information on systems and components for the cumulative period from July 1974 through December 1979. NUREG/CR-1637 Canadian' Seismic Agreement. September 1980. Canadian Commercial Corp, Ottawa ORES GPO. NTIS. This is the second annual progress report under terms of Contract No. NRC-04-79-180. This report gives the details of the developments during the first year of this agreement and plans for the expansion of the Eastern Canadian Telemetered Netwcrk during the coming year. This includes a-detailed description of the entire seismograph system, i.e., sensors, A/D conversion, modes of telemetry, etc. Also included are the descriptions of the present and proposed station deployment. P NUREG/CR-1638 Evaluation of In-Situ Soll Damping Characteristics. September 1980. Shannon & Wilson; Agbabian Assoc. ORES GPO. NTIS. This study consists of a development of procedures for estimating strain-dependent damping ratios from in-situ impulse test data. Two types of methods, both based on. phase differences between stress and strain,.have been developed. The fi rs t , te rmed loop methods, is based on estimation of the shape of the hysteresis loop, once the secant shear modulus is obtained from the in-situ impulse test. The second, termed equilibrium methods, utilizes dynamic equilibrium equations'to estimate shear stress which, when considering its phase relationship with shear strain, can be used to estimate the damping ratio. Careful assessments of the accuracy and applicability of the methods, using finite element calculations, laborator; tests, and field tests, shows that the third degree loop method is most promising for future applications. However, only limited cyclic shear and torsional shear laboratory test data were available at this time to carry out the assessments, and more such data is needed to carry out assessments for a wide range of soil types and strain levels. NUREG/CR-1639 Site-Dependent Effects at Strong-Motion Accelerograph Stations. September 1980. Shannon & Wilson; Agbabian Assoe ORES GPO. NTIS. This study consists of an assessment of a six-step procedure that comprises site-dependent and site-independent methods for developing seismic input criteria at sites of nuclear plants and other major structures. The assessment was based on an applica-tion of the procedure at three accelerograph sites in California; Ferndale, El Centro, and Taft, where strong earthquake motions have been recorded and where subsurface soil properties have been measured by the SV-AA joint venture. At each of these sites, reference earthquake events were identified which correspond to magnitudes and i distances identical to those for which strong ground motions had been recorded. The various site-dependent and site-independent techniques were then used to predict ground motions corresponding to these events. Comparisons between the recorded and predicted earthquake motions served as a basis for assessing the adequacy of the procedure. These comparisons showed that the six-step procedure provided reasonable estimates of the recorded motions, although no single site-dependent or site-independent method comprised by the procedure was superior in all cases. NUREG/CR-1641 Statistical Analysis of Earthquake Ground Motion Parameters. September 1980. Shannon & Wilson; Agbabish Assoc ORES GPO. NTIS. Several earthquake ground response parameters that define the strength, duration, and frequency content of the motions are investigated using regression analyses techniques;
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Report No. Bibliographic Data these techniques incorporate statistical significance testing to establish the terms in the regression *quations. Th parameters investigated are the peak acceleration, velocity, and displacement; Arias intensity; spectrum intensity; bracketed duration; Trifunac-Ptsdy duration; and response spectral amplitudes. The study provides insight into how these parameters are affected by magnitude, epicentral < distance, local site ( conditions, direction of motion (i.e., whether horizontal or vertical), and earthquake event type. The rosults are presented in a form so as to facilitate their use in the development of seismic input criteria for nuclear plants and other major structures. They are also compared with results from prior investigations that have been used in the past in the criteria development for such major tacilities. KUREG/CR-1643 Geotechnical Data f rom Accelerograph Stations Invest gated during the Period 1975-1979. Summary Report. September 1980. Shannon & Wilson; A l babian Assoc ORES GPO. NTIS. This report summarizes geotechnical data that was obtained in the investigation of 83 accelerograph stations located in the United Staten. These stations were studied during the period from 1975 to 1979 and the detailed findings are contained in nine data reports. Summary logs indicating subsurface soil conditions and material ~ properties have been prepared for each of the accelerograph stations. A classification system was devised for grouping the stations as either rock sites, stiff soil sites, or deep soil sites. Using this classification system, simple ground motion plots have been prepared which qualitatively indicate the dependency of earthquake motions on local site conditions. This classification system may be used in more elaborate and quantitative studies of the influence of local site conditions upon earthquake ground response. On a practical engineering level, the site classification system and the results of the individual site investigations may be used in selecting earth-quake records to establish seismic design criteria. Further research of the subsurface conditions at additional accelerograph stations is needed to increase the data base of earthquake records and recording stations, NUREG/CR-1648 A Probabilistic Evaluation of Earthquake Detection and Location Capability for Illinals, Indiana, Kentucky, Ohio, and West Virginia, September 1980. University of Michigan ORES GPO. NTIS. Probabilistic estimations of earthquake detection and location capabilities for the states of Illinois, Indiana, Kentucky, Ohio, and West Virginia are presented in this document. The algorithm used in these epicentrality and minimum-magnitude estimations is a version of the program KETWORTH by Wirth, Blandford, and Husted (DARPA Order No. 2551, 1978) which was modified for local array evaluation at the University of Michigan Seismological Observation. Estimations of earthquake detection capability for the years 1970 and 1980 are presented in four regional minimum ab magnitude contour maps. Regional 901 confidence error ellipsoids are included for mb magnitude events from 2.0 through 5.0 at 0.5 m bunit increments. The close agreement between these predicted epicentral 90% confidence estimates and the calculated error ellipses associated with actual earthquakes within the studied region suggest that these error determinations can be used to estimate the reliability of epicenter location. KUREG/CR-1649 Geophysical Investigation of the Anna, Ohio Earthquake Zone. September 1980. University of Michigan ORES GPO. NTIS. This report discusses the progress and achievements accomplished under KRC Contract No. NRC-04-76-192 during fiscal year 1980. The Anna, Ohio scismic array, converted i to solar recharge power systems, has been in continuous operation. No local earthquakes ' above a 1.5 have occurred. Near regional earthquakes from 1977 through 1980 supple-mentedb with quarry blast recordings have been used to determine the regional travel time curves. Theoretical estimates of earthquake detection and location capabilities for g 2.5, 7.0, and 1.5 earthquakes in the Anna, Ohio region are included to demon-- strate the coverage effectiveness of the network. Teleseismic P-wave residuals as a function of azimuth are included to demonstrate the lower crustal velocity variation for the region. Finally, an exhaustive catalog of water and gas well data is included from which a regional depth to bedrock map has been produced.
i I Report No. Bibliographic Data NUREC/CR-1656 Utility Management and Technical Resources. September 1980. Teknekron Research, McLean, VA ONRR GPO. NTIS. NRR contracted with Teknekron Research, Inc. to analyze and evaluate utility management and technical resources for dealing with events like that at Three Mile Island Unit 2.
- Teknekron (1) analyzed licensee submittals in response to an NRC request to identify managment and technical short-term and long-term resources for reacting to THI-2 type accidents, (2) developed acceptance criteria that specify minimum management and technical (onsite and offsite) resources, and (3) evaluated the adequacy of licensee management and technical resources (onsite and offsite). The general conclusion is that those resources described in licensee submittals may be capable of dealing with THI-2 type accidents.
NUREG/CR-16S7 Steam-Water Mixing and System Hydrodynamics Program - Task 4. Quarterly Progress Report, October-December 1979. August 1980. Battelle Columbus Lab BMI-2062 ORES GPO. NTIS. During this quarter, snalysis included development of a new correlation for air-water data in terms of the K* scaling parameter, and implementation of this correlation into the mechanistic model for ECC penetration. Experimental work this quarter included studies of annulus and break leg flow with neutral and equilibrium walls, hot wall tests in the rectangular test section, and development of a data reduction program for the void distribution measurerent (VDM) system. NUREG/CR-1660 Compilation, Assessment and Expansion of the Strong Earthquake Ground Hotion Data Base. September 1980. Lawrence Livermore Lab UCRL-15227 ORES GPO. NTIS. A catalog has been prepared which contains information for: (1) world-wide, ground-
- motion accelerograms, (2) the accelerograph sites where these records were obtained, and (3) the seismological parameters of the causative earthquakes. The catalog is limited to data for those accelerograms which have been digitized and published. In addition, the quality and completeness of these data are assessed. This catalog is unique because it is the only publication which contains comprehensive information on the recording conditions of all known digitized accelerograms. However, information for many accelerograms is missing. Although some literature may have been overlooked, most of the missing data has not been published. Nevertheless, the catalog provides a convenient reference and useful tool for earthquake engineering research and applicttions.
NUREG/CR-1661 Variability of Dynamic Characteristics of Nuclear Power Plant Structures. September 1980. Lawrence Livermore National Lab UCRL-15267 ORES GPO. NTIS. This report presents the results of an investigation of the sources of random vari-ability of the dynamic response of nuclear power plant structures. Sources affecting both the response frequencies and dynamic amplification of structures are identified. Numerical values developed for the Zion auxiliary building are presented for sources of inherent randomness. Several sources' of uncertainty resulting f rom lack of knowledge of material properties or approximations 'in analytical modelling are discussed but are not in general quantified. The dispersion in both the structure dynamic character-istics and the input to equipment as defined by the in-structure response spectra is addressed. The evaluation of the dynamic response variability is limited to elastic
- response levels.
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Report No. Bibliogrrphic Dtte NUREG/CR-1668 Advanced Mobile Multi-Processor Gamma-Ray Acquisition and Analysis System. September 1980. EG&G Idaho ORES GPO. NTIS. The report describes a new Gamma-Ray Acquisition and Analysis system which has been developed for the In-Plant Source Term Measurement Program. A new computer was added to the system described previously in Reference 5, " Procedures, Source Term Measurement Program," NUREG-0384. One computer as now used to acquire the data and the other (new computer) is used to analyze the resulting data. The throughput of the system has been dramatically isiproved. Data analysis times have been reduced by about a factor of 10. Moreover, the analysis procedure is much more complex and provides results which can be directly reported with minimal operator interpretation. The information contained in this report supersedes the description of the analysis package given in Appendix B of the procedures manual (NUREG-0384). NUREG/CR-1670, Vol 1 The Use of Process Monitoring Data for Nuclear Material Accounting: Volume 1. ' Summary Report. August 1980. Battelle Pacific Northwest Lab PNL-3396 ONMSS GPO. NTIS. A study was conducted for the Nuclear Regulatory Commission as part of a continuing program to estimate the effectiveness of using existing production control, process control, and quality control data to enhance strategic special nuclear material (SSNM) control and accounting of nuclear fuel manufacturing licensees. Two licensed SSNM fuel fabrication facilities with internal scrap recovery processes were examined. The loss detection sensitivity, timeliness, and localization capabilities of these techiques were evaluated for single and multiple (trickle) losses of material undergoing processing. The impact of records manipulation, mass and isotopic substitution, and collusion between insiders on these methods for detecting diversion were also studied. Volume 1 is an unclassified, nonproprietary summary. Volume 2 contains details on the mixed oxide fabrication process studied; its availability is restricted because it contains proprietary intormation. Volume 3 has details on an HEU fabrication process; it contains classified and proprietary information and so its availability is also limited. NUREG/CR-1671 Transportation of Radioactive Material in Kentucky. September 1980. Bureau of Health Services, Frankfurt, KY OSP GPO, NTIS. Shipments of radioactive materials into, within, or through Knatucky were surveyed to determine the types of materials, pattern of traceportation and wagnitude of activity, the extent of compliance wita shipping regulations, and the radiation exposure to persons handling the materials. The transported radioactive materiais were categorized as (1) local delivery service, (2) air carrier, (3) nuclear pharmacy, (4) highway carriers, (5) nuclear fuel cycle. The shipments with the most numerous packages were radiopharmaceuticals. The shipments iridicating the greatest volume or amount were those associated with the nuclear fuel cycle. The transportation workers whose radiation exposures were measured did not receive excessive doses from radioactive materials, but practices for reducing the rtJiation doses can be instituted and are discussed in the report. KUREG/CR-1676, Vol 1 Using Advanced Process Monitoring to improve Material Control. September 1980. , NUSAC, McLean, VA NUSAC-556 ONMSS GPO. NTIS. The MC&A Task Force Report (NUREG-0450) included the long-term recommendation that the NRC give the highest priority to development and demonstration of process monitoring techniques for timely detection of material losses. NUREG/CR-1676 and 1687 are the final reports of two contractors selected to competitively perform the concept definition phase of a project responsive to that recommendation. In this phase each contractor independently formulated and estimated the costs and effectiveness of alternative data generating systems for validating the presence of KEU fuel material undergoing processing. Existing production control, quality control, and process control data were supplemented and relationships between cost and level of safeguards
l l l l Report No. Bibliographic Data ' performance obtained. . Volumes I are unclassified, nonproprietary reports. Classified or proprietary detail 6 are in Volumes 2. NUREG/CR-1677, Vol 1 Piping Benchmark Problems. Dynamic Analysis Uniform Support Motion Response Spectrum Method.. August 1980. Brookhaven National Lab BNL-NUREG-51267 ONRR GPO. NTIS. A set of benchmark problems and solutions have been developed for verifying the adequacy of computer programs used for dynamic analysis and design of nuclear piping systems by the Response Spectrum Method. The problems range from simple to complex configurations which are assumed to experience linear <iastic behavior. The dynamic loading is represented by uniform support motion, assumed to be induced by ' seismic excitation in three spatial directions. The solutions consist of f requencies, participation factors, nodal displacement enmponents, and internal force and moment components. Solutions to associated anchor point motion static problems are not t?cluded. NUREG/CR-1683 Characterization of Existing Surface Conditions at Sheffield Low-Level Waste Disposal Facility. August 1980. Harding-Lawson Assoc, Oak Brook, IL HLA-9906-001-14 ONMSS uPO. NTIS. This report presents results of the lavestigation to characterize the existing surface conditions at the Sheffield Low-Level Waste Disposal Facility, Sheffield, Illinois. The investigation is based on visua. observations made in the field and detailed topographic surveying. The following information is presented: (1) Analyses of individual trench caps describing surf ace conditions and the ability of trench caps to minimize erosion and water infiltration into trenches; (2) A detailed survey of erosion at the site and a detailed description of the vegetation; and (3) A topographic rurvey. Numerous photographs are provided to document observations. Possible remedial actions, conclusion, and recommendations for improving surface conditions are presented. Field observations of the Shef field site presented in this report were made July 1980. NUREG/CR-1730 Data Summaries of Licensee Event Reports of Primary Containment Penetrations at U.S. Commercial Nuclear Power Plants from January 1, 1972 to December 31, 1978. September 1980. EG&G Idaho EGG-EA-5188 ORES GPO. NTIS. Thts repo-t describes the results of an analysis of nuclear plant primary containment penetration failure; The data used for this analysis were the Licensee Event Reports (LERs). The LERs are written reports filed with the NRC whenever certain f ailures or iticidents occur concerning nuclear plant safety systems. The prima ry containment penetration failures or incidents contained in the LERs were evaluated and categorized as to type of failure or problem and were used to calculate summary primary contain-ment penetrating failure rate statistics. The report includes a ve:1?*v of different statistics calculated to highlight or show important failure modes or uther failure information. In addition to the quantitative failure rate information, there is also considerable qualitative information tabulated to allow the user to make additional primary containment penetration failure rate calculations or inferences. l
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Keyword Index to Reports 1 l l l l l l l 4
Report Title Report No. Keyword List.ag A ABG Accidents Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents. Abnormal Occurrences Report to Congress on Abnormal Occurrences, January- NUREG-0090, Vol 3, No. 1 ' March 1980. Accelerated Predicting 1.ife Expectancy and Simulating Age of Complex 'NUREG/CR-1466 ' Equipment Using Accelerated Aging Techniques. Accelerograph Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, V. 2 App Accelerograph Stations in California. Accelerograph Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, Vol 2 Accelerograph Stations in California. Verification of Subsurf ace Conditions at Selected " Rock" NUREG/CR-0055, Vol 3 Accelerograph Accelerograph Stations in California. Accelerograph Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 3 Accelerograph Stations.
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1 Accelerograph Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 4 Accelerograph Stations. Accelerograph Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 5 Accelerograph Stations. Accelerograph Stations Site-Dependent Ef fects at Strong-Motion Accelerograph NUREG/CR-1639 Stations. Accelerograph Stations Geotechnical Data from Accelerograph Stations NUREG/CR-1643 Investigated during the Period 1975-1979. Summary Report. Acceptance Criteria MARK II Containment Lead Plant Progra,s Load Evaluation and NUREG-0487, Supp 1 Acceptance Criteria - Generic Technical Activities AB and A39. Acceptance Criteria for the Physical Protection Upgrade NUREG-0721 Acceptance Criteria Rule Requirements for Fixed Sites. Accident NRC Action Plan Developed as a Rehult of the TMI-2 NUREG-0660, Vol 1 Accident, Revision 1 Vol 1. 1 I 1
.. _ _ _ . . _ _ . ~ .. . . - - . . . ....~m.m.._.m. ._ . _. _ i Keyword Listing A . Report Title Report No.' ! Accident NRC Action Plan Developed as a Result of the THI-2 NUREG-0660, Vol 2 - Accident, Revision 1. Vol. 2. Accident Calculations of the Skyshine Gamma-Ray Dose Rates f rom NUREC/CR-0723- . Independent Spent Fuel Storage Installations (ISFSI) Under . i Worst Case Accident Conditions. [ i Accident Uncertainty Analysis for a IW Loss-of-Coolant Accident: NIREG/CR-1364 )
- 11. Alternative Core Damage Estimators.
Accident Drop-Size Estimates for a 1.oss-of-Coolant Accident. NUREC/CR-1607 i Accidents Review and Assessment of Package Requirements (Yellowcake) NUREG-05.o l and Emergency Response to Transportation Accidents, f Accidents A Risk Assessment of a Pressurized Water Reactor for Class NUREC/CR-0603 3-8 Accidents, i t Accounting Material Accounting as Required by the United States NUREG/CR-1192 ,; Wuclear Regulatory Commission: Capabilities and ( Vulnerabilities. Accounting The Use of Process Monitoring Data f or Nuclear Material NUREG/CR-1670, Vol 1 Accounting: Volume 1. Summary Report, f Accounting Program Safeguards Material Control and Accounting Program: NUREG/CH-1485, Vol 1, No. 1 l' Quarterly Report, October-December 1979. I I Acquisition Report to Congress on the Acquisition of Reactor Data NUREG-0730 ' for the NRC Operations Center, [ Acquisition System Advanced Mobile Multi-Processor Gamma-Ray Acquisition NUREG/CR-1668 + and Analysis System. [ i Action Plan NRC Action Plan Developed as a Result of the TMI-2 NUREG-0660, Vol 1 ! Accident, Revision 1, Vol 1.
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Action Plan NRC Action Plan Development as a Result of the THI-2 NUREG-0660, Vol 2 i Accident, Revision 1 Vol. 2. ' s Activated Carbon The Effects of Temperature, Moist.ure, Concentration, NUREG-0678 Pressure and Mass Transfer on the Adsorption of Krypton # and Xenon on Activated Carbon. r I i
_,5_.. ,a1 x- J 4 & <-e-4 s 1i .-..:- A e _. .-p4a 4 Report No.
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Keyword Listing A Report Title e Activities Potential Threat to Licensed Nuclear Activities from NUREG-0703 Insiders (Insider Study).
' i Acute Effects Acute Effects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF6/UO 2 2F Studies in Experimental Animals.
Acute Toxicity Acute Toxicity and Bioaccumulati Chloroform to Four NUREC/CR-0893 Species of Freshwater Fish. Adequacy Evaluation of Simulator Adequacy for the Radiation NUREG/CR-ll84' Qualification of Safety Related Equipment. i Adsorption The Effects of Temperature, Moisture, Concentration, NUREG-0678 Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon. Advanced Two-Phase Flow Measurements with Advanced Inc trumented NUREG/CR-1529 Spool Pieces. Advanced Using Advanced Process Monitoring to Improve Material NUREG/CR-1676, Vol 1 Control. Advanced Mobile Advanced Mobile Multi-Processor Gamma-Ray Acquisition NUREG/CR-1668 and Analysis System. Advanced Reactor Advanced Reactor Safety Research Division Quarterly NUREG/CR-1402 Progress Report. October 1-December 31, 1979. Advanced Reactor Advanced Reactor Safety Research Division Quarterly NUREC/CR-1505 Progress Report, January 1-Maren 31, 1980, Aging Techniques Pre 6 feting Life Expectancy and Simulating Age of Complex NUREG/CR-1466 Equipment Using Accelerated Aging Techniques. Agreement Canadian Seismic Agreement. NUREG/CR-1637 Air Measurement of XE-133, C-14 and Tritium in Air and 1-131 NUREG/CR-1195 i in Vegetation and Milk Around the Quad Cities Nuclear i Power Station. Airborne Solubility Classification of Airborne Uranium Products NUREG/CR-1428 i from LWR-Fuel Plants. 1
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- i. Keyword Listin_g A Report Title Report No.
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~ Alabama Power Co. Safety Evaluation Report Related to the Operation of NUREG-Oll7, Supp. 4 Joseph M. Farley Nuclear Plant, Unit 2. Docket No. 50-364, Alabama Power Company. Supplement 4 to NUREG-75/034.
{ l I Algorithm An Algorithm to Estimate Field Concentrations Under NUREG/CR-1474 Nonsteady Meteorological Conditione from Wind Tunnel Experiments. J 1 1 Alternative Uncertainty Analysis for a PWR Loss-of-Coolant Accident: NUREG/CR-1364 l
- 11. Alternative Core Damage Estimators, i 1
A' ernative Fuct :iechnical Safeguards lasues for Alternative Fuel Cycles. NUREG/CR-1048 Alternativea Psychological Stress for Alternatives of Decontamination NUREG/CR-1584 of THI-2 Reactor Building Atmosphere. a Alternatives Performance Testing of Personnel Dosimetry Services: NUREG/CR-1593 Alternatives and Recommendations for a Personnel 1 Dosimetry Testing Progsam. i Analyses Critical Experiments, Measurements, and Analyses to NUREG/CR-1601 ! Establish a Crack Arrest Methodology for Nuclear Pressure Vessel Steels. Analysis A Comparative Analysis of LWR Fuel Designs. NUREG-0559 Analysis Extended Analysis of Data from the 1/5-Scale MARK 1 NUREG/CR-0761 Bolling Water Reactor Pressure Suppression Experiment. Analysis Dynamic, Inelastic Buckling Analysis of NARK 1 Torus WUREG/CR-1038 Support Columns. 1 I Analysis Pip Inelastic Fracture Mechanics Analysis. NURCG/CR-lll9 l Analysis Vital Area Analysis Using SETS. FifREG/CR-1487 Analysis Models SCALE: A Modular Core System for ierforming Standardized NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSASI and CSAS2. Analysis System Flow Topography Instrumentation and Analysis System. NUREG/CR-1333
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Keyword Listing _a Report Title Rtnort No. Analysis System Advanced Mobile Multi-Processor Gamma-Ray Acquisition NUREG/CR-1668 and Analysis System. Animals Acute Effects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF6/D0 2 2F Studies in Experimental-Animals. Anna, OH Geophysical Investigation of the Anna, Ohio Earthquake NUREG/CR-1649' f Zone. '! Annual Reports Nuclear Plant Reliability Data System 1979 Annual Reports NUREG/CR-1635, of Cumulative System and Component Reliability.
- Answers Answers to Frequently Asked Questions about Cleanup NUREG-0732 Activities at Three Mile Island, Unit 2.
Anticipated Experiment Data Report for LOFT Anticipated Transient NUREG/CR-1520 Experiment L6-5. Appalachian A Characterization of Faults in the Appalachian Foldbelt. NUREG/CR-1621 Appendix A Technical Specifications, Sequoyah Nuclear Plant, Unit No. NUREG-0658 Rev 1. 1, Docket No. 50-377, Appendix "A" to License No. DPR-77. e Appendix A North Anna Power Station Unit 2 Technical Specifications .NUREG-0664, Rev 1 Appendix "A" to License No. NPF-7. Appendix E Environmental Assessment for Effective Changes to 10 CFR NUREG-0685 Part 50 and Appendix E to 10 CFR Part 50; Emergency : Planning Requirements for Nuclear Power Plants. Area Analysis Vital Area Analysis Using SETS. NUREG/CR 1487 Arrays Critical Experiments with Interstitially-Moderated NUREG/CR-1071 Arrays of Low-Enriched Uranium oxide. Topical Report on Reference Critical Experir <nts. , Assessment Review and Assessment of Package Requirements (Yellowcake) NUREG-0535 and Emergency Response to Transportation Accidents. Assessment Environmental Assessment for Effective Changes to 10 CFR. NUREG-0685 Part 50 and Appendix E to 10 CFR Pact 50; Emergency Planning Requirements for Nuclear Power Plants. '
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I Keyword Listing A heportTitle Report No. Assessment A Risk Assessment of a Pressurized Water Reactor for Class NUREG/CR-0603 3-8 Accidents. 4 Assessment Identification and Assessment of the Social Impacts of NUREG/CR-0744 Transportation of Radioactive Materials in Urban Enviranments. 4 m Assessment Licensability of CANDU-Type' Reactors in the United States. KUREG/CR-Ill3
- A Preliminary Assessment of the R and D Requirements.
4 Assessment Assessment of Current Onsite inspection Techniques for NUREG/CR-1380, Vol 1, ES Light Water Reactor Fuel Systems - Volume 1 - Executive Summary. x Assessment An Assessment of LWR Fuel-Failure Propagation Potential: NUREG/CR-1471 -l Literature Survey. Assessment Assessment of Core Penetration of a PWR. Reactor Vessel NUREG/CR-IS18 I and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents. . Assessment Compilation, Assessment and Expansion of the Strong NUREG/CR-1660 Earthquake Ground Motion Data Base. i Atmosphere Psychological Stress for Alternatives of Decontamination NUREG/CR-1584 i of TMI-2 P.r3.; or Building Atmosphere. Atmospheric Dirfusion Near Buildings as Determined from Atmospheric NUREG/CRal394 Tracer Experiments. 1 Availability COPS Model Estimates of LLEA Availability Near Selected Reactor Sites, NUREG/CR-ll66. k I 4 1 h
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Keyword Listing B Report Title Report No. 1 Battelle-Frankfurt Comparison of CONTEMPT-LT Containment Code Calculations NUREC/CR-1564 with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. Behavior Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 1 Plume Behavior from Cooling Towers: Vol. 1--Dispersion from Single and Multiple Source Natural Draft Cooling Towers. Behavior Evaluation of Mathematical Models for Characterizing KUREG/CR-1581, Vol 2 1 Plume Behavior from Cooling Towers: Vol.2--Salt Drift Deposition from Natural Draft Cooling Towers. Behavior Evaluation of Mathematical Models for Characterizing KUREC/CR-1581, Vol 3 Plume Behavior from Cooling Towers: Vol. 3--Plume Rise from Mechanical Draft Cooling Tovers. Benchmark Problems Piping Benchmark Problems. Dynamic Analysis Uniform KUREG/CR-1677, Vol 1 Support Motion Response Spectrum Method. Best Estimate Best Estimate Method vs Evaluation Method: A Comparison KUREG/CR-1489 of Two Techniques in Evaluating Seismic Analysis and , Design. Bioactumulation Acute Toxicity and Bioaccumulation of Chloroform to Four NUREG/CR-0893 + Species of Freshwater Fish. Blowdown LWR fuel Rod Post-Subcooled Blowdown Scoping Analysis. NUREG/CR-1568 Blowdown Tests Comparison of CONTEMPT-LT Containment Code Calculations KUREG/CR-1564 with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. Body Radioactivity Health Status and Body Radioactivity of Former Thorium KUREG/CR-1420 Workers. , f Boiling Pools Heat Removal Characteristics of Volume Heated Boiling KUREC/CR-1357 Pools with Inclined Boundaries. r Borehold Estimates of Uranium Content and Radon Flux for Uranium NUREG/CR-1549 Mine Dumps Based on Borehold Radioactivity Logs. Boundaries Heat Removal Characteristics of Volume Heated Boiling KUREG/CR-1357 Pools with inclined Boundaries. , Boundary Technical Report on Material Selection and Processing NUREG-0313, Rev. 1 Guidelines for BWR Coolant Pressure Boundary Piping. h m, - - - - -- , _ . - - - - - - - - - - - _ ----
- _. ..m . . . . __ .- . ._ m. . . .m - . _ - . . Keyword Listing B Report Title Report No. Boundary PRESBC: Pressure Boundary Conditions for the K-FIX Code., NUREG/CR-1536-Boundary Components Structural integrity of Water Reactor Pressure Boundary NUREC/CR-1472 Components. Quarterly Progress Report,' January-March 1980. Breathing Apparatus Respirator Studies for the Nuclear Regulatory Commission, NUREG/CR-1586' Evaluation and Performance of Escape-Type Self-Contained Breathing Apparatas, October 1,.1978-September 30, 1979. Buckling Analysis Dynamic, Inelastic Buckling Analysis of MARK I Torus NUREG/CR-1038 Support Columns. Budget Conunents on the NRC Safety Research Program Budget for NUREG-0699 Fiscal Year 1982. Buildings Diffusion Near Buildings as Determined from Atmospheric NUREG/CR-1394 Tracer Experiments. Burial Evaluation of Isotope Migration-Land Burial Water NUREG/CR-1513 Chemistry at Commercially Operated 1,ow-Level Radioactive Waste Disposal Sites. Quarterly Progress Report. January-March 1980. Burial Sites Vegetational Cover in Monitoring and Stabilization of NUREG/CR-135' Shallow Land Burial Sites. Burst Prompt Burst Energetics Experimentst ' Fresh Uranium NUREG/CR-1396 Carbide / Sodium Series. Burst Test Multirod Burst Test Program Progress Report for NUREG/CR-1450 July-December 1979. BWR Technical Report on Material Selection and Processing NUREG-0313, Rev. 1 Guidelines for BWR Coolant Pressure Boundary Piping. BWR Extended Analysis of Data from the 1/5-Scale MARK 1 NUREC/CR-0761 Boiling Water Reactor Pressure Suppression Experiment. J
Report No. Keyword Listing C Report Title Development and Verification of Fire Tests for Cable NUREG/CR-1552 Cable Systems Systems and System Components. Preparation of Working Reference Materials: Calcined NUREG/CR-1445 Calcined Waste Waste Recovery Products Containing Uranium or Plutonium. A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 Calculate the Motion of fuel and Coolant Subsequent to Pin Failure in a LMFBR. Calculating Heat User's Manual for USINT: A Program for Calculating NUREG/CR-1375 Heat and Mass Transfer in Concrete Subjected to High Heat Fluxes. Calculations of the Skyshine Gamma-Ray Dose Rates from NUREG/CR-0723 Calculations Independent Spent Fuel Storage Installations (ISFSI) Under Worst Case Accident Conditions. Summary of Thermal Hydraulic Calculations for a NUREG/CR-1480 Calculations Pressurized Water Reactor. Comparison of CONTEMPT-LT Containment Code Calculations NUREG/CR-1564 Calculations with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. NUREG/CR-0055, V. 2 App California Verification et Subsurface Conditions at Selected " Rock" i Accelerograph Stations in California. l l NUREG/CR-0055, Vol 2 l California Verification of Subsurface Conditions at Selected " Rock" Accelerograph Stations in California. j Verification of Subsurf ace Conditions at Selected " Rock" NUREG/CR-0055. Vol 3 California Accelerograph Stations in California. Canadian Seismic Agreement. NUREG/CR-1637 Canadian Licensability of CANDU-Type Reactors in the United States. NUREG/CR-all3 CANDU-Type A Preliminary Assessment of the R and D Requirements. Material Accounting as Required by the United States NUREG/CR-1192 Capabilities Nuclear Regulatory Comission: Capabilities and Vulnerabilities. Survey of Currant State Raliological Emergency Response NUREG/CR-1620 Capabilities Capabilities for Transportation Related Incidents. i l l l
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r Keyword Listing C Report Title Report No. Carbide /Sodima Prompt Burst Energetics Experiments: Fresh Uranium NUREG/CR-1396 Carbide / Sodium Series. 1 i Carbon-14 Measurement of XE-133, C-14 and Tritium in Air and'l-131 NUREG/CR-1195' in Vegetation and Milk Around the Quad Cities Nuclear Power Station. Changes Environmental Assessment for Effective Changes to 10 CFR NUREG-0685 ;
-Part 50 and Appendix E to 10 CFR Part 50; Emergency '
Planning Requirements for Nuclear Power Plants. l I Characteristics Heat Removal Characteristics of Volume Heated Boiling NUREG/CR-1357 Pools with inclined Boundaries.
' Characteristics Evaluation of In-Situ Soil Damping Characteristics. NUREG/CR 1638 l Characteristics Variability of Dynamic Characteristics of Nuclear Power NUREG/CR-1661' Plant Structures.
Characterization Gap Conductance Test Series Fuel Characterization Data NUREG/CR-1537-Report. Characterization' A Characterization of Faults in the Appalachian Foldbelt. NUREG/CR-1621 Characterization Characterization of Existing Surf ace Conditions at NUREG/CR-1683 Sheffield Low-Level Waste Disposal Facility. Characterizing Evaluation of Mathematical Models for Characterizing KUR2G/CR-1581, Vol 1 Plume Behavior from Cooling Towers: Vol. 1--Dispersion from Single and Multiple Source N?tural Draft Cooling Towers. Characterizing Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 2 Plume Behavior from Cooling Towers: Vol.2--Salt Drift Deposition from Natural Draft Cooling Towers. Characterizing Evaluation of Mathematical Models for CharacterizinC NUREG/CR-1581, Vol 3 Plume Behavior from Cooling Towers: Vol. 3--Plume Rise from Mechanical Draft Cooling Towers. Chemical Final Environmental Statement Related to Primary Cooling NUREG-0686 System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Docket No. 50-010. s Chlorinated Growth and Histological Effects to Protothaca Staminea NUREG/CR-1298 (Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water. > I l
Keyword Listing C Report Title Report No. .l Chloroform Acute Toxicity and Bioaccumulation of Chloroform to Four NUREG/CR-0393 Species of freshwater Fish. . Chugging Lateral Loads on Vent Pipe in Steam Chugging. NUREG/CR-1631 Class 3-8 A Risk Assessment of a Pressurized Water Reactor for C1r ss NUREG/CR-0603 3-8 Accidents. Classificaton Solubility Classification of Airborne Uranium Products NUREG/CR-1428 from LWR-Fuel Plants. Cleanup NRC Plans for Cleanup Operations at Three Mile Island NUREG-0698 Unit 2. Cleanup Activities Answers to Frequently Asked Questions about Cleanup NUREG-0732 Activities at Three Mile Island, Unit 2. Closed Containment Hydrogen-Mixing in a Closed Containment Compartment Based NUREG/CR-1575 on a One-Dimensional Model with Convective Effects. Clusters Criticality Experiments with Suberitical Clusters of NUREG/CR-1547 2.35 Wt% and 4.31 Wt% U-235 Enriched UO Rods in Water at a Water-to-Fuel Volume Ratio of 1.6.2 Coal A Methodology and a Preliminary Data Base for Examining NUREG/CR-1539 the Health Risks of Electricity Generation from Uranium and Coal Fuels. Coast-Down Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592 to Sodium Boiling in Fuel Subassemblies during Pump Coast-Down of an LMFBR, Code System SCALE: A Modular Core System for Perfarming Standardized NUREG/CR-0200 l Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSASI and CSAS2. i Combinatico Program Load Combination Program. Prograss Report No. 5. NUREG/CR-1624 Comments Comments on the NRC Safety Research Program Budget for NUREG-0699 Fiscal Year 1982. Commercial Power Plants Data Summarier of Licensee Event Reports of Primary NUREG/CR-1730 Containment Penetrations at U.S. Commercial Nuclear Power Plants from January 1, 1972 to December 31, 1978. 1
.- _-_ ,-- . . _ . - . . . - . . . .- . - - . ~ l Keyw.rd Listing C Report Title Report No. 1 1 Commercially Operated Evaluation of Isotope Migration Land Burial Water Chemistry KUREG/CR-1289
.at Commercially Operated Low Level Radioactive Waste Disposal Sites. r Commercially Operated Evaluation of Isotope Migration-Land Burial Water- NUREG/CR-1513 Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report, Janua ry-March 1980.
Communication Security Communication Systems for Nuclear Fixed-Site NUREG/CR-0508 Facilities. Communications Report to Congress on NRC Emergency Communications. NUREG-0729 Comparative Analysis A Comparative Analysis of LWR Fuel Designs. CREG-0559 COMPARE-MOD 1 COMPARE-MOD 1 Code, Addendum 1. NUREG/CR-1185 Comparement ifydrogen Mixing in a Closed Containment Compartment Based NUREG/CR-1575 on a One-Dimensional Model with Convective Effects. Comparison Best Estimate Method vs Evaluation Method: A Comparison NUREG/CR-1489 of Two Techniques in Evaluating Seismic Analysis and Design. Comparison Comparison of CONTEMPT-LT Containment Code Calculations NUREG/CR-1564 with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. Compilation Regulatory and Technical Reports Compilation for 1979. NUREG-0304, Vol. 4 Compilation Compilation, Assessment and Expansion of the Strong NUREG/CR-1660 Earthquake Ground Motion Data Base. Complex Predicting Life Expectancy and Simulating Age of Complex NUREG/CR-1466 > Equipment Using Accelerated Aging Techniques. Component Reliability Nuclear Plant Reliability Data System 1979 Annual Reports NUREG/CR-1635 of Cumulative System and Component Reliability. I Comp (nents Structural Integrity of Water Reactor Pressure Boundary NUREG/CR-1472 Components. Quarterly Progress Report, January-March 1 1980. '
Keyword Listing C Report Title Report No. Compressible Analysis Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592
' to Sodium Boiling in Fuel Subassemblies during Pump Coast-Down of an LMTBR.
Computer Analyses SCALE: A Modular Code System for Performing Standardized NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticalsty Safety Analysis Modules CSASI and CSAS2. Report on Nuclear Industry Quality Assurance Procedures NUREG-0653 Computer Code for Safety Analysis Computer Code Development and Use. Computer Code CONAN: An LMFBR Containment Response Computer Code. NUREG/CR-1355 SIMMER-II: A Computer Program for LMFBR Disrupted Core NUREG/CR-0453, Rev 1 Computer Program Analysis. Computer Program A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR 1504 the Motion of Fuel and Coolant Subsequent to Pin Failure in a LMFBR. CONAN CONAN: An LMFBR Containment Response Computer Code. NUREG/CR-1355 The Effects of Temperature, Moisture, Concentration, NUREG-0678 Concentration Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon. Concrete User's Manual for USINT: A Program for Calculating NUREG/CR-1375 Heat and Mass Transfer in Concrete Subject to High Heat Fluxes. Concrete Panels Strength and Stiffness of Tensioned Reinforced Concrete NUREG/CR-1602 Panels Subjected to Membrane Shear, Two-Way Reinforcing. Condensation-Induced An Evaluation of Condensation-Induced Water Hammer in NUREG/CR-1606 Preheat Steam Generators. Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, V. 2 App Conditions Accelerograph Stations in California. Verification of Sursurface Conditions it Selected " Rock" NUREG/CR-0055, Vol 2 Conditions Accelerographs Stations in California. Verification of S 4bsurface Conditions at Selected " Rock" NUREG/CR-0055, Vol. 3 Canditions Accelerograph Stations in California.
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Keyword LiatIng C Report Title Report Noi ' Conditions Calculations of the Skyshine Gamma-Ray Dose Rates from NUREG/CR-0723 Independent Spent fuel Storage Installations (11FSI) Under Worst Case Accident Conditions. ' Conditions An Algorithm to Estimate Field Concentrations Unde. NUREG/CR-1474 Nonsteady Meteorological Conditions from. Wind Tunn'i , Experiments. Conditions PRESBC: Pressure Boundary Conditions for the K-FIX Code. NUREG/CR-1536 ' l Conductance Gap Conductance Test Series Fuel Characterization Data- NUREC/CR-1537-Report. Congress Report to Congress on Abnormal Occu Tences, January- NUREG-0090, Vol 3, No. I March 1980. Congress Report to Congress: NRC Incident Response Plan. NUREG-0728 Congress Report to Congress on NRC Emergency Communications. NUREG-0729 Congress Report to Congress on the Acquisition of Reactor Data NUREG-0730 for the NRC Operations Center. Considerations Considerations on Nuclear Data Link Implementation. in NUREG/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System. , i Container Study of Plutonium 0xide Powder Emissions f rom Simulated NUREG/CR-1302 i Shipping Container Leaks. J Containers Shock Environments for Large Shipping Containers During ' NUREC/CR-1277 Rail Coupling Operations. Containers Properties of Radioactive Wastes and Waste Containers. NUREG/CR-1514 i Quarterly Progress Report, January-March 1980. Containment MARK 11 Containment Lead Plant Program Load Evaluation and NUREG-0487, Supp 1 Acceptance Criteria - Generic Technical Activities A8 and A39. Containment MARK I Containment Long Term Program Safety Evaluation NUREG-0661 Report. Resolution of Generic Technical Activity A-7; February 1977 to December 1979. i I I 1 1
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Keyword Listing C Report Title Report No. Containment CONAN: An LMFBR Containment Response Computer Code. NUREC/CR-1355 Containment Comparison of CONTEMPT-LT Containment Code Calculations NUREG/CR-1564' with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. Containment ifydrogen Mixing in a Closed Containment Compartment Based NUREG/CR-1575 on a one-Dimensional Model wita Convective Effects. Containment Data Summaries of Licensee Event Reports of Primary NUREG/CR-1730 Containment Penetrations at U.S. Commercial Nuclear Power Plants f rom January 1,1972 to December 31, 1978. Contaminant A Deterministic-Probabilistic Model for Contaminant NUREG/CR-1609 Transport. ; CONTEMPT-LT Comparison of CONTEMPT-LT Containment Code Calculations NUREG/CR-1564 with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. Control Control of Heavy Loads at Nuclear Power Plants NUREG-0612 Resolution of Generic Technical Activity A-36. Control Using Advanced Process Monitoring to Improve Material NUREG/CR-1676, Vol 1 Control. Convective Effects Hydrogen Mixing in a Closed Containment Compartment Based - NUREG/CR-1575 on a one-Dimensional Model with Convective Effects. i Coolability . Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents. Coolant Technical keport on Material Selection and Processing NUREG-0313, Rev. 1 Guidelines for BWR Coolant Pressure Boundary Piping. Coolant A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 the Motion of Fuel and Coolant Subsequent to Pin Failure in a LMFBR. , Coolant Flow Transient Analysis of Coolant Flow and Heat Transfer NUREG/CR-1404 in LMFBR Piping Systems. Cooling System Final Environmental Statement Related to Primary Cooling NUREG-0686 System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Docket No. 50-010.
Keyword Listing _C Report Title Report No. Cooling Towers Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 1 Plume Behavior from Cooling Towers: Vol. 1--Dispersion from Single and Multiple Source Natural Draft Cooling Towers. Cooling Towers Evaluation of Mathematical Models for Characterizing KUREG/CR-1581, Vol 2 Plume Behavior from Cooling Towers: Vol. 2--Salt Drift Deposition from Natural Draft Cooling Towers. Cooling Towers Evaluation of Mathematical Models for Characterizing KUREG/CR-1581, Vol 3 Plume Behavior from Cooling Towers: Vol . 3--Plume Rise from Mechanical Dcaft Cooling Towers. COPS COPS Model Estimates of LLEA Availability Near Selected NUREG/CR-Il66 Reactor Sites. Core Analysis SIMMER-II: A Computer Program for LMFBR Disrupted Core NUREG/CR-0453, Rev 1 Analysis. Core Damage Uncertainty Analysis for a PWR Loss-of-Coolant Accident: KUREG/CR-1364 II. Alternative Core Damage Estimators. Core Penetration Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents. Correlation Risk Methodology for Geologic Disposal of Radioactive NUREC/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. Coupling Operations Shock Environments for Large Shipping Containers During NUREG/CR-1277 Rail Coupling Operations. Crack Arrest Critical Experiments, Measureme.its, and Analyses to NUREG/CR-1601 Establish a Crack Arrest Methodology for Nuclear Pressure Vessel Steels. Cracking F4perience Pipe Cracking Experience in Light-Water Reactors, NUREG-0679 1967 through 1979. Cracking Incidents Investigation and Evaluation of Cracking Incidents in NUREG-0691 Piping in Pressurized Water Reactors. Criteria Functional Criteria for Emergency Response Facilities. NUREG-0696 Criteria Acceptance Criteria for the Physical Protection Upgrade NUREG-0721 Rule Requirements for Fixed Sites.
Keyword Listing _C Report Title Report No. Critical Experiments Critical Experiments with interstitially-Moderated NUREG/CR-1071 Arrays of Low-Enriched Uranium Oxide. Topical Report on Reference Critical Experiments. Critical Experiments' Critical Experiments, Measurements, and Analyses to NUREC/CR-1601 Establish a Crack Arrest Methodology for Nuclear Pressure' Vessel Steels. Criticality SCALEt A Modular Code System for Performing Standardized NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSASI and CSAS2. Criticality Experiments Criticality Experiments with Suberitical Clusters of NUREG/CR-1547 2.35 Wt% and 4.31 Wt% U-235 Enriched UO Rods in Water at a Water-to-Fuel Volume Ratio of 1.6.2 CSASl SCALE: A Modular Code System for Performing Standardized KUREG/CR-0200 Computer Analyses for Licensing Evaluation, SCALE System Criticality Safety Analysis Modules CSASI and CSAS2. CSAS2 SCALE: A Modular Code System for Performing Standardized NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSASI and CSAS2, Cumulative System Nuclear Plant Reliability Data System 1979 Annual Reports -NUREG/CR-1635 of Cumulative System and Component Reliability. Current Assessment of Current Onsite Inspection Techniques for NUREG/CR-1380, Vol 1, ES Light Water Reactor Fuel Systems - Volume 1 - Executive Summary. Current State Survey of Current State Radiological Emergency Response NUREG/CR-1620 Capabilities for Transportation Related incidents. , Cycle Fatigue High Cycle Fatigue Behavior of Incoloy 800H in a NUREG/CR-1356 Simulated High-Temperature Gas-Cooled Reactor Helium Environment. I i
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h Keyword Listing D Report Title Report No. Damping Evaluation of in-Situ Soil Damping Characteristics. NUREG/CR-1638 i Data Extended Analysis of Data from the 1/5-Scale MARK I NUREG/CR-0761 Boiling Water Reactor Pressure Suppression Experiment. [ f l Data Base Compilation, Assessment and Expansion of the Strong NL' REG /CR-1660 Earthquake Ground Motion Data Base. Data link Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System. Data Report Experiment Data Report for LOFT Anticipated Transient NUREG/CR-1520 Experiment L6-5. ; Data Report Gap Conductance Test Series Fuel Characterization Data NUREG/CR-1537 Report. Data Summaries Data Summaries of Licensee Event Reports of Primary NUREC/CR-1730 Containment Penetrations at U.S. Commercial Nuclear Power Plants from January 1, 1972 to December 31, 1978. Data System Nuclear Plant Reliability Data System 1979 Annual Reports NUREG/CR-1635 of Cumulative System and Component Reliability. Debris Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris coolability in TMLB, S20, and ABG Accidents. Decommissioning Plan for Reevaluation of NRC Policy on Decommissioning of NUREG-0436, Rev 1, Supp 1 Nuclear Facilities. December 1978 to' July 1930. Decommissioning Financing Strategies for Nuclear Power Plant NUREG/CR-1481 Decommissioning. Decontamination Final Environmental Statement Related to Primary Cooling NUREG-0686 System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Do'cket No. 50-010. Decontamination Psychological Stress for Alternatives of Decontamination NUREG/CR-1584 of TM1-2 Reactor Building Atmosphere. Deposition Acute Effects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF6 /UO2 F2Studies in Experimental Animals. F
. - - . . . . - ...~. - . . __ . . - - - - - . _ . . . . . - - . . . . - - . _ . - 1 I Keyword Listing D Report Title Report No. Deposition Evaluation of Kathematical Models for Charactrdizing NUREG/CR-1581, Vol 2 l Plume Behavior from Cooling Towers: Vol.2--Salt Drift ! Deposition from Natural Draft Cooling Towers. l t Design Guidance Design Guidance and Evaluation Methodology for Fixed KUREG/CR-1198, Vol 1 [ Site Physical Protection Systems, Volume:1. ; Design Guidance Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 2 Site Physical Protection Systems, Volume 2. ,[ Designs A Comparative Analysis of LWR Fuel Designs, NUREG-0559 Detection A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648 Location Capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia. i Determined Diffusion Near Buildings as Determined from Atmospheric NUREC/CR-1394 Tracer Experiments, i Deterministic A Deterministic-Probabilistic Mouel for Contaminant NUREG/CR-1609 Transport.
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t Development Report on Wuclear Industry Quality Assurante Procedures NUREG-0653 { for Safety Analysis Computer Code Development'and Use. r Development Development and Verification of Fire Tests for Cable- NUREG/CR-1552
~
Systems and System Components.
.l Development Scenario Development and Evaluation Related to the Risk NUREG/CR-1608 Assessment of High Level Radioactive Waste Repositories. l t
Diablo Canyon Safety Evaluation Report Related to Operation of Diablo KUREG-0675, Supp 10 Canyon Nuclear Power Station, Units I and 2, Pacific Gas > and Electric Company, Docket Nos. 50-275 and 50-323, Supplement No. 10. Diffusion Measurement of Radon Diffusion from Uranium Mill Tailing NUREC/CR-1109 i Piles. , k P Diffusion Diffusion Near Buildings as Determined from Atmospheric NUREC/CR-1394 Tracer Experiments. Dispersion Wind-Tunnel Measurements of Dispersion and Turbulence NUREG/CR-1475 in the Wakes of Nuclear Reactor Plants. r
r I l Report Title Report No. Keyword Listing _D Dispersion Evaluation of Mathematical Models for Characterizing KUREG/CR-1581, Vol 1 Plume Behavior from Cooling Towerst Vol. 1--Dispersion from Single and Multiple Source Natural Draf t Cooling ! Towers. Dispersion Radon Release and Dispersion from an Open Pit Uranium NUREG/CR-1583 Mine. Display System Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 , Relation to the Technical Support Center, Emergency Dperations Center and Safety Parameter Display System. i Disposal- ' Risk Methodology for Geologic Disposal of Radioactive KUREG/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank , Correlation Among Input Variables for Simulation Studies. Disposal Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1377 Waste: Transport Model Sensitivity Analysis. l Disposal Facility Characterization of Existing Surface Conditions at KUREG/CR-1683 Sheffield Low-Level Waste Disposal Facility, t Disposal Sites Evaluation of Isotope Migration-Land Burial Water Chemistry NUREC/CR-1289 at Commercially Operated Low Level Radioactive Waste Disposal Sites. Disposal Sites Evaluation of lootope Migration-Land Burial Water NUREG/CR-1513 Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report, January-March 1980. ! Disrupted Core SIKKER-II: A Computer Program for LMFBR Disrupted Core KUREG/CR-0453, Rev 1 ; Analysis. Distribution-Free Risk Methodology for Geologic Disposal of Radioactive NUREC/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. Dose Rates Calculations of the Skyshine Gamma-Ray Dese Rates from KUREG/CR-0723 r Independent Spent Fuel Storage Installations (ISFSI) Under , Worst Case Accident Conditions. t F Dosimetry LWR Pressure Vessel Irradiation Surveillance Dosimetry NUREG/CR-0720 Quarterly Progress Report, October-December 1978. Performance Testing of. Personnel Dosimetry Services: t Dosimetry Testing NUREC/CR-1593 Alternatives and Recommendations for a Personnel 1 Dosimet ry Testing Program. Dresden Final Environmental Statement Related to Primary Cooling KUREG-0686 System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Docket No. 50-010.
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$3yword Listin d Report Title Report No.
Drill lioles Monoclinal Structure and Shallow Faulting of the Reelfoot NUREG/CR-1501 Scarp as Estimated from Drill lloles with Variable Spacings. Drop Deposition Film Entrainment and Drop Deposition for Two-Phase Flow. NUREG/CR-1634 Drop-Size Drop-Size Estimates for a Loss-of-Coolant Accident. NUREC/CR-1607 Dynamic Dynamic, Inelastic Buckling Analysis of MARK 1 Torus NUREG/CR-1038 Support Columns. Dynamic Analysis Dynamic Analysis to Establish Normal Shock and Vibration NUREG/CR-1484 of Radioactive Material Shipping Packages - Quarterly Progress Report October 1-December 31, 1979. Dynamic Analysis Piping Benchmark Problems. Dynamic Analysis Uniform NUREG/CR-1677, Vol 1 Support Motion Response Spectrum Method. Dynamic Characteristics - Variability of Dynamic Characteristics of Nuclear Power NUREG/CR-1661 Plant Structures. Dynamics Modeling Tornado Dynamics. NUREG/CR-1585 4 J ar s -
Keyword Listing E Report Title ~ Report No. i Earthquake State-of-the-Art Study Concesaing Near-Field Earthquake NUREG/CR-1340 Ground Motion, t Earthquake Regional Relationships Among Earthquake Magnitude. NUREC/CR-1457 Scales. Earthquake Statistical Analysis of Earthquake Ground Motion NUREG/CR-1641 t Parameters. Earthquake Compilation, Assessment and Expansion of the Strong NUREG/CR-1660 Earthquake Ground Motion Data Base. Earthquake Data Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 3 Accelerograph Stations. Earthquake Data Geotechnical and Strong Motion Earthquake Data from U.S. KUREG/CR-0985', Vol 4 Accelerograph Stations. Earthquake Data Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985 Vol 5 Accelerograph Stations. r Earthquake Detection A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648 Location Capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia. Earthquake Zone Geophysical Investigation of the Anna, Ohio Earthquake NUREG/CR-1649. Zone. East Seasonal Variation of 10-Square Mile Probable Maximum NUREG/CR-1486 3 Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. 53). Eastern U.S. Seismic Hazard Analysis. A Methodology for the Eastern NUREG/CR-1582, Vol 2 United States.
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Eddy Current Statistical Analysis of Steam Generator Inspection Plans NUREG/CR-1282 and Eddy Current Testing. Eddy-Current Eddy-Current inspection for Steam Generator Tubing Program. NUREG/CR-1563 Annual Progress Report for Period Ending December 31, 1979. Effective Changes Environmenta'l Assessment for Effective Changes to 10 CFR NUREG-0685 Part 50 and Appendix E to 10 CFR Part 50; Emergency Planning Requirements for Nuclear Power Plants.
1 Keyword Listing E Report Title Report No. Effects The Ef fects of Temperature, Moisture, Concentration, NUREG-0678-Pressure and Mass Transfer on the Adsorption of Krypton j and Xenon'on Activated Carbon. ' Effects The Effects of Natural Phenomena on the Exxon Nuclear NUREG-0722 Company Mixed Oxide Fabrication Plant at Richland, Washington. Effects Site-Dependent Effects at Strong-Motion Accelerograph EUREG/CR-1639 l Stations, f I Electric Fuel Thermocouple Signal Sensitivity to the Sheath Thickness KUREG/CR-1347 of Thermal Hydraulic Test Facility Indirectly Heated Electric Fuel Pin Simulators. Electricity Demand The ORNL State-Level Electricity Demand Forecasting NUREG/CR-1295 Model. Electricity Generation A Methodology and a Preliminary Data Base for Examining NUREG/CR-1539 the llealth Risks et Electricity Generation from Uranium - and Coal Fuels. 1 l Emergency Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System. Emergency Communications Report to Congress on NRC Emergency Communications. KUREG-0729 ) i Emergency Planning Summary of Public Comments and NRC Staff Analysis Relating KUREG-0684' to Rulemaking on Emergency Planning for Nuclear Power Plants. Emergency Planning Environmental Assessment for Effective Changes to 10 CFR NUREG-0685 Part 50 and Appendix E to 10 CFR Part 50; Emergency Planning Requirements for Nuclear Power Plancs. Emergency Response Review and Assessment of Package Requirements (Yellowcake) KUREG-0535 and Emergency Response to Transportation Accidents. Emergency Response Functional Criteria for Emergency Response Facilities. NUREG-0696 Emergency Response Survey of Current State Radiological Emergency Response KUREC/CR-1620 Capabilities for Transportation Related locidents. Emissions Study of Plutonium oxide l'owder Emissions from Simulated NUREG/CR-1302 Shipping Container leaks. l
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.t Report No.
' Keyword Listing E: Report Title Energetics Experimentsi Prompt Burst Erergetics Experiments: Fresh Uranium NUREG/CR-1396 ,
Carbidv/ Sodium Series. Enhancement Enhancement of the Nuclear Materials Management _and NUREG/CR-1527 Safeguards System. Enriched 00 2 Criticality Experiments , wit h Subcritical Clusters of NUREC/CR-1547 ' 2.35 Wt% and 4.31 Wt% U-235 Enriched U0 Rods in Water. at a Water-to-Fuel. Volume Ratio of .l.6.2 i
')
Enriched Uranium Critical Experiments with Interstitially-Moderated Arrays KUREC/CR-1071 of Low-Enriched Uranium oxide. Topical Report on Reference Critical Experiments. Entrainment Film Entrainment and Drop Deposition for Two-Phase Flow. .NUREG/CR-1634 t Environmental Assessment Environmental Assessment for Effective Changes to 10 CFR KUREG-0685 Part 50 and Appendix E to 10 CFR Part 50; Emergency Planning Requirements for Nuclear Power Plants. Environmental Impact Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 1 Milling Project H-25: Volume 1 - Summery and ' Text. Environmental Impact Final ~ Generic Environmental Impact Statement on Uranium KUREG-0706, Vol 2 Milling Project M-25: Volume II - Appendices A-F. Environmental Impact Final Generic Environmental Impact Statement on Uranium KUREG-0706, Vol 3 Milling Project M-25: Volume Ill - Appendices G-V. Environmental Statement Final Environmental Statement Related to the Operatfor. of KUREG-0134, Add. 2 North Anna Power Station, Unit 7 and 2, Docket No. 50-338 and 50-339. Virginia Electric and Power Company. Environmental Statement Final Environmental Statement Related to Primary Cooling NUREG-0686 System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Docket No. 50-010. Environmental Statement Final Enviroemental Statement Related to Steam Generator- KUREG-0692 Repair at Surry Power Station, Unit No. 1. Virginia Elect ic and Power Cmpany, Docket No. 50 280. Environmental Statement Final Environmental Statement Related to the Dperation KUREG-0702 , of Gas Hills Uranium Project, Docket No. 40-299,' Union Carbide Corporation. L Environmental Statement Final Environmental Statement Related to the Operation of NUREG-0727, Add. the Joseph M. Farley Nuclear Plant, Units I and 2, Docket Nos. 50-348 and 50-364.
Keyword Listing _2 Report Title Report No. Environments Shock Environments for Large Shipping Containers During NUREG/CR-1277 Rail Coupling Operations. EPIC A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 the Motion of Fuel and Coolant Subsequent to Pin Failure in a LMFBR. Equipment Evaluation of Simulator Adequacy for the Radiation NUREG/CR-ll84 Qualification of Safety Related Equipment. Equipment Predicting Life Expectancy and Simulating Age of Complex NUREG/CR-1466 Equipment Using /.ccelerated Aging Techniques. Escape-Type Respirator Studies for the Nuclear Regulatory Commission, NUREG/CR-1586 Evaluation and Performance of Escape-Type Self-Contained Breathing Apparatus, October 1,1978-September 30, 1979. Establish Dynamic Analysis to Establish Normal Shock and Vibration NUREG/CR-1484 of Radioactive Material Shipping Packages - Quarterly l Progress Report October 1-December 31, 1979. Estimate An Algorithm to Estimate Field Concentrations Under KUREG/CR-1474 Nonsteady Meteorological Conditions from Wind Tunnel Experiments. Estimate Monoclinal Structure and Shallow Faulting of the Reelfoot MUREC/CR-1501 Scarp as Estimated from Drill Holes with Variable Spacings. Estimates COPS Model Estimates of LLEA Availability Near Selected NUREG/CR-ll66 Reactor Sites. Estimates Seasonal Vibration of 10-Square Mile Probable Maximum NUREG/CR-1486 Precipitation Estimates, United States East of the 105th Meridian (Hydrometeosological Report No. 53). Estimates Estimates of Uranium Content and Radon Flux for Uranium NUREG/CR-1549 Mine Dumps Based on Borehold Radioactivity Logs. Estimates Drop-Size Estimates for a Loss-of-Coolant Accident. NUREG/CR-1607 Estimators Uncertainty Analysis for a PWR Loss-of-Coolant Accident: NUREC/CR-1364
- 11. Alternative Core Damage Estimators.
Evaluation Investigation and Evaluation of Cracking Incidents in NUREG-0691 Piping in Pressurized Water Reactors.
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l Keyword Listing E ' Report Title Report No. Evaluation Evaluation of Simulator Adequacy for the Radiation NUREG/CR-ll84 Qualification of Safety Related Equipment. I i
~ Evaluation Evaluation of Isotope Migration-Land Burial Water ' Chemistry . NUREG/CR-1289 !
at Commercially. Operated Low Level Radioactive Waste Disposal Sites. Evaluation Evaluation of Isotope Migration-Lead Burial Water NUREG/CRal513 Chemistry at Commercially Operated Low Level Radioactive Waste Disposal Sites. Q urterly Progress Report, January-March 1980. Evaluation Evaluation of Mathematical ML4els for Characterizing NUREG/CR-1581, Vol 1 J Plume Behavior from Cooling Towerat Vol. 1--Dispersion from Single and Multiple Source Natural Draft Cooling ; Towers. Evaluation Evaluation c. Mathematical Models for Characterizing NUREC/CR-1581, Vol'2 Plume Behavior from Cooline Towers: Vol.2--Salt Drift Deposition f rom Natural Draf t Cooling Towers . Evaluation Evaluation of Mathematical Models for Characterizing NUREC/CR-1581, Vol<3 ; Plume Behavior from Cooling Towers: .Vol. 3--Plume Rise from Mechaniral Draft Coo.'.ing Towers. Evaluation Respirator Studies for the Nuclear Regulatory Commission, NUREG/CR-1586 Evaluation and Performance of Escape-Type Self-Contained + Breathing Apparatus, October 1, 1918-September 30, 1979. Evaluation Evaluation Methodology for Fixed-Site Physical Protection NUREG/CR-1590 Systems, i
' Evaluation An Evaluation of Condensation-Induced Water Hammer in NUREG/CR-1606 {'
Preheat Steam Generators. k E Evaluation Scenario Development and Evaluation Related to the Risk NUREC/CR-1608 Assessment of High Level Radioactive Waste Repositories. i Evaluation Evaluation of In-Situ Soil Damping Characteristics. NUREG/CR 1638 Evaluation Method Best Estimate Method vs Evaluation Method: A Comparison NUREG/CR-1489 [ of Two Techniques in Evaluatiri Scismic Analysis and ; Design, f Evaluation Methodology Design Guidance and Evaluation Methodology for Fixed NUREG/CR-Il98, Vol 1 Site Physical Protection Systems, Volume 1. I L Evaluation Methodology Design Guid.ance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 2 Site Physical Protection Systems,' Volume 2.
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. . . - . -. ~. . . - Keyword Listing _E Report Title Report No. Evaluation Program Qualification Testing Evaluation Program Light-Water NUREG/CR-1492-Reactor Safety Research Quarterly Report, July- ti September 1979. Evaluation Report Safety Evaluation Report Related to Operation of Sequoyah NUREG-00ll, Supp. 2 Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328. Tennessee Valley Authority, Supp. No. 2. Evaluation Report Safety Evaluation Report Related to Oparation of Sequoyah NUREG-00ll, Supp. 3. Nuclear Plant, Units I and 2, Docket Nos. 50-327/328. Tennessee Valley Authority, Supp. 3. Evalur-~ Report Safety Evaluation Report Related to the Operation of NUREG-0053, Supp. 11 North Anna Power Station, Unit 2, Virginia Electric and-Power Company, Docket No. 50-339. Supplement No. 11. Evaluation Report Safety Evaluation neport Related to the Operation of ~ NUREG-0053, Supp.12 North Anna Power Station, Unit 2, Docket' No. 50-339. Supplement No. 12. Evaluation Report Safety Evaluation Report Related to the Operation of NUREG-0117, Supp. 4 Joseph H. Farley Nuclear Plant, Unit 2. Docket No. 50-364, Alabama Fower Company. Supplement 4 to NUREG-75/034. Evaluation Report Safety Evaluation Report Related to Operation of Diablo NUREG-0675, Supp 10 Canyon Nuclear Power Station, Units I and 2, Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Supplement No. 10. Event Lisc Safeguards Summary Event List (SSEL). KUREG-0525, Rev 2 Examining A Nethodology and a Preliminary Data Base for Examining NUREG/CR-1539 the licalth Risks of Electricity Generation from Uraatum and Co.vl Fuels. Existing Review and Integration of Existing Literature Concerning NUREC/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. Existing. Characterization of Existing Surface Conditions at NUREG/CR-1683 Sheffield Low-Level Waste Disposal Facility. Expansion Compilation, Assessment and Expansion of the Strong NUREG/CR-1660 Earthquake Ground Motion Data Base. Expertment Extended Analysis of Data from the 1/5-Scale MARK l' NUREG/CR-0761 , Boiling Water Reactor Pressure Suppression Experiment. e Experiment L6-5 Experiment Data Report for LOFT Anticipated Transient NUhEG/CR-1520 Experiment L6-5. s l l i l l i
.Seport Title Report No.
gcyword Listing,E Experimental Anicials Acute Effects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. L'F6 /UO 2 2 F Studies in Experimental Animals. _j Experiments Critical Experiments with Interstitia'ly-Moderated 'NUREG/CR-1071 Arrays of Low-En iched Uranium Oxide. Topical Report '; on Reference Critical Experiments, r Experiments Diffusion Near Buildings as Determined from Atmospheric NUREG/CR-1394 , Tracer Experiments, 3 Experiments Criticality Experiments with Suberitical Clusters of NUREG/CR-1547 2.35 Wt% and 4.31 WL% U-235 Enriched UO 2 Rods in Water at a Water-to-Fuel Volume Ratio of 1.6 , t Seismic llazard Analysis, Solicitation of Expert NUREG/CH-1582, Vol 3 Expert Opinion Opinion. Exposure Acute Effects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF6/UO 2 2 F Stadies in Experimental Animals. Exposure Growth and Ifistological Effects to Protothaca'Staminea NUREG/CR-1298 (Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water. Extended Analysis Extended Analysis of Data from the 1/5-Scale MARK I NUREG/CR-0761 , Boiling Water Reactor Pressure Suppression Experiment. The Effects of Natural Phenomena on the Exxon Nuclear NUREG-0722 Exxon Nuclear Co. Company Mixed Oxide Fabrication Plant at Richland, Washington. t 9 r
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Keyword Lieting F Report Title Report No. Fabrication Plant the Effects of Natural Phenomena on the Exxon Nucleer NUREG-0722 Company Mixed Oxide Fabrication Plant at Richland, Washington. Facilities Functional Criteria for Emergency Response facilities. NUREG-0696 l Failure An Assessment of LWR Fuel-Failure Propagation Potential: NUREG/CR-1471 Literature Survey. Failure A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 the Motion of Fuel and Coolant Subsequent to Pin Failure in a LMFBR. Farley, Joseph M. Safety Evaluation Report Related to the Operation of . NUREG-Oll7, Supp. 4 Joseph M. Farley Nuclear Plant, Unit 2. Docket No. 50-364, Alabama Power company. Supplement 4 to NUREG-75/034. b Farley, Joseph M. Final Environmental Statement Related to the Operation of NUREG-0727, Add. Lhe Joseph M. Farley Nuclear Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364. Fatigue Behavior High Cycle Fatigue Behavior of Incoloy 800H in a inDtEG/CR-1356 Simulated High-Temperature Gas-Cooled Reactor Helium Environment'. , Faulting Monoclinal Structure and Shallow Faulting of the Reelfoot NUREG/CR-1501 Scarp as Estimated from Drill Holes with Variab1r. Spacings. Faults A Characterization of Faults in the Appalachian Foldbelt. NUREG/CR-1621 i Field Concentration An Algorithm to Estimate Field Concentrations Under NUREG/CR-1474 Nonsteady Meteorological Conditions from Wind Tunnel Experiments. Film Entrainment Film Entrainment and Drop Deposition for Two-Phase Flow. NUREG/CR-1634 Financing Financing Strategies for Nuclear Power Plant NUREG/CR-1481 Decommissioning. Fire Tests Development and Verification of Fire Tests for Cable NUREG/CR-1552 Systems and System Components. Fires The NACOM Code for Analysif of Postulated Sodium Spray RUREG/CR-1405 Fires in LMFBRs. l l 1
Keyword Listing 7 Report Title Report No. Fiscal Year 1982 Comments on the NRC Safety Research Program Budget for NUREG-0699 Fiscal Year 1982. Fish Acute Toxicity and Bioaccumulation of Chloroform to Four KUREC/CR-0893 Species of Freshwater Fish. Fixed Site Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 1 Site Physical Protection Systems, Volume 1. i Fixed Site Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 2 Site Physical Protection Systems, Volume 2. ., Fixed Site Fixed Site Neutralization Model Programmer's Model. NUREG/CR-1308, Vol 2 Fixed Sites Acceptance Criteria for the Physical Protection Upgrade NUPEG-0721 Rule Requirements for Fixed Sites. -Fixed-Site Security Communication Systems for Nuclear Fixed-Site NURES/CR-0508 Facilities.
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Fixed-Site Evaluation Methodology for Fixed-Site Physical Protection NUREG/CR-1590 Systems. Flow Film Entrainment and Drop Deposition for Two-Phase Flow. KUREG/CR-1634 Flow Measurements Two-Phare Flow Measurements with Advanced Instrumented NUREG/CR-1529 Spool Pieces. Flow Topography Flow Topography Instrumentation and Analysis System. NUREG/CR-1333 Flows On the Motions of Particles in Turbulent Flows. NUREG/CR-1554 Foldbelt A Characterization of Faults in the Appalachian Foldbelt. NUREG/CR-1621 Forecasting Model 'The ORNL State-Level Electricity Demand Forecasting KUREG/CR-1295 Model. _ ~
Keyword'Licting_F Report Title Report No. Fracture Piping Inelastic Fracture Mechanics Analysis. KUREG/CR-1119 i Frequently Asked Answers to Frequently Asked Questions about Cleanup NUREG-0732 Activities at Three Mile Island, Unit 2. Freshwater Fish Acute Toxicity and Bioaccumulation of Chloroform to Four NUREG/CR-0893 Species of Freshwater Fish, Fuel A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 the Motion of Fuel and Coolant Subsequent f.o Pin Failure ' in a LNFBR. Fuel Gap Conductance Test Series Fuel Characterization Data NUREG/CR-1537 Report. Fuel LWR Fuel Rad Post-Subcooled Blowdown Scoping Analysis. EUREG/CR-1568 Fuel Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592 to Sodium Boiling in Fuel Subassemblies during Pump Coast-Down of an LNFBR. I Fuel Centerline Qualification Test Results on 1550*C and 2200'C 1/16-Inch NUREG/CR-0961 0.D. Fuel Centerline Thermocouples for the LOFT Program. Fuel Cycles Technical Safeguards Issues for Alternative Fuel Cycles. NUREC/CR-1048 Fuel Designs A Comparative Analysis of LWR Fuel Designs. NUREG-0559 Fuel Pin Thermocouple Signal Sensitivity to the Sheath Thickness NUREG/CR-1347 of Thermal-Hydraulic Test Facility Indirectly Heated Electric Fuel Pin Simulators, r Fuel Plants Solubility Classification of Airborne Uranium Products EUREG/CR-1428 from LWR-Fuel Plants. Fuel Systems Assessment of Current Onsite Inspection Techniques for NUREG/CR-1380, Vol 1, ES Light Water Reactor Fuel Systems - Volume 1 - Executive Summary. Fuel-Failure An Assessment of LWR Fuel-Failure Propagation Potential: NUREG/CR-1471 Literature Survey.
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%eyword LA', tins F Report Title I Report No. 1 Fuels A Methodology and a Preliminary Data Base for Examining NUREC/CR-1539 the Health Risks of Electricity Generation from Uranium and Coal F els.
Functional Criteria Functional Criteria for Emergency Response Facilities. NUREG-0696 a 0 ? J 1
Keyeard Listing G Report Title Report No. Gamma Ray Calculations of the Skyshine Gamma-Ray Dose Rates from NUREG/CR-0723
' Independent Spent Fuel Storage Installations (ISFST) Under Worst Case Accident Conditions.
Gamma-Ray Advaaced Mobile Multi-Processor Gamma-Ray Acquisition NUREG/CR-1668 and Analysis System. Gap Cap Conductance Test Series Fuel Characterization Data NUREG/CR-1537 Report. Gas Hills Final Environmental Statement Related to the Operation- NUREG-0702 of Gas Hills Uranium Project, Docket No. 40-299, Union Carbide Corporation. Gas-Cooled High Cycle Fatigue Behavior of Incoloy 800H in a NUREG/CR-1356 Simulated High-Temperature Gas-Cooled Reactor Helium Environment. Gas-Cooled High-Temperature Gas-Cooled Reactor Safety Studies for the NUREG/CR-1521 Division of Reactor Safety Research Quarterly Progress Report , Janua ry 1-March 31,1980. Generator Statistical Analysis of Steam Generator Inspection Plans NUREC/CR-1282 and Eddy Current Testing. Generator Eddy-Current Inspection for Steam Generator Tubing Program. NUREG/CR-1563 Annual Progress Report for Period Ending December 31,
- 1979.
Generator An Evaluation of Condensation-Induced Water Hammer in NUREG/CR-1606 Preheat Steam Generators. Generator Repair Final Environmental Statement Related to Steam Generator NURG-0692 Repair at Surry Power Station, Unit No. 1. Virginia Electric and Power Cmpany, Docket No. 50-280. Generic MARK 11 Containment Lead Plant Program Load Evaluation and NUREG-0487, Supp 1 Acceptance Criteria - Generic Technical Activities A8 and A39. Generic Control of Heavy Loads at Nuclear Power Plants NUREG-0612 Resolution of Generic Technical Activity A-36. Generic MARK I Containment Long Term Program Safety Evaluation NUREG-0661 Report, Resolution of Generic Technical Activity A-7. February 1977 to December 1979. Generic Fin 1 Generic Environmental Imp--t it tement on Uranium NUREG-0706, Vol 1 Milling Project MT1 volume I - Summary and Text. 4 b
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Keyword Listing G Report Title Report No. Generic Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 2 Milling Project M-25: Volume II - Appendices A-F. Generic Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 3 Milling Project M-25: Volume III - Appendices C-V. Geologic Risk Methodology for Geologic Disposal of Radioactive KUREG/CR-1377 Waste: Transport Model Sensitivity Analysis. Geologic Disposal Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1261 , Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. Geophya: cal Investigation Geophysical' Investigation of the Anna, Ohio Earthquake NUREG/CR-1649 Zone. i Geotechnical Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 3-Accelerograph Stations. Geotechnical Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR 0985, Vol 4 Accelerograph Stations. Geotechnical Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 5 Accelerograph Stations. Geotechnical Data Geotechnical Data from Accelerograph Stations NUREG/CR-1643 Investigated during the Period 1975-1979. Summa ry Report. Great Lakes Seismicity and Tectonic Relationships for Upper Great NUREG/CR-1569 Lakes Precambrian Shield Province. Grotwh Growth and Histological Effects to Protothaca Staminea NUREG/CR-1298 (Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water. Ground Motion State-of-the-Art Study Concerning Near-Field Earthquake NUREG/CR-1340 Ground Motion. Ground Motion Statistical Analysis of Earthquake Ground Motion NUREG/CR-1641 Parameters. Ground Motion Compilation, Assessment and Expansion of the Strong NUFEG/CR-1660 Earthquake Ground Motion Data Base.
Keyword List ing G Report Title Report No. Guide A User's Guide to EPIs', a Computer Program to Calculate NUREG/CR-1504 the Motion of Fuel and Coolant Subsequent to Pin Failure in a 1.MFBk. Guidelines Technical Report on Material Selection and Processing NUREG-0313, Rev. 1 Guidelines for BWR Coolant Pressure Boundary Piping. d a b 1 i e
Keyword Listing H Report Ti .,l e Report No. Hazard Analysis Seismic Hazard Analysis. A Methodology for the Eastern NUREG/CR-1582, Vol 2 United States. Hazard Analysis Seismic Hazard Analysis. Solicitation of Expert NUREC/CR-1582, Vol 3 f opinica. Health Risks A Methodilogy and a Preliminary Data Base for Examining NUREG/CR-1539 the Healta Risks of Electricity Generation from Uranium and Coal Fuels. Health Status Health States and Body Radioactivity of Former Thorium NUREC/CR-1420 Workers. Heat Fluxes User's Manual tor USINT: A Program for Calculating NUREG/CR-1375 Heat and Mass Transfer in Concrete Subjected to High Heat Fluxes. Heat Removal Heat. Removal Characteristics of Volume Heated Boiling NUREG/CR-1357 Pools with Inclined Boundaries. Heat Transfer Transient Analysis of Coolant Flow and Heat Transfer NUREG/CR-1404 in LMFBR Piping Systems. Heavy Loads Control of Heavy Loads at Nuclear Power Plants NUREG-0612 Resolution of Generic Technical Activity A-36. Heavy-Section Heavy-Section Steel Technology Program Quarterly NUREG/CR-1477 Progress Report for January-March 1980. Helium Environment High Cycle Fatigue Behavior of Incoloy 800H in a NUREG/CR-1356 Simulated High-Temperature Gas-Cooled Reactor Helium Environment. Hexafluoride Acute Effects of inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF F Studies in Experimental 6 2 2 Animals. High Cycle High Cycle Fatigue Behavior of Incoloy 800H in a NUREG/CR-1356 Simulated High-Temperature Gas-Cooled Reactor Helium Environment. High Level Scenario Development and Evaluation Related to the Risk NUREG/CR-1608 Assessment of High Level Radioactive Waste Repositories. High Temperature An Ultrasonic Thermometry System for Measuring Very High fmREG/CR-1488 Temperatures in Reactor Safety Experiments.
Keyword Listing H Report Title Report No. High-Temperature High Cycle Fatigue Behavior of Incoloy 800H rn a NUREG/CR-1256 Simulated High-Temperature Gas-Cooled Reactor Helium Environment. High-Temperature High-Temperature Gas-Cooled Reactor Safety Studies for the NUREG/CR-1521 Division of Reactor Safety Research Quarterly Progress Report , Janua ry 1-Karch 31,1980. Histological Effects Growth and Histological Effects to Protothaca Staminea NUREG/CR-1298 (Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water. Holes Honoclinal Structure and Shallow Faulting of the Reelfoot ! NUREG/CR-1501 Scarp as Estimated from Drill Holes with Variable Spacings. Hydraulic Summary of Thermal Hydraulic Calculations for a NUREG/CR-1480 Pressurized Water Reactor. Hydrodynamics Hydrodynamics of a Vapor Jet in Subcooled Liquid. NUREG/CR-1632 Hydrodynamics Program Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-1557 Task 4 - Quarterly Progress Report, April 1-June 30, 1979. Hydrodynamica Program Steam-Water Mixing and System Hydrodynamics Program - NUREGicR-1625 Task 4 - Quarterly Progress Report, July 1-September 30, 1979. Hydrodynamics Program Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-1657 Ta6k 4. Quarterly Progress Report October-December 1979. Hydrogen Alloys Instrumented Impact Properties of Zircaloy-Oxygen and NUP.EG/CR-1408 Zircaloy-Hydrogen Alloys. Hydrogen Mixing Hydrogen-Mixing in a closed Containment Compartment Based NUREC/CR-15?5 on a One-Dimensional Model with Convective Effects. Hydrometeorological Rpt 53 seasonal Vibration of 10-Square Mile Probable Maximum NUREG/CR-1486 Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. 53). i
I I Keyword Listing I Report Title Repo,rt No. . Identification 16entification and Assessment of the Social Impacts of NUREG/CM-c746 ; Transportation of Radioactive Materials in Urban J Environments. 1 l Illinois A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648 Location Capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia, , impact Properties Instrumented Impact Properties of Zircoloy-Oxygen and NUREG/CR-1408 Zirealoy-Hydrogen Alloys. Impact Statement Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 1 Milling Project M-25: Volure I - Summary and Text. Impact Statement Final Gener(c Environmental Impact Statement on Urwniam NUREG-0706, Vol 2 Milling Project M-25: Volume II - Appecdices A-F. Impact Statement Final Generic tovironmental Impaci Stateoent on Uranium NUREG-0706, V i 3 it lling Project M-25: Volume lif - Appendices G-V. Irplementation Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System. Improve Using Advanced Process Monitoring to Improve Material NUREG/CR-1676, Vol 1 Control. In-Plant In-Plant Source Ters Measurements at Turkey Point NUREG/CR-1629 Station - Units 3 and 4. In-Situ Evaluation of In-Situ Soil Damping Characteristics. NUREG/CR-1638 Incident Response Report to Congress: NRC Incident Response Plan. NUREG-0728 Incidents Survey of Current State Radiological Emergency Response NUREG/CR-1620 Capabilities for Transportation Related Incidents. Inclined Boundaries Heat Removal Characteristics of Volume Heated Boiling NUREG/CR-1357 Pools with Inclined Boundaries. Incoloy 800H High Cycle Fatigue Behavior of Incoloy 800H in a NUREG/CR-1356 Simulated High-Temperature Gas-Cooled Reactor Helium Environment.
Keyword Listing I Report Title Report No. Independent Calculations of the Skyshine Gamma Ray Dose Rates from NUREG/CR-0723 Independent Spent fuel Storage Installations (ISFSI) Under Worst Case Accident Conditions. ,
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Indian Point Task Force Report on Interim Operation of Indian Point. NUREG-0715 l i Indian Point Report of the Zion / Indian Point Study: Volume 1. NUREG/CR-1410 Indian Point Report of the Zion / Indian Point Study. NUREG/CR-1411, Vol 1 4 Indian Point Report of the Zion / Indian Point Study. NUREG/CR-1411, Vol 2 Indiana A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648 , Location Capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia. Inducing Rank Risk Methodology for Geologic Disposal of Radioactive NUREC/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. Industry Report on Nuclear Industry Quality Assurance Procedures NUREG-0653 for Safety Analysis Computer Code Development and Us . Inelastic Dynamic, Inelastic Buckling Analysis of MARK I Torus NUREG/CR-1038 Support Columns. Inelastic Piping Inelastic Fracture Mechanics Analysis. NUREG/CR-1119 inhalation Exposure Acute Effecta of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UT6/UO 2 2F Studies in Experimental Animals. Inlet Plenum Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592 to Sodium Boiling in Fuel Subassemblies during Pump Coast-Down of an LMFBR. Input Variables Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1262 Waste A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies, losiders Potential Threat to Licensed Nuclear Activities from NUREG-0703 Insiders (Insider Study).
Keyword Listin d Report Title Report No. Inspection Statistical Analysis of Steam Generator Inspection Plans NUREG/CR-1282 and Eddy Current Testing. Inspection Assessment of Current Onsite Inspection Techniques for NUREC/CR-1380, Vol 1, ES-Light Water Reactor Fuel Systems - Volume 1 - Executive Summary. Inspection Eddy-Current Inspection for Steam Generator Tubing Program. NUREG/CR-1563 Annual Progress Report for Period Ending December 31, 1979. Inspection Methods Inspection Methods for Physical Ptotection Project: NUREG/CR-1610, Vol 1, No. 1 Quarterly Report, March-May 1980. Inst 51ations Calculations of the Skyshine Gamma-Ray Dose Rates from NUREG/CR-0723 Independent Spent Fuel Storage Installations (ISFST) Under Worst Case Accident Conditions. Instrumentation Flow Topography Instrumentation and Analysis System. NUREG/CR-1333 lustrumented Instrumented Impact Properties of Zircaloy-Oxygen and NUREG/CR-1408 Zircaloy-Hydrogen Alloys. Instrumented Two-Phase Flow Heasurements with Advanced Instrumented NUREG/CR-1529 Spool Pieces. Integration Review and Integration of Existing Literature Concerning NUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. Integrity Structural Integrity of Water Reactor Pressure Boundary NUREG/CR-1472 Components. Quarterly Progress Report, January-March 1980. , Interim Operation Task Force Report on Interim Operation of Indian Point. NUREG-0715 Interstitial Critical Experiments with luterstitially-Moderated NUREG/CR-1071 Arrays of Low-Enriched Uranium Oxide. Topical Report on Reference Critical Experiments. Investigated Geotechnical Data from Accelerograph Stations NUREG/CR-1643 Investigated during the Period 1975-1979. Summary Report. Investigatice. Investigation and Evaluation of Cracking Incidents in NUREG-0691 Piping in Pressurized Water Reactors.
Keyword Listing I Report Title Report No. Iodine-131 Measurement of XE-133, C-14 and Tritium in Air and I-131 NUREG/CR-1195 in Vegetation and Milk Around the Quad Cities Nuclear Power Station. Irradiation LWR Pressure Vessel Irradiation Surveillance Dosimetry NUREG/CR-0720-Quarterly Progress Report, October-December 1978. ISFSI Calculations of the eshine Gamma-Ray Dose Rates from KUREG/CR-0723 Independent Spent FL 4 Storage Installations (ISFSI) Under Worst Case Accident Conditions. Isotope Evaluation of Isotope Migration-Land Burial Water Chemistry NUREG/CR-1289 at Commercially Operated Low Level Radioactive Waste Disposal Sites. Isotope Evaluation of Isotope Migration-Land Burial Water NUREG/CR-1553 Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report, Janua ry-March 1980.
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Keyword Listing J Report Title Report No. Joseph M. Faricy Safety Evaluatton Report Related to the Operation of NUREG-Oll7, Supp. 4 Joseph M. Farley Nuclear Plant, Unit 2. Docket No. 50-364, Alabama Power Company. Supplement 4 to NUREG-75/034. l Joseph M. Farley - Final Environmental Statement Related to the Operation of NUREG-0727,- Add, the Joseph M. Farley Nuclear Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364. 4
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1 Keyword Listing K Report Title Report No. K-FIX Code PRESBC: . Pressure Boundary Conditions for the K-FIX Code. NUREG/CR-1536 Kentucky A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648 Location capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia. Kentucky Transportation of Radioactive Material in Kentucky. NUREG/CR-1671 Krypton The Effects of Temperature, Moisture, Concentration, NUREG-0678 Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon. L
Keyword Listing L Report Title' Report No.
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Land Burial Evaluation of Isotope Migration-Land Burial Water Chemistry NUREG/CR-1289 at' Commercially Operated Low Level Radioactive Waste Disposal Sites. Land Burial Vegetational Cover in Monitoring and Stabilization of~ .NUREG/CR-1358 ) Shallow Land Burial Sites. [ i Land Burial Evaluation of Isotope Migration-Land Burial Water NUREG/CR-1513 i Chemistry at Commercially operated Low-Level Radioactive. Waste Disposal Sites. t lateral Loads Lateral Loads on Vent Pipe in Steam Chugging. NUREG/CR-1631 LCP Load Combination Program. Progress Report No. 5. NUREG/CR-1624 Lead PBF/ LOFT Lead Rod Test Series Test Results Report. NUREG/CR-1538 , Lead Plant MARK II Containment Lead Plant Program Load Evaluation and NUREG-0487, Supp 1 Acceptance Criteria - Generic Technical Activities A8 and A39. Leaks Study of Plutonium Oxide Powder Emissions from Simulated NUREG/CR-1302 - Shipping Container Leaks LER Data Sununaries of Licensee Event Reports of Primary _ NUREG/CR 1730 j Containment Penetrations at U.S. Commercial Nuclear Power Plants f rom January 1,1972 to December 31, 1978. Licensability Licensability of CANDU-Type Reactors in the United States. NUREG/CR-1113 A Preliminary Assessment of the R and D Requirements. License DPR 77 Technical Specifications, Sequoyah Nuclear Plant, Unit No. NUREG-0658, Rev 1. 1, Docket No. 50-327 Appendix "A" to License 'No. DPR-77. , License NPF-7 North Anna Power Station Unit 2 Technical Specifications NUREG-0564, Rev 1 Appendix "A" to License No. NPF-7. Licensed Potential Threat to k.icensed Nuclear Activities f rom NUREG-0703 Insiders (Insider Study). I Licensee Events Data Summaries of Licensee Event Reports of Primary NUREG/CR-1730 Containment Penetrations at U.S. Commercial Nuclear Power - Plants from January 1, 1972 to December 31, 1978. I
4 Keyword Listing _L Report Title Report No. Licensing Evaluation SCALE: A Modular Core System for Performing Standardized NUREG/CR-0200 Computer Analyses .for Licensing Evaluation, SCALE System Criticality Safety _ Analysis Modul s CSASI and CSAS2. Life Expectancy Predicting Life Expectancy and Simulating Age of Complex NUREG/CR-1466 > Equipment Using Accelerated Aging Techniques. Line Dynamics Steam Line Dynamics. KUREG/CR-1438 Liquid Hydrodynamics of a Vapor Jet in Subcooled Liquid. NUREG/CR-1632 Literature Review and Integration of Existing Literature Concerning NUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. Literature Survey An Assessment of LWR Fuel-Failure Propagation Potential NUREG/CR-1471 Literature Survey. Littleneck Clam Growth and Histological Effects to Protothaca Staminea- NUREC/CR-1298 (Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water. LLEA COPS Model Estimates of LLEA Availability Near Selected NUREG/CR-1166 Reactor Sites, LMFBR SIMMER-II: A Computer Program for LHFBR Disrupted Core NUREG/CR-0453, Rev 1 + Analysis. LMFBR CONAN: An LMFBR Containment Response Computer Code. NUREC/CR-1355 LMFBR Transient Analysis of Coolant Flow and Heat Transfer NUREG/CR-1404 , in LMFBR Piping Systems.
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LMFBR The NACOM Code for Analysif of Postulated Sodium Spray NUREG/CR-1405 Fires in LMFBRs. I 1 i LMFBR A User's Guide to EPIC, a Computer Program to Calculate NUREC/CR-1504 l the Motion of fuel and Coolant Subsequent to Pin Failure in a LMFBR, l LMFBR Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592 to Sodium Boiling in Fuel Subassemblics during Pump Coast-Down of an LMFBR.
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Keyword Listing L Report Title - Report No. Load . Load Combination Program. Progress Report No. 5. NUREG/CR-1624 [ Load Evaluation MARK 11 Containment Lead Plant Program Load Evaluation and NUREG-0487, Supp 1' - Acceptance Criteria - Generic Technical Activities A8 and A39. l loads Control of Heavy Loads at Nuclear Power Plants NUREG-0612' Resolution of Generic Technical Activity A-36. J. Loads Lateral Loads on Vent Pipe in Steam Chugging. NUREC/,CR-1631 i LOCA Uncertainty Analysis for a PWR Loss-of-Coolant Accident: , NUREG/CR-1364 II. Alternative Core Damage Estimators. LOCA Drop-Size Estimates for a Loss-of-Coolant Accident. NUREG/CR-1607 F Location Capability A Probabilistic Evaluation of Earthquake Detection and . NUREG/CR-1648 Location Capability for 1111nois, Indiana, Kentucky, Ohio, and West Virginia. j l LOFT Experiment Data Report for LO.FT Anticipated Transient NUREG/CR-1520 l f Experiment L6-5. LOFT PBF/ LOFT Lead Rod TestLSeries Test Results Report, NUREC/CR-1538 Comparison of CONTEMPT-LT Containment Code Calculations' LOFT NUREG/CR-1564 ~ with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests.
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LOFT Program Qualification Test Results on 1550'C and 2200*C 1/16-Inch NUREG/CR-0961 0.D. Fuel Centerline Thermocouples for the LOFT Program. ! Long Term Program MARK I Containment Long Term Program Safety Evaluation NUREG-0661 i. Report, Resolution of Generic Technical Activity A-7. ' February 1977 to December 1979. Long-Te rm Crowth and Histological Effects to Protothaca Staminea NUREG/CR-1298 j
.(Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water.
Los s-o f-Coolant Uncertainty Analysis for a PWR Loss-of-Coolant Accident: NUREG/CR-1364 II. Alternative Core Damage Estimators. 1 l.
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KeywordListind y R ort Title y ort No. R
' Loss-of-Coolant Drop-Size Estimates for a Loss-of-Coolant Accident, NUREG/CR-1607 Low-Enriched . Critical Experiments with Interstitially-Moderated NUREG/CR-1071 Arrays of Low-Enriched Uranium oxide. Topical Report on Reference Critical Experiments, t
Low-Level Evaluation of Isotope Migration-Land. Burial Water Chemistry NUREG/CR-1289 at Commercially Operated Low-Level Radioactive Waste Disposal Sites. < Low-Level Evaluation of Isotope Migration-Land Buri61 Water NUREG/CR-1513 ' Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report, January-March 1980. Low-1.evel Characterization of Existing L;f ace Conditions at NUREG/CR-1683 theffield Low-Level Waste Disposal Facility. LWR A Comparative Analysis of LWR Fuel Designs. NUREG-0559 LWR Pipe Cracking Experience in Light-Water Reactors, NUREG-0679 1967 through 1979. LWR LWR Pressure Vessel Irradiation Surveillance Dosimetry NUREG/CR-0720 Quarterly Progress Reports. October-December 1978. LWR Assessment of Current Onsite Inspection Techniques for NUREG/CR-1380', Vol'1, ES Light Water Reactor Fuel Systems - Volume 1 - Executive Summary. LWR Solubility Classification of Airborne Uraniun Products NUREG/CR-1428 from LVR-Fuel Plants. , LWR An Assessment of LWR Fuel-Failure Propagation Potential: NUREG/CR-1471 Literature Survey. 1 LWR Qualification Testing Evaluation Program I.ght-Water NUREG/CR-1492 Reactor Safety Research Quarterly Report July-September 1979. > LWR Light Water Reactor Safety Researth Trogram Quarterly NUREG/CR-1509 Report, January-March 1980. LWR LWR Fuel Rod Post Subcooled Blowdown Scoping Analysis. NUREG/CR-1568 ,
N Report No. Keyword Listing M Report Title MagnitLde Scales Regional Relationships Among Earthquake Magnitude NUREG/CR-1457 Scales. Management Enhancement of the Fuclear Materials danagement and KUREG/CR-1527 Safeguards System. Management Safeguards User's Manual fut duelear Materials Management NUREG/CR-1528 [ and Safeguards System. I Utility Management and Technical Resources. NUREG/CR-1656 Management Fixed Site Neutralization Model Programmer's Model. KUREG/CR-1308, Vol 2 Manual Manual User's Manual for USINT: A Program for Calculating NUREG/CR-1375 Heat and Mass Transfer in Concrete Subject to High l Heat Fluxes. Safeguards User's Manual for Nuclear Materials Management NUREG/CR-1528 Manual and Safeguards System. , MARK I Containment Long Term Program Safety Evaluation MUREG-0661 MARK I Report, Resolution of Generic Technical Activity A-7. February 1977 to December 1979. MARK I Extended Analysis of Data from the 1/5-Scale MARK I !!JREG/CR-0761 Boiling Water Reactor Pressure Suppression Experiment. MARK I Dynamic, Inelastic Buckling Analysis of MARK I Torus NUREG/CR-1038 Support Columns. 1 HARK 11 Containment Lead Plant Program Load Evaluation and NUREG-0487, Supp 1 MARK II Acceptance Criteria - Generic Technical Activities A8 and A39. Ma rviken Comparison of C0KTEMPT-LT Containment Code Calculations NUREG/CR-1564 with Marviken, LOFT, and Battelle-Frankfurt Blowdown Tests. The Effects of Temperature, Moisture, Concentration, NUREG-0678 Mass Transfer Pressure and Mass Transfer on the Adsorption of Krypton i and Xenon on Activated Carbon. i Mass Transfer User's Manual for USINT: A Program for Calculating NUREG/CR-1375 Heat and Mass Transfer in Concrete Subjected to High Heat Fluxes. t
l l Keyword Listin g Report Title R_eport e No. Materici Accounting Material Accounting as Required by the United States NUREG/CR-1192 Nuc lear Regulatory Commissiont Capabilities and ; Vulnerabilities. . Material Accounting The Use of Procesa Monitoring Data for Nuclear Material NUREG/CR-1670, Vol 1 Accounting: Voltune 1. Sunniary Report. i Material Control Safeguards Material Control and Accounting Program: NUREG/CR-1485, Vol 1, No. 1 I Quarterly Report, October-December 1979. Material Control Using Advanced Process honitoring to improve Material NUREG/CR-16/6, Vol 1 Control. j Material Selection Technical Report on Material Selection and Proccasing NUREG-0313, Rev. 1 Guidelines for DWR Coolant Pressure Boundary Piping. Materials Management Enhancement of the Nuclear Materials Management and NUREG/CR-1527 Safeguards System Materials Management Safeguards User's Manual for Nuclear Materials Management NUREG/CR-1528 and Safeguards Syates. Mathematical Model Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 3 Plume Behavior from Cooling Towers: Vol. 3--Plume Rise f rom Mechanical Draft Cooling Towers. Mathematical Models Evaluation of Mathematical Models for Characterizing NUKEG/CR-1581, Vol 1 Plume Behavior from Cooling Towers: Vol. 1--Dispersion from Single and Multiple Source Natural Draft Cooling Towers. Mathematical Models Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 2 Plume Behavior from Cooling Towers: Vol.2--Salt Drift L moition from Natural Draft Cooling Towers. Measurement Heasurement of Radon Diffusion from Uranium Hill Tailing NUREG/CR-1109 Piles. Measurement Measurement of XE-133, C-14 and Tritium in Air and 1-131 NUREG/CR-Il95 in Vegetation and Milk Around the Quad Cities Nuclear Power Station. I Measurements Wind-Tunnel Measurements of Dispersion and Turbulence. NUREG/CR-1475 in the Wakes of Nuclear Resetor Plants. Measurements' Two-Phase Flow Measurements with Advanced Instrumented NUREG/CR-1529 Spool Pieces.
Seport Titta . Report No. Keyword Listing M Measurements Critical Experiments, Measurements, and Analyses to. NUREG/CR-1601 Establish a Crack Arrest Methodology for Nuclear Pressure 1
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Vessel Steels. I i Measurements in-Plant Source Term Measurements at Turkey _ Point KUREG/CR-1629 Station - Units 3 and 4. Measuring An Ultrasonic Thermometry System for Measuring Very High NUREG/CR-1488 Temperatures in Reactor Safety Experiments. Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 3 Mechanical Draft Plume Behavior from Cooling Towers: Vol. 3--Plume Rise from Mechanical Draft Cooling Towers. Mechanics Piping Inelastic Fracture Mechanics Analysis. NUREG/CR-1119 Membrane Sbear Strength and Stiffness of Tensioned Reinforced Concrete NUREG/CR-1602 Panels Subjected to Membrane Shear, Two-Way Reinforcing. Meridian Seasonal Variation of 10-Square Mile Probable Maxieum NUREG/CR-1486 Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. 53). l Meteorological An Algorithm to Estimate Field Concentrations Under NUREG/CR-1474 Nonsteady Meteorological Conditions from Wind Tunnel Experiments. Design Guidance and Evaluation Methodology for Fixed NUREC/CR-1198, Vol 1 I Methodology Site Physical Protection Systems, Volume 1. Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 2 Methodoiogy Site Physical Protection Systems, Volume 2. Methodology Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. ! Methodology A Methodology and a Preliminary Data Base for Examining KUREG/CR-1539 the Health Risks of Electricity Generation from Uranium and Coal Fuels. Seismic Hazard Analysis. A Methodology for the Eastern NUREG/CR-1582, Vol 2 Methodology United States. Methodology Evaluation Methodology for Fixed-Site Physical Protection NUREG/CR-1590 Systems..
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Key _ word Listing M Report Title Report No. Microcomputer The N!yT-COMP 9 Microcomputer. NUREC/CR-1548 Migration Evaluation of Isotope Migration-Land Burial Water Chemistry NUREG/CR-1289 at Commercially Operated Low Level Radioactive Waste Disposal Sites. Migration Evaluation of Isotope Migration-Land Burial Water NUREG/CR-1513 Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites, Quarterly Progress Report, January-March 1980. Milk Measuremen of Xe-133, C-14 and Tritium in Air and I-131 NUREG/CR-il95 in Vegetation and Milk Around the Quad Cities Nuclear Power Station. Mill Measurement of Radon Diffusion from Uranium Mill Tailing NUREG/CR-1109 Piles. Mine 9ulps Estimates of Uranium Content and Radon Flux for Uranium NUREG/CR-1549 Mine Dumps Based on Borehold Radioactivity Logs. Mixed Oxide The Ef fects of Natural Phenomena on the Exxon Nuclear NUREG-0722 Company Mixed Oxide Fabrication Plant at Richland, Washington, Mixing Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-1625 Task 4 - Quarterly Progress Report, July 1-September 30, 1979. Mixing Steam-Water Mixing and System flydrodynamics Program - NUREG/CR-1657 Task 4. Quarterly Progress Report, October-December 1979. MOD 1 Code COMPARE-MOD 1 Code, Addendum 1. NUREG/CR-1185 Model Estimates COPS Model Estimates of LLEA Availability Near Selected NUREG/CR-1166 Reactor Sites. Modeliug Modeling Tornado Dynamics. NUREC/CR-1585 Moderated Arrays Critical Experiments with Interstitially-Moderated NUREG/CR-1071 Arrays of Low-Enriched Uranium Oxide. Topical Report on Reference Critical Experiments. Moisture The Effects of Temperature, Moisture, Concentration, NUREG-0678 Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon.
Report No. Keyword Listing _M Report Title Vegetational Cover in Monitoring and Stabilization of NUREG/CR-1358 Monitoring Shallow Land Burial Sites. Using Advanced Process Monitoring to Improve Material NUREG/CR-1676, Vol 1 Monitoring Control. The Use of Process Monitoring Data for Nuclear Material NUREG/CR-1670, Vol 1 Monitoring Data Accounting: Volume 1. Summary Report. Monoclinal Structure Monoclinal Structure and Shallow Faulting of the Reelfoot NUREG/CR-1501 Scarp as Estimated from Drill Holes with Variable Spacings. Monte Carlo Validation of a Monte Carlo Code for Radiation Streaming NUREG/CR-1334 Analyses. Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 3 Motion Accelerograph Stations. Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 4 Motion Accelerograph Stations. l l Geotechnical and Strong Motion Earthquake Data f rom U.S. NUREG/CR-0985, Vol 5 Motion Accelerograph Stations. State of-the-Art Study Concerning Near-Field Earthquake NUREG/CR-1340 Motion Gronnd Motion. l KUREG/CR-1504 l Motion A User's Guide to EPIC, a Computer Program to Calculate the M> tion of Fuel and Coolant Subsequent to Pin Failure in a LMFBR. Statistical Analysis of Earthquake Ground Motion NUREG/CR-1641 Motion Parameters, Compilation, Assessment and Expansion of the Strong NUREG/CR-1660 j Motion Data Earthquake Ground Motion Data Base. l NUREG/CR-1677, Vol 1 Motion Response Piping Benchmark Problems. Dynamic Analysis Uniform Support Motion Response Spectrum Method. On the Motions of Particles in Turbulent Flows. NUREG/CR-1554 Motions
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i Keyword Listing M Report Title Report No. Hulti-Processor Advanced Mobile Multi-Processor Gamma-Ray Acquisition NUREG/CR-1668 and Analysis System. Multirod Multirod Burst Test Program Progress Report for NUREG/CR-1450 July-December 1979. d o
5 Keyword Listing N Report Title Report No. I NACOM Code The NACOM Code for Analysti of Postulated Sodium Spray NUREG/CR-1605 g Fires in LMFBRs. Natural Draft Evalurtion of Mathematical Models for Characterizing NUREG/CR-1581, Vol 1
. Plume Behavior f rom Cooling Towers: Vol. 1--Dispersion from Single and Multiple Source Natural Draft Cooling Towers.
Natural Draft Evaluation of Mathematical Models for Characterizing NUREG/CR-1581, Vol 2 I Plume Beh.vior from Cooling Towers: Vol.2--Salt Drift Deposition from Natural Draft Cooling Towers. NUREG-0722 ; Natural Phenomena The Effects of Natural Phenomena on the Exxon Nuclear
. Company Mixed Oxide Fabrication Plant at Richland, Washington.
NDT-COMP 9 Tbt NDT-COMP 9 Microcomputer. NUREG/CR-1548 Nea r-Field State-of-the-Art Study Concerning Near-Field Earthquake NUREC/CR-1340 Ground Motion. Neutralization Model Fixed Site Neutralization Model Programmer's Model. NUREC/CR-1308, Vol 2 Nonsteady An Algorithm to Estimrte Field Concentrations Cader NUREG/CR-1474 No stecdy Meteorological Conditions from Wind Tunnel Experiments. Normal Shock Dynamic Analysis to Establish Normal Shock and Vibration NUREG/CR-1484 - of Radioactive Material Shipping Packages - Quarterly Progress Report. October 1-December 31, 1979. North Anna Safety F. valuation Report Related to the oper1 tion of NUREG-0053, Supp. 11 North Anna Power Station, Unit 2. Virginia Electric and Power Company, Docket No. 50-339. Supplement No. 11. North Anna Safety Evaluation Report st.ated to the Operation of NUREG-0053, Supp. 12 North Anna Power Statica _Jt 2, Docket No. 50-339. Supplement No. 12. l North Anna Final Environmental Statement Related to the Operation of NUREG-0134, Add. 2 North Anna Power Station, Unit 1 and 2, Docket No. 50-338 [ and 50-339. Virginia Electric and Power Company. > North Anna North Anna Power Station Unit 2 Tc hnical' Specifications NUREG-0664, Rev 1 Appendix "A" to Licanse No. NPF-7. ; t NRC United States Nuclear Regulatory Commiss.on Staff NUREG-0386, Supp. 2 i Practice & Procedure Digest. Supplemen' 2 to Digest No. 2. i
I A Keyword Listing N Report Titio Report No. NRC NRC Action Plan Developed as a Result of the TMI-2 NUREG-0660, Vol 1 Accident,. Revision 1 Vol 1. NRC NRC Action Plan Development as a Result of the TMI-2 NUREG-0660, Vol 2 Accident, Revision 1, Vol. 2 NRC Summary of Public Comments and NRC Staff Analysis Relating NUREG-0684 to Rulemaking on Emergency Planning for Nuclear Fower Plants. NRC Comments on the NRC Safety Research Program Budget for NUREG-0699 > Fiscal Year 1982. NRC Report to Congress: NRC Incident Response Plan. ITUREG-0728 NRC Report to Congress on NRC Emergency Communications. NUREG-0729 t NRC Report to Congress on the Acquisition of Reactor Data NUREG-0730 for the NRC Operations Center. NRC Material Accounting as Required by the United States NUREG/CR-1192 Nuclear Regulatory Coumission: Capabilities and Vulnerabilities. NRC Quarterly Technical Progress Report on Water Reactor NUREG/CR-1400 Safety Programs Sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research - April-June 1980. NRC Respirator Studies for the Nuclear Regulatory Commission, NUREG/CR-IS86 Evaluation and Performance of Escape-Type Self-Contained Breathing Apparatus, October 1,1978-September 30, 1979. NRC Plans NRC Plans for Cleanup Operations at Three Mile Island NUREG-0698 Unit 2. NRC Policy Plan for heevaluation of NRC Policy on Decommissioning of NUREG-0436, Rev 1, Supp 1 Nuclear Facilities. December 1978 to July 1980. Naclear Activities Potential Threat to Licensed Nuclear Activities from NUREG-0703 Insiders (Insider Study). Nuclear Data Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System. l i
Report No. Keyword Listing.N , Report Title NUREG-0436, Rev 1, Supp 1 Nuclear Facilities Plan for Reevaluation of NRC Policy on Decommissioning of Nuclear Facillites. December 1978 to July 1980, NUREG/CR-1448 Nuclear Facilities Physical Protection of Nuclear Facilities. Quarterly 4 Frogress Report, Jcnuary-March 1980. NUREG-0653 Nuclear Industry Report on Nuclear Industry Quality Assurance Procedures for Safety Analysis Computer Code Development and Use. NUREG/CR-1670, Vol 1 The Use of Process Monitoring Data for Nuclear Material Auclear Material j Accounting: Volume 1. Summary Report, i Enhancement of the Nuclear Materials Management and NUREG/CR-1527 Nuclear Materials Safeguards System. NUREG/CR-1528 Nuclear Materials Safeguards User's Manual for Nuclear Materials Management and Safeguards System. i 1 I e i
L Keyword Lioting 0 Report Title Report No. > Report to Congress on Abnormal Occurrences, January- NUREG-0090, Vol 3, No. 1 Occurrences t March 1980. I obio A Probabilistic Evaluation of Earthquake Detection 2nd NUREG/Ch-1648 Location Capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia. Oblo Earthquake Geophysical Investigation of the Anna, Ohio Earthquake NUREG/CR-1649 Zone. One-Dimensional Hydrogen-Hixing in a closed Containment Compartment Based KU4EG/CR-1575 on a one-Dimensional Model with Convective Effects. Onsite Inspection Assessment of Current On ite Inspection Techniques for NUREG/CR-1380, Vol 1 ES Light Water Reactor Fuel Systems - Volume 1 - Executive Summary.
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Open Pit Radon Release and Dispersion from an Open Pit Uranium KUREC/CR-1583 Mine. Safety Evaluation Report Related to Operation of Sequoyah NUREG-0011, Supp. 2 Operation ' Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328. Tennessee Valley Authority, Supp. No. 2. Safety Evaluation Report Related to operation of Sequoyah NUREG-0011, Supp. 3 Operation Naclear Plant, Units 1 and 2, Docket Nos. 50-327/328. Tennessee Valley Authority, Supp. 3. , Safety Evaluation Report Related to the Operation of NUREG-00$3, Supp. 11 Operation North Anna Power Station, Unit 2, Virginia Electric and Power Company, Docket No. 50-339. Supplement No. 11. j Safety Evaluation Report Related to the Operation of KUREG-0053, Supp. 12 Operation North Anna Power Station, Unit 2, Docket No. 50-339. Supplement No. 12. Safety Evaluation Report Related to the Operation of KUREG-0117, Supp. 4 Operation Joseph M. Farley Nuclear Plant, Unit 2. Docket No. 50-364, Alabama Power Company. Supplement 4 to NUREG 75/034. Safety Evaluation Report Related to Operation of Diablo KUREG-0675, Supp 10 Operation Canyon Nuclear Power Station, Units 1 and 2, Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Supplement No. 16. Operation Final Environmental Statement Related to the Operation NUREG-0702 of Gas Hills Uranium Project Docket No. 40-299, Union Carbide Corporation. Task Force Report on Interim Operation of Indian Point. RUREG 0715 Operation l l I 1 i
. . . . , , - - ~ - . . . . . .. I I i Keyword Listing,0 Report Title Report No. Operation Final Environmental Statement Related to the Operation of KUREG-0727, Adds the Joseph M. Farley Nuclear Plant, Units 1 and 2,
. Docket Nos. 50-348 and 50-364.
Operations NRC Plans for Cleanup Operations at Three Mile Island KUREG-0698 Unit 2. Operations Center Report to Congress on the Acquisition of Reactor Data KUREG-0730. for the NRC Operations Center. 4 Operations Center Considerations on Nuclear Data Link Implementation in KUREG/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System, Operator Training Nuclear Power Plant Simulators: Their Use in Operator KUREG/CR-1482 Training and Requalification. Opoeration Final Environmental Statement Related to the Operation of KUREG-0134, Add. 2 North Anna Power Station, Unit I and 2, Docket No. 50-338 and 50-339. Virginia Electric and Power Company. ORNL The ORNL State-Level Electricity Demand Forecasting KUREG/CR-1295 Model. Oxide The Effects of Natural Phenomena on the Exxon Nuclear NUREG-0722 ; Company Mixed Oxide Fabrication Plant at Richland, Washington. Oxide Etudy of Plutonium oxide Powder Emissions from Simulated NUREG/CR-1302 Shipping Container Leaks. Oxygen Alloys Instrumented Impact Properties of Zircaloy-Oxygen and KUREG/CR-1408 Zircaloy-Hydrogen Alloys.
- .. . . . - ~ . . . ... . . - . I Keyword Listing P Report Title Report No. Pacific Gas & Elect. Co. Safety Evaluation Report Related to Operation of Diablo NUREG-0675, Supp'10. Canyon Nuclear Power Station, Units 1 and 2, Pacific Gas and Electric Compary, Docket Nos. 50-275 and 50-113, ., Supplement No. 10. Package Review and Assessment of Package Requirements (Yellowcake) NUREG-0535 and Emergency Response to Transportation Accidentc. i Packages -Dynamic Analysts to Establish Normal Shock and Vibration KUREG/CR-1484 of Radioactive Material Shipping Packages - Quarterly trogress Report October 1-December 31, 1979. Parameters Statistical Analysis of Earthquake Ground Motion KUREG/CR-1641 Parameters,
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Particles On the Motions of Particles in Turbulent Flows. NUREG/CR-1554 Pa t t e rns Acute Ef f ects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF6/UO F Studies in Experimental 2 2 Animals. PitE Prompt Burst Energ-A ics Experiments: Fresh Uranium NUREC/CR-1396 Carbide / Sodium Series. PBF/ LOFT PBF/ LOFT Lead Rod Test Series Test Results Report. NUREC/CR-1538 Penetrations Data Summaries of Licensee Event Reports of Primary NUREG/CR-1730 Containment Penetrations at U.S. Commercial Nuclear Power Plants from January 1, 1972 to December 31, 1978. Performance Respirator Studies fer the Nuc]sar Regulatory Commission, KUREG/CR-1586 Evaluation and Performance of Escape-Type Self-Contained Breathing Apparatus, October 1, 1978-September 30, 1973. Performance Testing Performance Testing of Personnel Dosimetry Services: KUREG/CR-1593 Alternatives and Recommendations for a Personnel Dosimetry Testing Program, Performing SCALE: A Modular Core System for Performing Standardized NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSAS1 and CSAS2. Personnel Dosimetry Performance Testing of Personnel Dosimetry Services: KUREG/CR-1593 Alternatives and Recommendations for a Personnel Dcsimetry Testing Program. Pttenomena The Effects of Natural Phenomena on the Exxon Nuclear NUREG-0722 Company Mixed Oxide Fabrication Plant at Richland, Wa.hington, P
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Kep ord Listing P Report Title Report No. J l l Physical Evaluation Methodology for Fixed-Site Physical Protection NUREG/CR-1590 Systems. Ihystral Protection Acceptance Criteria for the Physical Protection Upgrade NUREC-0721 Rule Requirements for Fixed Sites. Physical Protection Design Guidance and Evaluation Methodology for Fixed NUREG/CR-il98, Vol 1 Site Physical Protection Systems, Volume 1. l Physical Protection Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 2 Site Physical Protection Systems, Volume 2. Physical Protection Physical Protection of Nuclear Facilities. Qua rte rly NUREG/CR-1448 Progress Report, January-March 1980. Physical Protection Inspection Methods for Physical Protection Project: KUREG/CR-1610 Val 1, No. 1 Quarterly Report, March-May 1980. 6 Physics Physics of Reactor Safety. -Quarterly Report for NUREC/CR-1526, Vol.1 January-March 1980, a Piles Measurement of Radon Diffusion from Uranium Mill Tailing NUREC/CR-1109 Piles. Pin Failure A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 the Motion of Fuel and Coolant Subsequent to Pin Failure in a LMFBR. Pin Simulators Thermocouple Signal Sensitivity to the Sheath Thickness NUREC/CR-1347 of Thermal-Hydraulic Test Facility Indirectly Heated Electric Fuel Pin Simulators. Pipe Lateral Loads on Vent Pipe in Steam Chugging. NUREG/CH-1631 Pipe Cracking Pipe Cracking Experience in Light-Water Reactors, NUREC-0679 1967 through 1979, l I J Piping. Technical Report on Material Selection and Processing NUREG-0313, Rev. I Guidelines for BWR Coolant Pressure Boundary Piping. Piping investigation and Evaluation of Cracking Incidents in NUREG-0691 Piping in Pressurized Water Reactors.
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E Keyword Listing P Report Title Report No. Pipir,g Pip'ing Inelastic Fracture Mechanics Analysis. NUREC/CR-Il19 , Piping Piping Benchmark Probleas. Dynamic Analysis Uniform HUREG/CR-1677, Vol 1 Support Motion Response Spectrum Method. Piping Systems Transient Analysis of Ceolant Flow and lleat Transfer NUREG/CR-1404 in LMFBR Piping Systems. Plant Reliability Nuclear Plant Reliability Data System 1979 Annual Reports NUREG/CR-1635 ; of Cumulative System and Component Reliability, r Plant Structures Variability of Dynamic Characteristics of Nuclear Power NUREG/CR-1661 Plant Structures. Plume Behavior Evaluation of Mathematical P.odels for Characterizing NUREG/CR-1581, Vol 1 Plume Behavior from Cooling Towers: Vol. 1--Dispersion from Single and Multiple Source Natural Draft Cooling Towers. Plume Behavior Evaluation of Mathematical Models for Characteriztng NUREG/CR-1581, Vol 2 Plume Behavior from Cooling Towers: Vol.2--Salt Drift Deposition from Natural Draft Cooling Towers. Plume Rise Evaluation of Mathematical Models for Characterizing KUREG/CR-1581, Vol 3 Plume Behavior from Cooling Towers: Vol. 3--Plume Rise from Mechanical Draft Cooling Towers. Plutonium Preparation of Working Reference Materials: Calcined NUREG/CR-1445 Waste Recovery Products Containing Uranium or Plutonium. Plutonium Oxide Study of Plutonium Oxide Powder Emissions from Simulated NUREG/CR-1302 Shipping Container Leaks. Policy Plan for Reevaluation of NRC Policy on Decommissioning of NUREG-0436, Rev 1, Supp 1 Nuclear Facilities. December 1978 to July 1980. i Pools lieat Removal Characteristics of Volume Heated Bolling NUREG/CR-1357 Pools with Inclined Boundaries. Post-Subcooled LWR Fuct Rod Post-Subcooled Blowdown Scoping Analysis. NUREG/CR-1568 Potential Review and Integration of Existing Literature Concerning NUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas.
I l Keyword Listi_ng P ' Report Title Report No. ! l j Potential An Assessment of LVR Fuel-Failure Propagation Potential: NUREC/CR-1471 i literature Survey. j Potential Threat Potential Threat to Licensed Nuclear Activities from NUREG-0703-Insiders (Insider Study). Powder Emissions Study of Plutonium Oxide Powder Emissions from Simulated FIREC/CR-1302 ; Shipping Container Leaks. ' Power Plant. . Financing Strategies for Nuclear Power Plant NUREG/CR-1481 ) Decommissioning. l Power Plant Nuclear Power Plant Simulators: Their Use in Operator NUREG/CR-1482 Training and Requalification. Power Plant Variability of Dynamic Characteristics of Nuclear Power NUREG/CR-1661 Plant Structures. Power Plants Control of IIeavy Loads at Nuclear Power Plants NUREG-0612 Resolution of Generic Technical Activity A-36. Power Plants Summary of Public Comments and NRC' Staff Analysis Relating NUREG-0684 to Rulemaking on Emerg+ncy Planning for Nuclear Power Plants.
' Power Plants Environmental Assessment for Effective Changes to 10 CFR NUREG-0685 Part 50 and Appendix E to 10 CFR Part 50; Emergency Planning Requirements for Nuclear Power Plants.
Power Station Measurement of XE-133, C-14 and Tritium in Air and 1-131 ' NUREG/CR-1195 in Vegetation and Milk Around the Quad Cities Nuclear Power Station. Precambrian Seismicity and Tectonic Relationships for Upper Great NUREG/0R-1569 Lakes Precambrian Shield Province. Precipitation Seasonal' Variation of 10-Square Mile Probable Maximum NUREG/CR-1486 Precipitation Estimates, United States East of the 105th Meridian (llydrometeorological Report No. 53). Predicting Predicting Life Expectancy and Simulating Age of Complex NUREG/CR-1466 j- Equipment Using Accelerated Aging Techniques. Preheat An Evaluation' of Condensation-Induced Water llammer in NUREG/CR-1606' Preheat Steam Generators.
i Keyword Listing P Report Title Report No. Prelimina ry Licensability of CANDU-Type Reactors in the United States. NUREG/CR-Ill3 A Prelir / Assessment of the R a7d D Requirements. Preliminary Data A Methodology and a Preliminary Data Base for Examining NUREG/CR-1539 the Health Risks of Electricity Generation from Uranium and Coal Fuels. Preparation Preparation of Working keference Materials: Calcined NUREC/CR-1445 , Waste Recovery Products Containing Uraniwn or Plutonium. PRESBC PRESBC: Pressure Boundary Conditions for the K-FIX Code. NUREG/CR-1536 Pressure Technical Report on Material Selection and Processing NUREG-0313, Rev. 1 Guidelines for BVR Coolant Pressure Boundary Piping. Pressure The Effects of Temperature, Moisture, Concentration, N'.' REG-067 8 Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon. Pressure Extended Analysis of Data from the 1/5-Scale MARK 1- KUREG/CR-0761 Boiling Water Reactor Pressure Suppression Experiment. Pressure PRESBC: Pressure Boundary Conditions for the K-FIX Code. KUREG/CR-1536 Pressure Boundary Structural Integrity of Water Reactor Pressure Boundary NUREG/CR-1472 Components. Quarterly Progress Report, January-March 1980. Pressure Rise Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592 to Sodium Boiling in fuel Subassemblies during Pump Coast-Down of an LMFBR. Pressure Vessel LWR Pressure Vessel Irradiation Surveillance Dosimetry NUREG/CR-0720 Quarterly Progress Report, October-December 1978. Pressure Vessel Critical Experiments, Measurements, and Analyses to NUREG/CR-1601 Establish a Crack Arrest Methodology for Nuclear Pressure Veseel Steels. Primary Cooling Final Environmental Statement Related to Primary Cooling NUREG-0686 System Chemical Decontamination at Dresden Nuclear Power Station, Unit No. 1. Docket No. 50-010. Probabilistic Evaluation A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648 Location Capability for Illinois, Indiana, Kentucky, Ohio, and West Virginia.
. . . . - .~, . .~ . -. _ _-.- . .-- - _- - - Keyword Listing P Report Title Report No. Probabilistic Model A Deterministic-Probabilistic Model for Contaminant NUREG/CR-1609' Transport. Procedure Digest United States Nuclear Regulatory Commission Staff NUREG-0386, Supp(.2 i Practice & Procedure Digest. Supplement 2 to Digest No, 2. Process Monitoring The Use of Process Monitoring Data for Nuclear Material NUREG/CR-1670, Vol 1 Accounting: Volume 1. Summary Report. 1 I Process Monitoring Using Advanced Process Monitoring to Improve Material NUREG/CR-1676, Vol 1 Control,
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Processing Technical Report on Material Selection and Processing NUREG-0313, Rev. 1 Guidelines for BWR Coolant Pressure Boundary Piping. Processor Advanced Mobile Multi-Processor Gamma-Ray Acquisition NUREG/CR-1668 and Analysis System. ,
- l Programmer's Manual Fixed Site Neutralization Model Progranner's Model. NUREG/CR-1308, Vol 2 i
Project M-25 Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 1 Milling Project M-25: Volume 1 - Sunnary and Text. l Project M-25 Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 2 Milling Project M-25: Volume II - Appendices A-F. P Project M-25 Final Generic Environmental. Impact Statement on Uranium NUREG-0706, Vol 3 Milling Project M-25: Volume Ill - Appendices G-V. Prompt Burst Prompt Burst Energetics Experiments: Fresh Uranium NtjREG/CR-1396 i Carbide / Sodium Series. Propagation An Assessment of LWR Fuel-Failure Propagation Potential: NUREG/CR-1471 Literature Survey. 4 Properties Properties of Radioactive Wastes and Waste Containers. NUREG/CR-1514 Quarterly Progress Report, January-March 1980. Protection Physical Protection of Nuclear Facilities. Quarterly NUREG/CR-1448 Progress Report, January-March 1980.
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. ~ m . . . . _ . _ . _ _ . _ _ ..m.. .- _ _ Keyword Listing P Report Title Report No. Protection Project Inspection Methods for Physical Protection Project: NUREG/CR-1610, Vol 1, No. 1 Quarterly Report, March-May 1980.
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Protection Systems Design Guidance for Evaluation Methodology for Fixed NUREG/CR-1193, Vol 1 Site Physical Protection Systems, Volume 1. Protection Systems Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 2 Site Physical Protection Systems, Volume 2. Protection Systems Evaluation Methodology for Fixed-Site Physical Protection NUREG/CR-1590 Systems. Protection Upgrade Acceptance Criteria for the Physical Protectica Upgrade NUREG-0721 Rule Requirements for Fixed Sites. Protothaca Staminea Growth and Histological Effects to Protothaca Staminea NUREG/CR-1298 (Littleneck Clam) of Long-Term Exposure to Chlorinated Sea Water. Psychological Stress Psychological Stress for Alternatives of Decontamination NUREG/CR-1584 of TMI-2 Reactor Building Atmosphere. Public Comments Summary of Public Comments and NRC Staff Analysis Relating NUREG-0684 to Rulemaking on Emergency Planning for Nuclear Power Plants. e Pump -Compressible Analysis of Inlet Plenum Pressure Rise due KUREG/CR-1592 to Sodium Boiling in Fuel Subassemblies during Pump Coast-Down of an LMFBR. PWR Standard Technical Specifications for Westinghouse NUREG-0452, Rev 3 Pressurized Water Reactors, Revision 3. , PWR Investigation and Evaluation of Cracking Incidents in NUREG-0691 Piping in Pressurized Water Reactors. PWR A Risk Assessment of a Pressurized Water Reactor for Class NUREG/CR-0603 3-8 Accidents. PWR Uncertainty Analysis for a PWR Loss-of-Coolant Accident: NUREG/CR-1364 II. Alternative Core Damage Estimators. PWR Summary of Thermal Hydraulic Calculations for a NUREG/CR-1480 Pressurized Water Reactor. > l
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~ E!p ord [.isting P- Report Title Report No.
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PWR Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518
.and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents.
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_m. . .. _ . _ . _ . . . _ _ . Keyword Listing Q Report Title Report No. QTE Qualification Testing Evaluation Program Light-Water NUREG/CR-1492 Reactor Safety Research Quarterly Report, July-September 1979. Quad Cities Measurement of XE-133, C-14 and Tritium in Air and I-131 NUREG/CR-1195 in Vegetation and Milk Around the Quad Citics Nuclear Power Station. Qualification Evaluation of Simulator Adequacy for the Radiation NUREC/CR-1184 Qualification of Safety Related Equipment. Qualification Qualification Testing Evaluation Program Light-Water NUREG/CR-1492 Reactor Safety Research Quarterly Report, July-September 1979. Qualification Test Results on 1550*C and 2200*C 1/16-Inch NUREG/CR-0961 Qualification Test 0.D. Fuel Centerline Thermocouples for the LOFT Program. Quality Assurance Report on Nuclear Industry Quality Assurance Procedures NUREG-0653 for Safety Analysis Computer Code Development and Use. Answers to Frequently Asked Questions about Cleanup NURdG-0732 Questions Activities at Three Mile Island, Unit 2. 1 i l 1 l
Keyword Listing R ReportTit1}g Report No.- R and D Liccasability of CANDU-Type Reactors in the United States. KUREG/CR-1113 A Preliminary Assessment of the R and D Requirements. Radiation Evaluation of Simulator Adequacy for the Radiation NUREG/CR-1184 Qualification of Safety Related Equipment. Radiation Validation of a Monte Carlo Code for Radiation Streaming KUREG/CR-1334 Analyses. Radioactive Material ' Dynamic Analysis to Establish Normal Shock and Vibration KUREG/CR-1484 . of Radioactive Material Shipping Packages - Quarterly i Progress Report October 1-December 31, 1979. Radioactive Material Transportation of Radioactive Material in Kentucky. KUREG/CH-1671 Radioactive Materials Review and Integration of Existing Literature Concerning NUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. r Radioactive Materials Identification and Assessment of the Social Impacts of KUREG/CR-0744 Transportation of Radioactive Materials in Urban Envi ronment s . Radioactive Waste Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank. Correlation Among Input Variables for Simulation Studies. Radioactive Waste Evaluation of Isotope Migration-Land Burial Water Chemistry KUREG/CR-1289 at Commercially Operated Low Level Radioactive Waste Disposal Sites. Radioactive Waste Risk Methodology for Geologic Disposal of Radioacti e KUREG/CR-1377 Waste: Transport Model Sensitivity Analysia. Radioactive Waste Evaluation of Isotope Migration-Land Burial Water NUREG/CR-1513 Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report, January-March 1980. Radioactive Waste Scenario Development and Evaluation Related to the Risk KUREG/CR-1608 Assessment of High Level Radioactive Waste Repositories. l Radioactive Wastes Properties of Radioactive Wastes and Waste Containers. NUREG/CR-1514 , Quarterly Progress Report, January-March 1980. Radioactivity Health Status and Rndy Radioactivity of Former Thorium NUREG/CR-1420 Workers, i l i i
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I Keyword Listing R Report Title Report No. Radioactivity Logs Estimates of Uranium Content and Radon Flux for Uranium KUREC/CR-1549 Mine Dumps Based on Borehold Radioactivity Logs. Radiological Emergency Survey of Current State Radiological Emergency Response KUREG/CR-1620 Capabilities for Transportation Related incidents. Radon Radon Release and Dispersion from an Open Pit Uranium KUREC/CR-1583 Mine. Radon Diffusion Measu;;<tra of Radon Dif fusion from Uranium Mill Tailing NUREC/CR-1109 Pil2s. Radon Flux Estimates of Uranium Content and Radon Flux for Uranium. NUREG/CR-1549 Mine Dumps Based on Borehold Radicartivity Logs. Rail Shock Environments for Large Shipping Containers During KUREG/CR-l?77 Rail Coupling Operations, Reactor Building Psychological Stress for Alternatives of Decoutsmination KUREG/CR-1584 of TMI-2 Reactor Building Atmosphere. Reactor Data Report to Congress on the Acquisition of Reactor Data NUREG-0730 for the KRC Operations Center. Reactor Plants Wind Tunnel Measurements of Dispersion and Turbulence NUREG/CR-1475 in the Wakes of Nuclear Reactor Plants. Reactor Pressure Structural Integrity of Water Reactor Pressure Boundary KUREG/CR-1472 Components. Quarterly Progress Report, January-March 1980. l l Reactor Safety Nuclear Reactor Safety Quarterly Progress Report, NUREG/CE-1516 October 1-December 31, 1979. Reactor Safety Reactor Safety Research Programs. Quarterly Report - KUREG/CR-1009 , July-September 1979. I Reactor Safety Reactor Safety Research Programs. Quarterly Report, KUREG/CR-1349 October-December 1979. 1 Reactor Safety Quarterly Technical Progress Report on Water Reactor KUREC/CH-1400 Safety Programs Sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research - April-June 1980. l i l
i Keyword Listing R Report Title Report No. i Reactor Safety Advanced Reactor Safety Research Division Quarterly NUREG/CR-1402 Progress Report. October 1-December 31, 1979. Reactor Safety Water Reactor Safety Research Division Quarterly NUREG/CR-1403 Progress Report. October-December 1979. Reactor Safety An Ultrasonic Thermometry System for Measuring Very High NUREG/CR-1488 Temperatures in Reactor Safety Experiments. Reactor Safety Qualification Testing Evaluation Program Light-Water NUREG/CR-1492 Reactor Safety Research Quarterly Report, July-September 1979. Reactor Safety Advanced Reactor Safety Research Division Quarterly NUREG/CR-1505 Progress Report, January 1-March 31,1980. Reactor Safety Light Water Reactor Sefety Research Program Quarterly NUREG/CR-1509 Report, January-March 1980. Peactor Safety High-Temperature Gas-Cooled Reactor Safety Studies for the NUREG/CR-1521 Division of Reactor Safety Research Quarterly Progress Report, January 1-March 31,1980. Reactor Safety Physics of Reactor Safety, Quarterly Report for NUREG/CR-1526, Vol 1 Janua ry-March 1980. Reactor Sites COPS Model Estimates of LLEA Availability Near Selected NUREG/CR-1166 Reactor Sites. . Reactor Vessel Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris Coolability in TKLB, SZD, and ABG Accidents. Reactors Licensability of CANDU-Type Reactors in the United States. NUREG/CR-1113 A Preliminary Assessment of the R and D Requirements. Recommendations Performance Testing of Personnel Dosimetry Servicesi NUREG/CR-1593 Alternatives and Recommendations for a Personnel Dosimetry Testing Program. s Recovery Products Preparation of Working Reference Materials: Calcined NUREG/CR-1445 Waste Recovery Products Crntaining a Uranium or Plutonium. Reelfoot Scarp Monoclinal Structure and Shallow Faulting of the Reelfoot NUREG/CR-1501 Scarp as Estimated f rom Drill Holes with Variable Spacings. i
Keyword Lioting R Report Title Report No. Reevaluation Plan for Reevaluation of NRC Policy on Decommissioning of NUREG-0436, Rev 1. Supp 1 Nuclear Facilities. December 1978 to July 1980. Reference Materials Preparation of Working Reference Materials: Calcined NUREG/CR-1445 Waste Recovery Products Containing Uranium or Plutonium. Regional Regional Relationships Among Earthquake Magnitude EUREG/CR-1457 Scales. Regulatory Reports Regulatory and Technical Reports Compilation for 1979. NUREG-0304, Vol. 4 i Reinforced Concre'e Strength and Stiffness of Tensioned Reinforced Concrete NUREG/CR-1602 Panels Subjected to Membrane Shear, Two-Way Reinforcing. Relation Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 Relation to uhe Technical Sup; ort Center, Emergency ' Operations Center and Safety Parameter Display System. Relationships Regional Relationships Among Earthquake Magnitude NUREG/CR-1457 Scales. Relationships Seismicity and Tectonic Relationships for Upper Great NUREG/CR-1569 Lakes Precambrian Shield Province. Release Radon Release and Dispersion from an Open Pit Uranium NUREG/CR-1583 Mine. Reliability Nuclear Plant Reliability Data System 1979 Annual Reports KUREG/CR-1635 of Cumulative System and Component Reliability. Repair Final Environmental Statement Related to Steam Generator NURG-0692 Repair at Surry Power Station, Unit No. 1. Virginia Electric and Power Cmpany, Docket No. 50-280. Repositories Scenario Development and Evaluation Relsted to the Risk NUREG/CR-1608 Assessment of High Level Radioactive Waste Repositories. Requalification Nuclear Power Plant Simulators: Their Use in Operator NUREG/CR-1482 Training and Requalification. 1 Required l Material Accounting as Required by the United States NUREG/CR-1192 i Nuclear Regulatory Commission: Capabilities and Vulnerabilities. J l
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Keyword Listing R Report Title Report No. f Requirements Review and Assessment of Package Pequirements (Yellowcake) NUREG-0535 and Emergency Response to Transrottation Accidents. Requirements Environmental Assessment for Effective Changes to 10 CFR NUREG-0685 Part 50 and Appendix E to 10 CFR Part 50; Emergency Planning Requirements for Nuclear Power Plants. Requirements Acceptance Criteria for the Pbysical Protection Upgrade NUREG-0721 Rule Requirements for Fixed Sites, e Requirements Licensability of CANDU-Type Reactors in the United States. NUREG/CR-lll3 A Preliminary Assessment of the R and D Requirements. Research Program Seismic Safety Margins Research Program (Phase 1). NUREG/CR-1120, Vol 3 Progress Report No. 7. Research Programs Comments on the NRC Safety Research Program Budget for NUREG-0699 Fiscal Year 1982. Research Programs Reactor Safety Research Programs. Quarterly Report - .NUREG/CR-1009 July-September 1979. Research Programs Reactor Safety Research Programs. Quarterly Report,- NUREG/CR-1349 October-December 1979. , Resolution Control of Heavy Loads at Nuclear Power Plants NUREG-0612' Resolution of Generic Technical Activity A-36. Resolution MARK 1 Containment Long Term Program Safety Evaluation NUREG-0661 Report, Resolution of Generic Technical Activity A-7. February 1977 to December 1979. Resources Utility Management and Technical Resources. NUREC/CR-1656 Respirator Studies Respirator Studies for the Nuclear Regulatory Commission, NUREG/CR-1586
- Evaluation and Performance of Escape-Type Self-Contained Breathing Apparatus, October 1, 1978-September 30, 1979.
Response Functional Criteria for Emergency Response Facilities. NUREG-0696 Response CONAN: An UfFBR Containment Response Computer Code. NUREG/CR-1355
Keyword Lioting R Report Title Report No. Response Piping Benchmark Problems. Dynamic Analysis Uniform NUREG/CR-1677, Vol 1 Support Motion Response Spectrum Method. Recponse Plan Report to Congress: NRC Incident Response Plan. NUREG-0728 Recult NRC Action Plan Developed as a Result of the TMI-2 NUREG-0660, Vol 1 Accident, Revision 1, Vol 1. 6 Recult NRC Action Plan Development as a Result of the TMI-2 NUREG-0660, Vol 2 Accident, Revision 1, Vol. 2. Revtew Review and Assessment of Package Requirements (Yellowcake) NUREG-0535. and Emergency Response to Transportation Accidents. Review Review and Integration of Existing Literature Concerning KUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. Richland, WA The Effects of Natural Phenomena on the Exxon Nuclear NUREG-0722 Company Mixed Oxide Febrication Plant at Richland, Washington. Rish Assessment A Risk Assessment of a Pressurized Water Reactor for Class NUREG/CR-0603 3-8 Accidents. Risk Assessment Scenario Development and Evaluation Related to the Risk NUREG/CR-1608 Assessment of High Level Radioactive Waste Repositories. Risk Methodology Risk Methodology for Geologic Disposal of Radioactive NUREC/CR-1262 Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. Rish Methodology Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1377 Waste: Transport Model Sensitivity Analysis. L Rishs A Methodology and a Preliminary Data Base for Examining NUREG/CR-1539 the Health Risks of Electricity Generation from Uranium and Coat Fuels. Rock Accelerograph Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, V. 2 App Accelerograph Stations in California. Rock Accelerograph Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, Vol 2 Accelerograph Stations in California.
l i Report No. .j l Keyword Listing _R Ry ort Title Rock Accelerograph Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055. Vol 3 Accelerograph Stations in California. l l l Rod LWR Fuel Rod Post-Subcooled Blowdown Scoping Analysis. NUREG/CR-1568 PBF/ LOFT Lead Rod Test Series Test Results Report. NUREG/CR-1538 Rod Test Rods Criticality Experimentu with Suberitical Clusters of NUREG/CR-1547 2.35 Vt% and 4.31 Wt1 U-235 Enriched UO Rods in Water at a Water-to-Fuel Volume Ratio of 1.6.2 I Summary of Public Comments and NRC Staff Analysis Relating NUREG-0684 Rulemaking to Rulemaking on Emergency Planning for Nuclear Power Plants. l l I
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t J F Keyword Listio_ Report' Title. Report No.
-Safegua rds Safeguards Summary Event List (SSEL). NUREG-0525, Rev 2 Technical Safeguards Issues for Alternative Fuel Cycles.
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Safeguards NUREG/CR-1048 i Safeguards Sateguards Material Control and Accounting Program: NUREG/CR-1485, Vol 1, No. 1 Quarterly Report, October-December 1979. Safeguards Safeguards User's Manual for Nuclear Materials Management NUREG/CR-1528 and Safeguards System. P Safeguards System Enhancement of the Nuclear Materials Management and NUREG/CR-1527 Safeguards System. Safeguards System Safeguards User's Manual for Nuclear Materials Management NUREG/CR-1528 f and Safeguards System. Safety Analysis Report on Nuclear Industry Quality Assurance Procedures NUREG-0653 for Safety Analysis Computer Code Development and Use. i i Safety Analysis SCALE: A Modular Code System for Performing Standardized NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticality Sefsty Analysis Modules CSASI and CSAS2. Safety Evaluation Safety Evaluation Report Related to Operation of Sequoyah NUREG-00ll, Supp, 2 [ Nuclear Plant, Units 1 and 2 Docket Nos. 50-327 and 50-328. Tennessee Valley Authority, Supp. No. 2. Safety Evaluation Safety Evaluation Report Related to Operation of Sequoyah NUREG-00ll, Supp. 3 N ' lear Plant, Units 1 and 2, Docket Nos. 50-327/328.
'.ennessee Valley Authority, Supp. ' 3.
Safety Evaluation Safety Evaluation b ort Related to the Operation of NUREG-0053, Supp.11 North Anna Power ih ion, Unit 2, Virginia Electric and Power Company, Docket No. 50-339. Supplement No. 11.
- Safety Evaluation Safety Evaluation Report Relatcd so tbe Operation of NUREG-0053, Supp. 12
( North Anna Power Station, Unit 2, Docket w;o. 50-339. Supplement No. 12. Safety Evaluation Safety Evaluation Report Relateu to the Operation of NUREG-0117, Supp. 4 Joseph M. Farley Nuclear Plant, Unit 2. Docket No. 50-364, Alabama Power Company. Supplement 4 to NUREG-75/034, Safety Evaluation HARK 1 Containment Long Term Program Safety Evaluation NUREG-0661 Report , Resolution of Generic Technical Activity A-7. February 1977 to December 1979. r 9
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- -. . , - . . , . - -. .- , =- 1 Keyword Listing S Report Title Report No. 4 Safety Evaluation Safety Evaluation Report Related to Operation of Diablo NU2EG-0675, Supp 10 i Canyon Nuclear Power Station, Units 1 and 2, Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, , Supplement No. 10. Safety Experiments An Ultrasonic Thermometry System for Measuring Very High NUREG/CR-1488 Temperatures in Reactor Safety Experiments. Safety Margins Seismic Safety Margins Research Program (Phase 1). NUREG/CR-1120. Vol 3 Progress Report No. 7. Safety Parameter Considerations on Nuclear Data Link Implementation in NUREG/CR-1579 Relation to the Technical Support Center, Emergency ; Operations Center and Safety Parameter Display System. Safety Programs Quarterly Technical Progress Report on Water Reactor NUREG/CR-1400 Safety Programs Sponsored by the Nuclear Regulatory , Commission's Division of Reactor Safety Research - April-June 1980. Safety Related Evaluation of Simulator Adequacy for the Radiation NUREG/CR-1184 Qualification of Safety Related Equipment, r Safety Report Physics of Reactor Safety. Quarterly Report for NUREG/CR-1526, Vol 1 January-March 1980. Safety Research Comments on the NRC Safety Research Program Budget for NUREG-0699 Fiscal Year 1982. { Safety Research Reactor Safety Research Programs. Quarterly Report - NUREG/CR-1009 July-September 1979. Safety Research Reactor Safety Research Programs. Quarterly Report, NUREG/CR-1349 f October-December 1979.
-Safety Research Advanced Reactor Safety Research Division Quarterly KUREG/CR-1402 1 Progress Report. October 1-December 31, 1979. j 1
Safety Research Water Reactor Safety Research Division Quarterly NUREC/CR-1403 Progress Report. October-December 1979. 1 Safety Research Qualification Testing Evaluation Program Light-Water NUREG/CR-149' Reactor Safety Research Quarterly Report, July-September 1979. Safety Research Advanced Reactor Safety Research Division Quarterly NUREC/CR-1505 q Progress Report, January 1-March 31,1980, i l 1
l Keyword Listing S Report Title Report No. Safety Research Light Water Reactor Safety Research Program Quarterly- .NUREG/CR-1509 Report, January-March 1980.
. Safety Research Nuclear Reactor Safety Qua*rly Progrest. Report, NUREG/CR-1516 October 1-December 31, 1979.
Safety Research High-Temperature Gas-Cooled Reactor Safety Studies for the NUREG/CR-1521 Division of Reactor Safety Reses.ch Quarterly Progress Report, January 1-March 31, 1980. j Salt Drift' Evaluation of Mathematica1'Models for Characterizing 'NUREG/CR-1581, Vol 2 Plume Behavior from Cooling Towers: Vol.2--Salt Drift Deposition from Natural Draft Cooling Towers. SCALE SCALE A Modular Core System for Performing Standardized -NUREG/CR-0200 Computer Analyses for Licensing Evaluation. SCALE System Criticality Safety Analysis Modules CSASI and CSAS2. Scarp Monoclinal Structure and Shallow Faulting of the Reelfoot NUREG/CR-1501 Scarp as Estimated from Drill Holes with Variable Spacings. Scoping Analysis LWR Fuel Rod Post-Subcooled Blowdown Scoping Analysis. NUREG/CR-1568 Sea Water Growth and Histological Effects to Protothaca Staminea NUREG/CR-1298 ; (Littleneck Clam) of Long-Term Exposure to Chlorinated ! Sea Water. Seasonal Variation Seasonal Variation of 10-Square Mile Probable Maximum NUREG/CR-1486 Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. 53). Security Security Communication Systems for Nuclear Fixed-Site NUREG/CR-0508 Facilities. Seismic Seismic Safety Margins Research Program (Phase I). NUREG/CR-1120, Vol 3 Progress Report No. 7. Seismic Seismic Hazard Analysis. A Methodology for the Eastern NUREG/CR-1582, Vol 2 United States. Seismic Seismic Hazard Analysis. Solicitat on of Expert NUREG/CR-1582, Vol 3 Opinion. Seismic Canadian Seismic Agreement. NUREG/CR-1637 7
KeywordListing}S Repert Title Report No. Seismit Analysis Best Estimate Method vs Evaluation Method: A Comparison NUREG/CR-1489 of Two Techniques in Evaluating Seismic Analysis and Design, Seismic Design Best Estimate Method vs Evaluation Method ' A Comparison i NUREG/CR-1489 of Two Techniques iu~ Evaluating Seismic Analysis and Design, Seismicity Seismicity and Tectonic Relationships for Upper Great t NUREG/CR-1569 Lakes. Precambrian Shield Province, Self-Contained Respirator Studies for the Nuclear Regulatory Commission, NUREC/CR-1586 Evaluation and Performance of Escape-Type Self-Contained. Breathing Apparatus, October 1, 1978-September 30, 1979 Self-Contained Respirator Studies for the Nuclear Regulatory Commission, NUREG/CR-1586 Evaluation and Performance of Escape-Type Self-Contained , Breathing Apparatus, October 1, 1978-September 30, 1979. Sensitivity Thermocouple Signal Sensitivity to the Sheath' Thickness NUREG/CR-1347 of Thermal-Hydraulic Test Facility Indirectly Heated ; ' Electric Fuel Pin Simulators. Sensitivity Analysis Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1377 Waste: Transport Model Sensitivity Analysis. ' Sequoyah Safety Evaluation Report Related to Operation of Sequoyah NUREG-0011, Supp. 2 Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328. Tennessee Valley Authority, Supp. No. 2. Sequoyah Safety Evaluation Report Related to Operation of Sequoyah ' NUREG-00ll, Supp. 3 Nuclear Plant, Units 1 and 2, Docket. Nos. 50-327/328. Tennessee Valley Authority, Supp. 3. Sequoyah
.Technicat Specifications, Sequoyah Nuclear Plant, Unit No. NUREG-0658, Rev l-1, Docket No. 50-327, Appendix "A" to License No. DPR-77.
SETS Vital Area Atulysis Using SETS. NUREG/CR-1487 Shallow Vegetational Cover in Monitoring and Stabilization of NUREC/CR-1358 Shallow Land Burial Sites. Shallow Faulting Monoclinal Structure and Shallow Faulting of the Reelfoot NUREC/CR-1501 Scarp as Estimated f rom Drill Holes with Variable Spacings. Sheath Thickness Thermocouple Signal Sensitivity to the Sheath Thickness I NUREG/CR-1347 of Thermal-Hydraulic Test Facility Indirectly Heated Electric Fuel Pin Simulators.
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Report Title Report No. Yg word Listing a Characterization of Existing Surisce Conditions at NUREG/CR-1683 She f field Sheffield. Low-Level Waste Disposal Facility. Seismicity and Tectonic Relationships for Upper Great NUREG/CR-1569 Shield Province Lakes Precambrian Shield Province. Study of Plutonium Oxide Powder Emissions from Simulated 'NUREG/CR-1302 Shipping Container Shipping Container Leaks. Shock Environments for Large Shipping Containers During . NUREG/CR-1277 Shipping Containers Rail Coupling Operations. Dynamic Analysis to Establish Normal Shock and Vibration NUREC/CR-1484 Shipping Packages of Radioactive Material Shipping Packages - Quarterly Progress Report October 1-December 31, 1979. Shock Environments for Large Shipping Containers During NUREG/CR-1277 She:k Rail Coupling Operations. Dvnar.ic Analysis to Establish Normal Shock and Vibration NUREC/CR-1484 Shock of Radioactive Material Shipping Packages - Quarterly Progress Report, October 1-December 31, 1979. rhermocouple Signal Sensitivity to the Sheath Thickness NUREG/CR-1347 Signal of Thermal-Hydraulic Test facility Indirectly Heated Electric Fuel Pin Simulators. SIMMER-II: A Computer Program for LMFBH Disrupted Core NUREG/Cb 9453, Rev 1 SIMMER-Il Analysis. Predicting Life Expectancy and Simulating Age of Complex NUREG/CR-1466 Simulating Age Equipment Using Accelerated Aging Techniques. Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1262 Simulation Studies Waste: A Distribution-Free Approach to Inducing Rank Correlation Among Input Variables for Simulation Studies. Evaluation of Simulator Adequacy for the Radiation UUREG/CR-1184 Simulator Qualification of Safety Related Equipment. Simulators Nuclear Power Plact Simulators: Their Use in Operator NUREG/CR-1482 Training and Requalification. Evaluation of Mathematical Models for Characterizing NUREG/CR-15El, Vol 1 Single Source Plume Behavior from Cooling Towers: Vol. 1--Dispersion fro.4 Single and Multiple Source Natural Draft Cooling Towers.
Keyword Listing S Report Title Report No. Site l Design Guidance and Evaluation Methodology for Fixed NUREG/CR-1198, Vol 1 Site Physical Protectica Systems, Voiume 1. Site Fixed Site' Neutralization Model Programmer's Model. NUREG/CR-1308 Vol 2 Site-Dependent Site-Dependent Eff ects at Strong-Motion Accelerograph KUREG/CR-1639 Stations. Sites Acceptance Criteria for lue Physical Protection Upgrade NUREG-0721 Rule Requirements for Fiied Sites. Sites COPS Model Estimates of LLEA Availability Near Selected NUREG/CR-il66 Reactor Site.s. Sites Design Guidance and Evaluation Methodology for Fixed NJ' REC /CR-Il98, Vol 2 Site Physical Protection Systems,'Jolume 2. Sites Vegetational Cover in Monitoring and Stabilization of NUREG/CR-1258
-Shallow Land Burial Sites.
Sites Evaluation of Isotope Migration-Land Burial Water NUREG/CR-1513 Chemistry at Commercially Operated Low-Level Radioactive Waste Disposal Sites. Quarterly Progress Report, January-March 1980. Skyshine Calculations of the Skyshine Campt+ Ray Dose Rates from NUREG/CR-0723 Independent Spent Fuel Storage Installations (ISTSI) Under Worst Case Accident Conditions. Social Impacts Review and Integration of Existing Literature Concerning NUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. Social Impacts Identification and Assessment of the Social Impacts of MUREG/CR-0744 Transportation of Radioactive Materials in Urban Environments. Sodium Boiling Compressible Analysis of Inlet Plenum Pressure Rise due RUREG/CR-1592 to Sodium Boiling in Fuel Subassemblies during Pump Coast-Down of an IliFBR. Sodium Series Prompt Burst Energetics Experiments: Fresh Uranium NUREG/CR-1396 Carbide / Sodium Series. Sodium Spray The NACDM Code for Analysif of Postulated Sodium Spray KUREG/CR-1405 Fires in LMTBRs.
i Keyword Listing S Report Title Report No. . l Soil Evaluation of In-Situ Soil Damping Characteristics. NUREG/CR-1638 Solicitation Seismic Fazard Analysis. Solicitation of Expert NUREG/CR-1582, Vol 3 ; Opinion. g Polubility Solubility Classification of Airborne Uranium Produ:ts NUREG/CR-1428 from LWR-Fuel Plants.
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Source Term In-Plant Source Term Measurements at Turkey Point NUREG/CR-1629 ,
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Station - Units 3 and 4 Spacings Honoclinal Structure and Shallow Faulting of the Reelfoot NURE4/JR-1501 ' Scarp'at Estimated from Drill Holes with Variable Spacings. Species Acute Toxicity and Bioaccumulation of Chloroform to Four .NUREG/CR-0893 Opecies of Freshwater Fish. Specifications Standard Technical Specifications for Westinghouse NUREG-0452, Rev 3 Pressurized Water Reactors, Revision 3. Specifications Technical Specifications, Sequoyah Nuclear Plant, Unit No. NUREG-0658, Rev 1 1, Docket No. 50-327, Appendix "A" to License No. DPR-77. Specifications North Anna Power Station Uait 2 Technical Specifications NUREG-0664, Rev 1 Appendix "A" to License No. NPF-7. Spectrum Method Piping Benchmark Problems. Dynamic Analysis Uniform NUkEG/CR-1677, Vol I , Support Motion Response Spectrum Method. Spent Fuel calculations of the Skyshine Gamma-Ray Dese Rates from PJREG/CR-0723 Independent Spent Fncl Storage installations (ISFSI) Under Worst Case Accident Conditions. Spool Pieces Two-Phase Flow Measurements with Advanced Instrumented NUREG/CR-1529 Spool Pieces. SSEL Safeguards Summary Event List (SSEL). NUREG-0525, Rev 2 SSMRP Seismic Safety Margins Research Program (Phase I). KUREG/CR-Il20, Vol 3 Progress Report No. 7.
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Keyword Listing S Report Title Report No. Stabilization Vegetational Cover in Monitoring and Stabilization of NUREG/CR-1358 Shallow Land Burial Sites. Staff Anal! Tis Summary of Public Comments and NRC Staff Analysis Relating NUREG-0684 to Rulemaking on Emergency Planning for Nuclear Power Plants. Staff Practice United States Nuclear Regulatory Commission Staff NUREG-0386, Supp. 2 Practice & Procedure Digest. Supplement 2 to Digest No. 2. Standard Specifications Standard Technical Specifications for Westinghouse NUREG-0452, Rev 3 Pressurized Water Reactors, Revision 3. Standardized SCALE: A Modular Code System for Performing Standardized NUREG/CR-0200 j Computer Analyses for Licensing Evaluation. SCALE System-Criticality Safety Analysis Modules CSASI and CSAS2. State-Level The ORNL State-Level Electricity Demand Forecasting NUREC/CR-1295 Model. State-of-the-Art State-of-the-Art Study Concerning Near-Field Earthquake NUREG/CR-1340 Ground Motion. Stattatical Analysis Statistical Analysis of Steam Generator Inspection Plans NUREG/CR-12E2 and Eddy Current Testing. Statistical Analysis Statistical Analysis of Earthquake Ground Motion NUREG/CR-1641 Par: meters. Steam Chugging Lateral Loads on Vent Fipe in Steam Chugging. NUREG/CR-1631 Steam Generator Final Environmental Statement Related to Steam Generator NUREG-0692 Repair at Surry Power Station, Unit No. 1. Virgleia Electric and Power Cmpany, Docket No. 50-280. Steam Generator Statistical Analysis of Steam Generator Inspection Plans NUREC/CR-1282 and Eddy Current Testing. Steam Generator Eddy-Current In6pection for Steam Generator Tubing Program. NUREC/CR-1563 Annual Progress Report for Period Ending December 31, 1979. Steam Generators An Evaluation of Condensation-Induced Water Hammer in NUREG/CR-1606 Preheat Steam Generators.
Keyword Listing _S Report Title Report No. Steam Line Steani Line Dynamics. NUREG/CR-1438 Steam-Water Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-15'<7 Task 4 - Quarterly Progress Report , April 1-June 30,1979.
- Steam-Water Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-1625 Task 4 - Quarterly Progress Report, July 1-September 30, 1979.
Steam-Water Steam-Water Mixing and System Hydrodynamics Program - n'UREG/CR-1657 Task 4. Quarterly Progress Report, October-December 1979, Steel Technology Heavy-Section Steel Technology Program Quarterly NUREG/CR-1477 Progress Report for January-March 1980. F 1 Steels Critical Experiments, Measurements, and Analyses to NUREG/CR-1601' Establish a Crack Arrest Methodology for Nuclear Pressure Vessel Steels. Stiffness Strength and Stiffness of Tensioned Reinfor:ed Concrete NUREG/CR-1601 Panels Subjected to Membrane Shear, Twu-Way Reinforcing. Storage Calculations of the Skyshine Gamma-Ray Dose Rates from NUREG/CR-0723 Independent Spent Fuel Storage Installations (ISFSI) Under Worst Case Accident Conditions. Strategies Financing Strategies for Nuclear Power Plant NUREG/CR-1481 Decommissioning. Streaming Analysis Validation of a Monte Carlo Code for Radiation Streaming NUREG/CR-1334 Analyses. Strength Strength and Stiffness of Tensioned Reinforced Concrete NUREG/CR-1602 Pane?s Subjected to Membrane Shear, Two-Way Reinforcing, r Strong Motion GeotechnieL1 and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 3 Accelerograph Stat. ions, Strong Motion Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 4 Accelerograph Stations. Strong Motion Geotechnical and Strong Metion Earthquake Data from U.S. NUREG/CR-0985, Vol 5 Accelerograph Stations.
Keyword Listing S Report Title Report No. Strong-Motion -- Site-Dependent Ef fects at Strong-Motion Accelerograph . NUREG/CR-1639 Stations. Structural Structural Integrity of Water Reactor Pressure Boundary NURdG/CR-1472 Components, Quarterly Progress Report, January-March ! 1980. Structures Variability of Dynamic Characteristics of Nuclear Power - NUREC/CR-1661' Plant Structures. Subessemblies Compressible Analysis of Inlet Plenum Pressure Rise due NUREG/CR-1592 to Sodium Boiling in Fue! Subassemblies during Pump' O Coast-Down of an I M BR Subcooled Hydrodynamics of r. Vapor , in Subcooled liquid. NUREG/CR-1632 i. Suberitical 04usters Criticality Experiments with Subcritical Clusters of .NUREG/CR-1547 2.35 Wt% and 4.31 Wt% U-235 Enriched UO Rods in Water at a Water-to-Fuel Volume Ratio of 1.6.2 Subsurface Ccud,tions Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055 .V. 2 App Accelerograph Stations in California. Subsurface Conditions Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, Vol 2 Accelerogtaph Stations in California. , Subsurface Conditions Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, VA 3 Accelerograph Stations in California. Summary Event Safeguards Summary Event List (SSEL)/ NUREG-0525, Rev'2 Support Columns Dynamic, Inelastic Buckling Analysis of MARK I Torus NUREC/CR-1038 Support Columns. Suppression Extended Analysis of Data from the 1/5-Scale MARK 1 NUREG/CR-0761 Boiling Water Reactor Pressure Suppression Experiment. Surface Conditions Characterization of Existing Surf ace Conditions at NUREG/CR-1683 Sheffield Low-Level Waste Disposal Facility. Surry Final Environmental Statement Related to Steam Generator NUnEG-0692 Repair at Surry Power Station, Unit No. 1. Virginia Electric and Power Company, Docket No. 50-280. r I
Keyword Listing S Report Title Report No. Surveillance LWR Pressure vessel Irradiation Surveillance Dosimetry NURCC/CR-0720 Quarterly Progress Report, October-December 1978. System Corponents Development and Verification of Fire Tests for Cable NUREG/CR-1552 Systems and System Components. .[ S2D Accident Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris Coolability in TMLB, $2D, and ABG Accidents, t D L ( L
Keyword Listing T Report Title
- Report No. .
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Tailing Files' Measurement'of Radon' Diffusion.from Uranium Mill Tailing NUREC/CR-1109 Piles, r
. Task Force Task Force Report on Interim Operation of Indian Point. NUREG-0715 Task 4 Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-1657 Task 4. Quarterly Progress Report, October-December.1979.
I Technical Activities MARK II Containment Lead Plant Program Load Evaluation and NUREG-0487, Supp 1 Acceptance Criteria - Generic Technical Activities A8 and A39. Technical Activity Control of Heavy Loads at Nuclear Power Plants NUREG-0612 Resolution of Generic Technical Activity A-36. i MARK I containment Long Term Program Safety Evaluation Technical Activity NUREG-0661 Report, Resolution of Generic Technical Activity A-7. February 1977 to December 1979. Technical Issues Technical Safeguards Issues for Altern tve Fuel Cycles.- NUREG/CR-1048 5 Technical Progress Quarterly Technical Progress Report on Water Reactor XUREG/CR-1400 Safety Programs Sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research - April-June 1980. Technical Reports Regulatory and Technical Reports Compilation for 1979, NUREG-0304, Vol '4. Technical Resources Utility Management and Technical Resources. NUREG/CR-1656 , l Technical Specifications Technical Specifications, Sequoyah Nuclear Plant, Unit No. NUREG-0658 Rev 1 1 1. Docket No. 50-327, Appendix "A" to License No. DPR-77. Technical Specifications North Anna Power' Station Unit 2 Technical Specifications NUREG-0664, Rev 1' t Appendix "A" to License No. NPF-7. l Technical Specifications Standard Technical Specifications for Westinghouse NUREG-0452, Rev 3 Pressurized Water Reactors, Revision 3. l Technical Support Considerations on Nuclear Data Link Implementation in NUREC/CR-1579 Relation to the Technical Support Center, Emergency Operations Center and Safety Parameter Display System. I i I
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-Keyword Listing T Report Title Report No. o j i I Techniques Best Estimate Method vs Evaluation Method: A Comparison NUREG/CR-1489 of Two Techniques in Evaluating Seismic Analysim and Design. I Tectonic Seismicity and Tectonic Relationships for Upper Great -NUREG/CR-1569 t Lakes Precambrian Shield Province. - Temperature The Effects of Temperature, Moisture, Concentration, . NUREG-0678 Pressure and Mass Transfer on the Adsorption of Krypton and Xenon on Activated Carbon. Temperatures An Ultrasonic Thermometry System for Measuring Very High NUREG/CR-1488 Temperatures in Reactor Safety Experiments. Tensioned Strength and Stiffness of Tensioned Reinforced Concrete NUREC/CR-1602 - Panels Subjected to Membrane Shear, Two-Way Reinforcing. ! Test Facility Thermocouple Signal Sensitivity to the. Sheath Thickness. NUREG/CR-1347 of Thermal-Hydraulic Test Facility Indirectly Heated Electric Fuel Pin Simulators. Test Results Qualification Test Results on 1550*C and 2200*C 1/16-Inch NUREC/CR-0961 0.D. Fuel Centerline Thermocouples for the LOFT Program. Test Res',lts PBF/ LOFT Lead Rod Test. Series Test Results Report. NUREG/CR-1538 Test Series Gap Conductance Test Series Fuel Characterization Data NUREG/CR-1537 Report. Testing Statistical Analysis of Steam Generator Inspection Plans NUREG/CR-1282 and Eddy Current Testing. v Testing Qualification Testinh Evaluation Program Light-Water NUREG/CR-1492 Reactor Safety Research Quarterly Report, July-September 1979. Testing Performance Testing of Personnel Dosimetry Services: NUREG/CR-1593 Alternatives and Recommendations for a Personnel Dosimetry Testing Program. Thermal Seamary of Thermal Hydraulle Calculations for a NUREG/CM-1480 iressurized Water Reactor. r Thermal-Hydraulic Thermocouple Signal Sensitivity to the Sheath Thickness NUREG/CR-1347 of Thermal-Hydraulic Test Facility Indirectly Heated Electric Fuel Pin Simulators.
Keyword Listing T Report Title Report No. , Thermocouple Thermocouple Signal Sensitivity to the Sheath Thicknes's NUREC/CR-1347 of Thermal-Hydraulic Test facility Indirectly Heated' r Electric Fuel Pin Simulators.
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t Thermocouples Qualification Test Results on 1550'C'and 2200*C 1/16-Inch . >RREC/CR-0961 0.D.~ Fuel Centerline Thermocouples for the LOFT Program. s Thermome'.ry System An Ultrasonic Thermometry System for Measuring Very High' -NUREG/CR-1488l Temperatures in Reactor Safety Experiments. Thorium Health Status and Body Radioactivity of Former Thorium NUREC/CR-1420 . Workers. i Threat Potential Threat to Licensed Nuclear Activities from KUREG-0703 Insiders (Insider Study). P Three Mile Island NRC Plans for Cleanup Operations at Three Mile Island KUREG-0698 Unit 2. Three Mile Island Answers to Frequently Asked Questions about Cleanup NUREG-0732 Activities at Three Mile Island, Unit 2. THI-2 NRC Action Pl.an Developed as a Result of the TMI-2 KUREG-0660, Vol i Accident, Revtsion 1, Vol 1. , J THI 2 NRC Action Plan Development as a Result of the TMI-2 NUREG-0660, Vol 2 Accident, Revision 1, Vol. 2. THI-2 NRC Plans for Cleanup Operations at Three Mile Island NUREG-0698 Unit 2. l TMI-2 Answers to Frequently Asked Questions about Cleanup NUREC-0732 Activities at Three Mile Island, Unit 2. THI-2 Psychological Stress for Alternatives of Decontamination NUREC/CR-1584 of TMI-2 Reactor Building Atmosphere. TMLB Accident Assessment of Core Penetration of a PWR Reactor Vessel NUREG/CR-1518 and Particulate Debris Coolability in TMLB, S2D, and ABG Accidents. Topography flow Topography Instrumentation and Analysis System. NUREG/CH-1333 F i l l
Keyword Listing T Report Title Report No. Tornado Modeling Tornado Dynamics. NUREG/CR-1585 Torus Dynamic, Inelastic Buckling Analysis of MARK 1 Torus NUREG/CR-1038 Support Columns. Toxicity Acute Toxicity and Bioaccumulation of Chloroform to Four NUREG/CR-0893 Species of Freshwater Fish. Tracer Experiments Diffusion Near Buildings as Determined from Atmospheric KUREC/CR-1394 - Tracer Experiments. l Training Nuclear Power Plant Simulators: Their Use in operator NUREG/CR-1482 l Training and Requalification. Transient Experiment Data Report for LOFT Anticipated Transient KUREC/CR-1520 Experiment L6-5. Transient Analysis Transient Analysis of Coolant Flow and Heat Transfer NUREG/CR-1404 in LMFBR Piping Systems. Transport A Deterministic-Probabilistic Model for Contaminant NUREG/CR-1609 Transport. i Transport Model Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1377 Waste: Transport Model Sensitivity Analysis. Transportation Review and Assessment of Package Requirements (Yellowcake) NUREG-0535 and Emergency Response to Transportation Accidents. Transportation Review and Integration of Existing Literature Concerning NUREG/CR-0742 Potential Social Impacts of Transportation of Radioactive Materials in Urban Areas. Transportation Identification and Assessment of the Social Impacts of KUREG/C9-0744 Transportation of Radioactive Materials in Urban Environments. Transportation Survey of Current State Radiological Emergency Response NUREC/CR-1620 Capabilities for Transportation Related incidents. Transportation Transportation of Radioactive Material in Kentucky. EUREG/CR-1671
Keyword Listin_g_T Report Title Report No. Tritium Measurement of KE-133, C-14 and Tritium in Air and 1-131 NUREG/CR-1195 in Vegetation and Milk Around the Quad Cities Nuclear , Power Station. Tubing Program Eddy-Current Inspection for Steam Generator Tubing Program. NUREG/CR-1563 Annual Progress Report for Period Ending December 31, i 1979. Turbulence Wind-Tunnel Metiurements of Dispersion and Turbulence NUREG/CR-1475 in the Wakes of Nuclear Reactor Plants. Turbulent On the Motions of Particles in Turbulent Flows. KUREG/CR-1554 r Turkey Point in-Plant Source Term Measurements at Turkey Point NUREG/CR-1629 Station - Units 3 sud 4. TVA Safety Evaluation Report Related to Operation of Sequoyah KUREG-00ll, Supp. 2 Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328. Tennessee Valley Authority, Supp. No. 2. TVA Safety Evaluation Report Related to Operation of Sequoyah KUREG-00ll, Supp. 3 Nuclear Plant, Units 1 and 2, Docket Nos. 50-327/328. Tennessee Valley Authority, Supp. 3. Two-Phase Two-Phase Flow Measurements with Advanced Instrumented NUREG/CR-1529 Spool Pieces. Two-Phase Film Entrainment and Drop Deposition for Two-Phase Flow, EREG/CR-1634 Two-Way Reinforcing Strength and Stiffness of Tensioned Reinforced Concrete RUREC/CR-1602 9 Panels Subjected to Membrane Shear, Two-Way Reinforcing. '
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r Keyword Listing _U . Report Title Report No. U.S. Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 3 + Accelerogrpph Stations, r
. Geotechnical and Strong Motion Earthquake Data from U.S. NUREG/CR-0985, Vol 4 U.S.
Accelerograph Stations. Geotechnical'and Strong Motion Earthquake' Data from U.S. NUREG/CR-0985, Vol 5 U.S. Accelerograph Stations. i U.S. ' Data Summaries of Licensee Event Reports of Primary NUREG/CR-1730 Containment Penetrations at U.S. Commercial Nuclear Power Plants from January 1,1972 to December 31, 1978. Ultrasonic An Ultrasonic Thermometry System for Measuring Very High NUREG/CP-1488 Temperatures in Reactor Safety Experiments. Uncertainty Analysis Uncertainty Analysis for a PWR Loss-of-Coolant Accident: NUREG/CR-1364
- 11. Alternative Core Damage Estimators. i r
Uniform Piping Benchmark Problems. Dynamic Analysis Uniform NUREG/CR-1677: Vol 1 , Support Motion Response Spectrum Method. Union Carbide-Co. Final Environmental Statement Related to the Operation of NUREG-0702 Gas Hills Uranium Project, Docket No. 40-299, Union Carbide Corporation, i United States United States Nuclear Regulatory Commission Staff NUREG-0386, Supp. 2 , Practice & Procedure Digest. Supplement 2 to Digest No. 2. e
- United States Licensability of CANDU-Type Reactors in the United States. NUREG/CR-Ill3 A Preliminary Assessment of the R and D Requirements.
United States Material Accounting as Required by the United States NUREG/CR-Il92 Nuclear Regulatory Commission: Capabilities and Vulnerabilities. i United States Seasonal Variation of 10-Square Mile Probable Maximum NUREG/CR-1486 Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. 53). Upgrade Acceptance criteria for the Physical Protection Upgrade NUREG-0721 Rule Requirements for Fixed Sites. Uranium Preparation of Working Reference Materials: Calcined NUREG/CR-1445 Waste Recovery Products Containing Uranium or Plutonium. r
Keyword Listing _U Report Title Report No. t Uranium A Methodology and a Preliminary Data Base for Examining NUREG/CR-1539 the Health Risks of Electricity Generation from Uranium and Coal Fuels. Uranium Carbide Prompt Burst Energetics Experiments: Fresh Uranium NUREG/CR-1396 Carbide / Sodium Series. I Uranium Content Estimates of Uranium Content and Radon Flux for Uranium NUREG/CR-1549 Mine Dumps Based on Borehold Radioactivity Logs. Uranium Hexafluoride Acute Ef fects of Inhalation Exposure to Uranium Hexafluoride NUREG/CR-1045 and Patterns of Deposition. UF6/UO 2 2F Studies in Experimental Animals. Uranium Mill Measurement of Radon Diffusion from Uranium Mill Tailing NUREG/CR-Il09' Piles. Uranium Milling Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 1 Milling Project M-25: Volume 1 - Summary and Text. Uranium Milling Final Generie Environmental Impact Statement on Uranium NUREG-0706, Vol 2 Milling Project M-25: Volume II - Appendices A-F. Uranium Milling Final Generic Environmental Impact Statement on Uranium NUREG-0706, Vol 3 Milling Project M-25: Volume III - Appendices G-V. Uranium Mine Estimates of Uranium Content and Radon Flux for Uranium KUREG/CR-1549 Mine Dumps Based on Borehold Radioactivity Logs. " i e ' Uranium Mine Radon Release and Dispersion from an Open Pit Uranium KUREG/CR-1583 Mine. > Uranium Oxide Critical Experiments with Interstitially-Moderated NUREG/CR-1071 Arrays of Low-Enriched Uranium Oxide. Topical Report on Reference Critical Experiments. Uranium Products Solubility Classification of Airborne Uranium Products MUREG/CR-1428 . from LWR-Fuel Plants. ! Uranium Project Final Environmectal Statement Related to the Operation MUREG-0702 of Gas Hills Uranium Project, Docket No. 40~299, Union Carbide Corporation. Uranium-235 Criticality Experiments with Suberitical Clusters of NUREG/CR-1547 2.35 Wt% and 4.31 Wt% U-235 Enriched UO Roda in Water at a Water-to-Fuel Volume Ratio of 1.6.2 i
Report No. j Keyword Listing U Report Title
-)
- Review and Integration of Existing Literature Concerning NUREG/CR-0742~
Urban Areas Potential Social Impacts of Transportation of Pedioactive Materials in Urban Areas. Urban Environments Identification and Assessment of the Social Impacts of KUREG/CR-0744 Transportation of Radioactive Materials in Urban Environments. Report on Nuclear Industry Quality Assurance Procedures NUREG-0653 Use for Safety Analysis Computer Code Development .and Use. Use Nuclear Power Plant Simulators: Their Use in Operator KUREG/CR-1482 Training and Requalification, The Use of Process Monitoring Data for Nuclear Material NUREG/F'-1670, Vol 1 Use Accounting: Volume 1. Summary Report. A User's Guide to EPIC, a Computer Program to Calculate NUREG/CR-1504 User's Guide the Motion of Fuel and Coolant Subsequent to Pin Failure . in a LMFBR. User's Manual User's Manual for USINT: A Program for Calculating KUREG/CR-1375 Heat and Mass Transfer in Concrete Subjected to High Heat Fluxes. Safeguards User's Manual for Nuclear Materials Management NUREC/CR-1528 User's Manual and Safeguards System. User's Manual for USINT: A Program for Calculating """IG/CR-1375 USINT Heat and Mass Transfer in Concrete Subject to High Heat Fluxes. Utility Management and Technical Resources. KUREG/CR-1656 Utility Management O
i Keivord Listing V Report Title Report No. Validation Validation of a Monte Carlo Code for Radiation Streaming NUREG/CR-1334 Analyses. ( Vapor Jet Hydrodynamics of a Vapor Jet in Subcooled Liquid. NUREG/CR-1632 Variability Variability of Dynamic Characteristics of Nuclear Power NUREG/CR-1661 - Plant Structures. , Variation Seasonal Variation of 10-Square Mlle Probable Maximum NUREG/CR-1486 Precipitation Estimates, United States East of the
- 105th Meridian (Hydrometeorological Report No. 53).
Vegetation Measurement of KE-133, C-14 and Tritium in Air and 1-131 NUREG/CR-1195 in Vegetation and Milk Around the Quad Cities Nuclear Power Station. Vegetational Cover Vegetational Cover in Monitoring and Stabilization of NUREG/CR-1358 Shallow Land Burial Sites. Vent Pipe Lateral Loads on Vent Pipe in Steam Chugging. NUREG/CR 1631 Verification Verificatior, of Subsurface Conditions at Selected " Rock" NUREG/CR-0055. V. 2 App Accelerograph Stations in California. Verification Verification of Subsurface Conditions at Selected " Rock" NUREG/CR-0055, Vol 2 Accelerograph Stations in California. Verification Verification of Subsuctace Conditions at Selected " Rock" NUREC/CR-0055, Vol 3 Accelerograph Stations in California. Verification Development and Verification of Fire Tests for Cable NUREG/CR-1552 Systems and System Components. Vibration Dynamic Analysis to Establish Normal Shock and Vibration NUREG/CR-1484 of Radioactive Material Shipping Packages - Quarterly Progress Report October 1-December 31, 1979. Virginia Elec. & Power Final Environmental Statement Related to Steam Generator NURG-0692 Repair at Surry Power Station, Unit No. 1. Virginia Electric and Power Capany, Docket No. 50-280 Virginia Elect. & Power Final Environmental Statement Related to the Operation of NUREG-0134, Add. 2 North Anna Power Station, Unit 1 and 2, Docket No. 50-338 and 50-339. Virginia Electric and Power Company. j l l 1 I l
i Keyword Lieting V Report Title Report No. Virginia Electric & Power Safety Evaluation Report Related to the Operation of NUREG-0053, Supp. 11 North Anns Power Station, Unit 2, Virginia Electric and Power Company, Docket No. 50-339. Supplement No. 11. Vital Area Vital Area Analysis Using SETS. NUREC/CR-1487 Volume Heated Heat Removal Characteristics of Volume Heated Boiling NUREG/CR-1357 Pools with Inclined Boundaries. Vulnerabilities Material Accounting as Required by the United States NUREG/CF-1192 Nuclear Regulatory Commission: Capabilities and Vulnerabilities, o e
. - .. . - ~ , . .- . - . . .. - ~ . ~ . ~ - , Keyword Listing W Report Titje fReport No. , Wakes' Wind-Tunnel Measurements of Dispersion and Turbulence NUREG/CR-1475 'f in the Wakes of Nuclear Reactor Plants. , t Washington The Effects of Natural Phenomena'on the Exxon Nuclear KUREG-0722 1 Company Mixed Oxide Fabrication Plant at Richland, } Washington. Waste Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1262 i Waste A Distribution-Free Approach to Inducing Rank' ! Correlation Among Input Variables for Simulation Studies. . Vaste Risk Methodology for Geologic Disposal of Radioactive NUREG/CR-1377 l Wastes Transpart Model Sensitivity Asalysis. t r Waste Properties of Radiocetive Wastes and Waste Containers. NUREG/CR-1514 e
- Quarterly Progress Report, January-March 1980. l e
f Waste Containers- Properties of Radioactive Wastes and Waste Containersi NUREG/CR-1514 . i Quarte 1y Progress Report, January-March 1980, t f
~
Waste Disposal Characterization of Existing Surface Conditions at NUREG/CR-1683 Sheffield Low-Level Waste Disposal Facility. 4 e Waste Recovery Preparation of Working Reference Materials: Calcined NUREG/CR-1445 Waste Recovery Products Containing Uranium or. Plutonium. Water Chemistry ' Evalt ation of 1sotope Migration-Land Burial Water Chemistry KUREG/CR-1289 at Commercially Operated Low Level Radioactive Waste Disposal Sites. '; [ Water Chemistry Evaluation of Isotope Migration-Land Purial Water NUREG/CR-1513 Chemistry at Commercially Operated Low-Level Radioactive - Waste Disposal' Sites. Quarterly Progress Report, January-March 1980. Water Hammer An Evaluation of Condensation-Induced Water Hammer in NUREG/CR-1606 Preheat Steam Generators. i Water Mixing Steam-Water Mixing and System Hydrodynamics Program - NUREG/CR-1557 Task 4 - Quarterly Progress Report, April 1-June 30, 1979. , l Water Repetor Quarterly Technical Progress Report on Water Reactor NUREG/CR-1400- , Safety Programs Sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research - > April-June 1980. [ Water Reactor Water Reactor Safety Research Division Quarterly .NUREG/CR-1403 Progress Report. October-December 1979. 5
Keyword Listing W Rrport Title Rrport No. Water Reactor Structural Integrity of Water Reactor Presbure Boundary . NUREG/CR-1472 Components. Quarterly Progress Report, January-March 1980. Water-to-Fuel Criticality Experiments with Suberitical Clusters of NUREG/CR-1547 2.35 Wt1 and 4.31 Wt% U-235 ?,nriched UO Rods in Water at a Water-to-Fuel Volume Ratio of 1.6.2 West Virginia A Probabilistic Evaluation of Earthquake Detection and NUREG/CR-1648-Location Capability for Illinois, Indisas, Kentucky, Ohio, and West Virginia. Westinghouse Standard Technical Specifications for Westinghouse NUREG-0452, Rev 3 Pressurized Water Reactors, Revision 3.
' Wind Tunnel An Algorithm to Estimate Field Concentrations Under NUREG/CR-1474 Nonsteady Meteozelogical Conditions from Wind Tunnel Experiments.
Wind-Tunnel Wind-Tunnel Measurements of Dispersion and Turbulence NUREG/CR-1475 in the Wakes of Nuclear Reactor Plants. Workers 2talth Status and Body Radioactivity of Former Thorium NUREG/CR-1420 Workers. P Worst Case Calculations of the Skyshine Gamma-Ray Dose Rates from NUREG/CR-0723 Independent Spent Fuel Storage Installations (ISFSI) Under Worst Case Accident Conditions. i i I 1 l l l
. , . ~ . . .. . .- -
Report No. '! Keyword Listing X Report Title The Effects of Temperature, Moisture, Concentration, NUREG-0678- i Xenon Pressure and Mass Tran6fer on the Adsorption of Krypton l and Xenon on Activated Carbon. ! i NUREG/CR-1195 ; Xenon-133 Measurement of XE-133, C-14 and Tritium in Air and I-131 in Vegetation and Milk Around the Quad cities Nuclear > Power Station. > b I P Y f 6 L 9 1
Kryword Listing Y R7 port Title Repott No. i Yellowcake Review and Assessment of Package Requirements (Yellowcake) NUREG-0535 ' and Emergency Response to Transportation Accidents. i 1 P I i 1 i l
Keyeard Listing Z R ort Title y Report No. Zion Report of the Zion / Indian Point Study: Volume 1. hTREG/CR-1410 l Zion Report of the Zion / Indian Point Study. NUREG/CR-1411, Vol 1 4 Zion Report of the Zion / Indian Point Study. NUREG/CK-1411 Vol 2 f Zircaloy-Hydrogen Instrumented Impact Froperties of Zirealoy-Oxygen and NUREG/CR-1408 Zircaloy-Hydrogen Alloys. 1 , I i Zi rcalcy-Oxygen Instrumented Isrpact Properties of Zircaloy-Oxygen and. NUREG/CR-1408 Zircaloy-Hydrogen Alloys, i t
~
s p i e b I i i r t b f E
% I
, ,~ . - - - - - -. - ~ . . - - - I Numbeted Keywords Report Title Report No. 10 CFR Pt 50 Euvironmental Assessment for Effective Changes to 10 CFR NUREG-0685 ) Part 50 and Appendix E to 10 CFR Part 50; Emergency 4 Planning Requirements for Nuclear Power Plants. ) 1 105th Meridian Seasonal Variation of 10-Square Mile Probable Maximum KUREC/CR-1486-Precipitation Estimates, United States East of the 105th Meridian (Hydrometeorological Report No. 53). 1955-1979 .Geotechnical Data from Accelerograph S ations NUREC/CR-1643 .' Investigated during the Period 1975-1979. Summa ry Report. ! i 1979 Regulatory and Technical Reporta Compilation for 1979. NUREG-0304, Vol. 4 3 i s 1979 Nuclear Plant Reliability Data System 1979 Annual Reports NUREG/CR-1635 of Cumulative System and Component Reliability. 4 1982 Comments on the NRC Safety Research Program Budget for NUREG-0699 Fiscal Year 1982. G i n b
Cross-Reference from Contractor Report Numbers to NUREG/CR Numbers (This section contains an alpha-numerically arranged listing of contractor report numbers cross-referenced to their corresponding NRC report numbers.) E
- . .- -, . . . ~
1 i ANL-80 NUREG/CR-1408 IHEDL-TME 80-24 -NUREG/CR-1484.- ,
- -ANL-80-37 NUREG/CR-1420- HLA-9906-001-14 NUREG/CR-1683 l
ANL-80-47 NUREG/CR 1504' .LA-NUREG-6623, Supp 3 "hUREG/CR-1536 ANI 80-48 NUREG/CR-1592 LA-UR80-45-15' NUREG/CR-1471 P ANL-80-54, Vol 1 NUREG/CR-1526, Vol 1 LA-7199-MS, Add 1 NUREG/CR-1185 ANL/ES-97 NUREG/CR 1583 LA-7515-M, Rev NUREG/CR-0453, Rev l' ARAP Rpt 421 NUREG/CR-1585 LA-8299-PR NUREG/CR-1516 b BMI-2038 NUREC/CR-1!57 LA-8306-MS NUREG/CR-1411, Vol 1 BMI-2055 NUREG/CR-1601 LA-8506-MS NUREG/CR-1411, Vol '2 BMT-2062 NUREG/CR-1657 LA-8348 NUREG/CR-1445 BNL-NUREG-50950 NUREG/CR-0603 LA-8361-MS' NUREG/CR-1480-BNL+NUREG-51143 NUREC/CR-1289 LA-8423-MS NUREG/CR-1564' BNL-NUREG-51151 NUREG/CR-1355 LA-8429-MS NUREG/CR-1575 h BNL-NUREG-51156 NUREG/CR-1356 LA-8432-MS NUREG/CR-1586 BNL-NUREG-51157 NUREG/CR-1357 LA-8449-MS NUREG/CR-1607 BNL-NUREG-51177 NUREG/CR-1402 LA-8475-MS NUREG/CR-1634 BNL-NUREG-51178 NUREG/CR-1403 MR-7068 NUREG/CR-1334 BNL-NUREG-51179 NUREG/CR-1404 NOAA Tech Memo ERL ARL-84 NUREG/CR-1394 BNL-NUREG-51180 NUREG/CR-1405 NRL Memo Rpt 4254 NUP.EG/CR-1472 i BNL-NUREG-51182 NUREG/CR-1048 NRL Memo Rpt 4259 NUREG/CR-1119 , BNL-NUREG-51186 NUREG/CR-1438 NUSAC-556 NUREG/CR-1676, Vol 1 BNL-NUREG-51217 NUREG/CR-1505 ORNL/NUREG-62 NUREG/CR-1347 } BNL-NUREG-51219 NUREC/CR-1513 ORNL/NUREG-63 NUREG/CR-1295 i BNL-NUREG-51220 NUREG/CR-1514 ORNL/NUREG-72 NUREG/CR-1529 BNL-NUREG-51248 NUREG/CR-1606 ORNL/NUREG/CSD-2, Vol 1 NUREG/CR-0200 BNL-NUREG-51267 NUREC/CR-1677, Vol 1 ORNL/NUREG/TM-316 NUREC/CR-0723 CGS /NR85F060 NUREG/CR-1608 ORNL/NUREG/TM-390 NUREG/CR-1548 CGS /NR85UO60 NUREG/CR-1609 ORNL/NUREG/TM-392 NUREG/CR-1450 EGG-EA-5188 NUREC/CR-1730 ORNL/NUREG/TM-393 NUREG/CR-1477 EGG-2045 NUREG/CR-1520 ORNL/NUREG/TM-395 NUREG/rR-1482 EGG-2046 NUREG/CR-1537 ORNL/NUREG/TM-397 NUREG/CR-1521 EGG-2047 NUREC/CR-1538 ORNL/NUREG/TM-398 NUREC/CR-!M3 EGG-2043 NUREG/CR-1400 ORNL/Sub-7615/1 NUREC/CR-1539 ENICO-1023 NUREG/CR-1195 PNL-3040-3 NUREG/CR-1009 HEDL-TME 79-18 NUREG/CR-0720 PNL-3040-4 NUREG/CR-1349 HEDL-TME 79-50 NUREG/CR-0961 PNL-3046 NUREG/CR-0893
PNL-3158- NUREG/CR-1298-- UCLA-ENG-7853 .NUREC/CK-1113. PNL-3187 NUREG/CR-1109 UCRL-15227 NUREG/CR-1660 PNL-3278 NUREG/CR-1302 UCRL-15267 6 NUREC/CR-l'61 PNL-3314 .NUREG/CR-1547 UCRL-52707 NUREC/CR-0761 PNL-3325 NUREG/CR-1380, Vol 1, ES UCRL-52715-80-1 NUREG/CR-1485 Vol 1, No. 1 , PNL-3396 NUREG/CR-1670, Vol 1 LUCRL-52734 .NUREG/CR-1192 PNL-3411 NUREG/CR-1428 UCRL-52745 NUREG/CR-1457'
'RFP-3008 NUREG/CH-1071 UCRL-52746 NUREG/CR-1489 -SAND 78-7017 NUREG/CR-0742 UL-USNC 75 Q NUREC/CR-1552 SAND 79-0621 WURIG/CR-1488 URCL-52723 NUREG/CR-1038 SAND 79-1561 NUREG/CR-1466 Y/DW-128 ~ NUREG/CR-0508 SAND 79-1694 NUREG/CR-1375
[ SAND 79-1787 NUREG/CR-1184 - SAND 79-2168 NUREG/CR-1277 SAND 79-2242 NUREG/CR-1308, Vol 2-SAND 79-2300 NUREG/CR-1282 SAND 79-2378/l NUREG/CR-1198, Vol 1 SAND 79-23hs/2 NUREG/CR-1198, Vol 2 SAND 79-7032 NUREG/CR-0744 SAND 80-0157 NUREG/CR-1262 SAND 80-0505 NUREG/CR-1590 SAND 80-0617/1 NUREG/CR-1410 SAND 80-0629 NUREC/CR-1364 SAND 80-0644 NUREG/CR-1377 SAND 80-0701 NUREG/CR-1518 SAND 80-0820 NUREG/CR-1396 , SAND 80-1006 NUREG/CR-1448 SAND 80-1095 NUREC/CR-1487 FAND80-1117 NUREG/CR-1492 SANDBO-1304/1 of 4 NUREG/CR-1509 SAND 80-1618 NUREG/CR-1579 SSS-R-80-4217 NUREG/CR-1340 TN-314 NUREC/CR-1333 UCID-18123-80-1 NUREG/CR-1610, Vol 1, No. 1 UCID-18674 NUREC/CR-1624 UCLA 12-1235 NUREG/CR-1358
NRC roRu 335 U S. NUCLEAR REGULATORY COMMIS$ TON (7 7 " - NUREG-0304, Vol . S, No. 3 BIBLIOGRAPHIC DATA SHEET 4 TSTLE AND SUBTtiLE LAdd Volume Na. of worconaw) 2. (Leave blek} Regulatory and Technical Reports Compilation for July-September 1980 3 RECIPIENTS ACCESSION NO. . 7 ^"'" "* 5. ATE REPORT CWPLETED U. McKinney, Indexer W.E. Oliu and L. McKenzie, Compiler
.h nuwy I j *$"
- 9. PERFORMING ORGANIZATION N AME AND MAtuNG ADDRESS (Include le Coarl DATE REPORT ISSUED Division of Technical Information and Document Control "N" lYEAR '
Office of Administration January 1981 U.S. Nuclear Regulatory Connission 6 It * *e *** > Bashington, DC 20555 8 ILeave w r *1
- 12. SPONSORING ORG ANIZ ATION N AME AND M AluNG ADDRESS //rtetum Ic Com/ ,
p , Same as 9 11 CONTRAcr NO. 13 TYPE OF BEPORT PE RIOD "r3V'. RE D //nctupwe dJes) Bibliographical July-September 196Q
- 15. SUPPLEMENTARY NOTES 14 (Leere aral
- 16. ABSTR ACT 20G w wds or sen)
This compilation lists all NRC regulatory and technical reports published under the NUREG series during the third quarter of 1980. 17 AEY Yv0RDS AND DOCUMENT AN ALYSfs 174 Ot SC m P T O RS 17b IDENTIFtE RS OPEN ENDE D TE RMS 18 AV AIL ABILITY ST ATE MENT (9 SECu4sTV CLASS tre s recrues 21. N O OS P A Gt s Unclassified Unlimited eg . , sSirrap ei 22 emcE NncFORM '35 s1 in
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