ML20073B363

Computer Codes and Mathematical MODELS.January-December 1993

ML20073B363
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Issue date:	08/31/1994
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NUREG-BR-0083, NUREG-BR-0083-V09, NUREG-BR-83, NUREG-BR-83-V9, NUDOCS 9409210320
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{{#Wiki_filter:l NUREG/Pi.-0083 Volume 9 f** ""%, -United States l Nuclear Regulatory Commission '%,,,,/ Computer Codes and Mathematical Models January-December 1993 August 1994 End User Support Services Branch Omce ofInformation Resources Management 9409210320 940831 B -000 PDR

NUREG/BR-0083 l Volume 9 ..u, United States f 1 Nuclear Regulatory Commission \\,, / Computer Codes and Mathematical Models January-December 1993 Energy Science and Tbchnology Software Center P.O. Box 1020 Oak Ridge, TN 37831-1020 August 1994 End-User Support Services Branch Office ofInformation Resources Management US. Nuclear Regulatory Commission Washington, DC 20555-0001 i o

ABSTRACT This report contains citations of NUREG-series documents issued in calendar year 1993 relating to computer software and mathematical models for scientific, engineering, or technology-related programs performed or sponsored by the U.S. Nuclear Regulatory Commission (NRC). It is intended as a reference tool to assist the scientific and technical analyst in obtaining information on NRC computer-related activities. iii NUREG/BR-@83, VoL9 .-----------3

TABLE OF CONTENTS Page ABSTRACT iii 7 FOREWORD vii CITATIONS 1 APPENDIX A: Index by NUREG-Series Report Number A-1 APPENDIX B: Index by Software Identification B-1 APPENDIX C: Index by Contractor Report Number C-1 APPENDIX D:Index by Keyword D-1 I J i i l l i v NUREG/BR--0083, Vol.9

i FOREWORD The citations in this document appear in NUREG-series document order. Citations of h1C staff-generated reports designated NUREG-xxxx are listed first, followed by any conference proceedings identified as NUREG/CP-xxxx, contractor-generated reports published as NUREG/CR--xxxx docu-ments, grant repons published as NUREG/GR-xxxx documents, and International Agreement repons issued as NUREG/IA-xxxx publications. Each citation contains the following NUREG-series repon number; software identificadon (where applicable); contractor repon number; report title; a description of the repon contents; publication date; names of the individuals responsible for preparing, compiling, or editing the repon; contractor name and location; sponsoring NRC organizatior.; and keywords or descriptors. Indexes by NUREG-series report number,softwareidentification, contractorrepon number, and keyword are included in the Appendixes. Specific code names and software identification appear in the heading of those citations with primary emphasis on specific mathematical models, computer codes, or databases.The term " General" is used in the heading of those citations which contain significant information on many models, computer codes, or databases. i j i 1 l i vii NUREG/BR--0083, Vol.9

NUREG-0713-Vol.12 i REIRS

Title:

Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities,1990. Twenty-Third Annual Report. Volume 12

== Description:== This report summarizes the occupational radiation exposure informadon that has been reported to the US Nuclear Regulatory Commission's (NRC's) Radiation Exposure Informa-tion Reporting System (REIRS) by nuclear power facilities and cenain other categories of NRC licensees during the years 1969 through 1990. He bulk of the data presented in the report was obtained from annual radiation exposure repons submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuctcarpowerplants. Da'a on workers terminating their employment at certain NRC licensed facilides were obtained from reports submitted pursuant to 10 CFR 20.408. He 1990 annual reports submitted by about 443 licensees indicated that approximately 214,568 individuals were monitored, 110,2M of whom were monitored by nuclear power facilities. They incuned an average individual dose of 0.19 rem (cSv) and an average measurable dose of about 0.36 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,361 individu-als completed their employment with one or more of the 443 covered licensees during 1990. Some 77,633 of these individuals terminated from power reactor facilities, and about 11,083 of them were considered to be transient workers who received an average dose of 0.67 rem (cSv). Publication Date: January 1993 1 Prepared by: Raddatz, C.T. [ Nuclear Regulatory Commission, Washington, DC (United States) Div. of Regulatory Applicadons]; Hagemeyer, D. [ Science Applications International Corp., Oak Ridge,TN (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications: Science Applications International Corp., Oak Ridge,TN (United States) Keywords: compiled data, industrial radiography. nuclear facilities. nuclear power plants. occupadonal exposure, personnel, progress report, radiation dose distributions, radiation doses, REIRS, whole-body irradiation 1 NUREG/BR-4)083, Vol.9 L

NUREG--0713-Vol.13 REIRS

Title:

Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities,1991. Twenty-Fourth Annual Report. Volume 13

== Description:== This report summarizes the occupational radiation exposure information that has been reported to the US N uclearRegulatory Commission's (NRC's) Radiation ExposureInforma-tion Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1991.The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications ofnuclear powerplants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1991 annual reports submitted by about 436 licensees indicated that approximately 206,732 individuals were monitored, 182,334 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.15 rem (cSv) and an average measurable dose of about 0.31 (cSv). Termination radiation exposure reports were analyzedto reveal that about96,231 individuals completed their employment with one or more of the 436 covered licensees during 1991. Some 68,115 of theseindividuals terminated from power reactor facilities, and about 7,763 of them were considered to be transient workers who received an average dose of 0.52 rem (cSv). Publication Date: July 1993 Prepared by: Raddatz, C.T. [ Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications];Hagemeyer,D.[ Science ApplicationsInternationalCorp., Oak Ridge,TN (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory - Applications; Science Applications Intemational Corp., Oak Ridge, TN (United States) Keywords: compiled data, industrial radiography, nuclear facilities. nuclear power plants. occupational exposure, personnel, progress report, radiation doses, REIRS, reporting requirements NUREG/BR--0083, Vol.'/ 2

NUREG--0713-Vol.14 REIRS

Title:

Occupational Radiation Exposure at Commemial Nuclear Power Reactors and Other Facilities 1992. Twenty-Fifth Annual Report. Volume 14 i

== Description:== This report summarizes the occupational radiation exposure information that has been - ) reponed to the US Nuclear Regulatory Commission's (NRC's) Radiation ExposureInforma-tion Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. *Ihe bulk of the data presented in the repon was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv). i Publication Date: December 1993 Prepared by: Raddatz, C.T. [US Nuclear Regulatory. Commission, Washington, DC (United States). Division of Regulatory Applications]; Hagemeyer, D. [ Science Applications International j Corp., Oak Ridge, TN (United States)] i Prepared for: - Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications Keywords: commercial sector. data analysis, database management, nuclear power plants. occupational exposure, radiation doses, radiation protection, REIRS, statistical data t 1 i 3 NUREG/BR--0083, Vol.9 i

_ _ ~ _ -._ NUREG-1473 EDSFI

Title:

Electrical Distribution System Functional Inspection (EDSFI) Data Base Prog am

== Description:== This document describes the organization, installation procedures,and operating instructions I for the database computer program containing inspection findings from the US Nuclear Regulatory Commission's (NRC's) Electrical Distribution System Functional Inspections i (EDSFIs). The program enables the user to search and sort findings, ascertain trends, and obtain printed reports of the findings. The findings include observations, unresolved issues, or possible deficiencies in the design and implementation of electrical distribution systems in nuclear plants. This database will assist those preparing for electrical inspections, searching for deficiencies in a plant, and determining the corrective actions previously taken i for similar deficiencies. This database will be updated as new EDSFIs are completed. Publication Date: January 1993 Prepared by: Gautam, A. [ Nuclear Regulatory Commission, Washington, DC (United States), Div. of Reactor Inspection and Licensee Performance] Prepared for: NuclearRegulatoryCommission, Washington,DC(UnitedStates),Div ofReactorInspec-tion and LicenseePerformance Keywords: computer program documentation, EDSFI, electrical equipment, inspection, installation, nuclear power plants, operation, power distribution systems, reactor safety, US NRC I r P k i ? i b s NUREG/BR--0083, Vol.9 4

y-NUREG/CP--0040 - VTOUGH, TOUGH i

Title:

Proceedings of Workshop 5: Flow and Transport Through Unsaturated Fractured Rock Related to High-Level Radioactive Waste Disposal

== Description:== The Workshop on Flow and TransportThrough Unsaturated Fractured Rock Related to High-IevelRadioactive Waste Disposal was cosponsored by the US Nuclear Regulatory Commis-sion (NRC), the Center for Nuclear Waste Regulatory Analyses, and the University of Arizona (UAZ) and was held in Tucson, Arizona, on January 7-10,1991. He focus of this workshop, similar to the earlier four (the first being in 1982), related to hydrogeologic - technical issues associated with possible disposal of commercial high-level nuclear waste (HLW)in a geologic repository within an unsaturated fractured rock system which coincides L with the UAZ field studies on HLW disposal.The presentations and discussions centered on - flow and transport pmcesses and conditions, relevant parameters, as well as state-of-the-art j measurement techniques and modeling capabilities.ne workshop consisted of four half-day i technical meetings, a one-day field vis't to the Apache leap test site to review ongoing field studies that are examining site characterization techniques and developing data sets formodel validation studies, and a final half-day session devoted to examining research needs related I to modeling groundwater flow and radionuclide transport in unsaturated, fractured rock. nese proceedings provide extended abstracts of the technical presentations and short summaries of the research group reports. Publication Date: June 1993 Prepared by: Evara,D.D. [ed.] [ Arizona Univ., Tucson, AZ (United States), Dept. of Hydrology and Water Resources]; Nicholson, TJ. [ed.] [ Nuclear Regulatory Commission, Washington, DC-(United States), Div. of Regulatory Applications] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications: Southwest ResearchInst.. San Antonio,TX (United States), Center forNuclear Waste Regulatory Analyses: University of Arizona, Tucso,n, AZ (United States), Dept. of Hydrology and WaterResources Keywords: Darcy's taw, Equivalent Continuum Model (ECM), fluid flow, fractured reservoirs. ground-water, hydrology, mathematical models, meetings, Monte Carlo method, proceedings, radioactive waste disposal, radionuclide migration. TOUGH, VTOUGH j i 5 NUREG/BR--0083, Vol.9 e

i NUREG/CP--0126-Vol.1

General-

-l ,~

Title:

Proceedings of the US Nuclear Regulatory Commission Twentieth Water Reactor Safety - ' Information Meeting. Volume 1, Plenary Session Advanced Reactor Research, Advanced. Passive LWRs, Advanced Instrumentation and Control Hardware, Advanced Control .i System Technology, Human Factors Research, EPRI's Nuclear Safety Research and Devel-i opment

== Description:== This three-volume report contains papers presented at the Twentieth Water Reactor Safety f ' Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the = week of October 21-23,1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchers from the Commission of the European Communities (CEC), China, Finland, France, Germany, Japan, Spain and Taiwan. This document, Volume 1, presents papers on Advanced Reactor Research Advanced passive LWR's; advanced instrumentation and control hardware, advanced control system technology; human factors research; and EPRI's nuclear safety research and development. f Publication Date: March 1993 Prepared by: Weiss, A.J. [ comp.] [Brookhaven National Lab., Upton, NY (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Brookhaven National Lab.,Upton,NY (United States) - Keywords: ARRO'ITA, BETA, BWR type reactors, CAFTA, CHEKWORK, COMMIX, CONTAIN, CORB H, CORCON-MOD 2,EPRI,EQHAZARD,FREY,G OTHIC, human factors, MAAP, meetings, MELCOR, MINET, NPRDS, NUMARC, ORAM-TIP, PIPA, PWR type reactors, RAPID, reactor accidents, reactor components, reactor control systems, reactor instrumen-tation, reactor safety, RELAP5/ MOD 3, reliability, research programs, RETRAN-3, risk i assessment, SSC,'IRACG, VIPRE l i l t i b r NUREG/BR-0083, Vol.9 6

NUREG/CP -0126-Vol.2 General

Title:

Proceedings of the US Nuclear Regulatory Commission Twentieth Water Reactor Safety Information Meeting. Volume 2 Severe Accident Research, Hermal Hydraulics

== Description:== nis three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21-23,1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchers from the Commission of the European Communities (CEC), China, Finland, France, Germany, Japan, Spain,and Taiwan. Publication Date: March 1993 Prepared by: Weiss, AJ. [ comp.] [Brookhaven National Lab., Upton, NY (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear. Regulatory Research: Brookhaven National Lab., Upton, NY (United States) Keywords: ABAQUS, AEROSOL, APRIL. MOD 3, ATHLET-SA, BUSCA, BUTRAN, BWR type reactors, CATHARE2, CHARM, CHEMKIN, CHIP, CONTAIN, containment, corium, CORMLT, CORSOR-M, COUPLE, DEIMOS, ESCADR. ESPROSE, ESTER, explosions, FASTORASS, FIPLOC, fission product release, FLECHT-SEASET, FLOW 3 D, FPRATE, FRAP-T6, heat transfer, hydraulics, ICARE-1, ICARE-2, INSPECT, IODE, JERICO, KES S, MAAP 3.0, MARCH 3, meetings, MELTSPREAD, MELPROG, MITRA, ORIGEN2, PAT 2SR5,PATRAN,PECLOX,PM-ALPHA.PRUEP PWRtypereactors RAFT RALOC j MOD 2.2,reactoraccidents reactorsafety,RELAP/SCDAP5,researchprograms RSYGAL, STCP, TRAPFRANCE. TRAPMELT, VICTORIA, WECHSL 1 i i 7 NUREG/BR--0083, Vol.9

NUREG/CP--0126-Vol.3 General

Title:

Proceedings of the US Nuclear Regulatory Commission Twentieth Water Reactor Safety - Information Meeting. Volume 3, Aging Research Developments, Primary System Integrity, Structural and Seismic Engineering,Eanh Sciences,Probabilistic Risk Assessment Topics

== Description:== This three-volume repon contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21-23,1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchers from the Commission of the European Communities (CEC), China, Finland, France,G ermany, Japan, Spain and Taiwan. This document, Volume 3, presents papers on Aging research development; primary system integrity; structural and seismic engineering, canh sciences, and probabilistic risk assess-ment topics. Publication Date: March 1993 Prepared by: Weiss, AJ. [ comp.] [Brookhaven National Lab., Upton, NY (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Brookhaven National Lab.,Upton, NY (United States) - Keywords: aging, BWR type reactors, CART, canhquakes, EMTP, ground motion, LERS, MACCS, meetings, MELCOR, NPRDS, NUDOCS/AD, ODRPACK2.01, ORIGEN2, PR-EDB, PRAISE, primary coolant circuits, PWR type reactors, reactor accidents, reactor safety, research programs, risk assessment, seismic effects Structural Materials Electronic Data-base, SURFIT, TAC,TACMVS, TR-EDB ,= 8 NUREG/BR--0083, VoL9

i [E i NUREG/CP--0130-Vol.1 DEPOSITION 1 'l

Title:

Proceedings of the 22nd DOE /NRC Nuclear Air Cleaning Conference, Sessions 1-8. Volume 1 ) l

== Description:== This document contains the papers and the associated discussions of the 22nd Department of. j Fnergy/ Nuclear Regulatory Commission (DOE /NRC) Nuclear Air Cleaning Conferenm. Major topics are advanced reactors, reprocessing, filter testing, waste management, instru-i ments and sampling, reactor accidents, filters and filter performance, adsorber testing and performance, carbon testing, and ventilation systems. This document, Volume 1, contains Sessions 1 through 8. Publication Date: July 1993 Prepared by: First, M.W. [ed.] [ Harvard Univ., Boston, AM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Harvard Univ., Boston, MA (United States), Harvard Air Cleaning - 1 Lab.; USDOE Assistant Secretary for Nuclear Energy, Washington, DC (United States). Office ofNuclear Safety;IntemationalSociety ofNuclear Air Treatment Technologies.Inc., Columbus,OH (United States) Keywords: aerosols, air cleaning, air filters, decontamination, DEPOSITION, fuel reprocessing plants, meetings, nuclear facilities, nuclear power plants, performance, proceedings. radiation monitoring, radiation transport, radioactive effluents, radioisotopes, reactor accidents, j reactor safety, standards, testing, ventilation systems i i l i l 9 ' NUREG/BR--0083, Vol.9 4

I 1 NUREG/CP--0130-Vol.2 LAFIS, ACRITH, COSYMA, SOFIRE-M II, SPRAY-rf, FIRAC - i

Title:

Proceedings of the 22nd DOE /NRC Nuclear Air Cleaning Conference, Sessions 9-16. Volume 2 i

== Description:== This document contains the papers and the associated discussions of the 22nd Department of Energy / Nuclear Regulatory Commission (DOE /NRC) Nuclear Air Cleaning Conference. Major topics are advanced reactors, reprocessing, filter testing, waste management, instru-ments and sampling, reactor accidents, filters ar.d filter performance, adsorber testing and performance, carbon testing, and ventilation systems. This document, Volume 2, contains sessions 9 through 16. Publication Date: July 1993 Prepared by: First, M.W. [ed.] [ Harvard Univ., Boston, MA (United States)) Nuclear Regulatory Commission, Washirigton, DC (United States), Office of Nuclear Prepared for: Regulatory Research; Harvard Univ., Boston, MA (United States), Harvard Air Cleaning Lab.: USDOE Assistant Secretary for Nuclear Energy, Washington, DC (United States), Office of Nuclear Safety; International Society of Nuclear Air Treatment Technologies,Inc., Columbus, OH (United States) ACRITH, adsorbents.aircleaning airfilters, carbon,COSYMA, decontamination,FIRAC, Keywords: LAFIS, meetings,nuclearfacilities nuclearpowerplants, performance. proceedings radia-tion monitoring, radiation transport, radioactive effluents, radioactive waste management, radioisotopes.reactoraccidents,reactorsafety SOFIRE-MII, SPRAY-II, standards, testing, ventilation systems t a t ? i NUREG/BR--0083, Vol.9 10 [ 7

NUREG/CP--0131 General

Title:

Proceedings of the Joint IAEA/CSNI Specialists

• Meeting on Fracture Mechanics Verifica-tion by Large-Scale Testing Held at Pollard Auditorium, Oak Ridge, Tennessee

== Description:== This report contains 40 papers that were presented at the Joint International Atomic Energy Agency / Committee on the Safety of Nuclear Installation (IAEA/CSNI) Specialists' Meeting -- of Fracture Mechanics Verification by Large-Scale Testing held at the Pollard Auditorium, Oak Ridge, Tennessee, during the week of October 26-29,1992. The papers are printud in the order of their presentation in each session and describe recent largesale fracture (brittle and/or ductile) experiments, analyses of these experiments, and comparisons between predictions and experime:.tal results. The goal of the meeting was to allow international experts to examine the fracture behavior of various materials and structures under conditions. relevant to nuclear reactor components and operating environments. The emphasis was on the ability of various fracture models and analysis methods to predict the wide range of experimentaldata now available. Publication Date: October 1993 Prepared by: Pugh, C.E.; Bass, B.R.; Keeney, J.A. [ comps.] [ Oak Ridge National Lab., TN (United - States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div.'of Engineering; Oak Ridge National Lab., TN (United States) Keywords: AB AQUS, ADINA, ADINA FE, ADINA-T, CASTEM 2000, COMMIX IB, crack propa-gation, cracks, fracture mechanics, fracture properties, FRACTURE.TWO, IAEA, IWM-CRACK, IWM-VERB, meetings, NEWMIX, nuclear power plants, OCA-P. ORMGEN. ORVIRT, PATRAN, pipes, pressure dependence, pressure vessels, proceedings, reactor com ponents, reactor materials, REMIX, Rousselier model, SOLA-PTS, temperature depen-dence, TEMPEST, testing, thermal shock, VISCRK, VVTVIRT 1 11 NUREG/BR--0083, Vol.9 i

L NUREG/CP -0134 i LER,NPRDS 'l

Title:

Intemational Atomic Energy Agency Specialists' Meeting on Experience in Ageing, Maintenance, and Modemization of Instrumentation and Control Systems for Improving -l NuclearPowerPlant Availability

== Description:== This report presents the proceedings of the Specialists' Meeting on Experience in Ageing, Maintenance, and Modemization of Instrumentation and Control Systems for Improving Nuclear Power Plant Availability that was held at the Ramada Inn in Rockville, Maryland, on May 5-7,1993. 'Ihe meeting was presented in cooperation with the Electric Power Research Institute, Oak Ridge National Laboratory, and the Intemational Atomic Energy - Agency.There were approximately 65 participants from 13 countries at the meeting. Publication Date: October 1993 Prepared by: [ Nuclear Regulatory Commission, Washington, DC (United States)] t Prepared for: Nuclear Regulatory Commission, Washington, DC (United States); International Atomic Energy Agency, Vienna (Austria); Electric Power Research Inst., Palo Alto, CA (United States); Oak Ridge National Lab., TN (United States) Keywords: aging, availability, LER, maintenance, meetings, modifications, NPRDS, nuclear power l plants,reactorcomponents,reactorcontrolsystems.reactorinstrumentation reactorlicens-i l ing, reliability, software l i I i i t t h l 12 NUREG/BR -0083,Vol.9 l t -

1 1 BNL-NUREG -51708-Vol.7 NUREG/CR--3469-Vol.7 ACE, ESTS J F

Title:

Occupational Dose Reduction at NuclearPowerPlants: Annotated Bibliography of Selected Readings in Radiation Protection and ALARA. Volume 7

== Description:== The ALARA Center at Brookhaven National Labomtory publishes a series of bibliographies i ofselected readingsin radiation protection and ALARA in the continuing effort to collect and l disseminate information on radiation dose reduction at nuclear power plants. This is volume 7 - of the series.The abstracts in this bibliography were selected from proceedings of technical meetings and conferences, journals,research reports,and searches of the Energy Science and Technology database of the US Department ofEnergy.The subject material of these abstracts .I relates to radiation protection and dose reduction and ranges from use of robotics to l operational health physics, to water chemistry. Material on the design, planning, and I management of ntalear power stations is included, as well as information on decommission-ing and safe storage efforts. Volume 7 contains 293 abstracts, an author index, and a subject index. The author index is specific for this volume.The subject index is cumulative and lists all abstract numbers from volumes 1 to 7. The numbers in boldface indicate the abstracts in this volume: the numbers not in boldface represent abstracts in previous volumes. Publication Date: July 1993 I Prepared by: Kaurin, D.G.; Khan, T.A.: Sullivan S.G.: Baum, J.W. [Brookhaven National Lab., Upton, NY (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications: Brookhaven National Lab., Upton, NY (United States) Keywords: ACE, ALARA, bibliographies,ESTS,nuclearpowerplants.occupationalexposure person-nel, radiation doses, radiation protection, reactor safety 13 NUREG/BR-4)083, Vol.9. l

7 i LMF--136. NUREG/CR-4214-Rev.1-Pt.2-Add.2 'ORIGEN, MACCS Tiile: Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. Modifica-tion of Models Resulting from Addition of Effects of Exposute to Alpha-Emitting Radionu-i clides: Revision 1,Part 2, Scientific Bases for Health Effects Models, Addendum 2

== Description:== De Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. De Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts b NRC-y supported groupsresultedinimprovedhealtheffectsmodelsthatwerepublishedinthe1985 report entitled " Health Effects Models for Nuclear Power Plant Consequence Analysis," l NUREG/CR-4214, and revised further in the 1989 report NUREG/CR-4214,Rev.1.Part2. De health effects models presentedin the 1989 NUREG/CR-4214 report were developed for exposure to low linear energy transfer (LET) (beta and gamma) radiation based ori the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information i related to low LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source, term. The first addendum report, entitled " Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports l on Health Effects of Ionizing Radiation, Low LET Radiation, Part 2: Scientific Bases for Health Effects Models," was published in 1991 as NUREG/CR 4214, Rev.1, Part 2, l - Addendum 1. This second addendum addresses the possibility that some fraction of the accident source term from an operating nuclear power plant comprises alpha-emitting. radionuclides. Consideration of chronic high-LET exposure from alpha radiation as well as acute and chronic exposure tolow-LET beta and gamma radiations is a reasonable extension of the health effects model. l t Publication Date: May 1993 Prepared by: Abrahamson,S.[WisconsinUniv Madison,Wl(UnitedStates)]; Bender,M.A.[Brookhaven [ i National tab.,Upton,NY (United States)]; Boecker,B.B.; Scott,B.R. [Lovelace Biomedical and EnvironmentalResearchinst., Albuquerque,NM (United States), Inhalation Toxicology Research Inst.]; Gilbert, E.S. [ Pacific Northwest Lab., Richland, WA (United States)] l Prepared for: Nuclear Regulatory Commission, Wastington, DC (United States), Div.' of Regulatory Applications; Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (United States), Inhalation Toxicology Research Inst.: Wisconsin Univ., Madison, WI I (United States); Brookhaven National Lab., Upton, NY (United States); Pacific Northwest Lab.,Richland,WA (United States) i . Keywords: BEIR-Vmodel biologicatradiationeffects,fissionproductrelease genetieradiationeffects, j health hazards, ICRP model, MACCS, maximum credible accident, NCRP model, nuclear power plants, ORIGEN, radiation effects, radiation hazards, radionuclide administration, release limits, somatically significant dose, US NRC l t 3 i NUREG/BR--0083,Vol.9 14 k

i I 4 ITRI--141 NUREG/CR--4214-Rev.2-Pt.1 CRAC, MACCS

Title:

Health Effects Models for Nuclear Power Plant' Accident Consequence Analysis. Part 1 Introduction, Integration, and Summary: Revision 2

== Description:== 'Ihis report is a revision of NUREQ/CR-4214, Rev.1, Part 1 (1990), Health Effects Models for Nuclear Power Plant Accident Consequence Analysis.This revision has been made to ' incorporate changes to the Heahh Effects Models recommended in two addenda to the NUREG/CR-4214, Rev.1, Part 2,1989 report. The first of these addenda provided recommended changes to the health effects models for low-LET radiations based on recent reports from UNSCEAR,ICRP, and NAS/NRC (BEIR V).The second addendum presented changes needed to incorporate alpha-emitting radionuclides into the accident exposure e source term. As in the earlier version of this report, models are provided for early and continuing effects, cancers and thyroid nodules, and genetic effects. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects--the hematopoietic, pulmonary, and gastrointestinal syndromes-are considered. Linear and linear-quadatic models are recommended for estimating the risks of seven types of' cancer in adults-leukemia, bone, lung, incast, gastrointestinal, thyroid, and "other." For most cancers, both incidence and mortality are j addressed. Five classes of genetic diseases -dominant, x-linked, aneuploidy, unbalanced - translocations, and multifactorial diseases-are also considered. Data are provided that should enable analysts to consider the timing and severity of each type of health risk. Publication Date: october 1993 Prepared by: Evans,J.S. [ Harvard School of Public Health, Boston,MA (United States)]; Abrahmson, S. [ Wisconsin Univ., Madison, WI (United S tates)]; B ender, M.A. [Brookhaven National Lab., Upton, NY (United States)); Boecker, B.B.: Scott, B.R. [ Inhalation Toxicology Research Inst., Albuquerque, NM (United States)]: Gilbert, E.S. [Battelle Pacific Northwest Lab., ' Richland,WA(United States)] i Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory j Applications: Inhalation Toxicology Research Inst., Albuquerque, NM (United States) Keywords: biological radiation effects, CRAC, health hazards, MACCS, mathematical models, mortal-ity, neoplasms, nuclear power plants, radicisotopes, reactor accidents, reactor safety 1 1 i a 15 NUREG/BR--0083, Vol.9 i

o I o L NUREG/CR--4219-Vol.9-No.2 ORNL/TM--9593-Vol.9-No.2 i FAVOR, PATRAN, ABAQUS, OCA-P, VISA-II 1

Title:

Heavy-Section Steel Technology Program. Semiannual Progress Report for April-September f 1992; Volume 9 No.2

== Description:== The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). 'Ihe program focus is on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in 11 tasks:(1) program management,(2) fracture methodology and analysis (3) material characterization and properties,(4) special technical assistance,(5) fracture analysis i computer programs, (6) cleavage < rack initiation, (7) cladding evaluations, (8) pressurized.' thermal-shock technology, (9) analysis methods validation, (10) fracture evaluation tests, and (11) warm prestressing.The program tasks have been structured to place emphasis on the resolution of fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation (HSSI) Program at ORNL and with related research programs both in the United States and abroad.This report provides an overview of principal developments in each of the 11 program tasks from April 1992 to September 1992. Publication Date: November 1993 Prepared by: Pennell, W.E. [ Oak Ridge National Lab., TN (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering: Oak Ridge National Lab.,TN (United States) Keywords: ABAQUS, analytical solution, computerized simulation, crack propagation, cracks, FA-l' VOR, finite element method, fracture properties,J-Q theory, OCA-P, PATRAN, pressure vessels, progress report, reactor materials, reactors, research programs, steels, testing, j thermal shock, VISA-II t P b NUREG/BR--0083, Vol.9 16 b

1 PNL--8454-Vol.1-Rev.1 NUREG/CR--5247-Vol.1-Rev.1: RASCAL 2.0, DECAY, FM DOSE, ST-DOSE i .l

Title:

RASCAL Version 2.0: User's Guide. Volume 1, Revision 1 l

== Description:== The Radiological Assessment System for Consequence Analysis, Version 2.0 (RASCAL ] 2.0) has been developed for use during response to radiological emergencies.'The system j supplements assessments based on plant conditions and quick estimates based on paper methods. The modelis designed to provide a rough comparison to Environmental Protection ] Agency (EPA) Protective Action Guidance and thresholds for acute health effects. RASCAL .- l will be used by Nuclear Regulatory Commission (NRC) personnel who report to the site of a nuclear accident to conduct an independent evaluation of dose and consequence projec-tions. The model was developed to allow consideration of the dominant aspects of source 8 term, transport, dose, and consequences. Graphics were designed for use on the COMPAQ Portable III microcomputer used by the NRC response personnel. The model can be run on any DOS system, and the results can be displayed as text or maps. Substantial revisions to RASCAL 1.3 have required the release of this new version of the system.Two new models have been added to RASCAL 2.0.The first,FM-DOSE, computes doses from environmental concentrations. The second, DECAY, computes radiologic decay and ingrowth over a selected time period.The modelthat previously was the whole of RASCAL has been renamed ST-DOSE. Source term, transpon, and dose calculations in ST-DOSE have all been modified, and the input and output screens have been revised based on user comments and suggestions. RASCAL is reviewed on an ongoing basis and will be funher revised to reflect advances in our understanding ofreactor accidents.This volumeis a user's guide to RASCAL 2.0 and includes complete instructions for its use and example problems. It includes only that technical background material that cannot be found elsewhere. References to the technical details are included. Publication Date: February 1993 Prepared by: Athey, G.F. [ Phoenix Associates, Bethesda, MD (United States)]; Sjoreen, A.L. [ Oak Ridge National Lab., TN (United States)]; Ramsdell, LV. [ Pacific Nonhwest Lab., Richland, WA (United States)); McKen T,L [NuclearRegulatory Commission, Washington, DC (United States)) Prepared for: Nuclear Regulatory Commission, Wastiington, DC (United States), Div. of Operational Assessment: Pacific Northwest Lab., Richland, WA (United States) Keywords: aerosols, biological radiation effects, computer graphics, DECAY, FM-DOSE, health hazards, inhalation, manuals, meteorology, plumes, radiation accidents, radiation doses, radiation transport, RASCAL 2.0, reactor accidents, source tenns, ST-DOSE i 17 NUREG/BR--0083, Vol.9 l

NUREG/CR--5247-Vol.2 RASCAL 2.0, DECAY, FM-DOSE, ST-DOSE

Title:

RASCAL Version 2.0 Workbook. Volume 2 i

== Description:== he Radiological Assessment System for Consequence Analysis, Version 2.0 (RASCAL 2.0) has been developed for use by the Nuclear Regulatory Commission (NRC) personnel who respond to radiological emergencies. This workbook is intended to complement the I RASCAL 2.0 User's Guide (NUREG/CR-5247, Vol.1. Rev.1 ). He workbook contains exercises designed to familiarize the user with the computer based tools ofRASCAL through hands-on problem solving.The workbook is composed of four major sections.he first part -- { is a RASCAL familianzation exercise to acquaint the us r with the operation of the forms, menus, on-line help, and documentation. The last three parts contain exercises in using the three tools of RASCAL Version 2.0: DECAY, FM-DOSE, and ST-DOSE. Each section of exercises is followed by discussion on how the tools could be used to solve the problem. Publication Date: May 1993 Prepared by: Athey, G.F. [Athey Consulting, Charles Town, WV (United frates)]; McKenna, TJ. [ Nuclear Regulatory Commission, Washington, DC (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Operational Assessment: Athey Consulting, Charles Town, WV (United States) j Keywords: computer program documentation, DECAY, documentation, educational tools, fission i product release, FM-DOS E, nuclear power plants, particulates, radiation accidents, radiation doses, radiation transport, RAS CAL 2.0, reactor accidents, source terms, ST-DOSE, training i '1 1 1 I l l l l NUREG/BR--0083,Vol.9 18

SAND--90-2765 Vo!.2-Pt.1 NUREG/CR-5305 Vol.2-Pt.1. General

Title:

Integrated Risk Assessment for the LaSalle Unit 2 Nuclear PowerPlant: ?henomenology and Risk Uncertainty Evaluation ProgTam (PRUEP). Appendices A-C: Volume 2, Part 1

== Description:== This volume contains a description of the codes and input / output files used to perform the LaSalle Level II/III Probabilistic Risk Assessment. A chart showing the process flow is presented and the relationship between the codes and the needed input and output data is discussed. Code listings for codes not documented elsewhere and complete orsamplelistings of the input and output files are also presented. Publication Date: May 1993 Prepared by: Brown,T.D.;Payne, A.C. Jr.; Miller L.A.: Shiver. A.W. Higgins,SJ.;Sype,T.T.[Sandia National Labs., Albuquerque, NM (United States)]; Johnson, J.D. [ Science Applications Intemational Corp., Albuquerque, NM (United States)]; Chanin, DJ. [Technadyne Engi-neering Consultants, Inc., Albuquerque, NM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States) Keywords: COMBIN, computer program documentation, EVNTRE, La Salle County-2 reactor, Latin Hypercube Sampling, MACCS, MASTERK, PARTITION, PRAMIS, PRUEP, PSTEVNT, reactor accidents, reactor components. reactor safety, risk assessment, S AS, STER.TEMAC, XSOR 1 I 19 NUREG/BR--0083, Vol.9

h NUREG/CR-5305-Vol.2-Pt.2 SAND--90 2765.Vol.2 Pt,2 PRUEP, PARTITION, MACCS, NEWPART, COMBIN, STER T

Title:

Integrated Risk Assessment for theLaSalle Unit 2 Nuclear Power Plant.Phenomenology and Risk Uncertainty Evaluation Program (PRUEP): Volume 2. Part 2, Appendices D-G

== Description:== This volume contains a description of the codes and input / output files used to perform the LaSalle Level II/III Probabilistic Risk Assessment. A chart showing the process flow is presented and the relationship between the codes and the needed input and output data is discussed. Code listings for codes not documented elsewhere and complete or samplelistings of the input and output files are also presented. Putlication Date: May 1993 Prepared by: Brown, T.D.; Payne, A.C., Jr.; Miller,L.A.; Shiver, A.W.; Higgins, SJ.; Sype.T.T. [Sandia National Labs., Albuquerque, NM (United States)); Johnson, J.D. [ Science Applications International Corp., Albuquerque, NM (United States)]; Chanin, DJ. [Technad>me Engi-neering Consultants,Inc., Albuquerque,NM(United States)] Prepared for: Nuclear Regulatory Commission. Washington, DC (United States), Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States) Keywords: COMBIN, computer program documentation, La Salle County-2 reactor, MACCS, NEWPART, PARTITION, PRUEP, reactor accidents, reactor safety, risk assessment, source terms, STER NUREG/BR--0083,Vol.9 20 + +

-l SAND--89-0943 NUREGICR-5360 i t XSOR

Title:

XSOR Codes Users Manual f

Description:

This report describes the source term estimation codes, XSORs. He codes are written for three pressurized water reactors (S urry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named "XSOR." The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. He codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NR.C)in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena j and their uncertainty which are not currently modeled by the mechanistic codes. The i uncertainty distributions ofinput parameters may be used by an XSOR code to estimate the 'I uncertainty of source terms. Publication Date: November 1993 Prepared by: Jow, Hong-Nian [Sandia National Labs., Albuquerque, NM (United States)]; Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States)]; Johnson, J.D. [ Science Applications Intemational Corp., Albuquerque, NM (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States) Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States); Technadyne Engi-- neering Consultants,Inc.. Albuquerque, NM (United States); Science Applications Interna-tional Corp., Albuquerque, NM (United States) Keywords: BWR type reactors, PWR type reactors, reactor accidents, source terms, XSOR 1 i l l I 21 NUREG/BR--0083, Vol.9 I

NUREG/CR--5471 SAND--89 2562 NPRDS, IPRDS, OPEC-2

Title:

Enhancements to Data Collection and Reponing of Single and Multiple Failure Events

== Description:== During the past few years, methods have been developed for quantifying and analyzing common cause failures (CCFs). These methods have outpaced current data co!!ection activities. This document discusses the collection and documentation of failure events at nuclear power plants with respect to these new CCFs methods. The report concentrates on the infonnation necessary to improve the parameter estimates for both independent and dependent events in probabilistic risk assessments (PRAs) and alludes to the fact that the same information can be used to enhance other nuclear power plant activities. Several existing databases are reviewed as to their adequacy for these new CCF methods, and areas where information is lacking, either because certain information is simply not required to be reported or because required information was simply not reponed, are identified. Finally, data needs identified from recent PRAs are discussed. Publication Date: March 1993 Prepared by: Whitehead, D.W. [Sandia National Labs., Albuquerque, NM (United States)]; Paula, H.M. [JBF Associates,Inc., Knoxville,TN(United States)]; Parry,G.W. [NUS Corp.,Gaithersburg, MD (United S tates)]; Rasmuson, D.M. [ Nuclear Regulatory Commission, Washington, DC (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States) Keywords: data acquisition, documentation. failures,IPRDS, NPRDS, nuclear power plants, OPEC-2, reactor components, reactor safety, recommendations, reliability, risk assessment i i . NUREG/BR483, Vol.9 22

ORNL/FM 11945 NUREGICR--5782 OCA P

Title:

Pressurized Thermal Shock Probabilistic Fracture Mechanics Sensitivity Analysis for Yankee Rowe Reactor Pressure Vessel

== Description:== The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss-of-coolant transient as the loading condition. Regions of the vessel with distinguishing features wem to be treated individually-upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Dermal Shock (IFTS) Program.The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded ) flaws.He results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. Publication Date: August 1993 Prepared by: Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.' [ Oak Ridge National Lab., TN (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering: Oak Ridge National Lab., TN (United States) Keywords: defects, embrittlement, fracture mechanics, loss of coolant, OCA-P, pressure vessels, pressurization, probability, radiation effects, steels, stress analysis, thermal shock, welded joints, Yankee Rowe reactor 3 l i '] i I i 1 23 NUREG/BR--0083 Vol.9 i --- ---J

NUREG/CR--5801 SAND--91-7087 - ' General

Title:

Procedure for Analysis of Common-Cause Failures in Probabilistic Safety Analysis j

== Description:== This report provides practical guidelines for treatment of common cause failures (CCF) in l i risk and reliability studies.The procedures outlined in this report are organized according to three phases of analysis, screening analysis, detailed qualitative analysis, and detailed - quantitative analysis. The results of the screening analysis phase include conservative identification of potential consmon-cause vulnerabilities and determination of the scope and focus for more detailed analysisinPhasesIIandIII. PhaseII,the detailed qualitative analysis, provides a better understanding of the plant-specific susceptibilities of the systems and components to causes and coupling mechanisms of CCF. The information from this phase can then be used as a basis for a plant-specific quantitative assessment of CCF frecuencies. l Detailed guidelines are provided for Phase III to aid the analyst in using this qualitative 1 information and generic data in developing a plant-specific CCF base. Depending on the overall objective of the study, CCF analysis can stop at the end of any of the three phases. Publication Date: April 1993 Prepared by: Mosleh, A. [ Maryland Univ., College Park, MD (United States), Dept. of Materials and l t Nuclear Engineering] Prepared for:- Nuclear Regulatory Commission, Washington, DC (United States), Div. of. Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States); Maryland Univ., College Park, MD (United States), Dept. of Materials and Nuclear Engineeri::: Keywords: failure mode analysis, failures, fault tree analysis, mathematical mode 9, nuclear power 1 plants, probability, reactor components, reactor safety, reliability, risk assessment, safety [ analysis j \\ - NUREG/BR--0083, Vol.9 24 o

CNWRA-92-02S-Vol.3-No.2 NUREG/CR--5817-Vol.3-No.2 - t BIGFLOW, MINTEQA2 i

Title:

NRC High-Level Radioactive Waste Research at CNWRA, July-December 1992. Volume 3, No. 2

== Description:== Progress from July I to December 31,1992, on the nine NRC-sponsored research projects conducted at the Center forNuclear Waste Regulatory Analyses is described. Ion-exchange experiments between clinoptilolite and aqueous solutions of Na+ and Sr + and three 2 applications of reaction-path modeling are described in the Unsaturated Mass Transport (Geochemistry) project. Numerical simulation of a laboratory-scale non-isothermal two-phase flow is discussed in the 'Ihermohydrology chapter. Methods for estimating rock joint roughness coefficient are the focus of the Seismic Rock MechancJ project for which the Tilt Test,Tse and Cruden's equations, and fractal-based equations were tested and found to be j unsatisfactory. In the Integrated Waste Package Experiments chapter, investigations of pit mitiation and repassivation potential for alloys 825 and C-22 and stainless steel 304L and - 316L are described. Testing of the BIGFLOW computer code and visualization of fracture topology is the theme of the Stochastic Hydrology project. Preliminary analysis of field data j from the Akrotiri site in Greece is developed in the Geochemical Analogs project. Mecha-nistic modeling of sorption using the MINTEQA2 code is investigated as part of the Sorption project. Adaptive gridding and "rnodified equations" methods for solving the flow and transport equations are described in the Performance Assessment chapter. Finally, the Volcanism chapter focuses on using nonhomogeneous Poisson processes for estimating - probability of volcanic events at the putential repository site. Publication Date: July 1993 Prepared by: Sagar, B. (ed.); Ababou, R.: Ahola, M. [ Southwest Research Inst., San Antonio,TX (United States), Center for Nuclear Waste Regulatory Analyses] i Prepared for: Nuclear Pegulatory Commission, Washington, DC (United States), Div. of Regulatory i Applications;SouthwestResearchInst. San Antonio,TX(UnitedStates),CenterforNuclear Waste Rerulatory Analyses - Keywords: BIGFLOW, containers, corrosion, failures, geochemistry, geologic fractures, high-level radioactive wams, hydraulics, hydrology, MINTEQA2, packaging, performance, porous materials, progreas report, radioactive waste management, radionuclide migration, research { programs, seismic effects, sorption,' nit Test, Tse and Cruden's equations, two-phase flow, zeolites ^ l p i 25 NUREG/BR--0083, Vol.9 i L u-

3 NUREG/CR-5818 ' EGG--2665 RELAP5/ MOD 3, CSAU Tille: Uncertainty Analysis of Minimum Vessel Liquid Inventory During a Smail. Break LOCA in a B&W Plant: An Application of the CSAU Methodology Using the RELAP5/ MOD 3 ' Computer Code j y

== Description:== ne Nuclear Regulatory Commission (NRC) revised the emergency core cooling system licensing rule to allow the use of best estimate computer codes,provided the uncertainty of l the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability, and Uncertainty (CS AU) to evaluate best estimate code uncertainties.The objective of this work was to adapt - I - anddemonstratetheCSAUmethodologyforasmall-breakloss-of-coolantaccident(SBLOCA) in a Pressurized Water Reactor of Babcock & Wilcox Company lowered loop design using RELAP5/ MOD 3 as the simulation tool.The CS AU methodology was successfully demon-strated for the new set of variants defined in this project (scenario, plant design, code). s However, the robusmess of the reactor design to this SBLOCA scenario limits the applica-bility of the specific results to other plants or scenarios. Several aspects of the code were not exercised because the conditions of the transient never reached enough severity. The plant - operator proved to be a determining factor in the course of the transient scenario, and steps were taken to include the operator in the model, simulation, and analyses. Publication Date: December 1992 Prepared by: Ortiz, M.G.: Ghan, L.S. [EG and G Idaho,Inc., Idaho Falls, ID (United States)] l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; EG and G Idaho, Inc., Idaho Falls, ID (United States) Keywords: computerized simulation, CSAU, data covariances, ECCS, loss of coolant, PWR type reactors, reactor cooling systems, reactor licensing, reactor operators, RELAP5/ MOD 3 ] 4 t A 1 i I 't NUREG/BR--0083, Vol.9 26 I

~ ANL--91/43 NUREG/CR--5822 COMMIX

Title:

Analysis of Thermal Mixing and Boron Dilution in a PWR

== Description:== nermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was a four-loop Zion reactor that was initially filled with hot boron-rich water. It was assumed that the reactor coolant pumps are tripped. Following the trip, cold unborated water from seal injection or other sources continuously - flows into the reactor coolant system and dilution takes place first in the pump suction line and then in the reactor vessel. The thermal mixing and boron dilution under these conditions were analyzed. For the analysis of thermal mixing, water at room temperature (referred to as i cold water) was fed into the cold leg of the reactor system at various flow rates. For the analysis of boron dilution, cold and hot unborated water was fed into the cold leg at a high I flow rate. He subsequent transient thermal mixing and boron dilution that would occur in the reactor system were simulated for l to 2 hours depending on the flow rate. A third analysis i was performed for the boron dilution after the start of the reactor coolant pump,which forces a slug of cold unborated water from the pump suction line into the reactor vessel. This transient was simulated for 20 seconds, by which time the slug has been pushed out of the l reactor core. The rates of reactivity insertion were evaluated for these analyses. Publication Date: February 1993 ] l Prepared by: Sun, J.G.; Sha, W.T. [Argonne National Lab., IL (United States)] { i Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Programs: Argonne National Leb., IL (United States) Keywords: baron additions, COMMIX, dilution, flow models, fluid flow, mixing, pumps, PWR type reactors, reactor cooling systems, water chemistry, Zion-1 reactor, Zion-2 reactor i-27 NUREG/BR--0083, Vol.9 )

NUREG/CR--5843 SAND--92-0167 CORCON-MOD 3

Title:

CORCON-MOD 3: An Integrated Computer Model for Analysis of Molten Core-Concrete Interactions. User's Manual

== Description:== The CORCON-MOD 3 computer code was developed to mechanistically model the impor-tant core-concrete interaction phenomena, including those phenomena relevant to the assessment ofcontainment failure and radionuclide release.The code can be applied to a wide range of severe accident scenarios and reactor plants.The code repn:sents the current state of the art for simulating core debris interactions with concrete. This document comprises the user's manualand gives a briefdescription of the models and the assumptions andlimitations in the code. Also discussed are the input parameters and the code output. Two sample problems are also given. Publication Date: October 1993 Prepared by: Bradley, D.R.; Gardner, D.R.: Brockmann, J.E.; Griffith, R.O. [Sandia National Labs., Albuquerque,NM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search Sandia National Labs., Albuquerque,hN (United States) Keywords: computerized simulation, containment, CORCON-MOD 3, corium, fuel-cladding intemc-r tions, manuals, meltdown, water cooled reactors i l l 1 i NUREG/BR--0083, Vol.9 28 l l

o l i i i NUREG/CR--5851 NPRDS, LER

Title:

Long-Term Performance and Aging Characteristics of NuclearPlant Pressure Transmitters

== Description:== This report presents the results of a comprehensive research and development project' conducted for the Nuclear Regulatory Commission (NRC) to study the effects of normal aging on calibration and response time of nuclear plant pressure, level, and flow transmitters o and to develop and validate new methods for testing the performance of the transmitters as ] installed in nuclear power plants.The projectinvolved reseamh in seven areas as follows: (1) aging tests of complete transmitter assemblies; (2) aging tests of critical components of transmitters; (3) testing the effects of sensing line length, blockages, and voids on the - response time of pressure sensing systems; (4) oil loss phenomenon in Rosemount and other transmitters; (5) validation of new methods for on-line testing of response times of pressure transmitters; (6) on-line detection of oil loss, in Rosemount transmitters, and (7) analysis of Licensee Event Report (LER)and Nuclear Plant Reliability Data System (NPRDS) databases for failures of pressure sensing systems in nuclear power plants. Publication Date: March 1993 Prepared by: Hashemian, H.M.; Mitchell, D.W.; Fain, R.E.; Petersen, K.M. [ Analysis and Measurement Services Corp., Knoxville, TN (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering: 3 Analysis and Measurement Services Corp., Knoxville, TN (United States) Keywords: aging, calibration, failures, LER, NPRDS, nuclear power plants, performance, pressure measurement, reactor instrumentatioh, reactor safety, remote sensing, service life, testing, transducers l 29 NUREG/BR-0023, Vol.9 i

EGG--2677 ~ NUREG/CR--5882 TRAC-BFI

Title:

TRAC-B Thermal-Hydraulic Analysis of the Black Fox Boiling Water Reactor

== Description:== Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BFI computer code.This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with futum analysis wo-k at the US Nuclear Regulatory Commis-sion Technical Training Center in Chattanooga, Tennessee. 'Ihese accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions. Publication Date: May 1993 Prepared by: Martin, R.P. [EG and G Idaho,Inc., Idaho Falls,ID (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search: EG and G Idaho, Inc., Idaho Falls, ID (United States) Keywords: BlackFox-1 reactor,BlackFox 2 reactor,heattransfer. hydraulics.reactoraccidents, reactor safety, TRAC-BF1, transients NUREG/BR-0083,Vol.9 30

SAND--92-1422 NUREG/CR--5901 General Tiile: A Simplified Model of Aerosol Scrubbing by a Water Pool Overlying Core Debris Interacting with Concrete. Draft Report for Comment

== Description:== A classic model of aerosol scrubbing from bubbles rising through water is applied to the decontamination of gases produced during core debris interactions with concrete.ne model, originally developed by Fuchs, describes aerosol capture by diffusion, sedimentation, and inertial impaction. This original model for spherical bubbles is modified to account for ellipsoidal distortion of the bubbles. Eighteen uncenain variables are identified in the application of the model to the decontamination of aerosols produced during core debris interactions with concrete by a water pool of specified depth and subcooling. Rese uncertain variables include properties of the aerosols, the bubbles, the waterand the ambient pressure. Ranges for the values of the uncenain variables are defined based on the literature and experience. Probability density functions for values of these uncertain variables are hypoth-esized. He model of decontamination.is applied in a Monte Carlo sampling of the decontamination by pools of specified depth and subcooling. Results are analyzed using a nonparametric, order statistical analysis that allows quantitative differentiation of stochastic and phenomenological uncertainty. He sampled values of the decontamination factors are used to construct estimated probability density functions for the decontamination factor at confidence levels of 50,90 and 95 E The decontamination factors for pools 30,50,100,200, 300, and 500 cm deep and subcooling levels of 0,2,5,10,20,30,50,and 70 C are correlated by simple polynomial regression. Rese polynomial equations can be used to estimate decontamination factors at prescribed confidence levels. Publication Date: October 1992 Prepared by: Powers, D.A.: Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution: Sandia National Labs., Albuquerque, NM (United States) Keywords: aerosols, bubbles. concretes, decontamination. diffusion, Fuchs model, gases, mathematical models, Monte Carlo method, reactor core disruption, reactor cores, scrubbing, sedimenta-tion, spherical configuration, water i 31 NUREG/BR--0083, Vol.9 I

NUREG/CR--5907 SAND-92-1563 WETCOR-1 ?

Title:

Core-Concrete Interactions with Overlying Water Pools. The WETCOR-1 Test

== Description:== The WETCOR-1 test of simultaneous interactions of a high-temperature melt with water and 1 a limestone / common-sand concrete is described.The test used a 34.1-kg melt of 76.8 w/o Al O,16.9 w/o CaO,and 4.0 w/o SiO heated by induction using tungsten susceptors. Once 2 3 2 quasi-steady attack on concrete by the melt was established, an attempt was made to quench. the melt at 1850 K with 295 K water flowing at 57 liters per minute. Net power into the melt i 0.19 W/cm. The test configuration used in the 3 at the time of water addition was 0.61 WETCOR-1 test was designed to delay melt freezing to the walls of the test fixture.This was done to test hypotheses concerning the inhemnt stability of crust formation when high-i temperature melts are exposed to water. No instability in cmst formation was observed. The j flux of heat through the crust to the water pool maintained over the melt in the test was found to be 0.5210.13 MW/m. Solidified crusts were found to attenuate aerosol emissions during 2 the melt concrete interactions by factors of 1.3 to 3.5.The combination of a solidified crust and a 30-cm deep subcooled water pool was found to attenuate aerosol emissions by factors of 3 to 15. Publication Date: November 1993 Prepared by: Blose, R.E. [Ktech Corp., Albuque que, NM (United States)]; Powers, D.A.: Copus, E.R.: Brockmann, LE.; Simpson, R.B.; Lucero, D.A. [Sandia National Labs., Albuquerque, NM (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; Sandia National Labs., Albuquerque, NM (United States); Ktech Corp., Albuquer-que, NM (United States) Keywords: aerosols, concretes, corium, emission, heat transfer, hydraulics, interactions, meltdown, quenching, reactor safety, water, WETCOR-1 i L i 1 j NUREG/BR--0083, Vol.9 32

- SAND--91-2802-Vol.1 NUREG/CR-5927-Vol.1 General i

Title:

Evaluation of a Performance Assessment of Methodology for Low-Level RadioactiveWaste } Disposal Facilities: Evaluation of Modeling Approaches. Volume 1

== Description:== This report represents an update to our earlier reports on low-level waste performance assessment.This update addresses needed improvements and recommended approaches to the existing state of the art in modeling, treatment of uncertainty, and use of data. Greater attention is paid to developing an integrated approach to performance assessment than was done in earlierdevelopments of the methodology.Furthermore, insights are being developed by participating in validation exercises and by evaluating which validation data are needed to improve confidence in the methodology. Itis emphasized that the performance assessment methodology update is a work in progress; the recommendations given here will form the general directions toward which the methodology is heading, but some of the specific approaches may continue to evolve as the research progresses. Publication Date: August 1993 Prepared by: Kozak,M.W. Olague,N.E.;Rao,R.R. McCord,J.T. [SandiaNationalLabs., Albuquerque, NM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory. Applications: Sandia National Labs., Albuquerque, NM (United States) Keywords: B ARRIER, B LT, containers, data covariances, dosimetry, environmental exposure pathway, FEMWATER, fluid flow, GENII, geochemistry, groundwater, impacts, Latin Hypercube - Sampling, leaching, low level radioactive wastes, MODFLOW, MODPATH, Monte Carlo q method, NEFIRAN, PAGAN, performance, radiation transport, radioactive waste disposal, 'I radioactive waste facilities, radionuclide migration, RAESTRICT, source terms, VAM2D, VS2DT l -l h 33 NUREG/BR--0083, Vol.9

i NUREG/CR--5936 SAND--92-2109 IRRAS

Title:

Enhancements to the Accident Precursor Methodology

== Description:== A feasibility study for developing an improved tool and improved models for performing event assessments is described. De study indicates that the IRRAS code should become the base tool for performing event assessments but that modifications would be needed to make it more suitable for routine use. Alternative system modeling approaches are explored and an approach is recommended that is based on improved train-level models.These models are demonstrated for Grand Gulf and Sequoyah.neinsights that can be gained from importance measures are also demonstrated. He feasibility of using Individual Plant Examination (IPE) submittals as the basis for train-level models for precursor studies was also examined. The level of reported detail was found to vary widely, but, in general, the submittals did not provide sufficient information to fully define the model. The feasibility of developing an industry risk profile from precursor results and the feasubulity of trending precursor results for individual plants were considered. De data sparsity would need to be considered when using the results from me " types of evaluations, and because of the extremely sparse data for individual plants, we fou ad that trending evaluations for groups of plants would be more meaningful than trending a "aluations for individual plants. Publication Date: January 1993 Prepared by: Bohnhoff, WJ.: Dingman, S.E.: Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States) Keywords: BWRtypereactors feasibilitystudies,GmndGulf-1 reactor,GrandGulf-2 reactor,IRRAS, modifications, precursor, PWR type reactors, reactor accidents, reactor operation, risk assessment, Sequoyah-1 reactor, Sequoyah-2 reactor I NUREG/BR--0083, VC 0 34 i

l EGG--2688 NUREG/CR--5937 SCDAP/RELAP5/ MOD 3

Title:

Intentional Depressurization Accident Management S trategy forPressurized Water Reactors

== Description:== In a previous investigation of the Surry nuclear power station, it was concluded that intentional depressurization of the reactor coolant system (RCS) could prevent or mitigate the effects of direct containment heating (DCH) during a station blackout tran.mt. Two strategies, early and late depressurization, were investigated as methods to mitigate DCH. The investigation concluded that since there are greater opportunities to recover plant ' functions before core damage occurs and operator response uncertainties are lessened, the strategy of late depressurization is preferred over early depressurization. The results of the ) Surry analysis were extended to other US pressurized water reactors (PWRs) in order to evaluate their capability to successfully employ the late depressurization strategy to prevent or mitigate DCH. By applying appropriate scaling factors to the selected key parameters, this evaluation resulted in the categorization of four PWR groups based upon their perceived late j depressurization capability. In this report, a PWR representative of each of the four PWR j groups was chosen for detailed analysis of its capability to intentionally depressurize employing the late depressurization strategy. The phenomenological behavior, hardware { performance, and operational performance of these PWRs during the intentional depressur- { ization strategy were considered. The phenomenological behavior was analyzed using the I SCDAP/RELAP5/ MOD 3 severe accident analysis code. The results of these evaluations were then extended to the remaining PWRs comprising each PWR group. Publication Date: April 1993 - Prepared by: Brownson, D.A.; Haney, L.N.; Chien, N.D. [EG and G ldaho, Inc., Idaho Falls ID (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; EG and G Idaho, Inc., Idaho Falls, ID (United States) Keywords: Calvert Cliffs-1 reactor, Calvert Cliffs-2 reactor, containment, depressurization, heat trans-i fer, heating, hydraulics, management, Oconee-1 reactor, Oconee-2 reactor, Oconee-3 reactor, pipes, PWR type reactors, reactor accidents, reactor cooling systems, reactor safety, reactor vessels, reliability, SCDAP/RELAP5/ MOD 3, sensitivity analysis, Sequoyah-I reac-tor, Sequoyah.2 reactor, Surry-1 reactor, Surry-2 reactor, Surry-3 reactor, Surry-4 reactor i 35 NUREG/BR- 0083, Vol.9

NUREG/CR--5942 ORNL/TM--12229 MELCOR, CORBH, STCP

Title:

Severe Accident Source Term Characteristics for Selected Peach Bottom Sequences Pre-dicted by the MELCOR Code

== Description:== The purpose of this reportis to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences.For this comparison, two short-term depressur-ized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that, for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the hELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environ-ment than STCP for the station blackout sequences. Publication Date: September 1993 Prepared by: Carbajo, JJ. [ Oak Ridge National Lab., TN (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States) Div. of Safety Issue Resolution: Oak Ridge National Lab., TN (United States) Keywords: blackouts, computer calculations, containment, containment systems, CORB H, ECCS, heat transfer, hydraulics, loss of coolant, MELCOR, Peach Bottom-2 reactor, Peach Bottom 3 reactor, reactor safety, source terms, STCP l l l i f NUREG/BR--0083, Vol.9 36 i i

1 BNL-NUREG-52346 NUREG/CR--5943 BLT, FEMWATER i

Title:

Sensitivity Analysis and Benchmarking of the BLT Low-Level Waste Source Term Code ]

== Description:== To evaluate the source term for low-level waste disposal, a comprehensive model had been. I developed and incorporated into a computer code, called BLT (Breach-Leach-Transport) s Since the release of the original version, many new features and impmvements had also been j added to the Leach model of the code.This report consists of two different studies based on j the new version of the BLT code: (1) a series of verification / sensitivity tests and (2) benchmarking of the BLT code using field data. Based on the results of the verification / ] sensitivity tests,the authors concluded that the new version represents a significantimprove-ment and it is capable of providing more realistic simulations of the 1-aching process. Benchmarking work was carried out to provide a reasonable level of confidence in the model l predictions. In this study, the experimentally measured release curves for nitrate, technetium-1 99 and tritium from the saltstone lysimeters opented by Savannah River Laboratory were i used. The model results are observed to be in general agreement with the experimental data, within the acceptable limits of uncertainty. Publication Date: July 1993 Prepared by: Suen, CJ.: Sullivan, T.M. [Brookhaven National Lab., Upton, NY (US)] i 'l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; Brookhaven National Lab., Upton, NY (United States) .i Keywords: advection, benchmarks, BLT, computerized simulation, containers, failures, FEMWATER, j finite element method, flow rate, leaching, low-level radioactive wastes, radioactive waste. - disposal, radionuclide migration, sensitivity analysis, verification t -) 1 -l 1 -)

j

-] 37 NUREG/BR-4X)83, Vol.9 l i

NUREG/CR--5949 EGG--2689 SCDAP/RELAP5/ MOD 3

Title:

Assessment of the Potential for High-Pressure Melt Ejection P asulting from a Surry Station BlackoutTransient

== Description:== Containment integrity could be challenged by direct heating associated with a high-pressure melt ejection (HPhE) of core materials following reactor vesselbreach during certain severe accidents. Intentional reactor coolant system (RCS) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management s'Jategy to reduce risks by mitigating the severity of HPhE. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural ci culation flows could heat ex-l vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before vessel breach, providing an alternate mechanism for RCS I dpressurization and HPME mitigation."Ihis report contains an assessment of the potential for HPAE during a Surry station blackout transient without operator action and without recovery.The assessment included a detailed transient analysis using the SCDAP/RELAP5/ MOD 3 computer code to calculate the plant response with and without hotleg countercurrent natural circulation, with and without reactor coolant pump seal leakage, and with variations on selected core damage progression parameters.RCS depressurization-related probabilities were also evaluated, primarily based on the code results. Publication Date: November 1993 Prepared by: Knudson, D.L.; Dobbe, C.A. [EG and G. Idaho, Inc., Idaho Falls, ID (United States)] I Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-i search; EG and G Idaho, Inc., Idaho Falls, ID (United States) Keywords: blackouts, computer calculations, containment, corium, depressurization, heat transfer, heating, hydraulics, meltdown, reactor cooling systems, reactor safety, SCDAP/RELAPS/ j MOD 3, Surry 1 reactor, Surry-2 reactor, Surry-3 reactor, Surry-4 reactor, transients 1 i l NUREG/BR--0083, Vol.9 38 i

1 IS--5083 NUREG/CR--5957 BOSOR4, BOSOR5 l

Title:

System 80+" Containment: Structural Design Review

== Description:== A review of the structural design of the Combustion Engineering (CE) System 80+Nsteel containment was completed. The stress analysis and the evaluation of the structure against buciding were performed by using BOSOR4 and BOSOR5 finite difference software, respectively. The CE System 80+" containment was modeled as an axisymmetric shell consisting of different segments and mesh points with the additional mass of the penetrations and appurtenance being smeared around the circumference. The transition region was modeled using elastic springs with a foundation modulus of 1 R0 lb/in.3. The stresses due to the individual loads (dead loads, intemal and external pressures and temperatures) were computed using the stress analysis option in the BOSOR4 program. The stresses from i individual loads were combined according to ASME Code into stress intensities. Service Level B loadings produced a 20 percent over-stress in a small zonejust above the transition region. All other stress intensities were within allowable limits. For the System 80+N, the perfect shell with an elastic material was initially analyzed. The calculated factors of safety values were 23 (Level B) and 1.59 (Levels C and D). Finally, sensitivity studies were conducted to investigate the effects of mesh size and transition zone stiffness on the controlling buckling load. Publication Date: May 1993 'l Prepared by: Greimann, L.: Fanous, F.: Challa, R.: Bluhm, D. [Ames Lab., IA (United States)) l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering: Ames 1.ab.1A (United States) Keywords: BOSOR4,BOSOR5,computerizedsimulation.containmentbuildings, deformation. design. finite difference method, mathematical models, nuclear power plants, numerical solution, j pressure vessels, reactor safety, static loads, stress analysis j 4 'i 1 l I I l i I ) 39 NUREG/BR--0083, Vol.9 l f:-

NUREG/CR--5958 CDNSWC/SME-CR--16 92 General ' t

Title:

Two-Parameter Fracture Mechanics: Theory and Applications

== Description:== A family of self-similar fields provides the two parameters required to characterize the full. range of high-and low-triaxiality crack tip states. The two parameters,J and Q, have distinct roles: J sets the size scale of the process zone over which large stresses and strains develop, while Q scales the near-tip stress distribution relative to a high triaxiality reference stress state. An immediate consequence of the theory is this: It is the toughness values over a range ofcracktipconstraintthatfullycharacterizethematerial'sfractureresistance.itisshownthat - Q provides a common scale for interpreting cleavage fracture and ductile tearing data, thus allowing both failure modes to be incorporated in a single toughness locus. The evolution of - Q, as plasticity progresses from small scale yielding to fully yielded conditions, has been quantified for several crack geometries and for a wide range of material strain hardening properties. An indicator of the robustness of the J-Q fields is introduced; Q as a field parameter and as a pointwise measure of stress levelis discussed. Publication Date: February 1993 Prepared by: O'Dowd, N.P. [ Imperial Coll. of Science, Technology and Medicine, London (United Kingdom), Dept. of Mechanical Engineering]; Shih, C.F. [ Brown Univ., Providence, RI (United States), Div. of Engineering] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States) Div, of Engineering; Naval Surface Warfare Center, Annapolis, MD (United States), Carderock Div.; Brown Univ., Providence, RI (United States), Div. of Engineering Keywords: crack propagation, fracture mechanics, fracture properties, J-Q theory, strains, stresses, theoretical data 1 ? i F i i i NUREG/BR--0083, Vol 9' 40

EGG--2692 NUREG/CR--5964 SAPHIRE,IRRAS, SARA

Title:

S APHIRE Technical Reference Manual: IRRAS/S ARA Version 4.0

== Description:== This report provides information on the principles used in the construction and operation of Version 4.0 of the Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) system. It summarizes the fundamental mathemati-cal concepts of sets and logic, fault trees, and probability. The nport then describes the algorithms that these programs use to construct a fault tree and to obtain the minimal cut sets. It gives the formulas used to obtain the probability of the top event from the minimal cut sets, and the formulas forprobabilities that are appropriate under various assumptions concerning repairability and mission time. It defines the measures of basic event importance that these programs can calculate. The report gives an overview of uncertainty analysis using simple Monte Carlo sampling orLatin Hypercube sampling and states the algorithms used by these programs to generate random basic event probabilities from various distributions. Further references are given, and a detailed example of the reduction and quantification of a simple fault tree is provided in an appendix. Publication Date: January 1993 Prepared by: Russell, K.D.; Atwood, C.L.; Sattison, M.B. [EG and G Idaho, Inc., Idaho Falls, ID (United States)]; Rasmuson, D.M. [ Nuclear Regulatory Commission, Washington, DC (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution; EG and G Idaho, Inc., Idaho Falls, ID (United States) Keywords: algorithms, fault tree analysis,IRRAS 12 tin Hypercube Sampling, manuals, mathematical logic, mathematical models, Monte Carlo method, probability, reactor safety, reliability, risk assessment, SAPHIRE, SARA 1 41 NUREG/BR--0083, Vol.9 t

NUREG/CR--5966 SAND--92-2689 CONTAIN, STCP

Title:

A Simplified Model of Aeroso! Removal by Containment Sprays

== Description:== Spray systems in nuclear reactor containments are described. The scrubbing of aerosols from containment atmospheres by spray droplets is discussed. Uncertainties are identified in the prediction of spray performance when the sprays are used as a means for decontaminating containment atmospheres. A mechanistic model based on current knowledge of the physical phenomena involved in spray performance is developed. With this model, a quantitative uncertainty analysis of spray performance is conducted using a Monte Carlo method to sample 20 uncertain quantities related to phenomena of spray droplet behavior as well as the initial and boundary conditions expected to be associated with severe reactor accidents. Results of the uncertainty analysis are used to construct simplified expressions for spray decontamination coefficients.Two variables that affect aerosol capture by water droplets are not treated as uncertain; they are (1) "Q," spray water flux into the containment, and (2)"H," the total fall distance of spray droplets. The choice of values of these variables is left to the user since they are plant and accident specific. Also, they can usually be ascertained with some degree of certainty.The spray decontamination coefficients are found to be sufficiently dependent on the extent of decontamination that the fraction of the initial aerosolremaining in the atmosphere, m, is explicitly treated in the simplified expressions. The simplified r expressions for the spray decontamination coefficient are given. Parametric values for these expressions are found for median,10 percentile, and 90 pen:entile values in the uncertainty distribution for the spray decontamination coefficient. Examples are given to illustrate the utility of the simplified expressions to predict spray decontamination of an aerosol-laden atmosphere. Publication Date: June 1993 Prepared by: Powers, D.A. [Sandia National Labs., Albuquerque, NM (US)]; Burson, S.B. [ Nuclear i Regulatory Commission, Washington, DC (US), Div. of Safety Issue Resolution] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States) Keywords: aerosols, BWR type reactors, CONTAIN, containment, containment spray systems, decon-tamination, loss of coolant, Monte Carlo method, PWR type reactors, reactor accidents, scrubbing, statistical models, STCP NUREG/BR--0083, Vol.9 42

) UILU-ENG--92-2016;CDNSWC/SME-CR.-18-92 NUREG/CR--5970 l General

Title:

Approximate Techniques for Predicting Size Effects on Cleavcge Fracture Toughness (1,)

== Description:== This investigation examines the ability of an elastic T-stress analysis coupled with modified i boundary layer (MBL) solution to predict stresses ahead of a crack tip in a variety of plaratr geometries.'Ihe approximate stresses are used as input to estimate the effective d2iving force for cleavage fracture (J,) using the micromechanically based approach introduced by Dodds and Anderson. Finite element analyses for a wide variety of planar cracked geometries are j conducted which have elastic biaxiality parameters ( ) ranging from -0.99 (very low constraint) to +2.96 (very high constraint). The magnitude and sign of indicate therateat which crack-tip constraint changes with inemasing applied load. All results pertam to a moderately strain hardening material [ strain hardening exponent (11) of 10]. These analyses I suggest that is an effective indicator of both the accuracy of T-MBL estimates ofJ and of applicability limits on evolving fracture analysis methodologies (i.e., T-MBL,J-Q, and J/J,). Specifically, when I l>0.4 these analyses show that the T-MBL approximation of J,is accurate to within 20% of a detailed finite-element analysis. As " structural type" configura-tions (i.e., shallow cracks in tension) generally have 1 1>0.4,it appears that only an elastic analysis may be needed to determine reasonably accurate Jevalues forstructuralconditions. Publication Date: July 1993 Prepared by: Kirk, M.T.: Dodds, R.H. Jr. [ University ofIllinois, Urbana,IL (United States), Dept. of Civil Engineering) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering: Naval Surface Warfare Center, Annapolis, MD (United States); University of Illinois, Urbana, IL (United States), Dept. of Civil Engineering Keywords: boundary layers, crack propagation, cracks, finite element method, fracture mechanics, fracture properties, mathematical models, size, steels, stress analysis 1 43 NUREG/BR--0083, Vol.9

I l EGG--2694 NUREG/CR--5976 ' PRA

Title:

Development and Use cf a Train-Level Probabilistic Risk Assessment

== Description:== The Idaho National Engineering Laboratory examined the potential for the development of - train-level probabilistic risk assessment (PRA) databases.These train-level databases will i allow the Nuclear Regulatory Commission to investigate effects on plant core damage frequency (CDF) given a train is failed or taken out of service.'Ihe intent of this task was to~ develop user-friendly databases that required a minimal amount of personnelinvolvement to be usable. It was originally intended that the train-level models would not be expanded to j include basic events below the top gate of a train, with the possible exception ofincluding j some of the major train-related components (e.g.,important pumps and motor-operated valves). It was found that a database similar to the original plant PRA provided the accuracy needed to measure the changes in plant CDF, The Peach Bottom Unit 2 NUREG-1150 PRA ~ l (a large fault tree model) and the Beaver Valley Unit 2 IPE (a large event tree model) were selected to demonstrate the feasibility of developing train-level databases. Five different methods for developing train-level databases were hypothesized and are examined. Uhi-mately, two train-level databases were developed using the Peach Bottom Unit 2 PRA and one train-level database was developed using the Beaver Valley Unit 2 IPE. The develop-ment, use, limitations, and results of these train-level databases are discussed. l Publication Date: April 1993 Prepared by: Smith, C.L.; Fowler, R.D.; Wolfram, L.M. [EG and G Idaho, Inc., Idaho Falls,ID (United l States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Safety Issue Resolution: EG and G Idaho, Inc., Idaho Falls,ID (United States) Keywords: . Beaver Valley 2 reactor, database management, failure mode analysis, fault tree analysis, nuclear power plants, Peach Bottom-2 reactor, PRA, reactor components, risk assessment k I t i 4 h i I i ' NUREG/BR--0083, Vol.9 44

d l ' SAND--92-2688 NUKEG/CR--5978 I - General ) i i

Title:

Source Term Attenuation by Water in the Mark I Boiling Water Reactor Drywell "

== Description:== Mechanistic models of aerosol decontamination by an overlying water pool dunng core debris / concrete interactions and spray removalof aerosols from a MarkI drywell atmosphere -i are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are ~ identified. Ranges for values of parameters that characterize these uncertain features of the ' j F models are established. Probability density functions for values within these ranges are assigned according to a set ofrules. A Monte Carlo uncutainty analysis of the decontami, i nation factor produced by water pools 30 and 50 cm' deep'and subcooled 0 to 70 K is - performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25,0.01, and 0.001 cm H 0/cm -s and decontamination factors of1.1, 3 2 2 2,3.3,10,100, and 1000. L Publication Date: September 1993 Prepared by: Powers, D.A. [Sandia National Labs., Albuquerque, NM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re. search; Sandia National Labs., Albuquerque, NM (United States) Keywords: aerosols, attenuation, BWR type reactors, concretes, corium, data covariances, decontami-nation,heattransfer. hydraulics interactions,MonteCarlomethod, ponds.reactoraccidents, reactor safety, sampling, source terms, sprays 'J i l ,i ) I 7 .) ] -l P l I p s 45 NUREG/BR-@83, Vol.9 l y

=.

r . NUREG/CR-5983 BNL-NUREG--52355 ' THATCH

Title:

Safety Aspects of Forced Flow Cooldown Transients in Modular High Temperature Gas-Cooled Reactors

Description:

During some of the design basis accidents in Modular High Temperature Gas-Cooled Reactors (MHTGRs), the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS) are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. Ifeither forced flow cooling system becomes available. during such a transient, its restart could significantly reduce the downtime. This report used the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures.lf the reactoris serammed, either system can apparently be restarted at any time without exceeding any safelimits.However,under unscrammed conditions a restart offorced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits. Publication Date: May 1993 : Prepared by: Kroger, P.G. [Brookhaven Nationa! Lab., Upton, NY (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; Brookhaven National Lab.,Upton, NY (United States) Keywords: after-heat removal, criticality, design basis accidents, failures, heat transfer, HTGR type reactors, hydraulics, reactor accidents, reactor cooling systems, reactor safety, scram, THATCH, transients NUREG/BR--0083, Vol.9 46

i BNL-NUREG--52356_

NUREG/CR--5984 THATCH i

Title:

Code and Model Extensions of the THATCH Code for Modular High Temperature Gas-Cooled Reactors

== Description:== This report documents several model extensions and improvements of the THATCH code, a code to model thermal and fluid flow transients in High Temperature Gas-Cooled Reactors. A heat exchanger model was added, which can be used to represent the steam generator of the main Heat Transport System or the auxiliary Shutdown Cooling System. This addition permits the modeling of forced flow cooldown transients with the THATCH code. An enhanced upper head model, considering the actual conical and spherical shape of the upper i plenum and reactor upper head, was added, permitting more accurate modeling of the heat ..l tansfer in this region. The revised models are described,and the changes and additions to the input records are documented. Publication Date: May 1993 1 Prepared by: Kroger, P.G.; Kennett, R.J. [Brookhaven National Lab., Upton, NY (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; Brookhaven National Lab., Upton, NY (United States) Keywords: fluid flow, heat transfer, HTGR type reactors, hydraulics, reactor cooling systems, steam generators,THA'ICH, transients j .i -l \\ J I i I ? 1 1 l 47 NUREG/BR--0083, Vol.9

NUREG/CR--5991 CNWRA.-92-003 PORFLOW-

Title:

PORFLOW: AMultifluidMultiphaseModelforSimulatingFlow HeatTransfer,andMass Transport in Fractured Porous Media. User's Manual, Version 2.41

== Description:== The PORFLOW software package is designed to simulate flow, heat transfer, and mass transport in three-dimensional heterogenous porous and fractured media. Phase change and i gas phase flow is included. Radionuclide decay chains of up to four members can be included l in transport analyses. The mathematical basis of the model is described in Chapters 2 and 3, the code structure is discussed in Chapters 4 and 5, detailed instructions for the user are in 1 Chapter 6, and a few test problems are in Chapter 7. Publication Date: February 1993 Prepared by: Runchal, A.K. [ Analytic and Computational Research,Inc., Bel Air, CA (United States)); Sagar, B. [ Southwest Research Inst., San Antonio, TX (United States), Center for Nuclear Waste Regulatory Analyses] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; Analytic and Computational Research, Inc., Bel Air, CA (United States); Southwest Research Inst., San Antonio, TX (United States), Center for Nuclear Waste Regulatory Analyses Keywords: flow models, fluid flow, geologic fractures, heat transfer, high-level radioactive wastes, manuals, mathematical models, PORFLOW, porous materials, radiation tansport, radioac-tive waste disposal, radioisotopes L NUREG/BR--0083, Vol.9 48 f

1 l BNL-NUREG--52362-Vol.1 NUREG/CR--5993-Vol.1 General

Title:

Methods for Dependency Estimation and System Unavailability Evaluation Based on Failure Data Stadstics. Volume 1, Summary Report

== Description:== His report introduces a new perspective on the basic concept of dependent failures where the definition of dependency is based on clustering in failure times of similar components. This perspective has two significant implications: first,it relaxes the conventional assump-tion that dependent failures must be simultaneous and result from a severe shock; second,it allows the analyst to use all the failures in a time continuum to estimate the potential for multiple failures in a window of time (e.g., a test interval), therefore arriving at a more accurate value for system unavailability. In addition, the models developed here provide a method forplant-specific analysis of dependency, reflecting the plant-specific maintenance practices that reduce orincrease the contribution of dependent failures to system unavailabil-ity. The proposed methodology can be used for screening analysis of failure data to estimate the fraction of dependent failures among the failures. In addition, the proposed method can evaluate the impact of the observed dependency on system unavailability and plant risk. The formulations derived in this report have undergone various levels of validations through computer simulation studies and pilot applications. The pilot applications of these method-ologies showed that the contribution of dependent failures of diesel generators in one plant was negligible, while in another plant it was quite significant. It also showed tha:,in the plant with significant contribution of dependency to Emergency Power System (EPS) unavailabil-ity, the contribution changed with time. Similar findings were reported for the Containment Fan Cooler breakers. Drawing such conclusions about system performance would not have been possible with any other reported dependency methodologies. Publication Date: July 1993 Prepared by: Azarm, M.A.: Hsu, F.; Martinez-Guridi, G. [Brookhaven National Lab., Upton, NY (US)]: Vesely, W.E. [ Science Applications International Corp., Dublin, OH (US)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search: Brookhaven National Lab., Upton, NY (United States): Science Applications International Corp., Dublin, OH (United States) Keywords: failure mode analysis, failures, nuclear power plants, probabilistic estimation, reactor components, reactor safety, reliability, risk assessment, statistical models, system failure analysis, time dependence L 49 NUREG/BR--0083 Vol.9 i me

NUREG/CR -5993-Vol.2 BNL-NUREG-52362-Vol.2 General

Title:

Muhods for Dependency Estimation and System Unavailability Evaluation Based onFailure Data Statistics. Volume 2, Detailed Description and Applications

== Description:== nis report introduces a new perspective on the basic concept of dependent failures where the definition of dependency is based on clustering in failure times of similar components. This perspective has two significant implications first,it relaxes the conventional assump-tion that dependent failures must be simultaneous and result from a severe shock; second,it allows the analyst to use all the failures in a time continuum to estimate the potential for multiple failures in a window of time (e.g., a test interval), therefore arriving at a more accurate value for system unavailability. In addition, the models developed here provide a method for plant-specific analysis of dependency, reflecting the plant-specific maintenance practices that reduce orincrease the contribution ofdependent failures to system unavailabil-ity. The proposed methodology can be used for screening analysis of failure data to estimate the fraction of dependent failures among the failures. In addition, the proposed method can evaluate the impact of the observed dependency on the system unavailability and plant risk. The formations derived in this report have undergone various levels of validations through computer simulation studies and pilot applications.He pilot applications of these method-ologies showed that the contribution of dependent failures of diesel generators in one plant was negligible,while in another plant it was quite significant. It also showed that,in the plant with significant contribution of dependency to Emergency Power System (EPS) unavailabil-ity, the contribution changed with time. Similar Endings were reported for the Containment Fan Cooler breakers. Drawing such conclusions about system perfonnance would not have l been possible with any other reported dependency methodologies. Publication Date: July 1993 Prepared by: Azarm, M. A.: Hsu, F.', Martinez-Guridi, G. [Brookhaven National Lab., Upton,NY (United States)); Vesely, W.E. [ Science Applications International Corp., Dublin, OH (Umted States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; Brookhaven National Lab., Upton, NY (United States); Science Applications International Corp., Dublin, OH (United States) Keywords: availability, containment systems, failure mode analysis, failures, maintenance, nuclear power plants, perfonnance, power supplies, probabilistic estimation, reactor components, reliability, risk assessment, systems analysis NUREG/BR--0083, Vol.9 50

.=. PNL -8496 NUREG/CR-5998 1 MSTS, VAM2D y \\

Title:

Simulation of Unsaturated Flow and Nonreactive Solute Transport in a Heterogeneous Soil at the Field Scale

Description:

A field-scale, unsaturated flow and solute transport experiment at the Las Cmces trench site in New Mexico was simulated as part of a " blind" modeling exercise to demonstrate the ability or inability of uncalibrated models to predict unsaturated flow and solute transport in spatially variable porous media. Simulations were conducted using a recently developed - multiphase flow and transport simulator. Uniform and heterogeneous soil models were tested, and data from a previous experiment at the site were used with an inverse procedure to estimate water retention parameters. A spatial moment analysis was used to provide a quantitative basis for comparing the mean observed and simulated flow and transport behavior.The results of this study suggest that defensible predictions of waste migmtion and ' fate at low-level waste sites will ultimately require site-specific data for model calibration. Publication Date: February 1993 J - Prepared by: Rockhold, M.L. [ Pacific Northwest Lab., Richland, WA (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; Pacific Northwest Lab., Richland, WA (United States) Keywords: computerized simulation, field tests, flow models, fluid flow, low-level radioactive wastes, MSTS, porous materials, radioactive waste disposal,radionuclide migration, soils, VAM2D .l 2 i 1 J m 1 51 NUREG/BR--0083, Vol.9

[ t NUREG/CR--6018 EPRI-TR--102106;SAIC--91/6660 { General i i

Title:

Survey and Assessment of Conventional Software Verification and Validation Methods l l

Description:

By means of a literature survey, a comprehensive set of methods was identified for the verification and validation of conventional software. The 134 methods so identified were i classified according to their appropnateness for various phases of a developmental life j cycle-requirements, design, and implementation; the last category was subdivided into two, static testing and dynamic testing methods. The methods were then characterized in termsofeightratingfactors,fourconcerningease of useof themethodsandfourconceming the methods' power to detect defects. Based on these factors, two measurements were d developed to permit quantitative comparisons among methods, a cost-benefit metric and an ) effectiveness metric. The effectiveness metric was further refined to provide three different estimates for each method, depending on three classes of needed stringency of V&V (determined by ratings of a system's complexity and required integrity). Methods were then rank ordered for each of the three classes in tenns of their overall cost-benefits and effectiveness. The applicability of each method was then assessed for the four identified components of knowledge-based and expert systems, as well as for the system as a whole. ) Publication Date: April 1993 Prepared by: Miller, L.A.; Groundwater, E.; Mirsky, S.M. [ Science Applications International Corp., l Reston,VA(United States)] l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div, of Systems Re-j search; Science Applications Intemational Corp., Reston, VA (United States) Keywords: cost-benefitanalysis defects design. evaluation.expertsystems.generallifecycle. nuclear power plants, programming, surveys, validation j q l 1 i 1 4 NUREG/BR -0083, Vol.9 52'

E CNWRA. 92-011 NUREG/CR--6021 General '

Title:

A Literature Review of Coupled Thermal-Hydrologic-Mechanical-Chemical Processes Peninent to the Proposed High-Level Nuclear Waste Repository at Yucca Mountain

== Description:== A literature review has been conducted to determine the state of knowledge available in the modeling of coupled thermal (T), hydrologic (H), mechanical (M), and chemical (C) processes relevant to the design and/or performance of the proposed high-level waste (HLW) repository at Yucca Mountain, Nevada. The review focuses on identifying coupling mecha-nisms between individual processes and assessing their importance (i.e., if the coupling is either important, potentially important, or negligible). The significance of considering THMC-coupled processes lies in whether or not the processes impact the design and/or perfonnance objectives of the repository. A review, such as reported here, is useful in identifying which coupled effects will be imponant, hence which coupled effects will need to be investigated by the US Nuclear Regulatory Commission in order to assess the assumptions, data, analyses, and conclusions in the design and performance assessment of a geologic repository. Although this work stems from regulatory interest in the design of the geologic repository,it should be emphasized that the repository design implicitly considers all of the repository performance objectives, including those associated with the time after permanent closure. The scope of this review is considered beyond previous assessments in that it attempts with the current state-of-knowledge to determine which couplings are imponant and identify which computer codes are currently available to model coupled processes. Publication Date: July 1993 Prepared by: Manteufel, R.D.: Ahola, M.P.: Turner, D.R.: Chowdhury, A.H. [ Southwest Research Inst., San Antonio,TX (United States), Center for Nuclear Waste Regulatory Analyses] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of High-Level Waste Management: Southwest Research Inst., San Antonio,TX (United States), Center for Nuclear Waste Regulatory Analyses Keywords: 3DEC, ABAQUS, ADINA, ANSYS, BEASY, chemical reaction kinetics, CHEMTRN, CHEQMATE, CTM, DYNAMIX, ECHEM, EQ 3/6, FASTCHEM, FEHMN, FLAC, fluid flow, FRAC-UNIX, GENASYS, GEOCHEM. GEOTHER, GPBEST 3D, high-level radio-active wastes,HYDRAQL,HYDROGEOCHEM, hydrology,mechanicalpropenies,MPATH, MSC/NASTRAN, NEFI'RAN, NORIA, PETROS, PHREEQE, PORFLOW, radioactive waste disposal, radioactive waste management,radionuclide migration,ROCMAS,SANGRE, SPECTROM-32, STEALTH, STRES3D, SUPCRT, THAMES, THCC, thermal analysis, thermochemical processes, thermodynamics, TOUGH, TRACR3D, TRANQL, UDEC, VTOUGH, WATEQ, Yucca Mountain 53 NUREG/BR--0083, Vol.9 i

l ) a l l- ' NUREG/CR--6026 CNWRA--92-006 - General

Title:

Theoretical and Experimental Investigation of Thermohydrologic Processes in a Partially Saturated, Fmetured Porous Medium

== Description:== The performance of a geologic repository for high-level nuclear waste will be influenced to a large degree by thermohydrologic phenomena created by the emplacement of heat-- generating radioactive waste.The importance of these phenomenais manifestin that the um greatly affect the movemcat of moisture and the resulting transport of radionuclides from the repository. Thus, these phenomena must be well understood prior to a definitive assessment - l of a potential repository site. An investigation has been undertaken along three separate l avenues of analysis: (I) laboratory experiments, (II) mathematical models, and (III) simili-i tude analysis. A summary of accomplishments to date includes (I) a review of the literature i on the theory of heat and mass transfer in partially saturated porous medium, (2) a development of the goveming conservation and constitutive equations, (3) a development of a dimensionless form of the governing equations (4) a numerical study of the importance and sensitivity of flow to a set of dimensionless groups, (5) a survey and evaluation of experimental measurement techniques, and (6) execution of laboratory experiments of nonisothermal flow in a porous medium with a simulated fracture. Publication Date: July 1993 Prepared by: Green, R.T.: Manteufel, R.D. [ Nuclear Regulatory Commission, Washington, DC (United l S tates), Div. of Regulatory Applications]; Dodge, F.T.: S vedeman, S J. [ Southwest Re search Inst., San Antonio,TX (United States), Center for Nuclear Waste Regulatory Analyses] l Prepared for: Nr: lear Regulatory Commission, Washington, DC (United States) Div. of Regulatory Applications: Southwest ResearchInst. San Antonio,TX(United States),CenterforNuclear - - Waste Regulatory Analyses Keywords: Darcy's Law, FEiniN, fluid flow, fractures, FRAC-UNIX, GEOTHER, GWHRT, heat transfer, high. level radioactive wastes, hydrology, LSODES, mathematical models, j MA'ITUM, NORIA, OCM3D, PETROS, PORFLOW, porous materials, radiation heating, j radioactive waste disposal, radioactive waste facilities, radioactive waste storage, sirnula-tion, thermodynamics, TOUGH, two-phase flow, VTOUGH l i a r l. NUREG/BR--0083, Vol.9 54

CNWRA--92-026 NUREG/CR--6028 BIGFLOW, DATAFLOW, PORFLOW, CMVSFS

Title:

BIGFLOW: A Numerical Code for Simulating Flow in Variably Saturated, Heterogeneous Geologic Media. Theory and User's Manual, Version 1.1

== Description:== This report documents BIGFLOW 1.1, a numerical code for simulating flow in variably saturated heterogeneous geologic media. It contains the underlying mathematical and numerical models, test problems, benchmarks, and applications of the BIGFLOW code. The BIGFLOW software packageis composed of a sim ulation and an interactive data processing code (DATAFLOW). 'Ihe simulation code solves linear and nonlinear porous media flow equations based on Darcy's law, appropriately generalized to account for 3D, deterministic, orrandom heterogeneity. A modifiedPicard Schemeis used forlineartzmg unsaturated flow - equations, and preconditioned iterative methods are used for solving the resulting matrix systems. The data processor (DATAFLOW) allows interactive data entry, manipulation,and analysis of 3D data sets. The report contains analyses of computational performance carried out using Cray-2 and Cray-Y/MP8 supemomputers. Benchmark tests include comparisons with other independently developed codes, such as PORFLOW and CMVSFS, and with analytical or semi-analytical solutions. Publication Date: June 1993 Prepared by: Ababou, R. [CEA Centre d' Etudes de Saclay,91-Gif-sur-Yvette (France)]; Bagtzoglou, A.C. [ Southwest Research Inst., San Antonio,TX (United States), Center for Nuclear Waste Regulatory Analyses] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; CEA Centre d' Etudes de Saclay,91-Gif sur-Yvette (France); Southwest Research Inst., San Antonio, TX (United States), Center for Nuclear Waste Regulatory Analyses Keywords: algorithms, benchmarks, BIGFJ OW, CMVSFS, computerized simulation, DATAFLOW, fluid flow, geologic formations, hydrology, iterative methods, PORFLOW, radioactive waste disposal, radioactive waste storage, saturation 55 NUREG/BR--0083, Vol.9

/ i i NUREG/CR--6032 ANL--93/9 f ~ CORCON-MOD 2, MELCOR l i Tille: Sv.idus and Liquidus Temperatures of Core-concrete Mixtures

== Description:== soldus and liquidas temperatures were measured for four types of concrete (limestone, limesme sand, basalt, and siliceous) and for their mixtures with urania and zirconia. Differential 6ermal analysis (DTA) was generally employed to determine the solidus and liquidus temperatures. However, some liquidus temperatures were also measured by rota-tional viscometry. The measured solidus temperatures for the urania-zirconia-conc tte mixtures were significantly lower (hundreds of degrees) than those employed in the CORCON-MOD 2 thermal hydraulic code, and the measured liquidus temperatur:nere significantly higher (also hundreds of degrees). The liquidus temperatures for urama-zirconia-concrete mixtures containing limestone or limestone-sand concrete were generally above 2850 K,which was the upper temperaturelimit ofour experiments.The revised solidus and liquidus temperatures are to be incorporated in the CORCON-MOD 3 thermal hydraulic code which is an integral part of the US Nuclear Regulatory Commission's MELCOR code. l t MELCOR computes the consequences of severe accidents at nuclear reactors. DTA was also employed to redetermine the calcia-urania (CaO-UO ) phase diagram. Earlier data were not 2 in agreement, and this binary phase diagram is required in computer programs that calculate ,t the phase diagrams (and solidus and liquidus temperatures) of urania-zirconia-concrete systems from the phase diagrams of simpler systems. A cutectic temperature and composi-tion of 221815 K and 37 mol% UO, respectively, were determined for the CaO-UO 2 2 system. The solubility of Ca0 in UO at the eutectic temperature was 30 mol% CaO. 2 Publication Date: June 1993 Prepared by: Roche, M.F.; Leibowitz, L.; Fink, J.K.; Baker, L. Jr. (Argonne National Lab., IL (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div.'of Systems Re-search: Argonne National Lab.,IL (United States) Keywords: concretes, CORCON-MOD 2 corium, interactions, MELCOR, mixtures, phase studies, phase transformations, reactor accidents, temperature distribution, thermal analysis, thermo-dynamics, uranium oxides, water cooled reactors, zirconium oxides i i l .NUREG/BR--0083 Vol.9 56 E

=.. NUREG/CR--6035 r RELAP5

Title:

Feasibility Study for Improved Steady-State Initialization Algorithms for the RELAPS Computer Code. Phase I Improved Str ly-State Initialization Algorithms for Computer Codes

== Description:== A design for a new steady-state initialization method is presented that represents an improvement over the current method used in RELAP5. Current initialization methods for RELAP5 solve the transient flu d flow balance equations simulating a transient to achieve i steady-state conditions. Because the transient solution is used, the initial conditions may change from the desired values requiring the use of controllers and long transient running times to obtain steady-state conditions for system problems. The new initialization method allows the user to fix thermal-hydraulic values in volumes andjunctions where the conditions are best known and have the code compute the initial conditions in other areas of the system. 'Ihe steady-state balance equations and solution methods an: presented. The constitutive, component, and special pu' pose models are reviewed with respect to modifications required for the new steady-state initialization method. The requirements for user input are defined and the feasibility of the method is demonstrated with a testbed code by initializing some simple channel problems.Theinitialization of the sample problems using the old and the new methods is compared. Publication Date: April 1993 Prepared by: Paulsen,M.P.:Peterson C.E.:Katsma,K.R.[ComputerSimulationandAnalysis,Inc Idaho Falls,ID (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search: Computer Simulation and Analysis, Inc., Idaho Falls, ID (United States) Keywords: BWR type reactors, computer program documentation, flow models, fluid flow, heat transfer, hydraulics, PWR type reactors, reactor cooling systems, reactor kinetics, reactor safety, RELAP5, steady-state conditions, steam generators, transients 4 57 NUREG/BR--0083 Vol.9

NUREG/CR--6041 BNL-NUREG--52375 DUST, DUSTIN, GRAFXT l

Title:

Disposal Unit Source Term (DUST) Data Input Guide

== Description:== Performance assessment of a low-level waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). He focus of this workis to develop a methodology for calculating the source term.In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). He computer code DUST (Disposal Unit Source Term) has been developed to model these processes. His document presents the models used to calculate release from a disposal facility, verification of the model, and instructions on the use of the DUST code. In addition to DUST, a preprocessor, DUSTIN, which helps the code user create input decks for DUST, r and a post-processor, GRAFXT, which takes selected output files and plots them on the computer terminal, have been written.Use of these codes is also described. Publication Date: May 1993 Prepared by: S ullivan, T.M. [Brookhaven National Lab., Upton, NY (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Low-Level Waste Management and Decommissioning: Brookhaven National Lab., Upton, NY (United States) t Keywords: containers, DUST, DUSTIN, GRAFXT, performance testing, radiation doses, radioactive waste disposal, radioactive waste facilities, radioactive waste management, radionuclide migration, source tenns, underground disposal t i' t ) NUREG/BR483, Vol.9 58

BNL-NUREG--52377 NUREG/CR--6049 PSAFE2

Title:

Piping Benchmark Problems for the General Electric Advanced Boiling Water Reactor l

== Description:== To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representa-tive piping systems subjected to representative dynamic loads with solutions developed using I the methods being proposed for analysis for the advanced reactor standard design. lt will be required that the combined license holders demonstrate that their solutions to these problems - are in agreement with the benchmark problem set. - Publication Date: August 1993 i Prepared by: Bezier P.;DeGrassi,G.;Braverman,J.; Wang,Y.K.[BrookhavenNationa! Lab.,Upton,NY 3 (US)] l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering; Brookhaven National Lab., Upton, NY (United S tates) Keywords: benchmarks, BWR type reactors, computer calculations, dynamic loads, pipes, PSAFE2, j -) reactor licensing, reactor safety, stress analysis i i 1 59 NUREG/BR--0083, Vol.9

= NUREG/CR--6050 SAIC--93/1310-01 REMIT

Title:

Radiation Exposure Monitoring and Information Transmittal (REMIT) System. User's Manual

== Description:== The Radiation Exposure Monitoring and Information Transmittal (REMIT) system is designed to assist US Nuclear Regulatory Commission (NRC) licensees in meeting the reporting requirements of the revised 10 CFR 20 and in agreement with the guidance contained in R.G. 8.7, Rev.1, " Instructions for Recording and Reporting Occupational Exposure Data." REMIT is a personal computer (PC) based menu driven system that facilitates the manipulation of database files to record and report radiation exposure information. REMIT is designed to be user-friendly and contains the full text of R. G. 8.7, Rev.1, on-line as well as context-sensitive help throughout the program. The user can enter data directly from NRC forms 4 or 5. REMIT allows the user to view the individual's exposure in relation to regulatory or administrative limits and alerts the user to exposures in excess of these limits. The system also provides for the calculation and summation of dose from intakes and the determination of the dose to the maximi.11y exposed extremity for the monitoring year. REMIT can produce NRC forms 4 and 5 in paper and electronic format and can import / export data from ASCII and database files. Publication Date: June 1993 Prepared by: Cale, R.; Clark, T.; Dixson, R.; Hagemeyer, D. [ Science Applications International Corp., Oak Ridge,TN(United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; Science Applications Intemational Corp., Oak Ridge,TN (United States) Keywords: computer graphics, data processing, dosimetry, licensing, occupational exposure, personnel monitoring, radiation doses, regulations, REMIT, reporting requirements, US NRC i NUREG/BR--0083, Vol.9 60

-l ORNL/Sub--93-SD684 NUREG/CR--6052 General i i

Title:

Methodology for Reliability Based Condition Assessment. Application to Concrete Struc-turesin Nuclear Plants

Description:

Structures in nuclear power plants may be exposed to aggressive environmental effects that - l cause their strength to decrease over an extended period of service. A major concern in evaluating the continued service for such struetures is to ensure that in their current condition they are able to withstand future extreme load events during the intended service life with a level of reliability sufficient for public safety. This report describes a methodology to ] facilitate quantitative assessments ofcurrent and fature structuralreliability and performance ] of structures in nuclear power plants. This methodology takes into account the natum of past and futureloads and randomness in strength and in degradation resulting from environmental -j factors. An adaptive Monte Carlo simulation procedure is used to evaluate time-dependent system reliability. The time-dependent reliability is sensitive to the time-varying load characteristics and to the choice of initial strength and strength degradation models but not to correlation in component strengths within a system. Inspection / maintenance strategies are identified that minimize the expected future costs of keeping the failure probability of a structure at or below an established target failure probability during its anticipated service period. Publication Date: August 1993 Prepared by: Mori, Y.: Ellingwood, B. [ Johns Hopkins Univ., Baltimore, MD'(United States), Dept. of CivilEngineering] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering; OakRidgeNationalLab.,TN(UnitedStates);JohnsHopkinsUniv., Baltimore,MD(United - States), Dept.of Civil Engineering Keywords: dynamic loads, general, inspection, maintenance, mechanical stmetures, Monte Carlo i method, nuclearpowerplants. performance,probabilistic estimation, reliability, staticloads, stresses, time dependence -l 1 i 61 NUREG/BR--0083, Vol.9

' NUREG/CR--6054 PNL--8497 CECP'

Title:

Estimating Pressurized Water Reactor Decommissioning Costs: A User's Manual for the PWR Cost Est mating Computer Program (CECP) Software. Draft Report for Comment

== Description:== With the issuance of the Decommissioning Rule (July 27,1988), nuclear power plant i lictnsees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user's manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to j the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point oflicense termination. S uch cost estimates includecomponent piping,andequipmentremovalcosts;packagingcosts; decontamination j costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning. Publication Date: October 1993 Prepared by: Bierschbach, M.C.; Mencinsky, GJ. [ Pacific Northwest Lab., Richland, WA (United States)] Papared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; Pacific Northwest Lab., Richland, WA (United States) Keywords: CECP, cost estimation, decommissioning, decontamination, PWR type reactors, radioactive waste disposal, reactor components 4 4 l t I NUREG/BR--0083.Vol.9 62 l t

NUREG/CR--6056 - General

Title:

A Framework for the Assessment of Severe Accident Management Strategies

== Description:== Severe accident management can be defined as the use of existing and/or altemative j resources, systems and actors to prevent or mitigate a core-melt accident. For each accident 1 sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and opera-tional considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model j uncertainty. Itis applied to several potential strategies, including PWR cavity flooding,BWR drywell flooding, PWR depressurization, and PWR feed and bleed. ' Publication Date: September 1993 . Prepared by: Kastenberg, W.E. [ed.]: Apostolakis, G.: Dhir, V.K. [ California Univ., Los Angeles, CA (United States), Dept. of Mechanical, Aerospace and Nucles Engineering] [and others] l 1 Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search:CalifomiaUniv.,IAs Angeles,CA(UnitedStates), Dept.ofMechanical, Aerospace and Nuclear Engineering Keywords: AB AQUS, BWR type reactors, diagrams, failure mode analysis, FEM, heat transfer, human factors, hydraulics, MACCS, management, MARCH, MELPROG/ MODI, meltdown, miti-gation, NASTRAN, PARTITION, PRAMIC, probability, PWR type reactors, reactor safety,-- risk assessment, SCDAP/RELAP, SUPERTREE, XSOR ) 63 NUREG/BR--0083, Vol.9

r NUREG/CR--6059 SAND--92 2146 MACCS

Title:

MACCS Version 1.5.11.1: A Maintenance Release of the Code

== Description:== ' A new version of the MACCS code (version 1.5.11.1)has been developed by Sandia National Laboratories under sponsorship of the US Nuclear Regulatory Commission. MACCS was developed to support evaluations of the off-site consequences from hypothetical severe accidents at commercial power plants. MACCS is the only current public domain code in the United S tates that embodies all of the following modeling capabilities: (1) weather sampling i using a year of recorded weather data; (2) mitigative actions such as evacuation, sheltering, relocation, decontamination, and interdiction; (3) economic costs of mitigative actions; (4) cloudshine, groundshine, and inhalation pathways as well as food and water ingestion; (5) i 4 calculation of both individual and societal doses to various organs; and (6) calculation ofboth acute (nonstochastic) and latent (stochastic) health effects and risks of health effects. All of the consequence measures may be generated in the form of a complementary cumulative distribution function (CCDF). The current version implements a revised cancer model consistent with recent reports such as BEIR V and ICRP 60. In addition, a number of error corrections and portability enhancements have been implemented.This report describes only the changes made in creating the new version. Users of the code willneed to obtain the code's original documentation. NUREG/CR-4691. i 4 Publication Date: October 1993 Prepared by: Chanin. D.; Foster, J. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States)]; Rollstin, J. [ GRAM, Inc., Albuquerque, NM (United States)]; Miller, L. [Sandia National Labs., Albuquerque, NM (United States)) Prepared for: Noclear Regulatory Commission, Washington, DC (United States), Div of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States) I Keywords: computer program documentation, cost, economics, emergency plans, evacuation, fission product release, health hazards, MACCS, maintenance, mitigation, modifications, nuclear power plants, radiation doses, radiation transport, reactor accidents, reactor safety L b NUREG/BR--0083, Vol.9 64

J L .LA--12593-M NUREG/CR--6060 HMS

Title:

Hydrogen Mixing Studies (HMS) Assessment Manual

== Description:== This report documents some calculations performed to assess the Hydrogen Mixing Studies (HMS) code. Results are presented first for some analytical test problems, including laminar. l flow and mass diffusion. The von Karman vortex street problem and the Sandia FLAhE Facility and Heiss Dampf Reaktor (HDR) containment facility test problems are then i discussed. For the analytical problems, the code gave results that agree exceptionally well ? with the analytical solutions. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations, specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. Calculations for the von J Karman vortex street problem were performed at selected Reynolds numbers for several I obstacle types. The computed flow pattems agree well with experimental observations, specifically the occurrence of a vortex street (double row of vonices) above a critical Reynolds number. The last assessment problem involves modeling the experiment T31.5. 'Ite experiment was carried out in the HDR containment building, which is a large, 3 multicompartment facility (11300 m free volume in 72 compartments). In the experiment, a steam-water mixture was first injected into the containment to simulate a large-break i blowdown of a pressure vessel, and then superheated steam was injected that was followed by a release of helium-hydrogen light gas.The calculated results (pressure, temperature,and gas concentrations) agree reasonably well with the experimental data. Publication Date: June 1993 Prepared by: Lam, K.L.; Wilson, T.L. [Los Alamos National Lab., NM (United States)); Travis, J.R. [ Science Applications Intemational Corp., Albuquerque, NM (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; Los Alamos National Lab., NM (United States) Keywords: blowdown. chemicaireaction kinetics.chemicai reactions. combustion.computenzed simu-teon, containment systems, diffusion, HMS, hydrogen, laminar flow, mixing, nuclear powt: plants, turbulent flow i 65 NUREG/BR--0083, Vol.9

l NUREG/CR--6061 EGG--2610 FRAP-T6 l

Title:

Determination of the Bias in LOFT Fuel Peak Cladding Temperature Data from the Blowdown Phase of Large-Break LOCA Experiments

== Description:== Data from the Loss-of-Fluid Test (LOFr) Program help quantify the margin of safety inherentin pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose conceming the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. His report artalyzes how well extemally mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding smface temperature during large-break LOCA experiments.In particular,the I validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the extemally mounted thermocouples might have influenced cladding temperature.The analy-sis makes use of data and information from several sources, including later, similar LOFT experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a eme-wide cladding quench did occur in expenments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 i 16.2K(20.5 29.2'F).Thebest<sumatevalueofpeakcladdmgtemperatureforLOFTLP&6 is 1,104.8 K. The best estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K. l Publication Date: May 1993 l Prepared by: Berta. V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R. [EG and G Idaho, Inc.. Idaho Falls, ID(United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search: EG and G Idaho, Inc., Idaho Falls,ID (United States) Keywords: accuracy, blowdown, data analysis, FRAP-T6, fuel cans, fuel rods, LOFT reactor, loss of j coolant,PWR type reactors, reactor safety, reactor safety experiments, temperature measure-1 ment, test facilities, thermocouples l l i l l 1 NUREG/BR--0083, Vol.9 66 4 J

i EGG--2701 NUREG/CR--6070 CEMENT t

Title:

Modeling Approaches for Concrete Barriers Used in Low-Level Waste Disposal

== Description:== A series of three NUREGs and several papers addressing different aspects of modeling performance ofconcrete barriers forlow-leve' radiosctive waste disposalhave been prepared previously for the Concrete Barriers Research Project. This document integrates the i information fmm the previous documents into a general summary ofmodels and approaches that can be used in performance assessments of concrete barriers. Models for concrete degradation, flow, and transport through cracked concrete barriers are discussed.The models for flow and transport assume that cracks have occurred and thus should only be used forlater times in simuladons after fully penetrating cracks are formed. Most of the models have been implemented in a computer code, CEMENT, that was developed concurrently with this document. User documentation for CEMENTis provided separate from this report. To avoid duplication, the reader is referred to the three previous NUREGs for detailed discussions of each of the mathematical models. Some additionalinformation that was not presented in the previous documents is also included. Sections discussing lessons learned from applications to actual performance assessments of low-level waste disposal facilities are provided. Sensitive design parameters are emphasized to identify critical areas of performance for concrete barriers, and potential problems in performance assessments are also identified and discussed. i Publication Date: November 1993 Prepared by: Seitz, R.R.: Walton, J.C. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications: EG and G Idaho, Inc., Idaho Falls,ID (United States) Keywords: CEMENT, concretes, containment, containment systems, cracks, low-level radioactive wastes, radioactive waste disposal i i 1 i i i l 67 NUREG/BR--0083, Vol.9

l K i 1. l ? - NUREG/CR.-6071 ORNLfrM -12406 r ENDF/B-VI e

Title:

Impact of ENDF/B-VI Cross-Section Data on R B. Robinson Cycle 9 Dosimetry Calcula-tions

== Description:== Dosimeters that were removed from the R B. Robinson reactor following Cycle 9 were analyzed and compared with calculated results in an earlier study. This work updates the. calculation using recently available ENDF/B-VI data in order to assess advantages to using the newer cross sections in reactor pressure vessel fluence calculations. A comparison is also made to determine the impact of various cross-section libraries on computed dosimeter activities.SignificantimprovementsareobtainedwiththeENDF/B VIcrosssections.Other factors, such as differences in group structures of multigroup libraries, may also affect the { calculated dosimeter activities. Publication Date: October 1993 Prepared by: Williams, M1.; Asgari, M. iLouisiana State Univ., Baton Rouge, LA (United States), Nuclear Science Center); Kam, F.B.K. [ Oak Ridge National Lab., TN (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Engineering; Oak Ridge National Lab., TN (United States) Keywords: cross sections, dosimetry, ENDF/B-VI, neutron fluence, nuclear data collections, pressure vessels, Robinson-2 reactor 1 l i 1 i NUREG/BR--0083, Vol.9 68 i 1

=.. RRCKI--80-05/3;VARGOS--93/1 NUREG/CR--6072 General

Title:

Experimental Study on the Combustion Behavior of Hydroger-Air Mixtures with Turbulent JetIgnition at Large Scale

== Description:== ne formation and release of large amounts of hydrogen can accompany the course of a severe accident in a nuclear power plant. After the hydrogen is released into the containment, - i it is ruixed and transported by natural and forced convection. If a sufficient amount of hydrogen accumulates and becomes flammable, different combustion modes may be possible: deflagrations, accelerated flames, and detonations. This report describes research i carried out in the KOPER facility on spontaneous detonation initiation in hydrogen-air - mixtures by turbulent jet ignition. The KOPER facility is a large semiconfined volume - designed to investigate conditions of explosion initiation and propagation in gaseous fuel. air mixtures. The effects of three variables were investigated: hydrogen concentration, jet orifice diameter, and the composition of combustion products in the turbulent jet. The possibility ofinitiation of two characteristic combustion regimes, turbulent combustion and local detonation, was demonstrated. Local detonation develops after a delay of 10 to 25 ms ' from ignition. Both high enough hydrogen concentration and large enough jet size are necessary for spontaneous detonation initiation. The minimum hydrogen concentration is within the range of 20 to 25 vol %, and the minimum jet orifice diameter lies in the range 100 to 200 mm for the given geometrical sizes and configuration of the facility.The existence of an optimum concentration of hot product gases was observed. A minimum ratio of turbulentjet size L and mixture detonation cellwidth A,IA = 12-13 is equired for detonatic:- initiation. This minimum value corresponds to that measured for other types of turbulentjet initiation experiments (closed volume and continuous venting) and is supported by theoreti-cal analysis. Publication Date: June 1993 Prepared by: Dorofeev, S.B.; Bezmelnitsin, A.V.; Efimenko, A.A.; Kochurko, A.S.; Sidorov, V.P.; Yankin, J.G.; Matsukov, I.D. [ Russian Research Center, Moscow (RU), Kurchatov Inst.) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Systems Re-search; Nuclear Regulatory Commission, Washington, DC (United States), Office of NuclearRegulatory Research Keywords: combustion kinetics, combustion products, data acquisition systems, detonations, experi-mental data, flammability, fuel-air ratio, general, hydrogen, hydrogen production, ignition, mathematical models, measurmg instruments, orifices, reactor accidents, ::pontaneous combustion, test facilities, turbulence, ultrahigh-speed photography 69 NUREG/BR--0083, Vol.9 r

D NUREG/CR--6082 UCRL-ID--114567 General

Title:

Data Communications

== Description:== The purpose of this paper is to recommend regulatory guidance for reviewers examining computer communication systems used in nuclear powerplants.Therecommendations cover three areas imponant to these communications systems: system design, communication protocols, and communication media. The first ama, system design, considers three aspects of system design-questions about architecture, specific risky design elements or omissions - to look for in designs being reviewed, and recommendations for multiplexed data commu-nication systems used in safety systems. The second area reviews pertinent aspects of communication protocol design and makes recommendations for newly designed protocols or the selection of existing protocols for safety system,information display, and non-safety control system use. The third area covers communication media selection, which differs significantly from traditional wire and cable. The acommendations for communication media extend or enhance the concerns of published IEEE standards about three subjects: data rate, imported hazards and mainta' ability. m Publication Date: August 1993 Prepared by: Preckshot, G.G. [ Lawrence Livermore National Lab., CA (United States)] i I Prepared for: NuclearRegulatoryCommission, Washington DC(UnitedStates),Div.ofReactorControls and Human Factors; Lawrence Livermore National Lab., CA (United States) Keywords: computerized control systems, data' transmission, data transmission systems, hazards. information systems,Intemet, maintenance, nuclear power plants, performance testing, reat 4 time systems I t NUREG/BR--0083, Vol.9 70

I l UCRL-ID--114565 NUREG/CR--6083 General

Title:

Reviewing Real-Time Performance of Nuclear Reactor Safety Systems

== Description:== The purpose of this paperis torecommend regulatoryguidance forreviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific probleras have been solved or that performance goals have been met.The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing - expected or actuai real-time performance before,during,and after developmentis completed. To ensure that the delivered real-time software product meets performance goals, the paper ~ recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time softwam development approaches, modeling and measurement, and real-time operating systerns. l Publication Date: August 1993 Prepared by: Preckshot, G.G. (Lawrence Livermore National Lab., CA (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United S tates), Di v. ofReactor Controls' and Human Factors; Lawrence Livermore National Lab., CA (United States) l Keywords: CoCoMo, computerized contro! systems, engineered safety systems, nuclear powerplants, on-line measurement systems, performance testing, real time systems, task scheduling i r 71 NUREG/BR--0083, Vol.9 - mwd L F

NUREG/CR--6090' UCRL-ID--112900 General } i

Title:

The Programmable Logic Controller and its Application in Nuclear Reactor Systems

== Description:== nis document provides recommendations to guide reviewersin the application of Program-I mable Logic Controllers (PLCs) to the control, monitoring and protection of nuclearreactors. The first topics addressed are system-level design issues, specifically including safety. The document then discusses concerns about the PLC manufacturing organization and the ~I protection system engineering organization. Supplementing this document are two appendi-ces. Appendix A summarizes PLC characteristics. Specifically addressed are those charac-l teristics that make the PLC more suitable for emergency shutdown systems than other - i electrical / electronic-based systems, as well as characteristics that improve reliability of a system. Also covered are PLC characteristics that may create an unsafe operating environ-i ment. Appendix B provides an overview of the use of programmable logic controllers in i emergency shutdown systems. He intent is to familiarize the reader with the design, development, test,and maintenance phases of applying a PLC to an ESD system. Each phase j is described in detail and information pertinent to the application of a PLC is pointed out. I Publication Date: September 1993 Prepared by: Palomar, J.; Wyman, R. [ Lawrence Livermore National Lab., CA (United States)) l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States),Div.ofReactor Controls and Human Factors; Lawrence Livermore National Lab., CA (United States) Keywords: computer architecture, computerized control systems, electronic equipment, engineered - safety systems, maintenance, nuclear power plants, PLC, programming, reactor monitoring i systems, reactor protection systems, recommendations i Y h l-t l i l. i ) 5 l NUREG/BR--0083,Vol.9 72 l

\\ l UCRL-ID--114839 NUREG/CR--6101 FMEA, FMECA

Title:

Software Reliability and Safety in Nuclear Reactor Protection Systems

== Description:== Planning the development, use, and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report: first, software considerations cannot be fully understood in isolation from computer hardware and applica-tion considerations; and second, the process of engineering reliability and safety into a computer syst.m requires activities to be carried out throughout the softwate life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is pnmarily that of the assessor or auditor. Publication Date: November 1993 Prepared by: Lawrence, J.D. [ Lawrence Livermore National Lab., CA (United States)] Prepared for: Nuclear Regulatory Commission, Washington,DC(United S tates),Div. ofReactor Controls - and Human Factors; Lawrence Livermore National Lab., CA (United States) Keywords: fault tolerant computers,FhEA,FMECA, life cycle models,Markov models, nuclear power plants, Petri Nets, reactor protection systems, reactor safety, reliability > I i 1 1 73 NUREG/BR--0083 Vol.9

m _. -_ . r. l wg l 4.- .NUREG/CR--6113. i General' l l i

Title:

Class IE Digital Systems Studies - l ?

== Description:== This document is furnished as part of the effort to develop NRC Class IE Digital Computer Systems Guidelines which is Task 8 of USAF Rome Laboratories Contract F30602-89-D-0100. The report addresses four major topics, namely, computer programming languages, software design and development, software testing and fault tolerance, and fault avoidance. The topics are intended as stepping stonesleading to a Draft Regulatory Guide document. As part of this task a small-scale survey of software fault avoidance and fault tolerance practices 4 was conducted among vendors of nuclear safety related systems and among agencies that develop software for other applications demanding very high reliability.The findings of the . j present report are in part based on the survey and in part on review of software literature . j relating to nuclear and other critical installations, as well as on the authors' experience in these areas.- -i Publication Date: october 1993-Prepared by: Hecht, H.1 Tai, A.T.: Tso, K.S. [SoHaR,Inc., Beverly Hills, CA (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Re. - search; SoHaR Inc.,BeverlyHills,CA(United States); AirForceRome Labs.,Griffiss AFB, l NY (United States) Keywords: array processors, comparative evaluations, computer-aided design, computer ' architecture,- - 2 computer codes, digital systems, fault tolerant computers, general, neural networks, parallel processing,programminglanguages l .i i i i 1 1 ) i ) i 1 i NUREG/BR--0083 Vol.9 74 - l l C -.. m.

t NUREG/GR--0006 DEPOSITION ' 1

Title:

DEPOSITION: Software To Calculate Panicle Penetration Through Aerosol Transpon Systems. Final Repon

== Description:== User-friendly software (DEPOSITION 24) has been developed which permits characteriza-tion of aerosol panicle losses in transpon systems. The sub-models which comprise the DEPOSITION code are presented and the limitations of these sub-models are noted. These sub-models have all been previously published in the pect-reviewed literature. 'Ihe software can be used to determine the penetration of aerosol through existing transpan systems; it will provide the optimal tube diameter for a transport system operated at a given flow rate and at a given panicle size; it will provide a value for the maximum penetration for a transpon system that would connect two points in three-dimensional space; and it will provide tables of data and create output files for parametric studies on the effects of varying particle size.. flow rate, and tube diameter. Use of this software for specific examples is given in an Appendix. Reference to this software is included in NRC Regulatory Guide 8.25 (1992) where it is considered to be an acceptable method for calculating the penetration of panicles through sampling systems. Publication Date: April 1993 Prepared by: Anand, NL; McFarland, A.R.; Wong, F.S.; Kocmoud. CJ. [ Texas A and M Univ., College Station, TX (United States), Dept. of Mechanical Engineering] ] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Div. of Regulatory Applications; Texas A and M Univ., College Station, TX (United S tates), Dept. of Mechani-1 cal Engineering Keywords: aerosols, air quality, DEPOSITION, ducts, flow rate, particle size, probes, progress report, radiation monitoring, sampling, stacks 1 I 75 NUREG/BR--0083, Vol.9 1 'l

NUREG/GR--0009 HEATING-6, SCDAP

Title:

Stepwise Integral Scaling Method and Its Application to Severe Accident Phenomena )

== Description:== Severe accidents in light water reactors are characterized by an occurrence of multiphase flow with complicated phase changes, chemical reaction and various bifurcation phenomena. Because of the inherent difficulties associated with full-scale testing, scaled down and simulation experiments are essential parts of the severe accident analyses. However, one of the most significant shoncomings in the area is the lack of a well-established and reliable scaling method and scaling criteria. In view of this, the stepwise integral scaling method is. developed for severe accident analyses. This new scaling method is quite different from the - conventional approach. However,its focus on dominant transport mechanisms and use of the 3 integral response of the system make.this method relatively simple to apply to very complicated multiphase flow problems. In order to demonstrate its applicability and .] usefulness, three case studies have been made. The phenomena considered are (1) corium dispersion in DCH,(2) corium spreading in BWR MARK-Icontainment,and(3)in-core boil-off and heating process.The results of these studies clearly indicate the effectiveness of their stepwise integral scaling method. Such a simple and systematic scaling method has not been previously available to seve:e accident analyses. Publication Date: October 1993 Prepared by: Ishii, M.: Zhang, G. [Purdue Univ., West Lafayette, IN (United States), School of Nuclear Engineering]; No, H.C. [ Korea Advanced Inst. of Science and Technology, Seoul (Korea, Republic of)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States) Div. of Systems Re-search; Purdue Univ., Lafayette, IN (United States), School of Nuclear Engineering j l Keywords: corium, energy transfer, HEATING-6, integral equations, loss of coolant, multiphase flow, l reactor accidents, scaling laws, SCDAP, water cooled reactors l. NUREG/BR--0083, Vol.9 76

i 4 l NUREG/GR--0010 EDDYANN, LOTUS 1-2-3

Title:

Hybrid Digital Signal Processing and Neural Networks for Automated Diagnostics Using NDE Methods

== Description:== The primary purpose of the current research was to develop an integrated approach by combining information compression methods and artificial neural networks for the monitor-ing of plant components using nondestructive examination (NDE) data. Specifically, data from eddy cunent inspection of heat exchanger tubing were utilized to evaluate this technology. The focus of the research was to develop and test various data compression l methods (for eddy current data) and the performance of different neural network paradigms for defect classification and defect parameter estimation. Feedforward, fully connected neural networks, which use the back-propagation algorithm for network training, were implemented for defect classification and defect parameter estimation using a modular network architecture. A large eddy current tube inspection database was acquired from the Metals and Ceramics Division of Oak Ridge National Laboratory. These data were used to - study the performance of anificial neural networks for defect type classification and for estimating defect parameters. A PC-based data preprocessing and display program was also 1 developed as pan of an expert system for data management and decision making.The results of the analysis showed that for effective (low-error) defect classification and estimation of parameters,it is necessary to identify proper feature vectors using different data representa-tion methods. The integration of data compression and artificial neural networks for information processing was established as an effective technique for automation of diagnos-tics using nondestructive examination methods. Publication Date: November 1993 I Prepared by: Upadhyaya, B.R.; Yan, W. [ Tennessee Univ., Knoxville, TN (United States), Dept. of Nuclear Engineering) Prepared for: Nuclear Regulatory Commiss,on, Washington, DC (United States), Div. of Engineering; Tennessee Univ., Knoxville, TN (United States), Dept. of Nuclear Engineering Keywords: automation, computer program documentation, data processing, decision making, defects, digital systems, eddy current testing, EDDYANN, expert systems, heat exchangers, LOTUS 1-2-3, neural networks, nondest uctive testing 77 NUREG/BR--0083, Vol.9

f r l + > NUREG/IA--0085 ICSP--TR-TTRIP-R RELAP5/ MOD 2

Title:

Assessment of Full Power Turbine Trip Start-Up Test for C. Trillo 1 with RELAPS/ MOD 2. International Agreement Report

== Description:== C. Trillo I has developed a model of the plant with RELAP5/ MOD 2/36.04. This model will be validated against a selected set of start-up tests. One of the transients selected to that aim is the turbine trip, which presents very specific characteristics that make it significantly different from the same transient in other pressurized-water reactors (PWRs) of different design. The main difference is that the reactor is not tripped; a reduction in primary power is carried out instead. Pretest calculations were done of the Turbine Trip Test and compared against the actual test. Minor problems in the first model, specially in the Control and Limitatio Systems, were identified and postlest calculations had been carried out. The results show a good agreement with data for all the compared variables. 1 Publication Date: July 1993 Prepared by: Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M. [Consejo de Seguridad Nuclear, Madrid (Spain)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear - Regulatory Research; Consejo de Seguridad Nuclear, Madrid (Spain) l Keywords: computer calculations, computerized simulation, heat transfer, hydraulics, international agreements, reactor safety, reactor start-up, RELAP5/ MOD 2, Spain, testing, transients, _ Trillo-1 reactor, turbines,US NRC t i f I a NUREG/BR--0083, Vol.9 78

NUREG/IA--0090 RELAPS/ MOD 2

Title:

Assessment of RELAPS/ MOD 2 Using the Test Data of REWET-II Reflooding Experiment SGI/R

== Description:== An analysis of a reflooding experiment with RELAP5/ MOD 2 cycle 36.04 is presented.The experiment had been carried out in the REWET-II facility simulating the reactor core with a bundle of 19 electrically heated rods. On the basis of the results of two calculations, recommendations for the core nodalization are presented, and a modification to the code is proposed. Publication Date: May 1993 Prepared by: Haemaelaeinen A. [Valtion Teknillinen Tutkimuskeskus, Helsinki (Finland). Nuclear Engineering Lab.) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Valtion Teknillinen Tutkimuskeskus, Helsinki (Finland), Nuclear Engineering Lab. Keywords: database management, experimental data, Finland, international agreements, Loviisa-1 reactor, Loviisa-2 reactor, reactor cores, recommendations, RELAP5/ MOD 2, rewetting, simulation, US NRC J 79 NUREG/BR--0083, Vol.9

NUREG/IA--0091 ECN--89 91 i RELAP5/ MOD 2

Title:

Assessment of RELAP5/ MOD 2 Against a Natural Circulation Experimentin Nuclear Power Plant Borssele. International Agreement Report

== Description:== As part of the ICAP (International Code Assessment and Applications Program) agreement i between ECN (Netherlands Energy Research Foundation) and US Nuclear Regulatory i Commission,ECN has performeda namberofassessment calculations for the thermohydraulic j system analysis code RELAP5/ MOD 2/36.05. This document describes the assessment of this computer program versus a natural circulation experiment as conducted at the Borssele 1 Nuclear Power Plant. The results of this comparison show that the code RELAP5/ MOD 2 predicts well the natural circulation behaviour of Nuclear Power Plant Borssele. 1 Publication Date: July 1993 l Prepared by: Winters, L. [ Netherlands Energy Research Foundation (ECN), Petten (Netherlands)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research: Netherlands Energy Research Foundation (ECN), Petten (Nether-lands) i Keywords: after-heat removal, Borssele reactor, computer calculations, computer program documenta-tion, computerized simulation, heat transfer, hydraulics, Netherlands, reactor cooling sys-tems, reactor safety, RELAP5/ MOD 2, scram, testing, US NRC j i -i J NUREG/BR--0083, Vol.9 80 1 i

NUREG/IA--0092 l RELAP5/ MOD 2

Title:

Assessment of RELAPS/ MOD 2 Computer Code Against the Net Load Trip Test Data from Yong-Owang, Unit 2

== Description:== The results of the RELAPS/ MOD 2 computer code simulation for the 100% Net Load Trip Test in Yong-Gwang Unit 2 are analyzed here and compared with the plant operation data. The control systems for the control rod, feedwater, steam generator level, steam dump, j pressurizer level, and pressure are modeled to function automatically until the power level i decreases below 30% nuclear power. A sensitivity study on control rod wonh w1Ls carried out and it was found that variable rod worth should be used to achieve good prediction ofneutron i power. The results obtained from RELAPS/ MOD 2 simulation agree well with the plant I operating data and it can be concluded that this code has the capability of analyzing the-transient of this type in a best estimate means. Publication Date: June 1993 Prepared by: Ame, N.; Cho, S, [ Korea Electdc Power Co., Taejeon (KR)]; Ize, S.H. [ Korea Inst of Nuclear Safety,Taejeon(KR)] Prepared for: Nuclear Regulatory Commission Washington, DC (United States), Office of Nuclear ~ -{ Regulatory Research; Korea Electric Power Co.,Taejon (Korea, Republic of); KoreaInst. of Nuclear Safety, Taejon (Korea, Republic of) Keywords: computerized simulation, power losses, PWR type reactors, reactor control systems, reactor cooling systems, reactor operation, reactor safety RELAPS/ MOD 2, Republic of Korea, sensitivity analysis, transients 1 i l 81 NUREG/BR--0083, Vol.9 - ) I 1

I l .I NUREG/IA--0094 STUDSVIK/NS--90/93 1 RELAP5/ MOD 2, RELAP5/ MOD 3 i

Title:

AssessmentofRELAPS/ MOD 3 AgainstTwenty FivePost-DryoutExperimentsPerformed l at the RoyalInstitute of Technology

== Description:== Assessment of RELAP5/ MOD 2 has been made against various experimental data, among other data from twenty-five post-dryout experiments conducted at the Royal Institute of Technology (RIT)in Stockholm. As the MOD 3 version of RELAPS has now been released, incorporating a different method of calculadng critical heat flux compared to RELAP5/ MOD 2,it seemedjustified to make another assessment against the same RIT data.The results show that the axial dryout position is generally better predicted by the MOD 3 than by the MOD 2 version. The prediction is, however, still nonconservative (i.e., the calculated dryout position falls in most cases downstream of the actual measured point).While the pre-dryout heat transfer seems to be equal for MOD 2 and MOD 3, both versions giving slightly higher l wall temperatures than the experiments, there is a considerable difference in the post-dryout heat transfer. The results of the RIT data comparison indicate that MOD 3 underpredicts the ] post-dryout wall temperatures remarkably while MOD 2 gave reasonable agreement. In this - respect RELAP5/ MOD 3 shows no improvement over RELAP5/ MOD 2. l Publication Date: May 1993 Prepared by: Nilsson, L. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear l Regulatory Research; Swedish Nuclear Power Inspectorate, S tockholm (S weden) Keywords: algorithms, computerized simulation, critical heat flux, data compilation, dryout, heat transfer, hydraulics, reactor safety, RELAP5/ MOD 2, RELAP5/ MOD 3, validation, verifica-tion i s .l j 'NUREG/BR--0083, Vol.9 82 l 4 t e =' ~ - >- ~-

NUREG/IA--0095 RELAP5810D2, RELAPSSIOD3 i

Title:

RELAPS Assessment Using LSTF Test Data SB-CL-18

== Description:== A 5% cold leg break test, run SB-CL-18, conducted at the Large Scale Test Facility (LSTF) was analyzed using the RELAPS/ MOD 2 Cycle 36.M and the RELAP5/ MOD 3 Version 5m5 j codes. The test SB-CL-18 was conducted with the main objective being the investigation of the thermal-hydraulic mechanisms responsible for the early core uncovery, including the manometric effect due to an asymmetdc coolant holdup in the steam generator upflow and downflow side.The present analysis, carried out with the RELAP5/ MOD 2 and MOD 3 codes, i demonstrates the code's capability to predict, with sufficient accuracy, the main phenomena occurring in the depressudzation transient, both from a qualitative and quantitative point of view. Nevertheless, several differences regarding the evolution of phenomena and affecting the timing order have been pointed out in the base calculations. The sensitivity study on the break flow and the nodalization study in the components of the steam generator U-tubes and the cross-over legs were also carried out. The RELAP5/ MOD 3 calculation with the nodalization change resulted in good predictions of the major thermal-hydraulic phenomena ] and their timing order. 1 Publication Date: May 1993 Prepared by: Lee, S.: Chung, B.D.: Kim, HJ. [ Korea Inst. of Nuclear Safety, Taejon (Korea. Republic of)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research: Korea Inst. of Nuclear Safety Taejon (Korea, Republic of) Keywords: computer calculations, depressr.rization, evaluation, heat transfer, hydraulics, loss of coolant, PWR type reactors, reactor cooling systems, reactor safety, RELAPS/ MOD 2, _l RELAP5/ MOD 3, steam generators, test facilities, transients, tubes l 1 83 NUREO/BR--0083, Vol.9

NUREG/IA--0096 AEA-TRS--1050:AEEW.R--2501 RELAP5/ MOD 3

Title:

Numerics and Implementation of the UK Horizontal Stratification Entrainment Off-Take Model into RELAP5/ MOD 3. International Agreement Repon

== Description:== His report presents the numerics and implementation details to add the same improved discharge quality correlations into RELAP5/ MOD 3. In the light of experience with the modified RELAP5/ MOD 2 code, some of the numerics have been slightly changed for RELAP5/ MOD 3.The description is quite detailed in order to facilitate change by some future code developer. A simple test calculation was performed to confirm the coding of the correlations implemented in RELAP5/ MOD 3. Publication Date: June 1993 Prepared by: Bryce, W.M. [AEA Thermal Reactor Services, Winfrith (United Kingdom), Physics and ThermalHydraulics] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; AEAThermal Reactor Services,Winfrith (United Kingdom), Physics and Thermal Hydraulics Keywords: computer calculations, entrainment, flow models, fluid flow, heat transfer, hydraulics, international agreements, loss of coolant, PWR type reactors, reactor safety, RELAP5/ i MOD 3, stability, stratification, United Kingdom NUREG/BR--0083, Vol.9 84

.,a A u .~.. + u -, .a j NUREG/IA.-0099 RELAP5/ MOD 2, RELAP5/ MOD 3 l

Title:

RELAPS Assessment Using Semiscale SBLOCA Test S-NH-1. Intemational Agreement Report i

Description:

A 2-inch cold leg break test S.NH 1, conducted at the 1/1705 volume scaled facility l Semiscale, was analyzed using RELAP5/ MOD 2 Cycle 36.04 and MOD 3 Version 5m5. Loss of high-pressure injection system (HPIS) was assumed, reactor trip occurred on a low. i pressurizer (PZR) pressure signal (13.1 MPa), and pumps began an unpowered coastdown on SI signal (12.5 MPa). The system was recovered by opening atmospheric dump valves (ADVs) when the peakcladding temperature (PCT) became higherthan 811 K. Accumulator - 'I was finally injected into the bystem when the primary system pressure was less than 4.0 MPa. 'Ihe experiment was terminated when the pressure reached thelow-pressure injection system 1 (LPIS) actuation set point. RELAP5/ MOD 2 analysis demonstrated its capability to predict,'. with a sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and a quantita:ive pint of view. Nevertheless, several differences were noted regarding the break ficw rate and irgentory distribution due to deficiencies in two-phase choked flow model, horizontal stratificition interfacial drag, and a countescurrent .l flow limitation (CCFL) model. 'Ihe main reason for the core to remain nearly fully covered j with the liquid was the underprediction of the break flow by the code. Several sensitivity calculations were tried using the MOD 2 to improve the results by using the different options of break flow modeling (downward, homogeneous, and area increase). The break area I compensating concept based on "theintegrated break flow matching"gave better results than downward junction and homogeneous options, and tuc MOD 3 showed improvement in predicting a CCFL in the steam generator and a heatup in the care. Publication Date: June 1993 l Peepared by: Lee, EJ.; Chung, B.D.: Kim, HJ. [ Korea Inst. of Nuclear Safety, Taejon (Korea, Republic I of)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research: Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of) Keywords: computer calculations, heat transfer, hydraulics,intemational agreements, loss of coolant, PWR type reactors, reactor cooling systems, reactor safety, RELAP5/ MOD 2, RELAPS/ MOD 3, scale models, test facilities F l ? 85 NUREG/BR-0083, Vol.9 -

.. p + NUREG/IA--0100 ~ RELAP5/ MOD 3

Title:

Assessment of CCFL Model of RELAP5410D3 Against Simple Vertical Tubes and Rod t Bundle Tests. Intemational Agreement Report

== Description:== The countercurrent flow limitation (CCFL) modelused in RELAP5/ MOD 3 version Sm5 has been assessed against simple vertical tubes and bundle tests performed at a facility of Korea 1 Atomic Energy Research Institute. The effect of changes in tub diameter and nodalization of tube section were investigated.*Ihe roles ofinterfacialdrags cn the flooding characteristics are discussed. Differences between the calculation and the experiment are also discussed. A f comparison between model assessment results and the test data showed that the calculated. 'i l-value lay weJ1 on tb: experimental flooding curve specified by user, but the pressure jump - l 1 before onset of Ilooding was not calculated. Publication Date: June 1993 f Prepared by: Cho,S.; Arne,N.[KoreaElectricPowerCorp..Taejon(KR),ResearchCenter]:Chung,B.D.; l Kim, HJ. [ Korea Inst. of Nuclear Safety, Taejon (KR)]. .t Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear l Regulatory Research; Korea E! ctric Power Corp.,Taejon (Korea, Republic of), Research - Center; Korea Inst. of Nuclear Safety,Taejon (Korea, Republic of) Keywords: computerized simulation, fluid flow, heat transfer, hydraulic transport, hydraulics, perfor. mance, PWR type reactors, reactor safety. RELAP5/ MOD 3, md bundles, thermal analysis. - + transients t i f I -{ a i i I i 3 i ?NUREG/BR--0083 Vol.9 86 l 1 i

NUREG/IA--0103 RELAP5/ MOD 3

Title:

Assessment of BETHSY Test 9.1.b Using RELAPS/ MOD 3. International Agreement Report

== Description:== The 2-inch cold leg break test 9.1.b, conducted at the BETHSY facility, was analyzed using l the RELAP5/ MOD 3 Version Sm5 code.'Ihe test 9.1.b was conducted with the main objective being the investigation of the thermal-hydraulic mechanisms responsible for the large core uncovery and fuel heat-up, requiring the implementation of an ultimate procedure. *Ibe present analysis demonstrates the code's capability to predict, with sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and quantitative point of view. Nevertheless, several differences regarding the evolution of phenomena and affecting the timing order have to be pointed out in the base calculation. Three calculations were carried out to study the sensitivity to change of the nodalization in l the components of the loop seal cross-overlegs,of the auxiliary feedwater controllogics,and of the break discharge coefficient. Publication Date: June 1993 Prepared by: Lee, S.; Chung, B.D.; Kim, HJ. [ Korea Inst. of Nuclear S afety, Taejon (Korea, Republic of)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office'of Nuclear-Regulatory Research; Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of) Keywords: computer calculations, depressurization, evaluation, heat transfer, hydraulics, international agreements, loss of coolant, PWR type reactors, reactor cooling systems, reactor safety, RELAPS/ MOD 3, Rept.blic of Korea, testing 87 NUREG/BR--0083, Vol.9 f

NUREG/IA--0104 RELAP5/ MOD 3

Title:

RELAPS/ MOD 3 Assessment Using the Semiscale 50% Feed Line Break Test S-FS-11

== Description:== The RELAP5/ MOD 35m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase,(b) stabilization phase,and (c) refill phase.The first objective was to assess the code applicability to 50% FLB situation; the second was to evaluate the fm' al safety analysis report (FSAR) conservatisms regarding steam generator (SG) heat transfer degra-dation, steam line check valve failure, break flow state, and peak primary system pmssure; and the third was to validate the emergency operating procedure (EOP) effectiveness. 'Ihe code was able to simulate the major thermal / hydraulic parameters except for the two-phase break flow and the secondary convective heat transfer rate.The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not conservative, the measured break flow experienced all types of flow conditions, and the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operations performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator. Publication Date: June 1993 Prepared by: Lee, EJ.; Chung, B.D.; Kim, HJ. [ Korea Inst, of Nuclear Safety, Taejon (Korea, Republic of)] Prepared !vr: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research: Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of) Keywords: blowdown, computer calculations, evaluation, heat transfer, hydraulics, loss of coolant, PWR type reactors, reactor safety, RELAP5/ MOD 3, stability, steam generators, test facili-ties, valves NUREG/BR -0083, Vol.9 88

a NUREG/IA--0105 RELAP5/ MOD 3

Title:

Assessment of RELAPS/ MOD 3 Version Sm5 Using Inadvertent Safety Injection Incident Data of Kori Unit 3 Plant

== Description:== This report discusses an inadvertent safety injection incident which occurred at Kori Unit 3 in September 6,1990, and was analyzed csing the RELAP5/ MOD 3 code. The event was initiated by a closure of main feedwater control valve of one of three steam generators. High pressure safety injection system was actuated by the low pressure signal of main steam line. The actual sequence of plant transient with the proper estimations of operator actions was investigated in the present calculation. 'Ihe asymmetric loop behaviors of the plant were also considered by nodalizing the loops of the plant into three. The calculational results are j compared with the plant transient data. It is shown that the overall plant transient depends strongly on the auxiliary feedwater flowrate controlled by the operator and that the code gives an acceptable prediction of the plant behavior with the proper ascumptions of the operator actions. The results also show that the solidification of pressurizer does not occur and the i liquid-vapor mixture does not flow out through pressurizer power-operated relief valve (PORV). The behavior of primary pressure during pressurizer PORV actuation is poorly predicted because the actual behavior of pressurizer PORV could not be modeled in the present simulation. Publication Date: May 1993 Prepared by: Kim, K.T.; Chung, B.D.; Kim,I.G.; Kim, HJ. [ Korea Inst. of Nuclear Safety,Taejon (Korea, Republic of)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Korea Inst, of Nuclear Safety, Taejon (Korea, Republic of) Keywords: computerized simulation, flow rate, fluid flow, heat transfer, high pressure coolantinjection, { hydraulics, international agreements, pressurizers, PWR type reactors, reactor cooling systems, reactor safety, RELAP5/ MOD 3, testing, transients, valves 89 NUREG/BR-4)083, Vol.9

k NUREG/IA--0106 TEC/L--0471/R91 l RELAP5/ MOD 2, RELAP5/ MOD 3 (

Title:

Assessment of PWR Steam Generator Modelling in RELAP5/ MOD 2. International Agree-ment Report

== Description:== An assessment of Steam Generator (SG) modelling in the pressurized-water reactor (PWR) thermal-hydraulic code RELAP5/ MOD 2 is presented.The assessment is based on a review of code assessment calculations performed in the United Kingdom and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR,and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed, including methods for compensating for the deficiencies through (. changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/ MOD 3. Publication Date: June 1993 Prepared by: Putney, J.M.: Preece, RJ. [ National Power, Leatherhead (GB), Technology and Environ-ment Centre] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; National Power, Leatherhead (United Kingdom), Technology and Environment Centre Keywords: computerized simulation, heat transfer, PWR type reactors, reactor safety,RELAP5/ MOD 2, RELAP5/ MOD 3, safety analysis, steam generators, Wolf Creek-1 reactor l l l NUREG/BR -0083, Vol.9 90

1 F ICSP--V2 R50-R NUREG/IA--0107 I RELAP5/ MOD 2 ?

Title:

Assessment of RELAPf/ MOD 2 Against a Load Rejection from 100% to 50% Power in the Vandellos 11 Nuclear Power Plant. International Agreement Repo;t

== Description:== An assessment of RELAPS/ MOD 2 cycle 36.04 against a load rejection from 100% to 50% power in Vandellos 11 Nuclear Power Plant (Spain)is presented. 'Ihe work is inscribed in the l framework of the Spanish contribution to the International Code Assessment and Applica-tions Program (ICAP) Project. The model used in the simulation consists of a single loop, a steam generator, and a steam line up to the steam header, all of them enlarged on a scale of l 3:1, and full-scaled reactor vessel and pressurizer.The results of the calculations have been in reasonable agreement with plant measurements. Publication Date: June 1993 Prepared by: Llopis, C. [Asociacion Nuclear de Vandellos, Madrid (Spain)]; Mendizabal, R.: Perez, J. (Consejo de Seguridad Nuclear, Madrid (Spain)] + Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Asociacion Nuclear de Vandellos, Madrid (Spain); Consejo de Seguridad Nuclear, Madrid (Spain) l Keywords: coordinated research programs, excursions, heat transfer, hydraulics, intemational agree-ments, reactor safety, RELAP5/ MOD 2, stability, transients, Vandellos II reactor l i a 1 i l 1 I i 91 NUREG/BR--0083, Vol.9 1 1 1

I l 1 NUREG/IA -0108 ICSP--V2-R100-R RELAP5/ MOD 2 ) Assessment of RELAP5/ MOD 2 Against ' Turbine Trip from 100% Power in the Vandellos

Title:

a IINuclear Poww Plant

== Description:== An assessment ofRELAP5/ MOD 2 cycle 36.04 against a turbine trip from 100% power in the Vandellos II Nuclear Power Plant (Spain) is presented. The work is inscribed in the framework of the Spanish contribution to the International Code Assessment and Applica-tions Program (ICAP) Project. The model used in the simulation consists of a single loop, a steam generator, and a steam line up to the steam header, all of them enlarged on a scale of 3:1, and full-scaled reactor vessel and pressurizer. The results of calculations have been in reasonable agreement with plant measurements. An additional study has been performed to .i check the ability of a modelin which all the plant components are full-scaled to reproduce l the transient. A second study has been performed using the Homogeneous Equilibrium i Model in the pressurizer, trying to elucidate the influence of the velocity slip in the primary depressurization rate. Publication Date: June 1993 ~ Prepared by: Llopis, C. [Asociacion Nuclear Vandellos, Barcelona (ES)]; Perez, J.: Mendizabal, R. [Consejo de Seguridad Nuclear, Madrid (ES)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; AsociacionNuclearVandellos, Barcelona (Spain);Consejo de Seguridad Nuclear, Madrid (Spain) Keywords: CDC computers, computerized simulation, evaluation, experimental data, reactor vessels, RELAP5/ MOD 2, Spain, steam turbines, transients, US NRC, Vandellos !! reactor l I l 1 i 4 - NUREG/BR--0083, Vol.9 92 i ) i I 1

e UNID--91-08 NUREG/IA--0109 RELAP5/ MOD 2

Title:

Assessment of RELAP5/ MOD 2 Against a 10% Load Rejection Transient from 75% Steady State in the Vandellos 11 Nuclear Power Plant

== Description:== The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained modelhasbeenassessedagainstthefollowingtransientsthatoccurredintheplant atripfrom the 100% powerlevel(CSN); aload rejection from 100% to 50% (CSN); aload rejection from 75 % to 65 % (ANV); and a feedwater turbopump trip (ANV). ~Ihis copy is a report of the load rejection from 75% to 65% transient simulation.This transient was one of the tests carried out in Vandellos II NPP during the startup tests. Publication Date: May 1993 Prepared by: Llopis, C.; Casals, A. [Asociacion Nuclear Vandellos, Barcelona (ES)]; Perez,J.; Mendizabal, R. [Consejo de Seguridad Nuclear, Madrid (ES)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear RegulatoryResearch; Asociacion Nuclear Vandellos, Barcelona (Spain);Consejo de Seguridad Nuclear, Madrid (Spain) i Keywords: computer program documentation, evaluation, international agreements, reactor control systems, reactor cooling systems, reactor start-up, RELAP5/ MOD 2, simulation, Spain, transients, US NRC, Vandellos II reactor l 93 NUREG/BR-4)083, Vol.9

NUREG/IA--0110 ICSP. V2 TURFW R RELAP5/ MOD 2 I

Title:

AssessmentofRELAP5/ MOD 2 AgainstaMainFeedwaterTurbopumpTripTransientin the - VandellosIINuclear Power Plant

== Description:== The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data and in the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients that occurred in the plant: A trip from the 100% power level (CSN): a load rejection from 100% to 50% (CSN); a load rejection from 75% to 65% (ANV); and a feedwater turbopump trip (ANV). This copy is a report of the feedwater turbopump trip transient simulation.~Ihis transient actually occurred in the plant on June 19,1989. Publication Date: December 1993 Prepared by: Llopis,C.;Casals, A. [Asociacion Nuclear Vandellos, Madrid (Spain)];Perez,J.; Mendizabal, - R. [Consejo de Seguridad Nuclear, Madrid (Spain)] l Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear I Regulatory Research: Asociacion Nuclear de Vandellos Madrid (Spain); Consejo.de Seguridad Nuclear, Madrid (Spain) Keywords: bilateral agreements, computerized simulation, RELAP5/ MOD 2, steam generators, tran-sients, Vandellos 11 reactor I L NUREG/BR--0083, Vol.9 94

ECN C--92-008 NUREG/IA--0112 -RELAP5/ MOD 2 4

Title:

Assessment of RELAP5/ MOD 2 Against ECN Reflood Experiments. International Agree-ment Report ..l

== Description:== As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN (Netherlands Energy Research Foundation) and the US Nuclear Regulatory Commission, ECN has performed a number of assessment calculations with the computer program RELAP5.This report describes the results as obtained by ECN from the assessment of the thermohydraulic computer program RELAP5/ MOD 2/CY 36.05 versus a series of reflood experiments in a bundle geomeny. A totai number of seven selected experiments ! have been analyzed from the reflood experimental program as previously conducted by ECN j under contract of the Commission of the European Communities (CEC). In this document, the results of the analyses are presented and a comparison with the experimental data is provided. Publication Date: July 1993 Prepared by: Woudstra, A.: Van De Bogaard, J.P.A.: Stoop, P.M. [ Netherlands Energy Research Founda-tion (ECN),Petten (Netherlands)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research: Netherlands Energy Research Foundation (ECN), Petten (Nether-lands) Keywords: computer calculations, coordinated research programs, experimental data, heat transfer, hydraulics, loss of coolant, Netherlands, PWR type reactors, reactor cooling systems, reactor safety, RELAP5/ MOD 2,US hTC t-95 NUREG/BR--0083, Vol.9 '

NUREG/IA--0113 ' RELPIN, RELAP5/ MOD 3

Title:

Preliminary Assessment of PWR Steam Generator Modelling in RELAPS/ MOD 3. Interna- - I tional AgreementReport )

Description:

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAPS/ MOD 3 is presented.The study is based on calculations against a ' series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests'are used to assess the modelling of primary to j secondary side heat transfer and, in particular, to examine the effect of reverting to the i standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/ MOD 2. Comparisons between the two versions of the code are also used to show how the new interphase drag modelin RELAP5/ MOD 3 affects the calculation of SG liquid i inventory and the void fmetion profile in the riser. j Publication Date: July 1993 Prepared by: Preece, RJ.: Putney, J.M. [ National Power, Leatherhead (United Kingdom), Technology _ and Environment Centre] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research: National Power, Leatherhead (United Kingdom), Technology and - Environment Centre Keywords: computer calculations, computerized simulation, heat transfer, hydraulics, PWR type reac-tors, RELAP5/ MOD 3, RELPIN, steam generators, void fraction, Wolf Creek-1 reactor i i l . NUREG/BR-0083, Vol.9 96 y r n-- m - i- " -

KWU--E412/91/E1002 NUREG/IA--0116 RELAP5/ MOD 2, RELAP5/ MOD 3

Title:

Assessment ofRELAP5/ MOD 3/V5M5 Against theUPTFTestNo.11 (Countercurrent Flow in PWR Hot Leg)

== Description:== Analysis of the UPTFTestNo.11 using the"best-estimate" computer code RELAP5/ MOD 3/ Version SMS is presented. Test No.11 was a quasi-steady state, separate effect test designed to investigate the conditions for countercurrent flow of steam and saturated water in the hot leg of a PWR. Without using the code's new countercurrent flow limitation (CCFL) model, RELAP5/ MOD 3/V5M5 overestimated the mass flow rate of back down flowing water up to 35 % (1.5 MPa runs) and 43 % (03 MPa runs).Thisis the most obvious difference to RELAP5/ MOD 2, which did not allow enough countercurrent flow. From the point of view of performing plant calculations, this is certamly an improvement, because the new junction-based CCFL option could be used to restrict the flows to a flooding curve defined by a user-supplied correlation. Very good agreement with the experimental data for 1.5 MPa--which are relevant for SBLOCA reflux condensation conditions-could be obtr.ined using the code's new CCFL option in the middle of the inclined part (riser) of the hot leg. Using the same CCFL correlation for the simulation of 0.3 MPa test series-typical for reflood conditions-the code underestimated by 44% the steam mass flow rate at which complete liquid carryover occurs. An unphysical result was received using a CCFL correlation of the Wallis type with the intercept C=0.644 and the slope m=0.8. The unphysical prediction is an indication ofpossible programming errorsin the CCFL modelof the RELAP5/ MOD 3/V5M5 computer code. Publication Date: May 1993 Prepared by: Cuma-Tivig, F. [Siemens AG-KWU Group, Erlangen (Germany)] j Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Siemens AG Untemehmensbereich KWU, Erlargen (Germany) Keywords: computerized simulation, experimental data, flow rate, loss orcoolant. prediction equations, prirnary coolant circuits, programming, PWR type reactors, reactor accidents, reactor vessels, RELAPS/ MOD 2, RELAP5/ MOD 3, specifications, steam, water l 97 NUREG/BR--0083, Vol.9

' NUREG/IA--0118 TD/SPB/ REP--0130 RELAP5/ MOD 2

Title:

Analysis of LOFT Test L5-1 Using RELAP5/ MOD 2

== Description:== REL APS/ MOD 2 is being used by Technology Division for the calculation of certain small break loss-of coolant accidents (SBLOCA) and pressurized transients in the Sizewell"B" presstdzed-water reactor (PWR). To assist in validating RELAPS/ MOD 2 for the above 1 application, the code is being used to model a number of smallLOCA and pressurized fault simulation experiments carried out in integral test fccilities. ne present report describes a RELAP5/ MOD 2 analysis of an intermediate break LOCA test in the loss of fluid test (LOFT) facility.ne test discussed in this report was designed to simulate the rupture of a single 14-inch diameter accumulator injection line in a commercial PWR with a 25% break in the j broken loop cold leg. Early in the transient the pumps were tripped and the high-pressure injection system (HPIS) injection initiated; towards the end of the transient, accumulator and low-pressure injection system (LPIS) injection began. RELAP5/ MOD 2 gave reasonably accurate predictions of the system thermal hydraulic behaviour but failed to accurately calculate the core dryout which occurred because of boil-off prior to accumulator injection. The error is the result of the failure to calculate the correct core void distribution during this period of the trans ent. A separate calculation using the RELPIN code using hydraulic data from the RELAP5 analysis gave significantly improved predictions of the core dryout. However, the peak clad temperature was underpredicted. It is believed that the error is due to the fact that the core liquid inventory in this boildown was overpredicted in the RELAP5/ MOD 2 calculation. 1 i Publication Date: May 1993 Prepared by: Cooper, S. [ Nuclear Electric plc, Barnwood (United Kingdom)] _j Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Nuclear Electric plc, Barnwood (United Kingdom) Keywords: computer calculations, computer program documentation, heat transfer, hydraulics, interna-tional agreements, LOFT reactor, loss of coolant, PWR type reactors, reactor safety, RELAP5/ MOD 2, testing, transients NUREG/BR--0083, Vol.9 98

ICSP--AS-BOUT-R NUREG/IA--0119 RELAP5/ MOD 2 i L

Title:

Assessment and Application of Blackout Transients at Asco Nuclear Power Plant with ~ RELAPS/ MOD 2. International Agreement Report

== Description:== The Asociacion Nuclear Asco has prer ired a model of the Asco Nuclear Power Plant using RELAP5/ MOD 2. This model, which includes thermal-hydraulics, kinetics, and protection and controls.has been qualified in previous calculations ofseveral actualplant transients.The l first part of the transie. ' presented in this reportis an actualblackout and one of the transients of the qualification process. The results are in agreement with plant data.The second part of the transient is a hypothetical case. It consists in restarting a primary pump and assuming a new blackout. The phenomenology prediction of this second part has been useful from the operation and safety point of view. Publication Date: June 1993 Prepared by: Reventos, F.: Baptista, J.S.: Navas, A.P.: Moreno, P. [Asociacion Nuclear Asco, Barcelona (Spain)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear-Regulatory Research; Asociacion Nuclear Asco, Barcelona (Spain) i Keywords: Asco-1 reactor, Asco-2 reactor, blackouts, computer calculations, coordinated research programs, heat transfer, hydraulics, international agreements, reactor safety, RELAP5/ I MOD 2, transients l I l ) l 'l ] .i 1 i l I' 99 NUREG/BR--0083, Vol.9 j i

1 1 NUREG/IA--0120 EST-SIAN--22 TRAC BF1 1

Title:

Assessment of the Turbine Trip Transient in Cofrentes NPP with TRAC-BF1

== Description:== This report presents the results of the assessment of TRAC-BF1 (G 1-J1) code with the model of C N. Cofrentes for simulation of the transient originated by the manual trip of the main turbire. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 MW, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola. S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbinedven feedwater (FW) pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator. Publication Date: June 1993 Prepared by: Castrillo, F. [Hidroelectrica Espanola, Madrid (Spain)); Gomez, A.; Gallego, I. (Union Iberoamericana de Tecnologia, Madrid (Spain)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Union Electrica, S A, Madrid (Spain) i Keywords: Cofrentes reactor, computerized simulation, reactivity coefficients, reactor safety, simula-tion, steam turbines, TRAC-BF1, transients i i 1 i i 1 NUREG/BR--0083, Vol.9 100

i: ICSP--AS-SPR-R NUREG/IA--0121 RELAPS/ MOD 2

Title:

Assessment of a Pressurizer Spray Valve Faulty Opening Transient at Asco Nuclear Power Plant with RELAPS/ MOD 2. Intemational Agreement Repon

== Description:== ne Asociacion Nuclear Asco has prepared a model of the Asco Nuclear Power Plant using RELAP5/ MOD 2. This model, which includes thermal-hydraulics, kinetics, and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification processis a" Pressurizer spray valve fauhy opening" presented in this repon. It consists of a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection.The results are in close agreement with plant data. Publication Date: December 1993 . Prepared by: Reventos, F.; Baptista, J.S.: Navas, A.P.; Moreno, P. [Asociacion Nuclear Asco, Barcelona (Spain)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Asociacion Nuclear Asco, Barcelona (Spain) Keywords: Asco-1 reactor, Asco-2 reactor, depressurization, evaluation, failures, heat transfer, hydrau-lies, pressurizers, primary coolant circuits, reactor safety, RELAP5/ MOD 2, transients, valves 101 NUREG/BR-@83, Vol.9

P NUREG/IA--0122 - L TRAC-BF1 1

Title:

Assessment of MSIV Full Closure for Santa Maria de Garona Nuclear Power Plant Using. TRAC-BF1 (G1J1)

== Description:== This document presents a spurious Main Steam Isolation Value (MSIV) closure analysis for - 1 Santa Mana de Garona NuclearPower Plant describing the problems found when comparing 4 calculatedandrealdata.TheplantisaGeneralElectricBoilingWaterReactor3 containment - type Mark 1. Itis operated by NUCLENOR. S.A. and was connected to the grid in 1971.The - analysis has been performed by the Applied Physics Depanment from the University of Cantabria and the Analysis and Operation Section from NUCLENOR, S.A. as a part of an. agreement for developing an engineering simulator of operational transients and accidents. 1 for Santa Maria de Garona Power Plant.The analysis was performed using the frozen version of TRAC-BFI (GlJI) code and is the second of two NUCLENOR contributions to the. International Code Assessment and Applications Program (ICAP). 'Ihe code was run in a Cyber932withoperatingsystemNOS/VE,propertyofNUCLENOR,S.A. Aprogramming effort was carried out in order to provide suitable graphics from the output file. Publication Date: June 1993 ] Prepared by: Crespo, J.L. [Cantabria Univ., Santander (Spain), Facultad de Ciencias]; Fernandez, R.A. j [Centrales Nucleares del Norte SA (NUCLENOR), Madrid (Spain)) j -Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Cantabria Univ., Santander (Spain), Facultad de Ciencias 1 Keywords: computer calculations, failures, Garona reactor, heat transfer, hydraulics, international -) agreements, reactor accidents, reactor safety, testing, TRAC-BF1, transients, valves l 'I i 1 !:i ~l NUREG/BR--0083, Vol.9. 102 )

ICSP--AL-B OUT-R - NUREG/IA--0123 RELAP5810D2, RELAP5M10D1

Title:

Application of Full Power Blackout for C. N. Almaraz with RELAP5/ MOD 2

== Description:== The analysis group of Almaraz NuclearPower Plant has developed a model of the plant with RELAP5/ MOD 2/36.04. This model is the result of the work experience on the code RELAPS/ MODI that was the standard code during the period 1984-1989. Different solutions were adopted in the network to implement the RELAP5/MODl/CY computer code. The first complete calculation ofan accident (normal power blackout) was undertaken. All calculations presented in this report were performed on a computer CDC Cyber 180/830. l The CPU time versus real time event was 22.6. Publication Date: June 1993 Prepared by: Lechas, A.L. [C. N. Almaraz I y 11. Madrid (ES)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Central Nuclear de Almaraz, Madrid '(Spain) Keywords: Almaraz 1 reactor, Almaraz-2 reactor.computerizedsimulation. outages reactoraccidents, reactor safety, RELAP5/ MODI, RELAP5/ MOD 2, theoretical data, transients 103 NUREG/BR--0083, Vol.9

NUREG/IA--0124 E RELAP5/ MOD 2 l

Title:

Assessment of RELAP5/ MOD 2 Against a Pressurizer Spray Valve Inadverted Fully Open-ing Transient and Recovery by Natural Circulation in Jose Cabrera Nuclear Station

== Description:== This document presents the comparison between the simulation results and the plant - measurements of a real event that took place in Jose Cabrera nuclear power plant August 30, 19&4. The event was originated by the total, continuous, and inadverted opening of the pressurizer spray valve PCV-400A. Jose Cabrera power plant is a single-loop Westinghouse pressurized-water reactor PWR belonging to Union Electra Fenosa, S.A. (Union Fenosa), a Spanish utility which participates in the International Code Assessment and Applications Program (ICAP) as a member of Unidad Electrica, S.A. (UNESA). His is the second of its two contributions to the Program: the first one was an application case and this is an assessment one. The simulation has been performed using the RELAPS/ MOD 2 cycle 36.04 code, running on a CDC Cyber 180/830 computer underNOS 2.50perating system.ne main phenomena have been calculated correctly and some conclusions have been made about the 3D characteristics of the condensation due to the spray and its simulation with a 1D tool. Publication Date: June 1993 Prepared by: Arroyo, R.; Rebollo, L. [ Union Electrica, S A, Madrid (Spain)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Union Electrica, SA, Madrid (Spain) Keywords: failures, heat transfer, hydraulics. international agreements,naturalconvection,pressurizers, reactercoolingsystems reactorsafety,RELAP5/ MOD 2, transients, valves,Zorita-1 reactor t NUREG/BR--0083, Vol.9 104

' NUREG/IA -0125 RELAPS/ MOD 2

Title:

Assessment of RELAPS/ MOD 2 Computer Code Agains't the Natural Circulation' Test Data - from Yong-Gwang Unit 2 -

== Description:== The results of the RELAP5/ MOD 2 computer code simulation for the NaturalCirculation Test in Yong-Gwang Unit 2 are analyzed here and compared with the plant operation data. "Ihe result of comparison reveals that the code calculation does present well the overall macro-scopic behaviors of thermal-hydraulic parameters in primary _and secondary systems compared with the plant operating data. The sensitivity study is performed to find out the i effect of steam dump flow rate on the primary temperatures and it is found that the primary. temperatures are very sensitive to the steam dump flow rate during the Natural Circulation. - Because of the inherent uncertainties in the plant data, the assessment work is focussed on phenon. ens whereby the comparison bete a plant data and calculated data is based more on trends than on absolute values. Publication Date: June 1993 Prepared by: Ame, N.; Cho, S. [ Korea Electric Power Corp., Taejon (Korea, Republic of), Research Center); Kim, HJ. [ Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Korea Electric Power Corp., Taejon (Korea, Republic of), Research Center; Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of) - Keywords: computer calculations. computerized simulation. flow rate, heat transfer, hydraulics, natural ' convection, PWR type reactors, reactor cooling systems, reactor safety, RELAPS/ MOD 2, sensitivity analysis, steam l 1 I l .] 4

1 1

l i E 105 NUREG/BR--0083, Vol.9 l l

NUREG/IA--0126 GRS--100;MPR--1345 TRAC

Title:

2DSD Program Work Summary Report, [ January 1988-December 1992). International Agreement Report

== Description:== The 2DS D Program was carried out by Germany, Japan, and the United S tates to investigate the thermal-hydraulics of a pressurized-water reactor (PWR) large-break loss-of-coolant accident (LOCA). A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UI'TF), and Japan constructed and operated the Cylindrical Core TestFacility (CCTF) and the Slab Core TestFacility (SCTF). The U.S. contribution consisted of provision of advanced instrumentation to each of the three test facilities and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2DSD Program in terms of the contributing efforts of the participuts. Publication Date: June 1993 Prepared by: Damerell,P.S.;Simons,J.W.[eds.][MPR Associates,Inc. Washington.DC(UnitedStates)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear Regulatory Research; Japan Atomic Energy Research Inst., Tokyo (Japan); Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) mbH, Koeln (Germany); Siemens AG, Berlin (Ger-many);Los AlamosNationalLab..NM(United States);MPR Associates,Inc., Washington, DC (United States) Keywords: coordinated research programs, Germany, heat transfer, hydraulics, Japan, loss of coolant, PWR type reactors, reactor instrumentation, reactor safety, test facilities, TRAC, USA NUREG/BR--0083, Vol.9 106

GRS-101;MPR--1346 NUREG/IA--0127 TRAC ~ l

Title:

Reactor Safety Issues Resolved by the 2DSD Program. International Agreement Report

== Description:== The 2DSD Program studied multidimensional thermal-hydraulics in a pressurized-water reactor (PWR) core and primary system during the end-of-blowdown and post-blowdown phases ofa large-breakloss-of-coolant accident (LBLOCA) and during selected small-break loss-of-coolant accident (S BLOCA) transients.The program included tests at the Cylindrical Cere Test Facility (CCTF), the Slab Core Test Facility (SCIF), and the Upper Plenum Test Facility (UPTF) and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior.while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydmulic behavior throughout the primary system in tests as well as in PWRs.This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2DSD Program. The discussion is organized in terms of the reactor safety issues investigated. Publication Date: July 1993 Prepared by: Damerell, P.S.; Simons, J.W. [eds.] [MPR Associates,Inc., Washington, DC (United States)) Prepared for: Nuclear Regulatory Commission, Washington, DC (United States); Japan Atomic Energy Research Inst., Tokyo (Japan); Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) mbH, Koeln (Gemany); Siemens AG Untemehmensbereich KWU, Erlangen (Germany); i Ims Alamos National Lab., NM (United States); MPR Associates, Inc., Washington, DC (United States) Keywords: coordinated research programs, depressurization, ECCS, Germany, heat tran sfer. hydraulics, international agreements, Japan, loss of coolant, PWR type reacton, reactor safety, steam, test facilities, testing, TRAC, two-phase flow [ 107 NUREG/BR--0083, Vol.9

5 NUREG/IA--0128 EGG. EAST--8719 RELAP5/ MOD 2, RELAP5/ MOD 3, TRAC-B

Title:

International Code Assessment and Applications Program: Summary of Code Assessment StudiesConcerningRELAP5/ MOD 2,RELAP5/ MOD 3,andTRAC-B. International Agree-ment Report s

== Description:== Members of the Intemational Code Assessment Program (ICAP) have assessed the U.S. Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at Idaho National Engineering Laboratory (IhTL), are summarized. Code deficiencies are discussed and user recommended noch1mtions investigated dtuing the course of conducting the assessment studies and reviews are listed. All the work that is sammarized was done using the RELAPS/ MOD 2, RELAP5/ MOD 3, and TRAC-B codes. Publication Date: December 1993 Prepared by: Schultz, R.R. [EG and G Idaho,Inc., Idaho Falls,ID (United States)] Prepared for: Nuclear Regulatory Commission, Washington, DC (United States), Office of Nuclear j Regulatory Research; EG and G Idaho, Inc., Idaho Falls, ID (United States) Keywords: BWR type reactors, coordinated research programs, ECCS, evaluation, heat transfer, hydraulics, loss of coolant, nuclear power plants, PWR type reactors, reactor cooling systems, reactor safety,RELAPS/ MOD 2,RELAP5/ MOD 3 TRAC-B l l i.; NUREG/BR--0083, Vol.9 108

i .i .?i f APPENDIX A: Index by NUREG-Series Report Number - i r 1 f il Report Number Page j 4 i NUREG-0713.Vol.12 1 1 NUREG--0713-VoL13 _ 2 l NUREG--0713-Vol.14..... . 3 NUREG-1473..... _4 NUREG/CP--0040 5 NUREG/CP--0126-Vol.1 ...............6 NUREG/CP-0126.Vol.2...._.......... 7 J 1 NUREG/CP--0126-VoL3. .. 8 NUREG/CP--0130-Vol.1... ....<.....9 NUREG/CP--0130.Vol.2...... 10 i NUREG/CP-0131 . 11 NUREG/CP--0134...._..._.. 12 NUREG/CR-3469.VoL7.... ... 13 NUREG/CR--4214.Rev.1-PL2-Add.2 ..... ~. 14 NUREG/CR--4214.Rcv.2.PL1.._ 15 NUREG/CR-4219-Vol.9.No.2.. ..... 15 'l NUREG/CR-5247.Vol.1.Rev.1 .17 NUREG/CR-.5247.VoL2.. 18 NUREG/CR-5305-VoL2-PL1.... ..19 J NUREG/CR-5305.Vol.2.PL2.. 20 NUREG/CR-5360. .21-NUREG/CR -5471.-.... ... 22 NUREG/CR-5782.......... 23 NUREG/CR-5801. ..........._.._.......24 NUREG/CR--5817-Vol.3.No.2.. 25 NUREG/CR-5818 -........... . 26 NUREG/CR--5822 27' NUREG/CR--5843.. 28 NUREG/CR-5851 .... ~.~. _.. 29 i 1 NUREG/CR--5882..... 30 ] NUREG/CR--5901 _........ _ . ~.31 NUREG/CR-5907._. ~ 32 -l NUREG/CR-5927-Vol.1. ..~...... 33 NUREG/CR-5936...... 34 i i NUREG/CR--5937.... ......... 35. NUREG/CR--5942.. 36' 1 NUREG,nR-5943.... 7 3 NUREG/CR--5949.. 38 NUREG/CR-5957. ... 39 l NUREG/CR--5958 40 NUREG/CR-5964 41 NUREG/CR-5966..... ~.... 42 - NUREG/CR-5970.. _ 43 NUREG/CR-5976 44 NUREG/CR-.5978. ~. 45 NUREG/CRa5983.._ _ 46 NUREG/CR-5984..... . _..._47 NUREG/CR-5991.. _ 48 A-1 NUREG/BR--0083, Vol.9

Report Number Page .... 49 NUREG/CR-5993.Vol.1 NUREG/CR-5993.Vol.2......... .........-.......50 MG/CR--5998.... ......_.....~_..........51 ... _.52 NUREG/CR-.6018 NUREG/CR--6021...~.._ - _ 53 NUREG/CR-6026 ..._........._54 NUREG/CR--6028.._. ......_........_55 NUREG/CR-62..,..._.. 56-NUREG/CR-.6035. 57 NUREG/CR-.6041...- 58 NUREG/CR--6@9 . ~ 59 NUREG/CR-.6050.. _....... - ._60 NUREG/CR--6052._............... _.. 61 NUREG/CR--6054.. ....._.-........62 MG 56 i ..~. .~ NUREG/CR--6059. ............-,.....~......_64 'i i NUREG/CR-.6060............. 65 NUREG/CR--6061.... . 66 NUREG/CR--6070... ._....._,~........._..67 1 .... ~. 1 0 71 __....~....._.....68 NUREG/CR.-6072...... 69 NUREG/CR--6082.. -......_...........70 NUREG/CR-.6083..... ..... ~ ......~... ~..... ... 71 i NUREG/CR-.6090. ...................._.72 NUREG/CR--6101. 73 NUREG/CR-.6113_. ~. 74 NUREG/GR--0006 _...~ 75 NUREG/GR--0009....... ...._....._ 76 NURE0/GR--M10.._. 77 NUREG/IA -0085._........ . 78 NUREGAA--009O .......-...........79 NUREGAA-0091....................... 80 MEGSA-0092........ .........~.~....81 NUREGAA--0094................ _....~. . ~ 82 NUREGSA--0095. .....~ ~.. 83 s l NUREG/lA-.0096. _ _..__ 84 ) NUREGAA--0099.. . 8 5 NUREGAA--0100...._._............._...... 86 NUREGAA-0103... 87 i NUREGAA-01N 88 i NUREG/1A-0105. 89 NUREGAA--0106. ..~._._.. . 90 NUREGAA--0107 ..~... ~... . 91 NUREG/IA--0108. .. 92 NUREGAA--0109... ~ ~.. ~ 93 NUREGA A--0110....... ..... ~ _... ~ _ 94 NUREGAA -0112... _ _ _ 95 NUREGAA-0113..~.................... 96 NUREGAA-0116.-..... ....~ _._ 97 NUREGSA--0118 .. 98 NUREG/IA-0119 .-_......_.._-....-.99 NUREG/BR--0083, Vol.9 A-2

Report Number Page NUREGAA--0120........ 100 NUREGAA--0121......... 101 NUREGAA--0122..... 102 NUREG/IA--0123....... 103 NURE A--0124. 104 NUREG/1A--0125... - 105 NUREG A--0126 106 NUREG A--0127... ........... 107 NUREG/IA--0128.... 108 t 1 I a i 1 l i a 1 1 A-3 NUREG/BR--0083, Vol.9 \\

~ I p. 1*; . APPENDIX B: Index by Software Identification Software Identification Page 3 DEL,...... 5 3 ABAQUS. 7,11,16,53,63 ACE 13 ACR1TH..._. . 10 A 1,5. ADINA FE. 11 ADINA-T.. 11 - .. ~.. ANS YS 53 APRIL. MOD 3._... . 7 ARRO'ITA. ........_...6 ATHLET-SA 7 BARRIER 33 BEASY 53 BETA.. 6. BIGFLOW 25,55 BLT. . 33,37 BOSOR4 .. ~......~. 39 BOSOR5 . 39 BUSCA.. _.._... 7 BUTRAN 7 CAFTA 6 CART.. 8 CASTEM2000 11 CATHARE2. 7 CECP. . 62 I CEMENT 67 CHARM 7 CIIEKWORK.._ 6, CM 7 CmiTRN _ ._..__.. 53 CHEQMATE. 53, CHIP 7 - CMVSFS 55 : COCOMO 71 ' COMBIN.. 19,20 COMMIX. 6,27 ' Q.DUX 1B _. 11.- CONTAIN. 6,7,21 CORBH 6,36 CORCOh,. MOD.2........ 6,56 CORCON-MOD 3... 28 . CORMLT 7 CORSOR M ._. 7 l COUPLE. 7 QSYMA 10 CRAC 15 CS AU... 26 ii l = B 1-NUREG/BR 0083,Vol.9 /L l r'

Software Identincation Page CTM.. .........~._ .............._....53 DATAFLOW..._........ _.....~ ..........._.........._.55 DECAY .. _...... _..._...... 17,18 DEIMOS..... ..._._.................._...7 9,75 : DEPOSITION......... ... 58 DU ST......_............. DUSTIN 58 DYNAMIX ......._............................53 ECHEM.. 53 77 EDDYANN. _....... EDSFI...... .........._4 ..8 EMTP ENDF/B.VI... ..... 68 ES CADR....._.._.. .. - 7 ..7 ESPROSE ESTER.......,_.. ....._.~......_.._......7 ... I 3 ESTS.. .... 53 EQ 3/6... 6 EQHA*l.ARD..... 19 EVNTRE.. FASTCHEM.... 53 i FASTG RAS S.. 7 16 FAVOR...._........... FEHMN... 53,54 FEM. _........_..._....._.........63 FEMWATER .. 33.37 FIPLOC .._.......7 FIRAC.. _... ~... 10 FLAC.. .....................53 FLECHT.SEASET............. .._..........._....7 i ............7 FLOW 3D.... FMEA. .. 73 FMECA....... 73 FM. DOSE................ 17,18 FPRATE...... .......~...7 FRA C'IURE.TWO......._. ......... 1 1 FRAC. UNIX .53,54 FRAP.T6.... -............................................~..7 FREY............_......~.. .....,...........................~_.....6 G ENA S Y S .. 53 Genera 1. _.. 6,7,8,11,19,24,31,33,40,43,45,49,50,52,53,54,61,63,69,70,71,72,74 G ENII...............,......_... _.. _.. 3 3 GEOCHEM.. -................._..... 53 GEOTHER m. 53,54 GOTHIC.......................... .....~..........._..6 GPBEST 3D ..... _53 G RAFXT.. ...... 58 GWHRT _ 54 HEATING.6....... 76 HMS ...-..._.._...._......~._....65 53 HYDRAQL... NUREG/BR-.0083, Vol.9 B.2

Software Identification Page ICARE.2. .._.....7 INSPECT ..~. ..._...._......7 IODE....... ......................._...7 IPRD S __..._.....................22 IRRAS - .. 34,41 IWM-CRACK..._... ................_......11 IWM.. VERB. ......._.-.......11 c JER1CHO.......... ._ 7 1 KESS ..~....................._....._.........7 LAFIS. .. 10 LER...._.. 12,29 LERS. ._........_................8 S 1 9 3 _..... .........-.....77 e I.SODES.. ..._........_....._.........54 MAAP.. _.... ......._...........6 i MAAP 3.0...... ..... 7 .... ~.... MACCS... 8.14,15,19,20,63,64 MARCH......... . 63 i l MARCH 3.... ... 7 MASTERK., _.................19 i I MATTUM....... . 54 j MELCOR.. 6,8,36,56 MELPROG: ...........7 MELPROG/ MODI...... ... 63 j MELTS PREAD........ . 7 MINET.......... 6 MINTEQA2.... . 25 MITRA....... ............_...........7 MODFLOW -- . 33 MODPAm. 33 MPATH ............_.......53 MSC/NASTRAN.. 53,63 M. STS...... .. ~.... 51 . NEFTRAN.......... _... .. 33,53 4 NEWMIX. . _ 1 1 NEWPART............. 20 NORIA.... 53,54 NPRDS... . 6.8,12,22,29 NUMARC. .6 NUDOCS/AD .............~ 8 OC A.P..... 11,16,23 OCM3 D............ 54 ODRPACK 2.01.. 8 OPEC-2.........~..... 22 ORAM-TIP 6 ORIG EN.... 14 ORIGEN2.............. 7,8 ORMGEN...........-.... 11 . ORVmT. 11 i PAGAN._.. ~ 33 PARTITION.. . 19,20,63 B-3 NUREG/BR--0083,Vol.9

~

i.

.i -t 4 Software Identification - Page ] k PAT 2SR5 7 7,11,16 L .... _.. ~... . PECLOX ....._._~..7 PETROS.. .~ 53,54 J t . PFA......__............._..._.....6 3 PHREEQE.. 53 PM-ALPHA 7 PORFLOW 4 8,53,54,55 PR.EDB.. 8 a PRA.... ..._............._....................44 PRAISE 8 PRAMIC. 63. PR AMIS..... 19 - l PRUEP...................... _.. 7.19,20 PS AFE2.... 59 PSTEVNr........... 19 Y i RAESTRICT....... 33 RAFT _ ...........7 RALOC MOD 2.2... 7 -i RAPID. 6 RASCAL 2.0........_.. 17,18 S 1 R 5.....-_._......._......7 R S .. 57 .f S OD1........... 03 RELAP5/ MOD 2......._.... 78,79,80.81,82,83,85,90,91,92,93,94,95,97,98,99,101,103,104,105,108 - RELAP5/ MOD 3...... _ 6,26,81,82,83,84,85,86,87,88,89,90,96,97,108 ' RELPIN.... e RF.Mrr _ ._.60 REMIX 11 RETRAN 3 6 ROCMAS. _. 53 RSY6, AL 7 S ANGRE.. 53 S APHIRE.. 41 SARA..._. .._.._,......._....41 F S AS....__.. . 19 a SCDAP 76 SCDAP/RELAP _..._. ._63. SCD 5 03.. 1 SOFIRE.M II 10 i ~ SOLA-PTS 11 l SPECTROM.32 53 SPRAY-11 lO SSC . _._ 6 ST. DOSE _..... _ ~ .17,18 - STCP ...~....... 7.36,42 e h STEALTH 53 STER - 19,20 STRES3D _._-. 53 SUPCRT._ . _53 5 1 'I NUREG/BR-.0083, Vol.9 B.4 1 -,u m

Software Identification Page SUPERTREE...... ... _. 63 S URFIT.. ....,.-._...._.8 TAL,._................................ .....~......_..8 TACMVS -- ..................~8 TEMAC. _....__....._............19 TEMPEST 1 1 THAMES ...~. 53 TH A'ICH... .. _ 46,47 THCC .. 53 TOUGH. 5.53.54 r 'lR.EDB_ 8 TRAC...._.. 106,107 TRAC.B...

........_.._..._.~..._....~.....108 1

TRAC-BF1._..... 30,100,102 TR A%...-.. _...._ _.....................6 TRACR3D... _ 53 TRANQL _~........._._.._..................53 TRAPFRANCE..... ~ _... 7 .. ~....... TRAPMELT - ... 7 UDEC....... 53 VAM2D.. _,3 3,51 - VICTORIA. ....7 ...... ~... - VIPRE._. 6 VISA.II.... ... _ 16 VISCRK.. -.... .... 1 1 VS2DT .. ~ 33 VTOUGH .. 5.53,54 VVTVIRT ....... 1 1 WATEQ. .... 53 WECHSL 7 r WETCOR-1...._. .. 32 XSOR 19,21,63 P 1 1 1 a B.5 NUREG/BR-.0083, Vol.9

4 4 APPENDIX C: Index by Contractor Report Number 1 i Report Number .Page AEATR.c-.1050. 84 AEEW-R--2501 . 84 ~.. - ANL-91/43 27 ANL--93/9.... 56 T BENUREG--51708.Vol.7 ....-............13 BENUREG-52346.. ...~... 3 7 f BENUREG-52355... 46 BENUREG-52356 ._ 47 BENUREG-52362.Vol.1........... '.49 - BENUREG--52362.Vol.2 50 BENUREG--52375.... ..~ ~. 58 BENUREG--52377 ... 59' CDNSWC/SME-CR.-16 92.. 40 CDNSWC/SME-CR-18-92......_ 43 i CNWRA-92-003 48 ' E CNWRA-92 006 ~. 54 CNWRA-.92-011...... 53 CNWRA-92-026.. ~ 55 CNWRA-.92-02S-Vol3-No.2.. ~. 25 l ECN-89-91.... . 80 ECN.C--92-008 95 EGG-2610 66 EGO-2 5 _. . 26 - .5 G-2677 0 EGG-2688..... 35 EGG 2689... 38 EGG-2692 41 EGG-2694.. 44 - EGG-2701 _ 67 EGG. EAST-8719........ 108 1 4 EPRI-TR-102106 _.... ~ 52 -l EST-SIAN-.22.. ...... 100 - GRS-- 100...... _._ . 106 GRS-101 107 ICSP-AL-BOUT-R........ . 103 ICSP-AS-BOUT.R.. 99 ICSP--AS-SPR-R. .~. 101 ICSP-TR-TTRIP.R ....... 78 ICSP--V2.R100-R... ~. 92 - ICSP--V2-R50-R. .... 91 - - ICSP_V2-TURFW-R _._ ..... 94 IS-5083.. .'. 3 9 ITRI-141 .... ~. . 15 KWU-.E412/91/E1002 97 LA-12593.M _65 LMF-136 . 14 MPR-1345... _106 MPR--1346 107 ORNI./Sub-93.SD684,_ _..~. .._61 C-1 NUREG/BR--0083, Vol.9 4 l l - i l

Report Nomber Page - ORNI/rM-11945...~. 23 ORNI/IM 12229 36 l ORN1/rM-12406 .................68 ORN!/rM-9593-Vol.9-No.2 16 PNL--8454-Vol.1-Rev.1 17 ........ ~.. _ PNL-8496........... _..............51 62 RRCKI--80-05/3 ....... 69 . ~........ 1 SAIC--91 ......._...............52 l. L SAIC-93/1310-01.. .._...........,........60 S AND-89-094 3.......... .... 21 SAND-89-2562 22 S AND-90-2765.Vol.2-PL1.. _.... ... 19 SAND-90-2765.Vol.2-PL2 .............._.........20 S AND-.91 2802-Vol.1.............. .....................33 SAND-.91-7087._. -- ...._..................24 ) S AND-92-0167.. . 28 S AND-92 1422._.. 31 32 S AND-.92-1563.. S AND. 92 2109...... -......... ...... 34 ..............~.......z SAND-92-2146 ... 64 SAND-92 2688 ._.....__........__45 SAND. 92-2689 _...._..........-.._.....42 . STUDSVIK/NS--90/93..... . 82 TD/SPB/ REP--0130.._.............. .. 98 TEC/L-0471/R91......... 90. UCRL.ID-.112900.. 72 UCRL.ID-114565 71 UCRL.ID--114567............... -... ~ 70 UCRL.1D-114839 . 73 UILU.ENG-92-2016. 43 UNID--91-08.... ... 93 VARGOS-93/1...... ...... 69 .-NUREG/BR--0083 Vol.9 C-2

APPENDIX D: Index by Keyword Keyword NUREG Report Number 3DEC NUREG/CR--6021 ABAQUS NUREG/CP--0126-Vol.2 NUREG/CP--0131 NUREG/CR--4219-Vol.9-No2 NUREG/CR--6021 NUREG/CR--6056 accuracy NUREG/CR--6061 ACE NUREG/CR--3469-VoL7 ACRITH NUREG/CP--0130-Vol.2 ADINA NUREG/CP--0131 NUREG/CR--6021 ADINA FE NUREG/CP-0131 adsorbents NUREG/CP--0130-VoL2 advection NUREG/CR-5943 AEROSOL NUREG/CP--0126-Vol.2 aerosols NUREG/CP-0130-Vol.1 NUREG/CR--5247-Vol.1-Rev.1 NUREG/CR-5907 NUREG/CR-5966 NUREG/CR-5978 I NUREG/GR--0006 after-heat removal NUREG/CR--5983 NUREG/IA--0091 aging NUREG/CP--0126-Vol.3 NUREG/CP-0134 NUREG/CR-5851 air cleaning NUREG/CP-0130-Vol.1 NUREG/CP-0130-Vol.2 air filters NUREG/CP--0130-Vol.1 NUREG/CP--0130-Vol.2 air quality NUREG/GR-0006 ALARA NUREG/CR--3469-Vol.7 D-1 NUREG/BR-0083, Vol.9

i ] Keyword' NUREG Report Number algorithms NUREG/CR--5964 NUREG/CR--6028 l NUREGAA--0094 l Almaraz-1 reactor NUREGAA-0123 Almaraz 2 reactor NUREGAA--0123 analytical solution NUREG/CR--4219-Vol.9 No.2 ANSYS NUREG/CR.-6021 APRILMOD3 NUREG/CP--0126-Vol.2 array processors NUREG/CR--6113 ARROTTA NUREG/CP--0126-Vol.1 l ASCO-1 reactor NUREGAA--0119 NUREG/1A--0121 ASCO-2 reactor NUREG/IA--0119 NUREGAA--0121 ATHLET-SA NUREG/CP--0126-Vol.2 attenuation NUREG/CR-5978 automation NUREG/GR--0010 availability NUREG/CP--0134 NUREG)CR-5993-Vol.2 BARRIER NUREG/CR-5927-VoL1 BEASY NUREG/CR--6021 Beaver Valley-2 reactor NUREG/CR-5976 BEIR-V model NUREG/CR--4214-Rev.1 Pt.2-Add.2 - benchmarks NUREG/CR-5943 NUREG/CR--6028 NUREG/CR--6049 . BETA NUREG/CP--0126-Vol.1 bibliographies NUREG/CR-3469-Vol.7 . BIGFLOW NUREG/CR-5817-Vol.3-No.2 NUREG/CR--6028 ' NUREG/BR--0083, Vol.9 - D 1

Keyword NUREG Report Number bilateralagreements NUREG/IA-0110 biologicalradiation effects NUREG/CR--4214-Rev.1-Pt.2-Add.2 NUREG/CR--4214-Rev.2-Pt.1 NUREG/CR-5247 Vol.1-Rev.1 Black Fox-1 reactor NUREG/CR-5882 Black Fox-2 reactor NUREG/CR--5882 blackouts NUREG/CR--5942 NUREG/CR-5949 NUREG/IA-0119 blowdown NUREG/CR--6060 NUREG/CR--6061 NUREG/IA--0104 BLT NUREG/CR-5927-Vol.1 NUREG/CR--5943 boron additions NUREG/CR--5822 Borssele reactor NUREG/IA-0091 BOSOR4 NUREG/CR-5957 BOSORS NUREG/CR-5957 boundarylayers NUREG/CR-5970 rubbles NUREG/CR-5901 BUSCA NUREG/CP-0126-Vol.2 i BUTRAN NUREG/CP--0126-Vol.2 BWR type reactors NUREG/CP--0126-Vol.1 NUREG/CP-0126-Vol.2 NUREG/CP--0126-Vol 3 1 NUREG/CR-5360 NUREG/CR-5936 NUREG/CR-5966 NUREG/CR--5978 t NUREG/CR--6035 ) NUREG/CR--6049 l NUREG/CR--6056 NUREG/IA-0128 l CAFTA NUREG/CP-0126-Vol.1 calibration NUREG/CR-5851 i l D-3 NUREG/BR--0083, Vol.9 1

Keyword NUREG Report Number Calvert Cliffs-1 reactor NUREG/CR-5937 Calvert Cliffs-2 reactor NUREG/CR--5937 I l carbon NUREG/CP -0130-Vol.2 l l CART NUREG/CP -0126-VoL3 ) 1 CASTEM 2000 NUREG/CP-0131 CATHARE2 NUREG/CP--0126-Vol.2 CDC computers NUREG/IA--0108 CECP NUREG/CR--6054 CEhENT NUREG/CR--6070 CHARM NUREG/CP -0126-Vol.2 CIEKWORK NUREG/CP -0126-Vol.1 chemicalreaction kinetics NUREG/CR--6021 NUREG/CR--6060 chemicaireactions NUREG/CR--6060 CHEMKIN NUREG/CP-0126-Vol.2 CIEMTRN NUREG/CR--6021 CHEQMATE NUREG/CR--6021 CHIP NUREG/CP--0126-Vol.2 CMVSFS NUREG/CR--6028 COCOMO HUREG/CR--6083 Cofrentes reactor NUREG/1A -0120 COMBIN NUREG/CR--5305-Vol.2-Pt.1 NUREG/CR--5305-Vol.2-Pt.2 t combustion NUREG/CR--6060 combustion kinetics NUREG/CR--6072 combustion products NUREG/CR--6072 commercialsector NUREG-0713-Vol.14 NUREG/BR-0083, Vol.9 D-4

i l Keyword NUREG Report Number COMMIX NUREG/CP-0126-Vol.1 NUREG/CR--5822 COMMIX IB NUREG/CP--0131 -l comparative evaluations NUREG/CR--6113 compiled data NUREG--0713-Vo?.12 NUREG--0713-Vol.13 computer-aided design NUREG/CR--6113 computer arehitecture NrJREG/CR--6090 NUREG/CR--6113 i computer calculations NUREG/CR-5942 NUREG/CR--5949 NUREG/CR--6M9 NUREGBA--0085 NUREGSA--0091 NUREG/IA--0095 NUREGAA--0096 NUREG/IA-0099 NUREG/IA--0103 NUREGSA--01M NUREGAA-0112 NUREG/IA-0113 NUREG/IA--0118 NUREGAA--0119 NUREGSA--0122 NUREG/IA--0125 computer codes NUREG/CR--6113 computer graphics NUREG/CR--5247-Vol.1 Rev.1 NUREG/CR--6050 computer program documentation NUREG-1473 NUREG/CR-5247-Vol.2 NUREG/CR--5305-Vol.2-Pt.1 NUREG/CR-5305-Vol.2-Pt.2 NUREG/CR--6035 NUREG/CR--6059 NUREG/GR--0010 NUREG/IA--0091 NUREGAA--0109 NUREG/IA--0118 computerized control systems NUREG/CR--6082 NUREG/CR--6083 NUREG/CR -6090 D-5 NUREG/BR--0083, Vol.9

~_ _ t t - Keyword ' NUREG Report number l l computerized simulation' NUREG/CR--4219 VoL9-No.2 NUREG/CR-5818 NUREG/CR-5843 NUREG/CR-5943 NUREG/CR-5957. NUREG/CR-5998 NUREG/CR--6028 { NUREG/CR-6060 NUREGAA-0085 NUREGAA--0091 NUREGAA--0092 NUREGAA--0094 NUREGAA--0100 NUREG/IA-0105 - NUREGAA-0106 NUREGAA--0108 NUREGAA--0110 NUREGAA--0113 NUREGAA--0116 NUREG/IA--0120 NUREGAA-0123 NUREGAA-0125 concretes NUREG/CR-5901 NUREG/CR-5907 - NUREG/CR-5978 NUREG/CR-6032 CONTAIN NUREG/CP-0126-VoL1' NUREG/CP--0126-VoL2 NUREG/CR-5966 containers NUREGKR-5817-VoL3-No.2 NUREG/CR--5927-VoL1 NUREG/CR-5943 NUREG/CR--6041 containment NUREG/CP--0126-VoL2 NUREG/CR--5843 NUREG/CR-5937 NUREG/CR-5942 NUREG/CR--5949 NUREG/CR--6070 j containment buildings NUREG/CR-5957 i containment spray systems NUREG/CR-5966 containment systems NUREG/CR-5942 NUREG/CR--5993-VoL2 NUREG/CR--6060 -j - NUREG/CR--6070 - NUREG/BR--0083,Vol.9 D-6

Keyword NUREG Report Number coordinated research programs NUREGAA-0107 NUREG/1A-0112 NUREGSA--0119 NUREG/1A-0126 t NUREGilA-0127 NUREGAA-0128 CORBH NUREG/CP--0126-VoL1 NUREG/CR-5942 CORCON-MOD 2 NUREG/CP-0126-Vol.1 NUREG/CR--6032 CORCON-MOD 3 NUREG/CR--5843 corium NUREG/CP--0126-Vol.2 NUREG/CR--5843 NUREG/CR-5907 NUREG/CR-5949 NUREG/CR-5978 NUREG/CR--6032 NUREG/GR--0009 CORMLT NUREG/CP--0126-VoL2 conosion NUREG/CR-5817-VoL3-No.2 CORSOR-M. NUREG/CP--0126-VoL2 cost-benefit analysis NUREG/CR--6018 cost NUREG/CR-6059 cost estimation NUREG/CR--6054 COSYMA NUREG/CP-0130-Vol.2 COUPLE NUREG/CP--0126-VoL2 CRAC NUREG/CR--4214-Rev.2-Pt.1 i crack propagation NUREG/CP--0131 NUREG/CR--4219-VoL9-No.2 NUREG/CR-5958 i NUREG/CR-5970 l cmeks-NUREG/CP-0131 NUREG/CR--4219-Vol.9-No.2 NUREG/CR-5970 NUREG/CR-6070 criticalheat flux NUREGAA--0094 l D-7 NUREG/BR--0083, Vol.9 i

3 Keyword NUREG Report Number criticality NUREG/CR--5983 cross sections NUREG/CR--6071 CSAU NUREG/CR-5818 CTM NUREG/CR--6021 Darcy's Law NUREG/CR--0040 NUREGEP-6026 ) data acquisition NUREG/CR-5471 data acquisition systems NUREG/CR--6072 data analysis NUREG--0713-Vol.14 NUREG/CR--6061 database management NUREG--0713-Vol.14 NUREG/CR-5976 NUREG/IA--0090 data compilation NUREG/IA-0094 data covariances NUREG/CR-5818 NUREG/CR--5927 Vol.1 j NUREG/CR-5978 data processing NUREG/CR--6050 NUREG/GR--0010 data transmission NUREG/CR -6082 data transmission systems NUREG/CR-6082 DATAFLOW NUREG/CR--6028 DECAY NUREG/CR-5247-Vol.1-Rev.1 NUREG/CR-5247.Vol.2 decision making NUREG/GR--0010 decommissioning NUREG/CR--6054 decontamination NUREG/CP--0130-Vol.1 NUREG/CP--0130-Vol.2 NUREG/CR-5901 NUREG/CR-5966 NUREG/CR-5978 NUREG/CR--6054 defects NUREG/CR-5782 NUREG/GR-0010 i NUREG/BR--0083, Vol.9 D-8

l Ke'yword NUREG Report Number l deformation NUREG/CR-5957 -l 2 DEIMOS NUREG/CP--0126-Vol.2 DEPOSITION NUREG/CP--0130-Vol.1 NUREG/GR-0006 depressurization NUREG/CR-5937 - t NUREG/CR-5949 NUREG/IA-0095 j NUREG/IA--0121 NUREG/IA-0127 design NUREG/CR-5957 NUREG/CR--6018 ] J design basis accidents NUREG/CR-5983 detonations NUREG/CR--6072 j diagrams NUREG/CR--6056 diffusion NUREG/CR-5901 NUREG/CR--6060 digital systems NUREG/CR--6113 NUREG/GR--0010 dilution NUREG/CR-5822 i 1 documentation NUREG/CR-5247-Vol.2 i NUREG/CR-5472 dosimetry NUREG/CR-5927-Vol.1 NUREG/CR--6050 NUREG/CR--6071 dryout ' NUREG/IA-0^94 ducts NUREG/GR-0006 ) DUST NUREG/CR--6041 DUSTIN NUREG/CR--6041 4') I dynamic loads NUREG/CR--6049 -j NUREG/CR--6052 1 DYNAMIX NUREG/CR--6021 .canhquakes NUREG/CP-0126-VoL3 i D-9 NUREG/BR--0083, Vol.9

Keyword NUREG Report Number ECCS NUREG/CR--5818 NUREG/CR-5942 NUREG/1A--0127 NUREG/IA--0128 ECHEM NUREG/CR--6021 economics NUREG/CR--6059 eddy current testing NUREG/GR-0010 EDDYANN NUREG/GR-0010 EDSFI NUREG-1473 educational tools NUREG/CR-5247-Vol.2 electrical equipment NUREG-1473 electronic equipment NUREG/CR--6090 cmbrittlement NUREG/CR-5782 emergency plans NUREG/CR--6059 emission NUREG/CR-5907 EMTP NUREG/CP--0126-VoL3 ENDF/B-VI NUREG/CR--6071 energy transfer NUREG/GR--0009 engineered safety systems. NUREG/CR--6083 NUREG/CR--6090 entrainment NUREG/IA--0096 environmentalexposure pathway NUREG/CR-5927-Vol.1 EPRI NUREG/CP--0126-VoL1 EQ 3/6 NUREG/CR--6021 EQHAZARD NUREG/CP-0126-Vol.1 Equivalent Continuum Model(ECM) NUREG/CP--0040 ESCADR NUREG/CP--0126-VoL2 j ESPROSE NUREG/CP--0126-Vot2 NUREG/BR--0083, VoL9 D-10

Keyword NUREG Report Number ESTER NUREG/CP--0126-Vol.2 ESTS NUREG/CR-3469-Vol.7 evacuation NUREG/CR--6059 evaluation NUREG/CR--6018 NUREGAA--0095 NUREGAA-0103 NUREGAA--0104 NUREG/IA-0108 NUREGAA--0109 NUREGAA--0121 ) NUREG/lA-0128 EVNTRE NUREG/CR-5305-Vol.2-Pt.1 excursions NUREGAA--0107 experimental data NUREG/CR--6072 NUREGAA--0090 NUREGAA-0108 NUREGAA-0112 NUREG/1A--0116 expert systems NUREG/CR--6018 NUREG/GR--0010 explosions NUREG/CP--0126-VoL2 failure mode analysir NUREG/CR-5801 NUREG/CR-5976 NUREG/CR-5993-Vol.1 NUREG/CR--5993-Vol.2 NUREG/CR--6056 failures NUREG/CR-5471 NUREG/CR-5801 NUREG/CR-5817-Vol.3-No.2 NUREG/CR-5851 NUREG/CR-5943 NUREG/CR-5983 NUREG/CR-5993-Vol.1 NUREG/CR-5993-Vol.2 NUREGAA--0121 NUREG/IA--0124 FASTCHEM NUREG/CR--6021 FASTGRASS NUREG/CP-0126-Vol.2 fault tolerant computers NUREG/CR--6101 NUREG/CR--6113 D-11 NUREG/BR--0083, Vol.9

.m Keyword N' UREG Report Number .t 3 fault tree analysis NUREG/CR-5801 NUREG/CR-5964 NUREG/CR--5976 e' FAVOR NUREG/CR--4219-Vol.9-No.2 ~ feasibility studies NUREG/CR-5936 l .i FEHMN NUREG/CR -6021 NUREG/CR--6026 i FEM A'JREG/CR--6056 FEMWATER NUREG/CR--5927-Vol.1 NUREG/CR-5943 a field tests NUREG/CR--5998 finite difference method NUREG/CR--5957 finite element method NUREG/CR--4219 VoL9-No.2 NUREG/CR--5943 NUREG/CR--5970 Finland NUREG/IA-0090 i FIPLOC NUREG/CP--0126-VoL2 FIRAC - NUREG/CP-0130-VoL2 ii fission product release NUREG/CP-0126 Vol.2 ~ NUREG/CR--4214-Rev.1-Pt.2-Add.2 NUREGfR-5247-VoL2 ~ NUREG/CR--6059 .i FLAC NUREG/CR--6021 flammability NUREG/CR-6072 l FLECHT-SEASET NUREG/CP-0126-Vol.2 flow models NUREG/CR-5822 NUREG/CR-5991 ' NUREG/CR-5998 NUREG/CR--6035 NUREG/IA--0096 : a o ~ . flow rate ' NUREG/CR-5943 NUREG/GR--0006 NUREG/IA--0105 NUREG/IA--0116 NUREG/IA--0125 - NUREG/BR--0083, Vol.9 D-12

q + c Keyword NUREG Report Number. FLOW 3D NUREG/CP--0126-VoL2 fluid flow NUREG/CP--0040 I NUREG/CR-5822 NUREG/CR-5927-VoL1 NUREG/CR-5984 NUREG/CR--5991 NUREG/CR-5998 NUREG/CR--6021 ' NUREG/CR-6026 NUREG/CR--6028 NUREG/CR--6035 j NUREG/IA -0096 NUREG/IA--0100 NUREG/IA--0105 FM-DOSE NUREG/CR--5247-Vol.1 Rev,1 l NUREG/CR--5247-VoL2 FhEA NUREG/CR--6101 l FMECA NUREG/CR-6101 1 FPRATE NUREG/CP--0126-Vol.2 ' FRAC-UNIX NUREG/CR--6021 NUREG/CR--6026 : .I 1 fracture mechanics - NUREG/CP--0131 NUREG/CR-5782 NUREG/CR-5958 NUREG/CR-5970 i i fracture properties. NUREG/CP -0131 NUREG/CR -4219-VoL9-No.2 NUREG/CR-5958 - NUREG/CR-5970 I i l FRACTURE.TWO NUREG/CP--0131' i fractured reservoirs NUREG/CP--0040 l fractures NUREG/CR--6026 FRAP-T6 NUREG/CP--0126-VoL2 NUREG/CR--6061 FREY.. NUREG/CP--0126-Vol.1 Fuchs model' NUREG/CR-5901 fuelcans NUREG/CR--6061 l - l D-13 NUREG/BR--0083,Vol.9

Keyword NUREG Report Number fuelreprocessing plants NUREG/CP-0130-Vol.1 fuelrods NUREG/CR--6061 fuel-air ratio NUREG/CR--6072 fuel-cladding interactions NUREG/CR--5843 Garona reactor NUREG/IA-0122 gases NUREG/CR-5901 GENASYS NUREG/CR--6021 general NUREG/CR--6018 NUREG/CR -6052 NUREG/CR--6072 NUREG/CR--6113 genetic radiation effects NUREG/CR--4214-Rev.1-Pt.2-Add.2 GENII NUREG/CR-5927-Vol.1 GEOCHEM NUREG/CR--6021 geochemistry NUREG/CR-5817-Vol3-No.2 NUREG/CR-5927-VoL1 i geologic formations NUREG/CR--6028 geologic fractures NUREG/CR-5817-Vol.3-No.2 NUREG/CR-5991 l GEOTHER NUREG/CR--6021 NUREG/CR--6026 Germany NUREG/IA--0126 GOTHIC NUREG/CP--0126-Vol.1 GPBEST 3D NUREG/CR--6021 GRAFXT NLTREG/CR--6041 Grand Gulf 1 reactor NUREG/CR-5936 Grand Gulf-2 reactor NUREG/CR-5936 ground motion NUREG/CP--0126-Vol.3 groundwater NUREG/CP--0040 NUREG/CR-5927-Vol.1 NUREG/BR--0083, Vol.9 D-14

Keyword NUREG Report Number GWHRT NUREG/CR--6026 hazards NUREG/CR--6082 health hazards NUREG/CR--4214-Rev.1 Pt.2-Add.2 NUREG/CR--4214-Rev.2-Pt.1 NUREG/CR--5247-Vol.1-Rev.1 NUREG/CR--6059 heat exchangers NUREG/GR--0010 heat transfer NUREG/CP--0126-Vol.2 NUREG/CR--5882 NUREG/CR--5907 NUREG/CR--5937 NUREG/CR-5942 NUREG/CR--5949 NUREG/CR-5978 NUREG/CR-5983 NUREG/CR--5984 NUREG/CR-5991 NUREG/CR--6026 NUREG/CR--6035 NUREG/CR--6056 NUREG/IA--0085 NUREG/IA--0091 NUREG/IA--0094 NUREG/IA--0095 NUREG/IA-0096 NUREGAA--0099 NUREG/tA--0100 NUREG/IA--0103 NUREGAA--0104 NUREG/IA--0105 NUREGBA--0106 NUREG/IA--0107 I NUREGAA--0112 ] NUREGAA--0113 NUREGAA--0118 NUREGAA--0119 NUREGAA-0121 NUREG/IA--0122 NUREG/IA--0124 NUREG/IA--0125 i NUREGAA--0126 NUREG/IA-0127 i NUREGAA-0128 heating NUREG/CR-5937 1 NUREG/CR-5949 HEATING-6 NUREG/GR--0009 l i D-15 NUREG/BR--0083, Vol.9 i-

Keyword NUREG Report Number high pressure coolant injection NUREGSA--0105 high-level radioactive wastes NUREG/CR-5817.Vol.3-No.2 NUREG/CR-5991 NUREG/CR--6021 NUREG/CR-4026 HMS NUREG/CR--6060 HTGR type reactors NUREG/CR--5983 NUREG/CR-5984 human factors NUREG/CP--0126-Vol.1 NUREG/CR--6056 hydraulic transport NUREGAA-0100 HYDRAQL NUREG/CR--6021 hydraulics NUREG/CP--0126-Vol.2 NUREG/CR-5817-VoL3-No.2 NUREG/CR-5882 NUREG/CR--5907 NUREG/CR-5937 NUREG/CR-5942 NUREG/CR-5949 NUREG/CR-5978 NUREG/CR-5983 NUREG/CR--5984 NUREG/CR--6035 NUREG/CR-4056 NUREG/IA--0085 NUREG/IA--0091 NUREG/IA--0094 NUREG/IA-0095 NUREGAA--0096 NUREG/IA--0099 NUREGAA--0100 NUREGBA--0103 NUREG/1A--0104 NUREG/IA--0105 NUREGBA--0107 NUREGBA--0112 NUREG/IA--0113 NUREG/IA--0118 NUREGAA--0119 NUREG/IA--0121 NUREGBA--0122 NUREGBA--0124 NUREGAA--0125 NUREG/IA--0126 NUREGAA--0127 NUREGAA--0128 NUREG/BR--0083, Vol.9 D-16

1 Keyword - NUREG Report Number - hydrogen NUREG/CR--6060 NUREG/CR--6072 ' hydrogen production NUREG/CR--6072 HYDROGEOCHEM NUREG/CR--6021 hydrology NUREG/CP--0040 NUREG/CR-5817-Vol.3-No.2 NUREG/CR--6021 - NUREG/CR--6026 NUREG/CR--6028 IAEA NUREG/CP-0131 ICARE 1 NUREG/CP.-0126-Vol.2 ICARE-2 NUREG/CP.-0126-Vol.2 ICRP model NUREG/CR--4214-Rev.1-Pt.2-Add.2 ignition NUREG/CR--6072 impacts NUREG/CR-5927-Vol.1 industrialradiography NUREG--0713-Vol.12 NUREG--0713-Vol.13 information systems NUREG/CR--6082 inhalation NUREG/CR--5247-Vol.1-Rev.1 INSPECT NUREG/CP--0126-Vol.2 ' inspection ' NUREG-1473 ~ NUREG/CR--6052 installation NUREG-1473 integral equations NUREG/GR--0009 interactions NUREG/CR--5907 ~ NUREG/CR-5978 NUREG/CR--6032 internationalagreements NUREGAA--0085 NUREGAA--0090-NUREGAA--0096 ' NUREG/IA-0099 NUREGAA -0103 NUREGAA--0105 - NUREGAA--0107 D-17 NUREG/BR--0083, Vol.9 -~

) l i Keyword NUREG Report Number international agreements (continued) NUREG/IA--0109 NUREG/IA-0118 NUREG/1A--0119 NUREG/IA--0122 NUREG/IA--0124 NUREG/IA--0127 Internet NUREG/CR -6082 IODE NUREG/CP--0126-Vol.2 IPRDS NUREG/CR-5471 IRRAS NUREG/CR-5936 NUREG/CR-5964 iterative methods hTREG/CR--6028 IWM-CRACK NUREG/CP-0131 IWM-VERB NUREG/CP--0131 J-Q lheory NUREG/CR--4219 VoL9-No.2 NUREG/CR-5958 Japan NUREG/IA--0126 NUREG/IA--0127 JERICO NUREG/CP-0126-Vol.2 KESS NUREG/CP-0126-Vol.2 La Salle County-2 reactor NUREG/CR-5305-Vol.2-Pt.1 NUREG/CR-5305 Vol.2-Pt.2 LAFIS NUREG/CP--0130-Vol.2 1 I laminar flow NUREG/CR--6060 l Latin Hypercube Sampling NUREG/CR-5305-VoL2-Pt.1 NUREG/CR-5927-Vol.1 NUREG/CR-5964 leaching NUREG/CR-5927 Vol.1 NUREG/CR-5943 LER NUREG/CP--0134 NUREG/CR-5851 LERS NUREG/CP--0126-Vol.3 f licensing NUREG/CR--6050 NUREG/BR--0083,Vol.9 D-18 = --- f

r 1 t f f Keyword NUREG Report Number life cycle NUREG/CR--6018 life cycle models NUREG/CR--6101 LOFT rt. actor NUREG/CR--6061 NUREGAA--0118 loss of coolant NUREG/CR-5782 NUREG/CR--5818 NUREG/CR-5942 NUREG/CR-5966 NUREG/CR--6061 NUREG/GR-0009 NUREG/IA-0095 NUREGAA--0096 NUREGAA-0099 NUREGAA--0103 NUREGAA--0104 NUREGAA--0112 NUREG/IA--0116 NUREGAA--0118 NUREGAA--0126 NUREGAA--0127 l NUREGAA--0128 LOTUS 1-2-3 NUREG/GR--0010 Loviisa-1 reactor NUREGAA-0090 Loviisa-2 reactor NUREGAA--0090 low-level radioactive wastes NUREG/CR-5927-Vol.1 NUREG/CR-5943 NUREG/CRe-5998 NUREG/CR--6070 LSODES NUREG/CR--6026 MAAP NUREG/CP--0126.Vol.1 MAAP3.0 NUREG/CP-0126-VoL2 MACCS NUREG/CP--0126-Vol.3 NUREG/CR--4214-Rev.1-Pt.2-Add.2 NUREG/CR-4214-Rev.2-Pt.1 NUREG/CR-5305-Vol.2-Pt.1 NUREG/CR-5305-Vol.2-Pt.2 NUREG/CR--6056 NUREG/CR--6059 maintenance NUREG/CP--0134 NUREG/CR--5993-Vol.2 D-19 NUREG/BR--0083, Vol.9

Keyword NUREG Report Number maintenance (continued) NUREG/CR--6052 NUREG/CR--6059 NUREG/CR -6082 NUREG/CR--6090 management NUREG/CR-5937 NUREG/CR--6056 manuals NUREG/CR-5247-Vol.1-Rev.1 NUREG/CR-5843 NUREG/CR-5964 l NUREG/CR-5991 MARCH NUREG/CR--6056 MARCH 3 NUREG/CP--0126-VoL2 Markov models NUREG/CR--6101 MASTERK NUREG/CR--5305-VoL2-Pt.1 mathematicallogic NUREG/CP-5964 mathematical models NUREG/CP--0040 NUREG/CR--4214-Rev 2-Pt.1 NUREG/CR-5801 NUREG/CR--5901 NUREG/CR-5957 NUREG/CR-5964 NUREG/CR-5970 NUREG/CR--5991 NUREG/CR--6026 NUREG/CR--6072 MATTUM NUREG/CR--6026 maximum credible accident NUREG/CR-4214-Rev.1-PL2-Add.2 measuringinstruments hWREG/CR--6072 mechanicalproperties hTREG/CR--6021 mechanicalstructures NUREG/CR--6052 meetings NUREG/CP-0040 NUREG/CP--0126-VoL1 NUREG/CP--0126-Vol.2 NUREG/CP--0126-VoL3 NUREG/CP--0130-Vol.1 NUREG/CP--0130-VoL2 NUREG/CP--0131 NUREG/CP-0134 NUREG/BR--0083, Vol.9 D-20

i i { Keyword NUREG Report Number MELCOR NUREG/CP--0126-Vol.1 NUREG/CP--0126-Vol.3 NUREG/CR--5942 NUPEG/CR--6032 - MELPROG NUREG/CP--0126-Vol.2 MELPROG/ MOD 1 NUREG/CR--6056 meltdown NUREG/CR-5843 NUREG/CR-5907 NUREG/CR-5949 NUREG/CR--6056 hELTSPREAD NUREG/CP--0126-VoL2 meteorology NUREG/CR-5247-Vol.1-Rev.1 MINET NUREG/CP--0126-VoL1 MINTEQA2 NUREG/CR-5817-Vol.3-No.2 mitigation NUREG/CR--6056 NUREG/CR--6059 MITRA NUREG/CP-0126-VoL2 mixing NUREG/CR-5822 NUREG/CR--6060 mixtures NUREG/CR--6032 MODFLOW NUREG/CR-5927-Vol.1 modifications NUREG/CP--0134 NUREG/CR--5936 NUREG/CR--6059 MODPATH NUREG/CR--5927-Vol.1 Monte Carlo method NUREG/CR-0040 NUREG/CR-5901 NUREG/CR--5927 Vol.1 NUREG/CR--5964 NUREG/CR-5966 }. NUREG/CR--5978 }i NUREG/CR--6052 mortality NUREG/CR--4214-Rev.2-Pt.1 MPATH NUREG/CR--6021 D-21 NUREG/BR--0083 Vol.9

Keyword NUREG Report Number MSC/NASTRAN NUREG/CR--6021 MSTS NUREG/CR--5998 multiphase flow NUREG/GR--0009 NASTRAN NUREG/CR--6056 natural convection NUREG/IA--0124 NUREG/IA--0125 l NCRP model NUREG/CR--4214-Rev.1-Pt.2-Add.2 NEFTRAN NUREG/CR-5927-VoL1 NUREG/CR--6021 neoplasms NUREG/CR--4214-Rev.2-Pt.1 Netherlands NUREG/IA--0091 NUREG/lA--0112 neural networks NUREG/CR--6113 NUREG/GR--0010 neutron fluence NUREG/CR--6071 NEWMIX NUREG/CP-0131 NEWPART NUREG/CR--5305-Vol.2-Pt.2 nondestmetive testing NUREG/GR--0010 NORIA NUREG/CR--6021 NUREG/CR--6026 NPRDS NUREG/CP--0126-VoL1 NUREG/CP--0126-VoL3 NUREG/CP--0134 NUREG/CR-5471 NUREG/CR--5851 nuclear data collections NUREG/CR--6071 nuclear facilities NUREG-0713 Vol.12 NUREG-0713-Vol.13 NUREG/CP--0130-VoL1 NUREG/CP-0130-Vol.2 nuclear power plants NUREG--0713-Vol.12 NUREG--0713-Vol.13 NUREG--0713-Vol.14 l NUREG-1473 NUREG/BR--0083, Vol.9 D-22

Keyword NUREG Report Number nuclear power plants (continued) NUREG/CP--0130-Vol.1 NUREG/CP--0130-VoL2 NUREG/CP--0131 i NUREG/CP--0134 NUREG/CR--3469-VoL7 NUREG/CR--4214-Rev.1-Pt.2-Add.2 NUREG/CR--4214-Rev.2-Pt.1 NUREG/CR-5247-VoL2 NUREG/CR-5471 NUREG/CR-5801 NUREG/CR-5851 NUREG/CR--5957 NUREG/CR-5976 NUREG/CR-5993 VoL1 NUREG/CR--5993-VoL2 NUREG/CR--6018 NUREG/CR--6052 NUREG/CR--6059 NUREG/CR--6060 NUREG/CR--6082 NUREG/CR--6083 NUREG/CR--6090 NUREG/CR--6101 NUREG/IA--0128 l NUDOCS/AD NUREG/CP--0126-VoL3 NUMARC NUREG/CP--0126-Vol.1 numerical solution NUREG/CR--5957 OCA-P NUREG/CP-0131 NUREG/CR--4219-VoL9-No.2 NUREG/CR-5782 occupational exposure NUREG-0713-Vol.12 NUREG--0713 Vol.13 NUREG--0713 Vol.14 NUREG/CR-3469-Vol.7 NUREG/CR -6050 OCM3D NUREG/CR--6026 i Oconee-1 reactor NUREG/CR-5937 Oconee-2 reactor NUREG/CR-5937 Oconee-3 reactor NUREG/CR--5937 ODRPACK2.01 NUREG/CP--0126 Vol.3 on-line measurement systems NUREG/CR-6083 I D-23 NUREG/BR--0083, Vol.9

l Keyword NUREG Report Number OPEC-2 NUREG/CR-5471 operation NUREG-1473 ORAM-TIP NUREG/CP--0126-VoL1 orifices NUREG/CR--6072 ORIGEN NUREG/CR--4214-Rev.1-Pt.2-Add.2 ORIGEN2 NUREG/CP -0126-VoL2 NUREG/CP--0126-Vol.3 ORMGEN NUREG/CP--0131 ORVIRT NUREG/CP--0131 outages NUREG/IA--0123 packaging NUREG/CR-5817 VoL3-No.2 PAGAN NUREG/CR-5927-VoL1 parallelprocessing NUREG/CR--6113 particle size NUREG/GR--0006 particulates NUREG/CR--5247-VoL2 PARTITION NUREG/CR-5305-VoL2-Pt.1 NUREG/CR-5305-Vol.2-Pt.2 NUREG/CR--6056 PAT 2SR5 NUREG/CP-0126-Vol.2 PATRAN NUREG/CP--0126-Vol.2 NUREG/CP--0131 NUREG/CR--4219-VoL9-No.2 Peach Bottom-2 reactor NUREG/CR--5942 NUREG/CR-5976 Peach Bottom-3 reactor NUREG/CR--5942 PECLOX NUREG/CP -0126-VoL2 performance NUREG/CP--0130-Vol.1 NUREG/CP--0130-Vol.2 NUREG/CR-5817-VoL3-No.2 NUREG/CR-5851 NUREG/CR-5927-VoL1 NUREG/CR-5993-Vol.2 NUREG/CR--6052 NUREG/1A--0100 NUREG/BR--0083, Vol.9 D-24

Keyword NUREG Report Number performance testing NUREG/CR--6041 NUREG/CR--6082 NUREG/CR--6083 personnel NUREG--0713-Vol.12 NUREG--0713-Vol.13 NUREG/CR--3469-Vol.7 personnel monitoring hTREG/CR--6050 I. PetriNets NUREG/CR--6101 L-PETROS NUREG/CR--6021 NUREG/CR--6026 i phase studies NUREG/CR--6032 phase transformations NUREG/CR--6032 PHREEQE hVREG/CR--6021 PIPA NUREG/CP-0126-Vol.1 pipes NUREG/CP--0131 NUREG/CR--5937 NUREG/CR--6049 PLC NUREG/CR--6090 plumes NUREG/CR-5247-VoL1-Rev.1 PM-ALPHA NUREG/CP--0126-Vol.2 ponds NUREG/CR-5978 PORFLOW NUREG/CR-5991 NUREG/CR--6021 NUREG/CR--6026 NUREG/CR--6028 porous materials NUREG/CR-5817.Vol.3-No.2 NUREG/CR-5991 NUREG/CR-5998 NUREG/CR--6026 power distribution systems NUREG-1473 powerlosses NUREG/1A--0092 power supplies NUREG/CR-5993-Vol.2 PR-EDB NUREG/CP--0126-Vol.3 PRA NUREG/CR--5976 D-25 NUREG/BR--0083, Vol.9 =

Keyword NUREG Report Number PRAISE NUREG/CP--0126-Vol.3 PRAMIC NUREG/CR-6056 PRAMIS NUREG/CR-5305-VoL2-Pt.1 Precursor NUREG/CR--5936 prediction equations NUREG/IA-0116 pressure dependence NUREG/CP--0131 pressure measurement NUREG/CR-5851 pressure vessels NUREG/CP-0131 NUREG/CR-4219-Vol.9-No.2 NUREG/CR-5782 NUREG/CR--5957 NUREG/CR--6071 pressurization NUREG/CR-5782 pressurizers NUREG/IA--0105 NUREG/IA--0121 NUREG/IA-0124 primary coolant circuits NUREG/CP--0126-Vol.3 NUREG/IA--0116 NUREG/IA--0121 probabilistic estimation NUREG/CR--5993-Vol.1 NUREG/CR-5993-Vol.2 NUREG/CR--6052 probability NUREG/CR-5782 NUREG/CR--5801 NUREG/CR-5964 NUREG/CR--6056 probes NUREG/GR--0006 proceedings NUREG/CP--0040 NUREG/CP--0130-Vol.1 NUREG/CP--0130-Vol.2 NUREG/CP--0131 programming NUREG/CR--6018 NUREG/CR--6090 - NUREG/IA--0116 Programminglanguages NUREG/CR--6113 i NUREG/BR--0083, Vol.9 D-26

l l Keyword NUREG Report Number progress report NUREG--0713-Vol.12 NUREG--0713-Vol.13 NUREG/CR--4219-Vol.9-No.2 NUREG/CR-5817-Vol.3-No.2 NUREG/GR--0006 PRUEP NUREG/CP--0126-VoL2 NUREG/CR-5305-Vol.2-Pt.1 NUREG/CR--5305-Vol.2-Pt.2 PSAFE2 NUREG/CR--6049 PSTEVNT NUREG/CR-5305-Vol.2-Pt.1 pumps NUREG/CR-5822 PWR type reactors NUREG/CP--0126-Vol.1 NUREG/CP--0126-Vol.2 NUREG/CP--0126-Vol.3 NUREG/CR-5360 NUREG/CR--5818 NUREG/CR-5822 NUREG/CR--5936 NUREG/CR-5937 NUREG/CR-5966 NUREG/CR--6035 NUREG/CR--6054 NUREG/CR--6056 NUREG/CR--6061 NtfitEGAA--UU92 NUREG/IA--0095 NUREGAA--0096 NUREGAA--0099 NUREGAA--0100 NUREGAA--0103 NUREGAA-01N NUREGAA--0105 NUREG/IA--0106 NUREG/IA--0112 NUREGAA--0113 NUREG/IA--0116 NUREG/IA--0118 NUREG/IA--0125 NUREGAA--0126 NUREGAA--0127 NUREGAA--0128 quenching NUREG/CR-5907 radiation accidents NUREG/CR-5247-Vol.1 Rev.1 NUREG/CR--5247-Vol.2 i D-27 NUREG/BR--0083, Vol.9 f = I

l Keyword NUREG Report Number radiation dose distributions NUREG--0713-Vol.12 radiation doses NUREG--0713 Vol.12 NUREG--0713-Vol.13 NUREG--0713-Vol.14 NUREG/CR-3469-Vol.7 NUREG/CR--5247 Vol.1-Rev.1 NUREG/CR--5247 VoL2 NUREG/CR--6041 NUREG/CR--6050 NUREG/CR--6059 radiation effects NUREG/CR--4214-Rev.1-PL2-Add.2 NUREG/CR--5782 radiation hazards NUREG/CR--4214-Rev.1-Pt.2-Add.2 radiation heating NUREG/CR--6026 L radiation monitoring NUREG/CP--0130-Vol.1 NUREG/CP-0130-Vol.2 NUREG/GR--0006 radiation protection NUREG-0713-Vol.14 NUREG/CR-3469-Vol.7 radiation transport NUREG/CP--0130-Vol.1 ( NUREG/CP-0130-VoL2 NUREG/CR-5247-Vol.1-Rev.1 NUREG/CR--5247-VoL2 NUREG/CR--5927-Vol.1 NUREG/CR-5991 NUREG/CR--6059 radioactive effluents NUREG/CP--0130-VoL1 1' NUREG/CP--0130-Vol.2 radioactive waste disposal NUREG/CP--0040 NUREG/CR-5927-VoL1 NUREG/CR-5943 NUREG/CR-5991 NUREG/CR-5998 [. NUREG/CR--6021 NUREG/CR--6026 I NUREG/CR--6028 NUREG/CR--6041 NUREG/CR--6054 NUREG/CR--6070 radioactive waste facilities. NUREG/CR-5927-VoL1 NUREG/CR--6026 NUREG/CR--6041 ( NUREG/BR--0083, Vol.9 D-28 .1 a

l Keyword NUREG Report Number radioactive waste management NUREG/CP--0130-Vol.2 NUREG/CR--5817-VoL3-No.2 NUREG/CR-6021 NUREG/CR--6(M1 radioactive waste storage NUREG/CR--6026 NUREG/CR--6028 radioisotopes NUREG/CP--0130-VoL1 NUREG/CP-0130-VoL2 NUREG/CR--4214-Rev.2-Pt.1 NUREG/CR-5991 radionuclide administration NUREG/CR-4214-Rev.1-Pt.2-Add.2 radionuclide migration NUREG/CP--0(M0 NUREG/CR--5817-VoL3-No.2 NUREG/CR-5927-Vol.1 NUREG/CR--5943 NUREG/CR-5998 NUREG/CR--6021 NUREG/CR--6041 RAESTRICT NUREG/CR-5927-Vol.1 RAFT NUREG/CP--0126-VoL2 RALOC MOD 2.2 NUREG/CP--0126-Vol.2 RAPID NUREG/CP--0126-Vol.1 RASCAL 2.0 NUREG/CR-5247-Vol.1-Rev.1 NUREG/CR--5247-VoL2 reactivity coefficients NUREG/IA--0120 reactor accidents NUREG/CP--0126-VoL1 NUREG/CP--0126-VoL2 NUREG/CP--0126-Vol.3 NUREG/CP-0130-VoL1 NUREG/CP-0130-Vol.2 l NUREG/CR--4214-Rev.2-Pt.1 l NUREG/CR--5247-Vol.1-Rev.1 NUREG/CR-5247-Vol.2 NUREG/CR--5305-VoL2-Pt.1 NUREG/CR-5305-VoL2-Pt.2. NUREG/CR-5360 NUREG/CR-5882 NUREG/CR-5936 ' NUREG/CR-5937 NUREG/CR-5966 NUREG/CR-5978 ? D-29 NUREG/BR--0083. Vol.9 l

Keyword NUREG Report Number reactor accidents (continued) NUREG/CR--5983 NUREG/CR--6032 NUREG/CR--6059 NUREG/CR-6072 NUREG/GR--0009 NUREG/IA--0116 NUREGAA--0122 NUREG/IA-0123 reactor components NUREG/CP--0126-Vol.1 NUREG/CP--0131 NUREG/CP--0134 NUREG/CR--5305-Vol.2-Pt.1 NUREG/CR--5471 NUREG/CR--5801 NUREG/CR-5976 NUREG/CR-5993-Vol.1 NUREG/CR-5993-VoL2 NUREG/CR--6054 reactor control systems NUREG/CP--0126-Vol.1 NUREG/CP--0134 NUREG/IA--0092 NUREG/IA--0109 reactor cooling systems NUREG/CR-5818 NUREG/CR-5822 NUREG/CR--5937 NUREG/CR-5949 NUREG/CR-5983 NUREG/CR-5984 NUREG/CR--6035 NUREG/IA--0091 NUREG/IA--0092 NUREG/IA-0095 NUREG/IA--0099 NUREG/IA--0103 NUREG/IA--0105 NUREG/IA--0109 NUREGAA-0112 NUREGAA--0124 NUREG/IA--0125 NUREG/IA--0128 reactorcore disruption NUREG/CR--5901 NUREG/CR--5901 reactor cores NUREG/IA--0090 - reactorinstrumentation NUREG/CP--0126-Vol.1 }- NUREG/CP--0134 NUREG/CR-5851 i NUREGAA-0126 f NUREG/BR--0083, Vol.9 D-30 1

Keyword NUREG Report Number i reactor kinetics NUREG/CR--6035 reactorlicensing NUREG/CP--0134 NUREG/CR-5818 NUREG/CR--6049 reactor materials NUREG/CP--0131 NUREG/CR--4219-Vol.9-No.2 reactor monitoring systems NUREG/CR-4090 reactor operation NUREG/CR-5936 NUREG/IA--0092 reactor operators NUREG/CR-5818 reactor protection systems NUREG/CR--6090 NUREG/CR--6101 reactor safety NUREG-1473 NUREG/CP -0126-Vol.1 NUREG/CP--0126-Vol.2 NUREG/CP--0126-Vol.3 NUREG/CP -0130-Vol.1 NUREG/CP--0130-Vol.2 NUREG/CR-3469-VoL7 NUREG/CR--4214-Rev.2-Pt.1 NUREG/CR-5305-VoL2-Pt.1 NUREG/CR-5305 VoL2-Pt.2 NUREG/CR--5471 NUREG/CR-5801 - NUREG/CR-5851 NUREG/CR-5882 NUREG/CR--5907 NUREG/CR-5937 NUREG/CR-5942 NUREG/CR-5949 NUREG/CR-5957 NUREG/CR-5964 NUREG/CR-5978 NUREG/CR--5983 NUREG/CR--5993-Vol.1 NUREG/CR-4035 NUREG/CR--6049 NUREG/CR -6056 NUREG/CR--6059 NUREG/CR--6061 NUREG/CR--6101 NUREG/IA--0085 NUREG/IA -0091 NUREG/IA--0092 NUREG/IA--0094 NUREG/IA-0095 p D-31 NUREG/BR--0083, Vol.9 . N

Keyword NUREG Report Number reactor safety (continued) NUREGAA--0096 NUREGAA--0099 NUREG/Lb0100 NUREGAA--0103 NUREGAA--0104 NUREGAA--0105 NUREGRA--0106 NUREGAA--0107 NUREGAA--0112 NUREGAA--0118 NUREGAA-0119 NUREG/IA-0120 NUREG/IA--0121 NUREG/IA--0122 NUREG/IA--0123 NUREGAA--0124 NUREGAA--0125 - NUREGAA--0126 NUREGAA--0127 NUREG/IA--0128 reactor safety experiments NUREG/CR--6061 reactor start-up NUREG/lA--0085 NUREG/IA--0109 reactor vessels NUREG/CR--5937 NUREG/IA--0108 NUREGAA--0116 reactors NUREG/CR--4219-Vol.9-No.2 realtime systems NUREG/CR--6082 NUREG/CR--6083 recommendations NUREG/CR-5471 NUREG/CR-6090 NUREG/IA--0090 regulations NUREG/CR-6050 REIRS NUREG--0713-Vol.12 NUREG--0713 Vol.13 NUREG--0713-Vol.14 RELAP/SCDAP5 NUREG/CP--0126-Vol.2 RELAP5 NUREG/CR--6035 RELAPS/ MODI NUREG/IA--0123 \\ NUREG/BR--0083, Vol.9 D-32 1

Keyword ' NUREG Report Number RELAP5/ MOD 2 NUREG/IA--0085 NUREGAA-0090 NUREGAA--0091 NUREG/IA--0092 NUREGBA--0094 NUREGAA--0095 NUREGAA--0099 NUREG/1A--0106 NUREGAA--0107 NUREGAA-0108 NUREGAA--0109 NUREGAA-0110 NUREGAA--0112 NUREGBA-0116 NUREG/IA--0118 NUREGAA--0119 NUREGSA--0121 NUREG/IA--0123 NUREG/IA--0124 NUREGAA--0125 NUREG/IA--0128 RELAP5/ MOD 3 NUREG/CP-0126-Vol.1 NUREG/CR-5818 NUREGAA-0094 NUREGAA--0095 NUREGAA--0096 NUREGAA--0099 NUREGBA--0100 NUREGAA--0103 NUREGAA-0104 NUREGAA--0105 NUREGAA--0106 NUREG/IA--0113 NUREGSA--0116 NUREGAA--0128 release limits NUREG/CR--4214-Rev.1-PL2-Add.2 reliability NUREG/CP--0126-Vol.1 NUREG/CP--0134 NUREG/CR--5471 NUREG/CR--5801 NUREG/CR-5937 NUREG/CR-5964 NUREG/CR--5993-Vol.1 NUREG/CR-5993 Vol.2 NUREG/CR--6052 NUREG/CR--6101 RELPIN NUREGAA-0013 ? D-33 NUREG/BR--0083, Vol.9 ---e.

Keyword NUREG Report Number REMIT NUREG/CR--6050 REMIX NUREGEP--0131 remote sensing NUREG/CR--5851 reporting requirements NUREG--0713-Vol.13 NUREG/CR--6050 Republic of Korea NUREG/IA--0092 NUREG/IA--0103 i l l research progmms NUREG/CP-0126-Vol.1 NUREG/CP-0126-Vol.2 NUREG/CP--0126-Vol.3 l NUREGER -4219-Vol.9-No.2 NUREG/CR-5817-Vol.3-No.2 RETRAN-3 NUREG/CP--0126-VoL1 rewetting NUREG/IA-0090 risk assessment NUREG/CP--0126-Vol.1 NUREG/CP--0126-Vol.3 NUREG/CR--5305-Vol.2-Pt.1 NUREG/CR-5305-Vol.2-Pt.2 NUREG/CR--5471 NUREG/CR -5801 NUREG/CR-5936 NUREG/CR-5964 NUREG/CR-5976 NUREG/CR--5993-Vol.1 NUREG/CR-5993-Vol.2 NUREG/CR-6056 Robinson-2 reactor NUREG/CR--6071 ROCMAS NUREGK%--6021 rod bundles NUREG/IA-0100 Rousselier model NUREG/CP-0131 RSYGAL NUREG/CP -0126 Vol.2 safety analysis NUREG/CR-5801 NUREG/1A-0106 sampling NUREG/CR--5978 NUREG/GR -0006 SANGRE NUREG/CR--6021 c NUREG/BR -0083, Vol.9 D-34

Keyword NUREG Report Number SAPHIRE NUREG/CR-5964 SARA NUREG/CR-5964 SAS NUREG/CR--5305-Vol.2-Pt.1 saturation NUREG/CR--6028 scale models NUREG/IA--0099 scaling laws NUREG/GR--0009 SCDAP NUREG/GR--0009 SCDAP/RELAP NUREG/CR--6056 SCDAP/RELAP5/ MOD 3 NUREG/CR--5937 NUREG/CR-5949 scram NUREG/CR-5983 NUREG/IA--0091 scrubbing NUREG/CR-5901 NUREG/CR--5966 sedimentation NUREG/CR--5901 seismic effects NUREG/CP-0126-Vol.3 NUREG/CR--5817-Vol.3-No.2 sensitivity analysis NUREG/CR-5937 NUREG/CR--5943 NUREG/IA--0092 NUREG/IA--0125 Sequoyah-1 reactor NUREG/CR-5936 NUREG/CR-5937 Sequoyah-2 reactor NUREG/CR-5936 NUREG/CR--5937 l. I service life NUREG/CR-5851 I simulation NUREG/CR--6026 NUREG/IA--0090 NUREG/IA-0109 NUREG/IA--0120 size NUREG/CR--5970 SOFIRE-M 11 NUREG/CP-0130-VoL2 D-35 NUREG/BR--0083, Vol.9

Keyword NUREG Report Number software NUREG/CP--0134 soils NUREG/CR-5998 SOLA-1rrS NUREG/CP-0131 l somatically significant dose NUREG/CR--4214-Rev.1-PL2-Add.2 sorption NUREG/CR-5817-Vol.3-No.2 soarce terms NUREG/CR-5247 Vol.1-Rev.1 NUREG/CR-5247-VoL2 NUREG/CR-5305-Vol.2-Pt.2 NUREG/CR--5360 NUREG/CR-5927-Vol.1 NUREG/CR-5942 NUREG/CR-5978 NUREG/CR--6N1 Spain NUREG/IA--0085 NUREG/IA--0108 NUREG/1A--0109 specifications NUREGFA--0116 SPEC 1 ROM-32 NUREG/CR--6021 sphericalconfiguration NUREG/CR-5901 spontaneous combustion NUREG/CR--6072 SPRAY-11 NUREG/CP--0130-VoL2 sprays NUREG/CR-5978 SSC NUREG/CP--0126-Vol.1 l ST-DOSE NUREG/CR-5247 Vol.1-Rev.1 ) NUREG/CR--5247-Vol.2 stability NUREG/IA-0096 NUREG/lA-01M NUREG/1A--0107 stacks NUREG/GR--0006 standards NUREG/CP-0130-Vol.1 NUREG/CP-0130-Vol.2 staticloads NUREG/CR--5957 NUREG/CR--6052 NUREG/BR--0083, Vol.9 D-36

' Keyword NUREG Report Number - statisticaldata NUREG--0713-Vol.14 statistical models. hTREG/CR-5966 NUREGER-5993-Vol.1 STCP NUREG/CP--0126-VoL2 NUREG/CR--5942 NUREG/CR--5966 steady-state conditions NUREG/CR--6035 STEALTH NUREG/CR--6021 steam NUREG/IA--0116 NUREG/IA--0125 NUREG/IA--0127 steam generators NUREG/CR--5984 NUREG/CR--6035 NUREG/IA--0095 NUREG/IA--0104 NUREG/IA--0106 NUREG/IA--0110 NUREG/IA--0113 steam turbines NUREG/IA-0108 NUREG/IA--0120 steels NUREG/CR--4219-Vol.9-No.2 NUREG/CR-5782 NUREG/CR-5970 STER NUREG/CR-5305-Vol.2-Pt.1 NUREG/CR-5305-Vol.2-Pt.2 ' strains NUREG/CR-5958 stratification NUREG/IA--0096 STRES3D NUREG/CR-4021 stress analysis NUREG/CR-5782 ' NUREG/CR-5957 NUREG/CR-5970 NUREG/CR--6049 stresses. NUREG/CR-5958 NUREG/CR--6052. Structural Materials Electronic Database NUREG/CP--0126-Vol.3 ' . SUPCRT ' NUREG/CR--6021 D-37 NUREG/BR--0083, Vol.9 L u

Keyword NUREG Report Number SUPERTREE NUREG/CR--6056 SURFIT NUREG/CP--0126-Vol.3 Surry-1 reactor NUREG/CR--5937 NUREG/CR--5949 Surry-2 reactor NUREG/CR-5937 NUREG/CR--5949 - Surry-3 reactor NUREG/CR-5937 NUREG/CR--5949 Surry-4 reactor NUREG/CR-5937 NUREG/CR-5949 surveys NUREG/CR--6018 1 system failure analysis NUREG/CR--5993-Vol.1 systems analysis NUREG/CR--5993-Vol.2 TAC NUREG/CP--0126-Vol.3 TACMVS NUREG/CP--0126 VoL3 task scheduling NUREG/CR -6083 TEMAC NUREG/CR-5305-Vol.2-Pt.1 temperature dependence NUREG/CP--0131 temperature distribution NUREG/CR--6032 temperature measurement NUREG/CR--6061 TEMPEST NUREG/CP--0131 test facilities NUREG/CR.-6061 NUREG/CR-6072 NUREG/IA-0095 NUREG/IA--0099 NUREG/IA--0104 NUREG/IA--0126 NUREG/1A--0127 l testing NUREG/CP -0130-Vol.1 1 NUREG/CP--0130-Vol.2 NUREG/CP--0131 NUREG/CR--4219-Vol.9-No.2 NUREG/CR-5851 NUREG/IA--0085 NUREG/IA--0091 NUREG/BR--0083, Vol.9 D-38 J

Keyword NUREG Report Number testing (continued) NUREG/IA--0103 NUREGAA--0105 NUREG/IA-0118 NUREGAA-0122 NUREG/1A-0127 THAMES NUREG/CR--6021 THATCH NUREG/CR-5983 NUREG/CR-5984 THCC NUREG/CR--6021 theoretical data NUREG/CR-5958 NUREGSA-0123 thermal analysis NUREG/CR--6021 NUREG/CR--6032 NUREG/IA--0100 thermal shock NUREG/CP--0131 NUREG/CR-4219-Vol.9-No.2 NUREG/CR--5782 thermochemicalprocesses NUREG/CR--6021 l thermocouples NUREG/CR--6061 i thermodynamics NUREG/CR-6021 NUREG/CR--6026 NUREG/CR--6032 Tilt Test NUREG/CR-5817-Vol.3-No.2 time dependence NUREG/CR-5993-Vol.1 NUREG/CR--6052 TOUGH NUREG/CP--0040 NUREG/CR--6021 NUREG/CR--6026 TR-EDB NUREG/CP--0126-Vol.3 TRAC NUREG/lA--0126 NUREGAA--0127 TRAC-B NUREGAA--0128 TRAC-BF1 NUREG/CR-5882 NUREGAA--0120 NUREG/1A--0122 TRACG NUREG/CP--0126-Vol.1 D-39 NUREG/BR--0083,Vol.9

Keyword NUREG Report Number TRACR3D NUREG/CR--6021 training NUREG/CR--5247-Vol.2 TRANQL NUREG/CR--6021 transducers NUREG/CR--5851 transients NUREG/CR--5882 NUREG/CR-5949 NUREG/CR--5983 NUREG/CR-5984 NUREG/CR--6035 NUREGAA--0085 NUREGAA--0092 NUREGAA--0095 NUREG/IA--0100 NUREGBA-0105 NUREGAA--0107 NUREG/IA--0108 NUREGSA-0109 q NUREGAA--0110 NUREGAA--0118 NUREGAA--0119 NUREG/IA--0120 NUREG/IA--0121 NUREGSA--0122 NUREGAA--0123 NUREGAA--0124 TRAPFRANCE NUREG/CP--0126-Vol.2 TRAPMELT NUREG/CP--0126-Vol.2 Trillo-1 reactor NUREGAA--0085 Tse and Cruden's equations NUREG/CR-5817-Vol.3-No.2 tubes NUREG/IA--0095 turbines NUREG/IA--0085 turbulence NUREG/CR--6072 turbulent flow NUREG/CR--6060 two-phase flow NUREGER-5817-Vol.3-No.2 NUREG/CR-4026 NUREGAA--0127 UDEC NUREG/CR--6021 1 l-NUREG/BR--0083, Vol.9 D-40 ~k _3

Is 1 U Keyword NUREG Report Number ultrahlgh-speed photography NUREG/CR--6072 underground disposal NUREG/CR-6041 United Kingdom NUREG/IA--0096 uranium oxides NUREG/CR--6032 US NRC NUREG-1473 NUREG/CR--4214 Rev.1-Pt.2-Add.2 NUREG/CR--6050 - ' NUREGAA--0085 NUREGAA--0090 NUREG/IA--0091 NUREGAA-0108 NUREGAA--0109 NUREGAA--0112 USA NUREGAA--0126 validation NUREG/CR--6018 NUREGAA--0094 valves NUREGAA--0104 NUREGAA--0105 NUREGAA-0121 NUREGAA--0122 NUREGAA--0124 VAM2D NUREGER--5927-Vol.1 NUREG/CR-5998 VandellosIIreactor NUREGAA--0107 NUREGAA--0108 NUREGAA-0109 NUREGAA-0110 ventilation systems NUREG/CP--0130-Vol.1 NUREG/CP--0130-Vol.2 . verification NUREG/CR--5943 NUREGAA--0094 VICTORIA - NUREG/CP.-0126-VoL2 VIPRE NUREG/CP--0126-Vol.1 VISA-II - NUREGER-4219-Vol.9-No.2 ' VISCRK NUREG/CP--0131 g void fraction NUREGAA--0113 D-41 NUREG/BR--0083, Vol.9 ?

Keyword NUREG Report Number VS2DT NUREG/CR-5927-Vol.1 VTOUGH NUREG/CR--0040 NUREG/CR--6021 NUREG/CR--6026 VVTVIRT NUREG/CP--0131 WATEQ NUREG/CR--6021 water NUREG/CR-5901 NUREG/CR--5907 NUREG/IA-0116 water chemistry NUREG/CR-5822 water cooled reactors NUREG/CR--5843 NUREG/CR--6032 NUREG/GR-0009 WECHSL NUREG/CP--0126-Vol.2 weldedjoints NUREG/CR-5782 WETCOR-1 NUREG/CR--5907 ] whole-body irradiation NUREG--0713-Vol.12 Wolf Creek 1 reactor NUREG/IA-0106 NUREG/IA--0113 XSOR NUREG/CR--5305-Vol.2-Pt.1 NUREG/CR-5360 NUREG/CR--6056 Yankee Rowe reactor NUREG/CR-5782 Yucca Mountain NUREG/CR--6021 zeolites NUREG/CR-5817-Vol.3-Nol Zion-1 reactor NUREG/CR-5822 Zion-2 reactor NUREG/CR-5822 zirconium oxides NUREG/CR--6032 Zorita-1 reactor NUREG/1A--0124 NUREG/BR--0083, Vol.9 D-42

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