ML20116P164

Computer Codes and Mathematical MODELS.January-December 1991

ML20116P164
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Issue date:	10/31/1992
From:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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References
NUREG-BR-0083, NUREG-BR-0083-V07, NUREG-BR-83, NUREG-BR-83-V7, NUDOCS 9211240373
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United c.'ates Nucis Regulatory Commission u-Computer Codes and Mathematical Models January-December 1991 i

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i AllSTRACT This report contains citations of NUREO-series documents issued in calerxiar year 1991 relating to computer Aoftware for scientinc, engincenng, or technology related prognuns perfonned or sponsored i

by the U.S. Nuclear Regulatory Comtnission (NRC). It is intended as a reference tool to assist the j

scientiGe and techrdcal analyst in obtaining infonnation on NRC coinputer related activities.

l I

iii NUREC/BR MiB3,Vol.7

TAllLE OF CONTENTS Page ADSTRACT ill FOREWORD vil CITATIONS 1

APPENDIX A:Irdex by NUREG Series Report Nurnber A1 APPENDIX B: Irxier by Software Identification is 1 APPENDIX C: Index by Contractor Report Nurnier C1 APPENDIX D:Irxlex by Keyword D-l l.-

v NUREO/BR-0083,Vol.7

FOltEWORD The citations in this document appeat in NUREGeries document order. Citations of NRC staff-generated repons desigrated NUREG--uu are listed first, follswed by any conference pn<ecdmgs identified as NUREO/CP uu, contractor generated trports published as NUREG/CR-nu docu rnents, grant reports pubhshed as NUREO!OR-uu documents, ard the International Agreement reports issued as NUREO/1A nu publications. Each citation contains the following; NUREG series report nurnber; software identifica6on; contractor report number; repon title; a descriptiorn of the r contents; publication date; names of the trxhviduals tesponsible for preparing, cornpihng, or editing t repon; contractor name and location; sponsoring NRC organizauon; and Leywords or descriptors.

Irxhecs by NUR EG series report numler, software klentilication, contractor tepon numter, atd areincludedinthe Appendices.

Specific code names and softw are identification appear in the headmg of those citations w emphasis on specific mathematical inodels, computer codes. or databases *Ihc term " General"is used the heading of those citauons which contain significant infonnation on many mtxiels, computer cod or databases.

1 vii NUREO/BR.-0083, Vol.7

NUREG.1435 Suppl.1 SIMS Database

Title:

Status of Safety issues at License 1 Power Plants: TMI Acdon Plan Requirements, Unresolved Safety issues, Gerenc Safety issues. Supplement 1

==

Description:==

As part of ongoing US Nuclear Regulatory Conunission (NRC) effons to ensure the quahty and acwuntabihty of safety issue infonnation, a program was established whereby an annual NUREO repon would be published on the shtus of licensee implementation ard NRC ven6 cation of safety issues in major NRC seguirements areas. %is infonnation was compiled and reponed in three NUREO volumes. Volume 1, published in March 1991, addressed de status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, pubitsbed in May 1991, a<kiressed the status of unresolved safety issues (USls). Volume 3, published in Jure 1991, addressed the implementation and verincation status of genene safety issues (OS!s). This annual NUREO report combines these volumes in o a single report and provides updated infonnation as of September 30,1991. %e data contained in these NUREO reports are a product cf the NRC's Safety lasues Management System (SIMS) database, which is inaintained by the Project Maragement Staff in the Of6ce of Nuclear Reactor Regulation and by NRC regional personnel. His repon is to provide a comprehensive desenption of the implementation and verification status of TMI Action Plan Requirements, safety issues designated as USIs, arid OSis that have beeri resolved and involve implementation of an action or actions by licensees. 71us report rnales the infonnation available to otter interested parties,incithhng the pubhc. An acLiitional purpose of this NUREO neport is to serve as a follow on to NUREO 0933, *A Prioritization of Generic Safety issues, which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees.

Publication Date:

Ikcember 1091 Prepared by:

NRC Ofnce of Nuclear Reactor Regulation Prepared for:

NRC Of6cc of Nuclear Reactor Regulation Keywords:

compiled data, database managemem, evaluation, information dissemination, nuclear power plants, planning, program management, reactor safety, regulations, SIMS database, %ree Mile Island 2 reactor 1

NUREO/BR- 0083,VoL7 l

NUlt EG/CP.-0114.Vol.1 4

TitAC,ItEt AP5, PlCASSO, CAltES

Title:

Eighteenth Water Reactor Safety information Mee ing, hoceedings: %me 1

==

Description:==

Separate abstracts have teen prepared for the papers presented at the meeting which was held October 22-24,1990, in Rockville, Maryland. The 40 papers in Ltds volutne cover the following aspects of light water reactor s aiety: human factors research; orgardrational factors and reliabihty assessment; radioactive waste snanagement research; earth sciences, specifically, seismicity studies in North America; reactos thertaal hydraulics; containment system testiag and structural engirecting; and seismic engineering.

Publication Date:

Apnl 1991 P4 epared by:

Weiss, A1 (compM Bnothaven National Lab., Upton, NY)

Prepared for:

NRC Oflice of Nuclear Regulatory Research Keywords:

aging, BWR type reactors, California CARES, computer codes, computerized simulation, containment systems, canhquakes, engineered safety systems, geologic formations, heat transfer, higti-level radmactive wastes, human factors, hydraulics, low level radioactive wastes, meetings, PICASSO, pressure vessels, primary coolant circuits, probabilistic estiination, PWR type reactors, radioactive waste managemern, reactor components, reactor safety, reactor ctabihty, regulations, RELAPS, reliability, research programs, risk susessment, seismicity, service hie, soil mechanics, TRAC, US NRC, USA u

NUREG/DR-0083,Vcl.7 2

NUltEGICib0ll4Wol.2 MELCOlt, VICTOltl A

Title:

Eighteenth Water Reactor Safety Infonnatum Meeting. Volume 2, Severe Accident Research Accident Management, Probabihstic Risk Assessm:nt Topics, hxhvidual Plant Examination Program ard Other 1ssues Descripilon:

Bus three volume repon contains 100 papeis out of the 128 that were presented at the Eighteenth Water Reactor Safety Infonnation Meeting held in Rockville, Marylard, dunng the week of Odoter 22-24,1990. ' Hie papers are pnnted in the order of their presentation in each session and descrite progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign panicipation in the meeting included 16 different papers presented by researchers from Derunark, Egypt, Gennany, IAEA, Itiy, Japan, Norway, Taiwan, UK, and USSR. ' Die irvividual papers have been cataloged separately. *ntis document Volume 2, addresses these topics: severe accident research, accident managament, probabilistic risk assessment, and imhvidual plant examination i

programs.

l'ublication Date:

April 1991 l'repared by:

Weiss, AJ. [ comp.)(Bnokhaven Natiortal Lab., Upton, NY)

I'repered for:

NRC Office of Nuclear Regulatory Research Keywords; sging, BWR type reactors, contai:. ment, corium, emergency plans, Europe, fission-product tricase, heat transfer, human factors, hydraulics, hydregen, inspection, Japan, meetings, MELCOR, nuclear power plants, PWR type reactors, radioactive waste management, reactor accxients, reactor components, reactor safety, reliability, reviews, nsk assessment, source terms, systems analysis, USA, valves, VICTORIA 3

NUREG/BR.-0083 Vol.7 I

.NUltEG/CI:. 0116-Vol.1 MINUIT

Title:

Proceedings of tic 21st DOE /NRC Nuclear Air Cleaning Conference. Sessions 14

==

Description:==

Separate abstracts have been prepared for the papers presented at the rnecting on nuclear facihty air clearung technology in the following specific areas of interest: air cleamng teduiologies for the inanagement and disposal of radioactive wastes; Canadian waste managernent program; iadiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off gas cleaning; incineration ard vitrification; atorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; rad nuclear air cleaning programs around the world.

Publication Date:

February IW1 Prepared by:

Fttst, M.W. [ed.) Olarvard Univ., Boston, MA. liarvard Air Cleaning Lab.)

Prepared for:

NRC Of fice of Nuclear Regulatory Research; IX)E Office of Nuclear Safety Keywords!

activated carbon, adsorption, aerosols, air cleaning systems, air filters, alpha detection, beta detection, BWR type reactors, containment systems, control systems, detection, emergency plans, Federal Republic of Germany, fires, France, fuel fatrication plants, fuel reprocessing plants, humidity, Idaho chemical pnicessing plant, impregnation, indnerators, iodine 129, rnathematical mo'lels, meetings, mercury, MINUIT. retrogen oxides, twJes, nucleat power plants, off gas systems, ORNL, penetrometers, PWR type reactors, radiation har.ards, radiation protection, radioactiv-waste facilities, radioactive waste management, raAcactive wastes, reactor safety, real time systems, recommendations, reproecssing, l

ruthenium, safety, safety standards, sampling, Sweden, SYVAC geosphere model, see facilities, US DOE, US NRC, ventilation systems NUREG/BR 0083,Vol.7 4

1 1

NUREG/CP--0116 Vol.2 -

SIMEVENT, CAIRE -

Title:

Proceedings of the 21st DOE /NRC Nuclear Air Cleaning Conference. Volume 2, Sessions 9-16 Ducription:

The 21st meeting of the Depanment of Energy / Nuclear Regulatory Commission (DOFJ NRC' Nuclear Air Cleaning Conference was held ia San Diego, California, on August 13-16,1990. *Ihe proceedmgs have been published as a two volume set. Volume 2 contains sessions coveri u' adsortents, nuclear codes and standards, modelling, filters, safety, containment venug, and a review of nuclear air cleaning propams around the world. Also included is the list of atterxlees and an index of authors and rpeakers.

Ptiblication Date:

Febmary 1991 Prepared by:

First, M.W. [ed.] (Harvard Univ., Boston, MA. Harvard Air Ocar' g 1.ab.)

Prepared for:

NRC Office of Nuclear Regulatory Research: DOE Office of Nuclear Safety Keywords:

activated carbon, adsorbents, aeros324, air cleaning systems, air. filters, CAIRE, computerized simulation, containment builsngs, dusts, ecological concentration, Gaussian plume model, ground release, international cooperation, magnetic filters, mathematical models, meetings, monitoring, nuclear facilities, pollution regulations, pressure selease, reconunendatims, regulations, safety, SIMEVENT, ventilation s>Trems 5

NUREG/BR--0083, Vol.7

l

- NUREG/CP--0118 MELCOR, CONTAZN, FRAPCON 1, RELAPS, FRAP-T6

Title:

Tansactions of the Nineteenth Water Reactor Safety Infonnation Meeting

==

Description:==

This repon contains summaries of papers on reactor safety tesearch to be presented at the 19th Water Reactor Safety Infonnahon Meeting at tir Betlesda Marriott Hotelin Betlesda, Maryland, October 28-30,1991. The summaries briefly describe the proFrams aid results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research,-

US NRC. Summaries of invited papers concerning nuclear safety issues from US govemment laboratorica, the electnc utihties, tie Electric Power Research lastitute (EPRI),

tle nuclear indw try, and from the govemments and mdustry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information r ' ange during the course of tle meeting and are given in the order of their present'

. in each session. The individual summaries have been cataloged separately.

Publication Date:

Octoter 1991 Prepared by:

Weiss, A.J. [ comp.) (Brookhaven Wational Lab., Upton, NY)

Prepared for:

NRC Office of Nuclear Regulatory Research Keywords:

after-heat removal, aging, annealios, BWR type reactors, concretes, CONTAIN, contaitunent, control elements, control rod drives, corium, crack propagation, cracks, canhquakes, electric cables, embrittlement, explosions, fatigue, fission-product release, fluid flow, FRAP-T6, FRAPCON-1, fuel elements, fuel rods, fueldoolant interactions, heat transfer, human factors, hydraulics, hydrogen, in-service inspection, irradiation, management, MELCOR, nondestructive analysis, pipes, pressure vessels, pumps, PWR type reactors, radioactive effluents, reactor accidentc, reactor components, reactor control systems, reactor cooling systems, reactor materials, reactor opert ors, reactor protection systems, reactor safety, regulations, RELAP5, reliability, risk assessment, seismic effects, seismicity, source terms, stainless steels, steels, thermal degradation, trermal shock, transformers, turbines, valves, welded joints NUREO/BR--0083, Vol.7 6

SAND--86 0196-Vol.2 NUREG/CR--3964 Vol.2 General

Title:

Techniques for Determining Probabilities of Events and Processes Affecting the Performance of Geologic Repositories. Volume 2 Suggestx! Approaches

==

Description:==

The US Environmental Protection Agency has established a standard for the performance of geologic repositories for the disposal of radioactive waste. This standard is probabilistic in nature, but the methods for determining probabilities of events and prmesses ofinterest in implementing such a standard am still being &vekped Decision Ticory, which involves Bayesian probability tediniques, can serve as a framework for estimating the probability of occurrence of processes and events that are likely to dicrupt a geologic repository, This report presents the mathematical basis for such a methodology and demonstrates ao application of it in tir.ee areas: climate change, rectonic events, and human intmsion.125 refs.,27 figs.,11 tabs.

Publication Date:

June 1991 Prepared by:

Apostolakis, G. (Califonna Univ., Los Angeles, CA); Bras, R. (Bras Consulting Engineers, Lexington, MA); Price, L (Sarxha National Labs., Albuquerque, NM); Valdes, J. (Texas A&M Univ., ColleFe Statioru TX); Wahi, K. (GRAM, Inc., Albuquerque, NM); Webb, E.

(Witconsin Univ., Madison, WI)

Prepared for:

NRC Division ofIligh-Levt.1 Waste Management, Office of Nuclear Material Safety and Safeguards Keywords:

alpha-bearing wastes, Bayes' theorem, climates, Fourier analysis, high-level radioactive -

wastes, human intrusion, mathematical models, probabilistic estimation, probability,.

radioactive waste disposal, radionuclide migration, risk assessment, spent fuels, tecttmics, time series analysis, underground disposal, US EPA r

i l

l 7

NUREG/BR--0083, Vol.7

NUREG/CR--4063 EGG--2636 TRAC-PF1/ MODI

Title:

An Investigation of Core Liquid Level Depressian in Small Break Loss-of-Coolant -

Accidents

==

Description:==

Core liquid level depression can result in panial core dryout and heatup early in a small break los of coolant accident (SULOCA)innsient. Such behavior occurs when Leam,.

trapped in the upper regions of the reactor primary system (between the loop seal and the.

cort.

uy), moves coolant out of the core region and uncovers the rod upper elevations.

The i mcsult is core liquid level depression. Core liquid level depression and rubsequent core heatups are investigated using subscale data from the ROS A IV Program's bscale Large Scane Test Facihty (LSTF) and the So-scale semiscale facility. Both facilities are -

Westinghouse-type, four loop, pressurized water scactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PFl/ MODI (Version 12.7) thennal-hydraulic code, are also described and summanzed. Finally, the res;xuse of a typical Westinghouse four-loop plant (RES AR-3S) was calculated to qualitatively study coai liquid level depression in a full-scale system. 31 reft,37 figs.,6 tabs.

Publication Date:

August 1991 Prepared by:

Schultz, R.R : Watkins, J.C. (EG&G Idaho, Inc., Idaho Fa'Is, ID); Motley. F.E.; Stumpf,.

11. (los Alamos National Lab., NM); Chen, Y.S. (Nudear Regulatory Commission, Washington. DC. Div. of Systems Research)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

computerized simulation, heat transfer, hydraulics, levels, liquids, loss of coolant, PWR type reactors, reactor cores, reactor safety, scale models, steam, TRAC-PFI/ MOD 1 NUREG/BR--0383, Vol.7 8

. LM F--132-Rev.1 -Pt.2-Add.1 NUREG/CR -4214 Rev.1-Pt.2-Add.1 General

Title:

Health Effects Models for Nuclear Power Plant Accident Consequence Analysis:

Mochfications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation. Low LET Radiatiorn Part 2. Scientific Bases for Health Effects Models: Revision 1, A(Llendum i

==

Description:==

'Ihe Nuclear Regulatory Commission has sponsored several studies to identify and quantify -

the potential health effects of accidental releases of radionuclides from nuclear power plants. The most recent health effects models resulting from these efforts were published in two reports, NUREG/CR-4214, Rev.1, Part I (1990) and Part 2 (1989). Several major heahh effects reports have been published recently that may impact tic heahh effects.

models presented in these repons. Tids addendum to the Part 2 (1989) report provides a review of the 1986 and 1988 reports by the United Nations Scientific Committee on the Effects of Atomic Radiation, the National Academy of Sciences / National Restarch Council BEAR 5 Committee report, and Publication 60 of the International Commission on Radiologic <l Protection as they relate to this report. The three r.;in sections of this adderxium discuss early occurnng and continuing effects, late somatic chects, and genetic effects. The major changes to the NUREG/CR-4214 health effects models recommended in this addendum are for late somatic effects. These changes reflect recent changes in cancer risk factots that have come from longer followup and mvised dosimetry in major studies like that on the Japanese A-bomb survivors. 'Ihe resuhs presented in this acklendum should be used with the basic NUREG/CR-4214 reports listed atrve to obtain the most recent views on the potential health effects of radmnuci, *, reler j a,;cidentally from -

nuclear power plants. 48 refs.,4 figs.,24 tabs.

Publication Date:

August 1991 Prepared by:

Abrahamson, S. (Wisconsin Univ., Madison, WI); Bender, M.A. (Brookhaven National Lab., Upton, NY); Boecker, B.B.; Scott, B.R. (Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM. nhalation Toxicology Research Inst.); Gilbert, E.S.

(Pacific Northwest Lab., Richland, WA)

Prepared for:

NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research Keywords:

A bomb survivors, acute exposure, age dependence, biological radiation effects, carcinogenesis, carcinomas, chronic exposarc delayed radiation effects, dose rates, dosimetry, early radiation effects, epitheliomas, gastrointestinal tract, genetic radiation -

effects, genetically significant dose, health hazards, lonizing radiations, LET, leukemia, mathematical models, melanomas, mortality, neoplasms, nuclear power plants, osteosarcomas, probabilistic estimation, radiation accidents, radiation doses, radiation hazards, risk assessment, sex dependence, somatically significa n dose, th>Toid l

l 9

NUREG/BR-0083,Vol.7

l l

NUREG/CR.-4269 NISTIR--4405 General

Title:

Models of Ttansport Processes in Concrete

==

Description:==

An approach being considered by the US Nuclear Regulatory Commission for disposal of low level radioactive waste is to place the waste forms in concrete vaults buried underground. The vaults would need a service hfe of 500 years. Approaches for pred cting the senice life of concrete of such vaults include the use of mathematical models.

Mathema6 cal models are presented in this report for the major degradation pocesses anticipated for the concrete vaults, which are corrosion of steel reinforcement, sulfate attack, acid attack, and leactung. De models mathematically represent rate controlling processes including diffusion, convection, and reaction and sorption of chemical species, hese models can form the basis fo: predicting the life of concrete under in-senice conditions. 33 refs.,6 figs.,7 tabs.

Publication Date:

January 1991 Prepared by:

Pommersheim, J.ht. (Bucknell Univ.,12wisterg, PA); Clifton, J.R. (National last.'of Standards and Technology, Gaittersburg, hiD)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

adsorption, alkali metal compounds, calcium hydroxides, calcium silicates, cements, chemical reaction kinetics, chlondes, concretes, convection. differential equations, d ffusion, gypsum, leaching, low-level radioactive wastes, mathematical models, porosity, radioactive waste disposal, radionuclide migration, sulfates, Ttdele modulus NUREG/BR--0083,Vol.7 10 l

1

y++

NUREG/CR -4295 General

Title:

Bond Strength of Cementitious Borehole Plugs in Welded Tuff

==

Description:==

Axial loads on plugs or seals in an underground repository due to gas, water pressures, and temperature changes induced subsequent to waste and plug emplacement lead to.

shear stresses at the plug / rock contact. 'Iherefore, the tend between the plug and rock is a -

critical element for the design and effectiveness of plugs in boreholes, shafts, or tunnels.

"his s,udy includes a systematic investigation of the tad strength of cementitious torehole plugs in welded tuff. Analytical and numerical analysis of borehole plug-rock stress transfer mecharues is perfonned. The interface stiength and deformation are studied as a function of Young's modulus ntio of plug and rock, plug length and rock cylinder outside-to-inside radius ratio. The tensile stresses in and near an axially leaded plug are analyzed. The frictional interface strength of an axially loaded borehole plug, the effect of axial stress and lateral extemal stress, and thennal effects are also analyzed. Irnplications for plug design are discussed. The main conclusion is a strong recommendation to design friction plugs in shafts, drifts, tunnels, or boreholes with a minimum length to diameter ratio of four. Such a geometrical design will reduce tensile stresses in the plug and in the host rock to a level which should minimize the risk of long term deterioration caused by excessive tensile stresses. Push-out tests have been used to determine the imd strength.

by applying an axial load to cement plugs emplaced in boreholes in welded tuff cylinders.

A total of 130 push out tests have teen performed as a function of borehole size, plug length, temperature, and degree of saturation of the host tuff. The use of four different borehole radti enables evaluation of size effects. I19 refs.,42 figs.,20 tabs.

Publication Date:

February 1991 Prepared by:

Akgun, H.: Daemen, J.J.K. (Arizona Univ., Tucson, AZ. Dept. of Mining and Geological Engineering)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

boreholes, cements, cracks, deformation, design, dimensions, friction factor, gases, interfaces, length, mathematical models, mechanical properties, mechanical tests, mine shafts, permeability, plugging, radioactive waste facilities, reservoir rock, saturation, sealing materials, seals, shear, size, stress analysis, stresses, temperature dependence, temperature gradients, tensile properties, tuff, tunnels, underground disposal, urxicrground storage,_

water, Young's modulus i

I I

[

11 NUREG/BR-0083, Vol.7

NUREG/CR--4599-Vol.1-No.1 BMI 2173-Vol.1 No.1 PIFRAC, NRCPIPE Tille:

Shon Cracks i i Piping and Piping Welds. Semiannual Repon, March-P,eptemter 1990:

Volume 1, No.1

==

Description:==

nis is tie nrst semiannual repon of the US Nuclear Regulatory Commission's Short Cracks in Piping and Pipmg Welds research program. The program tegan in March 1990 and will extend for 4 years. The intent of this program is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-tefore break analyses os in-service flaw evaluations. Only quasi-static loading rates are evaluated since the NRC's Intemational Piping Integrity Research -

Group (IPIRG) program is evaluating the effects of seismic loadmg rates on cracked piping systems. Additional efforts involve investigating phenomena discovered during the course of conducting the Degraded Piping program. nese include the evaluation of the occurrence of unstable crack jureps in ferritic steels at light-water reactor temperatures and the occurrence of anisotropic fracture properties causing helical crack growth. Both of these phenomena may affect the safety margins implicit in I:ak-before-bitak (LBB) analyses. Other investigations deal with the tracture behavior of bimetallic welds and improvements in crack opening area analyses used in 1.BB. Since much of the work in this program was just teginning during this Grst reporting period and progress is limited, a complete statement of work for the whole program is provided in this report,42 reis.,

14 Ggs., i1 tabs.

Publication Date:

' day 1991 Prepared by:

Wilkowski, G.M.: Ahmad, J.; Brust, F.; Ghadiali, N.; Krishnaswamy, P.; Landow, M.;

Marschall, C.W.; Scott, P.; Vieth, P. (Batteile, Columbus, OH)

Prepared for:

NRC Division of Engineenng, OfGee of Nuclear Regulatory Research Keywords:

BWR type reactors, crack propagation, cracks, fracture mechanics, mechanical propenies, NRCPIPE, PIFRAC, pipes, progress report, PWR type tract ors, reactor materials, reactor safety, research programs, seismic effects, steels, welded joints NUREG/BR-0083,Vol.7 12

a NUREG/CR--4735.Vol.7 PROTOCOL Ti!!e:

Evaluation wi Compilation of DOE Waste Package Test Data.- Biannual Repon, _

F6ary 1989-Jui.,1989: Volume 7 '

==

Description:==

This repon summarizes cvaluations by the Nation 4 lastitute of Starxiards and Technology (NIST) of Department, f Energy (DOE) activiti:s on waste packages designed for 1

I contamment of radioacth e high-level nuclear waste (HLW) for the six-month period, February through July IH9. 7his includes reviews.of related materials research and 1

plans; information on the Yucca Mountain Nevada, disposal site activities; and other information regardmg sup} orting research and special assistance. Outlines for planned interpretative repons on the topics of aqueous corrosion of copper, mechanisms of stress corrosion crackings and inte nal failure modes of Zircaloy cladding are included. For tie publications reviewed durit t this reporting period, r 2 ort discussions are given to 5

supplement the completed rev.'ws and evaluations. Included in this repon is an overall review of a 1984 report on glass-caching mechanisms, as well as reviews for each of the -

seven chapters of this report.

Publication Date:

December 1991 Prepared by:

Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC. Division of High-Level Waste Management); Fraker, A.C.; Escalante, E. (National Inst. of Standards and Technology, Gaithersburg, MD Metallurgy Div.)

Prepared for:

NRC Division of High-level Waste Management, Office of Nuclear Material Safety and

- Safeguards Keywords:

containers, containment, copper, cracks, demonstration programs, dissolution, environmental effects, fuel cans, glass, high level radioactive wastes, leaching, progress report PROTOCOL, radiation effects, radioactive waste disposal, radionuclide migration, research programs, stress corrosion, Yucca Mountain, zircaloy 13 NUREG/BR-0083,Vol.7

NUREG/CR--4757 PNL-7597 General

Title:

Line-loss Detennination for Air Sampler Systems

==

Description:==

Indme deposition can potentially bias the resuhs of rad oioene air sampling systems. To develop guidance and acceptance criteria foi determinations ofline-loss correction factors.

the data on laboratory sampler simulations, field tests on samplers, and experimentally --

measured iodme deposition rates were reviewed. Sampling system design features and operating con &tions at several power reactors are discussed. Measurements of iodine deposition tates on various air sampler construction materials were reviewed, and predicted air sampler perfomiance based on the data was presented. Three examples of field tests of air sampler performance for rasoiodite were examined. A model of iodine disposition arx! resuspension was extensively reviewed, and suggestions were made for incorporating variable resuspension rates. ' Dure principal methods for determining radiciodine line-loss factors were defined arxi compared: in-place field tests, laboratory mock-up with modelled extrapolations to various release rate modes, and modelling based on lateratory data on similar materials. Guidelines for applying these methods were given. Research was recommended to detennine whether the three methods were comparable so the less-expensive method could be substituted for the preferred field tests.12 refs.,10 figs.,20 tabs.

Publication Date:

February 1991 Prepared by:

Glissmeyer, J.A.: Schmel, G.A. (Pacific Nonhwest Lab., Richland, WA)

Prepared for:

NRC Division of Radiation Protection and Ernergency Preparedness, Office of Nuclear Reactor Regulation Keywords:

aerosols, air samplers, BWR type reactors, contairunent systems, deposition, differential equations, experimental data, iodine, mathematical models, nuclear power plants, panicle resuspension, PWR type reactors, radiation transpon, radioactive effluents, radioactivity transport, ventilation systems NUREG/BR-0083,Vol.7 14

ORNL/TM--10328.Rev.1 -

NUREG/CR ;4816-Rev.I' PRiEDB, dB ASE_3 Plus -

Title:

PR EDB: power Reactor Embrittlement Data Base, Version 1: Program Description

==

Description:==

Data concerning radiation embrittlement of pressure vessel steels in U.S. commercial -

i power reactors have been ' collected from available surveillance repons. The purpose of.

this NRC sponsored program is to provide the._ technical bases for voluntary _ consensus standards, regulatory guides, standard review plata, and codes. The data can'also te used ~

for the explorauoa and verification of embuctlement prediction models. The data files are iF ven in dBASE 3 Plus format and can te accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data.,

including curve fitting and plotting facilhies. This seitware has drastically reduced tie time and effort for data processing and evaluation compared to previous databases. The current version of the Power Reactor Embrittlement Data Base (PR EDB) lists the testi results of 117 base niaterials (plates and forgings). 85 welds, and 88 heat affected-zone materials that were irradiated in 241 capsules of 82 reactors. Many capsules also contained _

correlation materials (standard reference materials, SRMs) from the ASTM plate and two HSST plates (01 and 02). Malenal from the Humboldt Bay reactor was used as an SRM ;

~- for some General Elecuic reactors. The Electric Power Research Inst tute (EPRI), reactor, i

vendors, and utilities have provided back up quality assurance checks of tlw PR EDB.

I3 figs.,37 tabs.

Publication Date:

July 1991

)

Prepared by:

Stallmann, F.W.; Kam, F.B.K,: Taylor,' BJ. (Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research' t

Keywords:-

BWR type reactors, compiled data, data acquisition, data processing, database management -

dBASE 3 Plus, embnttlement, information systems, PR EDB, pressure ' vessels, PWR _-

type reactors, quality assurance, aliation effects, steels L

15 NUREG/BR-0083.Vol.7.

l NUREG/CR--5128 BMi--2164 SQUIRT

Title:

Evaluation and Refmement of Leak Rate Estimation Models

==

Description:==

Leak-rate estimation models are important elements in developing a leak-before break methodology in piping integrity and safety analyses. Existing thermal--hydraulic and crack-operting-area models used in current leak rate estimations have bxn incorporated into a single computer code for leak-rate estimation. The code is called SQUIRT, which stands for Seepage Quantification of Upsets In Reactor Tubes. The SQUIRT proFram has been vahdated by companng its thermal-hydraube predictions with the hmited experimental data that have been published on two-phase flow through shts and cracks and by comparing i's crack-opening-area predictwns with data from the Degraded Piping Program. In adchtion leak rate experiments were conducted to obtain validation data for a circumferenti I fatigue crack in a carbon steel pipe girth weld. 56 refs.,30 figs.,4 tabs.

Publication Date:

April 1991 Prepared by:

Paul, D.D.: Ahmad. J., Scott, P.M.; Flanigan, LF.; Wilkowski, G.M. (Battelle Columbus Labs., OH)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research a

Keywords:

BWR type reactors, carbon steels, computer codes, computerized simulation, cracks, fatigue, fracture mechanics, heat transfer, hydraulics, leaks, loss of coolant, pipes, primary coolant circuits, PWR type reactors, reactol safety, ruptures, SQUIRT, two-phase flow, yalidation, weldedjoints NUREO/BR--0083,Vol.7 16 l

1

m

! HNL-NUREG--52181 NUREG/CR--5282-CONTAIN, CONTAIN-DCH;

Title:

Estimation t f Containment Pressure Loading Due to Direct Containment Heating for the -

ZmaPlant-s

==

Description:==

For some core meltdown accident sequences in light water reactors (LWRs) it is possible

. for the primary system to remain at high pressure. Under these circumstances, as thel

- molteu core debris penetrates the reactor vessel, the core debris would be ejected under i high pressure and, subsequently, dispersed into the contaltunent atmosphere. During the process, thermal and chemical energies m directly transfened from the core debris to the containment atmosphere. *lhis phenomenon has become known as airect. containment heating (DCH). 'lhis report presents the results of a series of calculations at Brookhaven '

a National laboratory (BNI.) to provide estimates of the DCH containment pressure loading in the Zion plant subject to a wide range of initial conditions and phenomenological ?

assumptions. The containment loading calculations were performed using a version of the-1 CONTAIN ' code with update modifications, which parametrically characterize DCH _

phenomena (CONTAIN DCH, Version 1.10). The range of calculation parameters was.

selected to represent many of the. current uncertainties in DCH initial conditions and t 3

uncenainties in modeling DCH phenomena. The choice of CONTAIN calculation input '

parameters is discussed and results are pn sented for both a seven-cell nodulization of the '

_ Zion containment buildmg. The seven. cell calculations incorporate all the features of the' CONTAIN-DCH model 'Ihe single-cell calculations are included only for comparison purposes and represent up;st. bound estimates of DCH loads, assuming complete mixing, adiabatic conditions, and thermal equilibrium between the gas and core debris; Calculation results are presented and discussed.16 refs.,12 figs.,8 tabs.

Publication Date:

March 1991

+

Prepared by:

Tutu, NX: Ginsberg, T. (Brookhaven National Lab., Upton, NY); Park, CX (Korea'

- Atomic Energy Research Inst., Daeduk);' Grimshaw, C.A. (Margrove Consulting Ltd..-

London) '

Prepared for:

NRC Division of Systems R~earch, Office of Nuclear Regulatory Research -

a Keywords:

adiabatic processes, aerosols, blowdown, BNL, combustion, computer codes, computerized Il simulation. CONTAIN, CONTAIN.DCH, containment systems, corium, droplets, flow:

rate, heat transfer, hydraulics, hydrogen,' meltdown, mitigation, snixing, particle size,7 pressure vessels, pressurization, primary coolant circuits, quenching, reactor accidents,-

reactor core disruption, reactor safety, steam generators US NRC, Zion-1 reactor. Zion-2 ;

reactor

[

l e

. ?'

i 1-L s

17

-NUREO/BR.-0083 Vol.7 l

. ~, - - - -

NUREG/CR--5300-Vol.1 EGG-2613-Vol.1 IRRAS

Title:

Integrated Reliability and Risk Analysis System (IRRAS), Version 2.5: Reference Manual.

Volume 1 Denription:

The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic rist assessment (iitA) model developmers and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the abihty to create and analyze fault trecs and accident sequences using a microcomputer. 'ntis program provides functions that range from graphical fault tree construction to cut set generation and quantification. Version 1.0 of the IRRAS program was releasedin February of 1987. Since that time, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 2.5 and is the subject of this Reference Manual. Version 2.5 of IRRAS provides the same capabilities as Version 1.0 and adds a relational database facility for managing the data, improved functionality, and imp oved algonthm performance. 7 refs.,348 figs.

Publication Date:

March 1991 Prepared by:

Russell, K.D.; McKay, M.K.; Sattison, M.B.; Skinner, NL; Wood, S.T. (EG&G Idaho, Inc., Idaho Falls,19); Rasmuson, D.M. (Nuclear Regulatory Commission, Washington, DC. Division of Systems Research)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

algorithms, computer codes, computer graphics, computer output devices, computerized simulation, database management, fault tree analysis, LtRAS, manuals, microprocessors, modifications, nuclear power p; ants, performance, probabilistic estimation, reactor accidents, reactor safety, rist assessment, systems analysis NUREG/BR-0083, Vol.7 18 l

l

SAND-90-7116 NUREG/CR--5304 MACCS

Title:

Radionuclide Behavior in the Envimnment

==

Description:==

The purpose of this report is to document the results of the following task: Review for quality and consistency the available data on measurements of initial ground contantination of Chernobyl radionuclides in various pans of Norway and subsequent concentrations of these radionuclides in various environmental media as functions of time and utilize the data obtained to verify the existing models, or to improve them, for describing radionuclide behavior in the environment. Some of the standard processas were; migration into soil; weathering; resuspension, food-chain contamination: and loss or reconcentration by run-off. The task performed within this contract has been to use post-Chernobyl data from Norway to verify or find areas for possible improvement in the chronic exposure pathway models utilized in MACCS. Work bas consisted mainly of collecting and evaluating post.

Chernobyl infonnation from Norway or other countnes when relevant but has also included experimental work performed specifically for the current task, la most connections the -

data available show the models and data in MACCS to be appropriate. A few areas where the data indicate that the MACCS approach is faulty or inadequate are, however, pointed out in the report. 'These should be examined carefully, and appropnate modifications should eventually be made.14 refs.,12 figs.,22 tabs.

Publication Date:

Sepcember 1991 Prepared by:

Tveten, U. (Institutt for Energiteknikk, Kjeller, Norway)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Resecrch Keywords:

bio!ogical accumulation, biological as allabihty, cesium 134, cesium 137, chronic exposure, compiled data, contamination, deposition, environmental transpon, fallout, food chains, MACCS, mathematical models, Norway, quality assurance, radiation accidents, radiation doses, radivecological concentration, radionuclide migration, reviews, risk assessment, root absorption, runoff, soils, surface waters, validatim, weathering i

l l"

19 NUREG/BR--0083, Vol.7

a

(

.i

~

l E

NUREG/CRd5312; S AND -88 3324!

AN

' General"

&ly

~

~

M

Title:

. A hermodynamic Model of Fuel Disruption in ST 1 l

Description:

Preliminary examination of the fuel cross-sections from the ST 1 test indicates that the':

portion of the irradiated fuel that appeared to le severely disrupted in the radi_ographs.

. may well have undergone some sort of partial melting process. his type of disruption /

+l has not been observed in any previous tests involving the BR-3 fuel, and _it is felt that the-highly reducing environment used in ST-1 may be a significant factor in the observed.

j disruption.- Significant fuel relocation was also observed in the Oak Ridge VI-4 test 0

which was perfonned using conditions as rnuch like ST-1 as possible. An uupection of =

the U-O phase diagram shows that one would not expect toLobserve molten urania at i

temperatures below 2693 K. %is temperature is considerably higher than the 2500 KL maximurn fuel temperature achieved in the ST 1 test. De lower temperature: ponion of

the phase diagram indicates that at temperatures extant in the ST 1 test one could find !

j liquid uranium if the O!M ratio of the fuel were sufficiently reduced. In the ST 1 test, tir coolant flow gas was a 1:2 mixture of hydrogen in argon. This mixture presents a; s

reducing environment for the fuel. The existence in the package of a large quantity off unoxidized zircortium adds to the overall reducing nature.of the test environmern.'A-

- thermodynamic model that qualitatively accounts for the observed fuel disruption in thcl ST 1 test is presented. he model is based on Winslow's equation for the oxygen pressure ;

over hypostoichiometric fuel and the reducing nature of the test atmosphere. %c' y

stoichiometry of the fuel is calculated as a function of temperature. This calculation-predicts panial liquefaction of the irradiated fuel in the test. 8 refsl,2 figs.,2 tabsi Publication Date:

February 1991-.

d

- Prepared by:

Grimley, AJ. (Sandia National l_ abs., Albuquerque, NM)

Prepared for:

NRC Dinsion of Systems Research, Office of Nuclear. Regulatory Research Keywords:-

BR 3 reactor, chemical reactions, entropy, flow rate, fuel elements, heat transfer, hydraulics, irradiation, melting, Newton method, oxidation, phase diagrams, phase transformations,1-post-irradiation examination, pressure dependence, quantity ratio reduction, research programs. Sandia laboratories, stoichiometrj', temperature dependence, test facilities, f

uranium dioxide, US NRC, zircanium -

o;

.1

~.

4

.NUREG/BR-0083, Vol.7 20

.L.

'^

SAND--89c0072

.NUREG/CR-5331 MELCOR, HECTR

Title:

MELCOR Analyses for Accident Progression issues i

==

Description:==

Results of calculations performed with MELCOR and HECTR in support of the NUREG.

I150 study are presented in this report. The analyses examined a wide range of issues.

The analyses included integral calculations covenny an entire accident sequence, as well I

as calculations that addressed specific issues that could affect several accident sequences.

The results of the analyses for Grand Gulf, Peach Bottom, LaSalle, and Sequoyah are described, and the major conclusions are summarized 23 refs.,69 figs.,8 tabs.

Publication Date:

January 1991 Prepared by:

Dingman, S.E.; Shaffer, CJ.; Payne, A.C.; Carmel, MK (Sandia National Labs.,

Albuquerque, NM)

Prepared for:

NRC Division of Systems Research, Of6ce of Nuclear Regulatory Research -

Keywords:

blackouts, computer calculations, containment, failure mode analysis, Grand Gulf l reactor, Grand Gulf-2 reactor, heat transfer, HECTR, hydraulics, La Salle County.1 reactor, La Salle County-2 reactor, MELCOR, Peach Bottom-1 reactor, Peach Bottom-2 reactor, Peach Bottom-3 reactor, reactor components, reactor safety, reactor vessels, sensitivity analysis, Sequoyah-1 seactor, Sequoyah-2 reactor 21 NUREG/BR--0083,Vol.7

NUREG/CR--5345 SAND--89-0308 VICTORIA

Title:

Fission Praiuct Release and Fuel Behavior of Irradiated Light Water Reactor Fuel Under Severe Accident Conditions.The ACRR ST 1 Experiment

==

Description:==

The Atmular Core Research Reactor (ACRR) Source Term (ST) Experiment program was designed to obtain tirne-resolved data on the release of fission products from irradiated fuels under well-controlled hght water reactor severe accident conditions. The ST-1 Experiment was the first of two experiments designed to investigate fission product release. ST 1 was conducted in a highly reducing environment at a system pressure of approximately 0.19 MPa and at maximum fuel temperatures of about 2490 K. 'Ihe data will be used for the development and validation of rnechanistic fission-product release computer codes such as VICTORIA.

Ptiblication Date:

November 1991 Prtpared by:

Allen, M.D.: Stockman, H.W.; Reil, K.O. (Sandia National Labs., Albuquerque, NM);

Fisk, J.W. (Tills and Associates. lac., Albuquerque, NM)

Prepared lor:

NRC Division of Systems Research. Office of Nuclear Regulatory Research Keywords:

ACRR reactor, BWR type reactors, experimental data, fission-product release, fission products, fuel elements, irradiation, post-irradiation examination, PWR type reactors, radioisotopes, reactor accidents, reactor safety, source terms, test facilities, VICTORIA NUREG/BR-0083,Vol.7 22

n NUREG/CR-5352-Rev.1 VAM2D

Title:

VAM2D: Variably Saturated Analysis Model in Two Dimensions. Version SIwith Hysteresis and Chained Decay Traruport: Documentation and User's Guide: Revision 1

==

Description:==

This repon documents a two-dimensional finite element model VAM2D, de'veJoped to simulate water flow and solute transpon in vanably satured porous media. Both ilow-and transport simulation can be handled concurrently or sequentially, The fonnubtion of tie goveming equations mi the numerical procedures used in the code are presented 1he flow equation is rnroximated using the Galerkin finite element method. Ncditear soil moisture charactensues and atmosplx-ic boutdary conditions (e.g.,infiltratier,bvaporation, and seepage far) are treated using Picard and Newton-Raphson iteratiota:. Hysteresis effects and anisotropy in the unsaturated hydraulic conductivity can be taken irJ6 ucemt if needed. The contaminant transpon simulation can account for advection, hydrodynenic dispersion, linear equilibrium sorption, and first-order degradation. Transpon of a single component or a multicomponent decay chain can be handled. The trampun equation is approximated using an upstream weighted residual method.' Several test problems are presented to verify the code and demonstrate it.; utility These problems range from simple one-dimensional to complex two-dimensional and axisymmetric problems. This document has been produced as a user's manual, it contains detailed infomiation on tle code structure along with instructions for input data preparation and sample input and printed output for selectef ut problems. Also included are instructions for job set up and restaning procedes. 44 refs.,54 figs.,24 tabs.

Publication Dr.te:

October 1991 Prepared by:

Huyakom, P.S.; Kool, J.B.t Wu, Y.S, (HydroGeoLogic, Inc., Hemdon, VA)

Prepared for:

NRC Division of Regulatory Applict.tions, Oiace of Nuclear Regulatory Research Keywords:

boundary conditions, computer program documentation, computerized simulation, finite -

clement method, fluid fLw, groundwater, hydrology, leaciung, mathematical models, numerical solution, radioactive waste disposal, radioactive waste facilities, radionuclide migration, site characterization, soils, transpiration, VAM2D ll l

23-NUREG/BR -0083 Vol.7

p y

UNUREG/CRJ-5377:

MACCS, ARANO, CRAC, NECTAR,-NUCRAC, UFOMOD:

Title:

- Review of the Chronic Exposure Pathway Models in MACCS and Several Other Well-

- Known Probabilistia RisbWoneni !.Stis

==

Description:==

The purpose of this report is to dacumna the resnits of the work perfonned by the author.

in connection with the following task, performed for US Nuclear Regulatory Commission =

f x (USNRC), Office o Nuclear Regulatory Research, Division of Systems Research: MACCS '

' Chronic Exposure Pathway M&iels: Review the chronic exposure. pathway models

. implemented in the MELCOR Accident Consequence Code System (MACCS) and compare.

those models to the chronic exposure pathway models' implemented in'similar' codes.

-i developed in countries thii et inunbets of the Orgardsation for Economic Co-operation =

and Development (OECD). The chronic exposures concemed are via the tenestrial food

- pathways, the water pathways, the long-tenn groundshine pathway, and the inhalation of.

~ I resuspended rat.ionuclides pathway. The USNRC has indicated during discussions of the

> )"

task that the major effort should be spent on the terrestrial food pathways. There is one '

chapter for each of the categories of chronic exposure pathways listed above,-

Publication Date:

June 1990 Prepared by:

Tveten, U,(Institutt for Energiteknikk, Kjeller, Norway) '

N-Prepared for:

NRC Divisioc of Systems Research, Office of Nuclear Regulatory Research,

Keywords:

aquatic ecosystems, ARANOi-biological pathways, chronic exposure,' comparative -

evaluations, computer codes, CRAC, danking water, environmental exposure pathway, food chams, inhalation, MACCS, NECTAR, NUCRAC, plants; radioactivity, radioisotopes,i radionuclide migntion, risk assessment, terrestrial ecosystems, UFOMOD i

3 4

1 J

1 NUREO/BR--0083 Vol.7 24 T

J EPRI-NP-6480 Vol.1aUAW--2023-Vol.l NUREG/CR--5395-Vol.1 RELAP5/ MOD 2, TRAC '

Title:

Multiloop Integral Systeso Test (MIST): Final Report. Volume 1, Summary

==

Description:==

The Multiloop Integral System Test (MIST) is pan of a mukiphase program started in 1983 to addr-ss small break loss-of-coolant accidena (SBLOCAs) specific to Babcock &

Wdcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock & Wilcox Owners Group, the Elecnic Power Research Institute, and Babcock '

& Wilcox. The unique features of the Babcock & Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facihties to ackiress the thermal 4ydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility---the Once Through Integral System (OTIS)-was also used. Data from MIST and OTIS are used 3.o benchmark the adequacy of system codes, such as RELAP5 and TRAC, for.

predicting c.bnont.al plant transients. The MIST program is reported in 11 volumes.

Volumes 2 through 8 pertain to groups of Phase 3 tests by type; Volume 9 presents intergroup comparisons; Volume 10 provides compansons between the RELAPS/ MOD 2 calculations and MIST observations, and Volume 1I (with addendum) presents tir later Phase 4 tests.This is Volume 1 of the MIST final report, a summary of the entire MIST program, Major topics include Test Advisory Group (TAG) issues, facility scaling and design, test matrix, observations, comparison of RELAP5 calculations to' MIST observations, and MIST versus the TAG issues. MIST gene ated consistent integral.

system data covering a wide range of transiern interactions. MIST provided insight into integral system behavior and assisted the code effort. The MIST observations addressed each of the TAG issues. I1 refs,29 6gs.,9 tabs.

Publication Date:

April 1991 Prepared by:

Gloudemaas, LR. (Babcock & Wilcox Co., Lynchburg, VA. Nuclear Power Division)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research; Electric Power Research Institute; Babcock & Wilcox Owners Group Keywords:

blackouts. computerized simulation, experimental data, heat transfer, hydraulics, loss of-coolant, pumps, PWR type reactors, reactor cooling systems, reactor safety, RELAPS/

MOD 2, ruptures, scale models, steam generators, test facilitier IRAC, transients, tubes,-

two-phase flow l

25 NUREG/BR--0083,Vol.7 f

NUREG/CR~5423

-MELTSPREAD, CORCON

Title:

Tle Probability of Liner Failure in a Mark ! Containtnent -

==

Description:==

An integrated analysis of Mark.1 liter anack in a postulated core melt accident is presented.- -

The approach consists of the mechanistic treatment of the sequence (,f physical phenomena that lead to liner contact by corium debris and their coupling through a probabilistic framework that allows representation of uncertainties. We emphasize a physically consistent treatment in eadi sequence but allow for qualitatively different scenario ' s represent the range of behavior due to model uncertaintics 'Ihe results are presented in a format that allows their direct use in PRAs, and, in panicular, expert opinion is incorporated by a new methodological approach (first applied in our study of tr-mode failut< > 'UREO/

- CR 50") that involves expert wwiew of, and comment on, a fully documern o study all under one cover. 34 refs.,90 figs.,17 tabs.

Publication Date:

August 1991 Prepared by:

Theofvous, T.G.; Amarasooriya, W.11; Yan, ILt Ramam, U. (Califomia Univ., Santa Barbara, CA, Dept. of Chemical and Nuclear Engineering)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

BWR type reactors, CCI model, concretes, containment, CORCON, corium, failures, fluid flow, heat transfer, hydraulics, hydrodynamics, liners, meltdown, MELTSPREAD, probabilistic estimation. probability, reactor safety, reactor vessels, risk assessment 1

f l

I I

NUREG/BR-0033,Vol.7 26_

ANL-91/6 NUREG/CR-5456 COMMIX-IC

]

Title:

Analysis of Flow Stratification in the Surge Line of the Comanche Peak Reactor 1

==

Description:==

A numler of nuclear power plants have reported failure of reactor components due to flow rtratification. Tierefore, a fun < lamer'tal enderstarKimg of, and a capability to predict, flow strad'ication in a reactor system is critically important to reactor performance and safety. The work presented here is the first step in this direction and will contritmte to the resolution of the issue of flow stratification. An analysis is perfonned using the COMMIX-IC computer program for the surge line of the Comanche Peak reactor, A comparison is made between the calculated results from the COMMIX code and the plant-measured data, and the agreement is good.1I refs.,36 figs., I tab.

Publication Dete:

April 1991 Prepared by:

Sun, J.O.; Shen. Y.IL; Sha, W.T. (Argonre Nt.tional Lab., IL)

Prepared for:

NRC Division of Safety Programs, Office for Analysts and Evaluation of Operational Data Keywords:

algorithms, blackouts, Comanche Peak 1 reactor, Comanche Peak 2 reactor, COMMIX-1C failures, Duid flow, fluid mechanics, heat transfer, hydraulics, Navier-Stokes equations, reactor accidents, teactor components, reactor coohng systems, reactor safety, stratification, temperature distribution, thermal analysis i

27 NUREG/BR--0083 Vol.7

NUREG/CR-5518 S AND-90-0011 CONTAIN

Title:

Quality Assurance Procedures for the CONTAIN Severe Reactor Accident Computer Code

==

Description:==

The - )NTAIN quality assurance program follows a strict set of procedures designed to ensure the integnty of the code, to avoid errors in the code, and to prolong the life of the code. The code itself is maintained under a code-configuration coutrol system that provides a listorical record of changes. All changes are incorporated using an update process 3r 4

that allows separate identification ofimprovements made to each successive code version.

Code modifications and improvements are formally reviewed and clecked. An exhaustive,

~f multilevel test program validates the theory and implementation of all code changes through assessment calculations that compare the cole-predicted results to standard handbooks of ideahzed test. ases. A document trail and archive establish the problems solved by the software, the verification and validation of the software, software changes and subsequent teverification and revalidation, and the tracking of sof ware problems and t

actions taken to nesolve those problems. 'Ihis document describes in detail the CONTAIN quality assurance procedures. 4 refs.,21 figs.,4 tabs.

Publication Date:

January 1991 Prepared by:

Russell, N.A.; Washing *on, K.E.; Bergeron, K.D.; Murata, K.K.; Carroll, D.E. (Sandia National Labs., Albuquerque, NM); Harns, C.L (los Alamos Technical Associates,Inc.,

Albuquerque, NM)

Prepared for:

NRC Division of Systems Research. Office of Nuclear Regulatory Research Keywords:

CONTAIN, containment, containment buildmgs, documentation, errors, maintenance, nuclear power plants, performance testing, quality assurance reactor accidents 4

L NUREG/BR-0083,Vol.7 28 l

EGG--2630

- NUREG/CR--5520 IRRAS, MAR D. Database

Title:

Procedures Guide for Extracting and Loading Probabilistic Risk Assessment Data into MAR D Using IRRAS 2.5

==

Description:==

The Models and Results Database (MAR D) can be used to organize infonnation from probabilistic risk assessments (PRAs). Data may be entered into M AR D electrordcally or manually. T1us report concentrates on manual data-entry methods and documents the use -

of the Integrated Reliability and Risk Analysis System (IRRAS), Version 2.5 and ASCII text editors to load level 1 (intemal event) PRA models into MAR D. Step-by-step instructions for using IRRAS 2.5 are provided, which will help the user transfer data from a printed (hardecpy) source to MAR D.

Publication Date:

November 1991 Prepared by:

Fowler, R.D.: Judd D.L; Pham, M.; VanHom, R.L.; Wolfram, LM. (EG&O Idaho, Inc.,

Idaho Falls,ID)

Prepared for:

NRC Division of Safety issue Resolution, Office of Nuclear Regulatory Research Keywords:

computer program documentation, data processing, database management IRRAS, MAR D database, nuclear facilities, nuclear power plants, risk assessment -

29 NUREO/BR--0083,Vol.7 l

l

NUREG/CR--5522 SAND--90 0128 MODFLOW, INVFD

Title:

A Comparison of Parameter Estimation and Sensitivity Analysis Techniques and Their impact ci the Uncertainty in Grout.d Water Flow Model Predictions

==

Description:==

This work documents a comparison of sensiuvity and uncertainty analysis techniques that are likely to be used in support of repository performance assessments to determine compliance with the Nuclear Regulatory Commission (NRC) and the Environmental Protection Agency (El A) regulations for lugh-level radioactive waste (HLW) repositories.

A variety of parameter estimation and sensitivity analysis techniques welc applied to a model of the Avra Valley aquifer, Arizona. Two approaches to sensitivity analyses were used, statistical and detennittistic; these were applied to evaluate the sensitivity of the groundwater travel time to changes in transmissivity. We effect of different boundary conditions on tic calculated sensitivity derivatives was also evaluated. Parameter estimates and estimation errors were obtained via geostatistical and inverse techniques. The

throughput" of the kriging techniques suggests that the mean estimates cenved from these techniques are frequently "off the mark" or inconsistent with the conceptual model.

With no screening of the input parameter estimates for realism, non-conservative travel time estimates were obtai~d. The differential analysis sensitivity technique is shown to be dependent on the cb of design point, providing only a local measure of the sensitivity, 'nw statistical approach to sensittvity identifies parameters which are both sensitive and uncertain, i.e., it shows when the uncertainty in a model parameter is important. Sensitivity estimates are also shown to be dependent on the choice of toundary conditions used. 92 refs.,55 figs.,13 tabs.

Publication Date:

May 1991 Prepared by:

2.immennan, D. A. (GRAM, Inc., Albuquerque, NM); Hanson, R.T. (Geological Survey, San Diego, CAK Davis, P.A. (Sandia National Labs., Albuquerque, NM)

Prepared for:

NRC Division of High-Level Waste Management, Office of Nuclear Material Safety and Safeguards Keywords:

aquifers, clays, computerized simu ation, data analysis, data compilation, data covariances, i

errors, flow models, forecasting, groundw ater, Fydrology, INVFD, kriging, MODFLOW, monitoring, Mmte Carlo methd origin, secommendations, regn:ssion analysis, regulations, response functions, sensitivity analysis, silt, site characterization, stratigraphy, welllogging, wells NUREG/BR -0083,Vol.7 30 1

g.

E

?

NUREG/CR-5531 l

SAND '-90 0364 L

o l;

MELCORy 4

Title:

_ MELCOR 1.8.0: A Computer Co; for Nuclear. Reactor Severe Accident Source Term

[

- and Risk Assessment Analyses z_:

1 i

==

Description:==

MELCOR is a fullyintegrated. engineering-level computer code that models the progression l of severe accidents in light water reactor nuclear power plants. MELCOR is '3ing.

b developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and tie successor to tie Source Term ~

Code Package. The entire spectmm of severe accident phenomena, including reactor l coolant systern cad containment thennal-hydraube response, core heatup, degradation.

.?

and relocation, and fission-product release and transport, is treated in MELCOR in a-unified framework for both boiling-water reactors (BWRs) and pressurized. water reactors (PWRs). MELCOR has been especially designed to facilitate sensitivity and uncertainty =

analyses, its current uses include estimation of severe accident source terms and their -

sensitivities and uncenainties in a variety of applications. This report is a summary of MELCOR 1.8.0, the code version released in March 1989. Condensed information is -

~,

presented on its developmental history, structure, modeling features and capabilities,-

verification and validation, and quality assurance. Detailed documentation on these aspects -

of MELCOR, including users' guides, a:ference m?.~ alsi programmers' guides,' and assessment and application repons,is avauable in draft form and is distnteed to MELCOR users.

Pullication Date:

January 1991 Prepared by:

Summers, R.M.; Cole, R.K., Jr.: Boucheron, E.A.: Carmel, M.K.: Dingman, i,*.L.; Kelly,.-

J.E. (Sandia National Labs., Albuquerque, NM) --

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research,

i Keywords:

BWR type reactors, contrunment, fission.psoduct release, heat transfer,' hydraulics, -

MELCOR, meltdown, PWR type reactors, reactor safety, risk assessment, source terms N

F 31 NUREG/BR-0083,Vol.'7 r-t' ier y--

y d

m.

f t "

NUREG/CR--5536 -

l S AND-o90-7015 -

DCM3D w

L

Title:

_ : DCM3DL A Dual-Continuum, Three-Dimensional,l Ground Water Flow Code for t

'- Unsaturated. Fractund. Porous Mediai

==

Description:==

This report constitutes the user's manual for DCM3D; DCM3D is a computer code for solving three-dunensional, groundwater flow problems in variably saturated,' fractured

. porous media. The code is based on a dual-continuum model with porous media comprising i one continuum and fractures comprising tie other, The continua are connected by a1 transfer term that depends on the unsaturated permeability of the porous medium. An :

integrated finite-difference scheme is used to discretize the goveming equations iu space.

The time-dependent term is allowed to remain continuous. The resulting set of ordinary differential equations (ODE's) is solved with a Feneral ODE solver, LSODES. *lhe code '

is capable of harding transient, spatially dependent source terms and boundary conditions.

The boundary conditions can be either prescribed head or prescribed flux. 24 refs., -

22 figs.,5 tabs.

Publication Date:

February 1991 4

Prepared by:

Updegraff, C.D. (GR\\M, Inc..-Albuquerque, NM); lee, C.E. (Applied Physics, Inc,

Albuquerque, NM): Gallegos, D.P. (Sandia National Labs., Albuquerque, NM)

Prepared for:

. NRC Division of Engineering, Office of Nuclear Regulatory Research.

.i Keywords:

DCMoD, differcatial equations, environmental transport, flow rate, fluid flow, fractures, groundwaten hydrology, porous materials. pressure dependence, space dependence,-

subsurface environments, three-dimensional calculations, time dependence, water saturation 3

F 9

4 -.

J l

p NUREG/BR--0083.Vol.7 32-

SAND 90 0575 NUREG/CR 5537 Geperal I

Title:

Approaches for the Validation of Models Used for Perfonnance Assessment of Iligh-level Nuclear Waste Repositones Descriptiott:

The purpose of this report is to provide general approadrs and concepts that can le applied in vahdation of models used in performece assessment of hir.h level waste (IILW) repositories nic approaches are based on a validatic rtrategy that Sarxtia National Laloratories has implemented as panicipants in the International Traraport Validation Study (INTRAVAL). This strategy focuses on the demonstration that jerformance assessment rnodels are adequh;e representations of the real systems they are intended to represent, given the pertinent regulatory requirements raniet than proving absolute conectness inen the purely scientific point of view. positforu that are taken comast of the following: due to the relevet time and space scales, models that att used to assess the perfonnance of an IILW repository can never le validated; t!erefore, validation is

• pro ess that consists of building confidence in these models and not provichng " valid.ated-tmxtets; in this content, model validation includes comparisons to " reality." however, adequacy for the given purpose is the overall goal; comparisons to *rcality" consist of comparing matel predictions against laboratory and field experiments, natural analogues, and site specific infonnation; when comparing experimental data lo model predictions, a model can te either " invalid" or "not invalid." based on the null hypothesis concept, however, conGdence in the model arises frorn fitxhng a model to le "not invalid" over a wide range of corxhtions; an attempt should te made to consider in the validat on process i

all plaudble conceptual rnodels; and when comparing experimental data to model predictions, a logical systematic approach should le followed. This report discust.es the definition of validation in the context of performance assessment for HLW repositories, the need for validation, an approach to validation, and an approach to cornparing rnodel predictions wita experimental data proposed by the authors. 21 refs.. I fig.

Publicallon Date:

March 1991 Prepared by:

Davis, p. A.: Otague, N.E. (Sandia Nationd I ms, ' - iguerque, NM); Goodrich, M.T, (CiRAM,Inc., Albuquerque,NM)

Prepared for:

NRC Division of High-Level Waste Management, Otfree of Nech erial Safety and Safeguards Keywords:

benchacale experiments, compliance. CRSTER model, design, field tests, high-level l

radmactive wastes, licensing, mathematical models, performance testing, radiation protection, radioactive waste disposal, radioactive waste facilities, recomrrendations, tegulations, site approvals, specificatmns, undergroutxi disposal, validation l

33 NUREG/BR UB3,Vol.7 e

e

NUREG/CR. 5539 SAND. 90 05H5 VAM2D, GENil

Title:

A self Teaciung Curnculum for the NRC/SNLlew-level Waste Perfonnance Assestment Methodology Descripiir..

A perfonnance assessment methodology has teen de* 'opeo for use by the US Nuclear Regulatory Commission in evaluaung license applicanora for low level waste disposal facilities. This report provides detailed guidance on input and output procedures for the computer ecxtes recommended for use in the methodology, Seven sample problems are provided for various aspects of a perfonnance amssmer. andysis of a simple hypothetical conceptual model. When combined, these sample problems demonstrate how the methmlology is used to produce a dose history for the site under ncnnal conditioru and to dernonstrate an analysis of an intruder scenano,20 refs.,26 figs. 4 tabs.

Publication Date:

January 1991 Prepared by:

Chu, M.S.Y.: Kozak, M.W.; Camplell, J.E (Sarnha Nabonal Labs. Albuquerque, NM);

'Thempson, B.ll. (INTERA, Inc., Albuquerque, NM)

Prepared for:

NRC Division of lew Level Waste Management and Deccenmissioning,0ffice M Nuct;ar Matenal Safety and Safeguards Keywords:

alpha learing wastes, dosimetry, environmental exposure pathway, environmental polky.

flow models, fluid flow, GEN!!, groundwater, license applications, low level radioactive wastes, nuclear waste policy acts, performance testing, radiation doses, rad:oactive waste facilities, radioecological concentration, radionuclide migration, reviews, naurce tenns, surface waten, undergmund disposal, US NRC. VAM2D NURiiO/BR--0083, Vol.7 34

1 SAND-90 7020 NUREG/CR 5561 General

Title:

Analysis of Itellows Expansion Joints in the $cquoyah Containment

==

Description:==

15ellows exparsioc joints are an integral part of the containment buildmg pressure tardary -

in some nuclear power pianu. They are 'used at piping penetrations to ruirdrnize the loadmgs on the contairmeent shell 9ue to differential movement telween the shell and piping. Tie purpose of this study was to investigate tellows behavior in _the unlikely j

event of a severe acci&nt inside the contaitunent building. The study trgan with a i

survey of available infonnation on tellows design, analysis, and past test prograins, This information was then used to assess the ultimate behavior of the tellows in the Sequoyah containment. It was detennined that the tellows at penetration X 47 in the Sequoyah containtnent would experience the worst loadmg conditiota during a severe accident.

Firste element calculatiorn ofI ellows X-47 were conducted to examme the deformation and resulting strains caused by the cornbination of axial compression, lateral offset, tendmg, and internal pressure that would te applied to the bellows during a severe accident. Ilecause of convergence problerns, the analyses could not le continued to a point of obvious tellows failure. Ilowever, large inelastic bernbng strains, up to 8%.

were calculated. A test prograin to determine the ultimate tellows tehavior ar.d develop data for validation of analytical methods is secommirxled.

Publication Date:

Decernter 1991 Prepared by:

Onimann, L: Wassef, W.t Fanous, F.; Bluhm, D. (Arnes Lab., IA)

Prepared for:

NRC Livision of Engineering, Office of Nuclear Regulatory Research Keywords:

bellows, ctanpression, containment systems, deformation, expansion joints, firste elemern method, pipes, pressure elbets. Sequoyah-1 reactor, Sequoyah 2 reactor 1

l~

l 35 NUREO/BR-0083,Vol 7 h

~

NUREG/CR. 5565 ORNLff'M.11548 IlWR LTAS,llWRSAR, MELCOR

Title:

The Response of BWR Mark II Contaitunents to Station Blackout Severe Accident Sequences

==

Description:==

This report describes the results of a series of calculations conducted to investigate the response of BWR hiark 11 containments to shM term arx1 long term station blackout severe accident sequences. The BWR LTAS, SWRSAR, and MELCOR codes were employed to conduct quantitative accidern sequence progression and containmers response analyses for several station blacLout scemirios. The accident initigation effectiveness of autornatic depressuritation system actuation, drywell flooding via containment spray operation, and debns quenching in Mark 11 suppression pool. is assessed. 27 refs.,

16 figs.,21 tabs.

Publication Date:

May 1991 Prepared by:

Oreene. S.R.; lk>dge, S.A.; ilyman, C.R.; Tobias, M.L. (Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Safety issue Resolution, Office of Nuclear Regulatory Research Keywords:

blackouts, BWR type reactors, BWR-LTAS, BWRSAR, containment, depressurization systems, evaluation. MELCOR, mitigation, pressure suppression, reactor accidents l

NUREO/BR 0083,Vol.7 36

l ORNL/TM 11549 NURF.G/CR. 5571 IlWR LTAS, llWRSAR. MELCOR

Title:

The Response of fiWR hiark 111 Containtnents to Short. Term Station Blackout Severe Accident Sequerres Descriptiott:

This repon describes the results of a series of calculations conducted to invettigate the i

response of BWR hiark 111 contaisunents to short term station blackcut severe accident i

sequences. The BWR-LTAS, BWRS AR, and hiELCOR codes were etaployed to conduct j

quantitauve accideni sequence progression and contaisunent response analyses for t,cvera!

I station blackout scenarios.1he accident snitigatien effectiveness of containment venting and bac-Lup cruerpency pow,.t for contaitunent hydrogen ignitors and drywell vacuum I

breaLers is assessed.18 refs.,142 figs.,12 tats.

Publication Date:

June 1991 Prepared by:

Greene, S,R. Ikidge, S.A.; Hyman, C.R.: Partm, B.W.: Tobias, bl.L (Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Safety Issue Resolution, Office of Nuclear Regulatory Research Keywords:

blackouts, bWR type reactors, BWR LTAS, BWRSAR, Clinton 1 reactor, Ginton 2-reactor, cornbustion, computer calculations, computer prograrn documentation, containment, containment buildmgs, containment systems, corium, Ota covariances, detonations, fue' cans, fuel elements, Grand Gulf l reactor, Grand Gulf-2 reactor, heat transfer, hydraulic %

hydrogen, loss of coolant, hfELCOR, meltdown, Perry 1 reactor, Perry 2 reactor, reactor components, reacte* core disruption, reactor cores, reactor safety, reactor vessels, River Bend l reactor, River Bend-2 reactor, steam 37 N tfREG/BR- 0083, Vol.7

~-

NUlt EG/ Cit-55M General

Title:

Extemion and Extrapolation of J R Curves ux!Their Apphcation to tir 1.ow Upper Shelf Toughness issue

==

Description:==

This document develops metints of measuring cyrsinen' ally the lunits of valid fracture mecharucs data that can be obtamut from srnall haeture mecharucs r.pecimens. The pr9poted technique generally simws that present ASTM limits are overly corue:vative and the new technique would allow almost a ttueefold increase in the amount of crack extension allowed in the testing of a surveillance specimen. Analytic relationships are then developed to allow use of the new expenmentally measured limit to J controlled crack growth for design or f ailure analysis apphcations to pressure vessel structures. The new region of J controlled crack growth is shown to correlate test with the omega criterion which defines hmits on leth the madmum J level and the maximum crack extension allowable for a particular specimen sire and material toughness combination.

The final section looks at the problem of e strapolation of J.R curve data when needed for a structure fracture analysis. Several forms of strapolation relauonships are compared from the point of view of accurate and conservative extrapolation, particularly from the standpoint of tearing instability analysis of a growmg ductile crack on the material upper shelf. 35 refs.,38 figs.,12 tabs.

Publication Date:

March 1991 Prepared by:

Joyce, J.A. (Naval Academy, Annapolis, MDx llackett, E.M. (David Taylar Research Cetiter, Annapohs, MD)

Prepared for:

NRC Division of Engineering. Office of Nuclear Regulatory Research Keywords:

s.ging, BWR type tractors, Charpy test, crack propagation, Eason equauon, elasticity, fracture mechanics, mathematical models, omega enterion, physical radiation effects, plasticity, prvssure vessels. PWR type reactors, reactor accidents, reactor safety, service life, surveillauec, tensile properties, validation NUREO/BR--0083, Vol.7 38

ORN1JTM-I1581 NUREG/CR-!592 General

Title:

Analytical Studies of Transverse Strain Fjfects on Fracture Tougiriess for Ctrcumfuentially Oriented CracLA

==

Description:==

1he objective of tlus report is to describe the developrnent of analysis metints for estrrnating the decrease in crack irutiation toupluess, from a reference plane strain value, due to positive straitung along the crack front of a circumferenual flaw in a reactor pres <.ure vessel. The analysis inethods are Sa'ed on two different approaches that are currently being developed to analyze and to explain the effects of transverse strain and stress states on fracture toughness. The first approach is a micro-rnecharucal approach that provides a relation letween fracture toughness and more fundamental material propenies that can be determined experimentally. The second approch focuses on the development of correlation parameters that relate fracture toughness with nominal stress ard r. train states. In tie first phase of this work, the scope of the investigation is limited to car.ck front constraint conditions that can le descriled in tenns of conventional one-parameter (K or J)in plane near tip fields and the transverse strain. Validation checks of the analysis methods against existing fracture data for conditions of contained crack tip yieldings are promising but incomplete. Recommendations for subsequent phases of the work considered necessary to provide more precise estimates on the effects of positive out-of plane straining on the crack irutiation toughness of circumferentially oriented flaws are included. 91 refs.,53 figs. 7 tabs.

Publication Date:

April 1991 Prepared by:

shum, D.K.f,t; Merkle, J.O.; Keeney. Walter, J.; Itass, D.R. (Oak Ridge Nat!onal Lab.,

1N)

Prepared for:

NRC Divisien of Engineering, Office of Nuclear Regulatory Research Keywords:

conclation functions, correlations, crack pigagation, cracks, futite element rnethod, fracture mechanics, fracture propernes, mathematical models, notches, pressure effects, pressure vessels, pWR type reactors, recommendations, steels, strains, stress analysis, tensile properties, thennal shock, welded joints 39 NUREO/BR--0083, Vol.7

)

NUREG/CR-.5595 SEA. 89 461 11.All FORECAST

Title:

FORECAST: Regulatory Effects Cost Analysis Software Annual. Version 3.0

==

Description:==

Over the past several years the NRC has developed a generic cost methodology for the quantification of cost / economic impacts associated with a wide range of new or revisod regulatory requirements. This methodology has tren develojed to aid tle NRC in preparing Regulatory Irnpact Analyses (RIAs). 'llsse generic coving rnettxdi can le useful in quantifying impacts toth to industry and to the NRC. 'fhe FORECAST program was developed to facihtate the use of the generic costing r:tethmlology Thie PC prograrn integrates the major cost considerations that may tw required lecause of a regulatory change. FORECAST automates inuch of the calculations typically needed in aa RIA and thus reduces the time and later required to perfonn these analyses. More importantly, its integrated and consistent treatinent of tie different cost elements should help assure compreheroiveress, maiformity, and accuracy in the preparation of needed cost estimates.

Publicallon Date:

Nevernier 1991 Prepared by:

Lopez, IL, Sciacca. F.W. (Science and Engineering Associates, Inc., Allmquerque, NM)

Prepared for:

NRC Division of Regulatory Applications Of0cc of Nuclear Regulatory Research Keywords:

administrative procedures, computer program dxumentation, cost, cost estimation, economic analysis, economic impact, FORECAST, forecasting, industry, nuclear power plants, radioactive waste disposal, reactor fueling, reactor shutdown, regulations, US NRC NUREG/BR--0083,Vol.7 40

IINL NUREG 52251 NUREG/CR. 5611 General

Title:

!ssues and Approacts s for Using Equipment Reliabihty Alert Levels

==

Description:==

This repon descrites wont accomplished to idenufy issues and approaches to establish alert levels for cosuponent reliabihty. Reliability alert levels are established on standby component counts of s.uc4ess and failure, where equipruent demands are mortitored and counted to ascertain if as.wmptions about acceptable rebatulity are hicly to le correct. A l

Monte Carlo simulation was used to determine the detection responses and false alarm I

rates of several alert level systems. The detectiori responses were obtained in response to a specified reliabibty degradation. Two of the alert systerns were demonstrated with actual failun: data on the Emergency Diesel Generator for five plants.11urden and risk measures of effectiveness were developed to compare different alert level schemes having -

different detection responses and false alann rates. 7 refs.,32 figs.,6 tabs.

Publication Date:

June lo91 Prepared by:

Lofgren. E.V.; Gregory, S.H. (Science Applications Intemational Corp., Mclean, VA)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

alann systems, equipment, failu e rmxte analysis, failures, Monte Carlo method, nuclear power plants, perfonnance, reactor components, reactor safety, reliability 4i NUREG/BR.-0083, Vol.7 j

NUREG/CR. 5614 EGG. 2614 PORF1,0

Title:

Peiformance of Intaa and Partially Derraded Concreie 11aniers in Limiting fluid Flow Descriptlun:

Concrete bamers will play a citical role in de long-term isolation oflow level radioactive wastes. Over time the barriers will degrade, ar'd in muy cases the fundaruental processes controlhng performance of the barriers will be different for intact and degraded conditions.

11us dxument examines factors c<sitrolling fluid flow through intad ard degraded corerete disposal facilities. Simplifiad models are presented for predicting build up of fluid above a vault; fluid flow through and around intact vaults, through flaws in coatings / liners applied to a vault, and through cracks in a concrete vault; and die influence of different backfill materials around the outside of the vault. Exarnple calculations are presented to -

illustrate the parameters and processes that influence fluid flow,46 reis.,49 figs.,2 tabs.

Publication Date:

July 1991 1

Prepared by:

Walton, LC.: Seitz, R.R. (EG& G Idaho, Inc., Idaho Frdis, ID) i Prepared for:

NRC Division of Engirwering, OfGee of Nuclear Regulatory Research I

Keywords:

computerired simulation, concretes, cracks, flow rate, fluid flow, hydraulics, leaks, low-level radwactive wastes, mathematical models, membranes, permeability, PORFLO, PSOR method, radioa tive waste storage, saturation, soils, storage facilities, underground storage, water NUREG/BR--0083, Vol.*1 42

i IINL.NUREG. 52297 NUREG/CR. 5620 i

TilATCil l

Title:

THATCil: A Computer Cale for Modelling Thermal Netuorks of liigh Temperature Gas Cooled Nucient Reactors Descripilon:

This report documents the TilATCil code, winch can te u'ed to rnalel gerwral tlwrmal and now networks of solids and coolant charuels in two-dunensional r geometries. The mam application of TilATCil is to model reactor thenno hydraulic trainients in liigh-Temperature Gas-Cooled Reactors (IITGRs). The available anodules simulate pressurized or depressurized core h:atup trarnients, heht trarnier to general extenor sinks or to specific passive Reactor Cavity Cooling Systerns, which can te air or water cooled.

Graphite otidation dunng air or water ingress can le matelled, includtog the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance, due to water ingress, and also for hypothetical no-scram scenarios. For most liTOR transients, which generally range over hours, a user selected nmiallration of the core in t z reornetry is used. Ilowever, a separate model of heat transfer in the symmetry element of each fuel element is al.so available for very rapid trarnients. This imxtel can be applied coupled to the traditional coarser t.

nodalization. This repon descrites the mathematical inmiets.

uwd in the cale and the snethod of solution it descrites the code and its various sub-clements. Details of the input data and file usage, with file formats, are p 'en for the cmle, as well as for several preprocessing and postprocessing options. The T11ATCH mmiel of the currently applicable 3$0 MWm reactor is descrited. Input data foi four sarnple cases are given with output available in fiche form. Irutallation requirernents and cale limitatiora, as well as the most common error indications, are listed. 31 refs.,

23 figs.,32 tabs.

Publication Date:

Octoter 1991 Prepared by:

Kroeger, P.O.; Kennett, RJ.; Colman, L; Girubcrg,"J. (Brookhaven National Lab., Upton, NY)

Prepared for:

NRC Division of Regulatory Applicatiota, Office of Nuclear Regulatory Research Keywards:

air, comtation products, computer program docurnentation, graphite, beat trarsfer, HTGR.

type reactors, hydraulics, mathematical inmlels, oxidation, tractivity, reactor cooling systerns, TilATCll, thermal analysis, transierus, watet l

l 43 NUREO/BR--(K)S3, Vol,7

NUREG/CR. 5623 ORNL/TM.11644 MELCOR

Title:

IlWR Mark 11 Ex Vessel Corium Interaction Analyses Descripilort:

This repon descrites the results of c series of studies conducted to investigate the lehavior of core debns within a BWR Mark 11 conttinment.*Ihese stuies focused on the interaction of core debris with concrete and stcel structures (downcomers and in pedestal floor drains) within the dry,vell, the transport of debns through these drains and downcorners into the wet well, and on debns-wates reactions within the wetwell. Estunates of the conditions under which debns would lenetrate the in. pedestal drain lines, the time-depernic nt lehavior of the debns wittun the dram bues, and the amount of debris which might enter tie suppression pool via these drain lires are pro' tided. An assessment of the corxhtions under u hich the upperlip of the downcorners would te expected to fail (i.e., meft) due to exposure to hot core debns is presented. Finally, the ursque characteristics of debris-water interactions in Mark 11 containments are discussed, the existing knowledge base irgarding core-concrete debris-water interactiota is summarized, and an evaluation of the applicabihty of the MELCOR 1.80 code's debns-water interaction model to BWR Mark II's is presented.

Publication Date:

Novemtet 1991 Prepared by:

Greene,S.R.:1.evin A.E.;Hyman,C.R.tSoter A.;Taleyarkhan R.P.(Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Safety issue Resolution, Office of Nuclear ReFulatory Research Keywords:

blackouts, BWR type reactors. clernicai reactions, concretes.contairunent. corium.cxplosiors, fluidflow, fuel-coolantinteractions,heattransfer bydraulics,MELCOR, meltdown,rcactor safety, steam, water

. NUREG/BR--0083, Vol.7 44

SAND. 90 2339 NUREG/CR. 5630 CONTAIN

Title:

pwu Dry Contaimneni parametiic stuches Descriptlon:

suiry was used as a representative dry containment plant foi tie evaluation of por.sible ways that containment performance could be irnproved. Sensitivity - udies using the NUREO.

1150 models and inethodologies were used to estimate the ;ducuon of risk potentials resulting hom bypass scrubbing and DCil paitial depressuritation. These studies showed that the greatest reduction of risk occurs when bypass releases are mitigated by scrubbmg or puvented. Iligh-pressure DCllis also important. The CONTAIN code was used to esumale reduction in peak contamment pressure resulting from mitigation actions including venting, partial depressurization,inerting andigniters. Specifically,tle reductiors were approximately 2 bar with early depressurization and approximately 3 bar with igniters. Limited studies of the tenefits of venting and inerting were inade, but additionalinvestigations are needed to complete thir, area ofinvestigation. A brief discussion regarding concepts to mitigate the consequences of bypass is presented. CONTAIN code calculatiora were performed to investigate the possible overpressurization of the containment for the station blackout scer.r.rio. 30 re(s.,24 figs.,17 tats Publication Date:

April 1941 Prepared by:

Oido, R.O.t Williams. D.C.; Gregory, J.1 (Sandia National Lats, Albuquerque, NM)

Prepared for:

NRC Division of Safety Issue Resolution. Office of Nuclear Regulatory Research Keywords:

aerosols, blackouts, computes calculatiom, CONTAIN containment, depressurization, fission-product releaw, fission products, health hazards, hydrogan, neoplasms, pararnetric analysis, pWR type react ors, t adioisetopes, reador safety, recommendations, risk assessment, sensitivity analysis, source tenns, Surry 1 reactor, Surry 2 reactor, Surry.3 reactor, Surry-4 reactor l

l l

l 45 NUREG/BR -0083,Vol.7 l

1

NUltEG/ Cit-5648 OltNIJTM-11686 ENDFlll Y Data File,Itevised ENDF/Il-Y Data File, ENDF/Il VI Data File Tille;

'Ir nspon Calculations of NeutronTransnunionThrough Stecl Using ENDF/lbV.Revbed ENDF/lbV, and ENDIVD-VI hon Evaluatmra

==

Description:==

Tim ENDF/lbVI evaluated nuclear data file was recently tricased by the US National Nuclear Data Center dunng 1990. Arnong the rnost eagerly awaited new cross section evaluations m tius data collecuon are thor,e for the natural iton isotopes, due to their irnportance in nuclear systems analysis aral be:ause the pree.ous ENDF/B data (version V.

v,hich was released in 1979) are known to uralerestirnate die tratarnission of fast rieutrons through steel structures s udi as reac tor prtr sure vessels and uxh arion shieldmg, in this paper, a companson is made of results obtaired f rom neutron trarnport calculations perfonned with these 1wo ENDF/B ve rxions (V arx1 VI) ofiron data as well as an inte rmediate, revised version V evaluation that was proposed in 1986. Several different response parameters that are seraitive to high energy neutrora are examined for a variety of geornetrical configurations arnt source sgretra. It is 1ourw! that the two rrwer iron evaluations substantially increase the transmission of high eterFy neutrons through steel components with an incident fission spectrum source. Preliminary estimates irxbcate that the version VIiron evaluauon will considerably improve the agreernent between calculations and expenmental dosimeter measurements used in bght water reactor prer,sure vessel fluence analysis.

• Die calculated leakare spectrum of D-T fusmo neutrons from aniron sphere is also stuproved for energies above 4 MeV, but large mscrepancies with the measured spectrum are stdl observed at lower energies. 30 refs.,12 figs.,11 tabs.

Publication Dalel April 1991 Prepared by:

Wuluuns, M.I.; Aboughantous, C.; Asgari, M. (Louisiana State Univ., Baton Rouge, LA.

Nuclear Science Cenier); Wtute, J.E.; Wnyht, R.Q.; K am, F.D.K. (Oak Ridge N ational Lab.,

TN)

Prepared for:

NRC Division of Engineering. Office of Nuclear Regulatory Research Keywords:

Arkansas 1 reactor, carbon steels, cross sections, discrete ordinate method, dosimeters, embrittlement. ENDF/lbV data file, ENDF/B-VI data file, evaluated data, fis.sion products, iron, mmhfications, neutron fluence, neutron leakage, neutron spectra, neutron transport, neutrora, nuclear data collections, pressure vessels, PWR type reactors, reactor vessels, revised ENDF/lbV data file, steels, thickness, transmission

(

NUREO/BR-0083, Vol.7 46

ORNL/TM.11692 NUREG/CR. 5651 i

General 1

Title:

An lavestigauon of Crack Tip $ tress Field Criteria of Pinhaing Ocavage-Crack initiation 1

1 1)tscriptlon:

Deavage-crack initiation in large-scale wide plate (WP) r.pecimens could not le accurately 2

predicted itorn sm all. compact ( CT) specimera by using a line ar-clasne fracture.tnechanics, Ku,inctlalology,inthe wide platetestsconductedbythelleavy SectionSteelTechnology Program at Oak Ridge National Laixiratory, crack initiation has cornistently occurred at stress intensity (K ) values ranging frorn two m four times those precheted by the CT t

specimens. Studies were irutiited to develop crack tip stress field enteria incorporating efrects of georoetry, sire, arid constraint that willlead to improved predictions of cleavage initiation in WP specimera from CT specimera. The work centers around nonline at two-and three-dirnensional finite-elernent analyses of the crack-tip stress fields in these geometrier;.

Analyses were cortiucted on Cr and WP specimens for which cleavage initiation fracturc had been rneasured in taloratory tests. The local crack tip fields generated for these specimens were then used in the evaluation of fracture correlation pararneters to augment the Kt parameter for predicting cleavage initiation. Parameters of hydrostatic constraint and of maximum principal stress, measured volumetrically, are included in ibese eyaluations. The results suggest that the cleavage irdtiation process cari be correlated with the kical crack tip__

fields via a manmum principal stress criterion based on acideving a critical area witidn a critical stress contout. This criterion has tren successfully applied to conelate cleavage initiation in 2T CT and WP specsmen geometries. 23 refs.,16 figs., $ tabs.

Publication Date:

September 1991 Prepared by:

Keency Walker, J.; Dass,11.R.; Landes, J.D. (Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research i

Keywords:

crack propa g ation, finite c tement method, fracture rnechanics. fract ure properties. inechanical properties, plates, pressure vessels, steel ASTM-A533, strerae:

47 NUREG/BR -0083, Vol.7

NUREG/CR 5656 PNL-.7510 EXTRAN

Title:

EXTRAN: A Computer Code for Dtimating Concerurations of Toxic Substr.oces at Control Room Air intales Descripilon:

This repon presents the NRC staff with a tool for assessing the potential elfects of accidental releases of radioactive materials and toxic substances on habitability of nuclear facility control rooms. Tte tool is a cornputer code that estimates concentrations at nuclear facility controt room ais intakes given infoi mation about the rclease ard the envirotunental cerxhtions; The name of the computer code is EXTRAN. EXTIUW combines proctr*.utes for estimating the amount of airborne snaterial, a Gaussian puff dispersion, model, arxl tie most recent algonthms for estirnanng diffusion coefficients in ht11 ding wakes. II :s a modular computer code, wntte n in FORTRAN 77,that mru on personal computers. It uses a m ath coprocessor, if present, but does not require one. Code output may te directed to a printer or disk files.

25 refs.,3 figs.,4 Iabs.

Publication Date:

March le?!

l Prepared by:

Ramsdell, J.V. (Pacific Northwest Lab., Richland, WA)

Prepared for:

NRC Division of Safety issue Resolution, Office of Nuclear Regulatory Research Keywords:

air flow, computer prog *am documentation, control rooms, ecological concentration, craironmental transport, EXTRAN, FORTRAN, gaseous wastes, Gaussian processes, he alth hazards, indoor air pollution. liquid wastes, m athematical models, radioactive clouds, radioactive materials,spaa &peryknee, spatial distribution, theoretical data, time deperxlence, ioxic materials, ver.;ilation systems s.

i l

NUREG/BR-0083,Vol.7 48 g

l'NL 7513 NUltEG/ Cit. 5658 g.gigali2

Title:

ITIT 2: A Code for Followmg Airbome F stion Pro 3 acts in Genenc Nuck at Plant Flow Patto

==

Description:==

In order to assure that a nuclear power plant control room remens habitute dunnt cena n types of postulated accidents, Pacific Northwest Latuatory (PN1J bas urdertateu a special study for the US Nucleat Regulatory Commission. The purpose of this stt:dy is to develop softwan that can aid in the analyses of control roorn habuability dunng accidents in which airborne fission pnducts could challenge internal air pathways to the cotit rol room. PNL has

,:ompteled an frutial version tITIT) ard fittal version (ITIT 2) of a software package that can estimate the unsteady state invasion of quantities of fission puducts into the control room or any other destination within the nuclear plant via generic internal flow paths. Ttus repon consists of three pans: Section 2.0, Technical Bases, describes the flow path components ard mecharusms of natural fission-puduct deposition; Section 3.0, FPIT 2 Ctde Description, describes code organir.ation and the functions of the subroutines; and Section 4.0, Code Operation, discusses detatis of input requirements, code output, and a sample case demonstration. *Ihe apperdices consist of an ITIT 2 FORTRAN code listing, a listing of a code for iniilding input files, famis for buildmg input files, and the sample case input and output files. 7 reis.,3 figs.

Publication Date:

March 1991 Prepared by:

Owcanki, P.C.: Burk, K.W.: Ramsdell, JN.; Yasuda, D.D. (Pacific Northwest Lab.,

Richland, WA) l'repared for:

NhC Division of Safety issue Resolution, Office of Nuclear ReFulatory Research Keywords:

B atte11e Pacific Nonhwest Laboratories. controt rooms. deposition, diffusion, environn.cnt, fission product release, fission products, flow rate,ITFP 2, human iactors, nuclear pow et plants, programming, radioactivity transpon, reactor accidents, reactor safety US NRC, ventilation systems 49 NUREG/BR.-0083. Vel.7 i

l

NU: LEG /CR. 5662 IINL-NUREG 52271 CONTAIN Title-liydroren comtwsuon, Control, and value.bnpact Analysis for PWR Dry Contairunents Descriptlont flydrogen issues appheable to PWRt. with dry contairunent desigra are reviewed based on cunaginfonnatin1frornticNRC'ssevereaccidentresearch egum.Additionalcalculatio;ts l

were perionned uting the CONTAt.N code for a multi-computment imdel of the Zion plant.

The review includes in vessel and ex vessel hydrogen generation, tirne and modes of hydrogen release, bydrogen mixing and transport in the contamment, hydrogen combustion snechanistn, hydrogen control methods and tie equipment survivability. A cost-lenefit analysis of the hydrogen ignition system was performed for de 7jon and Surry plants.

Potendal for hydrogen detonation in these plants was evaluated. 47 figs.,36 tabs.

l'ublication Date:

June 1991 I'repared by:

Yang, J.W.; Musicki, Z.; Nimnual, S. (Droolhaven National Lab., Upton, NY)

I'repared for:

NRC Division of Safetyissue Resolution Office of Nuclear Regulatory Research Keywords:

combusdon, combustion propernes, CONTAIN, containment, containment systems, cost--

benefit analysis, detonatiota, hydrogen, PWR type reacton, reactor accidents, reactor components, reactor core disruption, Surry.1 reactor, Suny.2 reactor Zion 1 reactor, Zion 2 reactor NUREO/BR. 0083,Vol.7

$0

EGG-2633 NUREG/CR. 566'4 RELAP5/ MOD 3

Title:

ret.AP5 Thennal-llydraulic Analysis of the WNP1 Pressuriied Water Reactor.

Revismn 4 Descriptlon:

nermal-hydraulic arealyses e t nve hypothetical accecni scencios were perfonned with the REl>> P5/ MOD 3 co uputer code for the llabcod; & Wilcox Company Washington Nuclear hoject Unit 1 (WNP1) pressurized water reactor. *Ihis work was sporaored by the US Nt. clear Regulatory Comrnh.sion (NRC) and is being performed in conjunction with future a,nalysis work at the NRC Technical Training Center in Chattanooga, Tennessoc. *Ihe accident scenanos were chosen to assess and benchmark the therrnal-hydraulic caphbihties of the Technical Training Center WNP1 simulator to model abnonnal trainient conditkirs.

6 refs.,45 ngs.,9 tabs.

l l'uhlicalloti Date:

May 1991 l

l l'repared by:

Martin, R.P. (EO& O idaho, Inc., Idaho Falls. ID) l'reparcd for:

NRC Dwision of Systems Research, Office of Nuclear Regulatory Research Keywords; blackouti,computercalculatiorn computeriredsamulation,heattransfer. hydraulics. reactor accidents, re actor coohng rystems, teactor safety, RELAP5/ MOD 3, WNP.1 reactor 51 NUREO/BR.-0083, Vol]-

NUREG/CR 5667 EGG 2634 MACCS

Title:

INEL Pen,onal Cornputer Version of h1 ACCS 1.5

==

Description:==

The hiELCOR Accident Consequence Code System, Version 1.5 (hiACCS 1.5), calculates potential consaquences tesulting irom atmaspheric eleases of radioactive mnicria'. Sarda National Laboratories developed the code for the US Nuclear Re gulatory C snission on a VAX/VhtS mini computer.This report documents the Idale National Engirrering Laborcry conversion of bl ACCS 1.5 for cempilation and esecutica on ardO386 base <1 IBM or IDM-cornpatible personal comranter TC). Tte resulting PC version of the code is available through the Energy Science and Technology Sof tware Center.P.O. Box 1020, Oa1. Ridge,TN 37831.

2 refs., I fig., I tab.

Publication Date:

March 1991 Prepared by:

Jores, K.R. Doble, C.A.t Knudson, D.L (EO&O Idaho, Inc., Idaho Falls,ID)

Prepared for:

NRC Division of Systems Rescuch, Of5cc of Nuclear Regulatory Research Keywords:

ANL compuwr codes, computedred simulation, corium, DEC computers, heat transfer,

.I hydraulics !!!M computers, Idaho National Engineering Laboratory, hiACCS, meltdown, mahfications,personalcomputers, programming publichealth,quahtyassunmce, radiation hazards, tadioactivity traruport. ne actor accidents. reactor core disniption, Sarda taboratories, source terms, US NRC NUREO/BR--0083,Vol.7 52

r ORNL/TM 11743 NUREG/CR 5668 i

CORSOR.M, ORIGEN2, VICTORI A

Title:

Data Summary Repon for Fission Praluet Release Test VI $

Descriptlon:

Test VI-5, the fifth in a senes of high temperature fisrion product release tests in a vertical test apparatus, was corducted in a flowiry mixture of hydroyen and helium. The test specunen was a 15.2 cm long se etion cf a fuel tod from the liR3 scactor m Belgium which hacibieninadidled to a buruup of approxw.tely 42 MWWkg Using a hot cell-mounted test apparatus, the fuel rod was tested in an induction fumare uruler simulated LWR accident conditions to two test temperatures,2000 K 1o-20 min and then 2700 K for an albtional 20 min The released fission pnxtucts were collected in three sequentially operated collection trains on components designed to measure tission-product transport characteristics and facibtate sampling and analysis.The te.iutts from trus test were compared with those obtained in previous tests in this senes and o. h tla CORSOR M and ORNL diffusion nelease rncdels for fission product release. 21 refs.,19 figs.,12 tats.

l'ublication Date:

October 1991 l'repared by:

Osbome, M.F.; Lorenz, R. A.; Tra vis, J.R.; Wetster, C.S c Collins, J.L (Oak Ridge National Lab., TN)

I'repared for:

NRC Division of Systems Research Office of Nuclear Regulatory Research Keywords:

BR 3 reactor, bumup, computerized simulation, CORSOR M. deposition, fission-product release, fission products, fuel elements, fumaces, gamms spectroscopy. ORIGEN2, post.

irradiation cramination, quantitative chemical analysis, raunisotopes, reactor accidents, stainless stects, temperature gradients, test facilities, vicrORIA. water cooled reactors l

l l-53 hUREO/LRu0083,Vol.7 1

NUREG/CR-5670 EPRI.NP--7165 RELAP5, TRAC

Title:

Multilaop Integral System Test (MIST): MIST Facihty Functional Specification

==

Description:==

The Multiloop Inierral System Test (MIST)is pot of a multipinse program staned in 1983 to address small break loss-of-coolant accidents (S BLOCAs) specific to Babcock & Wilcox designed pl ants. MISTis sponsored by the US Nucleat Regulatory Commission, the B attock

& Wticox Ow r.ers Group, the Elecuic Power Research trutitute, and Babcock & Wilcox.The unique iestures of the Babcock & Wilcon design, specifically the hot leg U-tends andste am p nerators, prevented the use of e xisting integral system datn or exisung integn 1 facilities to address Oc thennal-hydraulic SBLOCA questions. MIST was specifically destpwd and constructed for this program, and an existing facility--the Once Through Integral System (OTIS)-was also used. Data frorn MIST and OTlS are used to benchmark the adequacy t' system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. 'Ihe MIST Functional Specification documents arbuilt design features, dimensions, instrumentation, and lest approach it also presents the scaling basis forthe facility and serves todefine tir scope of work ior the f acihtydesign and c(natruction.13 reis.,112 figs. 38 tabs.

Publicallon Date:

April 1991 Prepared by:

liabib T.F.;Koksal C.O.;Moskal.T.E.; Rush,0.C.;Oloudemans,LR.(Babcock & Wilcox Co.,Lynchburg,VA and Alliance 011)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research: Electric Power Research hisutute; Babcock & Wilcox Owners Group Keywords; auxiliarysystems computerizedcontrolsystems,computerizedsimulation,dataacquisition, experunent planning, feedwater, fuel elements, heat transfer, hydraulics, loss of coolant, pumps, PWR type tractors, reactor components, reactor cooling systems, reactor safety, reactor vessels, RELAP5, suptures, scale models, specificatiora, steam generatots, test facilities, TRAC, transients, tules, two phase flow, valves, water chemistry NU!LT M h3, Vol.7 54 P

REMIX. NEWMIX NUltEG/CR 5677 TitL:

A Unified Interpretatmn of One-Fifth to Full 5cale 'lksmal Miting Expenments Related to Pressurir.ed Thennal Shack Descripilon:

Therm al minng in s e lation t o Pre ssu riecd Therm al S h<d. has bee n exarnined e xpe rimentally throughout the world in a variety of scales These include the CREARE 1/5, the TVO/

IVO(NRC)-2/5,the PURDUE(UCSD Fl/2,the CREARE.1/2 the llDR.1/l, ard the UPTF.

1/1 test iacibues. 'Ihe Regional Miung Model and the associated computer programs REMIX and NEWMIX are used to interp:et ther e data,in this report,in a comprehensive fashien These interpretations indicate that cooldown transients and degree of stratification can te predicted with confi&nce. Univen.al stratification solutions are also provided, in graplucal fomi, and a simple procedure for hand-calculation is also described. 40 refs.,

20 figs.,17 tabs.

l'ublication Date:

April 1991 Prepared by:

'theofanous. T.G.: Yan. H. (Califomia Univ., Santa Barbara, CA. Dept. of Chemical and Nuclear Engineering)

Prepared I'or:

NRC Division of Systems Research, Office of Nuclear Regulatory Reseach Keywords:

BWR type reactors, entrainment, fluid fiow, Duid mechanics, heat transfer, high pressure coolant injection, hydraulics, loss of coolant, mathernatical models, rniting, NEWMIX.

PWR type reactors, react or coohng systems, reador safety, REMIX, rese arch reactors, scale models, scaling laws, stratification, test facilities, thermal, hock 55 NUREG/BR 0083,Vol,7

NUREG/CR. 5681 IINL.NUREG 52280 FEMWATER Tille:

low-level Waste source Term Mc let Development and Testing Descripilon:

The low-level waste source tenn model deselopment project has adapted / developed two computer codes to predict the rnigration of radionuclides emplaced in shallow land terial fadhties. The computer code FEMWATER h med to predict water flow and moisture content. The computer code BLT is used to predict container Dreach, waste form leaching, and contaminant Transport. Recent work on this project focused on two are as. One involved improvements to the teaching models incorporated in BLT. Inparticular, this report descrites an ailitional model that wat added to BLT which simulates the waste form using the rnethod of finite diffe ences and treats the contacting solution as a mixing bath.This modelimproves upon the prevmus moods in BLT in three areas:(a)it treats the release processes of diffusion, dissolution, and surface rinse simultaneously;(b)it allows for partitioning between the waste form ain! solution; and (c)it pennits solution feedback effects toin0uence diffusive releases.

Venfication studies of the finir differenec/ mixing bath model are discussed in detail. The second area of research involved companng BLT rnodel predictions to experimental data.

This report presents the results of rnodeling laboratory scale wet / dry cycle teach experiments and lysimeter experiments conducted at Pacific Northwest Laboratories. Based on this modeling work, recommendations for future areas of study are given. 23 refs.,19 figs.,

15 tabs.

1 Publication Date:

May 1991 Prepared by:

allivan, T.M.: Suen, CJ. (Brookhaven National Lab., Upton, NY)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

analytic functions, benchmarks, BLT rnodel, cements, cesium, diffusion, dissolution, experimental data, FEMWATER, finite difference method, ground release, leaching, low.

level radioactive wastes, radioactive efnuents, radioactive waste disposal, radionuclide.

migration, recommendations, source terms, underground disposal, waste forms NUREO/BR--0083,Vol.7 56

-.. ~

NUREG/CRw56M FLUMPS

Title:

Analyses and Field Tests of the liydraulic Performance of Cement Grout Borehole Seals Desciiptlon:

Three tests are presented ard analyred in detail for detenntning the hydrau!!c properties of torehole seals as applicable to disposal of high level radioactive wastes. Two consist of rnonitonng the rate ofinjection of water at comtant pressure into an injection tone at one end of a r, cal and mordtoring the collection rate or rate of flovr into a free-draining collection zone at the other erd The third test is performed by shutting in the collection rone and morutoring the buildup in hydraulic head. One-dimensional ard adsymmetric three dimemional flow models are presented for analyztrig test results. In the one-danensional rnodels, the sealis assumed to le a homogeneous and isotropic porous mediurn, and the rock is assumed to te impermeable. In the axisymmetric models, the seal and the sunounding rock mass are taken ashomogeneousandisotmpicporousmedia Theequationforsaturated.confitedgmurdwtuer flow is assumed to apply. The hydraulic propenies of the se al are expressed by its hydraulic conductis ity and specific storage. in the axisymmetric mtdels,the comiuctivity and specific storage of the rock mass are included in the formulation. A fourth test, a tracer travel time test,is presented as a means for detecting any Idgh velocity flow path through or around the r,eal. Detailed and specific recommendations are given for corducting borehole seal tests. In pnnciple, these methods also thould te applicable to testing Gaft seals and tunnel dams. In -

prac ice, complicatiom will be encountered for the implenken.ation of the tests on a much larges scale, but these complications should te sesolvable,154 refs.,79 figs.,38 tats.

Publication 1) ate:

April 1991 Prepared by:

Orcer, W.B.: Daemen, J.JK (Arizona Univ., Tucson, AZ. Dept. of Mining and Geological Engineering)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

toreholes.tourdtryconditions, cements,computercodes computerizedsimulation. drainage, environmental exposure pathway, fission products, flow rate, FLUMPS, geology, ground water, grouting, high-level radioactive wastes, hydraulics, hydrology, one-dimemional calculations, perfonnance, porosity, radioactive waste disposal, radioactive waste facilities, radioactive waste management, radionuclide migration, research programs, rock mechanics, Sandia laboratories,scals, steady statecornbtions,three-dimensionalcalculations,timedependence, transients, verification, WIPP l.

r 57 NUREO/BR -0083 Vol.7 1

1 i

NUHEG/CR. 5701 SAND 910539 DCM3D, NEITRAN 2 4

Title:

A Perfonnance Assessinent Methodology for liigh level Radioactive Wasie Disposal in Unsaturated, Fractured Tuff Descripilon:

Sandia National Laboratories has developed a methodology for perfonnance assessment of deeppeolt daposal of high level nuclear waste. The applicabihty of this perfonnance assessmet. aethmlology has h en demonstrated for disposalin ledded salt and basalt; it has since teen mahfied for asrsment of tepositories in unsaturated, fractured tuff. Changes to the me%dology are primarily in the fonn of new or rncdified groundwater flow and radionu 'ite transport codes. A new computer cale, DCM3D,has teen developed to mmici j

three tb ansional groundwater flow in unsaturated, fractured rock using a dual continuum appr:

The NEITRAN 2 code has been developed to cificiently roodel radionuclide trt _, or n time dependent velocity fields.has the abibty to use extemally calculated port velocities and sat urations, and includes the e ffect of sat uration deperxlent te tardation iactors.

In order to uw these codes together in ;wifonnance-assessment type analyses, cale-coupler programs were developed to translate DCM3D output into NEFTRAN 2 input. Other portions of the perfonnance assessment methodology were evaluated as part of mmlifying the methmlology for tuff, The scenario methodology developed under the ted3ed salt program has teen applied to tuff. An ir:vestigation of the applicability of uncertainty and sensitivity analysis techniques to non linear models indicates that Monte Carlo girautation remains the most robust technique for these analyses. No changes have teen reconunended for the dose and health effects models or the biosphere transport inodels. $2 refs., I fig.

Publication Date:

July 1991 Prepared by Gallegos, D.P. (Sandia National Lats., Albugaerque, NM)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regtilatory Rexarch Keywords:

biosphere, DCM3D, fluid flow, groundwater, health hazards, high level radioactive w astes, Monte Carlo method, NEFTRAN 2, perfonnance testing, radionuclide migration, time dependence, tuff, t irxlerground disposal, Yucca Mountain NUREO/BR-4083.Vol.7 58

NUlmG/CR-5711 General

Title:

Assessment of Unce rtainties in Measurernent of pliin llostile Envirotonents Otaracteristic of Nuclear Repositones

==

Description:==

This rtpon focuses on evaluation and characteristics of sputtered thin fitrn pli electrodes which can te used to assess the corrosivity of hat (100*C) aqueous solutions present in nuclear repositonn Sputtered thin filrus have the advantages of high temperatury capabihty, ruggedness, and low cost. The iridium oxide films were found to have a linear,58..N/pli, response to changes in pil. They had little hysteresis but drifted approxirnately 0.2 V over a period ofIwo days cxposurr to pli 2-12 solutions. The films were found tole insensitive to inter fen nce f rom most ionuuch as allali ions but had redox sensitivity to ferri /ferrocya' tide ions. Ahhough special surface treatrnents were needed for the films for good adherence at 200 C, the filtns were not degraded after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> exposure at pli 4,7 and 10 at 26, J.

R utleru um o xule sputten ed filins perf ormed equally well to the iridium othie films in parallel tests. The report also contains information on electrochemistry and testing of thin film electrodes and the characteriration of the thin fihus by x ray photoemission spectroscopy, uluaviolt t phottemission specuoscopy, andinn scattenng spectroscopy 123 refi.. 29 figs.,

7 tals.

l'uhlication Date:

Octoler 1991 l'repared lpy:

Kreider, K.G.; Tarlov, M.J.: liuang, P.ll. (National Inst, of Standards and Technology, Gaithersburg, MD) l'repared for:

NRC Division of Engineering, Office of Nuclear Reguir

' arch Keywords:

corrosion, design, electrochemistry, electrodes, Gouy Chaptan-Stern-Grahame mtxiel, hightemperature, hysteresis,indiumoxies,ironcanplexes mathematicalmodels, measuring methods, medmm temperat ure, pe rform ance testmg, pil value, Poisson-11oltun ann equ at ion, radioactive waste facilities, redox potential, ruthenium oxides, thin filtns 59 NUREG/IlR 4083,Vol.7

r NUREG/CR 4712 ORNUTM 11823 MORECA

Title:

MORECA: A computer ccde for simulating Modular liighTemperatuie Gas-cooled Reactor Core lleatup Accidents Descriptlon:

The de sign featur, of the m(dular high temperature gas-cooled reactor (MitTOR)have the potentittl to inake it essentially invulnerable to damate from postulated core heatup accidents. This report descrites the ORNL MORECA code, which was developed for analyring postulated long tenn core heatup scenarios for which active cooling systems used to remove ahetheat following the accidents can te assumed to te unavailable. Simulauons oflong-termloss of forced-convectionaccidents,bothwithardwithoutdepressurizationof the pnmary coolant, have shown that maumom core temperatures stay telow the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have teen done to detennine the effects of enors in the predictions due both to ureertainties in the muling ard to the assumptions about operational paramet ers. M ORECA mociels the US Department of Energy reference design of a standard M11 TOR, Publica!!on Date:

Octoler 1991 Prepared by:

Ball, SJ. (Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Regulatory Applications, Office of Nucler.r Regulatory Research Keywords:

after-heattemoval.depressurization.trattransfer,llTORtypereactors hydraulics,MORECA, re rtc,r accide nts, re actor cooling systems. rs actor safety, safety aaalysis, sensitivity analysis d

NUREO/DR-0083,Vol.7 60

)

I

S AND-91-0835 NUREG/CR--5715 CONTAIN

Title:

Reference Manual for the CONTAIN 1.1 Code for Containment Severe Acci.ient Analysis

==

Description:==

Riis tepon describes the phenomenological equations and the numerical procedures used by the CONTAIN 1.1 code to determine the conditior;s within nuclear power plant contairunent dunng a severe accident. 'Ihe CONTAIN detailed modi ovide the capabihty to i

mechanistically calculate '6e containment intemal thermalm, raulic corxhtions and tie amount of radioactive matter that wo.ld be released to the environment if there were a leak from the containment. Note that the CONTAIN models can be venfied by comparing the code calculauons to experimente' esults. The models descriled include those to account for the flows o. mass and energy ictween contairunent compa'tments, th: exchange of energy letween the atmosphere and heat stmetures,the tierrne varnic conditions,the distributions of aerosols, the decay and transpon of fission prod..

kilagration of hydrogen and carbon monoud,, boiling water reactor suppressic:4 avior, and engineering safety features,includmg a spray, fan cooler - and an ice cc.nenser.These models are solved with implicit coupling, where appropriaic, to obtain a stable and computationally efficient solution 52 refs.,36 figs.,9 tabs.

Pitblication Date:

July 1991 i

Prepared by:

Washington, K.E.; Murata, K.K.; Gido, R.O.; Gelbard, F.; Russell, N.A.; Billups, S.C.;

Carroll, D.E.; Grif0th, R.O. (Sandia Na'ional Labs., Altmquerque, NM); leuie, D.LY. (Los Alamos Technical Associates,Inc., Albuquerque, NM)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

aemsols, carbon monoxide, combustion, concretes, CONTAIN, containment, containment systems, conium, engineered safety systems, fission-product release, fission products,leat tians fer, hydraulics, hydrogen, ice condensers, nuclear power ptants, radimetive effluents, themiod>rties 61 NUREG/BR--0083, Vol.7

NUREG/CR--5716 BIGFLO, VAM2D, TH'1ST, TRUMP, UNSATII, UNSAT2, TRACER 3D, PORFLO3

Title:

Model Validatior; at the Las Cruces Trench Site

==

Description:==

A series of dynamic field experiments have been performed at the las Cruces Trench Site to provide data to test deterministic and stochastic models for water flow and solute transpon in spatially variable unsaturated soils. Two expenments were performed to provide suppon for model validation effons during Phase 1.of thTRAVAL, and a third experiment is currently wxier way to suppon the INTRAVAL Phase 2 effons. *Ihe third experimern utilizes different boundary and initial conditions and additional chemical tracers. The data from tie third experiment along with model predictions from several modeling groups will be used to test moils for water flow and solute transpon during infiltration and redistribution. This report summarizes the 1.as Cruces Trench Site model validation effons atxt presents the INTRAV AL Phase 2 validation plans.The Phase 2 validation strategy is discu+ sed in detail.

29 refs.,1i figs.,4 tabs.

Publication Da,.

June 1991 Pr: pared by:

IIllls, R.O. (New Mexico State Univ., Las Casces, NM. Dept. of Mechanical Engineeringh Wierenga. P.J. (Arizona Univ., Tucson, AZ Dept. of Soil and Water Science)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

animals, BIG FLO, environmental trarr:pon, fluid flow, groundwater, by draulic conductivity, low-level radioactive wastes, m athematical models, New Me xico, plants, PORFLO3, porous materials, ratioactive waste facilities, radionuclide migration, hite characterization, soils, solutes, TRACER 3D, TRUMP, TRUST, UNSAT2, UNSATH. VAM2D l

NUREG/BR--0083, Vol.7 62 2

MCS--910401 NUREG/CR--5729 l

TACMVS, CAR '. TAC

Title:

Multivariabh Modeling of Pressure Vessel and Piping J-R Data Descriptlon:

Multivariable models were developed for tuedicting J.R curves from available data, such as material chemistry, radiauon exposure, temperature, and Charpy V-notch energy. The present work involved collection of pt.blic test data, application of advanced pattem recognition tools. and calibration ofimproved multivariable mcdels. Separate models were fitted for different material groups,includmg RPV w elds, Linde 80 welds, RPV base metals, piping w elds, pipic; base me t als, and the combined database.Three different types of models were developed, involving different combinations of variables that might be available for applications: a 01arpy model, a pa irradiation Charpy model, and a copper. fluence model.

In general, the best results were obtained with the preirradiation Charpy model. The copp'r-fluence modelis recornmended only if Charpy data are unavaitable, and then only for Linde 80 welds. Relatively good fits were obtained, capable of predicting the values of J for pressure vessel steels to within a standard deviation of 13--18% over the range of test cata.

The models were qualified for predictive purposes by d'monstrating their ability to predict validation data not used for fitting. 20 refs.,45 flgs.,16 tabs.

Publication Date:

May 1991 Prepared by:

Eason, E.D.: Wright, J.E.: Nelson E.E. (Modeling and Computing Services, Newark, CA)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

CART. Charpy tes, cracks. dat abase m anagement, fracture properties, inte grals, irradiation, y

mathematical models, r..wls, neutmn fluence, physical radiation effects, pipes, pressure i

vessels, reactor safety, steels, surveillance, TAC, TACMT _,, welded joints at\\

l s

(

i

NUREG/CR--5740 ORNL/RSIC--49/R1 PALLAS, ADJMOM-I, AC"IT, PIlOTX Data Library, NIST Database

Title:

New Gamma-Ray Buildup Factor Data for Point Kemel Calculations: ANS-6.4.3 Standard

.1 Reference Data

==

Description:==

An American Nuclear Society Standard.1 Committee Worting Group,ikntified as ANS-6.4.3, has developed a set of eyalu ated gamma-ray isotropic point-source buildup factors arx1 attenuation coefficients for a standard reference database. The largely unpublisted set of buildup factors calculated with the moments method has been evaluated by recalculating key valucs with Monte Carlo, integral transport, and discrete ordinates methods. Additional buildup factor data were obtained froin PAI LAS code results. Attention has been given to frequently neglected processes such as brenisstrahlung and the e ffect ofintroducing a tissue phantom behind the shield. The proposed draf t standard, provided as an appendix, contains data for a source energy range from 15 kev to 15 MeV and for 22 e;ements and 3 mixtures (wa,cr, a'r, and concrete). The buildup f acto: data are also represented as coefficients for the G-P fittirig function. Tables giving correction factors for multiple scattering in tissue are also provided. 26 refs.,4 tabs, Publication Date:

August 1991 Prepared by:

Tmbey, D.K. (Oak Ridge National Lab., TN)

Prepared for:

NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research Keyworos:

ADJMOM-1, air, annihilation, ASF1T, attenuation, bremsstrahlung, buildup, coherent scattering.computercalculations. concretes, database management,dtscrete ordinate method, elements, energy spectra, fluorescence, gamma radiation, materials, Monte Carlo method, NIST database, PALLAS, PHOTX data library, point kernels, indiation doses, response functions,shieldmg standards,Three MileIsland-2 reactor, tissues,transporttheory, water, xenon 133 NUREG/BR-0083,Vol.7 64

NUREG/CR-5743 TRUST

\\

Title:

Approaches to Large Scale Umaturated Flow in lieterogetrous, Stratified, and Fractured Geologic Media

==

Description:==

Tlus report develop a broad review and assessment of quantitative modchng approaches and data requirements for large scale subsurface flow in radioactive waste geologic repository. The data review includes discussions of controlled field experitnents, existing contamination sites, arri site-specific hydrogeologic corxhtions at Yucca Mountain. Local-scale constitutive models for tie unsaturated hydrodynarnic properties of geologic media are analyzed, with particular emphasis on the effect of stmetural tharacteristics of the medium. The report further reviews and talyics large-scale hydrogeologic spatial vanabihty from aquifer data, unsaturated soil dr.a. and fracture network data gathered from the hterature. Finally. various modeling strategies towardlarge-scale flow simulations are assessed, including direu high resolution simulation, arxl coarse-scale simulation based on auxiliary h,drodynamie models such as single equivalent continuum and dual-porosity continuum. 'Ihe roles of animtropy, fracturing, and broad-band spatial variability air emphasized. 252 refs.

Publicatien Date:

August 1991 Prepared by:

Ababou, R.

Prepared for:

NRC Division of Engmeering Office of Nuclear Regvlatory Research Keywords:

alpha beanng wastes, anisotropy, aquifers, computerized simulation, Darcy law, data analysis, dual absorption model, dual continuum model, flow models, fluid-structure interactions, Gauss-Markov process, geologic fonnations, geologic fractures, geologic surveys, groundwater, high level radioactive wastes, hydraulic conductivity, hydraulic fractures, hydrodynamic model, Koreny-Carmen formula, Navier-Stokes equauons, numerical analysis, porosity, radioactive waste dispoul, radionuclide migration, randomness, regulations, retention, soil mechanics, spatial detribution, stochastic processes, stratigraphy, subsurface structures, surface contan ' nation, TRUST, Van Genuchten.

Mualem model, Wang Narasimhan model, water satumtion, Yucca Mountain 65 NUREG/BR--0083 Vol.7

NUREG/CR-5757 ANL--91/25 1

L SMACS

Title:

verification of Piping Response Calculations of ShiACS Code with Data from Seismic Testing of an in-Plant Piping System

==

Description:==

The objective of this effort was to evaluate the piping analysis part of the ShiACS code for estimating the response of realistic piping systems subjected to seismic excitation, given as m ultiple. independent support acceleration histories. The e xperimental dat a from the seismic testing of an in-plant piping system at the HDR Test Pacility in Gennany were used for this pmpose. Of the six different support systems test ed, two were selected for the evaluation: or a " stiff" configuration containing both struts and snubbers, and the other a more flexible configuration with no snubbers. Described are th analytical mode'ing, calculations, and results of tin posnest simulation of two sets each for both suppon configurations, with excitations at 100% and 200/300% of safe shutdown-eanhquake loading. Almost all the calculated peak response quantities were sm aller (by different amounts) than the correspondm g ttst measurements. However, pipe displacements and bendmg stresses were better estimated than the pipe accelerations and suppon forces. The discrepancies are mainly attributable to the inability of the lineat analysis to model the nonlinear behavior of the VKL piping system, characterized by gaps in support connections and the friction at pipe clamps. A similar trend of underestimating test responses was observed in the hnear analyses performed by other investigatort 13 refs.,14 figs.,8 tabs.

Publication Date:

September 1991 Prepared by:

Srinivasan, hi.G.; Kot, C.A.: Hsieh, B.J. (Argonne National Lab.,IL)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

acccleration, bending.computercalculations computerizedsimulation, damping, dynamic loads, e xperimental data, nuclear power plants, pipes, response functions, restraints, seismic effects, ShiACS, strains, stress analysis, suppons NUREG/BR 0083,Vol.7 66

---

____-_-__--___mu-s

PNL--7723 NUREG/CR-5765 SPARC-87. SPARC-90

Title:

SPARC 90: A Code for Calculating Ftssion Product Capture in Suppression Pools Descriptlon:

This report describes the technical bases and use ofIwo updated versions of a computer code initially &veloped to serve as a tool for calculating aerosol panicle retention in boiling water reactor (BWR) pressure suppression pools daring severe accidents, SPARC-87 and SPARC-

90. The most recent version is SPARC-90. The initial or protoiype version was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth,both mechanistic and empirical models for vent e tit region scrubbing, specific models for hydrodynamics ofbubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC 90 (along with SPARC-87). A code description, code operating instructions, partial code listing. etamples of the use of SPARC-90, and summaries ofexperimental data comparison studies also support the use of SPARC 90. 29 refs.,4 figs.,11 tabs.

Publication Date:

October 1991 Prepared by:

Owczarski, P,C.; Burk, K.W. (Pacific Northwest Lab., Richland, WA)

Prepared for:

NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research Keywords:

aerosols, bubbles, BWR type reactors, computer calculations, computer program documentation containment.contairanentsystems. fission-productrelease,fissionpmducts, fx.attransfer hydraulics, hydrodynamics,iodme,mathematicalmodels,paniculates, reactor accidents, reactor safety, SPARC-87, SPARC-90, vapors 67 NUREG/BR--0083, Vol.7

' NUREG/CR-.5768 PNL-.7765 ICEDF

Title:

Ice. Condenser Aerosol Tests

==

Description:==

This report presents the results of an experimental investigation of aerosol panicle transpon and capture usirag a full-scale height and reduced-scale cross section test facility based on the.

design of the ice companment of a pressunzed water reactor (PWR) ice-cor>tnser containment

.jf system. Resnits of 38 tests included thermal-hydraulic as wc!! as aerosol particle data, Particle retention in the test section was greatly influenced by thermal-bydraulic and aerosol test parameters. Test-average decontamination factor (DF) ranged tetween 1.0 and 36 (retentions tetween approximately 0 and 97.2%). The measured test. average panicle retentions for tests without and with ice and steam ranged between DF = 1.0 and 2.2 atxi DF = 2.4 and 36, respectively. In onlet of ar, arent imponance, parameters that caused panicle retention in the test section in the presence ofice were steam mole fraction (SMF),

noncondensible gas flow rate (residence time), panicle solubility, and inlet particle size. Ice.

basket section concondensible flows greater than 0.1 m?/s resulted in staole thermal.

3 stratillCation,whereasilowsless than0.1 m /sresultedinthermalbehaviortermedmeandering with frequent temperature crossovers between flow channels.10 refs.,66 figs.,16 tabs.

Publication Date:

September 1991 Prepared by:

Ligotte, M.W.: Eschbach, E.J.; Wine gardner, W.K. (Pacific Nonhwest Lab., Richland, WA)

Prepared for:

NRC Division of Regulatory Applications, OfGee of Nuclear Regulatory Research Keywords:

aerosols, containment. containment s ystems, decontamination, fluid mechanics, heat transfer,.

hydraubcs, ice conderw.rs, ICEDF, panicle size, paniculates, performance testing, PWR type reactors, reactor accidents, reactor cooling systems, reactor safety, ruptures, steam, temperature measurement, test facilities NUREG/BR--0083,Vol.7 68

BNL NUREG--52295 NUREG/CR--5773 I

VAM2D, PRESTO, PAG AN 1

l

Title:

Selection of Models to Calculate the LLW Source Term

==

Description:==

Perfonnance assessment of an LLW disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facihty (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source tenn is influenced by the radionuclide inventory, the waste forms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid now, container degradation, waste form leaching, and radionuclide uansport). In tum, many of diese physical processes are influenced by the design of the disposal facility (e.g., infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. ' Utis document psovides a brief overview of disposal practices and reviews existing source term models as background for selecting appropnate m odels for estimating the source term.The selection rationale and the mathematical details of the models are presented. Finally, guidance is presented for combir.ing the inventory data with appropriate mechanisms describing release from the disposal facility.

44 reis.,6 6gs., I tab.

Publication Date:

October 19'll Prepared by:

Sullivan, T.M. (Brookhaven National Lab., Upton, NY)

Prepared for:

NRC Division of Low-1xvel Waste M anagement and Dec ommissioning, Office of Nuclear Material Safety and Safeguants Keywords:

BLT mtxlel, calculation methods, computer codes, containers, failures, Buid flow, leaching, low-level radioactive wastes, mathematical models, PAG AN, PRESTO, radioactive waste disposal, radioactive waste frilities, radioisotopes, radionuclide migration, source terms, VAM2D, waste fonns 69 NUREG/hR--0083 Vol.7

NUREG/CR--5778-Vol.1 General

Title:

New York /New Jersey Regional Seismic Network. Annual Report for April 1989-March u

1990: Volume i

==

Description:==

Lamont-Doberty Geological Observatory bDOO) continued operating a 31. station seismic network covering parts of Ne w York and New Jersey.'Ihe network is being transformed into sut>networkswithstationsradiotelemeteredto sman"reconhngstations.Tlesutwnetwort a

spproach is capable of providing improved data at reduced cost. The major research effort during the penod of this repon was centered about the Saguenay earthquake sequence in Quebec.1-DGO collaborated with the Canadian Geologic Survey in monitoring aftershocks with temporary local stations. Analysis of data from the 1985 Ardsley earthquake in Westchester county continued with a Green's function deconvolution approach to resolve the dimensions of the rupture of the main shock (Mb = 4.0) and of the largest aftershock (Mb =

3.0). 'lhe results corroborate the 1/2-I km diameter inferred for the rupture and suggest that the segmentation of the Doobs Ferry fault and of similar faults in the Manhattan Prong may be controlling the size ofhistanc earthquakes in the New York Cityregion. Finally,a ponable seismograph survey was camed out in Palco, Kansas, which showed clearly that seismicity at Palco was induced. $1 refs.,8 figs.,2 tabs.

Publication Date:

September 1991 Prepared by:

seeber. L: Simpson, D.: Johnson. D.: Armbruster, J. (Columbia Univ., Palisades, NY, Lamont-Dohetty Geological Observatory)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

aftershocks, attenuation, budgets, communications, cost, ranhquakes, Green's function, in form ati on system s, Kans as, m athem atical models, m odificati ons, m onitoring, Ne w J ersey, New York, Quebec, seismic arrays, seismic detection, seismic events, seismic surveys, seismic waves, seismi-ity, seismographs, signals, stresses NUREG/BR--0083,Vol.7 70 t

l l

NUREG/CR--5781 MELCOR, CONTAIN

Title:

Summary of a Workshop on Severe Accident Management f or PWRs

==

Description:==

Severe accident management can be defined as the ut.e of existing and/or alternative resources, systems, and actions to pre vent ormitigate a core-melt accident. For e ach accident sequence and each cornbination of strategies, there may te seveial options available to tie operator, and each involves phenomenological and operational considerations regardmg uncertainty. Operational uncertainty includes operator, system, and instrument tchavior during severe accidents. During the period May 15-17,1990, a workshop was held at the University of Califontia, Los Angeles, to address these uncenaintier 'or pressurized water reactors (PWRs). This repon contams a summary ofine worksinp proceedings.

Publication Date:

Novemter 1991 Prepared by:

Kastenberg, W.E. [ed.]; Apostolakis, O.; Jae, M.; Mibci T.: Park,11.; Xing, L.; Dhir, V.K.

Lim, H.: Okrent, D.; Swider, J.: Yu, D. (Califomia Univ., Los Angeles, CA. Dept of Mechanical, Aerospace arxl Nuclear Engineeritig)

Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

CONTAIN, comainment, data covariances, decision making, emergency plans, fission-product release, heat transfer, hum an fact ors, hydra ulics, m anagement, MELCOR, meltdown, PWR type reactors, reactor operaton, reactor safety, risk assessment i

l l

l l

t l

71 NUREG/BR--0083 ',ol.7 l

.. m

-m k-f 3;q;Y

~

's lNL7REGICR. 5794 VAM2D

)

Title:

Ground-Water Flow and Transport Modeling of the NRC Licensed Waste DisposalFacility,7 West Valley, New York Descripiion:

This report describes a simulation study of groundwater flow and radionuclide transport from.

disposal at the NRC licensed waste disposal facility in West Valley, New York; A transient,

-i precipitation driven lilow model of the near-surface fractured till layer and underlying

]

unweathered till was developed and calibrated against observedinflow data into a recently ;

q 1

constructedinterceptor trench for th: period Marcle-May 1990. The results suggest that j'

lateral flow through the upper, fractured tilllayer may be more signi6 cant thanirxticated by -.

previous, steady state flow modeling studies; A conclusive assessment of the actual.

magnitude oflateral flo. through the fractured till could,however,not be made; A primary -

8 w

factor contributing to this uncertainty is the unknown contribution of vertical infiltration.

t through the interceptor trench cap to the total trench inflow. The second part of the;

' investigation involved simulation of the migration of St 90, Cs-137, and Pu 239 from one '-

of the fuel hull disposal pits. A first order radionuclide leach rate with rate' coefficient of -

10 / day was assumed to describe radionuclide rele ase into the disposal pit. The simulations 4

indicated that for w astes buried below the fractured till zone, no significant migration would -

occur. However, under the assumed conditions, significant lateral migration could occur for radionuclides present in the upper, fractured till zone. 23 refs.,68 figs.,12 tabs.-

Publication Date:

October 1991 -

]

Prepared by:

Kool, J.B.; Wu, Y.S. (HydroGeologic, Inc., Herndon, VA) 2 Prepared for:

NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research '

{

Keywords:

cesium 137, computerized simulation, disposal wells, Dow models, geology, groundwater,- -

motutoring, plutonium 239, radioactive waste chsposal,t adioactive waste facilities, radioactive wastes, t dioisotopes, radionuclide migration, stratigraphy, strontium 90, underground -

- disposal,VAM2D, wells-c NUREG/BR-0083,Vol.7 72

- NUREG/CR 5795 VAM2D

Title:

Validation and Testing of the VAM2D Computer Code

==

Description:==

Tlus document describes two modeling studies conduaed by HydroGeologic, Inc., for the US NRC under contract no. NRC-04089-090, entitled, " Validation and Testing of the VAM2D Cot,suter Code." VAM2D is a two-dimensional, variably saturated flow and.

transp in code, with applications for performance assessment of nucleat waste disposal.The computer code itselfis documented in a separate NUREG document (NUREG/CR-5352,.

1989). The studies presented in this report involve application of the VAM2D code to two di,erse subsurface modeling problems. The first one it: >olves inodeling of infiltration and redistribution of water and solutes in an initially-dry, heterogeneous field soil. This appiication involves detailed modeling over a relatively shon. 9-month time period. Tir second problem penains to the application of VAM2D to the modeling of a waste disposal facility in a fractured clay, over much larger space and time scales and with panicular emphasis on the applicabihty and reliabihty of using equivalent porous medium approach for _

dmulating flow and transport in fractured geologic ' media. Reflecting the separate and distinct nature of the two problems studied, this repon is organized in two separate pans.

(l reis.,31 figs.,9 tabs.

Publication Date:

October 1991 Prepared by.

Kool, J.B. Wu, Y.S. (HydroGeoLogic, Inc., Herndon, VA)

Prepared for:

NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research Keywords:

contamination, Dow models, fluid Dow, fractures, groundwater, hydraulic conductivity, hydrology, porosity, radioactive waste disposal, radionuct'Je migration, soils, solutes.

surface waters, VAM2D 1

'l

-l t

j l

1 73 NUREG/BR--0083, Vol.7 a

.N..UREG/CR-5808 ORNL/TM--11970 General Tille:

Calculation of Absorbed Doses to Water Tools in Severe Accident Sequenses Descripti9n:

A methodology is presented for calcula'ing die radiation dost to a watet pool from the decay of uniformly distributed nuclides in that pool. Motivated by the need to accurately model radmlysis seactions ofiodine, direct application is tuade to fission-product sources dissolved or suspended in containment sumps or pools during a severe nuclear reactor accident. Two methods ofcalculating eamma alsogdon are discussed-one based onpoint-kemalintegration and the otter based on Monte Carlo techniques. Using least-squares minimization, the computed resul's are used to obtain a correlation that relates absorted dose to source ergergy and surface-to-volume ratio of the pool. This corTelation is applied to most relevant fission-product nuclides and used to actually calculate transient sump dose rate in a preuurized-water reactor (PWR) severe accident si,aence.

Publication Date:

Decemter 1991 Prepared by:

Weber, C.F. (Oak Ridge National Lab., TN)

Prepared for:

N 'C Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

BWR type reactors, containment, cooling ponds, fission-product telease, fission products, integrals, iodine, mathematical models, Monte Carlo method, point kemels, PWR type re actors, radiation absorption analysis, radiation doses, radiation effects, radiation transport, reactor accidents, n: actor safety NUREG/BR--0083,Vol.7 74

3

'NUREG/GR.-0003 :

CCS Database

Title:

Effect of Prior Deformation on Semitizatiou Development' in Stainless Stet! During Con:inuous Cooling

==

Description:==

Centinuous cooling semitization (CCS) occurs in austenitic stainless steel (SS) weldment ilA2s where the material is subjected to weld induced plastic deformation, and non-linear heating and cooling cycles. The primary purpose of this investigation y as to quantitatively determine the effects of prior deformation on CCS. In addition, these results were used to develop a CCS database for comparison to a recently published sensitization prediction model (SSDOS). Cominuous coohng thermal cycling of specunens from high-carbon Type 316 SSs was performed in a computer-controlled Gleeble thermal simulator. Tir degree of sensitization (DOS) of thermally treated specimens was quantitatively measured using the electrochemical potentiokineoc teactivation (EPR) tert. Prior defomiation significantly enhanced :he rate of CCS development in the Type 316 SS material. *Ihe DOS increased with increasing amounts of prior strain and decreasing cooling rates. Sensitization response was abo sensitive to peak cycle temperatures. Continuous cooling sensitization development occurredprimarilyinthecriticaltemperaturerangebetueenabout900and750*C Peakcycle temperatures of 1000 and 1050*C retarded sensitization development during subsequent continuous cooling. Strain recovery at elevated temperatures played an important role in reducing the effectiveness of prior deformeion in accelerating sensitization kinetics. Due to the effects of recovery,in certain cases, pnor strain values of 20% were only as effective as 10% in increasing the rate of sensitization development. Limited transgranular cartsde precipitation was observed in 20% prior strain samples depending on specific thermal cycle -

parameters but was not a significant factor in the present work. The SSDOS model -

consistently overpredicted the CCS development in both heats of 316 SS studied, regardless of material condition (i.e., mill-annealed, solution-annealed, and pre-strained materials).

Publication Date:

September 1991 Prepared by:

Simmons, J.W.; Attetidge, D.O.; Bru;mmer, S.M. (Oregon Graduate Inst. of Science and Technology, Beaverton, OR. Dept. of Matenals Science and Engineering)

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

BWR type reactors, carbides, CCS database, chromium, corrosion, deformation, heat treatments, precipitation, sensitivity, SSDOS prediction model, stainless stxl-316, thermal cycling, weldedjoints 75 NUREG/BR 0083,Vol.7 i:

0

m f

NUREG/GR--0(M-Draft DEPOSITION

Title:

- DEPOSITION: Software to Calculate Panicle Penetration Through Aerosol Transpon Lines. Draft Repon for Commeet

==

Description:==

In this repon, models are presented for calculating aerosol panicle penetration through straight tutes of ad>itrary orientation, inlets, and elbows. An expre;sion to calculate effeetive depositional velocities of panicles on tube walls is derived. The concept of " maximum penetration"is introduced, which is the maximum possible penetration through a sampling lir.e connecting anytwopoints in a three-dimensional space. A procedure to predict optimum tube diameter for an existing transport line is developed. An interactive menu driven software enutled DEPOSITION has been developed to perform above said tasks. This code can either te used on a PC or on a mainframe. The use and illustration of the software is described in Appendix A of the report.

Publication Date:

October 1991 Prep. ired by:

Anand, NX:McFarland, A.R.(Texas A&M Univ, College Station,TX Dept.of Mechanical Digineering)

Prepared for:

NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research i

Keywords:

aerosols, computer program document ation, computerized simulation, DEPOSITION, flow models, joints, radiation monitors, sampling tubes NUREG/BR--0083, Vol.7 76 A

}

APPENDIX A:Index by NUREG Series Report Number Report Number Page NUREG--!435-Suppl.1 1

NUREG/CP--0114-Vol.1 2

NUREG/CP--0114-Vol.2 3

NUREG/CP--0116-Vol.1 4

NUREG/CP--O116-Vol.2 5

NUREG/CP.-Oll8 6

NUREG/CR-3964-Vol.2 7

NUREG/CR--4063 8

NUREG/CR -4214-Rev.1-Pt 2-Add.1 9

NUREG/CR--4269 10 NUREG/CR -4295 11 NUREG/CR-4599 Vol.1-No.1 12 NUREG/CR-4735 Vol.7 13 NUREG/CR-1757 14 NUREG/CR--4816-Rev.1 15 NUREG/CR--5128 16 NUREG/CR--5282 17 NUREG/CR--5300-Vol.1 18 NUREG/CR--5304 19 NUREG/CR--5312 20 NUREG/CR -5331 21 NUREG/CR--5345 22 NUREG/CR-5352-Rev.1 23 NUREG/CR.-5377 24 NUREG/CR--5395-Vol.1 25 NUREG/CR--5423 26 NUREG/CR.-5456 27 NUREG/CR--5518 28 NUREG/CR--5520 29 NUREG/CR--5522 30 NUREG/CR-5531 31 NUREG/CR--5536 32 NUREG/CR--5537 33 NUREG/CR--5539 34 NUREG/CR-5561 35 NUREGICR--5565 36 NUREG/CR--5571 37 NUREG/CR--5577 38 NUREG/CR--5592 39 A1 NUREG/BR--0083, Vol.7

Report Numlier

.Pcge NUREG/CR--5595 40-i NUREG/CR.-5611 41-

. NUREG/CR.. Sold 42 NUREG/CR 5620 43.

]

NUREG/CR--5623 44 J

NUREG/CR 3630 45 NUREG/CR -5648 46 NUREG/CR-5651 47 NUREG/CR--5656 48

j NUREG/CR--5658 49 NUREG/CR--5662 -

50 NUREG/CR 5663 51 NUREG/CR--5667 52 NUREG/CR--5668 53

-1 NUREG/CR-5670 54 NUREG/CR--5677 55 j

NUREG/CR -5681

. 56 NUREG/CR--5684 57 -

NUREG/CR 5701 58-NUREG/CR--5711 59 NUREG/CK 5712 60 NUREG/CR--5715 61-NUREG/CR -5716 62 NUREG/CR-5729 63 NUREG/CR-5740 64 NUREG/CR -5743 65 NUREG!CR--5757 66 NUREG/CR--5765 67 NUREG/CR--5768 C8.

NUREG/CR-5773 69 NUREG/CR-5778 Vol.1 70 NUREG/CR--5781 71 NUREG/CR-5794 72-NUREG/CR-5795 73 NUREG/CR--5808 74' i

NUREG/GR--0003 75 h

NUREG/GR--0006-Draft 76 NUREG/BR-0083 Vol.7 A-2 J

-APPENDIX 11: Index by Software Identification Software Identification

.Page

. ADJMOM-1 64 ARANO 24 ASFIT 64 BIGFLO 62 DWF i TAS 36,37 g

B%L.a 36,37 CAIRE 5

CARES 2

CART 63 CCS Database 75 COffTAIN 6,17,28,45,

-i 50,61i71 CONTAIN-DCli 17 COMMIX.IC 27 CORCON 26 CORSAR M 53 CRAC 24 dB ASE 3 Plus 15 DCM3D 32,58 1

DEPOSITION 76 ENDF/B V Data File 46 ENDF/B-V Data File (Revised) 46 q

ENDF/B VI Data File 46-j EX~rRAN 48 FEMWATER

- 56 FLUhU'S 57 FORECAST 49 FPFP 2 49 FRAPCON1 6

FRAP-T6 6'

l General 7,9,10,11, 14,20,33,35,.

38,39,41,47, 59,70,74 GENII 34 HECTR 21 ICEDF 68 INVFD 30 IRRAS 18,29-B-1 NUREG/BR-0083 Vol.7

Software Identification Page MACCS 19,24,52 29 MAR D Database MELCOR 3,6,21,31, 36,37,44,71 MELTSPREAD 26 4

MINUIT MODFLOW 30 MORECA 60 NECTAR 24 NLFTRAN 2 58 NEW MIX

$5 NIST Database 64 NRCPIPE 12 NUCRAC 24 ORIGEN2 53 PAGAN 69 PALLAS 64 64 PilOTX Data Library PICASSO 2

PIFRAC 12 PORFLO 42 PORFLO3 62 PR-EDB 15 PRESTO 69 PROTOCOL 13 RELAPS 2,6,54 RELAP5/ MOD 2 25 IELAP5/ MOD 3 51 REMIX 55 SIMEVENT 5

SIMS Database 1

SMACS 66 SPARC-87 67 SPARC-90 67 16 SQUIRT TAC 63 TACMVS 63 THATCH 43 TRAC 2,25,54 TRACER 3D 62 TRAC-PFl/ MOD 1 8

TRUMP 62 NUREG/BR--00%, Vol.7 B-2

Softwere Identification -

-Page TRUST

- 62,65 UFOMOD

- 24' UNSAT2 62 UNSATH 62 VAM2D.

23,34,62,69, 72, 73 V1CTORIA 3,22,53-B3 NUREG/BR--0083, Vol.7

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. 3 APPENDIX C: Index by Contrcetor Report Number Report Number Page ANL-91/25 66 ANL-91/6 27 BAW-2023-Vol 1 25 BMI-2164 16 BMI -2173 Vol.1 No.1 12 BN1-NUREG-521' 1 17 6

BN1 NUREG-52251 41 BN1 NUREG-52271 50 -

BN1 NUREG-52280 56 BN1 NUREG-52295 69 BN1-NUREG-52297 43.

EGG-2613-Vol.1 18 EGO--2614 42 EGG-2630 29 EGO--2633 51 EGO-2634 52 EGG-2636 8

EPRI-NP-6480-Vol.1 25 EPRI-NP. 7165 54 LMF -132 Rev.1-Pt.2-Ack11 9

MCS-910401 63 NISTIR--4405 10 ORNL/RSIC--49/R1 64 ORN1/rM-Il548 36-ORN1/nt-103. Rev.1 15 ORNL/nt-11549 37 ORNL/ni-11581 39 ORNI/rM-11644 44 l

ORNI/TM-11686 46 ORNL/mi-11692 47 ORNL/nt-11743 53 ORNI/ni-il823 60 ORNI/ni-11970

'74 I

PNL-7510 48 PNL-7513 49 PNL--7597 14 PNL-7723 '

67.

PNL-7765 68 SAND-86-0196-Vol.2 7

i C-1 NUREG/BR-0083, Vol.7 s

Report Number l' age SAND-88-3324 20 SAND-89-0072 21 SAND-89 0308 22 SAND-90-00ll 28 S AND-90-0128 30 SAND-90-0364 31 SAND-90-0575 33 SAND-90-0585 34 SAND-90-2339 45 SAND-90-7015 32 SAND-90 7020 35 S AND-90-7116 19 S AND-91-0539 58 SAND--91-0835 61 SEA--89-461 11-A:1 40 l

NUREG/BR--0083,Vol.7 C-2 1

l

Keyword NUREG Repon Numtwr A-bomb survivors NUREG/CR--4214 Rev.1 Pt.2 Add.1 acceleration NUREG/CR-5757 ACRR reactor NUREG/CR--5345 activated carten NUREG/CP--O116-Vol.I NUREG/CP -0116 Vol.2 acute exposure NUREG/CR.-4214-Rev.1-Pt.2-Add.1 alabatic processes NUREG/CR--5282 ADJMOM-1 NUREG/CR-5740 adtmnistrative procedures NUREG/CR--5595 adsorbents NUREG/CP-C116-Vol.2 adsorption NUREG/CP--O116-Vol.I NUREG/CR--4269 aerosols NUREG/CP.-0116-Vol.1 NUREG/CP--O116-Vol.2 NUREG/CR.-4757 NUREG/CR--5282 NUREG/CR -5630 NUREG/CR-5715 NUREG/CR-5765 NUREG/CR-5768 NUREG/GR--0006-Draft after-heat tw wal NUREG/CP--0118 NUREG/CR 5712 aftershocks NUREG/CR--5778-Vol.1 age dependence NUREG/CR--4214 Rev.1-Pt.2-Add.1 aging NUREG/CP-0114 Vol.1 NUREG!CP-0114-Vol.2 NUREG/CP-Oll8 NUREG/CR 5577 air NUREG/CR -5620 NUREG!CR-5740 air cleaning systems NUREG/CP-0116-Vol.1 NUREG/CP--0116-Vol.2 air tilters NUREG/CP--0116-Vol.1 NUREG/CP--O116-Vol.2 air flow NUREG/CR--5656 air samplers NUREG/CR--4757 alarm systems NUREG/CR--5611 D-1 NUREG/DR--0083,Vol,7

Kryword.

NUREG Report Number algorithms -

NUREG/CR-5300-Vol.1 NUREG/CR -5456.

alkalimetal compounds

' NUREG/CR.-4269.

alpha detection NUREG/CP-Ollt>Voll alpha-bearing wastes -

NUREG/CA-3964 Vol.2 NUREG/CR-5539 NUREG/CR-5743 analytic functions NUREG/CR -5681 animals NUREG,CR-5716 anisotropy NUREG/CR-5743 ANL NUREG/CR-5667 annealing NUREG/CP.-Oll8 annihilation NUREG/CR--5740 aquatic ecosystems NUREG/CR--5377 aquifers NUREG/CR--5522 NUREG/CR-5743 ARANO NUREG/CR-5377 Arkansas 1 reactor NUREG/CR-5648 ASFIT NUREG/CR -5740 attenuation NUREG/CR-5740 NUREG!CR-5778 Vol.1 auxilimj systems NUREG/CR -5670 Battelle Pacific Nonhwest Laboratories NUREG/CR--5658 Bayes' theorem NUREG/CR-3964 Vol.2 bellows NUREG/CR-5561 hench-scale experiments NUREG.K'R-5537 benchmarks NUREG/CR-5681

' bendmg NUREG/CR-5757 beta detection NUREG/CP--0116-Vol.1 BIGFLO NUREG/CR--5716

~

biological accumulation NUREG/CR--5304 biological availability NUREG/CR-5304 biological pathways NUREG/CR-5377

)

- biological radiation effects NUREG/CR-- 214-Rev.1-It2-Add.1 biosphere NUREG/CR--5701 bladouts NUREG/CR--5331

NUREG/CR--5395-Vol.1 NUREG/BR--0083, Vol.7 D-2

Keyword NUREG Report Number blackouts (continued)

NUREG/CR-5456 NUREG/CR--5565

. NUREG/CR.-5571 NUREG/CR-5623 NUREG/CR--5630 NUREG/CR--5663 blowdown NUREG/CR-5282 BLT model NUREG/CR.-5681 NUREG/CR -5773 BNL NUREG/CR-5282 toreholes NUREG/CR--4295 NUREG/CR-5684 boundary corditions NUREG/CR--5352-Rev.1 NUREG!CR--5684 BR-3 reactor NUREG/CR--5312 NUREG/CR -5668 bremsstrahlung NUREG/CR-5740 bubbles NUREG/CR-5765 budgets NUREG/CR -5778-Vol.1 4

buildup NUREG/CR--5740 bumup NUREG/CR -5668 BWR type tractors NUREG/CP.-0114 Vol.1 NUREG/CP--0114-Vol.2 NUREG/CP--O116 VolI NUREG/CP-0118 NUREG/CR--4599-Vol.1-No.1 NUREG/CR--4757 NUREG/CR-4816-Rev.1 NUREG/CR--5128 NUREG/CR-5345 NUREG/CR--5423 NUREG/CR--5531 NUREG/CR-5565 NUREG/CR--5571 NUREG/CR--5577 NUREG/CR-5623 NUREG/CR--5677 NUREG/CR-.5765 NUREG/CR--5808 NUREG/GR-0003 BWR-LTAS NUREG/CR--5565 NUREG/CR--5571 BWRSAR NUREG/CR--5565 NUREG/CR -5571 CAIRE NUREG/CP--O116-Vol.2 D-3 NUREG/BR. 0083,Vol.7

Keycord NUREG Report Number calcium hydroxides NUREG/CR -4269 calcium silicates NUREG/CR--4269 calculation methods NUREG/CR-5773 California NUREG/CP 0114-Vol.1 carbides NUREG/GR-0003 carbon monoxide NUREG!CR--5715 cartx, steels NUREG/CR--5128 NUREG/CR-5648 carcinogenesis NUREG/CR--4214-Rev.1-Pt.2-Ad11 carcinomas NUREG/CR--4214-Rev.1 Pt.2 Adi!

CARES NUREG/CP-0114 Vol1 CART NUREG/CR-5729 CC1 rnodel NUREG!CR--5423 CCS database NUREG/GR--0003 cements NUREG/CR -4269 NUREG/CR--4295 NUREG/CR--5681 NUREG/CR-5684 cesium NUREG/CR--5681 cesium 134 NUREG/CR 5304 cesmm 137 NUREG/CR 5304 NUREG/CR-5794 Charpy test NUREG/CR-5577 NUREG/CR--5729 chemical reaction kinetics NUREG/CR--4269 chemical reacticas NUREG/CR-5312 NUREG/CR--5623 chlorides NUREG/CR--4269 chromium NUREG/GR--0003 chronic exposure NUREG/CR--4214.Rev.1-Pt.2-Add.1 NUREG/CR -53'M NUREG/CR--537' clays NUREG!CR-5522 climates NUREG/CR-3964-Vol.2 Clinton-1 reactor NUREG/CR--5571 Clinton-2 reactor NUREG/CR--5571 coherent scattering NUREG/CR--5740 Comancht Peak-1 reactor NUREG/CR--5456 NUREG/BR--0083, Vol.7 D-4

Keycord NUREG Hepo 1 Numl>cr l

Comande Peal. 2 :eactor NUK"OER.-5456 comtution NURLO/CR-5282 NUREGER. 5571 NUREO/CR- $662 NUREGER. 5715 comtetion pnduds NUREGER-5620 combustion properties NUREGER-5662 COMMIX 1C NUREGER-5456

(

communications NUREOCR. 5778.Vol.1 i

cornparative evaluations NUREO/CR-5377 compile data acquisition NUREO/CR.-4416 Rev.1 cornpiled data NUREO-1435. Suppl.1 NUREOCR-5304 cornpF :ce NUREGER-5537 compression NUREGER-5561 compo'at calculations NUREO!CR-5331 NUREO/CR-5571 NUREGER 5630 NUREO/CR. 5663 NUREGER-5740 NUREORR -5757 NUDEO/CR-5765 computer codes NUREOCP. 0114 Vol I NUREO/CR 5128 NUREOCR-5282 NUREGER-5300-Vol.1 NUREGER--5377 NUREO/CR- *667 NUREO/CR 5684 NUREO/CR-5773 computer graphics NUREO/Cl 5300-Vol.1 computer output devices NUREO/CR.-5300 Vol.1 computerprogram documentauon NUREO/CR-5352 Rev.1 NUREGER-5520 NURBOCR-5571 NUREO/CR-5595 NUREO/CR-5620 NUREGER 5656 NUREO/CR-5765 NUREO/OR-0006-Draft computerized control systems NUREO/CR--5670

. computerhed simulation NUREO/CP 0116-Vol.2 NUREO/CR--4063 D-5 NUREO/BR-0083,Vol.7

Keyword NUREG Heport Numlier cornputerir.ed simulation (continued)

NUREO/CR 512B NUREO/CR-5282 NUREO/CR-5300-vol.1 NUREOCR. 5352 Rev.l.

i NUREOCR 5395 Vol.1 NUREO/CR -5522 NUREO/CR-5614 NUREGER -3663 NUREO/CR--5667 NUREOCR 5668 NUREO/CR 5670 NUREO/CR $684 NUREO/CR 5743 NUREO/CR. 5757 7

NUREO/CR. 5794 NUREO/OR-0006-Drah concretes NUREO/CP-0118 NUREO/CR 4269 NUREO/CR-5423 NUREO/CR-5614 NUREORR 5623 NUREO/CR. 5715 g

NUREG/CR.-$740 CONTAIN NUREO/CP 9118 NUREO/CR-5282 NUREO/CR. 5518 NUREO/CR. 5630 NUREO/CR-5662 NUREO/Ck -5715 NUREO/CR. 5781 CONTAIN DCli NUREO/CR-5282 containers NUREO/CR -4735 Vol.7 NUREO/CR-5773 contaitanent NUREO/CP--0114.Vol.'

NUREO/CP.-0118 Ni'REO/CR.-4735 Vol.7 NUREG/CR 5331 NUREO/CR 5423 j

NUREO/CR -5518 i

NUREO/CR 5531 NUREO/CR-5565 NUREO/CR 5571 NUREO/CR -5623 NUREG/CR -5630 NUREO/CR 5662 NUREO/CR-T715 NUREO/CR-765

. NUREO/CR,768 NUREO/CR. "81 NUREO/CR 3 '08 NUREO/BR-0083, VoL7 D-6

Ke,5 word NUltEG Report Numt.cr contaitunent imildmp NUREG/CP.-0116 Yol.2 NUREG/CR. 5518 NUREGER 5571 contaitunent sptems NUREG/CP.-Oll4-Vol.1 NUREG/CP--O116-Vol.I NUREG/CR -4757 NUREG/CR 5282 NUREG/CR -5561 NUREG/CR 557i NUREG/CR 5662 NUREG/CR 5715 NUREG/CR 5765 NUREG/CR--5768 contamination NUREG/CR -5304 NUREG/CR.-5795 control elernents NUREGK:P Oll8 controt tod dnves NUREG/CP--0118 control rooms NUREG/CR. 5656 NUREG/CR 5658 control systerns NUREG/CP Oll6 Vol.)

convectiot NUREG/CR.-4269 cooling porvis NUREG/CR 5808 copper NUREG/CR 4735-Vol.7 CORCON NUREG/G *t'!3 corium NUREG/CP--Oll4 Vol.2 NUREG/CP.-Oll8 NUREG/CR 5282 NUREG/CR 5423 NUREG/CR 5571 NUREG/CR-5623 NUREG/CR-5667 NUREG/CR--5715 conclation functions NUREG/CR -5592 correlatiorts NUREG/CR 5592 ccr Jon NUREG/CR-5711 NUREG/GR- 0003 CORSOR M NUREG/CR -5668 cost NUREG/CR -5595 NUREG/CR 5778 Vol.1 cost estimadon NUREG/CR--5595 cost tenent analysis NUREG/CR-5662 CRAC NUREG/CR -53U D-7 NUREG/BR- 0083,Vol.7

c Keynord NUREG Report Number crack propagation NUREO/CP. Oll8 NUREG/CR--4$99 Vol.1 No.1 NUTIEO/CR-5$77 NUREO/CR 3592 NUREO/CR $6$1 crat NUREO/CP.-0118 NUREO/CR 4295 NUREO/CR 4599-Vol.1 No.1 NUREO/CR -473$.Vol.7 NUREO/CR 5128 NUREO/CR $$92 NUREO/CR $614 NUREO/CR-$729 cross sections NUREO/CR 5648 CRSTER nodel NUREO/Cis-$$37 damping NUREO/CR 5757 Darcylaw NUREG/CR. $743 data acquisition NUREO/CR- $670 data analysis NUREO/CR 5522 NUREG/CR 5743 data compilation NUREO/CR-$$22 data covariances NUREO/CR-5522 NUREO/CR-5$71 NUREO/CR 5781 data procet 's NUREO/CR -4816-Rev.!

NUREO CR-5520 database management NUREG-143$-Suppl.1 NUREO/CR-4816-Rev.1 NUREO/CR $300-Vol.1 NUREO/CR $$20 NUREO/CR $75 NUREG/CR--5)w dBASE 3 Plus NUREO/CR -4816-Rev.!.

DCM3D NUREO/CR -$$36 NUREO/CR.-5701 DECcomputers NUREO/CR. $667 decision making NUREO/CR -$781 decontamination NUREO/CR-$768 deformation NUREG,CR--429$

NUREO/CR 5561

- NUREO/GR-0003 delayed radiation effects NUREG/CR.-4214 Rev.1 Pt.2. Add.1 -

demonstration programs -

NUREO/CR.-473$-Vol.7

- NUREO/BR- 0083,Vol.7 D-8

Keyword NUREG Report Nucl>cr del'OSITION NUREO/OR -0006-Draft deposition NUREGER -4757 NUREOCR-5304 NUREORR-5658 NUREO/CR. 5668 depressurir.ation NUREGER. 5630 NUREO/CR 5712 depressuritation systems NUREO/CR-5565 design NUREO/CR -4295 NUREO/CR. iS37 NUREOCR-5711 detection NUREO/CP--Oll6 Vol.1 detonations NUREO/CR -5571 NUREG/CR.-5662 dif ferentialequations NUREO/CR -4757 NUREO/CR 4269 NUREGER-5536 diffusion NUREO/CR-4269 NUREO/CR-5658 -

NUREO/CR--5681 dimensions NUREO/CR.-4295 disente onhnate method NUREO/CR-5648 NUREO/CR--5740 disposal wells NUREO/CR 5794 diss slution NUREO/CR-4735 Vol.7 NUREGER-5681 documentation NUREO/CR-5518 dose rates NUREO/CR -4214 Rev.1 Pt.2-Add.1 dosimeters NUREO/CR-5648 dosimetry NUREO/CR--4214.Rev.1 Pt.2-Add.1 NUREO/CR -5539 drainage NUREOCR-5684 drinking water NUREO/CR-5377 droplets NUREO/CR -5282 dual absorption model NUREO/CR-5743

. dual continuum model NUREO/CR--5743 l

dusts NUREO/CP--0116-Vot2 L

dynamicloads NUREG/CR-5757 eatly radiation effects NUREO/CR -4214-Rev.1 Pt.2 Add.1 D-9 NUREO/BR--0083,Vol.7 -

l l

l

Keyword -

NUREG Report Number cartiquates NUREGEP.-0114.Vol.1 NUREOCP-0118 NUREO/CR-5778 Vol.1 Eason equation NUREGER-5577 ecologicalconcentraticm NUREO/CP-0116-Vol.2 NUREOCR-5656 cconomic atu ysis NUREO/CR-$.195 economicimpact NUREO/CR-5595 elasticity NUREO/CR-5577

. electric cables NUREO/CP- 0118 electrochemistry NUREOCR-$711 electrtxtes NUREGER-571i elements NUREO/CR-$740 ernbrittlernent NUREO/CP -Oll8 NUREO/CR 4816-Rev.1 NUREONR -5648 emergency pituis NUREO/CP-0114 Vol2 NUREO/CP 0116 Voi.1 NUREO/CR-5781 ENDF/B V data file NUREO/CR 5648 ENDF/D VI data file NUREGER-5648 energy spectra NUREO/CR -5740 engineered safety systems NUREO/CP-0114 Vol.1 NUREGER 5715 entrainment NUP&O/CR-5677 entropy NUREO/CR-5312 environment NUREGER 5658 environmentaleffeets NUREO/CR -4735 Vol.7 environmentalexposure pathway NUREO!CR -5377 NUREO/CR 5539 NUREO/CR. 5684 envirorvuentalpolicy NUREO/CR-5539 environmentaltransport NUREO/CR-5304

- NUREO!CR 5536 NUREO/CR 5656

- NUREO/CR 5716 epitheliomas NUREO/CR-4214 Rev 1-Pt.2 Adi!

equipment NUREO/CR. 5611-enors NUREO/CR 5518 NUREO/CR 5522 -

NUREO/BR.-0083,Vol.7 D-10

Keyn ord NUni:G Report Number 1i81"Pc NUREO/CP- 0114-Vol.2 evalualed data NUREO/CR 5648 evaluation NUREG--1435 Suppl.1 NUREO/CR 5565 expansionjoints NUREG/CR 5561 eaperiment platuting NUREOCR. 5670 experimental data NUREO/CR--4757 NUREO/CR 5345 NUREO/CR-5395 Vol.1 NUREO/CR-5681 NUREO/CR 5757 esplosions NUREO/CP- 0118 NUREO/CR.-5623 EXTRAN NUREO/CR-5656 failure mode analysis NUREO/CR-5331 NUREO/CR $611 failures NUREO/CR 5423 NUREO/CR-5456 NUREOTR--5611 NUREO/CR -5773 fallout NUREO/CR-5304 fatigue NUREO/CP.-0118 NUREO/CR -5128 fault tire analysis NUREO/CR-5300 Vol.1 Federal Republic of Germany NUREO/CP. 0116-Vol 1 feedw ater NUREO/CR 5670 FEMWATER NUREO/CR-5681 Deld tests NUREOCR 5537 fimte difference tnethod NUREO/CR -5681 6 nite element method NUREG/CR-5352-Rev.1 NUREO/CR -5561 NUREO/CR-5592 NUREO/CR-5651 fires NUREO/CP-O116-Vol.1 fission products NUREO/CR 5345 NUREO/CR. 5630 NUREO/CR--5648 NUREO/CR -5658 NUREO/UR--5668 NUREO/CR-5684 NUREO/CR 5715 NUREO/CR-5765 NUREO/CR-5595 NUREO/CR 5808 D NUREO/BR -0083,Vol.7

Keyw ord NURI:G ltepor1 Number Onion-product release NUREG/CP -0114-Vol.2 NUREG/CP--OII 8 NUREG/CR. 5345 NUREG/CR 5531 NUREG/CR-5630 NUREC/CR-.5658 NUREG/CR $b68 NUREG/CR -5715 NUREG/CR 5765 NUREG/CR 5781 NUREG/CR--5808 Dow imdels NUREG/CR 5522 NUREG/CR -5539 NUREG/CR. 5743 NUREG/CR-5794 NUREG/CR 5795 NUREG/GR 4XXK>-Draft Dow rate NUREG/CR-5282 NUREG/CR. 5312 NUREG/CR 5536 NUREG/CR-5614 NUREG/CR 5658 NUREG/CR 5684 fluid flow NUR EG/CP -0118 NUREG/CR-5352.Rev.)

NUREG/CR. 5423 NUREG/CR 5456 NUREG/CR -5536 NUREG/CR 5539 NUREG/CR 5614 NUREG/ Cit 5623 NUREG/CR--5677 NUREG/CR 5701 NUREG/CR--5716 NUREG/CR 5773 NUREG/CR -5795 o

Duid rnechardes NUREG/CR -5456 NUREG/CR -5677 NUREG/CR -5768 Ou;d-stn, tureinteractierts NUREG/CR--5743 FLUMPS NUREU/CR. 5684 fluorescence NUREG/CR.-5740 foal chains NUREG/CR-5304 NUREG/CR. 5377 FORECAST NUREG/CR. 5595 forecasting NUREG/CR-5522 NUREG/CR -5595 NUREG/BR.-0083. Vol.7 D-12

Kepord NUMEU Report Number FORTRAN NUREO/CR -5656 Fourier analyns NUREOCR-3964 Vol.2 ITFP 2 NUREO/CR-5658 fracture mechanics NUREORR--4599 Vol.1 No.1 NUREORR-5128 NUREO/CR-5$77 NUREORR 5592 NUREOCR-5651 fracture properties NUREO/CR -5592 NUREOCR 5651 NUREO/CR-5729 fractures NUREO/CR-5536 NUREO/CR-5795 France NUREO/CP- 0116-Vol.1 FRAP-T6 NUREO/CP -0118 FRAPCON1 NUREGEP-Oll8 friction iactor NUREO/CR 4295 fuel cans NUREO/CR -4735 Vol.7 NUREO/CR -5571 fuelelements NUREO/CP. Oll8 NUREO/CR-5312 NUREO/CR-5345 NUREO/CR 557 A.

NUREO/CR-5668 NUREO/CR--5670 fuel fabrication plants NUREO/CP--0116-Vol.1 fuel reprocessing plants NUREO/CP- 0116-Vol.1 fuel rods NUREO/CP. 0118 fuel. coolant interactions NUREO/CP-Oll8 NUREO/CR--5623 fumaces NUREO/CR-5668 gamma radiation NUREO/CR 5740 gamma spectroscopy NUREO/CR--5668 gaseous wastes NUREO/CR--5656 gases NUREO/CR.-4295 gastrointestinal tract NUREO/CR--4214.Rev.1 l't.2 Add.1 Oauss-Markov process NUREO/CR -5743 Gaussianplume model NUREO/CP-0116-Vol.2

- Oaussian processes NUREO/CR-5656 genetic radiation effects NUREO/CR-4214 Rev.1 Pt.2 Achi1 -

D-13 NUREO/BR -0083,Vol.7

Reywori NUR EG lleport Number genetically signincant dow NUREG/CR--4214 Rev.1 Pt.2. Add 1 GEN!!

NUREG/CR-5539 geologic fonnations NUREG/CP 0114 Vol.1 NUREG/CR 5743 geologic fractures NUREG/CR-$743 geologic surweys NUREG/CR-5743 geology NUREG/CR. 5684 NUREG/CR $794 glau NUREG/CR--4735 Ves' 7 Gouy Chapinan Stem-Graharneinekl NUREG/CR-5711 Grand Gulf 1 reactor NUREG/CR 5331

}

NUREG/CR 557i G:and Gulf 2 cactor NUREG/CR-5331 NUREG/CR-5571 graphite NUREG/CR 5620 Green's function NUREG/CR 5778 Vol.1 ground release NUREG/CP 0116.Vol.2 NUREG/CR -5681 proundwater NUREG/CR-5352 Rev.1 0,

NUREG/CR 5522 NUREG/CR -5536 NUREG/CR--5539 NUREG/CR. 5684 NUREG/CR. 5101 NUREG/CR 5716 NUREG/CR -57 3 NUREG/CR 5794 NUREG/CR. 5795

g. routing NUREG/CR 5684 gyissum NUREG/CR--4269 health hnards NUREG/CR -4214.Rev.1-Pt.2-Addl NUREG/CR-5630 NUREG/CR-5656 NU REG /CR-5701 heat trartsfer NUREG/CP--0114-Vol.1 NUREG/CP.-0114 Vol.2 NUREG/CP.-0118 NUREG/CR -4063 NUREG/CR 5128 NUREG/CR -5232 NW.EG/CR. 5312

.UREG/CR 5331 NUREG/CR-5395-Yol.1 NUREG/BR-(CR 3, Vol.7 D-14

\\

K' yw ord NURI:G Heport Number heat tiarnfer (conunued)

NUREG/CR. 542.4 NURI.0/CR. 5456 NUREGCR-5531 NUREOCR. 5571 NUREG/CR -5620 NUREG/CR-5623 NUREG/CR 5663 NUREG/CR 5667 NUREG/CR-5670 NUREG/CR 5677 NUREG/CR -5712 NUREG/CR -5715 NUREG/CR--5765 NUREG/CR 5768 NUREG/CR-5781 heat treatments NUREG/GR -0003 I; ECTR NUREG/CR 5331 Idgh pressure coolant injection NUREG/CR 5677 tugh ternperature NUREG/CR-5711 int -level radioactive wastes NUREG/CP -0114 Vol.1 h

NUREGCR. 3964-Vol.2 NUREG/CR -4735.Vol.7 NUREG/CR 5537 NUREG/CR 5684 NUREG/CR-5701 NUREG/CR -5743 IITGR type reactors NUREG/CR-5620 NUREG/CR. 5712 human factors NUREG/CP. 0114 Vol.1 NUREG/CP -OI14 Vol.2 NUREG/CP. Oll8 NUREG/CR.-5658 NUREGIR -5781 human intrusion NUREG/CR 39%Vol.2 humidity NUREG/CP--O116-Vol.I hydraulic conductivity NUREG/CR -5716 NUREG/CR--5743 NUREG/CR 5795 hydraulic fractures NUREG/CR--5743 bydraulia NUREG/CP -0114-Vol 1 NUREGCP--Oll4-Vol.2 NUREG/CP- 0118 NUREGER -4063 NUREG/CR-5128 NUREG/CR-5282 D-15 NUREG/IIR. 0083,Vol 7

Keyword NUREG Report Number hydrau'ics (continued)

NUREO/CR-$312 NUREO/CR 5331 NUREO/CR -5395 Vol.1 NUREO!CR-5423 NUREO!CR 5456 NUREO/CR--5531 NUREO/CR 5571 NUREO/CR--5614 NUREO/CR-5620 NUREO/CR. 5623 NUREO/CR-5663 NUREO/CR-5667 NUREO/CR-5670 NUREO/CR-5677 NUREO/CR. 5684 NUREO/CR -5712 NUREO/CR 5715 NUREO/CR -5765 NUREO/CR -5763 NUREO/CR 5781 hydrodynamic :miel NUREO/CR-5743 bydmJymunics NUREO/CR-5423 NUREO/CR-5765 NUREO/CP -O114 Yol.2 hydiogen NUREO/CP.-0118 NUREO/CR-5282 NUREO/CR--5571 NUREO/Cx 5630 NUREO/CR-5662 NUREO/CR -571$

hydmlogy NUREO/CR-5352 Rev.1 NUREO/CR 5522 NUREO/CR--5536 NUREO/CR -5684 NUREO/CR 5795 hysteresis NUREO/CR-5711 IBM computers NUREO/CR-5667 ice condemers NUREO/CR 5715 NUREO/CR -5768 1CEDF NUREO/CR--5768 Idaho National Engineering Lateratory NURhu/CR-5667 klaho chemical processing plant NUREO/CP--0116-Vol.1 impregnation NUREO!CP--0116-Vol.1 in serviceinspection NUREO/CP--0118 incinerators NUREO!CP.-Oll6 Vol.1 NUREO/BR -0083,Vol.7 D-16

Keyw ord NUREG Heport number inJoer air pollution NUREG/CR-5656 irtiustry NUREG/CR.-5595 infonnation dissemination NUREG-1435-Suppl.1 infonnation rystems NUREG/CR-.4816 Rev.1 NUREG/CR 5778-Vol.1 irdialation NUREG/CR--5377 inspemon NUREG/CP-Oll4 Vol.2 interrals NUREG/CR-5729 NUREG/CR--5808 interfaces NUREG/CR.-4295 intemational cooperation NUREG/CP -0116 Vnl.2 INYFD NUREG/CR -5522 ialme NUREG/CR--4757 NUREG/CR--576$

NUREG!CR-5808 imime 129 NUREG/CP O116-Vol.1 ioniring radiations NUREG/CR -4214-Rev.1-Pt.2.Acki.1 iridiurn oxides NUREG/CR 5711 iron NUREG/CR 5648 iron complexes

'!UREG/CR-5711 litadiation NUREG/CP-Oll8 NUREG/CR--5312 NUREG/CR 5345 NUREG/CR -5729 IRRAS NUREG/CR-5300 Vol.1 NUREG/CR--5520 Japan NUREG/CP--Oll4-Vol.2 joints NUREG/GR -00(%-Draft Kansas NUREG/CR -5718-Vol.1 Koreny-Cannen fommla NUREG/CR--5743 triging NUREG/CR 5522 La Salle County-1 reactor NUREG/CR 5331 La Salle County-2 reaetor NUREG/CR--5331 leaching NUREG/CR -4269 NUREG/CR-4735 Vol.7 NUREG/CR--5352 Rev.1 NUREG/CR--5681 NUREG/CR -5773 leaks NUREG/CR--5128 NUREG/CR--5614 D 17 NUREG/BR--0083,Vol.7

\\

Keywerd NUREG Repor1 Number length NUREO/CR.-4295 leukemia NUREO/CR--4214 Rev.1 Pt.2-Ad(i1 1-.ET NUREG/CR--4214 Rev.1 Pt.2 ' 111 levels NUREOCR--4063 license applications NUREO!CR-5539 licensmg NUREO/CR. 5537 liners NUREO/CR. 5423 liquid wastes NUREO/CR-5656 liquids NUREO/CR--4063 loss of coolant NUREO/CR 4063 NUREO/CR 5128 NUREO/CR 5395 Vol.1 NUREO/CR 5571 NUREO/CR 5670 NUREO/CR-5677 low-level radicadve wastes NUREO/CP. 0114 Vol.1 NUREO/CR-4269 NUREO/CR 5539 NUREO!CR-5614 NUREO/CR 5681 NUREG/CR--5716 NUREO/CR. 5773 hiACCS NUREO/CR -5304 NUR20/CR-5377 NUREO/CR 5667 magnetic filters NUREO/CP-0116-Vol.2 maintenance NUREO/CR 5518 NUREO!CP--0118 management NUREO/CR 5781 manuals NUREO/CR 5300-Vol.1 MAR-D database NUREO/CR--5520 materials NUREO/CR -5740 mathematical models NUREG/CP-O116-Vol.1 NUREO!CP.-0116-Vol.2 NUREO/CR 39M Vol.2 NUREO/CR -4214-Rev.1 Pt.2-Ad(11 NUREG!CR -4269 NUREG/CR -4295 NUREO/CR--4757 NUREO/CR-5304 NUREO/CR 5352-Rev.1 NUREO/CR--5537 NUREO/BR--0083,Vol.7 D-12

Keyword NURI:G Report Number mathematical matels (continued)

NUREO/CR 5577 NUREO/CR-5592 NUREO/CR 5614 NUREO/CR 5620 NUREO/CR -5656 NUREOCR-5677 NUREO/CR 5711 NUREO/CR-5716 NUREORR-5729 NUREO/CR-5765 NUREO/CR-5773 NUREO/CR-5778 Vol.1 NUREO/CR 5808 measuring methmis NUREO/CR 57II mechanicalproperties NUREO/CR -4295 NUREO/CR -4599-Vol.1 No.1 NUREO/CR 5651 mechardeal tests NUREO/CR. 4295 medium temperature NUREO/CR-5711 meetings NUREO/CP Oll4 Vol.1 NUREO/CP-Oll4-Vol.2 NUREO/CP.-Oll6-Vol.1 NUREO/CP Oll6-Vol.2 melanomas NUREO/CR.-4214 Rev.1 Pt.2-Adil MELCOR NUREO/CP -Oll4-Vol.2 NUREO/CP 0118 NUREO/CR-5331 NUREO/CR-5531 NUREO/CR-5565 NUREO/CR 5571 NUREO/CR -5623 NUREO/CR-5781 meltdswn NUREO/CR-5282 NUREO/CR 5423 i

NUREO/CR -5531 NUREO/CR-5571 NUREO/CR. 5623 NUREO/CR-5667 NUREO/CR-5781 melting NUREO/CR 5312 MELTSPREAD NUREO/CR -5423 membranes NUREO/CR 5614 mercury NUREO/CP.-Oll6-Vol.1 metals NUREO/CR-5729 micrc, processors NUREO/CR -5300-Vol.1 D-19 NUREO/BRLOO83,Vol.7

. - _ -.=

A

Keyword NUREG Report N;:mler mine shafts NUREO/CR. 4295 MINUIT NUREO/CP-0116-Vol.1 mitigation NUREO/CR 5282 NUREO/CR-5565 miting NUREO/CR-5282 NUREO/CR -5677 MODFLOW NUREO/CR 5522 m<dificatiora NUREO/CR. 5300 Vol.1 -

NUREO/CR. 5648 NUREO/CR-5667 NUREO/CR-5778 Vol.1 monitoring NUREO/CP-0116 Vol.2 NUREO/CR. 5522 NUREO/CR 5778 Vol.1 NUREO/CR-5794 l

Monte Carlo method NUREO/CR-5522 NOREO/CR-5611 NUREO/CR-5701 NUREO/CR-5740 NUREO/CR 580S MORECA NUREO/CR 5712 mortality NUREO/CR 4214 Rev.1 Pt.2-Addt!-

Navier Stokesequations NUREO/CR 5456 NUREO/CR-$743 NECTAR NUREO/CR-5377 NEFTRAN 2 NUREO/CR-5701

- neoplasms NUREO/CR--4214 Rev.1 Pt.2-Add.1 NUREO/CR 5630 neutron fluence NUREO/CR-5648 NUREO/CR 5729 neutron leakaFe NUKEO/CR-5648 neutron spectra NUREO/CR 5648 neutron transport NUREO/CR-5648 neutrons NUREO/CR -5648 New Jersey NUREO/CR-5778 Voll New Mexico NUREO/CR-5716 -

New York NUREO/CR 5778-Vol.1 NEWMIX NUREO/CR -5677 -

=

3 Newton method NUREO/CR-5312 NIST database NUREO/CR -5740 NUREO/IlR -0083,Vol.7 D-20

Keyword NUHEG Report Number introgen oxides NUREO/CP- 011(*Vol.1 nan:lestructive analysis NUREO/CP. 0118 Nors ay NUREO/CR-5304 notcbes NUREO/Ck 5592 noules NUREO/CP--0116 Vol.I NRCPIPE NUkEO/CR--4599-Vol.1 No.1 nuclear d.sta collections NUREO/CR-5648 nuclea. facilnies NUREO/CP 'll16-Vol 2 NUREO/CR-5520 nuclear power plants NUREG 1435 Suppl.1 NUREO/CP Oll4-Vol.2 NUREO/CP. 0116-Vol.1 NUkEO/CR.-4214-Rev 1-Pt.2 Add 1 NUREO/CR -4757 NUREO/CR. 5300 Vol.1 NURE'3/CR 5518 NUREO/CR 5520 NUREO/CR 5595 NUREO/CR. 5611 NUREO/CR-5658 NUREO/CR-5715 NUREO/CR-5757 nuclear waste pol.cy acts NUREO/CR-5539 NUCRAC NUREO/CR.-5377 4

numencal analysis NUREO/CR--5743 numerical solution NUREO/CR--5352-Rev.1 off gassystems NUREO/CP--Oll6-Vol.1 omega criterion NUREG/CR--5577 ene-dunensional calculations NUREO/CR. 5684 ORIGEN2 NUREO/CR-5668 origin NUKEO/CR--5522 ORNL NUREO/CP.-0116 Vol.1 osteosarcomas NUR'!O/CR--4214.Rev.1-Pt.2-Add 1 oxidation NUREO/CR -5312 NUREO/CR-5620 PAGAN NUREG/CR-5773 PALLAS NUREO/CR--5740 parametric analysis NUREO/CR--5630 panicle irsuspension NUREO/CR--4757 D-21 NUREG/BR--0083,Vol.7

Keym or d NUREG Report Number particle sire NUREG/CR--5282 NUREG/CR -5768 particulates NUREG/CR -5765 NUREG/CR--5768 Peach Bottom 1 reactor NUREG/CR 5331 Peach bottorn 2 rea: tor NUREG/CR--5331 Peach Dottom-3 reactor NUREG/CR-5331 peret ometers NUREG/CP 0116-Vol.1 perfonnance NUREG/CR 5300-Vol.1 NUREG/CR 5611 NUREG/CR--5684 performance testing NUREG/CR-5518 NUREG/CR 5537 NUREG/CR-5539 NUREG/CR--5701 NUREG/CR 5711 NUREG/CR -5768 penneability NUREG/CR. 4295 NUREG/CR 5614 Perry-l reactor NUREG/CR. 5571 Perry 2 reactor NUREG/CR 5571 personal cornputers NUREG/CR-5667 pH value NUREG/CR--5711 phase diagrams NUREG/CR -5312 phase transfonnations NUREG/CR--5312 PHOTX datalibrary NUREG/CR 5740 physical radiation effects NUREG/CR 5577 NUREG/CR -5729 PICASSO NUREG/CP-O114 Vol.1 4

PIFRAC NUREG/CR -4599 Vol.1 No.1 pipes NUREG/CP.-Oll8 NUREG/CR. 4599-Vol.1 No.1 NUREG/CR-5)N NUREG/CR -5564 NUREG/CR.-5729 NUREG/CR--5757 planinng NUREG--1435-Suppl.1 plaats NUREG/CR.-5377 NUREG/CR 5716 piasticity NUREG/CR--5577 plates NUREG/CR.-5651 NUREG/BR--0083, Vol.7 D-22

Keym ord NUREG lteport Number pluggmg NUREO/CR -4295 plutotuurn 239 NUP.EORR-5794 point it rnels NUREORR-5740 NUR EG/CR.-5808 Poisson Bnitnuannequation NUREORR 5711 pollution rc rulations NUREO/CP--0116-Vol.2 PORILO NUREO/CR 5614 PORILO3 NUREO/CR-5716 pormity NUREO!CR -4269 NUREO/CR 5684 NUREO/CR -5743 NUREO/CR--5795 porous ruaterials NUkEO/CR 5536 NUREO/CR 5716 post irradiationexamination NUREG/CR 5312 NUREO/CR. 5345 NUREO/CR -5668 PR EDB NUREO/CR--4816 Rev.1 precipitauon NUREO/OR -0003 pressure depen &nce NUREO/CR 5312 NUREO/CR 5536 pressure effects NUMEO/CR-5561 NUREO/CR. 5592 ptrssure release NUREO/CP- 0116-Vol 2 pressure suppressicn NUREO/CR-5565 pressure vessels NUREO/CP--0114-Vol.1 NUREO/CP 0118 NUREO/CR--4816-Rev.1 NUREO/CR-5282 NUREG/CR-5577 NUREO/CR-5592 NUREO/CR--5648 NUREO/CR -5651 NUREO/CR.-5729 pressurization NUREO/CR. 5282 PRESTO NUREO/CR -5773 prirnary coolant circuits NUREO/CP-0114.Vol.1 NUREO/CR--$128 NUREO/CR. 5282 probabilistic estimation UUREO/CP.-Oll4 Vol.1 NUREO/CR-3964 Vol.2 NUREO/CR-4214 Rev.1.Pt.2-Adil D-23 NUREO/BR--0033, Vol.7 4

Keyw ord NUREG luport Number probabilistic estiination (continued)

NUREO/CR-5303 Vol.1 NUREO/CR--$423 probabihty NUREO!CR -3964 Vol.2 NUREO/CR -5423 proparn managernent NUPEO !435-Suppl.1 proyramanng NUREO/CR 5658 NUREO/CR-5667 propens report NUREO/CR 4599.Vol.1 No I NUREO/CR 4735 Vol.7 PROTOCOL NUREO/CR--4735 Vol.7 PSOR inethm!

NUREO/CR-5614 public he Alth NUREO/CR-5667 NURE'NCP -0118 pumps NUREO/CR 5395 Vol.1 NUREO/CR 5670 PWR type reactors NUREO/CP-(

Vol.1 NUREO/CP Oll4 Vol.2 NUREO/CP-OI16-Vol.1 NUREO/CP--0118 NUREO/CR -4063 NUREO/CR-4599 Vol.1 No.1 NUREO/CR -4757 NUREO/CR -4816-Rev.1 NUREO/CR $128 NUREO/CR -5345 NUREO/CR -5395 Vol.1 NUREO/CR 5531 NUREO/CR-5577 NUREG/CR. 5592 NUREO/CR -5630 NUREO/CR-5648 NUREO/CR-5662 NUREO!CR-5670 NUREO/CR-5677 NUREO/CR 4768 NUREO/CR -5781 NUREO/CR 5808 quahty assurance NUREO/CR -4816-Rev.1 NUREO/CR 5304 NUREO/CR--5518 NUREO/CR-5667 quantitative chemical analysis NUREO/CR--5668 quantity natio NUREO/CR 5312 Quebec NUREO/CR 5778 Vol.1 s.

quenching NUREO/CR. 5282 o

NUREO/11R-4K)83, Vol.7 D-24

l Ec p ord NURI:G Report Numlier ra&ation absorption analpis NUREO/CR-5808 ra&ation accidents NUREO/CR -4214 Rev.1 Pt.2-Adil NUREO/rR 5304 4

raiation doses NUREO/CR--4214-Rev.1 Pt.2-Ad11 NUREO/CR-5304 NUREO/CR 5539 NUREO/CR.-5740 NUREO/CR-5808 ra&ation effects NUREO/CR--4735-Vol.7 NUREO/CR--4816 kev.1 NUREO/CR. 5808 ralation harards NUREO/CP--0116-Vol.1 NUREO/CR -4214 Rev.1 Pt.2 Add i NUREO/CR--5667 raintion inonitors NUREO/OR-4XX4DrafI ra&ation protection NUREO/CP. 0116 Vol.I NUREO/CR--5537 raaati; n transport NUREO/CR -4757 NUREO/CR-5808 ra&oactive clouds NUREO/CR-5656 radioactive effluents NUREO/CP--Oll8 NUREO/CR--4757 NUREO/CR-5681 NUREO/CR-5715 radioactive rnaterials NUREO/CR-5656 raioactive waste disposal NUREO/CR-3964 Vol.2 NUREO/CR 4269 NUREO/CR--4735 Vol.7 NUREO/CR--5352 Rev.)

NUREO/CR--5537 NUREO/CR 5595 NUREO/CL-5681 NUREO/CR -5c"4 NUREO/CR 5743 NUREG/CR -5773 VUREO/CR--5794 NUREO/CR--5795 raicactive waste facilities NUREO/CP.-0116-Vol.1 NUREO/CR.-4295 t UREO/CR-5352-Rev.1 NUREO/CR. 5537 NUkEO/CR. 5539 NUREO/CR -$684 NUREO/CR--57i!

NUREG/CR 5716 NUREOC 5773 NUREO!L,- 5794 D-25 NUREO/BR--0083, Vol.7

Keyword NURI:G Report Nurni>er radioactive waste managernent NUREORP -0114 Vol.1 NUREO/CP.-0114-Vol.2 NUREG/CP-Ollb Vol.1 N1iREO/CR 5684 radioactive w aste f torate NUREG/CR 5614 radmactive wastes NUREO/CP -0116-Vol.1 NUREO!CR 5794 radioactivity NUREO/CR. 5377 radioactivity transport NUREO/CR -4757 NUREONR 5658 NUPEO/CR 5667 raduecologicalconcentration NUREO/CR. 5304 NUREG/CR -$$39 radmisotopes NUREO/CR-5345 NUREO/CR-5377 NUREG/CR 5630 NUREO/CR-5668 NUREG/CR-5773 NUREO/CR -$794 radionuclide :nigration NUREO/CR-3964-Vol.2 NUREO/CR 4269 NUREG/CR -4735-Vol.7 NUREO/CR. 5304 NUREO/CR-5352-Rev.1 NUREG/CR--5377 NUREG/CR--5539 NUREO/CR-5681 NUREG/CR-5684 NUREO/CR-5701 NUREG/CR-5716 NUREG/CR. 5743 NUREG/CR $773 NUREG/CR 5794 NUREO/CR. 5795 randornness NUREG/CR-5743 reactivity NUREO/CR-5620 reactor accidents NUREG/CP. 0114-Vol.2 NUREG/CP.-Oll8 NUREG/CR -5282 NUREO!CR-5300 Vol.1 1

NUREG/CR. 5345 NUREG/CR-5456 NUREO/CR--55;8 NUREG/CR 5565 NUREG/CR--5577 NUREG/CR--5658 NUREO/BR-0083,Vol.7 D-26 c ___-__________ - _ -

Keyn ord NURI:G Report Number teactor accidents (contmued)

NUREG/CR -5662 NUREG/CR--5663 NUREGER 5667 NUREGER-5668 NUREG/CR-5712 NUREG/CR. 5765 NUREG/CR-5768 NUREG/CR-5R08 reactor components NUREG/CP--O114-Vol.I NUREG/CP.-0114 Vol.2 NUREG/CP--Oll 8 NUREG/CR. 5331 NUREG/CR-5456 NUREG/CR 5571 NUREG/CR-5611 NUREG/CR 5662 NUREG/CR--5670 reactor control systems NUREG/CP.-0118 reactor cooling systems NUREG/CP- 0118 NUREG/CR-5395-Vol.1 NUREG/CR -5456 NUREG/CR-5620 NUREG/CR. 5663 NUREG/CR--5670 NUREG/CR 5677 NUREG/CR-5712 NUREG/CR-5768 reactor core disruption NUREG/CR 5282 NUREG/CR 5571 NUREG/CR 5662 NUREG/CR-5667 reactor cores NUREG/CR -4063 NUREG/CR--5571 rc actor fuelmg NUREG/CR 5595 reactor materials NUREG/CP- 0118 NUREG/CR-4599 Vol.1 No.1 reactor operators NUREGRP--0118 NUREG/CR--5781 i

reactor protect on systems NUREG/CP.-0118 reactot safety NUREG-I435-Suppl.1 NUREG/CP--Oll4.Vol.1 NUREG/CP.-0114 Vol.2 NUREG/CP -Oll6-Vol.1 NUREG/CP 0118 NUREG/CR -4063 NUREG/CR--4599-Vol.1 No.1 NUREG/CR -5128 D-27 NUREG/BR. 0083,Vol.7

j thy

• ord NUREG Repo 1 Numler reactor safety (contmued)

NUREO/CR 5282 NUREO/CR 5300 Vol.1 NUREOCR-5331 NUREO/CR-5345 NUREO/CR -5395 Vol.1 NUREO/CR--5423 NUREO!CR-5456 NUREO/CR. 5531 NUREO!CR 5571 NUREG/CR -5577 NUREO/CR-5611 NUREO/CR-5623 NUREO/CR-5630 NUREO/CR 5658 NUREO/CR 5663 NUREO/CR.-5670 NUREO/CR-5677 NUREO/CR-5712 NUREO/CR 5729 NUREO/CR -5765 NUREO/CR. 5768 NUREO/CR--5781 NUREO/CR-5808 reactor s,hutdown NUREO/CR-5595 reactor stability NUREO/CP -Oll4-Vol.1 reactor vessels NUREO/CR 5331 NUREO/CR.-5423 NUREO/CR-5571 NUREO/CR--5648 NUREO/CR--5670 real time systems NUREO/CP--Oll6-Vol.1 recommendations NUREO/CP -0116-Vol.1 NUREO/CP-O116-Vol.2 NUREO/CR-5522 NUREO/CR--5537 NUREO!CR--5592 NUREO/CR-5630 NUREO/CR-5C 81 redox potential NUREO/CR-5711 ton NUREO/CR--5312 regression analysis NUREO/CR-5522 regulations NUREG-1435 Suppl.1 NUREO/CP-Oll4 Vol.1 NUREO/CP--Oll6-Vol.2 NUREG/CP- 0118 NUREO/CR-5522 NUREO/CR 5537 NUREO/CR--5595 NUREO/CR 5743 NUREO/BR.-0083,Vol.7 D 28

Keycord NURI:G Report Number RLLAPS NUREO!CP--Oll4-Yol.1 NUREO/CP.-0118 NUREO/CR 5670 RELAP5/ MOD 2 NUREO/CR-5395 Vol.1 RELAP5/ MOD 3 NUREO/CR 5663 relialulity NUREO/CP-Oll4-Vol.1 NUREG/CP--Oll4.Vol.2 NUREO/CP-Oll8 NUREO/CR -5611 REMW NUREO/CR 5677 reprocessing NUREO/CP-Oll6-Vol.1 reser?- programs NUREO/CP. 0114 Vol.1 NUREO/CR -4599-Vol.1 No.1 NUREO/CR. 4735 Vol.7 NUREO/CR -53!2 NUREO/CR. 5684 research reactors NUREO/CR-5677 reservott rock NUREO/CR.-4295 response functions NUREO/CR 5522 NUREO/CR--5740 NUREO/CR 5757 restraints NUREO!CR 5757 rnention NUREO/CR 5743 reviews NUREO/CP--0114-Vol.2 NUREO/CR 5304 NUREO/CR 5539 revised ENDF/B-V Data File NUREO/CR 5648 rist assessment NUREO/CP.-0114-Vol.1 NUREO/CP.-0114-Vol.2 NUREO/CP -O118 NUREO/CR 39M-Vol.2 NUREO/CR.-4214-Rev.1 Pt.2-Adi1 NUREO/CR--5300-Vol.1 NUREO/CR-5304 NUREO/CR-5377 NUREO/CR--5423 NUREO/CR--5520 NUREO/CR--5531 NUICO/CR--5630 NUREO/CR--5781 River Bend l reactor NUREO!G 5571 River Bend.2 reactor NUREOTR-3571 rock mechanics NU REO!CR--5684 DJ9 NUREO/BR-0083,Vol.7 l

Kiyword NUREG Repor1 N.

root alisorption NUREG/CR. 5304 i

runoff NUREG/CR 5304 ruptures NUREG.CR 5128 NUREG/CR. 5395 Vol.1 NUREG/CR 5670 NUREG/CR-5768 ruthenium NUREG/CP.-0116-Yol.1 ruthutium oddes NOREG/CR -5711 safety NUREG/CP--0116-Vol.1 NUREG/CP.-0116-Vol.2 safety analysis NUREG/CR-5712 safety standards NUREG/CP. 0116-Vol.1 sampling NUREG/CP-Oll6 Vol.1 NUREG/GR -00(4 Draft Sandia laboratories NUREG/CR 5312 NUREG/CR--5667 NUREG/CR. 5684 saturation e UREG/CR -4295 NUREG/CR.-5614 scale models NUREG/CR.-4063 NUREG/CR-5395.Vol.1 NUREG/CR. 5670 NUREG/CR-5677 scaling laws NUREG/CR 5677 sealing materials NUREG/CR--4295 seals NUREG/CR -4295 NUREG/CR-5684 seismic arrays NUREG/CR -5778-Vol.1 seismic detection NUREG/CR. 5778 Vol.1 se aiic effects NUREG/CP--0118 t

NUREG/CR--4599 Vol.1.No.1 NUREG/CR 5757 seismic events NUREG/CR--5778-Vol.1 seismic surveys NUREG/CR-5778-Vol.1 seismic waves NUREG/CR.-5778-Vol.1 seismicity NUREG/CP 0114-Vol.1 NUREG/CP--0118 NUREG/CR-5778-Ve.l.1 seismographs NUREG/CR--5778-Vol.1 sensitivity NUREG/GR -(003 NUREG/BR--0083,Vol.7 D-30 i

Keyw or d NUIEG Heport Number sensitivity analyas NUREO!CR 5331 NUREO!CR 5522 NUREO/CR. 5630 NUREO/CR-5712 Sequoyah-1 reactor NUREO!CR.-5331 NUREOCR-5561 Sequoyate2 reactor NUREO!CR 5331 NUREO/CR -5561 service life NUREOCP -0114 Vol.1 NUREO/CR.-5577 sex deperence NUREO/CR -4214 Rev.1 h.2-Adi!

shear NUREO/CR -4295 thielchrg NUREO/CR-5740 sigrub NUREO/CR-5778 Vol.1 silt NUREO/CR--5522 r

SIMEVENT NUREO/CP--0116-Vol.2 SIMS database NUREG 1435 Suppl.1 site approvals NUREO/CR -5537 ute characterization NUREG/CR 5352 Rev.!

NUREGER--5522 NUREG/CR.-5716 sire NUREO/CR-4295 SMACS NUREO/CR--5757 soil mechanics NUREO!CP-0114 Vol.I NUREO/CR. 5743 soils NUREO/CR--5304 NUREO/CR-5352-Rev.1 NUREO,CR -5614 NUREO/CR.-5716 NUREO/CR-5795 solutes NUREO/CR -5716 NUREO/CR-5795 somatically signincant dose NUREO/CR--4214 Rev.1-Pt.2 A&tt source terms NUREO/CP--0114-Vol.2 NUREO/CP--O118 NUREO/CR--5345 NUREG/CR-5531 NUREO/CR-5539 NUREO/CR--5630 NUREO/CR -5667 NUREO/CR 5681 NURTCR--5773 D-31 NUREO/DR 0083,Vol.7

5

___.__._-__m

1;eyword NUREG Report Nur l cr rpace depemience NUREG/CR-5536 NUREGER-5656 SPARC-87

?!UREG/CR 5765 SPARC 90 NUREG/CR 5765 spatial dtstntmtion NUREG/CR 5656 NUREG/CR 5743 specificatioits NUREGER-5537 NUREG/CR. 5670 spent fuels NUREG/CR -3964 Vol.2 SQUIRT NUREG CR-512*.

SSDOS prediction model NiMEG/GR-(003 stuuleas stecl 316 NUREG/GR--(003 stainless steels NUREG/CP -0118 NUREG/CR-5668 starklanis NUREG/CR 5740 steady state con &tions NUREG/CR-5684 stearn NUREG/CR -4063 NUREG/CR 5571 NUREG/CR. 5623 NUREG/CR-5768 stearn generators NUREG/CR 5282 NUREG/CR-5395-Vol.1 NUREG/CR -5670 steel-ASThl A533 NUREG/CR-5651 steels NUREG/CP-Oll8 NUREG/CR.-4599 Vol.1 No.1 NUREG/CR. 4816-Rev.1 NUREG/CR-5592 NUREG/CR-5648 NUREG/CR-5729 stochastic processes NUREG/CR--5743 stoichiometry NUREG/CR-5312 storage facihties NUREG/CR-5614 strains NUREG/CR--5592 NUREG/CR--5757 stratification NUREG/CR--5456 NUREG/CR-5677 stratigraphy NUREG/CR-5522 NUREG/CR-5743 NUREG/CR--5794 NUREG/BR--0083, Vol.7 D-32

.)-

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Keyword NUREG Report Number stress analysis NUREG/CR -4295

.i NUREG/CR-5592 NUREGER-5757.

stress corrosion NUREG/CR 4735 Vol.7 stresses NUREGER--4295 NUREGO-5651 NUREG/CR-5778 Vol.1 stronnum 90 NUREG/CR-5794 suhrface environments NUREG/CR-5536 subsurface structures NUREG/CR-5743 sulfa'es NUREG/CR.-4269 supports NUREG/CR 5757 surface contandnation NUREG/CR--5743 surface waters NUREG/CR-5304 NUREG/CR--5539 NUREG/CR-5795 Surry-1 reactor NUREG/CR--5630 NUREG/CR.-5662 Sur y-2 reactor NUREG/CR 5630 NURF.GER-5662

-i Surry-3 reactor NUREG/CR-5630 Surry-4 rtactor NUREG/CR -5630 surveillance NUREG/CR-5577 NUREG/CR -5729 Sweden NUREG/Cl--01h G.

systems analysis NUREG/CP-0114-Vol.2 NUREG/CR-5300 Vol.1 SYVAC geosphere model NUREG/CP--0116-Vol.1 TAC NUREG/CR -5729 TACMVS NUREG/CR-5729 tectonics NUREG/CR-3964-Vol.2 temperature dependence NUREG/CR--4295 NUREG/CR-5312 l

temperature distribution NUREG/CR-5456 temperature gradients NUREG/CR -4295 NUREG/CR-5668 temperature measurement NUREG/CR-5768 -

ter.sce propernes NUREG/CR--4295 NUREG/CR-5577 '

NUREG/CR-5592 D-33 NUREG/BR-0083, Vol.7 1

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h Keyword NUREG Report Numtwr -

- terrestrialecosystems NUREO/CR--5377 test facilities NUREO/CP-0116-Vol.1 NUREO/CR-5312 NURBO/CP-5345 NUREO/CR--5395 Vol.1 NUREO/CR-5668 '

NUREO/CR-5670 NUREO/CR -5677 -

5'REO/CR 5768 THATCH NirREO/CR--5670 theoretical data NUREO/CR-5656 thermal analysis NUREO/CR--5456 NUREO/CR 5620 thermal cycling NUREO/GR-0003 thermaldegradation NUREO/CP -0118 thermal shock NUREO/CP--0118 NUREO/CR-5592 NUREO/CR-5677 thermodynamics NUREO/CRA715 -

thickness NUREO/CR-5648 Thiele modulus NUREO/CR -4269 thin films NUREO/CR-5711 Three Mile Islan i ) reactor NUREO-1435 Suppl.1 NUREO!CR 5740 three-dimensional calculations NUREO/CR-5536 NUREO/CR -5684 thyroid NUREO/CR -4214-Rev.1 Pt.2-Add.1 time dependence NUREO/CR 5536 NUREG/CR-5656 NUREO/CR-56S4 NUREO/CR-5701 time series analysis NUREO/CR-3964-Vol.2 tissues NUREO/CR--5740 toxic materials NUREO/CR--5656 TRAC NUREO/CP--O114-Vol.I NUREO/CR-5395-Vol.1 NUREG/CR-5670 TRAC-PF1/ MOD 1 NUREO/CR -4063 TRACER 3D NUREO/CR--5716 transformers NUREO/CP--0118 -

NUREO/BR -0083,5'ol.7 D-34

Keyrord NUREG Report Number transients NUREG/CR--5395-Vol1 NUREG/CR--5620 NUREG/CR--5670 NUREG/CR--5684 transmission NUREG/CR -5M8 trarapitation NUREG,CR-5352-Rev.1 transport theory NUREG/CR-5740 TRUMP NUREG/CR-5716 TR11ST NUREG!CR--5716 NUREG/CR--5743 tubes NUREG/CR-5395-Voll 3

NUREG/CR--5670 NLREG/GR--0006-Draft tuff NUREG/CR--4295

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NUREG/CR--5701 tunnels NUREG/CR--4295 turbines NUREG/CP--0118 two-phase flow NUREG/CR -5128 NUREG/CR-5395-\\..

NUREG/CR-5670 UFOMOD NUREG/CR--5377 undergrounddisposal NUREG/CR--3964-Vol.2 NUPIG!CR -4295 NUREG/CR-5537 s

NUREG/CR-5539 NUREG/CR--5681 NUREG/CR--5701 NUREG/CR-5794 t'nderground storage NUREG/CR.-4295 NUREG/CR--5614 UNSAT2

't.UREG/CR-5716 UNSATH NUREG/CR--5716 uranium dioxide NUREG/CR--5312 US DOE NUREG/CP--0116 Vol.1 US EPA NUREG/CR--3964-Vol.2 US NRC NUREG/CP--0114-Vol1 NUREG/CP-0116-Vol.I NUREG/CR--5282 NUREG/CR--5312 NUREG/CR-5539 NUREG/CR-5595 NUREG/CR--5658 NUREG/CR-5667 D-35 NUREG/DR--0033,Vol.7

Keyword NUREG Report Number USA NUREG/CP--0114 Vol.1 NUREG/CP--O114-Vol.2 validation NUREG/CR-5128 NUREG/CR-5304 NUREG/CR--5537 NUREG/CR--5577 valves NUREG/CP--O114-Vol.2 NUREG/CP--O118 NUREG/CR--5670 VAM2D NUREG/CR-5352-Rev.1 NUREG/CR-5539 NUREG/CR--5716 NUREG/CR-5773 NUREG/CR--5794 NUREG/CR-5795 Van Genuchten-Mualem medel NUREG/CR -5743 NUREG/CR--5765 vapors ventilation systems NUREG/CP--0116-Vol.1 NUREG/CP--Oll6-VoL2 NUREG/CR-4757 NUREG/CR--5656 NUREG/CR-5658 verification NUREG/CR-5684 VICTORIA NUREG/CP--0114-Vot2 NUREG/CR--5345 NUREG/CR--5668 Wang-Natasimhan model NUREG/CR--5743 waste forms NUREG/CR--5681 NUREG/CR 5773 NUREG/CR--4295 water NUREG/CR -5614 NUREG/CR--5620 NUREG/CR -5623 NUREG/CR-5740 water chemistry NUREG/CR--5670 water cooled reactors NUREG/CR--5668 water saturation NUREG/CR-5536 NUREG/CR--5743 weatherin NUREG/CR--5304 e

weldedjoints NUREG/CP -0118 NUREG/CR-4599-Vol.1-No.1 NUREG/CR.-5128 NUREG/CR--5592 NUREG/CR -5729 NUREG/GR-0003 NUREG/BR--0083, Vol.7 D-36

Keyword NUREG Report Numher we11 logging NUREG 5'R--5522 wells NUREG/CR-5522 NUREG/CR-5794 WIPP NUREG/CR-5684 WNP 1 reactor NUREG/CR-5663 xenon 133 NUREG/CR-5740 Young's modulus NUREG/CR -4295 -

Yucca Mountain NUREG/CR--4735 Vol.7 NUREG/CR.-5701 NUREG/CR-5743 Zion-1 reactor NUREG/CR-5282 NUREG/CR--6662 Zion-2 reactor NUREG/CR-5282 NUREG/CR--5662 zircaloy NUREG/CR -4735-Vol.7 zirconiurn NURE&f%5312 I

D-37 NUREG/BR.-0083, Vol.7

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