ML20079G124

Computer Codes and Mathematical MODELS.January-December 1990

ML20079G124
Person / Time
Issue date:	09/30/1991
From:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
To:
References
NUREG-BR-0083, NUREG-BR-0083-V06, NUREG-BR-83, NUREG-BR-83-V6, NUDOCS 9110080420
Download: ML20079G124 (168)

Text

.

NUREG/BR-0083 Volume 6

p...n,,

k United States

%,,,j/

Nuclear Regulatory C(unmission Computer Codes and Mathematical Models January-December 1990 September 1991 Division of Information Support Senices Omce of information Resources Management l

g col g 910930 BR-OOB3 R PDR

i NUltEG/ Bit-4)083 Volume 6

,... m.,,

I United States Nuclear Regulatory Commission

\\,,,,,,.

Computer Codes and Mathematical Models January-December 1990 Compiled by C. E. Eyberger, M. K. llutter, M. Ilirgersson, l'. K. Degges National Energy Software Center Argonne National Iaboratory 9700 South Cass Avenue Argonne,IL 60439 September 1991 Division of Information Support Services Office ofInformation 1(esources Management

TAllLE OF CONTENTS l's l'OREWORD v

AllSTRACTS 1

APPENDIX A: Index by NUREG Report Number A-1 APPENDIX 11: Index by Software identification IL1 A"PENDIX C: Index by Contractor Report Number C1 APPENDIX D. Index by Kr.'yword D-1 iil

FOREWORD This report contains 4bstracts of NUREG drcuments issued in calendar year 1990 relating to computer zwftware for scientific, engineering, or tecimology-related programs performed or spon cred by the U. S. Nuclear Regulatory Commission, it is intended as a reference tool to assist the scientific and technical analyst in obtaining infornution on NRC computer-relatcd activities.

The abstracts appear in NUREG Jocument order. Abstracts of NRC staff generated reports doignated NUREG-xxxx are listed first, followed by those for conference proceedings identified as NUREG/CP-xxxx, contractor-generated reports published as NUREG/CR xxxx documents, and the International Agreement reports issued as NUREG/IA-xxxx publications.

Each abstract entry contains the following: NUREG repott number; software identification; contractor report number; report title; a desc/iption of the report contents; publication date; names of the individuals responsible for prepadng, compiling or edhing the report; contractor name and address; sponsoring NRC organization; and keywords or descriptors.

Indices by NUREG report number, software identification, contractor report number, and keyword are included as Appendices.

The abstracts aplear in two formats: D for those with primary emphasis on specific mathematical models, computer codes, or databases, and 2) for those that cont.3in s!gruficant information on many.

Specific code names and software identification appear in the heading of entries in the first format. The term " General" is used in the heading of those entries which refer to many models, computcr codes, or databases.

The Keywords section of the abstract contains the names of codes referred to in General entries and thou of evondary importance in other reports.

.l 1

l

NOREG-1150, Vol.1 General

Title:

Severe Accident Risks An Assessment for Five U. S. Nuclear Power Plants Final Summary Report Descriptioru This report summarizes an assessment of the risks from severe acddents in five commercial nuclear power plants in the United States. These risks are measured in a number of ways, including: the estimated frequencies of mre damage accidents from internally ir.itiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of raJionuclide releases and offsite consequenas of such acddents; and the overall risk (the product of accident frequencies and conseqqci N. Supporting this summary report are a large number of reports writt wt contract to NRC which provide the detailed discussion of the methn a<

nd results obtained in these risk studies.

Volumt

...t.is report us three parts. Part I provides the background and objectives of the assessment and summarizes the raethods used to perform the risk studies. Part 11 provides a summary of results obtained for each of the five plants studied. Part lit provides perspntives on the results and discusses the role of this work in the larger context of the NRC staff's work.

Accident progression event trees developed for this study made extensive use of available severe accident experimental and calculational databases. The analysis staff made use of calculational results from a number of accident simulation computer codes, including STCP, CONTAIN, MELCOR, and MELPROC. Source terms were calculated (one for each plant) by the,et of codes known collectively as the "XSOR* codes. These codes are parametric in nature; that is, they are designed to use the results of more detailed mechanistic codes or analyse as input. Release terms are divided into two time periods, an early release end a delayed relcan. Operator actions were evaluated using the SLIM MAUD method for transforming man-man and man-machine information into probability statements, and the MELCOR Accident Consaluence Code System (MACCS) was used to perform the offsite consequenm analysis. n the MACCS analysis, the protective actions to mitigate the chronic exposure pathways are largely confined to the 50-mile region of the site.

Publication Date December 1990 Prepared by:

NRC Division of Systems descarch, Office of Nuclear Regulatory Research Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Pescarch Keyworils:

nuclear power plants, acetdents, risk assessment, probabilistic estimation, reactor cores, containment, radionuclide migration, STCP codes, MELCOR codes, MELPROC codes, CONTAIN codes, SLIM MAUD codes, MACCS codes 1

I NUREG-1150, VCl. 2 G:nstal i

Title:

Severe Accident Risks An Assessment for Five U. S. Nuclear Power Plants Appendices A, B, and C l'inal Report Descriptioru This report summarizes an assessment of the risks from severe accidents in five commercial nuclear pcwcr plants in the United States. These risks are measured in a number of ways, induding: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionudide releases and offsite consequences of such acddents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC which provide the detailed discussion of the methods used and results obtained in these risk studies.

Volume 2 of this report contains three appendices, providing greater detail on the methods used, an example risk calculation, and more detailed discussion of particular technical issues found important in the risk studies.

Appendix A provides an overview of the NUREG-1150 risk analysis process, describing the different steps in the calculational process and the interrelationships among steps. Models used in calculation of risk include SETS and TEMAC for the systems analysis, EVNTRE snd PSTEVNT for accident progression analysis, XSOR for source term analpis, and PARTITION and MACCS for conquence analysis. The intermediate mformation obtained is then processed by RISQJE/PRAMIS, a matrix manipulation code, to calculate risk.

Appendix B follows an example calculation through the entire analysis from the initiating event in the accident frequency analysis through to the offsite risk.

CONTAIN, MAAP, and SURSOR were used by members of the containment loadings expert panel considering the pressure rise at vessel breach at Surry.

Other codes mentioned as the basis for the experts' t imates are EVNTRE, PSTEVNT, NAUA, STCP, MELCOR, PARTITION, and MALCS.

Appendix C summarizes the way in which a few issues important to risk uncertainty were treated in the five PRAs addressed in this report. Computations performed with MARCH 2, MARCH 3,

HECTR, TRAC /MELPROC, RELAPS/SCDAP, CORMLT/PSAAC, MAAP, STCP, the XSOR set, COMPBRN, and TEMAC are cited by experm. In this section of the report.

Publication Date:

December 1990 Prepared by:

NRC Division of Systems Research, Office of Nudcar Regulatory Research Prepared for:

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

nudcar power plants, accidents, risk assessment, probabilistic estimation, reactor cores, containment, radionudide migration, SETS codes, TEMAC codes, EVNTRE codes, PSTEVNT codes, XSOR codes, PARTITION codes, MACCS codes, RISQUE /PRAMis codes, CONTAIN codes, MAAP codes, SURSOR codes, NAUA codes, STCP codes, MELCOR codes, MARCH codes, HECTR codes, TRAC /MELPROG codes, RELAPS/SCDAP codes, CORMLT/PSAAC codes, COMPBRN codes 2

s

I NUREG-1266, Vcl. 4 Gen:ral l

~

Title:

NRC Safety Research in Support of Regulation - FY 1989

==

Description:==

This report, the fifth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during IY 1989. This research is necessary to make certain that the regulations that are imposed on licensees provide an adequate margin of safety so as to protect the health and saftty of the public. This report describes both the direct contributions to scientific and technical knowledge with regard to nuclear safety and their regulatory applications.

Acromplishments discussed indude studies performcd at the Idaho National Engineering Laborrtory using SCDAP/RELAP5 to evaluate potential depressurization strategies to minimize direct containment heating in a station blackout accident scenario; a research program developed to support the review of BWR Owners C:oup solutions for prevention and/or mitigation of power oscillations and review of emergency procedures using the RAMONA 3B and HIPA codes; development of a direct link between the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and computational codes, such as 'he lategrated Reliability and Risk Analysis System (IRRA5); field testing of an artificial intelligence-based Cognitive Environment Simulation (CES) and its reliability assessment application by means of a Cognitive Reliability Evaluation Technique (CREATE); field-evaluation at lleensed facilities of the user friendlinese.

and utility of the Maintenance Personnel Performance Simulation StAPPS).

Also described are SCDAP/RELAP5 and MELPROG/ TRAC calculations related to natural circulation in RCS (reactor coolant systems); receipt of plans for the final development of the VICTORIA in-vessel fission product release and transport code; incorporation into CORCON of new models for interfacial heat transfer, use of HMS BURN both as an experimental design tool and to perform blind post-test calculations of hydrogen distribution experiments in the Federal Republic of Germany's full scale HDR containment; publication of a description of the mechanistic core melt progression code, MELPROC, as part of a MELPROC user's manual for MELPROG/ TRAC Mod 1; direct containment heating uncertainty analyses performal foi the Surry, Sequoyah, and Grand Gul8 nudear plants using an enhanced edition of CONTAIN; completion of Version 1.8 of MELCOR, which analyzes accidents frem. initiating event through core degradation and vessel and containment failure, Version 1.5 of the MACCS code, which estimates the post-accident release of radioactive material and health and economic consequences to the public, and of the final production versions of SARA, the System Analysis and Risk Assessment system, and IRRAS, the Integrated Reliability and Risk Assessment System.

In addition, the report mentions the development of BLT, a low-level waste shallow land burial disposal source term code and its use in conjunction with FEMWATER and FEMWASTE to model groundwater flow and contaminant transport in unsaturated media and of SADDE (Scaled Absorbed Dose Distribution Evaluator) which calculates the necessary VARSKIN input data for any radionuclide, including daughters.

Publication Date:

April 1990 Prepared by:

NRC Office of Nuclear Regulatory Research Prepared for:

NRC Office of Nudear Regulatory Research Keywords:

research programs, reactor safety, regulations, nuclear power plants, earthquakes, ri3k assessment, failures, human factors, reactor cooling systems, reactor cores, radiation doses, skin, radionuclide migration, SCDAP/RliLAPS codes, RAMONA codes, HIPA codes, NUCLARR codes, IRRAS codes, CES codes, CREATE codes, MAPPS codes, MELPROG/ TRAC codes, CONTAIN codes, MELCOR codes, CORCON codes, MACCS codes, BLT codes, FEMWATER codes, FLMWASTE codes, SADDE codes, VARSKIN codes i

3

l NUREG-1272, Vcl. 4, No.1 G:n rcl

Title:

Office for Analysis and Evaluation of Operational Data 1989 Annual Report -

Power Reactors

==

Description:==

The annual report of the U. S. Nuc! car Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1989. The report is published in two separate parts. Volume 4, Number 1 covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report alu includes the principal findlags and issues identified in AEOD studies over the past year and summarizes information from such sources as lice..aee event reports, diagnostic evaluations, and reports to the NRC 5 operations Center. This report also compiles the status of staff actions resultin tom previous incident Investigation Team (llT) reports. Volume 4, Number : overs nonreactors and presents a revkw of the events and concerns during : 89 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Each volume contains a list of the AEOD reports issued for 1980-1989.

AEOD uses the Sequence Coding and Search System (SCSS) for storage and retrieval of licensee event report (LER) data. This system, developed in the early 1980s and maintained under contract at the Oak Ridge National Laboratory (ORNL), contains, on an average, ISO related pieces of data for each LER submitted since 1980. Its primary p(urpose is to facilitate the storage and retrieval of information about each event e.g., causal and time aspects of occurrences within the event sequence).

Subsequent to the accident at the Three Mile Island Nuclear Station (TMI) in 1979, international agencies developed an incident Reporting system (IRS) for the exchange c' information on events of particular safety significance. In 1989, the Paris based Nuclear Energy Agency (NEA) and the NRC finalized an agreement authorizing the NRC to assume the responsibility for th management and operation of NEA's IRS database. Also sub uent to the IMI accident, AEOD contracted with the Nudcar Operations Anal s Center at GRNL to develop a computer-based Foreign Event File (FEF). Th FEF, developed la 1980, is limited to events at foreign light water reactors with power output greater than 200 MWe.

Publication Date:

July 1990 Prepared by:

NRC Office for Analysis and Evaluation of Operaticnal Data Prepared fon NRC Office for Analysis and Evaluation of peranonal Data Keywords:

nudear power plants, reactor operation, :caeter safety, rer.aor accidents, performance, engineered safety systems, water cooled react:xs, water moderated reactors, SCSS codes, FEF codes, IRS codes 4

^

4 NUREG-1410 G:ntral j.

Title:

Special' Committee Review of the Nuclear Regulatory Commission's Severe

' Accident Risks Report (NUREG 1150)

==

Description:==

- In April 1989, the Nuclear Regulatory Commission's (NRC) Office of Nudear -

Regulatory Research (RES) published a draft report " Severe Accident Risks: An Assessment. for Five U.S. Nudcar Power Plants," NUREG 1150. This report 4

updated, extended,- and improved upon information presented in the 1974 "Keactor Safety Study," WASH 1400. Because NUREG-1150 will play a significant P

role - in implementing the NRC's Severe Accident Policy, its quality and credibility are of critical importance, Accordingly, the Commission requested a peer review of NUREG-1150 to ensure that the methods, safety insights, and conclusions presented are appropriate and adequately reflect the current state of knowledge with respect to reactor safety.

To this end, PES formed a special committee in June of 1989 under the provisions of the Federal Advisory Committee Act. The Committee, composed of a group of recognized national and international experts in nuclear reactor safety, was charged with preparing a report reflecting their review of NUREG 1150 with respect to the adequacy of the methods, data, analysis, and conclusions it set forth Probabilistic safety assessment of a nucicar plant can be done at three icvels, in level 1, the probability of severe damage to the reactor core, often equated to substantial or complete melting, is calculated. The results of a level 1 analysis d

are, then, principally the dominant accident sequences and the probabilities of different plant damage states, each of which cou!d vise from more than one accident sequence.

A level 2 PSA-tracks the fission products released from the different sequences

+

or damage states, to determine - the quantities, physical and chemical characteristics, and timing of their release from the containment building. These l

data are collectively called the source term.

A_ level 3 PSA continues the calculation through the dispersion of fission products through the available pathways, and calculates the consequences in such terms as damage to human health, land contamination and interdiction,.

and effects on the food chain.

A key objective of NUREG-1150 was to determine the level 2 uncertainties.

Unfortunately, the codes normally used ere large, detailed, and very expensive to run (e.g., the Source Term Code Package (STCP)). In fact, these codes were used for _ only a few calculaticens for each plant. Very simplified parametric codes, the XSOR codes, were used for the remainder.

Before NUREG-1150, the major tool for analysis of consequences in almost all risk assessments was the CRAC series of codes developed for WASH 1400.

NUREG-1150 employed the MELCOR Accident Consequence Code System (MACCS), a relatively new model still under development. CORCON 'was used to model the crosion of concrete by molten corium, and the Accident Sequence Evaluation Program (ASEP) procedure, based -heavily on the THERP methodology, was used to model the thinking processes of operators and their interaction with the plant system.

Publication Date:, August 1990 Prepared by:

Herbert J. C. - Kouts, George Apostolakis, E, H. Adolf Birkhofer, Lars G.

Hoegberg, William E. Kastenberg, Leo G. LeSage, Norman C. Rasmussen, Harry J. Teague, John J. Taylor

- Prepared for:

NRC Office of NucSar Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor safety, accidents, fission product release, source terms, reviews, CRAC codes, MELCOR codes, MACCS codes, XSOR codes, CORCON codes 5

NUREG/CP-0105, Vcl.1 G:nercl -

Title:

Seventeenth Water Reactor Safety Information Meeting Volu.ne 1 Held at Holiday Inn Crowne Plaza, Rockville, Maryland, October 23 25, 1989 Descriptioru This three-volume report contains 84 papers out of the 111 that were presented at the Seventeenth Water Reactor Safety Information Meeting held at the Holiday Inn Crowne Plaza, Rockville, Maryland, during the week of October 23-25, 1999. Volume 1 covers the following topics: Luncheon and Dinner Talks, Equipment Qualification of Valves, Generic Safety Issues Resolution, Human / System Interface and Persoenel Research, and Organization and Reliability Research.

Papers included w'. Ich contain information on computer applications are:

"Beyond Design B'. sis Accidents in Spent Fuel Pools - Cencric Issue 82", which describes SFUEllW and CLAD evaluation of spent fuel cladding failure and use of ORIGEN2, CRAC2, and MACCS in estimatmg releases and consequences for spent fuelpol accidents; "The Effects of Local Control Station Design Variation on Plant Kt,k", which discusses an Oak Ridge National Laboraty determination of LCS (1/, cal Control Station) design variations on human error probabilities by use of the Success Likelihood Index/ Multi-Attribute Utility. Decomposition (SLIM /MAUD) method; "A Method to Integrate Human Factors Expertise into the PRA Process", whi-h presents Lawrence Livermore National laboratory's experience in combining the NUCLARR database, SLIM /MAUD, CES, and MAPPS software, and THERPP and CREATE techniques as tools for integrating human factors expertise into the probabilistic risk assessment (PP A) process; "The Cognitwe Environmert Simulation as a Tool for Modeling - Human Perfo mance and Reliability", which evaluates Cognitive Environment Simulation (CES, and the Cognitive Reliability Assessment hchnique (CREATE) as means of 6mulating how people form intentions to act in nuclear power plant emergencies and measuring the human contribution to risk in PRA studies; and

- Extending the Evaluation of MAPPS: Results of User Participation", which discusses the results of a review and evaluation of the Maintenance Personnel Performance Simulation Model (MAPPS) conducted by the Idaho National Engineering Laboratory.

Publication Date:

. March 1990 Compued by:

A11cn J. Weiss Contractor:

Brookhaven National Laboratory, Upton, NY 11972 Prepared for:-

NRC Office of Nuclear Regulatory Research Keywords:

reactor safety, research programs, reactor accidents, reactor components, valves, reliability, human factors, risk assessment, systems analysis, performance, water cooled reactors, water moderated reactors, SFUEL1W codes, CLAD codes, ORIGEN2 codes, CRAC2 codes, MACCS codes, SLIM-MAUD codes, NUCLARR codes, CES codes, MAPPS codes a

6

i NUFEG/CP-0105, VCL 2 Gencrcl

Title:

Seventeenth Water Reactor Safety Information Meeting Volume 2 Held at Holiday Ir, Crowne Plaza, Rockville, Maryland, October 23 25, 1989 Descriptit,n:

This three-volume report contains 84 papers out of the 111 that were presented at the Seventeenth Water Reactor Safety Information Meeting held at the Holiday Inn Crowne Plaza, Rockville, Maryland, during the week of October 23-25, 1989. Volume 2 covers the following topics: Accident Manageraent, Severe Accident Research, Earth Sciences, Probabilistic Risk Assessment, and Seismic and Structural Engineering.

Papers included which contain information on computer applications are:

"Recriticality in a BWR Following a Core Damage Accident", which describes PNL MARCH calculations for station blackout scenarios and the use of NITAWL and XSDRNPM-S for determination of k-infinity and MCDAN for the Dancoff self-shielding correction; " Mitigation of Direct Containment Heating by Intentional Reactor Coolant S stem Depressurization", which discusses EC&G analyses based on SCDAI /RELAPS, MELPROC/ TRAC, and MELCOR;

" Reactivity Accidents

• A Reassessment of the Design Basis Events", which mentions BNL use of RELAP5/ MOD 2 and the BNL Plant Analyzer to assess event consequences; " Hydrogen Mixing Experiments in the HDR-Facility", in which hydrogen distribution analysis codes are treated; " Multi Compartment Hydrogen Deflagration Experiments in the Battelle-Frankfurt Model Containment", incorporating results of CONTAIN 1.10 calculations; " Bottom Head Failure Program Plan", noting the application of the fast reactor safety PLUGM code to the analysis of flow of molten materials into penetrations: " Fuel-Coolant hiteractions and Vapor Explosions; Recent Results and Related hsues", which discusses application of the two-fluid codes PHOENICS and CHYMES, the three-fluid model PM-ALPHA, and the TEXAi multi-fluid model; " Status of Cooperative Efforts on the VICTORIA Code for Fission Product Release and Transport", which describes VICTORIA's predecessors and sub-models;

" Assessment of Ex Vessel Steam Pressure Spikes in BWR Mark 11 Containments",

addressing BWRSAR calculations; and " Seismic Margin Assessment of Hatch Nuclear Power Plant", which comments on use of the CLASSI system of programs for soil structure interaction analysis and the SHAKE code to obtain strain-compatible dynamic soil properties.

Publication Date:

March 1990 Compiled by:

Allen J. Weiss Contractoc Brookhaven National laboratory, Upton, NY 11972 Prepared fon NRC Office of Nuclear Regulatory Research Keywords:

reactor safety, research programs, reactor accidents, carthquakes, seismic events, probabilistic estimation, risk assessment, water cooled reactors, water moderated reactors, nuclear power plants, soil-structure interactions, flow models, MARCH codes, NITAWL codes, XSDRNPM-S codes, MCDAN codes, SCDAP/RELAPS codes, MELCOR codes, RELAPS/ MOD 2 codes, COBRA-NC codes, HECTR codes, RALOC codes, WAVCO codes, CONTAIN codes, PLUGM codes, PHOENICS codes, CHYMES codes, PM-ALPHA codes, TEXAS codes, VICTORIA codes, BWRSAR codes, CLASSI codes, SHAKE codes I

l 7

NUREG/CP-0105, Vcl. 3 Genarcl

Title:

Seventeenth Water Reactor Safety Information Mectng Volume 3 Held at Holiday Inn Crowne Plaza, Rockville, Maryland, October 23-25, 1989 Descriptioru This three-volume report contain$ M papers out of the lit tha, were presented at the Seventeenth Water Rea: tor Safety Information Meeting held at the Holiday Inn Crowne Plaza, Rockville, Maryland, dunng the week of October 23 25. 1989. Volume 3 covers the followir$ topics: Primary Systems Integrity; Plant Performance, Testing and Analysis; Pipmg and NDE; and Plant Aging.

Papers included which contain information on computer applications are:

" Heavy-Section Steel Technology Program Fracture issues", which mentions use of OCA-P to analyze flaw depths of ductile tearing; " Power Reactor Embrittlement Data Base", a discussion of the PR-EDB database; " Future Plans for NRC Thermal-Hydraulle Research", which describes the Commission's plans for LWR codes RELAP, TRAC-PWR, RAMONA, COBRA, TRAC-BWR, and HIPA: "'Ihe U.K. Contribution to Improvements in TRAC and RELAP5"; " Code improvements Based on Results from the 2D/3D Activities", which covers JAERI's develorment of the REFLA core model and its installation in TRAC-PF1/ mob 1; and " Instrument Air System-Aging impact on System Availability", which documents the development of the PRAACE-IA system model for time-dependent probabilistic risk assessment calculations.

Publication Date:

March 1990 Compiled by:

Allen J. Weiss Contracion Brookhaven National Laboratory, Upton, NY 11972 Prepared fon NRC Office of Nuclear Regulatoiy Research Keywords:

reactor safety, research programs, primary coolant circuits, nuclear power plants, performance testing, planning, pipes, aging, nondestructive testing, water cooled reactors, water moderated reactors, reactor cores, probabilistic estimation, risk assessment, OCA-P codes, PR-EDB codes, TRAC-PWR codes, TRAC-BWR codes, RELAPS codes, RAMONA codes, HIPA codes, COBPA codes, REFLA codes, J-TRAC codes, PRAAGE-IA codes u

o 8

NUREG/CP-0110 Gensral UNL-NUPoEG-52226

Title:

Proceedings of the International Workshop on New Developments in Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities Held at Brookhaven National Laberatory, Upton, Long Island, New York September 18-21, 1989

==

Description:==

This report contains summaries of papers and discussions prosented at the Commission and the U. ponsored jointly by the U. S. Nuclear Regulatory International Workshop s S. Department of Energy, in cooperation with the Organization for Economic Cooperation and Development, Nuclear Energy Agency, and held at Brookhaven National Laboratory, Upton, New York, September 18 21, 1989. The workshop brought together scientists, engineers, regulators, and admin!strators involved with occupational dose control at nuclear facilities. The eleven countries represented were: Canrda, Finland, France, Germany, Italy, Japan, Luxembourg, Sweden, Switzerland, the United Kingdom, and the United States of America.

The following computer software was described:

DOSI-ANA, a microcomputer code, currently in use in French PWRs, incorporating a database management system (EXADATA) and consisting of a set of easy to-use program moduler to plan, follow up, and evaluate operations in nuclear power plants from an occupational dose control perspective.

ORE (Operator Radiation Exposure) database, which contains collective dose statistics from PWRs around the world. The ORE database was developed by NNC Ltd, England on a personal computer using dBASE III Plue.

DYDAS, a code to estimate gamma ray sou ces contained in a complex physical structure from measured dose rate surveys and from these calculate gamma dose rates for different plant states and locations. DYDAS was developed by Rolls-Reyce and Associates Limited, Derby, England RPS, Radiological Plant Status, developed jointly by Combustion Engineenng and Duke Power Company to automate documentation of survey records. RPS allows users to display radiation survey results graphically.

The Dynamic Dose Tracking Code, a PC-based program designed by tr"rence Livermore National laboratory to track the radiation exposuras of operabrs in a nudcar fuel cycle or plutonium manufacturing facility.

The NEA ISOE (Information System on Occupational Exposure) database, being developed at the CEPN in France for a PC or compatible microcomputer, based en the EXADATA database management system. It will hold information on commercial nuclear power plants of the type existing in NEA/OECD member coumries. In a NEA ISOE pilot study, the BNL ALARA Center has focused on dose control collaborative information, while CEPN has concentrated on the dosimetry data collection aspect.

Publication Date:

February 1990 Compiled by:

J. W. Baum, B. J. Dionne, T. A. Kahn Contractor:

Brookhaven National Laboratory, Upton, NY 11973 Prepared for:

NRC Office of Nuclear Regulatory Resecrch, U. S. Department of Energy, and OECD Nuc! car Energy Agency Keywords:

ALARA, occupational exposure, radiation doses, dow rates, nuclear power plants, NEA, intemational cooperation, proceedings, DOSI-ANA codes, DOSI-ECO codes, ORE codes, DYDAS codes, RPS codes, Dynamic Dose Tracking

< oles, ISOE codes t

l 9

r NUREGICP-0113 G:n:r-1

Title:

Transactions of the Eighteenth Water Reactor Safety Information Meeting To Be Held at Holiday Inn Crowne Plara, Rockville, Maryland, October 22-24, 1990 Descriptione nis report contains summaries of papers on reactor safety research to be presented at the 18th Water Reactor $afety Information Meeting at the Holiday mn Crowne Plaza in Rockville, Maryland, October 22-24, 1990. The summaries briefly describe the programs and results of nuclean safety research sponsored by the Office of Nuclear Regulatory Research, U.

S.

Nuclear Regulatory Commission. Summaries of invited papers concerning nuclear safety issues from the ciectric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summarier have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the Meeting, and are given in the order of their presentation in each session.

Among computer codes and databases discussed are the following:

The DEBRIS module prepared at Sandia National laboratories to solve the two-dimensional (r,z) momentum equation to account for rc. cit relocation due to both gravitational and capillary forces, the continuity equation to assure mass conservati*. ad the energy equation.

CARES (Computer Analysis for Rapid Evaluation of Structures), a microcomputer-based integrated system being developed by Brookhaven National Laboratory to determina the validity and accuracy of analysis methodologies used for structural safety evaluations of nuclear power plants.

MICROMAPPS, an Idaho National Engineering Laboratory microcomputer version of the comput?r model MAPPS (Maintenance Personnel Performance Simulation), designed to improve maintenance practices and procedures at nuclear power plants. A number of user enhancements were identified and implemented as part of the conversion process, VICTORIA, a Sandia National laboratories mechanistic computer code designed to analyze fission product behavior within the reactor coolant system during a severe accident, it proddes detailed prediction of the release and transport of radionuclides and non-radioactive materials during core degradation.

A databam of seismic fragility of various equipment classes, especially those that are important for safety of a nuclear power plant and, at the sanie time, have low seismic capacities compiled by Brookhaven National Laboratory. Equipment included is the motor control center, switchgear, panclboard, switchboard, power supply, NSSS 1&C panels, transmitters, indicators, switches, transformers, BOP I&C panels, miscellaneous instruments, Latteries, battery chargers, inverters, motors, and electrical penetration assemblics.

Other computer codes mentioned in the summaries are: OCA-P, MELCOR, SAFT-UT,

STCP, VAM2D, VAN ESA,

CORCON, TRAC-PF1/ MOD 2, RELAPS/ MOD 3, TRAC-BF1/ MOD 1, Nuclear Plant

Analyzer, LAPUR, Engineering Plant Analyzer, NRCPIPE, SQUIRT, and RAMONA-3B.

Publication Date:

October 1990 Cornpiled by:

Allen J. Weiss Prepared fon NRC Office of Nuclear Regulatory Research Keywords:

reactor safety, research programs, radiation protection, DEBRIS codes, CARES codes, MICROMAPPS codes, VICTORIA codes, OCA-P codes, MELCOR codes, SAFT-UT codes, STCP codes, VAM2D codes, VANESA codes, CORCON codes, TRAC-PF1/ MOD 2 codes, RELAP5/ MOD 3 codes, TRAC-DF1/ MODI, NPA codes, EPA todes, LAPUR codes, NRCPIPE codes, SQUIRT codes I

10

=

- NUREG/CR-2331, VcL9 Nc3 G:n::r.cl

'BNL-NUREG-51454, VcL9 Nc3 Tlile:

Safety Research Programs Sponsored by Office of Nuclear Regulatory Research Progress Report July 1 September 30,1989

==

Description:==

This progress report describes current activities and tedsnical progress in the programs at Brookhaven National Laboratory sponsored by the Divisions of Regulatory Applications, Engineering, Safety Issue Resolution, and Systems Research of the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research followin the reorganization in July 1988. Topics treated are the following: Accident Ana sis and Safety Review 'of Liquid Metal and High Temperature Gas Reactors (

Rs and HTCRs); Surveillance of Industry / DOE Dose-Redaction Research and ALARA Engineering; Technical Assistance for Llansing Irradiated Cemstenes; Quality Assurance Pilot Program in Medical Use of Byproduct Material; Severo Acddent Policy Implementation Support; Hot Particle ' Production, Mitigation and Dosimetry; - TA Licensing Monoclonal Antibodies; impact of Reduction of Occupational Dow Limits on NRC Licensecs; Computcr Evaluation of Structural Behrvior and Capability; Aging Components and System 11; Comparison of Foreign and Domestic Processes Which Contribute to Occupational Dose; Essential Service Water Pump Failures at Multiplant Sites (Cencric issue 130); Ice Condenser Containment Performance; Dry Containment Performance; IPE Submittal Overview / Review Guidance and Coordination; Direct Containment Heating Experiments and Model Development; MELCOR Verification and Benchmarking; Thermodynamic Core-Concrete Interaction Experiments and Analysis; Interfadal Heat Transfer for Core-Concrete Interactions; Safety Evaluation of Core-Melt Accidents; SP-90 PRA; Review of Diablo Canyon PRA: Maintenance of BWR Plant Analyzer; Procedures for Evaluating Technical Specifications (PETS;; Operational Safety Reliability Research; Risk Based Performance Indicators; Influenm of Supervisor / Manager Factors on Performance Reliability; Evaluation of Severe Accident Phenomena; Reactivity Accidents; Generic Issue 43: Air System Reliability; Annunciators; Local Control Stations; Application of RAMONA-3B and tl e BNL Engineering Plant Analyzer to BWR Stability; Risk Review of CESSAR 80 Plus Design; Analysis of Containment Protective Action: Containment and Release Management; Severe Accident Scaling Methodology Program; Organizational and Management Research Support to Accident Management; Risk Review of ABWR Design.

Previous reports covered the period October 1,1976 through June 30, 1989.

Comments are induded on the following codes: SYMELAN, THATCH, SSC, BNL

- Hot Particle, CARES, PSD, CONTAIN-DCH, MELCOR, SCDAP/RELAPS, CORCON, BWRSAR, CONTAIN, CORCON MOD 2, VANESA, RELAPS/ MOD 2, SLIM-MAUD, and the BWR Engineering Plam Analyzer software.

Publication Date:

February 1990 Contplied by:

Allen J. Weiss Contractor:

Brookhaven National Laboratory, Upton, NY-11973 Prepared for:

NRC Divisions of Regulatory Applications. Engineering. Safety issue Resolution, and Systems Research, Office of Nuclear Regulatory Research Keywords:

reactivity insertions, reactor safety, research programs, LMFBR type reactors, BWR type reactors, PWR type reactors, reactor accidents, radiation doses, dose limits, irradiation procedures, by-products, monoclonal - antibodies, structural models, aging, reator components, ALARA, pumps, ice condensers, containment systems, reliability, human factors, probabilistic estimation, risk assessment, heat transfer, management, CARES codes, PSD codes, MELCOR codes, RAMONA-3B cJ,les, BNL Engineering Plant Analyzer codes a

11

e NUREG/CR-2331, Vol.9. No.4-G:ntral BNL-NUREG-51454,-Vol.9 No.4

Title:

=

Safety Research Programs Sponsored by Office of Nudear Regulatory Research Progress Report October 1 December 31, 1989

==

Description:==

This progress report describes current activities and technial progress in the programs at Brookhaven National Laboratory sponsored by the Divisions of Regulatory Applications, Engineering, Safety Issue Resolution, and Systems 4

Research of the U. S. - Nuclear Regulatory Commission, Office of Nudcar

-o' Regulatory Research following the reorganization in July 1988. Topics treated are the following: Accident Analysis and Safety Review of Liquid Metal and High Temperature Gas Reactors; Surveillance of Industry / DOE Do e-Reduction Research.and ALARA Engineering; Technical Assistance for Licensing Irradiated Gemstones; Quality Assurance Pilot Pmgram in Medical Use of Byproduct Material; Sevem-Accident Policy Implem".tation Support; Hot Particle Production, Mitigation and Dosimetry Technical Assistance for Licensing Monoclonal Antibodies; Impact of Reduction of Occupational Dose Limits on f

NRC Licensees; Computer Evaluation of Stru-tural Behavior and Capability; Aging Components and System 11; Evaluation of the Adequacy of Curcent Reactor Coolant Pump Seal Instrumentation and Operator Responses to PosstNe Reactor Coolant Pump Seal Failures; Essential Servicu Water Pump Failures at Multiplant Sites; Ice Condenser Containment Performance, Dry Containment Performance, IPE Submittal Overview / Review Guidance and Coordination; Direct Containment - Heating; MELCOR Verification and Benchmarking:

Development - of an Advanced Control Room - Design Review Guideline; Interfadal Heat Transfer for Core Concrete Interactions; Safety Evaluation of Core-Melt Accidents: SP-90 PRA: Review of Diablo Canyon PRA: Maintenance of BWR Plant Analyzer; Procedures for Evaluating Technial Specifications (PETS);

Operational Safety Reliability Research; Risk-Based Performance Indicators; Influence of Organizational Factors on Performance Reliability; Evaluation of

- Severe Accident Phencmena; Reactivity Accidents; PWR Lower and Shutdown Accident. Frequencies; Annunciators; Le 1 Control Stations; Application of the BNL Engineering Plant Analyzer and mAMONA.3B to BWR Stability; Risk Review of CESSAR 80 Plus Design; Containment and Rebase Management; Severe Accident Scaling Methodology Program; Organizational and Management Research Support to Accident Management; and Risk Review of ABWR Design.

Previous reports covered the peri <xl October 1,1976 through September 30, 1989.

Comments are included on :he ALARA Center's databases and the following codes: CORCON, MELCOR, RETRAN, HIPA-PB2, HIPA BWR4, TRAC-BDI, RELAP5, RAMONA 3B, RELAPS/ MOD 2, SISOR, and BWR Engineering Plant Analyzer software.

~

Publication Date:- June 1990.-

1 Compiled by:

Allen J. Weiss L

Contractor:

Brookhaven National Laboratory, Upton, NY 19973 L

Prepared lor:

NRC Office of Nuc! car Regulatory Research l

- Keywords:

reactivity insertions, display devices, control rooms, computerized simulation, i

human factors, BWR type reactors, HTGR type reactors, LMFBR type reactors,-

PWR type reactors, reactor safety, research programs. radiation doses, dose

-limits, - irradiation procedures, ALARA, - structural models, aging, reactor p

components, reactor cooling systems, pumps, failures, reliability, progress report, meltdown, corium, risk assessment, probabilistic estimation, spent fuel storage,-

fuel pools, ice condensers, containment systems, licensing, monoclonal antibodies, occupational safety, design basis accidents, maximum credible accident, reactor accidents, radioactive effluents, heat transfer, management, MELCOR codes, BNL g

Engineering Plant Analyzer codes, RAMONA-3B codes 12

NUREG/CR-4214, Rev.1, Pt.I MACCS SAND 85-7185, R:v.1; Pt.1

- 1

Title:

~

' Health Effects Models for Nuclear Power Plant Accident Consequence Analysis

- Low 1.ET Radiation Part I: Introduction, Integration, and Summary Descriptioru This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided.

Two-paramevr Weibull hazard functions are recommended far estimating the risks of carly and continuing health effects. Three potentially lethal early effects-the hematopoietic, pulmonary, and gastrointestinal syndromes--are considered, in addition, models are induded for assessing the risks of several non-lethal early and continuing effects -including prodromal vomiting and diarrhea, hypothyroidism and radiation thyroiditis, skin burns, reproductive effects, and spontaneous abortions.

Linear and linear-quadratic models are recommended fcr estimating cancer risks.

Parameters are given for analyzing the risks of seven types of cancer in adults-leukemia, bone, lung, breast, gastrointestinal, thyrold, and "other" The category "other" cancers is intended to reflect the combined riskt of multiple myeloma, lymphoma, and cancers of the bladder, kidney, brain, ovary, uterus, and cervix. Models of childhood cancers due to in utero exposure are also provided. For most etnce-s, both incidence and mortality are addressed. The models of cancer risk are derived largely from informatior summarized ir BEIR lil-with some adjustment to ref!cct more recent studies. The effect of thc revised dosimetry in Hiroshima and Nagasak, has not been considered.

Linear and linear-quadratic models are also recommended for assessing genetic risks. Five classes of genetic disease--dominant,.x linked, ancuploidy, unbalanced translocations, and multifactorial diseases-are considered. In addition, the impact of radiation induced genetic damage on the incidence of peri-implantation embryo losses is discussed, i

i The uncertainty in modeling radiological health risks is addressed by providing central, upper, and lower estimates of all model parametera. Data are provided which should enable analysts to consider the timing and severity of each type of health risk.

A section of the report covers issues related to the computer implementation and mathematical derivation of certain health effects models. The structure of -

the MACCS nuclear power plant accident consequence code is based on the health effects models recommended in the first edition of NUREC/CR-4214. The

- risks of - all early and-continuing effects - are computed indirectly using two-parameter -Weibull hazard functions. The effect of dose rate on risk is accommodated using different values of the median lethal or effective dose to compute the risk from dose received in different time intervals following the accident. Approaches are outlined for implementing the new heahn effects:-

models in MACCS.

_ Publication Date:

January M Prepared by:

J. S, Evans Contractoc Harvard University, Harvard School of Public Health, 665 Huntington Avenue, Boston, MA 02115 under contract to Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185

- Prepared fon NRC Division of Regulatory Applications, Office of Nuclear Regulatory Research Keywords:

nuclear power phnts, accidents, dose-response relationships, health, dose L

equivalents, LET; carly radiation effects, delayed tadiation effects, genetic radiatica cifects, CRAC codes g

NUREG/CR-4469, Vel 10 NPRDS Drtabase PNL-5711, Vcl.10

Title:

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors Semi Annual Report October 1983. March 1989 Descriptforu The Evaluation and improvement of NDE (r.ondestructive examination)

Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability)

Program at the Pacific Northwest Laboratory was established by the Nudear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of 151 yerformed on the primary systems of commercial light water reactors (LWKs); using probabilistic framire mechanics analysis to determine the impact of NDE unreliability en system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and regulatory requirements based or. n terial prooerties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems inNM the piping, vessel, and other components inspected in accordance with f.

9 of the ASME Code.

covering the programmatic work from October 1988 This is a progress re n

through March 1989. In response to the recommendation from a 1987 workshop that past operating experience be utilized as a basis to develop improved inspection criteria, an evaluation was completed luring this period of a set of data on pi in failures extracted from the Nuclear Power Plant Reliability System (N

3) database. The evaluation began with a listing from a computerized search-for all the reported piping system failures in the NPRDS database. This provided some 412 individual reports of failures. Since utilities report informatton for entry into the NPRDS database on a purely voluntary basis, the level of completeness of reporting varies from entry to entry. Each individual report was reviewed and interpreted, and then essential information was coded and entered into a compute- '

A FORTRAN program'was written to generate a number of useful tab #d the coded data.

-In spite of significant limitations, -

RDS database is regarded as the best available compilation of informatfori. 5ts type at this time. This report includes various tabulations of the NPRDS data relating to piping system failures, e.g.,

date of discovery for incident reports; number of reports filed per plant; number of reports of each pip;cg failure mode, material :ype, vendor, discovery method, PWR system, BWR system, or pipe diameter.

Publication Date:

September 1990 '

Prepared by:

. S. R. Doctor, J. D. Deffenbaugh, M. S. Good, E. R. Green, P. G. Heasier,~ F. A.

Simonen, J. C. Spanner, T. T. Taylor, T. V. Vo Contractorf Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352 Prepared for:

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

nondestructive analysis, nondestruct ve testing, water cooled reactors, water i

moderated reactors, stainless stects, welded joints, fracture mecFanics, pipes, ultrasonic testing, performance testing, reliability, reactor safety 14

i NUREGICR4550,V3,R1,Pil TEMAC, LHS, SETS SAND 86-2084,V3,R1,Pt1

Title:

Analysis of Core Damage Frequency: Surry, Unit 1 Internal Events Descriptioru This document contains the accident sequence analyses of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nudear Regulatory Commission. NUREG-1150 documents the nsk of a selected group of nuclear power plants. The work performed and described here is an extensive reanalysis of that published in November 1986 as NUREG/CR-4550, Volume 3. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA (Probabilistic Risk Assessment) practitioners who need to know how the work was performed and the detalls for use in further studies. The mean core damage frequency at Surry was calculated to be 4.0E-5 per year, with a 95% upper bound of 1.3E-4 and 5% lower bound of 6.8E-6 per year. Station olackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were loss of coolant accidents (LOCAs). These sequences account for 15% of cure damage frequency. No other type of sequence accounts for more that 10% of core damage frequency.

The primary objective was to perform an analysis to support the NUREG-1150 project that is as near to a state-of the-art Level 1 Probabilistic Risk Assessment (PRA) as possible. A standard but focused Level 1 PRA approach formed the basis for this analysis. Event trees were constructed; the top events were modeled using large fault trees, and the results were quantified using the Set Equation Transformation System (SETS) and the Top Event Matrix Analysis Code (TEMAC) computer codes. TEMAC uses the parameter samples generated by the Latin Hypercube Sampling (LHS) code and the accident sequence equations (cut sets) as input to quantify the core damage estimates. TF.MAC generates a sample of the accident sequence frequency, a point estimate of the frequency, and various imnortance measures and ranking for the basa events.

LHS calculates probability distributions to the 99th percentile.

Publication Date:

April 1990 Prepared by:

R. C. Bertucio, J. A. Julius Contractor:

E.1. Services,1851 South Central Place, Suite 201, Kent, WA 98031 under con'ract to Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185 Prepared for:

NRC Divi <lon of Systems Rescatch, Office of Nuclear Regulatory Research Keywords:

probabilistic estimation, risk assessment, systemt analysis, reactor accidents, reactor cores, blackouts, loss af coolant, failure mode analysis, Surry-1 reactor 1

15

~..

NUREG/CR-4550, V3,R1,Pt3 G:n :r:1 SAND 86-2084, V3,R1,Pt3

Title:

Analysis of Core Damage Frequency; Surry Pawer Station, Unit 1 External Events -

Descriptioru This report presents the analysis of external events (earthquakes, fires, floods, etc.) performed for the Surry Power Station as part of the USNRC sponsored NUREG 1150 program. Both the internal and external events analyses make full us, of recent insights and developments in risk assessment methods, in addition, the external event analyses make use of newly-developed simplified methods.

As a first step, a screening analysis was performed which showed that all external events were negligible except for fires and seismic events. Subsequent detailed analysis of fires resulted in a total (mean) core damage frequenej of 1.13E-5 per year. The seismic analysis resulted in a total (mean) core damage frequency of 1.16E-4 per year using hazard curves developed by L.awrenm Livermore National Laboratory and 2.50E 5 per year using hazard curves developed.by the Electric Power Research Institute. Uncertainty analyses were rformed, and dominant components and sources of uncertainty were-Fdentified.

To obtain estimates of the scismic response of structures and components, SHAKE code calculations were performect to assess the effect of the local soil column (if any) on the surface peak ground acceleration and soll structure interactions, Fixed base mass-spring (eige ystem) models obtained either from the plant's architect /enginect, or develo from the plant drawings as needed, were used to compute the floor slab ace crations with the CLASSI code.

Quantification of the accident sequences is a multi step procedure involving several levels of screening. In the first step, the SETS code was used to evaluate

- all potential accident sequences using pomt estimate input screening values for all seismic failure evants (and using the internal events point estimate failure values for all random events).

The COMPilRN fire growth code was used to calculate fire propagation and equipment damage. COMPBRN, developed specifically for use in nuclear power plant fire probabilistic risk assessments, calculates the time to damage critical equipment given that a fire has started. Th.s failure time is then used in conjunction with experimental information on fire suppression in nuclear power plants to obtain the probability or frequency that a given fire will cause damage which leads to core damage before the flre can be suppressed.

Publication Date:

December 1990 Prepared by:

M. P. Bohn, J. A. Lambright, S. L Daniel; J. J. Johnson, M. K. Ravindra, P. O.

Hashimoto, M. J. Mraz, W. H. Tong Contractor:

Scadia Nationa Laboratories, P. O. Box 5800, Albuquerque, NM 87185: EQE, incorporated, San Francisco, CA Prepared for:

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

Surry-1 reactor, probabilistic estimation, risk assessment, fire hazards, scismic events, safety, accioents, reactor cores, soil structure interactions, SHAKE codes, CLASSI codes, SETS codes, COMPBRN codes 16

'NUREG/CR-4550, V4.R1,Pt3 G:nzr:1 -

SAND 86-2084, V4,R1,Pt3

Title:

. Analysis of Core Damage Frequency: Peach Bottom, Unit 2 External Events

==

Description:==

This report presents the analysis of external events (earthquakes, fires, floods, etc.) performed - for the Peach Bottom Atomic Power Station as part of the USNRCeponsored NUREC 1150 program. Both the internal and external events analyses make full use of recent insights and developments in risk assessment methods, in addition, the external event analyses make use of newly-developed simplified methods.

As a first step, a screening analysis was performed which showed that all external events were negligible except for fires and seismic events. Subsequent detailed analysis of fires resulted in a total (mean) core damage frequency of 1.95E 5 per year. The scismic analysis resulted in a total (mean) core damage frequency of 7.66E 4 per year using hazard curves developed by Lawrence

- Livermore National Laboratory and 3.09E-6 per year using hazard curves developed by the Electric Power Research Institute. Uncertainty analyses were performed, and dominant components and sources of uncertainty were identified.

To obtain estimates of the scismic response of structures and components, SHAKE code calculations were performed to assess the effect of the local soli column (if any) on the surface peak ground acceleration and soll structure interactions. Fixed base mass-spring (eigensystem) models obtained either from the plant's architect / engineer, or developed from the plant drawings as needed, were used to compute the floor slab accelerations with the CLASSI code. The SETS code and mean basic event frequencies were used to calculate nican probabilities of failure, core damage, etc.

De COMPBRN fire growth code was used to calculate fire propagation and equipment damage. COMPBRN, developed specifically for use in r.uclear power plant fire probabilistic risk assessments, calculates tne time to damage critical equipment give.4 that a fire has started. This failure time is then used in conjunction with experimental information on fire suppression in nuclear power plants to obtain the probability or frequency that a given fire will cause damage whleh leads to core damage before the fire can be suppressed.

Fire core damage uncertainty analyses were performed using two computer codes:. LHS-(Latin Hypercube Samoling) to generate the samples for all of the parameter values and TEMAC (Top' Event Matrix Analysis Code) to quantify the uncertainty of the accident se generated by the LHS code. quence equati_on using the parameter value sa LHS is a con., trained Monte Carlo technique which forces all parts of the distribution to be sampled. Using LHS parameter samples

~

and 1.he accident sequcnce equations (cut sets) as input to quantify the core damage estirrates, TEMAC generates a sample of the accident - sequence frequency, a point estimate of the frequency, and various importance measures and rankinb or the base events.

f Publication Date:

December 1990 Prepared by: -

J. A. bmbright, M. P. Dohn, S. L. Daniel; J. J. Johnson, M. K. Ravindra, P. O.-

Hashimoto, M. J. Mraz, W. H. Tong; D A. Brosseau Contractor:

Sandia National bboratories, P. O. Box 5800, Albuquerque, NM 87185; EQE, Incorporated, San Francisco, CA; ERCE, Inc., Albuquerque, NM

. Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatury Research

. Keywords:

Peach Bottom-2 reactor, probabilistic estimation, risk assessment, fire hazards, seismic events, safety, accidents, reactor cores, soil-structure interactions, SHAKE codes, CLASSI codes, SETS codes, COMPBRN codes, LHS codes, TEMAC codes 17

m NUREG/CR-4550,V5,R1,Pt1 TEMAC, LHS, SETS SAND 86-2084,V5,R1,Pt1

Title:

Analysis of Core Damage Frequency: Sequoyah, Unit 1 Internal Events Descriptioru This document contains the accident sequence analyses of internally initiated events for the Sequoyah Ualt I nudcar power plant. This is one of the live plant analyses conducted as part of the NUREG-1150 effort by the Nudcar Rcgulatory Commission. NUREG 1150 documents the risk of a selected group of nuclear power plants, The work performed and des.cribed here is an extensive reanalysis of that published in February 1987 as NUREC/CR-4550, Volume 5. It

- addresses comments from numerous reviewers and significant changes to the plant systems and procedures made sincn the first report. The uncertainty analysis and presentation of results are also much improved.

The mean core damage frequency at Sequayah was calculated to be 5.7E 5 per year, with a 95% upper bound of 1.8E-4 and 5% lower bound of 1.2E-5 per year.

less of coolant type accidents (LOCAs) were the largest contributors to the core damage frequency, accounting for approximately 62% of the total. The next most dominant type of accidents were station blackout (loss of all AC power), which accounted for 26% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency.

The numerical results are dominated by failure to initiate high pressure redrculation - due to operator error - following loss of coolant accidents.

Considerable effort was expended on the modeling of very small LOCAs and station blackout sequences, including the development of a reactor coolant pump sen! LOCA model through elicitation of expert opinion.

This: report evaluates core damage frequency from internally inititated events.

The consequences of these accidents are evaluated and reported under separate cover.

The primary objective was to perform an analysis to support the NUREG 1150 project that is as near to a state-of the-art Level 1 Probabilistic Risk Assessmant 1PRA) as possible. The standard Level 1 PRA approach was used in the analys s.

Event trees were constructed, the top events were modeled using large fault trees, and the results were quantified using the Set Equation Transformation System (SETS) and the Top Event Matrix Analysis Code (TEMAC) computer codes. TEMAC uses the parameter samples generated by (the Latin Hypercube Sampling (LHS) code and the accident sequence equations cut sets) as input to quantify the core damage estimates. TEMAC generatas a sample of the accident sequence frequency,kinga point estimate of the frequency, and various impo measures and ran for the base events. LHS calculates - probability distributions to the 99th percentile.

Publication Date:

April 1990 Prepared by:

R. C. Bertucio, S. R. Brown -

Contractor:

E.1. Services, 151 South Central Place, Suite 201, Kent, WA 98031 under contract to Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185 Prepared for:

NRC Division of Systems Research, Office of Nuc! car Regulatory Research EKeywords:

probabilistic estimation, risk assessment, systems analysis, reactor accidents, reactor cores, blackouts, loss of coolant, failure modo analysis, Sequoyah 1 reactor 18

NUREG/CR-4551, V2,R1,Pt1 G nzrni SAND 86-1309, V2,R1,Pt1

Title:

' Evaluation of Severe Accident Risks: Quantification of Major input Parameters Expert Opinion Elicitation on in Vessel issues

==

Description:==

In support of the Nudcar Regulatory Commission's (NRC's) assessment of the risk from sevue accidents at commercial nudear power plants in the U.S.

reported in NUREG 1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from f.evere accidents at five nuclear power plants: Surry, Sequoyah, Zion, Peach Bottom, and Grand Gulf.

The emphasis in this risk analysis was not on determining a "so-called" point estimate of risk. Rather, it was to determine the distribution of risk and to discover the uncertainties that account for the breadth of this distribution.

Off-site risk initiation by events, both internal to the power station and external to the power station, was assessed.

Much of the important input to the logie models was generated by expert panels. This document presents the distributions and the rationale supporting the distributions for the quesuons posed to the In-Vessel Expert Panel.

In part, panel experts based their input on review of the results of calculations performed with a

number of ' mmputer codes.

Each expert's ratianale/ methodology is expressed in det.1 in Section 5 of the report entitled, "Results of the Elicitation on Each In-Vese.1 Issue." Codes described as producing results pertinent to the aperts' analyser meluded MELPROG, TRAC /MELPROC, CORMLT/PSAAC, RELAP5/SCDAP, MAAP, BWRSAR, APRIL, and MARCH.

Publication Date:

December 1990 Prepared by:-

F. T. Harper, R. J. Breeding, T. D. Brown, J. J. Gregory, A. C. Payne, E. D.

Gorham; C. N. Amos Contractor:

Sandia National laboratories, P. O. Box 5800, Albuquerque, NM 87185; Science Applications International Corporation, 2109 Air Park Road SE, Albuquerque, NM 87106 Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywneds: -

probabilistic estima9on, risk assessment, reactor safety, accidents, reactor vessels, failures, hydrogen production, steam generators, ruptures, reactor cooling systems, MELPROC codes, TRAC /MELPROG - codes, RELAP5/SCDAP codes,

~ CORMLT/PSAAC codes, MAAP codes, BWRSAR codes, MARCH codes

'n+

19

i NUREG/CR-4551, V2,R1,Pt7 MACCS

- S AND864309, V2,R1,Pt7 -

Title:

Evaluation of Severe Accident Risks: Quantification of Major input Parameters

--MACCS Input

==

Description:==

Estimation of offsite accident consequenas is the customary final step in probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five U.- S. power reactors (NUREG-1150). Offsite accident consequenms for NUREG-1150 source terms were - estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value and an uncertainty range were recommended.

This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and ' biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report.

MACCS calculations require the following data The inventory at accident initiation (reactor scram) of those radioactive isotopes important for the calculation of ex-reactor core contains about 10 Ci of " plant consequences (e.g., an end-of-cycle 8

1).

The atmospheric source - term produced by the accident (number of plume segments released, sensible heat content of each segment, time and duratlon of release, time when offsite officials are warned that an emergency response -

i should be initiated, and the fraction of each important nuclide's scram inventory i

released with each segment).

Meteorolo cal data characteristic of the site region - usually one year of hourly winds

, atmospheric stability, and rainfall readings recorded at the site or a

' nearb National Weather Service station.

The population distribution about the reactor site.

Emergency response assumptions.

Land usage -(habitable land fractions, farmland fractions) and economic data (worth of crops, land, and buildings) for the region about the reactor site.

Publication Date:

December 1990

- Prepared by:

J. l Sprung, H N Jow; J. A. Rollstin; J. C. Helton Contractor: -

Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185; CRAM, Inc.,1709 Moon NE, Albuquerque, NM 87112; Arizona State University, Tempe, =

AZ 85287 Prepared for:

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

reactor accidents, earth atmosphere, deposition, dosimetry, economics, scram, radionuclide migration, evacuation, mitigation, environmental exposure pathway, nuclear power plants, radioactive materials, fission product release, emergency plans, radiation protection -

20

NUREG/CR-4551, V3,R1,Pt1 G!n rel SAND 86-1309, V3,R1,Pt1

Title:

1 Evaluation of Severe Accident Risks: Surry Unit 1 Main Report Descriptioru In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from r,cvere accidents at commercial nuclear power plants in the U.S.

reported in NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) completed a revised calculation of the risk to the general public from severe accidents at the Surry Power Station, Unit 1. This power plant, located in southeastern Virginia, is operated by the Virginia Electric Power Corporation.

Off site risk initiation by events, both internal and external to the power station, was assessed.

The risk assessment, on which NUREG 1150 is based can generally be 1

characterized as consisting of four analysis steps-accident frequency, accident progression, source term, and consequence analysis; a risk integration step; and an uncertainty analysis step. This volume presents the details of the last five of the six steps for the Surry Power Station, Unit 1. The first step is describ..! in NUREC/CR-4550.

4 The accident progression analysis uses large, complex event trees to determine i

the possible ways in which an accident might evolve from each plant damage stata (PDS). The definition of each PDS provides enough information to define t.

.itial conditions for the accident progression event tree (APET) analysis. The Ar T is evaluated by the EVNTRE code. Since there are far too many. paths through the APET to permit individual. consideration in subsequent source term and consequence analysis, the paths are grouped into accident progression bins (APBs) each defining a similar set of conditions for source term analysis. The source terms are calculated for each AFB with a non zero conditional probability by SURSOR. The process of determining which APBs go to which source term group is performed by the PARTITION program.

Offsite consequences were calculated with the MACCS code for each of the source term groups defined in the partitioning process Results for early fatalities, total latent cancer fatalities, population dose within 50 miles and for the entire region, early fatality risk within one mile, and latent cancer fatality risk within 16 miles are given in thi, report. The integration of the NUREG-1150

^

i probabilistic risk assessments uses MACCS results in two forms. The mean risk ' quence measures are used by PRAMIS and RISQUE in determining mean conse results - for - internal - initiators. Cencrated complementary cumulative distribution functions (CCDFs) are used to create CCDFs for risk with the

. PRPOST code.

This report is-published in seven volumes. The first volume describes the methodology. Volume 2 describes the results of convening expert panels to determine those input parameters thought to be the most important contributors to uncertainty in risk. Volumes 3 through 7 present the results of the accident progression, source term, and consequence analyses and the combined risk results for Surry, Peach Bottom, Sequoyah, Grand Gulf, and Zion.

- Publication 'Date:

October 1990 Prepared by:

R. J. Breeding; J. C. Helton; W. B. Murfin; 1. N. Smith Contracion Sandia National laboratories, P. O. Box S800, Albuquerque, NM 87185; Arizona State University, Tempe, AZ 85287; Technadyne Engineering Consultants, Inc.,

Albuquerque, NM; Science Ap-lications International Corporation, 2109 Air Park

-. Road SE, Albuquerque, NM 87406 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor safety, accidents, contahment systems, source terms, Surry-1 reactor, EVNTRE cades, SURSOR codes.

PARTITIO!4 codes, MACCS codes, PRAMIS codes. RISQUE codes, PRPOS'.

codes 21

~ NUREG/CR-4551, V3,RLPt2 -

G::nsr:1 SAND 86-1309, V3,RLPt2

Title:

. Evaluation of Severe Accident Risks: Surry Unit 1 Appendices

==

Description:==

In support of the Nudear Regulatory Commission's (NRC's) assessment of the

' risk from severe accidents at commerdal nuclear power plants in the U.S.

reported in NUREC 1150, the Severe Accident Risk Reduction Program (SARRP) completed a revised calculation of the risk to the general public from severe

. accidents-at the Surry Power Station, Unit 1. This-power plant, located in southeastern. Virginia, is operated by the Virginia Electric Power Corporation.

Off site risk initiation by events, both internal to the power station and external to the power station, was assessed.

This document consists of five appendices: Appendix A Supporting information for the Accident Progression Analysis, Appendix B Supportu.g information for the Source Term Analysis, Appendix C Supporting information for the Consequence Analysis, Appendix D Risk Results, and Appendix E Sampling Information.

Appendix A.1 contains a detallui description and listing of the Accident Progression Event Tree (APET) and the binner (computer input) that instructs EVNTRE oi. how to group the outcomes from evaluating the APET. Appenclix A.2 contains' a description and listing of UFUN, a FORTRAN function subprograrn linked with EVNTRE after compilation to create an executable-module of EVNTRE specific for Surry. UFUN is used to perform two tasks: 1) to determine whether the containment falls and the mode of containment ' failure and 2) to determine the pressure rise due to hydrogen deflagrations in the

- containment. ' Appendix A.3 contains additional-information about the -analysis i.ey basic plant information and a-discussion of how the APET initialization

_ cluestions were quantified. Appendin B includes a FORTRAN source listing, data files, and source term results for the SURSOR code as well as information used in source term ' partitioning. Appendix C provides _ supporting detail for - the consequence anafysis, while Appendix D presents risk results for-Surry for internal initiators including PRAMis output. A ndix E consists of listings of latin Hypercube Sampling (LHS) input of FO N source and data for the related USRDST subroutine, the EXTLHS extender code, and the MODEL program used to calculate power recovery curves.

Publication Date:

October 1990 Prepared by:

1 R. J. Breeding: J. C. Helton; W, B. Murfin; L. N. Smith Contractor:

Sandia National laboratories, P. O. Box 5800, Albuquerque, NM 8718S; Arizona State University, Tempe, AZ 85287; Technadyne Engineering-Consultants, Inc.,

Albuquerque, NM; Science Applications International Corporation,2109 Air Park Road SE, Albuquerque, NM 87106 Prepared for:

NRC Division of Systems Research, Office of Nuc! car Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor-safety, accidents, containment 1 reactor, UFUN codes, EVNTRE codes,.SURSOR systems, source terms,- Surry des, EXTLHS codes, MODEL codes I

codes, LHS codes, USRDST co L

L L

L l

l l}

l l

^

l-22 l

NUREG/CR-1551, V4,R1,Pt1 Gen:rcl SAND 86-1309, V4,R1,Pt1

Title:

Evaluation of Severe Accident Risks: Peach Bottom, Unit 2 Main Report

==

Description:==

In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the U.S.

reported in NUREC 1150, the Severe Accident Risk Reduction Program (SARRP) completed a revised calculation of the risk to the general pubde f om severe accidents at the Peach Bottom Atomic Power Station, Unit 2. This power plant, located in southeastern Pennsylvania, is operated by the Philadelphia Electric Company. The emphasis in this risk analysis was on determining the distribution of risk and discovering the uncertainties that account for the breadth of this distribution. Off-site risk initiation by events, both internal to the power station and external to the power station, was assessed.

De risk assessments on which NUREG-1150 is based can generally be characterized as consisting of four analysis steps--acrident frequency, accident progression, source term, and consequence analysis; a risk integration step; and an uncertainty analysis step. This volume presents the details of the last five of the six steps for the Peach Bottom Atomic Power station, Unit 2. The first step is described in NUREC/CR-4500.

De accident progression analysis is performed by means of a large and detailed event tree, the accident event tree (APET). The computer code EVNTRE, in addition to evaluating the APET, sorts the myriad possible paths through the tree into a manageable number of outcomes, denoted accident progression bins (APBs).

The source term analysis is performed by PBSOR, a relatively small, fast running, parametric code which calculates sourm terms for each APB for each observatwn for Peach Bottom. PARTITION is used to group the source terms with similar health effects weights so that a single consequence calculation can be performed for the mean source term for each group.

O'fsite consequences are calculated with MACCS for each of the source term groups defined in the partitioning process. Results for early fatalities, total latent cancer fatalities, population dose within 50 miles and for the entire region, early fatality risk withln one mile, and latent cancer fatality risk within 10 miles are given in this report. The integration of the NUREC-1150 probabilistic risk assessments uses MACCS results in two forms. The mean consequence measures are used by PRAMIS and RISQUE in determining mean risk results for internal initiators. Generated complementary cumulative distribution functions (CCDFs) are used to create CCDFs for risk with the PRPOST code.

This report is published in seven volumes. The first volume describes the methodology. Volume 2 describes the results of convening expert panels to determine distributions for the variables thought to be the most important contributors to uncertainty in risk. Volumes 3 through 7 present the results of the accident progression, source term, and consequence analyses and the combinect risk results for Surry, Peach Bottom, Sequoyah, Grand Gulf, and Zion.

Publication Date:

December 1990 Prepared by A. C. Payne, R. J. Breeding,11.-N. Jow, A. W. Shiver; J. C. tielton; L N. Smith Contracton Sandia National laboratories, P. O. Box 5900, Albuquerque, NM 87185; Arizona State ' University, Tempe, AZ 85287; Science Applications International Corporation,2109 Air Park Road SE, Albuquerque, NM 87106 Prepared fon NRC Di"ision of Systems Research, Office of Nuclear Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor safety, accidents, containment systems, source terms, Peach Bottom-2 reactor, EVNTRE codes, PBSOR codes, PARTITION codes, IMCCS codes, FRAMis codes, RISQUE codes, PRPOST codes 23

NUREG/CR-4551, V4,R1,Pt2 G:n:ral SAND 86-1309, V4,R1,Pt2

Title:

Evaluation of Severe Accident Risks: Peach Bottom, Unit 2 Appendices Descriptiort:

In support of the Nudcar Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the U.S.

reported in NUREG 1150, the Severe Accident Risk Reduction Program (SARRP) complettd a revised calculation of the risk to the general public from severe accidents at the Peach Bottom Atomic Power Station, Unit 2. This wer plant, located in southeastern Pennsylvania, is operated by the Philadel hia Efectric Company.

The emphasis in this risk analysis was not on determining a "so-called' point estimate of risk. Rather, it was to determine the distribution of risk and to discover the uncertainties that account for the breadth of this distribution.

Off-site risk initiation by events, both internal to the power station and external to the power station, was assessed.

This document consists of five appendices: Appendix A Accident Progression Event Tree, Appendix B Supporting Information for the Source Term Analysis, Appendix C Supporting Information for the Consequence Analysis, Appendix D Risk Results, and Appendix E Sampling Information.

Appendix A.1 contains a detailed description snd listing of the reach Bottom APET and the binner (computer input) that instructs EVNTRE on how to group the outcomes from evaluating the APET. Appendix A.2 contains a description and listing of PSUFUN, a FORTRAN function subprogram linked with EVNTRE after compiktion to create an executable module of EVNTRE specific for Peach Bottom. IBUFUN performs general calculations to determine such information containment failure pressure and mode of failure, pressure rise during core as:

damage and after vessel breach, level of reactor building b ss with and without hydrogen burns, base containment pressure before vesse reach, amount of hydrogen released at vessel breach, and amount of gases produced during core-concrete interaction (CCI). Appendix A.3 contains additional information for the accident progression analysis. Appendix B contains a FORTRAN source listing, data files, and source term results for the PBSOR code as well as information used in source term partitioning. Appendix C provides supporting detail for the consequence analysis, while Appendix D presents a detailed risk result example for the Peach Bottom internal events analysis including PRAMIS output for internal initiators, and Appendix E consists of listings of Latin Hypercube Sampling (LHS) input of FORTRAN source and data for the related U$RDST subroutine, the EXTLHS extender code, and the MODEL and SEILHS programs.

Publication Date:

December 1990 Prepared by:

A. C. Payne, R. J. Breeding, H.-N. Jow, A. W. Shiver: J. C. Helton; L N. Smith Contractor:

Sandia National laboratories, P. O. Box 5800, Albuquerque, NM 87185; Arizona State University, Tempe, AZ 8S287; Science Applications international Corporation,2109 Air Park Road SE, Albuquerque, NM 87106 Prepared for:

NRC l'ivision of Systems Research, Office of Nudcar Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor safety, accidents, containment systems, source terms, Peach Bottem-2 reactor, PBUFUN codes, PBSOR codes, LHS codes, USRDST codes. EXTLHS codes, MODEL codes, SEILHS ctxtes 24

t NUREGICR-4551, VS,R1,Pt1 G:n:rcl SAND 86-1309, VS,R1,Pt1 l

Title:

Evaluation of Severe Accident Risks: Sequoyah, Unit 1 Main Report Descrip'doru In support of the Nudear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commerdal nuclear power plants in the U.S.

reported in NUREG 1150, the Severe Accident Risk Reduction Program (SARRP) completed a revised calculation of the risk to the general public from severe accidents at the Sequoyah Power Station, Unit 1. This power plant, located in southeastern Tennessee, is operated by the Tenneswe Valley Authority. Off-site risks from initiating events interna! to the power station were assessed.

The risk assessments on which NUREG 1150 is based can generally be characterized as consisting of four analysis steps--accident frequency, accident progression, source term, and consequence analysis; a risk integration step; and an uncertainty analysis step. This volume presents the details of the last five of the six steps for the Sequoyah Nuclear Station, Unit 1. The first step is described in NUREG/CR-4550.

The accident progression analysis is performed by means of a large and detailed event tree, the accident progression event tree (APET). The computer code EVNTRE. In addition to evaluating the APET, sorts the myriad possible paths through the tree into a manageable number of outcomes, denoted accident propession bins (APBs).

The source term analysis is performed by SEQSOR, a relatively small, fast-running, parametric code which calculates source terms for each APB for each observation for Sequoyah. PARTITION is used to group the source terms defined to have similar properties, and a single consequenm calculation is completed for the mean source term for each group.

Offsite consequences were calculated with MACCS for each of the source term groups defined in the partitioning process. Results for early fatalities, total latent cancer fatalities, population dose within 50 miles and for the entire region, early fatality risks within one mile, and latent cancer fatality risk within 10 miles are given in this report.

The integration of the NUREG-1150 probabilistic risk assessments uses the MACCS results in two forms. The mean consequence measures are used by PRAMIS and RISQUE in determining mean risk results for intertial initiators.

Cencrated complementary cumulative distribution functions (CCDFs) are used to create CCDFs for risk with the PRPOST code.

This report is published in seven volumes. The first volume describes the methodology. Volume 2 describes the results of convening expert panels to determine distributions for the variables thought to be the most important contributors to uncertainty in risk. Ve!umes 3 through 7 present the results of the accident progression, source term, and consequence analyses and the combined risk results for Sutry, Peach Bottom, Sequoyah, Grand Gulf, and Zion.

Public..on Date:

December 1990 Prepared by:

J. J. Gregory, S. J. Higgins, R. J. Breeding, A. W. Shiver; W. B. Murfin; J. C.

Helton Contractor:

Sandia National Laboratories, P. O. Box 5S00, Albuquerque, NM 87185; Technadyne Engineering Consultants, Inc., Albuquerque, NM; Arizona State University, Tempe, AZ 85287 Prepared for:

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

probabilisuc estimation, risk assessment, reactoi safety, accidents, containment systems, ice condensers, source terms, Se reactor, EVNTRE codes, MACCguoyah-1 a codes, PRAMis codes, RISQUE SEQSOR codes, PARTITION codes, codes, PRPOST codes j

1 2S

NUREG/CR-4551, V5,R1,Pt2 General Se ND86-1309, V5,R1,Pt2 l

Title:

Evaluation of Gevere Aeddent Risks: Sequoyah, Unit 1 Appendices

==

Description:==

In support of the Nudcar Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nudear power plants in the U.S.

reported in NUREC-1150, the Severe Accident Risk Reduction Program (SARRP) completed a revised calculation of the risk to the general public from severe accidents at the Sequoyah Power Station, Unit 1. This power plant, located in southeastern Tennessee, is operated by the Tennessee Valley Authority.

The emphasis in this risk analysis was not on determining a "so-called" point estimate of risk. Rather, it was to determine th( distribution of risk and to discover the uncertainties that account for the breadth of thir, distribution.

Off site risks from initiating events internal to the pwer station were assessed.

Th;s document consists of five appendices: Appendix A Supporting Information for the Accident Progression Analysis, Appendix B Supportmg Information for the Source Term Analysis, Appendix C Supporting Information for the Consequence Analyds. Appendix D Risk Results, and Appendix E Sampling Information.

Appendix A.1 contains a detailed description and listing of the Accident Progression Event Tree (APET) and the binner (computer input) that instructs EVNTRE on how to group the outcomes from the evaluation of the APET.

Appendix A.2 contains a description and listing of SEQUFUN, a FORTRAN function subprogram linked with EVNTRE after compilation to create an executable module of EVNTRE specific for Sequoyah. SEQUFUN performs general calculations to detet aine such information as: the amount and distribution of hydrogen in the containment, the concentration and flammability of the atmosphere in the containment during the various time periods, the pressure rise due to hydrogen burns and adjustment of the amounts of gases consumed in the burns according,1y, and determining whether the containment fails and the mode of failure. dubsection A.3 provides additional infornution including basic plant and plant damage state detail and a description of the AC power racovery data used in the analysis. Appendix B includes a FORTRAN source listing, data files, and results from the source term analysis of the SEQSOR code as well as information used in source term partitioning. Appendix C provides supporting detail for the consequence analysis, while Appendix L presents risk results for Sequoyah for internal initiators including PRAMIS output. Appendix E consists of listings of Latin Hypercube Sampling (LHS) input of FORTRAN source and data for the related USRDST subroutine, EXTLHS extender code, and EXTSEQS, LOSP, and MODEL programs.

Publication Date:

Deccmber 1990 Prepared by:

J. J. Gregory, S. J. Higgins, R. J. Breeding, A. W. Shiver; W. B. Murfin; J. C.

Helton Contracten Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185; Technadyne Engineering Consultants, Inc., Albuquerque, NM; Arizona State University, Tempe, AZ 85287 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor safety, accidents, containment systems, ice condensers, source terras, Sequoyah-1 reactor, SEQUFUN codes, SEQSCR codes, LHS codes, USRDST codes, EXTLHS codes, EXTSEQ8 codes, LOSP codes, MODEL codes 1

26

l NUREG/CR-4551, V6,R1,Pt1 G:n:r:1 SAND 86-1309, V6,R1,Pil

Title:

Evaluation of Severe Accident Risks: Grand Gulf, Unit 1 Main Report

==

Description:==

In support of the Nudcar Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the U.S.

reported in NUREC-1150, the Severe Accident Risk Reduction Program (SARRP) completed a revised calculation of the risk to the general public i.om severe accidents at the Grand Gulf Nuclear Station, Unit 1. This power plant, located in Port Gibson, Mississippi, is operated by System Energy Resources, Inc. (SERD.

Off site risk initiation by events internal to the power plant was assessed.

The risk assessments on which NUREC-1150 is based can generally be characterized as consisting of four analysis steps-accident frequency, accident progression, sourm term, and conscauence analysis; a risk integration step; and an uncertainty analysis step. This volume presents the details of the last live of the six steps for the Grand Gulf Nuclear Station, Unit 1. The first step is described in NUREC/CR 4550.

The accident progression analysis uses large, complex event trees to determine the possible ways in which an accident might evolve from each plant damage state (PDS). The definition of each PDS provides enough information to define the initial (unditions for the accident progression event tree (APET) analysis. The APET is evaluated by the EVNTRE code. Since there are far too many paths through the APET to permit inoividual consideration in subsequent analyses, the paths are grouped into accident progres"on bins (APDs) each defining a similar set of conditions for source term analysts. Source terms are calculated for each APB for each observation by CCSOR. PARTITION is used to group sourm terms with similar properties, so that a single consequence calculation can be performed for the mean sourm term for each group.

Offsite consequences are calculated with MACCS for each of the sourm term groups. Resufts for six of these-carly fatalities, total latent cancer fatalities, population dose within 50 miles and for the entire region, carif fatality risk within one mile, and latent cancer fatality risk within 10 miles-are given in this report. The integration of the NUREC-1150 probabilistic risk assessments uses MACCS results in two forms. The mean consequence measures are used by PRAMIS and RISQUE in determinhig mean risk results for internal initiators.

Cencrated complementary cumulative distribution functions (CCDFs) are used to create CCDFs for risk with the PRPOST code.

This report is published in seven volumes. The first volume describes thc methodology. Volume 2 describes the results of convening expert panels to determine distributions for the variables thought to be the most important contributors to uncertainty in risk. Volumes 3 through 7 present the results of the accident progression, source term, and consequence analyses and the combined risk results for Surry, Peach Bottom, Sequoyah, Grand Gulf, and Zion.

Publication Date:

December 1990 Prepared by:

T. D. Brown, R. J. Breeding, IL-N. Jow, S. J. Higgins, A. W. Shiver; J. C. liciton; C. N. Amos Contractor:

Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185; Ar.' zona State University, Tempe, AZ 85287; Science Applications International Corporation,2109 Air Park Road SE, Albuquerque, NM 871(%

Prepared for:

NRC Division of Systen s Research, Office of Nuclear Regulatory Research Keywords:

probabilistic estimation, risk assessment, reactor safety, accidents, containment systems, source terms, BWR type reactors, Grand Gulf l reactor, EVNTRE codes, CGSOR codes, PARTITION : odes, MACCS codes, PRAMIS codes, RISQUE codes, PRPOST codes 1

27

WUREG/CR-4551, V6,R1,Pt2 General SAND 86-1309, V6,R1,Pt2

' fille:

Evaluat,on of Severe Accident Rals: Grand Culf, Unit 1 Appendices

==

Description:==

In support of the Nudcar Regulatory Commission's (NRC's) assessment of the risk froca severe acridents at com.ncrcial nucinar power plants in the U.S.

reported in NUkEG 1150, the Seven* Accident Risk Reduction Program (SARRP) completed a reviwd calculation of the risk to the general pub:ic from severe accidenti, at the Grand Gulf Nudcar Station, Unit 1. This power plant, located in Port Gibson, Mississippi, is operated M %jstem Energy Resourm, Inc. (SERI).

)

'the emphasis in this risk analyst was not on determining a "so-called' print estimate of r'sk. Rather, it was to determme the distribution of risr. and to dir. cover 6e uncertainties that account for the breadth of this distribution.

Off site risk inPstion by events internal to the power station was asseswd.

(c.t the Accident Progrwsion Analysis, Appendix B Supporting In'g Informat This dexment consists of five appendices: Appendix A Supportin formation fc.r the Source Term Analysis, Appendix C Supporting Information for the Consequence Analysis, Appendix D R. t Results, and Appendix E Sampling Information.

i Appendix A.1 contains a dett M A.ription and listing of the Grand Gulf ApFT and the binner (computer input) that instructs EVNTRE on how to group the APET pathways. Appendix A.2 contains a description and listing of CGUFUN, a FORTRAN function subprogram !!nked with EVNTi E after compilation to create an executable module of EVNTTE specific for Gra,.d Gulf.

GGUlVN performs general calmlations to determine the containment baseline pre,suto during the variens time periods, compute the amount of hydrogen released to the containmet.t at vessel breach (VD) and during core-concrete interaction (CCI), cori.pute the concentration snd flammability of the atmosphere in the containment and drywell during the various time periods, calculate the pressure rise due to hydrogen burns, determine whether the containment falls and the mode of failure, and determine whether the drywell falls and the mcdc of failure. Appendix A.3 contains additional information concerning the accident progression analysis. Appendix B contains a FORTRAN source listing, data files, and source term results for the GCSOR code. Appendix C provides supporting detail for the consequence analysis, while Appendix D presents detailed risk results for Crand Gulf internal Inlators including PRAMis output. Appendix E consists of listings of latin flypercube Sampling (LHS) input of FORTRAN source and data for the related USRDSTCC subroutine, EXTLHS extender code, and the MODEL and LOSP programs.

Publication Date:

December 1990 Prepare.d by:

T. D. Drown, R. J. Breeding, IL N. Jow, S. J. Higgins, A. W. Shiver; J. C. Helton; C. N. Amos ContracttJ Sandia National laboratories. P. O. Box 5800, Albuquerque, NM 87185; Arizona State University, Tempe, AZ SS2S7; Science Applications international Corporation,2109 Air Park Road SF, Albuquerque, NM 87106 Prepared for:

NRC Division of Systems Researr.h, Office of Nuclear Regulatory Research Keywords:

probabilistic estimatic,n, risk assessment, reactor safety, accidents, containmcat systems, source terms, BWR type reactors, Crand Gulf 1 reactor, GCUFUN codes, CCSOR codes, LHS codes, !!SRDSTCG codes, EXTLHS codes, MODEL codes, LOSP codes i

28

NUREG/CR-4554, Val 6 SCANS UCID-20674, Vol 6 Title.

SCANS Gh:pping C<r Analysis System) A Micerwmputer Based Analysis System for Shipping wk Design Review Volume 6-Theory Manual Buckling of Circular Cylmdrical Shells Descript.oru A computer system called SCANS (Shipping Cask Analysis Systerr) was developed for t!.e staff of the U. S. Nuclear Regulatory Commission to perform confirmator) licensing review analyses. SCANS caa handle problems associated with impact, heat trant.fer, ther nal stress, internal or external pressure loads, and Icad slump..A new upability. implemented in " CANS is buckling analysis of the steel shells of a spent fuel shipping cask during a postulated impact with an unyielding surface.

This ciocument (Volume 6) covers three sets of buckling analysis formulas. The first wt is based on Code Case N 284 of the ASME Boiler and Pressure Vessel Code expanded to include stainless steel shells, in addition to the existi.ig carbon steel shcIl capability. The second set is based on American Petroleum Institute Bulletin 2U, an upgrade of N 284 that includes test-results available after N 284 was written in 1979. The third set is based on formulas frequently used by the piping and pressure vessel industry and formulas recommended by the Structural Stability Research Council. To be compatible with the ASME Code, the first set is recommended for use in shipping cask evaluation: this. set is implemented in SCANS. The second and third sets are recommended references for SCANS users.

7 A buckling study was conducted for two typical shipping casks to evaluate various formulas and provide some insight about shell bucklirig of shipping casks. One typical rail cask and one typical truck cask were used in this study as

• the finite difference analysis codes, DOSOR4 and DOSORS.

i Publication Date:

Februs.y 1990 Prepared by:

T. Io, G. C. Mok, D. J. Chinn

' Contractor:

Lawrence Livermore National bboratory, P. O. Ocx 808, Livermore, CA 94550 Prepared for:

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

spent - fuel - casks, de'ormation, chelir,, impact strength, thermal stresses, heat-trsnsfer, pressure, DOSOR codes t

i 29

_.~.;_

NUREG/CR 4554, Vcl 7 SCANS UCID-20674, VCl. 7

Title:

SCAN 3 (Shipping Cask Analysis System) A Microcomputer Eaud Analyits S

• stem for Shipping Cask Design Review Volume 7-Theory Manual Ptmeture o$, Shipping Casks Descriptforu A computer system called SCANS (Shipping Cask Analysis System) was developed for the staff of the U. S. Nudcar Regulatory Commission to perform confirmatory licensing review analyses. SCANS can handle problems assodated lead slupact, heat transfer, thermal stress, internal or caternal pressure loads, with im np, and buckling analysis.

Under current regulatory requirements, a shipping cask should be designed for a series of hypothetical acrident conditions. These test conditions include a 40 inch free drop of the cask into a 6-inch diameter puncture pin. In this study, existing puncture test data were examined. Simple formulas based on test data were proposed for puncture evaluation of shipping casks. Dynamic and static nonlinear finite element analyses were performed to currelate analyses with the existing test data. In this analytical approach, three puncture frilure prediction methode were pmposed, and their applicability was evaluated. The analytical approach provides an alternative to testing. Both laminated and solid wall shipping casks were analyzed. in the study of laminated casks, the effects of the inner shell on the puncture of the outer shell were examined, as were the effects of material strength of the puncture pin. The study of geometric scaling of casks indicatcd that the normalized incipient puncture energy is insensitive to variations in the scale factors. This conclusion indicates that the proposed analytical approach of combining finite element analysis and failure prediction methods is consistent with the similarity principles it tests, except when local vibration is excited during the puncture process. Fu4ther study of this local vibration is needed. Future rewarch directions were also explored in this study.

A stat'stical analyd of existing puncture test data was performcd. A simple design formula for selecting shell thicknest.cs was derived from the test data.

This simple formula is adequate for design and evaluation purposes, and it is recommended for implementation in SCANS.

Publication Date:

February 1990 Prepared by:

Ting Yu to Contracton lawrence Livermore National Laboratory, P. O. Box 808, Livermore, CA 94550 Prepared fon NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

spent fuel casks, acddents, testing, shells, plates, failures, finite elemer.) method, DYNA 2D codes, NIKE2D codes l

i l

l 30 l

l NUREC/CR-1624, Vcl. 6 STCP, MARCIl3 UMI-2139, Vol. 6

Title:

Radlonuclide Release Calculations for Selected Severe Accident Scenarios Supplemental Calculations Descriptforu This report provides the results of source term calculattens performed in support of the NUREC 1150 study " Severe Accident Risks: An Aswssment for Pive U. S.

Nudcar Power Plants." This is the sixth volume of a series and supplements results presented in the earlier volumes. Analyses were performed for thrw of the NUREC 1150 plants Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Aquoyah, an ice condenser containment, pressurized wate reactor.

Complete wurce term results are presented for the following wquences short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA in the Sorry plant; station blackout with a punip seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant.

In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCIO analyws were performed for the Sorry and Sequoyah plants to evaluate the effeca of alte native emergency operating procedures involving primary and secondary de provided for these analyses. pressurization. Only thermal bydraulle results are In addition, three accident sequences were analyzed for the Surry plant for accident induced failure of ricam generator tubes. In these analyses, only the transport of radionudides within the primary system and fallec steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough.

Publication Date:

August IWO Prep.tred by:

R. S. Denning, M. T. Leonard, P. Cybulskis, K. W, Lee, R. F. Kelly,11. Jordan, P.

M Schumacher, L A. Curtis Contractor:

Battelle Columbus Division,505 King Averme, Columbus, OH 43201 2693 Prepared fon NRC Division of Systems Research, Office of Nudear Regulatory Research Keywords:

nuclear power pimts, source terms, blackouts, loss of coolant, radionuclide migration, reactor accidents, emergency plans, depressurization, PWR type reactors, DWR type reactors, ECCS, pumps, seals, MARCil2 codes, CORSQR codes, CORCON codes, VANESA codes, TRAP MELT codes, MERGE codes 31

NUREG/CR 4639, Vct.1, Rev.1 NUCLARR EGG-2458, Vct.1, Rev.1

Title:

Nudear Computerized Library for Assessing Reactor Reliability (NUCLARR)

Summary Description Descriptforu -

The Nuclear Computerited 1.ibrary for Assessing Reactor Reliability (NUCLARR) is an automated database management system for storing and processing human error probability and hardware component failure rate data. The NUCLARR system software resides on an IBM (or compatible) personal microcomputer.

NUCIARP can be accessed by the end user to furnish data suitable for input in hume.n p., t hardware reliability analysis to support a variety of risk assessment activities.

The NUCLARR system is documented in a five-volume wrics of reports. Volume I of this series is the Summary Description, which presents a general overview of the data management system, including a description of data collection, data qualifkation, data, structure, and tar.onomics. Human error probability (HEP) data are hierarchically organized by system /sobsystem/cumponent and human action. Hardware component failure data are organized by equipment type, component de lgn, and failure mode. Both parts of the NUCLARR system make use of a matrix structure and independent aggregation methods appropriate for the data types. Programming activities, procedures for processing data, a user's unde, and - a hard copy data manual are presented in the accompanying olumes 11 through V The NUCLARR Cleuinghouse has been established at the INEL to maintain the NUCLARR database, to distribute software diskettes, and to assist users of +hr system.

14UCLARR is designed to run on an IBM or compatible microcomputer (e.g.,

IBM XT/AT, or 15/2, Compaq 286/386) with 640 Kbytes of random access memory (RAM) for systems with DOS 3.1 through 3.3 or 1 Mbyte of RAM for 15/2 configurations with DOS 4.01. A 20 Mbyte hard disk, math co-processor, Borland's Sidekick, can be installed when running NUCLARR. programs, s No memory resident and EGA card are also required.

Publication Date:

May 1990 f

Prepared by:

D. l.. Certman, W. E. Cilmore, W. H. Calycan, M. R. Croh, C. D. Centillon, B.

D. Gilbert, W. J. Reece Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415 Prepared for:

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

data base managemem, risk assessment, nuc! car power plants, human factors, r

reactor components, reliability, failures, nuclear data collections, data processing, information retrieval

+

t 32

-]

I

- NUREG/CR-4639, VcL4, Rev.2 NUCLARR EGG-2458, VcL4, RCv.2

Title:

Nudcar Computerized Library for Assessing Reactor Reliability (NUCLARR)

User's Guide i

i

==

Description:==

The Nuclear Computerized Library for Assessing Reactor Reliabihty (NUCLARR)

{

is an automt.tal database management system for processing and storing human error probability (HEP) and hardware mmponent failure data (HCFD). The

+

NUCl.ARR system software resides on an IBM (or com,2atible) personal microecomputer and can be used to furnish input data for both human and hardware reliability analysis in support of a variety of risk assessment activities.

The NUCLARR system is documented in a fivcuvolume series of reports. Volume IV of this series is the User's Guide for operating the NUCLARR software. It is presented in three parts. Part 11 Overview of NUCLARR Data Retrieval provides an introductory overview to the system's capabilities and procedures for data retrieval. Methods and criteria for selection of data r.ourms and for enterin6 them into the NUCLARR system are also described. Part 2: Culde to Operations contains the instructions and basic procedures for using the NUCLARR software.

Guidance and information for getting startal, performing the desired functions, and making the most efficient use of the system's features are provided. Part 3:

NUCI ARR System Description provides an irwicpth discussion of the design characteristics and presents special features of the NUCLARR software including -

the organization of the datatuse structures and techniques used to manipulate the data.

HEP data that are accepted for inclusion into the NUCLARR system must meet the following criteria: Involve a human at on that was performed or was supposed to be performed, describe the equipment that was the object of the human action, and provido quantitative values of error probability in the form of an HEP point value, probabillt" distribution, or a ratio of errors to estimated opportunities for error. The HCFD requirements for NUCLARR apply to component failure rates and probabilities. To be included in the NUCLARR -

database the component failure data must have a description of the component and failure mode and probability values ti.e., number of failures and number of operating hours or demands, or the rate itself).

Use of the NGCLARR system requirca an IBM PC or PC. compatible computer (IBM PC/AT or S access memory (5$ystem 2 preferred) with a minimum of 640 Kbytes of rand graphics card, color monitor, math co-processor, and DOS operating system (3.x or 4.0). The user documentat on was developed for software version 4.0. New editions of the User's CulA will be issued periodically to reflect major software changes.

Publication Date:

October 1990 Prepared by:

W. E. Gilmore, C. D. Centillon, D.1. Certman, C. H. Deers, W. J. Calycan, D. C.

Gilbert, W. J. Reece Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415 Prepared for:'

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

Keywords:-

data base manage acnt, risk assessmeet, nuclear power plants, human factors, reactor components, reliability, failures, nuclear data collections, data processing, inforrmation retrieval

.I 33 I

NUREC/CR-1639, VcL5,Pt.1,R3 NUCLARR EGG 2458, Vol.5,Pt.1,R3

Title:

Nudcar Computerized Library for Awessing Reactor Reliability (NUCl ARR) taata Manual Part 1: Summary Description

==

Description:==

he Nuclear Computerized Library for Anessing Reactor Reliability (NUCLARR) is a mmputer-based data management system used to promss, store, and retrieve human error probability (llEP) and hardware component failure data (HCFD) in a ready to.use format. De NUCLARR system software allows analysts with an IBM (or cornpatible) personal computer with math co-processor, 20 Mbyte hard disk, and 640 Kbytes of available random access memory to use the system at their own location to supply data for both human and hardware reliability analysis in support of risk assenment activities.

The NUCLARR system is documented in a five-volume series of reports. Volume V: Data Manual provides a haid-copy representation of all data and related information available within the NU(X.ARR system software The Data Manual is organized to provide the exprienced analyst with a read a reference of data and information otherwise available from the NUCLARR so tware. All sections of this document are explained in sufficient detail such that familiarity with NUCLARR coding and output is 1.ot necessary. It is not meant, however, to serve as a guide to performing risk aneament or data analysis. Readers must use their own judgment and prior understanding of HRA/PRA to interpret the NUCLARR data.

The Data Manual is arranged in three parts. Part 1: Summary Descriptio,r,l orovides an overview of the NUCLARR system and data processing procedures.

?prt 2: Human Error Probability (HEP) Data contains all data and Information relevant to the human error probabihty side of NUCLARR. An overview of the NUCL/,RR IIEP data treatment is presented, and an explanation of the data presentation format is provided as a preface to the actual data records (Appendices A, B, and C). Further explanation of NUCLARR codes and other supporting information can be found in Appendix D. Data and information for the hardware component failure data (HCFD) side are presented in J4rt 3:

Hardware Component Fathy,rc Data (HCFD), which is organized in ye same manner as Part 2. The ll' FD treatment in NUCLARR is explained, as is the C

format of the data record presentation (Appendices A, B, C, and D). Appendix E of Part 3 provides the supporting information necessary to interpret the hardware failure data.

Publication Date:

December 1990 Prepared by:

B. G. Gilbert, W. J. Reece, D. l. Certman, W. E. Cilmore, W. !. Calycan Contractor:

EC&G Idaho, Inc.. ". O. Box 1625, Idaho Falls, ID 8M15 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

data base management, risk assessment, nucle:.r power plants, human factors, reactor components, reliability, failures, nuclear data collections, data processing, information retrieval 34

~ - - -

- - - - - _ ~

NUREG/CR-4639, VCL5,Pt.2,R3 NUCLARR EGG-2458, Vol.5,Pt.2,R3

Title:

Nuclear Computerized Library for Assessing Reactor Reliability ('NUCLARR)

Data Manual Part 2: Human Error Probability (IIEP) Data f

Descriptioru The Nuclear Computerized Library for Assessing Reactor Rc!! ability (NUCLARR) is a computer based data inanagement system used to process,- store, and retrieve human error probability (HEP) and hardware component failure data (HCFD) in a ready to use format. The NUCLARR system softwero allows analysts with an IBM (or compatible) personal computer with math co-processor, 20 Mbyte hard disk, and 640 Kbytes of ivallable random acmss memory to use the system at their own location to supply data for both human and hardware reliability analysis in support of risk assessment activities.

The equipment taxonomics and data structures for NUCLARR were designed specifically to support probabilistic risk assessment (PRA) techniques currently used by the nucioar power industry. The NUCLARR system aids the risk analysis promss by providing the analyst with accurate and relevant data from an on-line databau. All data and related information available from this PC-based system are also accessible in hard copy format. The NUCLARR system is documented in a five volume series of reports. Volume V: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software.

The Data-Manual is arranged in three parts. Part 1: Summary Demiotion orovides an overview of the NUCLARR system and data processing prot dures.

' art 2: Human Error Probability (llEP) Data contains all data and information relevant to the human error probabihty side of NUCLARR. An overview of the NUCLARR HEP data treatment is presented, and an explanation of the data presentation - format in provided as a preface to the actual data records (Appendims A, B, and C). Each Appendix contains an introduction followed by data pages. The introductory sections provide an outline of the information presented on the data pages, including definitions for each data ficid. All codes found in this report are identified in Appendix D, along with a listing of the data reference documents. This release of the Data Manual contains 1,212 HEP records. Data and information for the hardware component failure data (llCIV) side are presented in Part 3: Hnrdware Component Failure Data (HCFD), which is organized in the same manner as Part 2. The HCFD treatment in NUCLARR is explained, as is the f rmat of the data record presentation (Appendices A, D, C, and D). Appendix E of Part 3 provides the supporting information nomssary to interpr;t the hardware failure data.

Publication Date:

December 1990 Prepared by:

B. C. Gilbert, W. J. Reem, D. I. Certman, W. E. Cilmore, W. J. Calycan Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 8}t15 Prepared for:

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

data _ base management, risk assessment, nuclear power plants, human factors, reactor components, reliability, failures, nurlear data collections, data processing, information retrieval 35 i

NUREG/CR 4639, VCl.5,Pt.3,R3 NUCLARR EGG-2458, Vol.5,Pt.3,R3 I

Title:

Nudcar Computerized Library for Assessing Reactor Reliabillty (NUCLARR)

Data Manual Part 3: Hardware Component Failure Data (llCFD)

==

Description:==

The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is a computer-based data management system used to process, store, and retrieve human error probability (HEP) and hardware component failure data (HCFD) in a ready to-uw format. The NUCLAliR sys cm software allows analysts with an IBM (or compatible) personal computer with math co processor, 20 Mbyte hard disk, and 640 Kbytes of available random access memory to use the system at their own location to supply data for both human and hardware reliability analysis in support of dsk assessment ac*ivities.

The equipment taxonomics and data stmetures for NUCLARR were designed specifically to support probabilistic risk assessment (PRA) techniques currently used by the nuclear power industry. The NUCLARR system aids the risk analysis process by providim the analyst with accurate and relevant data from an on line database. All data and related information available from the PC-based system are also accessible in hard copy format. The NUCLARR system is documented in a five-volume sneles of reports. Volume V: Data Manual provides a hard-copy represemation of all dat i and related information available with(n the NUCLARR system software.

i The Data Manual is arranged in three parts. Part 1: Summary Description arovides an on-rview of the NUCLARR system and data processing procedures.

' art 2: Human Error Probability (HFP) Data contains all data and Information

relevant to the human error probabliity side of NUCLARR, An overview of the NUCLARR HEP data treatment is presented, and an explanation of the data p.'esentation format is provided-as a preface ' to the actual data records (Appendices A, B, and C). Further explanation of NUCLARR codes and other supporting information can be found in Appendix D. Part 3: Hardwagg pically Compor.mt Fadurc Data (HCFD) contains failure data for components ty'all the used at nuclear power plants. The system is configured to indude component codes and failure mode codes currently defined in "Rcquirements for Entry of Component Failure Data in NUCLARR." The taxonomy is hierarchically configured in terms of five basic event levels and event source record level data.

The hardware component failure data records are assembled in a series of four appendices (A, B, C, and D). Each appendix contains an introduction followed by data pages. The introductory sectiors provide an outline of :he information presented on the data pages, induding definitions for each ileid. All data codes found in this report are identified in Appendix E, along with a listing of the 3

reference documents. This release of the Data Manual contains a total of 1,442 HCFD records (629 electrical component failures and 813 mechanical component failures).

~ Publication Date:

December 1990 Prepared by:

. B. G. Gilbert, W. J. Reece, D.1. Certman, W. E. Gilmore, W. J. Calycan Contractor:-

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 8M15 Prepared for:

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

data base management, reactor components, failures, nuclear power plants, accHents, safety, electilcal faults, mechanical properties, data compilation 1

36

.-=;-----

NUREG/CR-4668 DFRMODX SAND 86-1030

Title:

Damagcd Fuel Experiment DF 1 Results and Analyses Descriptian:

A series _of in pile esperiments addressing LWR severe fuel damage phenomena was conducted in the Annular Core Research Reactor (ACRR) at Sandia National bboratories. The ACRR Debds Formation and Relocation (DF) experiments are quasi se urate effects tests that provide a database for the development and verification of raodels for LWR severe core damage accidents, ne first experiment in this series, DF 1, was perfctmed on March 15, 1984, and the results are presented in this report. The DF 1 experiment examined the effects of low Initial dad oxidation conditions on fuel damage and relocation procesus.

The DF 1 tnSt assembly consisted of a nine rcxs square-matrix bundle that employed PWR-type fuel tr<!s with a 0.5-m fissile length. The fuel rods were composed of 10% enriched UO pellets within a zircaloy-4 cladding. Steam flowed through the test bundle at flow rates varying between 0.5 and 3 g/s, and the ACRR maintained a peak power level of 1.5 MW during the high temperature oxidation phase of the test indudng approximately 8.5 kW fission power and approximately 20 kW peak oxidation power in the assembly. Visual observation showed early clad relocation and partial blockage formation at the grid spacer location accompanied by productbn of a dense acrosol. Posttest cross sections show li uefaction losses of fuel in excess of 10 volume, percent, as well as large tractiona losses of cladding material from the upper two thirds of the bundle. The quantity of hydrogen measured dunng the test was consistent with the observed magnitude of cladding oxidation Oxidation driven heating rates of 25 K/s and peak tempcratures in excess of 2525 K were observed. The analyses, inte.pretation, and application of :hese results to severe fuel damage accidents are discussed.

De DFRMODX code was dcHgned to provide voping and analysis capabilities for DF type esperiments. It models a one, four, or nine-fuel rod bundle, with coolant channefs and surrounding materials. The geometry is considered fixed in time, and there is no provision for material motion other than gas flow. Decay or fission heating, zircaloyfersteam oxidation reduction reaction, and tanvective and rautative heat trans from the cladding surfaces to a flowing steam hydrogen mixture are taken into account.

Other codes used in conjunction with the DF 1 experiment include: SCDAP, a state of the-ait compu; r model for a degiaded fuel rod bundle used in the early analysis without knowledge of the steam condensation problem so based on inaccurate steam flow rates and power coup?ing; the heat condaction/ convection code COPOX-R, with an added inversion procedure to infer *he hydrogen production history from reactic.n tube temperature profiles produced by the exothermic reduction of copper oxide by hydrogen; and MAEh0S, a convection and diffusion model of particle agglomeration, deposition, and aerosol transport, l

which was used to obtain the data needed to estimate the light extinction properties of the Sn acrosol.

Publication Date:

January 1990 Prepared by:

R. D. Casser, C. P. Fryer, R. O. Cauntt, A. C. Marshall, K. O. Reil, K. T. Stalker Contractor; Sandia Nat:c,nal bboratories, P. O. Box 5800, Albuquerque, NM 87185 Prepared for:

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

nuclear fucts, fuel rods, damage, reactor experimental.'acilities, reactor accidents, meltdown, PWR type reactors, fuel cladding interac+ ion, oxidation, hydrogen production, COPOX R cafes, MAEROS codes, SCDAP codes 37

NUREG/CR-4691, Vel 1 MACCS SAND 86-1562, Vol.1

Title:

MELCOR Accident Consequence Code System (MACCS) Ve.cr's Guide

==

Description:==

This report describes the MACCS code, which performs probabilistic calculations of potenti.d off. site consequences of the atmospheric releases of radioactive material in reactor aeddents. The report consists of three volumes. Volume 1, the User's Culde, des:ribes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

The purpose of the code is to simulate the impact of severe acddents at nudcar power plants on the surrounding environment MACCS was developed to replace the previous CRAC2 code and incorporates many im rovements in modeling flexibility in comparison to CRAC2. The prind I phenomena considered in MACCS are atmospheric transport, environmenta contamination, emergency response, long-term mitigative actions bawd on dose projection, dow accumulation by a number of pathways induding food and water ingestion, early and latent health effects, and economic costs. The time scale after the aeddent is divided in'.o three phases: emergency, intermediate, and long term.

The region surrounding the reactor is divided into a polar-coordinate grid, with the reactor located at the center, for the calculations.

MACCS can be used for a variety of appdcations induding probabilistic risk assessment (PRA) of nudear power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost-benefit analysis.

The preprocessor, MAXCC, generates the maximum allowable - ground concentrations based un protective action guide (PAG) dose levels, while another preprocessor, DOSFAC, generates the dose conversion data uwd by MACCS.

Publication Date:

February 1990 Prepared by:

D.1. Chanin; J. I., Sprung, L. T. Ritchie, H N Jow Contractor:

Technadyne Engineering Consultants, Inc., p. O. Box 13928, Albuquerque, NM 87192; Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

nudcar power plants, reactor acdden%, probabilistic estimation, nsk assessment, sensitivity analys:3, cost benefit analysis, emergency plans,- atmospheric circulation, dosimetry, economics, evacuation, health - hazards, mitigaHon, radiation hazards, weather 38

NUREG/CR 4691, Vol. 2 MACCS SAND 86-1562, Vol. 2

Title:

~

MEl.COR Accident Consequence Code System (MACCS) Model Description Descriptioru This report describes the MACCS code, which performs probabilistic calculations of potential off site conwquences of the atmosphede releases of radioactive material in reactor meddents The report consists of three volumes. Volume 2, the Model Description, describes the underlying models that are implemented in the code.

The purpose of the code is to simulate the impact of severe acddents at nuclear power plants on the surrounding environment. MACCS was devefoped to replam the previous CRAC2 code and incorporates many im rovements in modeling -flexibility in cortparison to CRAC2. 1hc prind I phenomena considered in MACCS are atmospherie transport, environr.senta contamination, emergency response, long term mitigative actions based on dose proJoction, dose accumulation by a numkr of pathways induding food and water ingestion, r

- carly and latent health effects, and economic cosa 4he time scale after t!.e accident is divided into three phar.es: emergency, intermediate, and long term.

The region surrounding the reactor is divided into a polar coordinate grid, with the reactor located at the center, for the calculations.

MACCS can be used for a variety of applications induding probabilistic risk assessment (PRA) of nudcar power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost-benefit analysis.

o The preprocessor, MAXCC, generates the maximum allowable ground concentrations based on protective action guide (PAC) dose levels, wh4e another

. preprocessor, DOSFAC, generates the dose conversion data used by MACCS.

Publication Date:

February 1990 Prepved by:

H N Jow, J. L Sprung, L T. Ritchle; J. A. Rollstin; D. I. Chanin Contractor:.

Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185; CRAM, Inc., 1709 Mnon NE, Albuquerque, NM 87112; Technadyne Engineering Consultants,-Inc., P. O. Box 13928, Albuquerque, NM 87192 Prepared for:

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

_nudcar power plants, reactor accidents, probabilistic estimation, risk assessment, sensitivity analysis, cos: - benefit - analysis, emergency plans, atmospheric circulation, dosimetry, economics, evacuation, health hazards, mitigation, radiation hnards, weather l

l

i.

~

l

[-p i

NUREG/CR 4691, Vol. 3 MACCS SAND 86-1562, Vol. 3

Title:

MELCOR Acrident Conwquence Code System (MACCS)

Programmer's Reference Manual Descriptiott This report describes the MACCS code, which performs probabilistic calculations of potential off site consequences of the atmospheric releases of radioactive material in reactor accidents. The te ert consists of three volumes. Volume 3, the Programmer's Reference Manual, cIescribes the code's structure and database management features.

The purpose of the code is to simulate the impact of r,cvere accidents at nudcar power pl:.nts or the surrounding environment. MACCS was developed to replace the previous CRAC2 code and incorporates many im rovements in modeling flexibility in comparir,on to CRAC2. The princt i phenomena censidered in MACCS are atmospheric transport, environmenta contamination, emergency response, long term mitigative actions based on dose projection, dose accumulation by a number of pathways including focxi and water ingestion, early and latent health effects, and economic costs. The time scale af ter the accident is divided into three phases: emergency, intermediate, and long term.

The region surrounding the reactor is divided into a polar-<nordinate grid, with the r2 actor located at the center, for the calculations.

MACCS can be used for a variety of applications including probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost benefit at.alysis.

The preprocessor, MAXCC, generates the aaximum allowable ground concentrati nn based on protective action guide (PAC) dose levels, while another preprocesor, DOSFAC, generates the dose conversion data used by MACCS.

Publication Date:

February 1990 Prepared by:

J. A. Rollstin; D.1. Chanin; if N Jow Contractor:

CRAM, Inc.,1709 Moon NE, Albuquerque, NM 87112; Technadyne Engineering Consultants, Inc., P. O. Box 13928, Albuquerque, NM 87192: Sandis National laboratories, P. O. Box 5800, Albuquerque, NM 87185 Prepared for:

NRC Division of Systems Research, Office of Nucicar Regulatory Rcratch Keywords:

nudcar power plants, reactor accidents, probabilistic estimation, risk assessment, sensitivity analysis, cost benefit analysis, emergency plans, atmospheric cinculation, dosimetry, economics, evacuation, health hazards, mit:dation, radiation hazards, weather 40 i

~

NUREC/CR-4735, Vcl 6 EQ3/6, PANDORA l

Title:

Evaluation atd Compilation of COE Waste Package Test Data hiannual Report August 1988. January 1989 g

Descriptforu This documer.t -summarizes evaluations by the National in$titute of Standards and Technology (NIST) of Department of Energy (DOE) activities on warte packages designed for containment of radioactive higblevel nylear waste (liLW) for the six month period August 19&S through January 1989. Includai are reviews of related materials research arwl plans, activities for the DOE Materials Characterization Center, informatic n the Yucca Mountain Project, and other information regarding supporting emarch and special assistance. NIST on the. ucca Motmtain Consultation Draft Site piven comments are Characterization llan (CDSCP) and on the Waste Compliance Plan for the West Valley Demonstration Project OVVDP) liigh-level Waste Form.

The NIST/NRC Database for Reviews and Evaluations on High Level Waste uses a new database management system, Advanced Revelation, and the necessary conversion frorn the previously uwd Revelat;on system was completed. In addition, two data elements were added. Sort and search operations can now tw performed on the prindpal author and the publication datr.

Reviews of rep /6 computes equilibsvorts on the EQ3/6 ro *puter program included. EQ3 models of aqueous geochemical systemt.

I The pa:kage contains two-ma.,c.

programs:

EQ3 which. performs distribution of-species calculations for natur I water compositions, and EQ6 which uses the results of EQ3 to predict the consequences of heating and cooling aqueous solutions and of irreversible reaction in rock water systems. The calculation predicts in detail the changes in fluid composition, the identity, appearance, and disappearance of secondary minerals, and the values of reaction progress at which the fluid saturates with reattants. The programs are vahiable for studying such thenomena as the formation of core bodies, scaling and L

plugging in geothermal development, and the long term disposal of nuclear Waste.

PANDORA is a computer basal model of nuc! car waste package performance r

and the processes affeoing it over the long term, specific to conditions at the proposed Yucca Meuntain, Nevada site. The pronsses PANDORA models include: changes in inventories due to radioactive decay, gamma radiation dose rate in and near the package, heat transfct, mechanical behavior,

- coch-t, corrosion, waste form altert. tion, and radionuclide release. groundwater The model.

tram the development and coupling of these prc, cesses over time.

Publication Date:

November 1990 t

Prepared by:

C. C. Interrante: E. Escalante, A. C. Fraker Contracton USNRC; U.S. Department of Commerce, Netional Institute of Standards and Technology, Caithersburg, MD 20899 Prepared for:

NRC Divirion of liigh Level Waste Management, Office of Nuclear Material Safety an.i Safeguards Keywords:

reviews, data base management, high level r dioactive wastes, radioactive waste disposal, corrosion, tuff, cpent fuels, Yucca mountain, salt caverns, basalt, water chemistry, austenitic steels, copper, zircoloy, borosilicate glass,- Irradiation, phase stability, teaching, carbon '.4

.i

. 41

.., = -

.. - ~. ~

l NUREG/CR-4753, VCl. 3 SAM, LOO CAUSMO

Title:

Canadian Seismic Agreement Annual Report: 1987 1988

==

Description:==

nis is the third annual progress report unda the terms of the Canadian Seismic Agreement betwcen the U. S. Nudcar Regulatory Commission and the Canadian i

Commercial Corporation. During the period of this report the contract resourecs were spent on operation and maintenance of the Eastern Canada Telemetted Network (ECTN), development of special purpose local network systems, Canada,g and maintenance of the strong motion seismograph network in eastern servicin operation and expansion of the Ottawa data lab and carthquake monitoring and reporting.

Of special note in this period was the final completion of both the Sudbury (SLTN) and Charlevoix (CLTN) local networks and the integration of their data

. processing and analysis requirements into the regular analysis stream for ECTN data. Rese actwor.u now acquire high quality digital data for detallcd analysis of seismic activity and r,ource properties from two very different but tectonically r

interesting rituations.

Regular analysis of castern Canadian seismicity is carried out on a MicroVAX computer where seismic events from CLTN, SLTN, and ECTN are systematically classified as Noise, Calibration, Telescism, or Local events. Noise files are eliminated, and the remaining files are archived to tape via OTTVAX, the central DEC VAX11/750 computer. Much of this processing is carried out by a discrimination package called CAUSMO which operates during off-hours and presents the analysts with a dassification of the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of events at the start of each working day. local events are analyzed using an :nteractive graphics disp *ay package (SAM) and an carthquake location package (LOC) both develcped by the Ceophysics Division, Geological Survey of Canada. Seismic data derived from the traces (i.e. phase picks and amplitudes) and locations and magnitudes calculated by LOC are stored in individual PlKFILES for each event.

PikFILE infor. nation is stored in an OTTVAX selational database facility. This facility, based on INGRES, is used to produce the quarterly reports and bulletins detaillng scismic activity.

Publicatlan Date:

April 1990 Prepared by:

R. J. Wetmiller, J. A. Lyons, W. E. Shannon, P. S. Munro, J. T. Thomas, M. D.

Andrew, M. I.,montagne, C, Wong, T M. Anglin, M. Plouffe, J. Adams, J. A.

Drysdale -

Contractor:

Geophysics Division, Geophysics and Terrain Sciences Branch, Geological Sun >cy of Canada, Department of Energy, Mines and Resources, Ottawa, Ontario K1A OY3 Canada Prepared for:

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

- seismic events, carthquakes, Cac-da, scismic detection, seismicity, telemetry i

42

l l

NUREG/CR-1816 PR-EDIl ORNI/rM-10328

Title:

PR EDD: Power Reador Embritt!cment Data Base, Version 1 Program Description Descriptioru Data concerning radiation embrittlement of pressure ves<ci steels in commercial power reactors have been collected from available surveillance reports. The purpose of this NRC sponsored program is to provide the tahnimi bases for voluntary consensus standards, regulatory guides, standard resiew plans, and codes..The data can also be used for the exploration and verification of embatt!cment prediction models. The data files are given in dBASE 111 Plus format and can be accessed with any personal computer using the DOS operating system. Menu.iriven software, provided for easy access to the data, includes curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evalu.ition compared to previous databases.

The current compilation of the Power Reactor Embrittlement Data liase (PR EDD, Version 1) contains results from surveillance capsule repcrts of 78 reactors with 381 data points from 110 different irradiatal base mt.terials (plates and forgings) and 161 data pcints from 79 different welds. Results from heat-affecte6 zone materials are also listed. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR.EDD and will be supplementing the database with additional data and documentation. Periodie updates of data and software will be released to authortred users. Future updates will also include results from irradiations in mateitals test reactors.

Publication Date:

June 1990 Prepared by:

F. W. Stallmann, F. D. K. Kam, D. J. Taylor Contracton Oak PJdge National 1.aboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared fon NRC Division of Engineering, Office of Nuclear Regulatry Rewcatch L ywords:

power reactors, data base management, data comp.htion, embrittlement, pressure vessels, stects, plates, forging, welded joints 43

NUREC/CR-4840 COMPBRN, LIIS, TEMAC SAND 88-3102

Title:

ProcWures for the External Event core Damage frequency Analym for NURTG 1150

==

Description:==

This domment presente procedures which can be used to assess extemal harards at nudear power plants. These methods were used to perform the external events risk assessments for the Surry and Peach Bottom nudear power tants as part of the NPC sponsored NURCG 1150 program. These methods app to the full - range of hatards such as carthquakes, firer, floods, etc. w ich are tollectively kncwn as external events and are bated on making full utilitetten of the power plant systems logic models developed in ths internal events analyses.

Hallmarks of the methods described include the use of extensive computer.alded scroening prior to the detailed analysis of each extemal event hazard to which the plant might conceivably be exposed. These screening procedures identify those external events which could contribute to the isk at the piant and significantly redure the number of events for which subuyent detailed analysis is requirM. Taken together, these techniques provide a relatively straightforward and, In some cases, simplified set of techniques for the analysis of the full range of external events and provide for both scrutability and reprodudbility of the final results.

The COMtURN fire growth code is used in fire assessment analysis to calcula e fire propagation and equipment damage. COMPBRN was developed speelfically for use in nudcar power plant fire probabilistic risk assessments (PRAs). The code calculates the time to damage critical equipment given that a fire has startec' This failure time is then used in conjunction with information on fire suppression to obtain the probability that a given fire will cause equipment failure which leads to core damage before the fire can be suppressed.

COMPDRN uses a zone model breaking the fire eiwironment into three zones:

flame / plume, hot gas layer, and ambient.

Distributions on fire frequency, fire suppression probability, fire code calculations, random failure probability, barrier failure probability, and operator recovery actions generate uncertainties on fire-induced core damage frequencies.

The uncertainty of these values is propagated through the accident sequenm models using two computer codes. A Latin flypercube Sampling (LHS) algorithm is used to generate the samples for all of the parameter values while the Top Event Matrix Analysis Code (TEMAC) is used to quantify the uncertainty of the accident sequence equation using the parameter value samples generated by the LHS code. LHS is a constra!ned Monte Carlo technique which forces all parts of the distribution to be t,ampled. TEMAC uses the LHS parameter samples and the accident sequence equations (cut sets) as input to quantify the core damage estimates. TEMAC generates a $ ample of the accident sequence frequency, a point estimate of the frequency, and various importance measures and rankmg for the base events.

Publicatlwi Date:

November 1990 Prepared by:

M. P. Dohn, J. A. bmbright Contractor:

Sandia National bboratories, P. O. Box 5800, Albuquerque, NM 87183 Prepared for:

N,<C Division of Systems Research, Office of Nuc! car Regulatory Kesearch

(,

Keywords:

nudest power niants, probabilistic estimation, risk assessment, hazards, fires, l

floods, carthquales, fauft tree analysis, frequency ana1 sis, reactor cores, reactor 1

acddents, Surry-1 teactor, Petch Bottomo reactor, SETS codes, SHAKE codes, CLASSI codes 44 L

1

NUREG/CR-4908 MRR and PJRR Databases PNL-6196

Title:

Ultrasonic Inspection Relianility for Intergranular Stress Corrosion Cracks A

Round Robin Study of the Effect'. ;f Perwnnel, Procedures, Equipment and Crack Characteristics DescripHoru A pipe inspection round robin entitlat *htini.Round Robin" was mnducted at Pacific Northwest Laboratory from hiay 1985 through October 1985. The research was sponsored by the U. S. Nudcar Regulatory Commission, Office of Nuclear Regulatory Rewarch under a program entitled Evaluation and improvement of NDE Reliability for Inservice Inspection of Light Water Reactors.*

The hiini Round Robin (h1RR) measu ed the intergranular stress wrrosion (ICSC) crack detection and sizing capabilities of inservice inspection (ISD inspectors that had paswd the requirements of IEB 83-02 and the Electric Powei Research Institute (EPRI) sizing training course. The h1RR database was compared with an earlier Pipe Inspection Round Robin (PIRR) that had measured effective detection prior to 1982. Comparison of the hiRR and PIRR databases indicated no significant change in the inspection capability for detecting intergranular stress currosion cracks (ICSCC). Also, when comparing detection of long and short cracks, nn difference in detestion capability was measured. An

'mprovement in the ability to differentiate between shallow and deeper ICSCC was found when the hiRR sizing capability was compared with an earlier sizing round *obin conducted by the EPRI.

In addition tc, the pipe inspection round robin, a human factors study was conducted in cocjunction with the hiRR. The most important result of the human factors study is that the Relative Operating Characteristics (ROC) curves provide a better methodology for describing insp/no crack data.ector perforrr.ance than on probability of detection (POD) or singkupoint crack Publication Date:

July 1990 Prepared by:

P. C. Hensler, T. T. Taylor, J. C. Spanner, S. R. Doctor, J. D. Delfenbaugh Contractor:

Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352 Prepared for:

NRC Division of Ergineerit.g, Office M Nuclear Regulatory Research Keywords:

nondestructive testing, austenitic steeis, ultrasonic testing, inspection, reliability, perfortnance, human factors, stress corrosion, intergranular corrosion, cracks i

45

}

NUREG/CR-5111 IRRAS2.0 EGG-2535 f

Title:

Integrated Reliability and Risk Analyals System (IRRAS)

Version 2.0 User's Guide Descriptioru The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of the-art microcomputer-bawd pmbabihstic risk assessment (PRA) mohl developni.' il

[

and analysis tool to address key nudcar plant safety issues. It gives the user se ability to create and analyze fault trees and event trees using a persoi.a!

computer. The program provides functions for fault tree and event tree construction and analysis. The fault tree functions range from graphical fault trec constsuction to fault tree cut $ct generation and quantification. The event tree functions include ihn linking of fault trees, defining seddent sequences, k

generating acddent sequence cut sets, and quantifying thent IRRAS2.0 %s all the capabilities and functions requhed to create, modify, reduce, and analyze r

fault tree models used in the analysis of complex systems and processes. Also pmvided in the system is. an integrated full screen editor for use when Interfadng with remote mainframe computer systems.

Version 1.0 of the IRRAS pro 6 ram was relet sed in February of 1987. Since that time, many urar comments and enhancements have been incorporated ir. the program providing a much more powerful and user friendly system. The subject of this user's guide, Versiot, 2.0 of IRRAS, provides all of the same capabilities as Version 1.0 and adds a relational database facility for managing the data and i

improved functiont.lity and algorithm performance.

IRRAS2.0 is written in Logitech MODULA 2. The system m;uires the following hardware: IBM PC/XT/AT, PS/2, or 100% compaSble system with 640 Kbpes main memory, a fixed disk with a minimum of 7.5 Mbytes free, a math m.

4 processor, and a 16-color enhanced monitor. The graphics input device can be the keyboard, a 2, 3, or 4 buttor moua, or a Summagraphics tablet.11 an t

enhanced graphics adapter (ECA) 3 wd, it must have the memory expansion option to extend the four standard

+s to 16. IRRAS2.0 does not support the four-color mode on the EGA adapter.

Publication Date:

June 1990

. Prepared by:

K. D Russell, M. B. Sattison; D. M. Rasmuson Contracton EG&G Idaho, ine., P._ O. Box 1625, Idaho Falls, ID 83415; U. S Nuclear

- Reg.ilatory Commission, Washington, DC 20555 Prepared for:

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

fault tree analysis, reactor safety, probabilistic estimation, failure mode analysis, risk assessment, reliability, system failure analysis, Monte Carlo method, computer graphics, IRRASt.0 codes l

46-

-.,,. _ _ m _. _

=

NUREC/CR-5213, Vol.1 CES

Title:

The Cognitive Environment Simulation as a Tool for Modeling Human Performance and Reliabdity Executive Summary Descriptioru improved methods to mcdel wgnitive behavior of nu personnel. A tool called Cognitive Environment Simulation (CES) was develop for simulating how people form intentions to act in NPP emergencies. CES situations. In addition, a methodology called Cognitive Technique (CREATE) was developed that describes how CES can be uwd to provide input to human reliability analyws (HRA) in probabili$ tic risk assessment (PRA) studies. T1,ls report describes the results of three activities that wr-e performed to evaluate CES/ CREATE: (1) a technical review was conduced by a panel of experts in cognitive modeling, PRA, and lira; (2) CES was performance esist; (3) a workshop with HRA practition

' worked example

• of the CREATE methodology. The intention fornution. Volume 1 provides a summary of the resu results of all three Publication Date:

June 1990 Prepared by:

D. D. Woods: H. E. Pople; E. M. Roth Contracion Cognitive Systems Enginating Laboratory, The Ohio State University, Cohrnbu OH 43210; University of Pittsburgh and Seer Systems, Pittsbur Westinghouse Science arid Technology Center,1310 Beulah Road,gh, PA 15260; 15235 Pittsburgh, PA Prepared fon NRC Division of Systems Research, Office of Nuclear i:e6ulatory Research Keywords:

artificial intelligence, reliability, probabilistic estimation, risk assessment, nudear power plants, human factors, errors, personnel, emergency plans 47

GS NUREG/CR-5213, Vol. 2 The Cognitive Environment Simulation as a Tool for Modeling 31uman

Title:

Performance and Reliability Mah Report The U. S. Nuclear Regulatory Commission is sponsoring a program to devel Improved methods to model cogritive behavior of nuc! car power plan Descriptforu personnel. A tool called Cognitive Environment Simu provides an analytic tool for exploring plausible human responws in emerge situatics in addition, a methodology called Cognitive provide input to human reliability analyses (HRA) in probabilistic risk assessment (PRA) studies. This report describes the results of three activitie d

were performed to evaluate CES/ CREATE: (1) a technical review was (1) (.ES was by a panel of experts in cognitive modeling, PRA, and HRA:

exercised on steam generator tube rupture inddents for which data on operator prformance ex,ist; (3) a workshop with HRA practitioners was h exampic" of the CRCATE methodology. The results of all three evaluations indicate that CES/ CREATE is a promising approach for modeli worke1.

intention formation. Volume 2 provides details on the three evaluations, including the CES computer output for the tube rupture events.

Publication Date; june 1990 D. D. Woods; H. E. Pople; E. M. Roth Prepared by:

Ccpitive Systems Engineering Laboratory, Tho Ohio State Univers 15260; 43210; University of Pittsburgh and Seer Systems, Pittsburgh, PA Contractor:

Westinghouse Science and Technology Center,1310 Deulah Road, P OH 15235 NRC Division cf Systems Research, Office of Nuclear Regulatory Research Prepared for:

artificial intelligence, reliability, probabilistic estimation, risk assessment, n power plants, human factors, errors, personnel, emergency plans, steam Keywords:

generators 48

NUREC/CR-5229, Vcl 2 DAS EGG-2577, Vol. 2

Title:

TMI2 EPICOR Il Resin / Liner Investigation: Low Level Waste Data Base Development Program for Fiscal Year 1689 Annual Report Descriptforu This report summarizes the accomplishments in fiscal year 1989 of the EPICOR !! Resin / Liner Investigation: Low Level Waste Data Baw Development Program funded by the U. S. Nudcar Regulatory Commission.

The 1979 accident at Three Mlle Island released approximately 560,000 gallons of mntaminated water to the Auxiliary and Tuct liandling Buildings. That water was decontaminated using a demlneralization system called EPICOR-!!. The contaminated water was cyded through three stages of organic and inorganic lon exchange media. The first stage of the system was designated the prefilter, and the $ccond and third stages were called demineralizers. Fifty EPICOR ll prefilters with high concentrations of radionuclides were trr sported to the Idaho National Engineering laboratory (INEL) for interim stor./,e before final disposal at the commercial facility in the State of Washington Research is be.1, conducted on materials from four prefilters under thme tasks of the TMI 2 EPICOR !! Resin / Liner Investigation: Low Level Waste Data Base Development Program. In the first task, Resin Degradation, equipment was fabricated and the third resin coring of PF 8 and 20 was made. For the s.ccond task, Rosin Solidification, Pcrtland Type 111 cement and vinyl ester styrene (VES) waste forms incorporating lon exchange resin waste from EPICORll! prefilters were subjected to tests to obtain waste form performance data. Information resulting from this task was presented and discussed at the NRC sponsored Workshop on Cement Solidiflcation. The third tatk, Field Testing, is an examination of the effect of disposal environments on solidified resin war.tes from EPICOR41 prefilters. The purpose of this task, using lysimeter arrays at Oak Ridge National Laboratory and Argonne National Laboratoryl, chemical, East, is to expose samples of solidified ion exchange resin to the actual physica and microbiological conditions of a disposal environment. The study is designed so that continuous data on nuclide release and movement, as well as environmental conditions, will be obtained over a 20 year pericxi, Each month, data stored on a cassette tape are retrieved from the data acquisition system (DAS) and translated into an IBM PC compatible disk file. At least quarterly, water is drawn from the porous cup soil water samplers and the lysimeter teachate collection compartment. Those water samples are analyzed for beta. and gamn # producing nuchdes.

Results of the fourth year of data acquisition are presented in this report. These results show tivt radionuclides are continuing to move from the waste forms and through the soil column. Also some data on waste form performance are presented. VES is comparab!c to cement in retaining Sr.90, unlike findings from savannah River Laboratory which found cement to be a better retainer than VES.

Publication Date:

February 1990 Prepared by:

J. W. McConnell, Jr., R. D. Rogers: E. C. Davis; J. D. Jastrow Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415; Oak Ridge National Laboracory, P. O. Box 2008, Oak Ridge, TN 37831; Arbonne National Laboratory,9700 South Cass Avenue, Argonne,11 60439 Prepared for:

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

Three Mile' Isttnd 2 reactor, resins, liners, waste forms, low level radioactive wastes, radioactive waste disposal, radionuclide migration, performance testing, Portland cement 49

~

~ -

. - ~ ~ - - - -

NUREG/CR-5253 PARTITION SAND 88-2940

Title:

PARTITION: A Program for Defining the Source Term / Consequence Analysis Interface in the NUREC 1150 Probabilistic Risk Assessments User's Guide i

Descriptiotu The individual plant analyses in the U. S. Nuclear Regulatory Commission's reassessment of the risk from commercial nuc! car power plants (NUREC 1150) consist of four parts: systems analysis, accident progression analysis, source term analysis, and consequence analysis. Careful definition of the interfaces between these parts is newssary fer both information flo'v and computational effidency.

This document was designed for users of the PARTITION computer program developed by the authors at Sandia National laboratories for defining the interface between the source term analysis (performed with the XXSOR program) and the consequence analysis (performed with the MACCS program).

The purpose of the PARTITION program is to form groups of source terms with similar properties, One wt of MACCS calculations la performed for each of these groups. The following operations are performed in PARTITION: (1) an early i

fatality weight and a chronic fatality weight are defined for each source term; (2) i the source terms are partitioned into groups of source terms with similar radiological -

tential on the basis of these weights, and a single frequency wel hied source term is calculated for each nource term group; 0) the source terms each source term group are divided into subgroups on the basis of evacuation timing, and a frequency weighted source term is calculated for each subgroup; (4) various summary plots are produced to aid in checking the adequacy of the partitioning, and (5) an ot,tput file that serves as input to the -

consequence analysis is generated. The result of the partitioning process to a subdivision of the source terms on the basis of three dimensions: carly fatality potential,'chrcnic fatality potential, and evacuation timing. To facilitate this division, PARTITION operates in an interactive mode.

This report provides a tutorial that details how the interactive partitioning is performed, along with detailed information on the partitioning process.

PARTITION contains extensive comments and was written in ANSI Standoed IDRTRAN 77 to make the code as machine-independent as possible; however, the code was developed in a VAX8650 computing environment, and the comments in the report assume familiarity with the VAX operating system. The program creates a number of output files, three of which are used by the proprietary SAS graphics software to generate plots of the two-dimensional source terms. Additional files contain bookkeeping information with respect to the pooling of subgroups, source term information required by MACCS for consequence calculations, and the frequency and the early and ' latent cancer

~ fatality weights for each source term.

Publication Date:

May 1990 Prepared by:

R. L Iman: J. C. Helton; J. D. Johnson Contracion Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 871?5; Arizona State University, Tempe, AZ 85287; Science Applications Intemational Corporation, Albuquerque, NM 87102 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulato y Research

= Keywards:

nudcar power plants, probabilistic estimation, risk assessment, source terms, -

reactor accidents, health hazards, delayed radiation effects, early radiation effects, statistics, neoplasms, MACCS codes, XXSOR codes f

i

/

l-i 50

.~..

.c

NUREC/CR-5254 TRAC-PF1/ MOD 1 UNL-NUREG-52168

Title:

Dias in Peak Clad Temperature Predictions Due to Uncertainties in Modeling of ECC Bypass and Dissolved Non Condensable Gas Phenomena Descriptbn:

The U. S. Nuclear Regulatory Commission (USNRC), its contractors and consultants have developed a methodology for evaluating Code Scaling, Applicability and Uncertainty (CSAU). The CSAU method has been demonstrated by applying it to the TRAC PF1/ MODI, Version 14.3 code and its analysis of a large break, loss-of-coolant accident (LBLOCA) for a Westinghouse four-loop plant. In applying the methodology, the accident course is divided into three different phases, namely: blowdown, refill, and reflood. There are two distinct peaks in the clad temp.'rature history, one in the blowdown phase and one in the reflood phase. The peak clad temperature (PCT) of the blowdown phase is governed by fuel characteristics. The peak clad temperature of the reflood phase is governed by the phenomena affectin the refill phase as the cl>d temperature continues to rise almost adiabaticall during the refill phase.

Specifically, the second PCT is affected by critical bre k flow, two phase pump degradation, and the phenomena related to the emergency core cooling system (ECCS) in the downcomer and lower plenum of the reactor vessel.

This report describes a ger.cral method for estimating the effect on the reflood phase PCT from systematic errors (blases) associated with the modeling of the ECCS and dissolved nitrogen, and the application of this method in estimating biases in the reflood phase PCT (second PCT) predicted by TRAC PFl/ MODI, Version 14.3. The bias in the second PCT due to the uncertainty in the existing code models for ECCS related phenomena is 19 K (44' F). The negative bias implies that the code models for this phenomena are conservative. The bias in the second PCT due to the lack of modeling of dissolved N, in the code is estimated to tv 9.9 K (17.8* F). The positive bias implies that the abwnee of a dissolved N model makes the code prediction of PCT non-conservative. The 2

bias estimated in this report is based on full-scale test data from the Upper Plenum Test Facility. Thus, the c>timates pres,nted are unaffected by scale distortions. Data from small size facilities were also available, and an estimate of bias based on these data will be conservative.

Publication Date:

September 1990 Prepared by:

U. S. Rohatgi, L Y. Neymotin, J. Jo, W. Wulff Contracion Brookhaven National 1.aboratory, Upton, NY 119'3 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

reactor accidents, PWR typ reactors, loss of coolant, blowdown, cladding, computerized simulation, ELCS, reactor safety, parametric analysis, test facilities, heat transfer 51

NUREG/CR-5256 Grntr:1 SAND 88-3020

Title:

Components of an Overall Performance Assessment Methodology Descriptioru Both the U. S. Environmental Protection Agency (EPA) and the U. S. Nudcar Regulatory Commission (NRC) have promulgated regulations tegarding the performance of geologic repositories for the disposal of high level nuclear waste.

Specifically, the EPA has promulgated three quantitative, postelosure requirements that apply m the entire disposal system, while the NRC'6 three quantitative, postclosure requirements apply only to particular subsystems of the repository. To assess compliance with all six of these quantitative requirements, the phenomena that can affect the performance of the repository, the processes by which thew phenomena are producai, and the parameters associated with these processes will have to be identified and quantified. In addition, the analyses performed to assess compliance will have to be conducted in accordance with a performance assessment methodology to ensure that all regulatory criteria are addressed. A performance asessment methodology proposed by Sandia National bboratories is composed of scenario development and screening, consequence analysis, uncertainty analysis, and sensitivity analysis. This methodology can be used to assess compliance with the EPA's and NRC's requirements.

Use of the 1.HS (l.atin 11 crcube SamplinM code in mnjunction with uncertainty analysis and ap ication of the CRESS (Gradient. Enhancement Software System) software an a similar system ADJEN when implementing sensitivity analyses basai on the deterministic approach are considered, Publication Date:-

February 1990 -

Prepared by:-

P. A. Davis, L L Price, D. P. Callegos, E. J. Bonano; K K Wahl, M. T.

Goodrich; R. V. Guzowski Contractor:

Sandia National bboratories, P. O. Box $800, Albuquerque, NM 87185; CRAM, inc., 1709 Moon, N.E.,

Albuquerque, NM 87112; Science Applications International Corporation,2109 Air Park Road, S.E., Albuquerque, NM 87106 Prepared for:

NRC Division of lil h imvel Waste Management, Office of Nuclear Material 6

Safety and Safeguards Keywords:

high level radioactive wastas, performance testing, underground disposal, sensitivity analysis, LHS codes, ADJEN mdes, GRESS codes

.- m ;

52

NUREG/CR-5262 PRAMIS SAND 88-3093

Title:

PRAMIS: Probability Risk Assessment Model Integration System Uwr's Guide

==

Description:==

'this document was designed for users of the Probabilistic Risk Assessment Model Integratinn System (PRAMIS) computer program developed by the authors at Sandia National laboratories for easy assembly of the individual parts of the NUREC.1150 plant analyses into overall risk results. PRAMis assembles the following files associated with the NUREC 1150 analyses in rnatrix format to obtain risk: the Latin hypercube sample, the results of the systems analysis, the results of the accident progression analysis, the results of the source term / partitioning analysis, and the results of the conwquence analysis. In addition, various intermediate and conditional quantities are available when requested by user specified inpet: the fractional contribution to risk of individual plant damage states, accident progression bins and source term groups, and c file containing the original Latin hypercube sample and user specified dependem variables for use as input to the proprietary SAS statistical package. This report provides a tutorial that details how to use she PRAMIS program.

The PRAMIS program is written in ANSI Standard PORTRAN 77 to make the code as machine-independent 0.c., portable) as possible. PRAMIS is flexible enough to pe form conditional analyses. For example, the analysis could promed through the various stages conditional upon various specified entities, such as certain plant damage states, or certain accident progression bins, or certain consequences. PRAhllS also allows for analyses based on calculations up through each stage of the analysis, such as after the source term calculations. That is, the calculations cnly have to progress sequentially through the stage of interest and do not have to proceed through consequences. The program is quite simple to use and executes quickly. The CPU times required to execute PRAMIS for all four of the plants considered in the NUREG-1150 analysis ranged from 1 to 20 tninutes on a VAX8650. Much of the processing requires use of the propriciary SAS r.oftware.

Publication Date:

May 1990 Prepared by:

R. L Iman; J. D. Johnson; J. C. llelton Contractor:

Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185; Science Applications International Corporation, Albuquerque, NM 87102; Arizona State University, Tempe, AZ 85287 Prepared for:

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

nuclear power plants, reaaor accidents, systems analysis, source terms, probabilistic estimation, risk aswssment, statistics, sampling, Lit, codes, MAPPER codes, PARTITION codes, SETS codes, XXSOR codes, MACCS codes, TEMAC codes, EVNTRE codes 53

NUREG/CR-5273, Vol. 4 MATPRO EGG-2555, Vol. 4

Title:

SCDAP/RELAPS/ MOD 2 Code Manual, Volume 4: MATPRO-A Library of Materials Properties for Light Water Reactor Accident Analysis Descriptforu The SCDAP/RELAPS code was developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The crde models the coupled behavior of the tractor coolant system, the core, and the fission products and acrosols in the system during a severe accident transient as wcil as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offs [tc power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as inuch of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, tu. bines, condensers, and secondary feedwater conditioning systems.

Volume 4 describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semlempirical In nature. The subcodes contained in this doeurnent are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuc.,

.a caloy zirconium

dioxide, stainless

steel, stainless steel

oxide, cladding, dium. cadmium

alloy, boron

carbide, inconel

718, silver in zirconium. uranium-oxygen melts, and fill gas mixtures. ne document also contains descriptions of the reaction and solution rate models that are needed to analyze a reactor accident.

The descriptive detail provided for the subcodes presented varies because the documentation came from many different resources, including the MATPRO-11 Revision 2 document, a series of informal reports dealing with materials properties subcodes that have been incorporated into SCDAP/RELAPS, and previously undocumented materials properties subcodes that are contained in the

$CDAP/RELAPS computer code or in the MATPRO library. The correlations usal in MATPRO-11 Revision 2 were developed using an extrosive. literature search, whereas later correlations were developed as their need became evident or new and relevant experimental data became available, such as the dissolution model for UO In zircaloy. A less extensive literature scarch was used to develop ihe correlations used to calculate the materials properties in the models developed after the publication of the MATPRO.1 Revision 2 document.

Publicatiori Date:

February 1990 Edited by:

J. M. Hohorst Contracton EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 8M15 Prepared for:

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

power reactors, reactor accidents, physical properties, chemical properties, nuclear fuels, reactor materials, uranium, stainless steels, SCDAP/RELAPS codes 54

NUREG/CR-5316 MELPROG SAND 88 3476

Title:

Melt Progression, Oxidation, and Natural Convection in a Severely Damaged ReaCor Core Descriptioru in the revised Severo Accident Research Program Plan, the U. S. Nuclear Regulatory Commission places a great emphasis on code documentation. This report, which describes work conducted over a period of several years in the MEl. PROG code develo ment project, is intended to support this goal. Of interest here is the late ase of core melt progression. A mWel that treats some of the important physi i procesws that can occur during this phase of the accident is described herein. A number of straightforward examples are given to illustrate the utility of the model and to identify the dominant physical promsses.

The focus of this study is on particle beds that fort. In the reactor core when the fuel rods fragment. Such fragmentation can occur,' hen damaged fuel pellet stacks collapse and when water added to a hot reactor core cau*.es the fuel rcds to shatter. Models discussed are also applicable to debris beds that form in the lower plenum as core structural materials relocate downward. The present report emphasites the effect of gas flow and oxidation.

Publication Date:

February 1990 Prepared by:

S. S. Dosanjh Contracion Sandia National Laboratories, P. O. Dox $800, Albuquerque, NM 87185 Prepared ion NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

reactor cores, reactor accidents, meltdo"n, oxidation, natural convection, fission products, gas flow, fuel rods, fra3 mentation, damage, SCDAP <rdes, CORMLT cod s 55

a c.

C Th

./CR-5366 IITAS2 ORNUCSDfrM-267

,* e]

j v.TAS2. A Three-Dimeasional Transient Shipping Cask Ar.alysis Tool o.

.~

in h..( 1 as M

a..ptioru This report describes the HTAS2 computer program which can be used to assess the thermal behavior of shipping casks.:entaining PWR type fuel asumblies.

s HTAS2 has two parts: a global cask aaalysis and a single assembly analysis, d...

w&4 rr The global cask analysis is a three dimensional lumped parameter model t'or the p ', '. t.

.,y, o

%g '

jf, g' entire saipping cask and its contents. The user has the option of simulating a prefire rttady state, a fire transient, a pottfire transient, or.: final steady state in PrJ -

.cceptable order. Detalls within the fuel assemblies are not resolved in this fg path,n of the analysis. The single assembly analysis is two-dimensional and p,21 s

+gg contains a detailed radiation and conduction mades which can be uwd to

[ps * '

p. -. f

• • present PWR equare pitched fu:1 assemblics. A simple convection model is

^

.t evallable. The boundary condita on the basket walls surrcundirig the y

V. -

assemM;. eu be directly specifica or may come from a previous global cak

~

analysis.

.a.

r (4q i

Good comparison was obtained with other comptational results for both parts of the analysis. Although good agreement was Dund between the angie

%:o assembly results and experimental data, the global cask analysis has not yet g

been validated against execriments. This task is recomtreded as a part of M ',, g future work.

pg HTAS2 is designed to be a control module in the SCALE (Standardized i

Coniputer Analyses for IRensing Evaluation) systan and uses 'wo other SCALE y

modules to perform the analysis: HEATING Dersion 6.1),

heat conduction

(

program used to wlve the single-assembly mcdel, and WHOCAM, a prag am bemg developed to perfo m the global cask analycis.

PuMication Date:

May 1990 Prepired by:

M. W. Wene. G. E. Giles Contradon 02k Ridge National Laboratory, P. O. Box 2006, Oak Ridgc, TN 37831 Prepared fon NRC Office of Nudcar Ma'erial Safety and Saf( tards V 'ywords:

PWR troe reacors. spent fuel casks, heat transfer, thertaal conduction, 3

convection, radiation heating, SCALE codes, HEATING todes, WHOCAM codes e

't 56

NUREG/CR-5376 MACCS EGG-2566

Title:

Quality Assurance and Verification of the MACCS Cod, Version 1.5

==

Description:==

An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consegwnce Code System (M,.C.$) was performed. The QA an<' veri %ation involved..

.! nation of the mde and associated documeetion

_.ne.irtent and correct implementc. tion of the models in an erro free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used in MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are 3

also presented. The QA and verification process roncluded that Version 1.5 of the MACCS code, within the scope and lln.itations of the models implemented in the coda, is essentially error free and ready for widespread use.

Publication Date:

February 1990 Prepared by:

. Dobbe, E. R. Carlson, N. H. Marshall, E. S. Marwil, I E. Tolli Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 8M15 Prepared for:

NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywore quality assurance, verification, inspection, Fortran, computerized simulation

(

1 57 p

v NUREG/CR-5377 MACCS -

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

Title:

Well-Known Probabilistic Risk Assessment Models

==

Description:==

This report documents the results of the work performed for the U. S. Nuclear.

Regulatory Commission in the review of the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and com rison of those models to the chronic exposure pathway models

-im emented in similar codes developed in <muntries that are members of the O

D. The chronic exposures concerned are via the te.Testrial food pathways, the water pathways, the loeghway. grout.dshine pathway, and the inhalatio term resuspended radionuclides pat As indicated by the NRC during discussion of the task, the major effort was on the terrestrial food pathways.

MACCS was compared to the following internationally well-known codes:

ARANO (Finland), CRAC/CRAC2 (USA), NECTAR (United Kingdom). NUCRAC (USA), UFOMOD (Federal Republic of Germany).

A-direct comparison has in many respects proved to be difficult to perform, because of the many basic differences between the approaches choser but the report contains comprehensive descriptions of the various approac.- 5 and default values of most of the important parameters, it also contams numerous remarks and comments at points where the approach chosen (by MACCS or any of the other codes) may have weaknesses or faults; er where the descriptions and manuals are incomplete, difficult to understend, or not consistent with information given in cd & documentation.

Publication Date:

June 1990

- Prepared by:

U. Tveten Contri.ctor:

Institutt for Ercrgiteknikk, Postboks 40, N-20007 Kjeller, Norway l

Prepared for:

NRC Division o' Systems Research, Office of Nuclear Regulatory Research Keywords:.

environmental exposure

pathway, tern... rial ecosystems, food

chains, radionuclide migration, particle resuspension, aquatic ecosystems, probabilistic estimation,- risk assessment, MELCCR codes, ARANO codes, CRAC codes, CRAC2 codes, NECTAR codes, NUCRAC codes, UFOMOD codes 58 4

i NUREGICR-5393 G:ncr:1 SAND 89-1432

Title:

A Review of Techniques for Propagating Data and Parameter Uncertainties in High Level Radioactive Waste Repository Performance Assessment Models

==

Description:==

Techniques for propagating data and parameter uncertainties in high-level waste (HLW) repository performance assessment model; are discussed. Uncertainty analysis techniques ascribe quantitatise measures of reliability to model predictions. Both 10 CFR Part 60 and 40 CFR Part 191 requh e consideration of unwrtainties, induding uncertainties in data and parameters, in the performance assessment of a HLW repository system.

Four categories of uncertainty analysis methods are discussed: Monte Carlo simulation, replacement models (response surface techniques), differential analysis techniques (direct, adjoint, and Creen's function), and geostatistical techniques (stochastic modeling using Monte Carlo simulation and spal analysis). Advantages, disadvantages, and applications of each technique are presented. Propagation of uncertainties through multiple, linked models is also discussed. Application of them techniques to sensitivity analysis is also presented. Sensitivity citalyses can be useful to uncertainty studies because the number of parameters included in the uncertainty anali Sis can be reducal by eliminating those parameters for which the uncertainty fias a minimal effect on the performance variabic(s).

A number of computer programs are described in the discussions of the various methods of analysis. The latin Hypercube Sampling code, LHS, is considered '.n the review of Monte Carlo simulation, and several other codes are mentioned in the discussion of <omputer based differential analysis methods including the Gradient-Enhanced Software System (GRESS), a FORTRAN precompilar which reads a FORTRAN source code to identify arithmetic statements and generate for each such statement a related statement representing its derivative. The derivative values are ti en propagated using the chain rule of differential calculus. The ORIGEN2 code, a point depletion and radioactive decay code, was enhanced by applying the GRESS automated procedure for deriveive calculation as was SWENT, a computer code created to analyze a three-dimensional, steady-state, ground water flow problem. The enhanced versions are named ORIGEN2G and SWENTG. The differential analysis apprcach was also considered for MAEROS, a model that represents multicomponent aerosol dynamics, in this case, the implementation was extremely difficult. Other codes

.tentioned in the review include SENCOF, which derives sc:nsitivity coefficients; ADJMIC, which solves the adjoint equation for transport; and the adjoint code SWATS, in which the canonical form of ordinary differential equations is used.

Publication Date:

March 1990 Prepared by:

D. A. Zimmerman, K. K. Wahl; A. L Gutjahr; P. A. Dwia Contractor:

GRAM, Inc.,1709 Mcon N.E., Albuquerque, NM 87112; New Mexico Institute of M!ning and 'lechnology, Socorro, NM 87801; Sandia National laboratories, P. O.

Box 5800, Albuquerque, NM 87185 Prepared for:

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

radioactive waste management, underground disposal, parametric analysis, sensitivity analysis, mathematical models, stochastic processes, performance teating, high-level radioactive wastes, Moete Carle method, GRESS codes, ORIGEN2G codes, SWENTG codes, MAEROS codes, LHS codes, SENCOF codes, ADJMIC codes, SWATS codes I

4 l

59

NUREG/CR-5398 General' SAND 89-1557

Title:

Tecr{nical Basis for Review of High. Level Waste Repository Mxieling

==

Description:==

Both the U.' S. Environmetital Protection Agency (EPA) :.nd the U. S Nuclear Regulatory Commission (NRC) have promulgated regulations _ regaramg the -

performance of geologic repositories for the disposal of high-level nuclear waste.

One of the-responsibilities of the U. S. Departrnent of Energy (DOE) is to demonstrate compliance with the apprcipriate regulations. The DOE will n'ost likely use extensive numerical modefing to show compliance with the various guantitative requirements. These analyses will then be evaluated by the NRC.

There are different levels of evaluation: peer review, conservatise estimates, use of existing models/ codes, and development of models/ codes by the NRC. The intensity of the review will vary from analysis to analysis, depending on the importance of the analysis, the acceptability of the conceptual model behind the analysis and the solution technique used, and the potential for increasing confidence in the system description, snould the NRC decide to develop its own models/cMes. A'n appropriate level :.f review can be determined by applying these four alteria in a specific manner.

Publication Date:

March 1990

- Prepared by:

L L Pdee, D. P. Galleges, N. E. Olague; K K. Wahl, M. T. Goodrich; D. A.

Brosseau Contracton Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185; GRAM, Inc.,1709 Moon, N.E., Albuquerque, NM 87112: ERCE, Inc., Albuquerque, NM Prepared for; NRC Division of High-Level Waste Management, Office of Nuclear Material

~ Safety and Safegua-ds Keywords:

high level' radioactive wastes, ' underground disposal, performance testing, licensing, regulations, mathematical models, reviews, numerical solution 60

l NUREG/CR-5104,V1 ORNL Fdl..re Record Database ORNL-6566,V1

Title:

Auxiliary Feedwa' m Aging Study Desc*iptioru TWs report docume....he results of an Auxiliary Feedwater (AFW) System ctudy that was conducted for the U. S. Nuclear Regulatory Commission's Nudear Plant Aging Research (NPAR) program. The study reviews historical failure data available from the Nudcar Plant Reliability Data System (NPRDS),

I the Licensee Event Report Sequence Coding and Search System (LER), and the Nudear Power Experience (NPE) databases.

The initial intent of reviewing all three databases was to provide an indication of the validity of relying on only the NPRDS data because it should be theoretically the tmst comprehensive; in fact, all failures that would W reported in LERs should alo be reported in NPRDS, whereas NPRDS should abo include many failures not reported in LERs. However, it became clear during the initial review of fauure data from these databases that a significant fraction of the failure records in the NPE and LER databases was not found in NPRDS. As a result, each record from all three databases was reviewed and combined to fonn a single ORNL database, thereby avoiding redundant entries while establishing a more thorough set of failure records. Although a significant number of failure records (1767 total) are in the ORNL Failure Database, it dearly does not include a large fraction of known historical failures, based on a review of some plant-specific failure data that were available for comparison.

The initial review piocess consisted of sorting each of the three databases by plant, failure date, and component. At this time, the ORNL component assignments were made. 'The combined ORNL database was then processed to climinate multiple entries for the same failure event. The resulting single set of failure records was reviewed again to evaluate and record the method of discovery, the impact on the AFW system, and other pertinent inf~mation. For each component type, the number of failure counts in the da aoase and the associated degradation effect upon the system as a whole were tallied. A total of 517 ceactor years of operating experience for the plants is included in the ORNL Failure Record Database. The plants included were Westinghouse and Babcocl-and Wilcox units that started in or before 1986. A total of 1767 fal;ure counts were reviewed.

The failure histories of AFW System components are considered from the perspectives of how the failures were detected and the significance of the fvlure.

Results of a detailed review of operating and monitoring practices at a plant owned by a cooperating utility are presented. General system configurations and

^ertinent data are provided for the Westingnouse and Babcock and Wilcox units.

/Lblication Date:

March 1990 Prepared by:

D. A. Cauda Contractor:

Oak Ridge National Laboratory, P. O. Bcx 2008, Oak Ridge, TN 37831 Prepared for:

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

auxiliary water systems, feedwater, failures, data base management, surveillance, monitoring, turbines, aging, inspection 61

w y

5 NUREG/CR-5405L MARC SAND 89-1650

Title:

' Analysis of Shell Rupture Failure Due to llypothetical Elevated Temperature Pressurization of the $equoyah Unit 1 Steel Containment Building.

Descriptioru Sandia National Iaboratories, as part of the Containment Integrity 190 grams under the sponsorsh!p of-the Nuclear Regulatory Commission (NRC), has z

developed analytical-techniques for predicting the performance of light water reactor stcei containment buildings subject to loads beyond the design basis. The analytical techniques are based on experience with large-scale steel containment model ter,ts that providM important insights and experimental validation of the analytical methods. As a means of demonstrating these analytical techniques, the NRC asked Sandia to conduct a structural evaluation of an actual steel containment building. The,bjective'of the analysis was to determine the actual pressure capadty and the mode, location, and size of fallere, where a functional definition of failure is used.

H The general purpose structurst' analysis program MARC was used to evaluate g

- two three-dimensional finite element models of the Sequoyah Unit I containment building in order to characterize its ;;cneral shell behavior. Each model consisted entirely of eight node bilinear constrained thin shell elements. The models were cons'ructal using - the three-dimensional mesh generation pre-and post p ncessing code PATRAN.

The _ purpose. of this report is to document the calculations performnd to determine - the pressure limits for the shell-repture mode of fallare. Cencral failure of the containment shcIl is predicmd by application of a failure criierlon to the results from finite element structurs *r.alyses. The failure criterion relates the -' calculated values of strain in the containment plates, due to.

to the ultimate strain limit of the steel. Included Internal pres.urization loading,djustr, ents for factors inherent in finite element in-the failure criterion are a analysis, such as level of detail and element size of the finite element model and variations in material property data. Separate finito element models were used to evaluate the overall free-field behavior of the structure andi the localized

- beh.svlor at a specific penetration location.

Three scenarios of static _ internal pressurization, based on gases building up slowly within the containment shell during a severe accident, were evaluated.

Tios scenarios included pressure loading with temperatures uniformly increasing -

in the model in correspondence to the properties of pressurized saturated steam, pressure _-_. loading with non uniformly increasing. temperatures, and pressure loading.without a corresponding temperature increase, which was done for comparison _with earlier published analyses. It-is cor&ded that thermal effects do not change the overall: response of the structure :;r the general shell failure mode, compared to the response due to pressurization at ambient temperature._

The ' reductmn in. the predicted internal pressure capacity of the containment building at temperature corresponds to the reduction in the ultimate strength of the A516 Crade 60 steel due to the temperature increase.

- Pub'llcation Date: -

February 1990 Prepared by:

J. D. Miller-Contractor: -

Sandia National Laboratories, P. O. Box 5800, Albuquerque, NM 87185 4

Prepared for:

NRC Division of Engineering,( unw of Nuclear Regulatory Research Keywords:

containment buildings, Nuop.1 reactor, PWR type reactors, mactor accidents, steel-ASTM-A516, temp. ratura dependence, pressurizing, finia element method, PATRAN codes i

62

.m m

m m

m

I NUREG/CR-5409 General ORNUTM-11267

Title:

Neutron Exposure Parameters for the Metallurgical Test Specimen in the Sixth Heavy-Section Steel irradiation Series l

==

Description:==

The goal of the Heavy-Section Steel Irradiation (HSSI) Program Sixth irradiation Series is to determine the effect of irra 'lation on the shape and shift of the crack arrest toughness versus temperature curve. Two capsules which contained crack-arrett and Charpy V notch test specimens were irradiated at the Oak Ridge Research Reactor located at the Oak Ridge National Laboratory. Tnese capsules have been disassembled, internal dosimeters have been analyzed, and exposure parameters are presented for each irradiation test specimen. This report describes the computational methodology for the least-squares adjustment of the dosimetry data with neutronics calculations, and presents exposure parameters at each test specimen location for the fluence rate greater than 1.0 MeV, fluence rate greater than 0.1 MeV, and displacements per atom. The specific activity of each dosimeter at the end of irradiation is listed in the Appendix.

The computational process for analyzing the dosimetry data involves a significant amount of data handling and reformatting. A computer program, CALACT, is used to obtain saturated activities and total fluences based on multiple irradiation positions, irradiation history, doslmetry data, and fluence rate spectra. The computer program, ACT, then prepares the dosimeter activity file for the least squares adjastment code, LSL-M2, which determines the damage functions (exposure parameter values) and displacements per atom (dpa) for neutron energies above 10' eV. After these results are obtained for each dosimeter location, they are fitted to an appropriate three-dimensional function tailored for rotated capsules. A second ancillary program, FLXPRO, was used to collapse 56-group fluxes to 20-group fluxes.

Publication Date:

Mzy 1990 Prepared by:

L F. Miller, C. A. Baldwin, F. W. Stallmann, F. B. K. Kam Contractor:

Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared for:

NRC Division of Engineering, Office of Nuclear Regulate < Research Keywords:

radiation effects, dosimetry, cracks, notches, steels, material testing, Charpy test, displacemenc rates, neutron fluence, pressure vessels, LSL-M2 codes, CALACT codes, FLXPRO codes, ACT codes 63

.-n-.-

m ENUREG/CR-5419 PRAAGE-IA-BNL-NUREG-52212-

Title:

! Aging Assessment of instrument Air Systems in Nuclear Power Plants -

j

==

Description:==

NRC. Generic issue 43, " Contamination of Instrument Air Lines," has' been i

unresolved since 1980. The potential seriousness of this issue was reinforced in a 1987 study by the Office for Analysis and Evaluation of Operational Data. Aging of components within compressed air systems leading to degraded function of e

the system, is the sub}ect of this study. This work was performed under the auspices of the NRC's Office of Nuclear Regulatory Research as part of the Nudear Plant Aging Research (NPAR) program.

The objective of this study was to identify all the aging modes and their causes, which should be mhigated to achieve a reliable operation of all safety related alt equipment. Also included is asinterim review of typical maintenance activities for air systems in the nuclear power industry. The Phase 2 effort of this study-

will make recommendations for developing an effective maintenance program industry wide to counter the effects of aging.

An analysis of system -operating experience was conducted on failure data stitained from 'the Nuclear Plant Reliability Data System (NPRDS), Licensee

' Event Reports (LER), and Plant Specific Failure databases. _ Each database was 7 analyzed to determine the predominant failure modes, causes, and mechanisms contributing to system failure. The operational stresses and other parameters to the aging of components were considered in assessing their contributing haractoristics. Other relevant factors such as failure rates, aging

' functional - c fractions, and time-to-failure were extracted for uw in the probabilistic models to predict the importance of particular components and system unavailability-with age. In parallel with the data analysis, a probabilistic analysis on a specif,c plant 4

Probabillstic Risk Assessment (PRA) model was performed. This assessment determined the-wmponents :which. nave the dominant effect on system availability ? The analysis of operating experience data revealed that aging

~

' degradation occurs in the compressed air system and becomes a factor as the system ager Normal wear of the system and contamination of the air dominate

' the problems of = system failure. Existing maintenance. programs. within the industry, lack uniformity, and. quality assurance is not rigorous because - the system is classified as non-safety..

' A PRA model and a computer program (FRAAGE) were developed to reflect the essential features of the Instrument Air (IA) system design and its failure rates.

> The fault tree model of the IA system is given in NSAC-60 along with the failure rate database. These were analyzed using the SETS code. The cutsets from this' analysis and the plant failure-rate data Tor all systems were processed using the IBM PC code,17AACE-1A, which was adapted specifically for this -

work from the PRAAGE-1988 set of codes used for aging analysis. The results.

showed that when the time-dependent effects of aging for the worst case are accounted for there are two significant system effects: 1) system unavailability increases moderately with-age, and 2) relative importances of components change with age. During early operation, leakage in both IA/SA (Service Air)-

piping and support system piping was the most important contributor to system.

unavailability;; during later years, aging _ an muse compressors - and air

- dryers / filters to become increasingly I.nportant.

PublNtice Date: - January 1990

- Prepared by:

M. Villaran, R. Fullwood, M. Subudhi Contractor:

Brookhaven National l_aboratory, Upton, NY 11973 j

Prepared for:

NRC Division of Engineering, Office of Nuclear Regulatory Research

)

%eywords: -

reactor components, aging, failures, reliability, system faibre analysis, failure mode analysis, reactor maintenance, reactor operation, air flow, compres3ed air, risk assessment, nuclear power plants, reactor protection systems, water cooled U

reactors, water moderate.d reactors, reactor instrumentation, SETS codes, NPRDS codes, LER codes 64

NUREG/CR-5421 LAPUR ORNIJTM-11285 Title * -

LAPUR User's Guide

==

Description:==

LAPUR, a computer program written in FORTRAN IV, is ' a mathematical description of the core of a boilleg water reactor (BWR). Its two linked modules, LAPURX, and LAPURW, respedively, solve the steady-state goveming equations for the coolant and fuel and the dynamics equations for the coolant, fuel, and neutron field in the frequency domain. General implementation of the modules is described followed by a detailed description of input and output parameters of LAPURX and LAPURW. Sample input da'a are included, and stability benchmarks are noted.

LAPUR was developed for an IBM PC or compatible miaocomputer mnning the MS-DOS 3.3 or higher operating system.

Publication Date:

January 1990 Prepared by:

P. J. Otaduy, J. March-Leuba Contracion Oak Ridge National Laboratory, P. O. Box 2008, Oak Fdge, TN 37831 Prepared for:

NRC Division of Systems Technology, Office of Nuclear Reactor Regulation Keywords:

BWR type reactors, hydrodynamics, reactivity, feedback, steady-state conditions, reactor cores, cactor stability, dynamics, decay i

65

NUREG/CR-5424 G:ner:1 LA-11667-MS

Title:

Eliciting and Analyr'ng Expert Judgment: A Pr2ctimi Guide Det,cription:

This report describes how to clicit and analyze expert judgment. Expert judgment is defined here to include both the exper's' answers-to' technical questions and their mental processes in reaching an answer. It refers specifically to data that are obtained in a deliber:te, structured manner that makes use of the body of research on human cognition and communication. The aim is to provide a guic.e for lay persons in expart judgment. These persons may be from physical and tagincenng sciencts, mathematics and statistics, business, or the miittary. %Qpund is provided on the uses of expert judgment and n the processes bf which bmans solve problems, induding those that lead to bias.

Detailed guidece is otfered on how to clicit expert judgment ranging from selecting the questions to be posed of the experts to se,ecting and motivating the experts to setting up for and conducting the clicitation. Analysis procedures arc introduced and guidance is given on how to understand the database structure, detect bias and correlation, form taodels, anci aggregate the exper, judgnients.

Four FORTRAN computer program listirgs are included as Appendices:

SAATY is a subroutine using Saaty's own 1982 routine for calculating the weights for a single level (matrix). For more than one icvet, weights can be for each matrix and then combined either by hand calculation or calculatect modification of this routine.

MCBETA is a Monte Carlo uncertainty analysis code for beta distributions. Betas are fit with two supplied estimates either as two percentile estimates and levels or as ona percentile estimate and level and a mean value.

EMPIRICAL is a program which forms empirical distribution functions for a given set of percentiles for multiple experts. Simulation is used to combine weighted aggregations of these distributions according to a specified weig)"ing function. Empirical cumulative distribution functions (for each expert) are samp ed in the simulation using lines connecting the individual points of the i

distribution. The more percentiles provided by the expens, the less influence this linear approximation has on the reselts.

BOOT constructs bootstrap ramples from the or.ginal sample of size n. Each sample is randomly formed independently.

Publicatica Date:

January 1990 Prepared by:

M. A. Meyer, J. M. Booker Contracton Los N.amos National laboratory, P. O. Box 1663, Los Alamos. New Mexico 87545 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

expert systems, knowledge base, sampling, statistics, weighting functions, distribution functions, SAATY codes, MCBETA codes, EMPIRICAL codes, BOOT codes t

l 66

NUREG/CR-5438 G:ncrd

Title:

Basic Considerations in Predicting Error Probabilities in 11uman Task Performance Descripilon:

It is well established that human error plays a major role in the mr.lfunctioning of complex systems. *Dus report takes a broad look at the study of human error and addresses the conceptual, methodological, and measurement issues involved in defining and describing errors in complex systems. In addition, a review of existing sources of human reliability data and approaches to human performance database development is presented. Alternative task taxonomics, which are promising for establishing the comparability of nuclear and non nuclear tasks, are also !dentified. Based on such taxonom'c schemes, various database prMotypes for generalizing human error ratc-s across settings are proposed.

Reviews of the following twelve human reliability databases are presented:

• AIR Data Store daveloped by the American Institutes for Research, which focuses on tasks perfoc.ned in the operation of electronic upipment.

• Aerojet Cencral Maintenance Performance Data Bank developed to predict reliabilities during checkout and maintenance performaam on Titan 11 rocket pmpulsion systems.

13unker Ramo Tables for Predicting Operatic.nal Performance containing probability estimates of operator performance de.ived from 37 experimental studies.

• TEPPS Gechnique for Establishing Personnel Performanm Standards) a small part of a larger set of analytic and probabilistic techniques designed to yield measures of stem effectiveness depending solely on expert judgments.

• OPREDS (

rational Recording and Data SyMems) developed at the Naval Dectronics la ioratory and installed in several shi s f:r sempling all console f

operator actions (e.g., switch activation, button manipulation, etc.).

• AFISC (Air Form inspection and Safety Center Life Scienms Accidant and incident Reporting System) containing data on each Air Force aircraft accident since 1971 resulting in an injury or fatality.

ASRS (Aviation Safety Reporting System) a voluntary reporting system

+

developed by NASA containing over 30,000 reports from pilots and controllers dealing with errors committed in the performance of their duties.

NPRDS (Nuclear Plant Reliability Data Systems) an initial attempt at developing a nuclear human reliability data bank, primarily orientcd toward operating equipment.

• The Safety helated Operator Action Program (SROA), developed at Oak Ridge National Laboratory, containing data obtained during the conduct of exercises in power plant control room simulators ctllected from operating records.

The El RI Operator Reliability Experiments (ORE) program of operator reliability experiments on nuclear power plant training simulators with the ctoperation of six U. S. utilities and Electricito de Franm.

The HEART (Human Errar Assessment and Reduction Technique) model which not only attempts te identify error-producing factors and the strength of their effect, but also offers an array of defensive measures which can be employed to counteract tMr effects.

• The Nuclect Computerized t.ibrary for Asse2 sing Reactor Reliability (NUCLARR) an automated database manager icnt system used to procc.,s, store, and retrieve human and equipment reliability data in a ready-to-use format. It provides human error and hardware failure rate data to support a variety of analytical techniques for assessing risk.

Publication Date:

Ap.il 1990 Prepared by.

E. A. Fleishman, L C. Buffardi, }. A. Allen, R. C. Cask.ns ill Contracion Center for Behavioral and Cognitive Studies, George Mason University, Fairfax, VA 22030 Prepared for:

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

human factors, reliability, performance, errors, data base management, taxonomy, comparative eve 2aiions, NUCLARR cades, NPRDS codes, SROA codes 67 m

i NUREG/CR-5447 SCDAP/RELAPS EGG-2574

Title:

Depressurization as an Accident Management Strategy to Minimize the Consequences of Direct Containment Heating Descriptforu

_ Probabilistic Risk. Assessments (PRAs) have identified severe accidents for noclea power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effesdveness of intentional depressurizadon during a station blackout TMLB' eequence was evaluated mnsidering the phenomenological behavior, hardware performance, SCDAP/RELAP5. performance. Phenomenological behavior was calculated and - operational RCfr were considered. One, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) e e latched open at steam generator dryout. The sc.cond, called late depressurir;. gry assumed that the head vent and 'PORVs were latched open at a core exit temperature of approximately 922 K (1200* F). Depressurization to a low value that may mitigate DCH was predicted prior +o reactor pressure vessel breach.

The strategy of late depressurization is preherred over early depressurization because there are greater opportunities to recover plant functions prior to core damage, and failure uncertainties are lessened.

SCDAP/RELAPS is a light water reactor transler.t analysis code capable of analyzing plant transients, such as large-and smdll-break loss-of-coolant accidents (LOCAs) and station blackouts (SBOs). The thermal-hydraulics portion -

of the code (RELAPS) uses a onedimensional, nonequilibrium hydrodynamic model. Each of the two phases, liquid and vapor, is represented by three I

equations-(energy, momentum, and continuity) for each control volume. Heat transfer between the coolant and solid sauctures, such as pipe walls, fuel rods, and steam generator tubes, is modeled as a source within the control ulume.

One-dimensional conduction is used to model heat transfer within solid.

structures. Specific models are included for plant components, such as valves, accumulators, turbines, and control systems.

- The severe core damage analysis package (SCDAP) models the energy generation and mechanical degradation of the fuel rods during a severe core damage accident quence. SCDAP includes models for fuel rod ballo ning and rupture, cladding oxidation for both fuel and control rods and control rod guide tubes, double-sided oxslation of claading if the fuel rod has ballooned and ruptured,

. candling ' of fuel rods,- dissolution of fuel by molten cladding,- formation of resolidified crusts, formation of molten pools, formation of rubble debris beds

caused - by fracture of embrittled cladding, failure of supporting crusts and relootion of molten corium to the lower plenum, and heatup and failure of the lower head due - to creep rupture or : melting using a two-dimensional, finiteelement heat conduction analysis, in addition to thermal-hydraulics and severe core damage analynis capabilities,

- the TRAP MELT fission product transport models have been integrated truo the code. However, the fission product models were not utilized in the transient analysis.

. Publication Date:

October 1990 Prepared by:

D. J. Hanson, D. W. Golden, R. Chambers, J. D. Miller, B. P, Hallbert, C. A.

Dobbe Contracton EC&G Idaho, Inc., P. O; Boy 1625, Idaho Falls, ID 83415 Prepared fon

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

probabilistic estirnation, risk assessment, nuclear power plaras, reactor accidees, depressurization, management, containmer.t, reactor cooling sy;tems,- heatine,,

pressure vesseb, relief valves, blackouts, loss of coolant, Surry-1 reac'or, TRAP-MELT codes, HECTR codes 68

NUREG/CR-5449 G:ncral OR'NUTM-11350 l

Title:

Determination of the Neutron and Gamma Flux Disnibution in the Pressure Vessel and Cavity of a Boiling Wster Reactor

==

Description:==

The Grand Gulf Boiling Water Reactor (BWR/6), owned and operated by Mississippi Power & Light Company, was analyzed to determine the neutron and gamma energy spectrum and flux levels m regions from the reactor vessel throughout the concrete shield wall.

Several two-dimensional and one-dimensional transport calculations were performed for the Grand Gulf reactos configuration. The results from these calculations were synthesized to obtain the threc<limensional neutron flux spectra and dosimeter activities. The results from the transport calculations indicate the flux above 1 MeV peaks near the axial mi<%1ano and azimuthal angle between 40* and 45*, depending on the radial locations. The peak flux above 1 MeV incident on the vessel and at midcavity is about 1.82 x 10' and 1.07 x 10' n cm'2 s, respectively. The vessel 4

fluence accumulated during Cycle 2 and after 32 effective full power years is about 4.41 x 10 and 1.M x 10 n cm s, respedively. The peak flux above 1 4 4 M3V at the front of the concrete shield wall,15.24 cm (6 in) into the concrete wall, and 30.48 cm (1 ft) into the concrete wall is about 7.91 x 10', 7.24 x 10',

and 6.44 x 10' n cm s. respectively.

i4 The results obtained from the gamma calculations show that the peak gamma at the 0-T location of the reactor pressure vessel has t. value of 2.54 x heating /g of stainless steel (SS 304). The peak gamma absorbed do<e rate at the 4

10 W midcavity is about 7.31 x 10' rad /h at full power operation.

A new computer code, 'IORT, has been developed for solving the threedimensional transport equation but has not been released. Computation costs for such 3-D calculations are very expensive and not practical for routine applicat'ans. In this study, the 3-D flux distribution was obtained by using the DOTSYN computer code, a module of the LEPRICON system. DOTSYN is based on the single-char.nel synthesis method, in which channel fluxes for r-theta, r-3, and r channels are used to obtain synthesized 3 D (r, theta,z) fluxes. Several discreto o,. inates transport calculations using the DOT-IV code were performed to obtain she channel fluxes that describe radial, azimuthal, and axial flux distributio-inside the reactor vessel and the reactor cavity of the plant.

Calculation of neutron source for DOT-IV r-the.a runs was performed using the DOTSOR code, another member of the LEPRICON system. DOTSOR generates the r-theta source for DOT-IV transport calculation based on the given x-y core power distribution.

A new version of the sat

• OR nudcar cross-section library was used to obtain all the nuclear data necesrary for the calculations in this study. In the r,cw SAILOR library, the thermal cross-section values for Fe, Cr, Mn, Ni, and H we'e modified from those in the original SAILOR to incorporate a more realist c thermal spectrum.

Publication Date:

June 1940 Prepared by:

M. Asgari, M. L Wiluams, F. B. K Kam Contractor:

Oak Ridge National laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared for:

NRC Division of F.ngineering, Office of Nuclear Pegulatory Research Keywords:

BWR type reactors, neutron transport theory, radiation monitoring, pressure vessels, DOTSTN codes, DOT-IV codes, DOTSOR codes, LEPRICON codes, TORT codes, SAILOR codes 69

NUREG/CR-5453, Vcl. 5 General SAND 89-2509, Vcl. 5

Title:

Background Information for the Development of a Low Level Waste Performance As<,essment Methodology Compute-Code implementation and Assessment De*cription:

This report documents the implementation and assessment of computer codes for a low-level waste performance assessment methodology. Computer codes and analytical. solutions are implemented for ground water flow and transport analyses, ource-term analyses, surface-water transport analyses, ah transport analyses, food-chain analyses, and dosimetry a.alyses. The capability has been retained to perform either simple or more complicated analyses of the source term and ground-water transport aspects of the performance essessment. The simple approaches consist of analytical and simple numerical analyses that are appropriate for relatively almple conceptual models. For fully multi-dimensional or transient problems, more complicated numerical solutions are recommended.

Details are given of the recommendet analytical methods, together with sensitivity analyses that demonstrate important aspects of the solutions. The problems that arose during implementation are discuswl.puter codes and th imp!cmentation processes for the more complianted com FinsJly, a comparison is given between the simple and complicated ground water transport analyses for a simplo conceptual model.

The computer progr ms implemented and discussed in this report are VAM2D, FEMWATER, BLT, GENil, DISPERSE, and SURFACE. Difficulties in achieving numerical convergence using FEMWATER for strongly nonlinear vado& zone soil properties were traced tc, the use of successive substitution (Picard iteration) in the FEMWATER solution algorithm. Neither FEMWATER nct VAM2D converged for the conceptual model when successiva substitution was used, but VAM2D converged using Newton-Raphson iteration.

Documentation -

.'iscrepancies and software bugs made VAM2D difficult to implement, but these were resolved, and the code is now considered satisfactory for use. When implementing BLT, first consider using FEMWATER for the flow field since the two - codes are appropriately coupad.

""he VAM2D flow field must be post rocesse1 to make it consistent with the BLT input format. GENil was easy

- to i lement, and its performance found to be satisfactory.

- Resul s from an analytical solution. as implemented in DISPERSE, are compared -

to FEMWATER/BLT predictic,ns of well concentrations for soil pro rties and found to bc qualitatlvely. similar, but the simple aoproach p uces more conservative results, in addition, results from DISP'ERSE and VAM2D are compared. Again, the results are qualitatively similar, with DISPERSE producing higher, mora conservative, well concentrations. Further : work is necessary to determine the conditions under whlen the simple analysis is not conservative.

Publication Date:

August 1990 Prepared by:

~ M. W. Kozak, M. S. Y. Chu, J. T. McCord; P. A. Mattingly; J. D. Johnson Contractor:

. Sandia National Laboratotics, P. O. Box 5800, Albuquerque, NM 87185; Science Applications laternational Corporation, Albuquerque, NM 87106; Applied -

Physics, Inc., Albuquerque, NM 87109 Prepared for:.

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

low-level radioactive wastes, waste management, performance testing, round water, surface waters, food chains, environmental transoort, dosimetry, "AM2D codes, FEMWATER codes, DLT codes, GENIl codes, DbPERSE codes, SURFACE codes 70

NUREG/CR-5168 QADS ORNIJCSD/TM-270

Title:

QADS: A Multidimensk.nal Point Kernet Analysis Module

==

Description:==

QADS is a multidimensional point kernel amputer code that utilizes the simplified free-form input of the SCALE rystem as.vell as compatibility with ORfCEN-S produced sources, SCALE cross section libraries, and standard composition data sets. QADS consists of a preprocessor that takes free-form input and prepares input specific to the widely available QAD-CCCP coac, which is then automatically executed by a driver module. Th's report describes the int kernel theory briefly, followed by numerous rips on successfully appl ng the theory to various types of shielding problems. The remainder of the ocunient is devoted to descriptions of QADS code input and output with several illustrative sample problems.

Publication Date:

May 1990 Prepared by:

'B. L Broad 5 cad Contracton Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared fon NRC Division of Safeguards and Trrnsportation, Office of Nuclear Material Safety and Safeguards Keywords:

point kemels, gamma radiation, dose rates, buildap, shielding, SCALE codes, MORSE codes, ORICEN codes l

71

NUREG/CR-5473 -

- OCA-P ORNLerM-11450

Title:

Inclusion of Unstable Ductile Tearing and Extrapolated Crack Arrest Toughness Data in PWR Vessel Integrity Assessment

==

Description:==

Over the past several years, the IInvy-Section Steel Technology Program at Oak l

Ridge. National Laboratory performed a

series of large-scale fracture-mechanics experiments. inese experirnents - have demonstrated that prototypical nudear reactor vessel steels can exhibit crack arrest toughness values considerably above 220 MPa SQRT(m) although arrest can be followed immediately by unstable ductile tearing. This report evaluates the influence of the crack-arrest toughness above 220 - MPa SQRT(m) on the integrity assessments of nuclear reactor pressure vessels for pressurized thermal shock (MS) loading conditions, taking into account the potential for unstable ductile tearing following arr The influence of : the high crack-arrest toughness data and unstable ductile tearing on pressurized water reactor vessel integrity assessment is FIS transient dependent. It appears that the potential benefit from crack-arrest events corresponding to toughness values above 240 MPa. SQRT(m) for low upper shelf weld (LUSW) material and above 370 MPa SQRT(m) lor those vessels not.containing LUSW material will usually be negated - by unstable ductile tearing.

Analyses were performed using the OCA-P computer program to determit.e the influence of the steepness of the Kla toughness curve on the cleavage fracture response of a nuclear reactor vessel subjected to a MS transient. OCA-P was developed specifically for simulating the cleavage fracture response of a reactor pressure vessel subjected to a PTS event. The program is based on linear clastic frauure-mechanics (LEFM)- theory and.is capable of performing both deterministic and probabilistic fracture-mechanics analyses.

Publication _ Date:

May 1990 Prepared by;

' T u Dickson, R. D. Cheverton, D. K. Shum Contractor:-

Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831

Prepared fow TNRC Division of Engineering, Office of Nuclear Regulatory Research Keywerds:

PWR type reactors, pressure vessels, fracture mechank:s, ductility, thermal shock, crack propagation, transients d

72 gd l

. a

f f

NUREGICR-5475 OPUS

Title:

Model Feasibility Study of Radioactive Pathways From Atmosphere to Sarface Water Descripfaru A feasibility study of the atmosphere to surface water radionudide pathways was performed for small catchments using a physically-based hydro-ecosystem model, OPUS. Detailed time-intensity (breakpoint) precipitation records from Arizona and Georgia were used as input to drive the model. Tests of model sens;tivity to distribution coefficients, Kd, for Cs-137, Cs-134, and Sr 90 illustrated different vegetation soil crosion-runoff pathways in response to agticultural management practices. Results reflected the fact that low k'd values allow a radionuclide to infiltrcte into the soil proPlc and isolate it from subsequent nmoff and crosion. Of the radionudides and physical settings studied, only the St-90 with low Kd values is sufficiently mobile and long-lived to be removed frcm the system via percolation below the root zone. Conversely, highly-adsorbed radianudides were subject to removal by adsorption to sediment particles and subsequent runoff. Comparison of different effective half lives of I-131 demonstrated the importance of the timing of an crosion-runoff storm event during or immediately af?ct a failout event. Seasonal timing of a fallout event and crop management also affect the fate of this short-lived radianudide. Removal by solution to surface-water runoff was negligible for all nuclides studied. Model simulation results for up to 10 half lives are corroborated by results from long term field studies. These results show the feasibility of modeling pathways in smali catchments using OPUS.

The hydro-ecosystem model OPUS simulates runoff and pollutant transport in and from an agricultural area. The model is assembled from detailed physically based process mtdels for surface-and soi! wate-flow, agricultural-management practices, crosion and sediment transport, plant growth, nutrient cycling, soil-heat flux, and chemicil transport. Currently, the use of the model is limited to areas that: 'a) have an areally homogeneous soil regime: (b) do not have extensive channel networks, (c) have a uniform cropping / management system; and (d) receive precipitation which can be described by the record of a single gage.

The principal processes which determine radionudide fate are largely analogous to those of pesticides: transport, adsorption, and decay, Thus, the OPUS pesticide submodel was cuended to encompass the physio-chemical and radiologicd behavior of radionudides. Major modifications to OPUS involved assignment of fixed adsorption coefficients 'o each radionuclide as contrasted with organic adsorption coefficients which depend on organic matter content.

Changes were required to hold decay rates constant, rather than allow these to vary as a function of soil moisture, temperature, end o ganic content, as is the case for pesdcide transport and associa'ed environmental degradation. A kinetic adsorption model capability was developui for processes where this is required.

The OPUS code is currently undergoing final testing and documentation at the Agricultural Research Seruce, Fort Collins, Colorado.

Publication Date:

March 1990 Prepared by:

R. E. Smith, R. M. Summer, V. A. Ferreira Contractor:

U.

S.

Department of Agriculture, Agricultural Research

Service, Hydro-Ecosystems Research Group, P. O. Box E, Fort Collins, CO 80522 Prepared for:

NRC Offices of Nudear Reactor Regulation anci Nuclear Material Safety and Safeguards Keywords:

surface waters, radionudide migration, environmental transport, runoff, atmospheric precipitations, ecosystems, adsorption 73

\\

8 NUREG/CR-5476 ABAQUS SAND 89-2603

Title:

_ Posttest Analysis _ of _ a 1:6-Scale Reinforced Concrete Reactor Containment Building-

==

Description:==

b an experiment conducted at Sandia Stional Laboratories, a 1:6-sale model of a reinforced concrete light water reactor containment building was pressurized I

with nitrogen gas to more than three times - its design pressurn. The pressurization produced one large tear and several smaller tears in the steel liner plate that functioned as the primary pneumatic seal for the structure. The data collected from the overpressurization test were used to evaluate and further refine methods of structural analysir that can be u:cd to predict the performance of containment buildings under cunditions produced by a severe accident.

This report descrites posttest finite element analyses of the 1:6-scale moc 31 test and compares pretest predictions of the structurul response to the experimental results. Strains and displacements calculated in axisymmetric finite element analyses of the 1:6-sale model are compared - to strains and displacements measured in the experiment. Detailed analyses of the liner plate are also described in the report. The results from these analyses indicate that the primary.

mechanisms that initiated the tear can be captured in a two-dimensional finite element model. Furthermore, the analyses show that studs which were used to anchor th3 liner to the concrete wall played an important role in initiating the liner-tear. Three-dimensional finite c!cment analyses of liner plates loaded by studs - are Mso presented. Re:ults from the three dimensional analyses are compared to msults from two-dimensional analyses of the same problems.

A tcries of four finite element ar.alyses was made of the 1:6-scale containment ustig an axisymmetric shell model. In each analys's, a different set of material parameters was used to define the stress-strain response of the concrete. Te

+

concret model available in Version 4-5-171 of the ABAQUS code was used in all four cases. In the first two shell analyses, the concrete was treated as an clastic perfectly-plastic material with a yield surface defined to reflect the difference between the tensile strength and the compressive strength of the concrete, in the first test, the tensife " yield" strength was set to 500 psi, the l

ul'imate tensile strength of the concrete as estimated from experiments; in the l

second, it was set to 10 psi to sintulate a no-tension material. In the third analysis,instead of assuming perfect plasticity, the concrete was ellowed to crack i

and subsequently soften using the smeared cracking approach lii isBAQUS, and l

In the fourth analysis, the convrete was veated as a material with no tensile streagth.

Publication Date:

February 1990 L

Prepared by:

J. Randall Weatherby l

l Contractor:

Sandia National Labora%1cs, P. O. Box 5800, Atbuqr,c, NM 87185

- Preparea for:

- NRC Division of Engineering, Office ut Nuclear Regula'ory Research p

- Keywords:

containment buildings, reactor components, ren:orced concrete, structural models,- finite element method, strains, stresses, water cooled reactors, water moderated reactors L

74

NUREG/CR-5677 COMPURN

Title:

An Evaluation of the Reliability and Usefulness of External-Initiator PRA Methodologies Descriptioru his report, prepared to assist policy-level decision-makers, evaluates the extent to which each category of external-initiators probabilistic risk assessment (PRA) methodology produces reliable and useful results and insights, at its current state-of-the-art level. The report addresses this need in the following five categories of external initiators: (1) carthquakes, (2) internal fires, (3) ext;rnal floods, (4) extreme winds, and (5) transportativn accidents. Each initiator is examined separately. The thrust is to describe the principal aspeds of the current state-of the-art PRA methodology, identify what aspects are less robust and therefore provide less reliable insights, and determine why.

In almost all fire PRAs to date, the COMPBRN code has been used. The original code has been modified twice, and the most recent version, which removes many of the conservatisms and corrects some of the known errors in the earlier versions, is COMP 3RN UL The code was developed to calculate scenarios involving an oil fire bencaO cable trays. It uses a zone model with three zones:

the flame and plurn, a hot gas layer, and the ambient surroundings. Models predict the growth of the fire and the thermal environment at various locations around the fire as a funcion of time.

Earthquake variabilities are usually accounted for by using several time histories, each of which captures the correlations properly for itself; the set o' time histories capture, as an casemble, the variability from earthquake to earthquake.

The Lawrence Livermore National Laboratory code SMACS was spccially developed for this analysis.

Publication Date:

January 1990 Prepared by:

R. J. Budnitz, H. E. Lambert Contracion Future Resources Associates, Inc.,2000 Center Street, Berkeley, CA 94704 Prepared fon NRC Division of Sys' ems Research, Office of Nuclear Regulatory Research Keywords:

pmbabilistic estimation, risk assessment, reliability, decision making, nuclear power plaats, fires, carthquakes, floods, wind, reactor accidents, SMACS codes I

75

NUREG/CR-5480 Grn ral ORNUTM-11399

Title:

Data Summary Report for Fission Product Release Test VI 3

==

Description:==

Test VI 3, the third in a series of high-temperature fission product release tests

- in the vertical test apparatus, was conducted in flowing steam. The test specimen was a 15.2.cm-long section of a fuel rod from-the BR3 reactor in Belgium, which had been irradiated to a burnup of 42 mwd /kg.' Using an induction furnace, it was hected under simulated LWR accident conditions to i

two test temperatures,20 minutes at 2000 K and then 20 minutes at 2700 K.

J l

.The dadding was completely oxidized during the test, and very little melting or fuel-dadding interaction had occurred. Based on fission product inventories measu.ed in the fuel or calculated by ORIGEN2, analyses of test components showed total releases from the fuel of 100% for "Kr,5% for *Ru, 99% for "5Sb, and 99% for both *Cs and ":s Small release fractions for many other fission products' were detected.

'n addition, _ver small amounts of fuel

' material-uranium and plutonium-were released.y The total mass released from the furnace to the ellection system was 3.17 g; 78% of which was collected on the filters. The results from this test were compared with previous tests in this series and'with a commonly used model for fission product release.

I The fission product ' inventories, as. measured in the fuel and calculated by -

ORIGEN2, and a description of the test procedure and conditions are included.

The specific objectives of this program are to obtain fission product release and behawr data applicable to the analysis of reactor accidents, and to apply these data to the development of VICTORIA and ocher release and transport models.

Comparison of the release data from this test with the results of previous tests

'l' provided evidence that the CORSOR-M medel overpredicts the release of volatile f;ssion products by factors of 3 to 10 in the temperature range of 2300 to 2700 i

K. SOLCASMIX calculations indicated that as much as a few percent of the UO3 could have been vaporized under the conditions of this test.

- Publication Date: - June 1990_

i Prepared by:

M. F. Osborne, R. A. Lorenz, J. L Collins, J. R. Travis, C. S. Webster, H. K. Lee.

T. Nakamura, Y.-C. Tong -

Contracton Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared for:

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

fiss!on products, fission product release, very high temperature, reactor accidents, performance testing, induction furnaces, nuclear fuels, BR 3 reactor, oxidation, ORIGEN2 codes, VICTORIA codes, CORSOR.M codes, SOLGASMIX codes 76

NUREG/CR-5506 General UCID-21831

Title:

Preliminary Structural Evaluation of Trojan RCL Subject to Postulated Kn'V Suppoit Failure

==

Description:==

This report describes a preliminary structural evaluation made to determine whether the reactor coolant loop (RCL) piping of the Trojan nuclear power plant is capable of transferrir g the loads normally carried by the reactor pressure vessel (RPV) supports to other component supports in the RCL system if the RPV supports should fail, say from radiation damage. For the evaluation, the computer model of the RCL systein of Unit 1 of the Zion nuclear power plant is used because of its availability; the RCL systems of the two plants closely resemble cach other. As a bounding case in the evaluation it is postulated that all four RPV supports have failed. Two load combinations are evaluated: (1) the combination af dead weight, operating pressure, and the safe-sbutdown earthquake, and (2) the combination of deid weight, operating pressure, and a lot.s-of-coolant acddent. Both load combinations are classified as Level D Service Limits in accordance with the ACME Boiler and Pressure Vessel Code. Static and i

dynamic linear clastic analyses are conducted to comply with rules specified by Subsection NB in conjunction with Appendix F, Division 1,Section III of the ASME Code Results c,f this prelin.! nary evaluation indicate that ASME Code Appendix F tequirements are satistfied by each of the load combinations

_ to the conc. sion that the Trojan RCL piping considered in the analysis, leading,V support loads to the steam generator and is capable of transferring the RP reactor coolant pump supports.

The Zicn Station RCL computer analysis model was originally developed for lawrence Livermore National laboratory's Load combination Program to be used to perform linear clastic analyses of the RCL system subject to either carthquake input motions or static loads such as dead weight, thermal loads, and intemal pressure. The input fortnat of the model is compatible with the finite-elemer* computer codes SAP 4 and CEMINL The original model has 339 nodes. The model utuizes beam clemens to mcVel componant supports, stiffness elements to represent nozzle effects, and pipe elements to si nulate piping, steam g nerators, reactor coolant pumps, the reactor pressure vessel, and the pressurizer For this analysis, the original model has been reduced by removing the surge line and the pressurizer. The reduced model has 282 nodes (234 unconstrained and 48 constrainal), 33 beam elements for static analyses or 37 for dynamic analyses,16 suffness elements, and 224 straight and beat pipe c!cments.

Publication Date:

January 1990

. Prepared by:

S.C.Lu Contractor:

lawrence Livermore National Laboratory, P. O. Box 808, Livermore, CA 9050 Prepared fon NRC Division of Engineering, Office of Nudcar Regulatory Re:carch Keyworda:

nuclear power plants, reactor cooling systems, Trojan reactor, structural models, supports, failures, Load Combination codes, SAP 4 codes, CEMINI codes l

n

.. -. - ~. - ~

i NUREG/CR-5510 -

TIRGALEX SAIC-89/1744

Title:

.- Evaluations of Core _ Melt-Frequency Effects Due to Component Aging And Maintenance -

==

Description:==

A methodology is developed to incorporate aging effects into Probabilistic Risk Analyses (PRAs). The methodology separates the PRA analyses from the aging models, allowing available PRAs to be used efficiently in evaluating risk effects of ag g. The methodology was applied to two NUREG-1150 PRAs using aging rate ata that was developed for active components. Various surveillance and maintenance programs were evaluated to cletermine their effects in controlling aging. Both point evaluations and uncertainty evaluations were carried out. The results of the applications showed the sensitivity of aging effects on core melt frequency to the efficiency of the maintenance and surveillance program in managing. aging effects. The dadled contributnrs to the aging effects showed relatively few _ components c,ntributing, implying that prioritized aging management programs would be most effective in controlling risk.

The core melt' frequancy as calculated in a PRA is a function of the component unavailabilities, structure failure probabilities, and initiating event frequencies. A standard PRA code, CAFTA, was specifim!!y modified to calculate the sensitivity.

coefficients for any PRA.

The aging of active components was modeled using the linear failure rate aging model developed in the U. S. Nuc! car Regulatory Commission's Nuclear-Plant

- Aging Research (NPAR) program. In the linear aging model, the component

=

failure rate linearly increases v.'ith age according to a charreteristic aging rate. To demonstrate the methodology, four aging rate databases were used titled TIRGALEX, TIRGALEX-MOD 1, TIRGALEX-MOD 2, and TIRCALEX-MOD 3.

. The TIRCALEX aging. rates were estimated in the NPAR program - to help prioritize' aging effects - for research. TIRGALEX-MODI - contains changes for-motor-driven pumps (a factor of 5 increase), motor-operated valves (a factor of 4 decrease), and check valves (a factor of 100 increase). TIRGALEX-MOD 2 contains modified aging rates for di-21s, pumps, and valves corresponding to doubling 4

j times of 20 -years? For TIRGALEX-MOD 3 aging rates of I x 10 per hour pt.r p

year were assigned to motor-dr.'ven pumps, turbine pumps, diesels, check valves, and motor-operated valvts. This aging _ rate corresponds to one failure occurring l

on the average after 5-years due to aging; it is elso equivalent to an.

unavailability _ contrbution - of approximately 0.1 after 18 months, - if no intermediate corrective measures are taken.

Publication Date:' June 1990 l

Prepared by:

W. M Vesely, R. E Kurth, S. M. Scalzo Contractor:.

Sdence, A >plications, mternational Corpo2ation,- 2941 Kenny Road,- Suite 210, k

Columbus, OH 43221 l

Prepared for:

NRC Division of Enginecting, Office of Nuclear Regulatory Research Keyword's:

aging, risk assessment, reliability, failures, probabilistic estimation, maintenance, BWR type reactors, PWR type reactors, reactor cores, nuclear power plants, CAFTA codes h

l L

78 r-~

L

- - ~ ~

NUREG/CR-5512 General PNL.7212 i

Title:

Residual Radioactive Contaminahon From Decommissioning Technical Easis for Translating Contamination Levels to Annual Dow Draft Report for Comment Descriptioru This report describes the generic modeling of the total effective dose equivalent (TEDE) to an individual in a population from a unit concentration of residual radioactive contamination. Radloactive contamination inside buildings and soil contamination are considered. Unit concentration TEDE factors by radionuclide, exposure pathway, and exposure scenario are calculated. Reference radiation exposure scenarios are used to derive unit concentration TEDE factors for about 200 individual radionuclides and parent-daughter mixtures. For buildings, these unit concentration factors list the annual TEDE for volume and surfan contamination situations. For soil, annual TEDE factors for unit concentrations of radionuclides in soil during residential use of contaminated land and the TEDE per unit total inventory for potential use of drinking water from a ground water source are presented. Because of the generic treatreent of potentially complex ground-water systems, the annual TEDB factors for drinking water for a given inventory may only indicate when additional site data or modeling sophistication are warranted. Descriptions are provided of the models, exposure pathways, exposure scenarios, parameter values, and assumptions used. An analysis of the potential annual TEDE resulting from reference mixtures of residual radionuclides is provided to demonstrate application of the TEDd factors.

To find a computerized model for use in calculating unit. concentration pathway dose conversion factors, an initial review of (xisting radiation pathway analysis models selected the ONSITE/MAXll computer program because of its flexibility in allowing the user to define various exposure scenarios. Its predecessor, GENil, is designed for estimating individual and population doses from releases of radionuclides to air, water, or soil and includes an enhanced user-defined scenarin capability.

PHASE 8, written in FORTRAN, uses output from GENIl to produce ' dose rate or committed effective dos 4: rate conversion factors and the TEDE rate factors.

Publication Date:

January 1990 Prepared by:

W. E. Kennedy, Jr., R. A. Peloquin Contractor:

Pacific Northwest Laboracory. P. O. Box 999, Richland, WA 99352 Prepared for:

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

decommissioning, dose equivalents, radioactivity, buildings, soils, contamination, radionuclide migration, ONSITE/MAXll codes, GENil codes, PilASE8 codes, ISOSHLD codes, DRALIST codes l

79

-~-

1

NUREG/CR-5514-MORECA ORNIIrM-11451

Title:

_ Modeling and Performance of the MHTGR Reactor Cavity Cooling System Descriptioru The Reactor Cavity Cooling System (RCCS) of the Modular High-Temperature Gascoled Reactor (MHTGP) proposed by the U. S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is ava;lable.

A computer dynamic simulation for the physical and mat.ematical modeling of an RCCS is described here. Two condusions can be made from computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and the RCCS panels. These emissivities should be chccked periodioilly to ensure the safety function of the RCCS. Second, the heat transfer trom the reactor vessel is reduced a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant reactor vessel temperature of 500 K. Thus, the presence of a medium that partidpates in the transtnission of radiant energy across the annulus can be expected to result in an increase in the - reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, induding small' partides, in the reactor cavity annulus is warranted.

The goal of the MHTGR safety analysis is to predict the temperatures of the fuel and reactor vessel that may follow postulated reactor transients and accidents.

The thermal performance of the reactor vessel and fuel depends on the ther.nal performance of the RCCS. Therefore, the reactor simulation must be coupled to the RCCS for an overall MHTGR dynamic analysis. The computed temperatures a.:.d flows of the MHTCR reactor vessel, fuel, and RCCS that may occur as a result of postulated transients were computed with MORECA, reactor simulation software, which includes the FCCS -simulation for prediction of overall plant b

transient response.

. Publication Date:

April 1990 Prepared by:

J. C. Conklin Contracion Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 l

Prepared fon NRC Division of Regulatory Applications, Office of Nudear Regulatory Research l _.

- Keywords:

. gas cooled reactors, reactor acddents,' transients, performance - testing, reactor vessels, high temperature, reactor cooling systems, cavities, after. heat semoval i

L I

I l-I 1:

80 1

U e) v,.h,, *e,h (g

o M',,

gf %Y gS A

\\o t,@"'<'V;'f9

((((@,4fg IMAGE EVALUATION

/

TEST TARGET (MT-3)

R}h9 h

,h

~

3 y

+

<a 1.0 ES a u :n un t-n1zs p,

r $ :~ @ 2.0 1.1 i;

=

11'I s=I.8 I!

==

1.25 L l.4 1.6

[

t_

4 150mm 4

6" 4%#r

++ A+

4:u,sy,,%'7///

r/

N,g

p ; M,::, f \\

e

sg Y Y*'

N

+p

?

3:

-,_______.______.m_

_\\_ _

....,aw.

..n,..

p

ch

.c% A

/.O

\\lt' x,_ y &y

_9 6

1 e

IMAGE EVALUATION

!Ib dir q

5 g@

v s \\g/

$8',

TEST TARGET (MT-3)

W

/p

///+

+ < > +

\\

l.0 F i"" f 24 E W pW 22

!! Rt l.\\

l '= R14

\\\\%8

' L25 u

r

~~

"!]a,3 4

150mm 4

6" 4%

f*

tly%> 475,,,%s

_ f,,p 4

////

R ";.

e p1:3 o/

l 4$ qf cQi>

l s,; j Jn _-

w 4

93 ~,p (g

o Y $t h'

[%

[ /

/ ];ff 49 IMAGE EVALUATION g p y @ #*

'k TEST TARGET (MT-3)

' p 4,*

F," p, = 2 f

a 7

l]

l,1

' u, [!!!ll m=e I,

p%

ts l.25

' l.4 I _1.6 g.

4 150mm -

P-l 4

6" 4R+4 e+ zz,o s

4Q,sy. '4)%j7/f,,, -

<S, Q/4

s. u og,,

vy 9

r t

...___-__.-___._i_,,

,4,,,,

I NUREG/CR-5517 IMPACTS-BRC SAND 89-3060

Title:

IMPAC'IS-BRC, Version 2.0 Program User's Manual q

Descriptlun:

This report describes the procedures for implementing IMPACTS BRC Version j

2.0, IMPACTS-BRC2 was designed for use by the Nuclear Regulatory Commission and :ndustry to evaluate petitions to classify specific waste streams as "below regulatory concern (BRC)." The code provides a capability for calculating radiation dose to a maximal Indnidual, critical group, and the general population as a result of transportation, s.2atment, disposal, and post-disposal activities involving low. level radioactive waste. Impacts are calculated for multiple waste streams, nudides, and pathways depending on the treatment / disposal options specified bf the user. Trer: ment and disposal options include onsite incineration, off-site incineration at raunicipal and hazardous waste facilities, and off site disposal at municipal (sanitary) and hazardous waste larvlfills, included within the disposal options is the ability to calculate impacts from the sorting and/or recycling of metal containers 'and metal and glas; L

tr.aterials. Nuclide-specific accounting is provided to facilitate the identification or critical nuclides and pathways that contribute significantly to the radiological l

impacts.

]

The cc.de is wriaen h. FORTRAN 77 and Turbo Pascal and runs on an IBM PC j

or PC-compatible microcomputer with 640 Kbytes e,f random access memory. A

[

nach co-processor and fixed disk drive are recommended but not required.

Publication Date:

W ! 1990 Prepared by:

E L 3'Neal; C. 5. Lee Contractor:

8.Mia r#tional 1.aboratories, P. O. Box 5800, Albuquerque, NM 87185; Applied 1

th sics. Inc.,5353 Wyoming Boulevard NE, Albtquerque, NM 87109 Prepared for:

NPC Divnion of Low l.evel Waste Msnagement and Decommissioning, Office of

,i Mudear Material Safety and Safeguards Keywords:

radketive waste management, radioactive waste disposal, low-level radioactive wastes, gamma radiation, human populations, ingestion, inhalation, risk 6.ssessment, radiation doses, sanitary landfills, IMPACTS coder IMPACTS BRC l

codes, RADTRAN codes s

li i

81 E

NUREG/CR-5521 SWIIT, LHS SAND 90-0127 r

s

Title:

Use of Performance Assessment in Assessing Compliance With the Cor.tainment Requirements in 40 CFR Part 191 Descriptionu nis document summarizes the role of performance assessment in assessing compliance with the containment requirements of 40 CFR Part 191, the Environmental Protection Agency's Standard for the disposal of spent nudear fuel, high level and transuranic radioactive wastes. In 1986, liuriter et al prepared a similar report (NUREC/CR-4510, SAND 860121) which provided an overview of the approach to assess compliance with this standard. The present report builds on its predecessor in that it incorporates subsequent advances in performance assessment. Its main purpose is to serve as a mechanism for transferring to the Nudcar Regulatory Commission (NRC) and its contractors the performance assessment methodologies (PAMs) developed by Sandia National Laboratories (SNL) for high level radioactive waste repositories.

The report starts with a discussion of the requirements in 40 CFR Part 191 and focuse, on the containmerit requirements (Section 191.13). A discussion of the role o performance assessment and its use in regulatory compliance fcilows, The report condudes with a discussion of sources of uncertainty, ticatment of uncern inties, and the cor.struction of the complementary cumulative distribution functiu of summed normalized total releases to the accessible environment for one ct more scenarios. Examples of the demonstration of performance assesstat methodologies for high-level waste disposal at two hypothetical sites-

-a bedad vlt site and a basalt site-are presented.

For the bedded salt formation a two-dimensional representation of the flow system was considered adequate. The SWIFT computer code was use.1 to establish regional flow characteristics such as hydrauhc heads and velocity field.

The results of the SWIFT analyses provided the boundary conditions for a local-scale flow model as well as justification to use a network flow and transport model with one-dimensional segments. A total of 33 input parameters were considered uncerta'n. Twenty-seven of these were common to all 12 of the scenarios selected for analysis. Input sectors (samples) were generated using LHS (latin Hypercube Sampling)

For the alt site, flow and transport tnrough fractured niedia were considered in addi wrous media flow and transport. Seven scenarios were selected egional ground-water flow model was constructed that for an.

s incorpora hydrogeologic features of the site. The uncertainty analysis was restr.c.- a the ' base-case' scenario only. The analysis was performed by generating 70 samples of the 57 uncertain parameters. These were obtained using LHS. For each sample, a regional flow model, a local flow model, and a radionuclide-transport model simulation was performed.

Publication Date:

September 1990 Prepared by:

E. J. Bonano; K. K. Wahl Contractor:

Sandia National Laboratories, P. O. Box 5600, Albuquerque, NM 87185; CRAM, Inc.,170 ivian Street NE, Albuquerque, NM 87112 Prepared fon NRC Division ot High-Level Waste Management, Office of Nudear Material Safety and Safeguards Keywords:

performance testing, containment, US EPA, standards, high-level radioactive wastes, radioactive waste disposal, basalt, salt caverns, regulations, compliana 82 l

l

NUREG/CR-5523 WGEN, UNSAT-H PNL-7356 -

Title:

Development of an-Infiltration Fvaluation-Methodology for Low-Level Waste Shallow Land Burial Sites -

.j

==

Description:==

An infiltration evaluation methodology (IEM) -was developed to provide a consistent, well formulated approach for evaluating ficid-sale infiltration and drainage at low level-waste sites. '02 IEM is designed to simulate significant factors and hydrologk conditions that determine or influence moisture infiltn.run, tedistribution, and drainage in engineered cover and barrier systems.

The IEM recognizes the sources of unwrtainty in estimating moisture movement through engineered covers and cjuantifies their influenas on infiltration and drainage estimates. The IEM is developed on the basis of regulatory l

requirements given in 10 CFR 61. Engineered cover design and data availability will be largely site spedfic. Because of the site-spedfic nature of the design and data availability, the IEM is a flexible framework that accepts various numerical models, dosed-form analytical models, and uncertainty approaches. Two e'

applications of the IEM (a closed-form analytical. model and a series of integrated numerical models) are demonstrated for a hypothetical waste s'tc.;

Both of these approaches are demonstrated by analyzing a hypothetical UMTRA l

barrier. The analytical approach is based on a dosed-form stochastic model of water flow developed by Dagan and' Bresler, representing a simplified analysis of the UMTRA barrier. The integrated numerical ' approach uses a climate generation model (WGEN), a one-dimensional variably saturated finite-difference model (UNSAT H), and a Monte Carlo sampling technique to address uncertainty.

WGEN is a dimate generation code that was -developed to produce daily weather data using historical weather records. WGEN generates a daily weather patte-n that reproduces the statistical frequency of the historical weather record.

The basic Input data requirements include average monthly mean, minimum,-

and maximum temperature; mean monthly precipitation; and daily precipitation values. The output from WGEN was used directly as input into UNSAT H.

- UNSAT H is a one-dimensional model that simulates the dynamic processes of infiltration, drainage, redistribution, surface evapration, and uptake of water from soil by plants. UNSAT H uses a fully-implidt, finite difference method for solving the water transport. equation. Features of UNSAT H that are-improvements over earlier codes, such as UNSAT, indude isothermal vapor Ellow,3 cheatgrass transpiration function, additional options for describing oil hydraulic properties, and reduction of mass-balance error. Since uncertainty exists in the areal spatial distribution of the saturated hydraulic conductivity

- within the day radon barrierc a Monte-Carlo sampling routine was used to Sencrate_ multiple realizations.of ' the saturated hydraulle conductivity. - These realizations were then used as part of the soils data in: the UNSAT H model.

UNSAT-H was run one _ time for ' cach of the 10 realizations generated. The analysis estimated average, minimum, and maximum values of drainage through the engineered barrier over 20 years. -

Publication Date:

- May 1990

. Prepared by:

J. D. Smyth,' G. W. Gee, C. T. Kincaid; E. Bres!ct Contracion Padfic Northwest Laboratory, P. O. Box 999, Richland, WA 99352; Department of Soil Physics,-Institute of Soils and Water, Agricultural Research Organization,-

P. O. Box 6, Bet Dagan, Israel Prepared fon

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

low level radioactive wastes, underground disposal, drainage, Monte ' Carlo method, dimate models, ground water, environmental transport, runoff, plants.

- soils, UNSAT2 codes, CREAMS codes, HELP codes, TRACR3D codes, VAM2D codes g

- NUREC/CR-5527 Humr_n Error Det:brse -

BNL-NUREG-52228

Title:

Risk Sensitivity to 11uman Error in the LaSalle PRA Descriptioru A sensitivity evaluation was conductni to assess the impact of human errors on the internal event risk-parameters in the LaSalle plant. The results provide the-variation in the risk parameters, namely, core melt frequency and accident sequence frequencies, due to hypothetical changes in human error probabilities.

Also provided are insights denved from the results, which highlight important areas for concentration of risk limitation efforts associated with human performance.

The liuman Error Database, containing the 83 human errors in the tasalle plant risk model, was constructed using dBASE 111 Plus on an IBM PC microcomputer.

Each category of the 14 clement categorization scheme was set up as a field with a predetermined size based on the coding descriptors of the categories. Each human error was defined as a record with 14 fields. The database of coded human errors provided the capability for convenient analysis and quick sorting of human errors for sensitivity study applications.

Publication Date:

March 1990

Prepared by:

S. Wong, J. Wiggins, J. O'Ibra, D. Crouch, W. L.uckas Contracton Brookhaven National Laboratory, Upton, NY 11973 Prepared fon NRC Division of Radiation Protection and Emergency Preparedness, Office of Nuclear Reactor Regulation Keywords:

risk assessment, sensitivity analysis, human factors, errors, failures, probabilistic estimation, reactor accidents, reactor cores, meltdown, system failure analysis l'

l' 84

~.

i NUREG/CR-5528 MACCS EGG-2593 i

Title:

An Assessment of BWR Mark !! Containment Challenges, Failure Modes, and Potential Improvements in Performance l

Description:

his report assesses chalknges to BWR Mark 11 containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative nsk/ benefit analysis ot a generic BWR/4 reactor with a Mark Il containment. Point estimate fraquencies of the dominant core damage sequences are obtalnai, and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term relcare categories, which provide input to the consequence analysis. The output of the consequence analysis is used to construct an overall base case risk profile.

Potential improvements and sensitivities are evaluated by modifying the event tree split fractions, thus generating a revised risk profile. Several imnortant sensitivity cases are examined in order to evaluate the impact of phenomcnological uncertainties on the final results.

The severe acddent consequence analysis was performed using a personal computer (PC) version of the MELCOR Accident Consequence Code System (MACCS). MACCS is composed of three modules: ATMOS, EARLY, and CHRONC, which are exercised in sequence by the code. This set of modules was developed for evaluating severe accident consequences at commercial LWR power plants. MACCS 1.5 incorporates several improvements over earlier codes like CRAC2 in the treatment of variable and long term releases, deposition modeling, dosimetry, emergency response, long tenn mitigative actions, radiological health effects, and economic tmpacts.

A number of other computer programs usal for primary cantainment response calculations for unmitigated short term station blackout at Peach Bottom and related assessment studies are mentioned in the document. These include MARCH, BWRLTAS/BWRSAR/MELCOR, MAAP, STCP, CONTAIN, and CORCON.

Publication Date:

July 1990 Prepared by:

D. L Kelly, K. R. Jones, R. J. Dallman, K. C. Wagner Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415 Prepared for:

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

BWR type reactors, reactor accidents, :ontainment, fi:sion product release, failure mode analysis, source *erms, risk assessment, CRAC2 codes, MARCH codes, BWRLTAS cod.s, BWRSAR codes, MELCOR codes, MAAP codes, STCP codes, CONTAIN cod.rs, CORCON codes l

85

NUREG/CR-5530-DOT 4.3, LEPRICON ORNI/rM-11476

. Title l

.' Analysis of H. B.i Robinson PWR Vessel Fluence for Cycle 10 Utilizing Partial 12ngth Shield Assemblies -

==

Description:==

- Neutron transport calculations were performed to determine the pressure vessel fluence and cavity dosimeter responses for cyde 10 of the H. B. Robinson pressurized water reactor. This cycle was the first to utilire " partial length shield assemblies" within the core to reduce the fluence rate at the critimi weld location in the vessel. This work is part of the ongoing surveillance of the Robinson plant to insure that the projected fluence rates are reliable.

1 The flux calculations utilize a "two-channel" synthesis approximation and recently processed iron cross r,ections based on a new evaluation for the inelastic data above 3 MeV. The methodology used to calculate this highly asymmetrical configuration is ' discussed in detail, and a comparison of the calculated and measured cavity-dosimetry results is presented. Discrepancies are observed in the computed and tr.casured results for the 2"Np dosimeter, and possible explanations are discussed. Calculated absolute neutron flux spectra, as well as radial, aximuthal, and axial variations in the fast flux and dpa within the pressure vessel, are given.

Channel fluxes were computed with the DOT 4.3 discrete-ordinates transport code, which required two r-theta, r z, and r runs, respectively. An S. angular quadrature - and - a P scatter cross-section - expansion were used - in the 3

calculations, and the energy variation was repretented by 47 neutron and 20 gamma groups.- The multigroup cross-section data were mostly obtained from-the SAILOR library.-

The effect of a least-squares consolidation of the measured and calculated results is - studied." This consolidation of the measured - dosimeter activities. and the transport calculations'was performed using the LEPRICON adjustment code. The present version or L"."RICON routinely considers only one cavity location at c

-l time. This means that even though measurements were made at four different.

1 azimuthal locations, a simultaneous adjustment was not - performed. Because some questions still remain conceming the appropriate adjustment procedure for this case, this report has only emphasized the reference (i.e., unadjusted) transport calculations.

. Publication Date:

September 1990

' Prepared by:

M. L Williams, M. Asgari; R. L. Childs Contractoc Louisiana State University Nudcar Science Center;. Oak Ridge National Laboratory, P;O. Box 2008, Oak Ridge, rN 37831 Prepared fon NRC Division of Engineering, Office of Nudcar Regulatory Research Keywards: -

surveillance, neutron fluence, dosimetry, reactor-vessels, radhtion effects,-

Robinson 2 reactor 86

L NUREG/CR-SM2 UARRIER EGG-2597

Title:

• dels for Estimation of Service Life of Concrete Barriers in Low-lxvel Radioactive Waste Disposal Desedption:

Concrete barriers will be used as intimate parts of systems for isolation of low-level radioactive wastes subsequent to dispor.al. nis work reviews mathematical models for eshmating the degradation rate of concrete in typical service environments. Ec models considered cover sulfate attack, reinforcement corrosion, calcium hydroxide leaching, carbonation, freeze / thaw, and cracking.

Additionally, fluid flow, mass transport, and geochtmical properties of concrete are briefly reviewed. Eumple calculations included illustrate the types of predictions expected of the models.

BARRIER is a computer program for estimating long. term degradation of concrete barriers. In the code, the coupled attack of all processes working simultaneously is simulated with annual changes in the properties of each concrete layer. Processes such as sulfate attack and freeze / thaw are as-

' I to lead to annual reductions in the effectim thickness and transport prop % of the concrete slab.

BARRIER models freeze / thaw durability by a fit of data that exprese the fractional decrease in dynamic moduks of clasticity (DME) of the concrete as a curve fit involving the percent of entrained air (AIR), water-to-cement ratio (WCR), and the number of freeze / thaw cycles. The variation of DME is taken to be linear when the number of freeze / thaw cycles is greater than 50 based on experimental results The code also computes the amount of time nyu red to reach a given level of damage for a given value of fractional decrease in DME as well as an annual rate of concrete loss.

BARRIER also models leaching, cr cking, end reinfoacment corrosion / chloride attack. Cracks from physical loading are aswmed initially to penetrate only to the neutral axis and not af fect the byJrologic properties ut the stab. Once the cracks reach a depth of thrw-fourths of th<s s!ao thickness or within 7.5 cm of the total slab thickness, they are assumed to fully penetrate the slab. Chloride attack is modeled as a two-stage process: (a) time to breakup of the passive layer and initiation of corrosion and (b) corrosion rate subsequent to breakup of the passive layer. The Clear empirical model is tacd for time to depassivation while the subsequent corrosion rate assuming limitation by oxyra diffusion is estimated by a one-dimensional steady-state oxygen diffusion equation with a zero concentration boundary condition at the metal surface.

Publication Date:

September 1990 Prepared by:

J. C. Walton, L E. Plansky, R. W. Smith Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 81115 Prepared for:

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

radioactive waste disposal, concretes, low level radioactive wastes, mathematical models, teaching, decomposition, cracking, corrosion 87

NUREG/CR-5M5 VICTORIA SAND 90-0756

Title:

VICTORIA: A Mechanistic Model of Radionuclide Behavior in the Reactor Coolant System Under Severe Accident Conditions Descriptioru This document provides a description of a model of the radionudide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident and serves as the user's manual for the computer code VICTORIA, based upon that model. VICTORIA predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating.

The report also contains a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

VICTORIA was developed on a CRAY X-MP at Sandia National Laboratory and a CRAY 2 and various SUN workstations at the Winfrith Technology Centre in England.

Publication Date:

Octoter 1990 Prepared by:

T. J. Heames; N. E. Bixler, A. J. Grimley; C. J. Wheatley; D. A. Williams, N. A.

Johns, N. At Chown Contractor:

Scence Applications International Corporation, 2109 Air Park Road, SE, Albuquerque, NM 87106; Sandia National Laboratory, P.

O.

Box 5800, Albuquerque, NM E7185; U. K. Atomic Energy Authority, Safety and Reliability Directorate, Wigshaw Lane, Culcheth, Warrington, Cheshire WA3 4NE, United Ki sgdom; U.

K.

Atomic Energy Authority, Winfrith Technology Centre, Dt rehester, Dorset DT2 SDH, United Kingdom Prepared fon NRC Division of Systems Research, Office of Nudear Regulatory Research Keywords:

reactor accidents, reactor cooling systems, water cooled reactors, water moderated reactors, fission products, fission product release, chemical reactions, radioactive acrosols, radioisotopes, building materials, after-heat I

88

NUREG/CR-5547 FITEQL PNL-7239

Title:

Application of Surface Complexation Models for Radionuclide Adsorption Sensitivity Analysis of Model input Parameters Descriptioru This report discusses activity in two areas: 1) an evaluation of methodologies currently used for the physical and chemical characterization of metal oxide and hydroxide adsorbents and 2) the sensitivity of the various surface complexation models' adsorbent input parameters for describing adsorption. The report describes the relative merits of three surface complexation models (SCMs),

procedures to estimate values or the model parameters from titration data. and what is required of experimental titration data sets. The ultimate goal is to determine how and whether SCMS can successfully desenbe adsorption of contaminants from disposed nuclear wastes in natural soils and sediment.

This study's results help clarify the applicability of SCMs, particularly with respect to their sensitivity to input parameters. A method is presented by which unique, best-fit valuca for these parameters may be obtained. An appendix is provided that reviews methods for determining surface area, site density, particle size distribution, and pore structure.

The three SCMs chosen for evaluating the sensitivity of model simulations to changes in parameter values are: 1) the diffuse-layer model (DLM), 2) the constant-capacitance model (CCM), and 3) the triple-layer model (TLM). In modeling titration data an im rtant attribute of a given model is its ability to account for the effects of nging background electrolyte concentration on surface protolysis reactions. Differences in accounting for electrolyte behavior distinguish each model and determine the relative ability of a given model to fit the titration data and therefore our ability to extract SCM constants for a given model from titration data.

SCM parameter values can be determined from titration data using a nonlinear least squares fitting program, FITEQL This program optimizes the values of adjustable parameters by changing their values until the sum of the squares of the residuals between the measured titration data and FITEQL calculated values is minimized. Because FITEQL will not routinely converge if more than two parameter values are declared as adjustabic, it is necessary to assume values for all but two of the SCM parameters and then let FITEQL optimize the two parameters which are chosen as the adjustable ones.

Publication Date:

April 1990 Prepared by:

K. F. Hayes, C. Redden, W. Ela, J. O. Lecxie Contracton Department of Civil Engineering, Stanford University, Stanford, CA 9430S Prepared f'or:

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

low-level radioactive wastes, adsorption, sediment water interfaces, radionuclide m ation, titration, sensitivity analysis, radioactive waste disposal, MINEQL 1

89 l

NUREG/CR-5548 G:ncral PNL-7285

Title:

Review of Geochemical Processes and Codes for Assessment of Radionuclide Migration Potential at Commercial LLW Sites

==

Description:==

Information on geochemical processes that control contaminant solution concentrators and migration at existing low-level radioactive waste (LLW) sites is summarized. The review identifies the current status and future information needs required for the development of effective performance assessment models for use in site license applications. Except for some reports on LLW disposal sites at Sheffield, Illinois and West Valley, New York, few references were identified that contained adequate geochemical data to model geochemical processes Giat affect migration. Tritium appears to be the most mobile radionuclide migrating from burial trenches at commercial LLW sites.

The review identified microbial-degradation induced anoxia, subsequent iron oxide precipitation during oxidation, alkalinity controlled pH changes, and organic complexation reactions as key controls of radionuclide mi6. tion. The quantity of experimental and field data against which to test geochemscal codes is very limited. All experimental work on radionuclide adsorption at commercial LLW sites relles upon the 14 concept. Studies of the effects of organics on radionuclide mobility suggest that a %EI7FA chelate formed in the original waste may persist indefinitely and lead to enhanced migration. Other organic-radionuclide complexes are less stable and do not significantly enhance mobility under conditions antidpated at commercial LLW sites.

Calculations of spedes distnbutions of contaminants are necessary to predict migration pa*e 11 in groundwater systems. Further, speciation can only be reliably c.letu d from a combination of accurate chemical analyses and use of chemical reaction rodes. Thus, the strengths and weaknesses of various reaction codes in describing chemical processes and the adequacy of available thermodynamic data are considered.

Brief descriptions and references to publications of.some of the more extensive reviews of geochemical reaction codes are given. Codes included are WATEQ, REDEQL, GEOCHEM, MINEQL, MINTEQ, PATHCALC, PHREEQE, EQ3/EQ6, CHEMIST, EQUILIB, FASTPATH, SIAS, and SOLMNEQ.

Publication Date:

April 1990 Prepared by:

R. J. Serne, R. C. Arthur, K. M. Krupka Contracton Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352 Prepared fon NRC Division of Low-Level Waste Management and Decommissioning, Office of Nuclear Material Safety and Safeguards Keywords:

geochemistry, chemical reactions, radioisotopes, radionuclide migration, low-level radioactive wastes, field tests, WATEQ codes, REDEQL codes, GEOCHEM codes, MINEQL codes, MINTEQ codes, PATHCALC codes, PHREEQE codes, EQ3/EQ6 codes, CHEMIST codes, EQUILIB codes, FASTPATH codes, SIAS codes, SOLMNEQ codes

l NUREG/CR-5553 G:n:ral ORNI/TM-11505

Title:

Computer Programs for Eddy-Current Defect Studies

==

Description:==

This report desenbes several computer programs developed to aid in the design of eddy-current tests and probes. The programs, written in FORTRAN, deal in various ways with the response to defects exhibited by four types of probes: the pancake probe, the reflection probe, the circumferential boreside probe, and the circumferential encirding probe. Programs are included which calculate the impedance or voltage change in a coil due to a defect, which calculate and plot the defect sensitivity factor of a coil, and which invert calculated or experimental readings to obtain the size of a defect. The theory upon which the programs are based is the Bunows point defect theory, and thus the results of the calculations will be more accurate for small defects.

The report is divided into five sections, one for each of the four coil types and a fifth for common subroutines. Pancake coil programs described and listed include:

PCBLDF builds a magnitude and phase lookup file PCDSF calculates magnitude and phase of defect sensitivity factor (DSF) for lattice of points PCDSFPLT generates a contour plot of magnitude of DSF PCFIX converts raw data to normalized impedance change PCAVZSCN calculates defect impedance change, average over depth PCAVVSCN calculates defect impedance change, average over volume PCCRAPH plots two sets of data on same graph PCINV inverts scan of pancake coil data to get depth and volume PCRISCAN converts raw voltage readings to impatance change Reflection coil programs described and listed indude:

RFBLDF builds a lookup file of magnitude and phase of the integral of the D'iF RFDSF calculates magnitude and phase of DSF for lattice of points RFDSFPLT generates a contour plot of magnitude of DSF RFAVZSCN calculates defect voltage change, average over depth RFCRAPH plots two sets of data on same graph RFINV inverts scan of defect using reflection coil data Circumferential boreside coil programs described and listcd include:

ABBORAR calculates defect impedance change for absolute coil Differential coil programs described and listed indude-DBDSF calculates DSF of a differential boreside coil at a lattice of points DBDSFPLT gene ates a contour plot fro n DBDSF data DFBORAR calculates defect impedance change, average over depth Circumferential encircling coil rograms described and listed indude:

ABENCAR calculates im nce change for absolute coil DFENCAR calculates de impedance change, average over depth These programs are written tr* run in Ryan-McFarland 10RTRAN on IBM PC/AT-compatible microcom; as, using either an Intel 80286 or 80386 microprocessor Grafmatic plo. ing software from Microcompatibles is used for the contour plots and must be installed on the machines.

Publication Date:

June 1990 Prepared by:

J. R. Pate, C. V. Dodd Contracton Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared fon NRC Division of Engineering, Office of Nudcar Regulatory Research Keywords:

eddy currents, point defects, probes, electric coils, impedance, sensitivity

{

91

NUREG/CR-5557 RELAP5 EGG-2599

Title:

REL@S Thermal Hydraulle Analysis of the SNUPPS Pressurized Water Reactor

==

Description:==

Thermal-hydraulle analyses of five hypothetical accident scenarios were performed with the RELAPS computer code for the Westinghouse Standardized Nudear Unit Power Plant System (SNUPPS) pressurized water reactor. This work was sponsored by the U. S. Nudcar Regulatory Commission and is being done in conjunction with future analysis work at the U. S. Nudear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. nese acrident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Technical Training Center Standardized Nuclear Unit Power Plant System simulator to mcxici abnormal transient conditions.

De model, a four loop pressurized water reactor (PWR), contained detailed thermal-hydraulic representations of the pertinent PWR grimary and secondary systems, including the feedwater train and steam lines. Detailed models of the key plant control systems were also induded.

He RELAPS model was used to analyze five separate transients, selected to cover a wide range of possible thermal hydraulic conditions that could occar in a reactor accident. De transients were: (a) loss of AC power, (b) small break loss-of-coolant accident with loss of AC power, (c) failed open pressurizer safety valve, (d) main steam line break with a steam generator tube rupture, and (e) loss of feedwater without scram.

In general, the calculated RELAPS trends were reasonable for the scenarios studied and will provide a good basis for :omparison with simulator data. Some uncertainties la boundary conditions and modeling options have not been resolved and could affect simulator /RELAPS comparisons.

Publication Date:

May 1990 Prepared by:

C. M. Kullberg Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415 Prepared for:

NRC Division of Systems Research, Office of Nudear Regulatory Research Keywordr reactor acddents, PWR type reactors, reactor cooling systems, loss of coolant, AC losses, pressurizers, steam generators, transients 92

NUREG/CR-5572 SLIM-MAUD UNL-NUREG-52236

Title:

An Evaluation of the Effects of Local Control Station Design Configuratior.. on Human Performance and Nuclear Power Plant Risk

==

Description:==

A human factors analysis was performed to assess how identified upgrades to local control stations (LCSS) in nudear power plants affect both human performance and plant risk. Upg.ades in the design of individual control panels and overall improvement of functional centralization were consideral. The analysis methodology was accomplished in four stages. First, a list of LCS human engineering design deficiencies was developed using data collected from a variety of sources including visits to nuclear power plants. From these data, a set of potential upgrades was defined to currect the deficiencies. Second, the effects of the upgrades on human error probabilities (HEPs) were determined using a computer-based methodology (SLIM-MAUD) for soliciting expert judgment. Third, the HEPs were propagated through a plant probabilistic risk assessment (PRA), and new core melt frequencies were established. A prelimbary, scoping value-impact assessment was performed to evaluate the regulat3ry nwd for further review of possible action to improve the human factors engineering aspects of local control stations. The results indicated that implementation of both types of upgrades would improve human performance and lower risk, but that only the panel design improvements would be cost beneficial.

SLIM-MAUD (Success Likelihond index Method / Multi. Attribute Utility Decomposition) utillies a consensus approach to discriminate among LCS configurations along several performance shaping factors (PSFs). In this study, three judges evaluated the panels along PSF dimensions of communications load, control panel configuration, training burden, and procedural complexity. The PSFs were weighted for relative importance by the judges. Weighted dimension ratings wue combined to produce a Success Likelihood Index (SLI) for each panel configuration. SLI: were converted to HEPs, and since nine panel configurations were evaluated, a total of nine unique sets of HEPs were determined. Each set of HEPs was then entered into the ppa, and th4. core melt frequency (CMF) was calculated. A total of nine runs of the pRA were completed to identify the variation m plant risk associated with LCS configuration vartation.

Publication Date:

Ceptember 1990 Prepared by:

J. O'Hara, C. Ruger, J. Higgins, W. Luckas. D. Crouch Contracton Brookhaven National Laboratory, Upton, NY 11973 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

human factors, nuclear power plants, man-machine systems, control rooms, display devices, reactor operators, performance, failures, errors, behavior, human factors engineering, reactor safety, risk assessment, prob 6ilistic estimation, expert systems 93

NUREG/CR-5573 RELAP5/ MOD 2 BNL-NUREG-52237

Title:

Boron Rushing During a BWR Anticipated Transient Without Scram Descriptforu This report documents a study of an accident sequence in a boiling water reactor (BWR) in which there is a large reactivity insertion due to the flushing of borated water from the core. T1us has the potential to occur dunng an anticipated transient without scram (ATWS) after the injection of borated water from the standby liquid control system. The boron shuts down the power, but if there is a rapid depressutization of the vessel (e g., due to the inadvertent actuation of the automatic depressurization system), large amounts of low pressure, relatively cold, unborated water enters the ussel causing a rapid dilution and cooling. This study was carried out to determine if the reactivity addition caused by this flushing could lead to a power excursion sufficient to cause catastrophic fuel damage. Calculations were carried out using the RELAPS/ MOD 2 computer code under different assumptions regarding timing and availability of low pressure pumps and with different reactivity coefficients.

The results shownf that the fuel enthalpy tise was insufficient to cause catastrophic fuel damage although less severe fuel damage might still be possible due to the overheating of the fuel claddin5-RELAPS/ MOD 2 was the computer code used for the calculations of the ATWS scenario. Thb code was chosen because of its well<stablished validity and for its capability at low conditions. Of the various especially modifications to the model made at BNL,pressurethe most expensive were the addition of a Mark I containment and a hot channel.

Publication Date:

June 1990 Prepared by:

D. Mirkovic, D. J. Diamond Contractor:

Brookhaven National laboratory, Upton, NY 11973 Prepared for:

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

BWR type reactors, excursions, reactor accidents, computerized simulation, depressunzation, reactivity, coolants, reactor vessels, pressure

vessels, thermodynamics, reactor core disruption, probabilistic stimation, boron, ATWS 94 l

l

NUREG/CR-5575 Genral EGG-2602

Title:

Quantitative Analysis of Potential Performance improvements for the Dry PWR Containment

==

Description:==

nis report calculates the risk benefit associated with potential performance improvements for the large dry pressurized water reac.or (PWR) containment.

The analysis is based on the June 1989 draft NUREC.1150 resalts for the Zion commercial nuclear reactor. Simplified containment event trees and the large accident progression event trees from draft NUREG-1150 are used to evaluate the effects of potential improvements on the response of the Zion containment to dominant severe acddent sequences. Source terms are generated parametrically using the ZISOR code and offsite consequences are calculated with the MELCOR Accident Consequence Code System (MACCS). These results give point estimates of the risk reduction associated with each containment improvement identified by Brookhaven National laboratory in their draft Issues Characterization Report.

The methodology used for the analyses in this report on be descibed briefly as follows. Simplified contairtment event trees (SCETs), each consisting of 10-15 top events, were developed from the large 72-question accident progression event trees (APETs) used to analyze the Zion containment response for draft NUREG-1150. He EVNTRE event progression analysis code provided the essential tool for developing SCETs from the large APETs. De SAIC code ET LOAD was used to read the sorted output file from EVNTRE and construct the SCET u ing the SAIC ETA-Il event tree code. Branching structure and split fractions are determined automatically by ET-LOAD. A base case SCET was constructed for each of the dominant Zion pt nt damage states: (1) loss-of coolant accidents, (2) station blackout, 0) transients (including anticipated transients without scram), and (4) containment bypass. To reduce the number of source term calculations, the end states of the SCETs were grouped into accident progression bins in accordance with the binning scheme for Zion in draft NUREC/CR-4551.

The ZISOR code was then used to calculate important characteristics of the containment release for each accident progression bin produced Examples of these characteristics are the time and duranon of the release, the release fracdons of the various nudide groups, and the energy release. In all of the cases in this report, point estimates of the source terms were obtained using ZISOR. The source terms generated by ZISOR were input to the MELCOR Accident Consequence Code System (MACCSL along with the site data from the Zion draft NUREG-1150 MACCS deck, to compute ofialte consequences for each set, generally for each accident progression bm. These conditional consequences are the last input needed to calculate risk. Five risk measures are reported: (1) the mean number of early (acute) fatalities per reactor-year of operation, (2) the mean munber of latent cancer fatahties per reactor-year of operation, 0) the mean dose (in person-rem per reactor-year) within 50 miles of the plant, (4) the mean dose (in person-rem per reactor-year) over the entire 1000-mile MACCS calculational grid, and (5) the mean offsite costs (5 per reactor-year).

Publication Date:

August 1990 Prepared by:

D. L Kelly, D, J. Pafford, J. A. Schroede, K. R. Jones Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415 Prepared for:

NRC Divisicn of Safety Issue Resolution, Office of Nuclear 'legulatory Research Keywords:

PWR type reactors, contain*nent, performance, failure modo analysis, risk assessment, depressurization, reacter a cidants, MACCS codes, ZISOR codes.

EVNTRE codes, ET-LOAD codes, ETA-II codes i

95

NUREG/CR-5584 CA-TEST ORNIRM-11575

Title:

Results of Crack Arrest Tests on Two Irradiated IUgh-Copper Welds

==

Description:==

he objeesve of this study was to determine the effect of neutrcn irradiation on the shift and shape of the lower-bound curve to crack arrest - data. Two submerged are welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plat.:. Crack-arrest specimens fabricated from these welds were irradlated at a nominal temperature of 288' C to an average fluence of 1.9 x 10" neutrons /cm (greater than 1 MeV). This report 8

com es the results of crack-arrest tests on 36 irradiated weld embrittled type ens with those from unirradiated cc.. trol specimens. Since this is only the s

- first phase of a two-phase program, the mnclusions presented are preliminary.

Evaluation of the results shows that the neutron-irradiation-induced crack-arrest toughness temperature shift is about the same as the Charpy V notch impact temperature shift at the 41-) energy level, ne shape of the lower-bound curves (for the range of test temperatures covered) did not seem to have been altered by irradiation compared to that of the ASME K, curve.

The Appendices document for archival and quality assurance purposes various aspects of the crack-arrest data. Appendix A traces the flow and processieg of data. It contains a sample of the data sheets on which the measured dimensions of each crack arrest spedmen are recordai and gives detailed specimen dimensions and results. The BASIC Hewlett Packard Series 200/300 computer code CA. TEST used to process the test data is listed in Appendix B, and the Young's moduli used are discussed in Appendix C. Typical CA.-TEST output for weld embrittled and duplex-type crack-arrest specimens is shown in Appendix D. He load vs. crack mouth cpening displacement charts obtained durin6 the test and a photograph of the fracture surface for each irradiated specimen are reproduced in Appendix E. All data (including that from the Hewlett Packard Series 200/300 computers) are maintained in the relational database system PARADOX on an IBM compattble computer.

Publication Date:

December 1990 Prepared by:

S. K. Iskander, W. R. Corwin, R. K Nanstead CoMracion Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared fon NRC Division of Engineering, Office of Nudear Regulatory Research Keywords:

welded joints, cracks, notches, copper, neutron fluence, irradiation, pressure vessels, transition temperature, submerge? are welding, water cooled reactors, water moderated reactors, testing, fracture properties, Basic 96

NOREG/CR-5586 CONTAIN, CONTAIN-DCII SAND 90-1102

Title:

Mitigation of Direct Containment Heating and Hydrogen Combustion Events in Ice Condenser Plants Analyses with the CONTAIN Code and NUREG-1150 PRA Methodology Descriptioru Using Squoyah as a representative plant, calculatic,ns were performed with a developmental version of the CONTAIN computer mde to assess the effecti eness of various possible improvements to ice condenwr containments in mitigating severe accident scenarios involving direct containment heating (DCH) and/or hydtcigen combustion. Mitigation strategies considered included backup power for igniters and/or air return fans, augmented igniter systems, containment venting, :ontainment inerting, subatmospheric containment

\\peration, reduced im condenser bypass, and primary system depressurization.

o arious combinations of these improvements were also considered. Only inerting the containment or primary system depressurization combined with backup power supplies for the igniter systems resulted in large decreases in the peak pressures calculated to result from DCH events. Potential hydrogen detonation threats were also assessed; providing backup power for both the igniter systems and the air return fans would significantly reduce the potential for detonation but might not totally climinate it. Sensitivity studies using the NUREG-1150 probabihstic risk assessment (FRA) methodology indicated that primary system depressurization combined with backup power for both igniters and fans could reduce the contribution to the mean risk potential of the class of events considered by about a factor of three.

The CONTAIN-DCH code, a developmental version of the CONTAIN 1.1 code which includes a combination of mechanistic and parametric models for DCH phenomena, was used to assess threats in the unmodified plant and to determine the effectiveness of possible improvements. Calculations were performed using 4<cll,6-cell, and 26-cell reprewntations of the containment, in the CONTAIN default model for hydrogen combustion (which is essentially the same as the default burn model in the HECTR code), hydrogen combustion is assumed to occur whenever hydrogen and oxygen mole fractions exceed 7%

and 5%, respectively, provided that the steam mole fraction is less than 55%.

The code does not include models for analyzing the in-vessel phase of the accident progression. Hence, sources of steam, hydrogen, acrosols, radionuclides, etc. entenng the containment from the primary system prior to vessel breach must be specified. Appropriate calculations for Sequoyah were nc.t availabic, and the primary system sources therefore were based upon Surry plant calculations.

For the fullppressurized case, a MARCH analysis was used, while SCDAP/RELAP analyses were employed for the intentionally depressurized case.

Publication Date:

October 1990 Prepared by:

D. C. Williams, J. J. Cregory Contractoc Sandia National Laboratories, P O. Box 5800, Albuquerque, NM 87185 Prepared fon NRC Division of Safety issue Resolution, Office of Nuclear Regulatory Research Keywords:

containment systems, ice condensers, performance, hydrogen, reactor accidents, mitigation, combustion, detonations, probabilistic estimation, risk assessment, sensitivity analysis, SCDAP/RELAP codes, MARCH codes 97

NUREG/CR-5588, Vcl 1 CARES BNL-NUREG-52241, Vol 1 l

Title:

- CARES (Computer Analysis for Rapid Evaluation of Structures) Version 1.0 J

Seismic Module Theoretim1 Manual

==

Description:==

Dudng fiscal years 1988 and 1989, Brookhaven National laboratory (BNL) developed the CARES system (Computer Analysis for Rapid Evaluation of Structures) for the U. S. Nudcar Regulatory Commission iNRC). CARES is a personal computer software system designed to perform structural response compu:ations similar to those encountered in licensing reviews of nuclear power

- plant structures. The documentation of the Seismic Module of CARES.conslsts of three volumes. This - report, Volume 1 of the three-volume Seismic Module documentation, concentrates on the theoretical basis of the system and presents modeling. assumptions and limitations as well as solution schemes and algorithms. The User's - Manual is published 'as Volume 2 of the CARES docunentation, while solutions and results from a set of sample problems are -

published as Volume 3.

The basic modules of the CARES system are associated with capabilities for static, seismic, and nonlinear analysis. The process of seismic analysis of nudear power plant structures and components generally involves definition of-the design seismic criteria at a given site, evaluation of the free-field motion, and evaluation of the structural response and = floor response spectra induding soil-structurc interaction. The scismic Module of CAlu.s was designed with specific capabilities needed to carry out the three-step process as outlined.

The d' evelopment of-the Seismic Module of CARES is based on the SIM and SLAVE computer codes developed at the City University of New York. SIM calculata the response of a structure to a given earthquake input through soil-structure interaction, while the SLAVE code computes a consistent set of

- time histories throughout a soil profile for a given motion at one loation (convdution/ deconvolution analysis). The SIM and SLAVE codes were modified to treat problems related to thi seismic evaluation of nuclear power plants and to indude the capability of generating spectrum-consistent acceleration time histories and some Power Spectral Density (PSD) operations in developing the Seismic Module of CARES.

CARES is designed for use on IBM PC and compatible microcomputers. The minimum hardware configuration is a 10 Mbyte hard disk drive, 720 Kbyte 3-1/2 inch flexible disk drive, 640 Kbytes random access memory, Hercules or EGA graphics adapter board, graphics monitor, and a math co-processor. CARES requires the DOS operating system version 3.1 or later.

1 Publication Date:-. ' July 1990 Prepared by:

J, Xu, A. J. Philippacopoulas, C. A. Miller, C. J. Costantino Contractor:

Brookhaven National Laboratory, Upton, NY 11973 l

Prepared for:

NRC Division of Engineering, Office of Nudear Regulatory Research l-

.-Keywords:.

seismic effects, site characterization, soil-structure interactions, nudear power L

plants, ground motion, mechanical structures, buildings, SIM codes, SLAVE codes 93

NUREG/CR-5588, VCL 2 CARES UNL-NUREG-52241, Vol 2

Title:

CARES (Computer Analysis for Rapid Evaluation of Structures) Version 1.0 Seismic Module Users Manual

==

Description:==

During fiscal years 1968 and 1%9, Brookhaven National laboratory (BNL) developed the CARES system (Computer Analvsis for Ragd Evaluation of Structures) for the U. S. Nuclear Regulatcry Co'mmission (NRC). CARES is a personal computer software ystem designed to perform structura_1 response computations similar to those encountered in licensing reviews of nuclear power plant structures he documentation of die Seismic Module of CARES consists of three volumes. This report, Volume 2 of the three-volume Seismic Module documentation, is the Ur. cts Manual aiid presents hardware and software reqoirements and details for the preparation of input data for use of the CARES system. The theoretical basis for the system is discussed in Volume 1, while results and printouts from a set of seis'nic problems inducEng soil-stmeture interaction effects are published as Volume 3.

We hasic modules of the CARES system are associated with cap 6ilitics for static, scismic, and nonlinear analysis. The process of seismic analysis of nudear power plant structures and components generally involves definition of the design scismic cnteria at a given site, evaluation of the free-field motion, and evaluation of the structural response and floor responsa spectra including soil-structure interaction. The seismic Module of CARES was designed with specific capabilities needed to carry out the three-step process as outlined.

The CARES system contains seven options with each of the options having one or more suboptRns. The options used for any particular problem depend on the type of problem to be run. The user must first decide which of the opdons are to be exercised and then provide the appropriate input data. Tbc main computational aspects of the Seismic Module are organized by the General Manager (Option 1). Options 6 and 7 deal with computational aspects asoociated with the definitio of the scismic input motion and the generation of Pcwcr Spectral Density funcuons. Option 2 provides the ttser with a capability to perform convolution / deconvolution analysis at a given uniform or layered site and computes strain compatible soil properties. The output consists of the free-field motion at the foundation level as well as at various depths below the grcund surface. Options 3, 4, and S perform seismic respoase computations by taking into accosmt soil ntructure effects. Option S provides the user with seismic floor response spectra at specified locations of a nudear plant.

CARES is designed for use on IBM PC and compatible microcomputers. The minimum hardware configuration is a 10 Mbyte hard disk drive, 720 Kbyte 31/2 inch flexible disk dnve, 640 Kbytes random access memory, Hercules or EGA graphics adapter board, graphics monitor, and a math co-processor. CARES requires the DOS operatmg system version 3.1 or later.

Publication Date:

July 1990 Prepared by:

J. Xu, A. J. Philippacopoulas. C. A. Miller, C. J. Costantino Cor tracton Brookhaven National Laboratory, Upton, NY 11973 Prepared fon NRC Division of Engineermg, Office of Nudear Regulatory Research Keywurds:

seismic effects, site characterization, soil-structure interactior, nudear power plants, ground motion, Tiechanical structures, buildings, SIM codes, SLAVE codes 99

i NUnEG/CR-5588, Vol. 3 CARES BNL-NUREG-52241, Vol. 3

Title:

CARES (Computer Analysis for Rapid Evaluation of Structures) Version L0 Seismic Module Sample Problems i

==

Description:==

During fiscal years 1988 and 1989, Brookhaven National Laboratory (BNL) developed the CARES system (Computer Analysis for Rapid Evaluation of Structures) for the U. S. Nuclear Regulatory Commission (NRC). CARES is a personal computer software system designed to perform structural response computations similar to those encountered in licensing reviews of nuclear power plant structures. The documentation of the Seismic Module of CARES consists of three volumes. This report. Volume 3 of the three-volume Seismic Module documentation, presents three sample problems typically encountered in Soil-Structure Interaction analye. The theoretical basis, modeling assumptions and limitations, as well as solution schemes and algorithms of the Seismic Module of CARES are given in Volume 1. The User's Manual is published as Volume 3. The software is made availabic to public us. cts through the National Energy Software Center (Argonne, Illinois).

De basic modules of the CARES system are associated with capabilities for static, seismic, and nonlinear analysis. De process of scismic analysis of nudear power plant structures and components generally involves definition of the design seismic criteda at a given t.ite, evaluation of the free. field motion, and evaluation of the structural response and floor response spectra including soil-structure interaction. The Seismic Module of CARES was designed with specific capabilities needed to carry out the thrco-step process as outlined.

The three sample problems presented are representative of those encountered in the seismic analysis of commercial nuclear plants. A typical plant containment embedded in a layered soil profile is chosen for the analysis. The first sample

. problem utilizes Option 6 to simulate a ground acceleration time history compatible to Reg. Guide 1.60 spectra. The second sample problem deals with seismic waves propagating through a layered soil medium. This sample problem is organize ( so that a complete deconvolution analysis is demonstrated in detail.

The third umple p oblem is run to perform soil-structure interaction analysis and develop floor response spectra in a plant containment building.

CARES is designed for use on IBM PC and cnmptible microcomputers. The minimum hardware configuration is a 10 Mbyte nard disk drive, 720 Kbyte 3-1/2 inch flexible disk drive, 640 Kbytes random access memory, Hevcules or ECA graphics adapter board, graphics monitor, and a math co-processor. CARES requires the DOS operating system version 3.1 or later.

Publication Date:

July 1990 Prepared by:

J. Xu, A. J. Philippacopoulas, C. A. Miller, C. J. Costantino Contractor:

Brookhaven National laboratory, Upton, NY 11973 Prepared for:

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

seismic effects, site characterization, soil-structure interactions, nuclear power plants, ground motion, mechanical structures, buildings, SIM codes, SLAVE codes 100

NUREG/CR-5590 IIECTR SAND 90-7080

Title:

Assessment of the Combustion Model in the HECTR Code Descriptforu HECTR (Hydrogen Event:

Containment ' Transient Response) is a

lumpM-parameter containment analysis code developed to model the containment atmosphere during a nuclear reactor accident involving the release, transport, and combustion of hydrogen. A new set of flame s wd and combustion completeness correlations has been included in HE The combustion model in HECTR was assessed against a simple two-compartment problem, as well as the NTS (Nevada Test Site) and VGES (Variable Geometry Experimental System) experiments.

The example using the two-compartment problem demanstrates that the combustion model in the modified HECTR code is capable of locating the flame position in a compartment with a hydrogen burn and convecting material with the proper composition instead of a mixture gar through now junctions. HECTR predictions compare reasonably well with the ricasured peak pressure ratios for 12 NTS premixed hydrogen experiments It is concluded that the capability of the new correlations for flame speed and combustion completeness is sulficient for the NTS experiments. Based on the an.,1yses of the NTS and VCES experiments, use of a single roorn as a node is recommended for the nodalization in the HECTR combustion calculation.

HECTR is used to guide and interpret experiments, provide combustion models for other accident analysis codes such as CONTAIN and MELCOR, and to analyze nudear reactor accidents involving transport and combustion of gases within a containment building.

Several modifications made to HECTR after the one-dimensional flame propagation model of Celler and Wong was incorporatal are discussed, inclusion of the new flame speed and combustion completeness correlations is a major improvement. Other modifications correct l'ORTRAN errors discovered in the one-dimensional flame propagation model. Using the same burn velocity, the mcdifications will yield a more complete burn compared to the results from the pr2vious version. ne modified code is referred to as HECTRI.8; it is mainly u*.ai to develop and assess combustion models which are considered for incorporation into CONTAIN and MELCOR, and will not be released offidally.

Publication Date:

l',ovember 19+0 Prepared by:

L T. Pong Contracton Scienm Applications International Corporation, 2109 Air Park Road, SE, Albuquerque, NM 87106; Sandia National laboratories, P.

O.

Box 5800, Albuquerque, NM 87185 Prepared fon NRC Division of Systems Research, Office of Nudear Regulatory Research Keywords:

hydrogen, combustion, reactor accidents, containment buildings, flame propagation, CONTAIN codes, MELCOR codes 101

NUREG/CR-5602 G:ncr:1 EGG-2606

Title:

Simplified Containment Event Tree Analysis for the Sequoyah Ice Condenser Containment Descriptioru An evaluation of a pressurized water reactor (PWR) ice condenser containment was performed. In this evaluation, simphfied containment event trees (SCETs) were developed that utilized information generated by the NRC's Draft NUREG-1150 effort. Specifimlly, the computer programs and data files produced by the NUREG-1150 analysis of Sequoyah were used to electronimlly generate SCETs, as opposed to the NUREG 1150 accident progression event trees (APETS).

SCETs were deve. loped for five of the seven plant damage state groups (PDSCs),

which indude: both short-and long term station blackout sequences (SBOs),

transients, loss of-coolant aeddents (LOCAs), and anticipated transient without scram (ATWS). Steam generator tube rupture (SGTR) and event V PDSCs were not analyzed because of their containment bypass nature. After being benchmarked with the APETs, in terms of containment failure mode and risk, the SCETs were used to evaluate potential containment modifications. The modifications were examined for potential to mitigate or prevent containment failure from hydrogen burns or direct impingemmt on the containment by the core (both factors identified as significant contributors to risk in the NUREG 1150 Sequoyah analysis). Ilowever, b&ause of the relatively low baseline risk postulated for Sequoyah (1,e.,12 person-rems per reactor year),

none of the potential modifications appear to be cost effective.

The codes used in the development of the SGITs are principally EVNTRE, PSTEVNT, SEQSOR, PARTITION, MACCS, and a number of undocumented translator codes. The checkout process begins with the EVNTRE code, which is used to evaluate the APET. The results from an EVNTRE run cannot be verified published resul s, so the first verification occurs after the accident against t

ro sion bins (APBs) are reduced to containment failure modes using the j

tiVNT code. PSTEVNT is used both to redum the APBs to containment failure modes and to source term bins. The output from the source term reduction step is passed to SEQSOR.

SEQSOR generates source term release information for each of the source term bins. 'The code is also capable of representing uncertainties in key source term issues. Direct computation of consequences for each of the source terms generated by SEQSOR would require excessive computer resources. Therefore, the SEQSOR output is passed on to a reduction code called PARTITION.

PARTITION identifies an early and chronic fot:lity weight for each source term.

releases, ptionally provide a summary of these fatality weights over all the It can o or continue with the reduction by locating each release on a two-dimensional plot of the early fatality weight.versus the chronic fatality weight. The output from PARTITION includes the averaged source term release information for each source term group (cell) and subgroup. This information, after some additional formatting, is used in the MELCOR Accident Consequence Code System (MACCS) analysis.

The offsite conseguences associated with each source term group are calculated using MACCS. The MACCS consequence information is the last data input required to complete the nsk calculation. After assembly in a risk matrix, the final risk numbers can be verified by comparison with published results.

Publication Date:

December 1990 Prepared by:

W. J. Calyean, J. A. Schroeder, D. J. Pafford Contractor:

EC&G Idaho, Inc., P. O. Box 1625, Idaho Falls, ID 83415 Prepared for:

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

failure mode analysis, containment systems, damage, PWR type reactors, uoyah-1 reactor, ice condensers, EVNTRE codes, PSTEVNT codes, SEQSOR es, PARTITION codes, MACCS codes 102

NUREG/CR-5605 LAPUR ORNUTM-n621 1

Title:

1APUR Benchmark Against In-Phase and Out-of Phase Stability Tests Descriplicn:

his report documents a benchmark of the LAPUR computer software versus expen.nents stability data collected during startup testing at the Oskarshamn-3 reactor. The data consist of decay ratios and natural frequencies of oscillation measured under several reactor operating conditions. Satisfactory agreement was found between the measured decay ratios and the ones calculated by IAPUR for both the in-phase and out-of-phase instability modes. The largest error m the calculated decay ratio was 0.11, which corresponded to test point 3. Because test point 3 corresponds to a very stable condition, the decay ratio was difficult to determine from the experimental measurements. The fact that 1.APUR predicts that in tests 7,8, and 9 the out-of phase mode should have been unstabic (decay ratio 1.03) when indeed a limit cycle was observed implies that the theoretical approach used in LAPUR to estimate the stability of the out of phase mode is not only physically sound but also able to produce numerically currect results.

He LAPUR code is a frequency domain representation of the dynamic behavior of boiling water reactors (BWRs). LAPUR sc2ves the reactor state equations to compute the open-loop transfer functions that mathematimlly represent the most relevant feedback paths between the different variables in a BWR vessel. The code forms the closed-loop transfer function by combining these individual feedback paths with a mathematical model of the neutron flux dynamics.

LAPUR estimates the reactor stability by locating the most unstable pair of complex poles in the closed loop transfer function.

Input data for the LAPUR benchmark were provided by ABB Atom and the Swedish Nuclear Inspectorate (SKl). The data consisted of operating conditions such as power, flow, and axial and radial distributions; gecmetric data such as number of bundles and dimensions; and cross section data that were used to estimate density reactivity coefficents. The main results are comparisons between calculated and measured values for decay ratio and oscillation frequency.

Excellent agreement was observed between the two sets of decay ratios; however, the oscillation frequences calculated by LAPUR were consistently underpredicted by 10 to 25 percent.

Publication Date:

October 1990 Prepared by:

J. March-Leuba Contractor:

Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831 Prepared fon NRC Division of Systems Technology, Office of Nuclear Reactor Regulation Keywords:

BWR type reactors, reactor stability, OKG-3 reactor, oscillations, reactor noise

NUREG/CR-5607 Las Cruces Trench Sita Drta Base Titic How and Transport at the Las Cruces Trench Site Experiments 1 and 2 Descriptiotu

-Two water flow and solute transport experiments were performed as part of a comprehensive - field trench - study - near Las Cruces, New Mexico. D ese experiments were designed to provide data to test determinir. tic and stochastic models of vadose zone flow and transport. In Experiment 1, a 4m by 9m area was irrigated for 10 days with water containing tritium. Thereafter, water was applied without tritium for an additional 76 days. Simple one. dimensional uniform and layered soll deterministic models for infiltration adequately predicted the overall movement of the wetting front during infiltration but poorly predicted point values for water content due to spatial variability. Use of the layered soil model, rather than the uniform soil moctel, did not consistently improve prediction accuracy for this particular field application.

In Experiment 2, a 1.22m by 12m area was irrigated for 11.5 days with water containing tritium and bromide. Thereafter, water was applied without tracers for an additional 64 days. Water and bromide moved fairly uniformly during infiltration, whereas high concentrations of tritium developed on one side of the irrigated area. During redistribution, tritium moved little, whereas bromide displayed significant movement both downward and to - one side.

A-two.dtmensional deterministic model for water flow showed qualitative, but not quantitative, agreement with observations. A two. dimensional deterministic model for solute transport poorly described trinum and bromide movement during redistribution, ne las Cruces Trench Site Data Base is divided into three subsets of data files:

1) Soil Property Data Files, 2) Plot #1 Data Files, and 3) Plot #2 Data files. The Soil Property Data Hles contain the soil property characterization data from the Las Cruces Trench Site. These data indude laboratory and in situ saturated hydraulic conductivities, soll particle size distribution data, measured water retention data, and estimates of the van Genuchten parameters.

The Plot #1 Data Files contain water and solute application rate informa>n and measurements of volumetric water contents, tensions, and relative tritium concentrations for the first trench experiment. De Plot #2 Data Bles contain water and solute application rate information and measurements of volumetric water contents, tensions, and relative tritium and bromide concentrations for the second trench experiment.

All files have the same general tabulated structure. The fields 0.e., columns) are tab delimited, and a carnage return is used to denote end of record 0.e., line).

De tecnrds are variable length;- the maximum is 87 bytes (or 87 ASCll characters). He first line of each file contains the column titles. The value 999 is used to denote missing data. All character strings are surrounded by single quotes. De maximum number of characters in a field is 15 (17 induding quotes). Descriptions of the Plot #1 and Plot #2 Data Files, as well as the data in -

the files, are given in Appendix A.

Publication Date:

August 1990 Prepared by:

P. J. Wierenga, D. B. Hudson: R. C. Hills, I. Porro, M. R. Kirkland, J. Vinson Contractor:

Department of Soil and Water Science, University of Anzona, Tucson, AZ 85721; onomy and Mechanical Engineering Departments, New Mexico State Universi, Las Cruces, NM 88003 Prepared fon NRC Division of Engineering, Office of Nudcar Regulatory Research Keywords:

hydraulic conductivity, soils, ground water, solutes, particle size, spatial distribution, fic!d ttsts, tritium, bromides, experimental data, data base management 104

i NUREG/CR-5622 Balence of ?lant Dctabase SAIC-89/1148 l

Title:

Analysis of Reactor Trips Originating in Balance of Plant Systems

==

Description:==

This report documents the results of an analysis of balanw-of-plant (BOP) related reactor trips at commercial U. S. nudear power plants over a five-year period, from January 1,1984 through Decemicr 31, 1988. He study was performed for the Plant Systems Branch, Office of Nudear Rmor Regulation, U. S. Nudcar Regulatory Commission. The objectives of the itudy were: 1) to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such ecors; 2) to prepare a computerized database of BOP-related reactor trip events and use the database to identify trends and patterns in the population of these events: 3) to investigate the risk implications of BOP events that challenge safety systems; and 4) to provide recommendations on how to address DOP related concerns in a regulatory context.

A database of BOP-related reactor trips was created with data drawn from th'e Licensee Event Report (LER) database maintained by Oak Ridge National laboratory. The Sequence Coding and Search System (SCSS) for the LER database was used to identify potentially relevant LERs. Approximately 2030 trips involving BOP systems were identified. The information collected from the LER search was analyzed to determ!nc whether the reactor trip was din'etly related to a failure of a BOP component or function. If so, the trip information was incorporated into the BOP database. Of the 2030 LERs reviewed,1405 BOP-related events were considered appropriate for entry into the BOP database.

The BOP database was written for use on a PC or compatible microcomputer using dBASE 111 Plus.

Supplementary databases were also found to be necessary for conducting analyses of trends and patterns. These databases contain plant and critical hours (number of hours the reactor was entical) data. The supplementary plant database indudet the following data elements: operating license data, nudcar steam supply system vendor, architect / engineer, and turbine-generator manufacturer, ne critical hours supplementary database indudes: critical hours per year for each plant for the years 1984 thmugh 1988 and total critical years accumulated during the period 19M through 1988. Appendix A contains a sample of 30 BOP database entries. Printouts of the supplementary databases are induded in Appendices J and K.

Pub!! cation Date:

September 1990 Prepared by:

F. T. Stetson, D. W. Callagher, P. T. Le, M. W. Ebert Contractor:

6 cnce Applications Intemational Corporation, P. O. Box 1303, 1710 GoodeMge Drive, McLean, VA 22102 Prepared for:

NRC Division of Systems Technology, Office of Nuc! car Reactor Regulation Keywords:

nudcar power plants, reactor safety, reactor shutdown, hazards, reactor licensing, criticality, data base management i

10S

NUREG/CR-5G14 GEMINI UCRL-ID-104845

Title:

Consequence Evaluation of Radiation Embritt!cment of Trojan Reactor Pressure Vessel Supports Descriptioru This document describes a consequence evaluation to address safety concems raised by the radiation embrittlerrent of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability.

The structural evaluation condudes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to t. 4 steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent analysis further demonstrates that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferrm to them through the RCL piping.

The effects evaluation, employing a systems analysis approach, investigates initiating events and tne reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns: (1) the RPV movements could cause multiple rupture of instrumeritation thimble tubes or the guide tubes that penetrate the bottom of the RPV and result in a loss of core coolant that may lead to core uncovery; (2) the deformation of the RCP casing may cause the impellers to bind 2.nd result in loss of natural circus Nn, and the tilting of the pump may affect.ts coastdewn ability; (3) the control rods could bend in the event of tilting of the RPV and the ability to insert control rods during a reactor trip may be affected; and (4) the rupture of the 10-in safety injection lines could impair the function of the emergency core cooling system.

Further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement wiH not result in Jonsequences of significant safety concern.

'Ihe basis of the geometrically correct Trojan model used for the analyses described in this report is a reduced model of the Zion plant, but with some significant changes. CEMINI is a finite element computer program for two-and three-dimensional linear, static, and seismic structural analysis. With the geometrically correct Trojan model, structural analyses performed wiih GEMINI predicted the forces, moments, and stresses in the Trojan RCL The predicted frequency response of the Trojan plant with and without RPV supports was also deterinined using CEMINL Publication Date:

October 1990 Prepared by:

S. C Lu, S. C. Sommer, C. L Johnson: H. E. Iambert Contracton Lawrence livermore National Laboratory, P. O. Box 808, Livermore, CA 94551; Fl'A Associates,3728 Brunell Drive, Oakland, CA 94602 Prepared fon NRC Division of Engineering, Office of Nuclear Regulatory Research Keywords:

Trojan reactor, pressure -essels, supports, embrittlement, reactor safety, structural models, reactor cooling systems 106

]

NUREGICR-5649, Vol.1 COMMIX-1C ANL-90/33, Vol.1

Title:

COMMIX lC: A Three-Dimensional Transient Single-Phase Computer Program for The~tnal-Hydraulic Analysis of Single-Component and Multicomponent Engineering Systems Equations and Numerics

==

Description:==

The COMMIX 1C computer program, an extended version of previous single-ohase COMMIX codes, is designed to analyze stead y-state / transient, single-phase, three-dimensional fluid now with heat transfer in reactor component and multicomponent systems. The fmr major improvements implemented to develop COMMIX-1C are:

(1) New finite-volume formulations for the mass, momentum, and energy equations to extend application to subsonic compressible Hows. The new momentum formulation employs the concept of a volume-averaged velocity.

It makes the numencal calculation more robust than in previous versions. It also makes the location of pressure change coincide with that of density change for one-dimensional Gows. In addition, the new discretized momentum equations satisfy the onedimensional Bernoulli equation.

(2) Addition of a new flow-modulated skew upwind discretization scheme in the energy equation to reduce numerical diffusion. This scheme is considered better because it not only reduces numerical diffusion but also has a theoretica: basis for not producing overshoots and undershoots that are physically unrealistic.

(3) Addition of two matrix solvers, the Yale Sparse Matrix Package and the preconditioned conjugate gradient method, for the discretired equations.

(4) An improved k-epsilon two-equation turbulence model that is more robust and better validated than that used previously.

Volume 1 of this report, entitled EcuaFons and Numerics, describes in detail the basic equations, formulation, solution procedures, flow-modulated skew-upwind discretization scheme, and models used for the following phenomena:

momentum, interaction between fluid and stationary solid structures, thermal interaction between fluid and stationary solid structures, and k-epsilon two-equation turbulence. Volume !!, entitled User's Guide and Manual, contains the flow charts, subroutine descriptions, geometry modeling, available options, input instructions, sample problems, etc.

Publication Date:

November 1990 Prepared by:

H. M. Domanus, Y. S. Cha, T. H. Chien, R. C. Schmitt, W. T. Sha Contractor:

Argonne National Laboratory,9700 South Cass Avenue. Argonne, IL 6&l39 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

Quid now, heat transfer, reactor components, numerical analysis, Guid. structure interactions, turbulence

)

1 107

NUREG/CR-5649, VCl. 2 COMMIX-1C ANL-90/33, Vcl. 2

Title:

COMMIX 1C: A nroe-Dimensional Transient Single-Phase Computer Program for Thermal Hydraulic Analysis of Single-Component and Multicomponent Engineering Systems User's Guide and hianual Descriptioru The COMMIX 1C computer program, an extended version of previous single-phase COMMIX codes, is designed to analyre steady state / transient, single-phase, three-dimensional fluid flow with heat transfer in reactor component and multicomponent systems. The four major improvements implemented to develop COMMIX 1C are:

(1) New finite-volume formulations for the mass, momentum, and energy equations to extend application to subsonic compressible flows. The new momentum formulation employs the concept of a volume-averaged velocity.

It makes the numerial calculation more robust than in previous versions. It also makes the location of pressure change coincide with that of density change for one-dimensional flows. In addition, the new discretized momentum equations satisfy the one-dimensional Bernoulli equation.

(2) Addition of a new flow-modulated skew-upwind discretization schece in the energy equation to reduce numerical diffusion. This scheme is considered better because it not only reduces numerical diffusion but also has a theoretical basis for not producing overshoots and undershoots that are physically unrealistic.

(3) Addition of two matrix solvers, the Yale Sparse Matrix Package and the preconditioned conjugate gradient method, for the discretized equations.

(4) An improved k-epsilon two-equation tuibulence model that is more robust and better validated than that used previously.

Volume 1 of this report, entitled Equations and Numerin, describes in detail the basic equations, formulation, soluton procedures, flow-modulated skew-upwind discretization scheme, and models used for the following phenomena:

momentum, interaction between fluid and stationary solid structures, thermal interaction between fluid and stationary solid structures, and k-epsilon two-equation turbulence. Volume II, entitled User's Guide and Manual, contains the flow charts, subroutine descriptions, geometry modelmg, available options, input instructions, sample pmblems, etc.

This volume dese:ibes the step-by-step procedure in suticient detail so that a reader unfamiliar with the COMMIX code an begin to use it with little difficulty. Sections on the structure of COMMIX-1C, geometry modeling, initialization, thermal-structure modeling, force-structure modeling, solution procedures, auxiliary input, steady-state calculations, transient calculations, and operating COMMIX-1C are included.

Publication Date:

November 1990 Prepared by:

H. M. Dornanus, Y. S. Cha, T. H. Chien, R. C. Schmitt, W. T. Sha Contractor:

Argonne National Laboratory,9700 South Cass Avenue, Argonne, IL 6G139 Prepared fon NRC Division of Systems Research, Office of Nuclear Regulatory Research Keywords:

fluid flow, heat transfer, mactor components, numerical analysis, fluid structum interactions, turbulence 108

L f

NUREG/CR=5653 MARCl!

PNL-7476

Title:

Reeriticality in a BWR l'ollowmg a Core Damage Event Dee.cription:

1his document describes the results of a sely mndutted by Pacific Northwest laboratory to assist 'he U. S. Nudcar Regulatory Commission in evaluatmg the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Bawd on a conwrvative bounding analysis, this report condudes that there is a potential for recritNtity in DWR$ il core refhwi occurs after control blade melting has begun but prior to signif cant fuel rod melting. Ilowever, a recriticality event will most likely not generate a pressure pulse significant enough to fail the veswl. Instead, a quast. steady power level would result, and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event.

Two strategies ara identified that would aid in regaining mntrol of the reaaor and terminate the recriticality event before containment failure pressures are reached. The first. > ategy involves initiating boration in ection at or before time of core reflood if the potential for control blade me ting esists. The second strategy involves initiating residual heat removal suppresuon pool cocling to remove the heat load genwated by the recriticality event and thus exter d the time available for boration.

The MARCil calculat ons for NUREG 1150 were perbimni using the Source Term Code P,ckage (STCP) or V192 version of trw code. The calculations performed spedfically for this study used a more licent veralen iversion V1941.

The Vl94 version contains a number of modeling enhancements, including a BWR control blade relocation model in which melted blade nodes fall either to the core plate below the cnre or into the water in the lower head. Version V194 also contains enhanced capability to calculate heat transfer and metal. water reaction during reflooding of a degraded core.

During a r.evere acddent, the grid spacers and end fittings which define the fuel rod assembly spadng may melt or collapse, fuel rod clad could melt or break releasing pellets in a random manner, or the fuel pellet:, tnemselves may shatter or molt, forming smaller or larger particles. To conservatively model the unknown, ill defined geometry, the partides were assumed to be spherical. To determine the optimum conditions, the spacing between the particles was varint Computer codes NITAWL and XSDRNPM S were used to calculate k infinity for each diameter, spacing, and boron cencentranon combination. The Dancoff self-shielding correction was calculated using the MCDAN program. A semi. automated spreadsheet was developed to generate the large number of inpst data sets required. Appendlas in this report show the spreadshcet and typical output files (the caw information print file and the NITAWL and XSDRNPM S input filed. An additional appendix contains a typical MCDAN output and the datated Dancoff corrections for all the cases. Results of the OPaCEN code, which computes time dependent concentrations and sourm terms of a large number of isotopes with generation and depletion through transmutation. fission, and radioactive decay, were used to estimate the spontaneous 'non and source strength of irradiated fuel. The initial power is estimated t -. I x 1(F M w.

Publicatior. Date:

December in Prepared by:

W. B. Scott. D. C. Harrison, R A. Libby, R. D. Tobra R D. Wocton, R. S.

Daning, R. W. Tayloe, Jr.

Contractor:

Padfic Nor+hwest Laboraory, P. O. Box 999, Richland, WA 99352; Battelle Memorial Institute,505 King Avenue, Columbus, OH 43201 Prepared for:

NRC Division of Sys. cms Research, Office of Nudcar Regulatory Rescatch Keywords:

reactor accidents, BWR type reactors, reactor cores, damage, excursions, failures, criticality, STCP codes, NITAWL codes, XSDRNPM mdes, MCDAN codes, OPICEN codes 109

)

NUREG/CR-5659 Control Room liabitability Systc;m SAIC-90M054

Title:

Control Room liabitability System Review Mcdels Descript!osu nie document desaibes a method for calculating contml room operator dows from postulated reactor acddents and toxic chemical spills as part of the resolutlon of TMI Action Plan Ill.D3A. he Control Room Hab!' ability System (CRHS) computer mdes pretiented in this te rt use source concentrations calculated by either TACTS, ITIT, or EXT and transport them via user defined flow :ates to the control room envelope. The CONHAll and CllEM codes compute doses to six organs from up to 150 radionuclides (or 1 tode chemical) for time steps es short as one second. The three supporting CRHS codes, written in Clipper, assist in data entry and manipulation and graphically display the roults of the FORTRAN calculations. There are two permanent data file libraries, ICRP.02 and ICRP.30.

The CRH code, written in Clipper, is a pop-up :Nnu driven program designed to aid in entering meteorological and system flow data. Its primary function is to allow input of CRHS parameters and meteorological paramete. 'or subsequent use by' t e CONHAB or CllEM cods A secondary functic be h

assemblage of all six input data files for CONHAd.

CONHAB is the FORTRAN program ftw computing estimated-doses due 19 radiological re! cases. Using the hfe CRHSYS.Dar created by CRH, it computes control room concentration and associated doses to six organs from up to 122 nuclides, each capable of being modeled as three chemical,orms (clemental, organic, and particulate), from up to four sources, and 2.6 million time steps

.(the number of seconds in a rnonth).

CHEM, written in FORTRAN, is the equivalent of CONilAB for use in the evaluation of toide gas concentrations rather than radionudides. De mathematical model used is identical to that in CONHAB; however, the only removal orocess modeled is dilution. The code reads an output file generated by EXTRAN'and applies this source to each of the four potential entry points to the control room.

De CRifPLOTR and CRHPLOTC codes plot the mults of a radiological or toxic gas analysis, respectively. Both are written in CPoper ustrig Flipper graphin

= libraries and r.re designed to run on most Pi display systems (VGA, ECA, CGA, and Hercules).

no CRHS codes require an EM PC or PS/2 compatible mt:m:omputer with a hard disk drive. A ver; ion utilizing a math co processor is.wallable.

Appendices A4 contain listings of the five codes, the two libraries, database file

- structures, and sample output for CONHAB and CHEM.

Publicatlast Date:

December 1990 Prepared by:

H. Gilpin Contractor:

Sdence Applications International Corporation.1710 Goodrid e Drive, McLean, S

VA 22102 Prepared for:-

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

control rooms, reactor operators, radiation doses, toxic materials, reactor

~

accidents, chemical spills, gas spills, CRH codes, CONHAB codes, CHEM codes, CRHPLOTR todes, CRHPLOTC codes, ICRP.02 codes, ICRP.30 codes t

110 dv-g-4a*

--r q

vy$-ye -ee

-w-'

v -se p =rm e -rww rayw*9 g-*PrrM y-gmuy-*=-N

- ee e-v y m--w=y v y wao pwey'w-wet---

u e-pw"e' m ea ev m-w ww-vvw*-

w-ww1w~%mv M J-am---me-G t"'wtr-e aww'wwr'-s--we-amw--ha---*--------a*

r

NUREG/IA-0011 TRAC-PF1/ MODI i

Title:

TRAC PF1 h10D1 Post Test Calculations of the OECD LOFT Esperiment LP 5B 1 Descriptioru An analysis of the small, hot leg break, OECD LOFT Esperirnent LP-SB 1 using the test-estimate mmputer code IRAC-PF1/htOD1 is presented.

Deptions of the LOFT facility and the LP-SB 1 esperiment are given, and development of the TRAC PF1/hiODI input model is detailed. The calculations performed in achieving the steady-state conditions, from whkh the eqvriment was initiated, and the specification of esperimental boundary cenditions are outlined.

Results of a "Baw Cast transient calculation are found to be generally consistent with thow reported by other OECD LOFT Program Review Group memben.

The experimental trends with respect to pressure histories and minimum system mass inventory are reasonably well reproduent by the TRAC PF1/ MODI calculation. However, the inability of TRAC PF1/htOD1 to account for main pipe stratification in determining Guld conditions in a side branch leads to significant discrepandes between the measured and predicted break line and hot leg densities and is identified as the main rason for the poorly predicted break mass-flow rate.

Implementation, via the TRAC-PF1/ MODI control system, of currelations for determining side branch quality as a function of main-pipe s ratified liquid level are shown to be effective in improving the predicted hot leg and break line dansities and break mass flow rate. The remaining differenas between measured and predicted data are considered to be due to deficiencies in the TRAC PF1/ MODI critical flow model and the sensitivitiy of the break flow to the hot leg liquid level twhavior.

TRAC (Transient Reactor Analysis Code) is an advanced best-estimate computer code, developed at the los Alamos National lateratory (LANL), for analyzing transients in thermel-hydraulic systems. Specifically, TRAC PF1/ MOD 1 was developed for analyzing postuleted accidents in pressurized water reacton (PWRs). Tne versions of the mde used for the calculatior s desenbed in this document were Venion B02A (Base Case Calculation) :nd Version B02C tBase Case Calculation with EPRI Correlation) which contain the LANL updates to TRAC-PF1/ MOD 1 Venion 12.7. The p,r: gram was run on a Cray X MP computer.

Publication Date:

Apnl 1990 Prepared by:

E. J. Allen Contractor:

United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset DT2 BDil, United Kingdom Prepared for:

NRC Office of Nuc! car Regulatory Research Keywords:

LOFT reactor, PWR type reactors, computerind simulation, intemational cooperation, UKAEA, flow models, OECD 9

4 111

NUREG/lA 4012 RELAP5/ MOD 2

Title:

RELAP/ MOD 2 Calculations of OECD LOIT Test LP.SB41

==

Description:==

To usist the Central Electricity Cencrating Board (CECB) in aswssing the capabilities and status of RELAPS/ MOD 2, the code was used to simulate SBLOCA test LP SB-01 carried out in the LOIT experimental reactor under the OECD LOIT program. This test simuhted a 1.0% hot leg break in a pressurized water reactor, with early tripping of the primary coolant circulating pumps. This report-compares the results of the RELAP5/ MOD 2 analysis with experimental measurements.

Comparison of the present calcul6 tion with earlier RELAPS/ MODI calculations shows that significant improvements have been made. Most notably, the horizontal stratification model in MOD 2 was found to enable improved calculation of fluid density dcme to the break in this test. In addition, mass conservation errors, numerical stability, and the computer run. time were all greatly improved, compared W ' en earlier CECB analysis using MODI In particular, MOD 2 which was ex, cuted on a Cray.1 at a CPU /real time ratio of 1.16, appears to 1m between 15 and 20 times faster than MOD 1, taking into account the fact chat RELAPS/ MODI was executed on a Cyber 176 at a CPU /real time rar, i 34.2.

The major difference between the RELAPS/ MOD 2 results and the experimental data is in the critical discharge flow rate. It is conduded that the error arises from thermal di uilibrium effects in the discharge nozzle which are not modeled in the

e. However, the discrepancies are not considered unduly significant for safety analysis of small break loss of coolant accidents in nudear power plants, since in this application such effects would normally be allowed for by rforming sensitivity studies to break size, orientation, etc. Activation of the RE PS/ MOD 2 vertical stratificatien model in the upper plenum was found to lead to the erroneous calculation of sudden draining of the hot legs. The current basis for general application of this model appears questionable.

Publicatlan Date:

January 1990 Prepared by:

P. C. Hall, C. Brown

.intractor:

Central Electricity Cencrating Board, Barnett Way, Barnwood, Gloucester CL4 7RS, United Kingdom Prepared for:

NRC Office of Nudear Regulatory Rewatch Keywords:

international cooperation, OECD, United Kingdom organizationa, LOIT reactor,

-loss of coolant, computerized simulation, PWR type reactors, RELAPS/ MODI codes 112

-....---~...----. -a.-

l l

NUREG/lA4013 REIAP5/ MOD 2

Title:

RELAP5/ MOD 2 Calculation of OECD LOIT Test LP SB4D Descriptfort in order to trnfirm the ability of RELAP5/ MOD 2 to describe a small break LOCA mjuency in a PWk type geometry, analysis of test LP SH40 of the OECD LOIT csperimental prograrn was carried out. Test LP-SB4D simulated a 0.4%

mld leg break in a PWR, with failure of high head safety injmion; moldown was achieved by feed and bleed of the wcondary side wme time af ter mre This report mmpares the results of the iiELAPS/ MOD 2 analyus with unmvarv esper8 stal measurements.

A Mrs 4.ation of test LP-SD4D was r,vviously arried out at the Central Electricity 14oard's Cencrition Evvelopment and Construction Division (CDCD) using the RELAPS/ MODI mde. RELAPS/ MOD 2 was developed from REtAPS/ MODI and mntains more wphisticated hydraulic models and constitutive relationships. Comparison of the RELAP5/ MOD 2 and MODI calculations show that RELAPS/ MOD 2 twrforms better than RELAPS/ MODI in a numler of key arcas: notably mass errors are much redund, there is improved numerical stability, and improved separator modeling and modeling of acmmutator injection.

Overall agreement with esperiment obtained using RELAPS/ MOD 2 is curllent, with all key phenomena wrrecily predicted, in the proper scsluence with accurate timing. In particular, RELAPS/ MOD 2 was found to tw more stable and robust, with an improved numetimi solution scheme temoving the mass error problems encountered in the earlier code version. The CPU to ral.tirne ratto was approximately 15:1 on the liarwell Cray1 computer.

Publication Date:

January 1990 Prepared by:

C. Harwood, G. Drown Omtractor:

Central thstricity Cencrating Board, Barnett Way, Barnwood, Cloucester CL4 7RS, United Kingdom Prepared for:

NRC Office of Nudear Regulatory Rencarch Keywords:

international emperation, OECD, United Kingdom organizations, LOIT reador, loss of molant, mmputerind simulation, PWR type reactors, RELAP5/ MODI codes 113

AdiD.t'3C~L.aL1F&

1 1

.,, _?. 6,..

RELAP5/ MOD 2 WRMh 1 SW RELAPS/ MOD 2 Aswssment, OECD LOIT Small Break Expenment LP.SB-03 Descriptforu An analysis of the esperimental results and post test calculations using RELAP5/ MOD 2 carned out for OECD-LOIT small break esperiment LP-SB-3 are prewnted. Experiment LP SB 3 was conducted on March 5, 19M in the less-of Fluid Test (LOIT) facility located at the Idaho National Engineering laboratory (INEL). The esperiment simulated a small mld leg break, with concunent loss of high pressure injection system, and cwidown and recovery by feed and bleed of the steam generator wcondary side and accumulator injection, respectively.

This document prewnts a short post test analysis of the esperiment emphasir.ing the results of additional analysis performed during the course of this task. The RELAPS/ MOD 2 input model and results of the post test calculation are documentcd. Included in the report are the results of a wnsitivity analysis which show the prtdicted thermal hydraulic response to a different input mal.

He RELAPS/ MOD 2 computer ccde was used for the post test calculation of Experiment LP sib 3. RELAPS/ MOD 2 is an advancul, best-estimate computer program developed at the INEL for the analysis of less of-Coolant. Accidents (LOCAs) and other pressurtred water reactor (PWR) transients. RELAPS/ MOD 2 employs a finite-difference fluid cell representation of the primary and secondary coolant systems. The input model consists of a total of 32 fluid alls for the and ECC vessel and 100 cells for the remainder of the primary, wmndary, computer.

systems. For this ttst, RELAPS/ MOD 2 was run on a CDC Cyber 1<6 The latest avallable version (Cycle 33 to 36.1) of the mde was used. The particular test selected for the analysis included wveral phenomena potentially relevant to any PWR plant operating in Switzerland.

De RELAPS/ MOD 2 code was shown to be valuable in understanding the physical phenomena in the experiment. Although differences in detail were observed between the calculational results and esperimental data, the cxdc generally performed well, predicting all the key events in the currect sequenm with reasonable timing.

Publication Date:

April 1990 Prepared by:

S. Cuntay Contractor.

Paul Scherrer Institute, Cil 5303 Wurentingen, Switzerland Prepared for.

NRC Office of Nuclear Regulatory Research Keywords:

loss of coolant, LOIT reactor, PWR type reactors, computerized simulation, sensitivity analysis, international cooperatton, Switzerland, OECD, reactor cores, steam generators 114 l

1

.i

NUREG/IA-0019 TRAC-PF1/ MODI

Title:

TRAC PF1/ MODI Post Test Calculations of the OECD Loft Experiment LP S!12 Descriptforu An analysis of the OECD-LOIT-LP SB-2 experiment making use of TRAC PF1/ MOD 1 is desenbed in this document. The LP SB2 experiment studies the effect of a delayed pump tnp in a small broak LOCA scenano with a 3-inch equivalent diameter break ln the hot hg of a commerdal pressurized water reactor (PWR) operating at full power. The experiment was performed on July 14, 1983 in the LOFT facility at the Idaho Natmal Engineering Laboratory (INEL) under the auspices of the Organization for Economic Co-operation and Development (OECD). The evolution of the experiment was determinal by several features, among the most important of which were the flow patterm present in the loop, vapor pull-through and liquid entrainment observed in the break line, and pump behavior.

This analysis prer.cnts ar evaluation of the code capability in reproducing the complex phenomena which determined the LP-SB-2 transient evolution. The analysis is comprised of the results obtained from two different ru s. The first run is described in detail analyzing the main variables over two time spans:

short and longer term. Several conclusions are drawn, and then a second run testing some of these conclusions is presented.

All calculations were perfonned on a Cray X MP computer using Winfrith TRAC PF1/ MODI Versmns B02A for RUN A and B02C for RUN B. Both versions contain los Alamos updates up to Version 12.7. A description of the differences between the Winfnth code versions and Version 12.7 is given in Appendix B. The code (CPU / Problem) time for RUN A was considered good, havmg a value of approximately 1.95, in RUN B this ratio increami to approximately 2.3.

Vapor pull-through and licluid entrainment were observed to occur at the offtake of the break line. TRAC 1 F1/ MODI was unable to recognize this phenomenon is no relevant model it, actually implemented relating quality in a branch to the thermal hydraunc condittoris of the fluid in the main pipe, as well as considering the geometne charactenstics of the break line junction to the main line. The TRAC built-in flow regime map perfor ns well in identifying fully stratified conditions in the hot leg. The mass loss predided for RUN A was large enough to provoke a mild uncovery of the top of the core after the pumps trip, contrary to experiment. In RUN B the break flow was adeguately predicted, and the mass loss dosely matched the experimental result. No core uncuvery was predicted.

Publication Date:

December 1990 Prepared by:

F. Pelayo Contractor:

United Kingdom Atomic Energy Authority. Atomic Energy Establishment, Winfrith, Dorchester, Dorset DT2 BDH, United Kingdom Prepared fon NRC Office of Nucicar Regulatory Research Keywords:

loss of coolant, LolT reactor, PWR type reactors, computerized simulation, international (ooperation, UKAEA, pumps, flow models, OECD 115

NUREG/1A 0021 RELAP5/ MOD 2

Title:

RELAPS/ MOD 2 Calculations of OECD toft Test LP.SB.2

==

Description:==

To help in assessing the capabilities of RELAPS/ MOD 2 for pressurized water reactor (PWR) fault analysis, the code is being usai by the Central Electricity Generating Board (cecil) to simulate several small LOCA and pressurized transient experiment, in the LOFT experimental reactor. This report describes an analpis of small LOCA tmt LP.SB-02, which simulated a 1% hot leg break LOCA in a PWR, with delayed tripping of the primary coolant pumps. This test was can;ed out under the OECD LOFT Program.

An important deficiency identified in the code is inadequate modeling of the quality of the Guld discharged from the hot leg into the break pipework. This gives rise to irrge errors in the calculatni system mass inventory. The effect of using an improved model for vapor pull through into the break is desenbai.

A second significant code deficiency identified is the failure to-predict the occurrence of stratified flow in the hot leg at the correct time in the test. It is believed that this error contributal to gross errors in the loop flow conditions after about 1300s.

Additional separate effects experimental data necessary to resolve the code defidendes encounteral are identified as the following:

(a)- Transition to stratified flow in geometries resembling a PWR hot-le.

(b) Flow quality in an offtake pipe connected to a larger horizont4 pipe in which there is a two. phase flow with a mass velocity of more than 1000 kg i4 ms.

Experiments designed to provide this information are currently in preparation at AERE, Harwell.

Publicatlan Date:

April 1990 Prepared by:

P. C. Hall Contractor.

Centrol Electricity Generating Board, Cencration Development & Construction Division, Barnwood, Gloucester CL4 7RS, United Kingdom Prepared for:-

NRC Office of Nudear Regulatory Research Keywords:

international cooperation, OECD, United Kingdom organiutions, LOFT reactor, loss of coolant, computerized simulation, PWR type reactors, Sizewell B reactor, RELAPS/ MODI codes 116

NUREG/lA 0022 TRAC-PF1/ MOD 1

Title:

TRAC PF1/hiODI Post Test Calculations of the OECD LOIT Experiment LP-SB-3 Descriptiort:

An analysis of the small, mld leg break, OECD LOIT Experiment LP SB 3 using the best-estimate cumputer code TRAC PFl/h10D1 is presented.

Descriptions of the LOIT facility and the LP SB-3 expenment are given, and development of the TRAC-PF1/h10D1 input model is detailal. The calculations performed in achieving the steady state conditions, frum which the experiment was initiated, and the speafication of esperimental boundary condinons are outlined.

TRAC (l'ransient Reactor Analysis code) is an advancwl best estimate computer code, developed at the los Alamos National laboratory (LANL), for analyzing transients in thermal hydraulic systems. Specifically, TRAC-PF1/ MODI was developed for analyzing postulated acridents in pressurized water reactors (PWRs). The version of the code used for the calculations described in this document was AEEW Version B02A which contains the LANL updates to TRAC PF1/htODI Version 12.7. A total of 41 components,47 junctions, and 182 cells were used in the model, and the runs were made on a Cray X MP.

Overall trends with respect to pressure histories, minimum primary system mass inventory, and accumulator behavior are reasonably well reproduced by TRAC-PF1/htODI. Prior to break uncovery, the break mass-flow rate is sli htly over-predictcd by the TRAC critical flow model. The most signi.icant discrepancy is in the rate with which the fuel rod cladding temperature rises during the core uncovery phase of the transient. TRAC PFl/hiODI, in common with other codes, significantly over predicts the rate with which the core heats up. Contrary to experimental observations, conditions for reflux condensation ce not prwileted by TRAC PF1/h10D1 during this part of the transient. Tb.s.

together with the under prediction of cure density by TRAC's interphase d,ag rnodel, is considered a potential contributor to the pcotly predicted rate of core heat up.

Publication Date:

April 1990 Prepared by:

E. J. Allen, A. P. Neill Contractor:

United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset DT2 BDil, United Kingdom Prepared fon NRC Office of Nudcar Regulatory Research Keywords:

loss of coolant, LOFT reactor, PWR type reactors, computerizei simulation, internattorud cooperation, UKAEA, flow models, reactor cores, drag, OECD I

i 117

I NUREC/ZA-0030 RELAP5/ MOD 2

Title:

Asst tsment of RELAP5/ MOD 2 Cole Using 1.oss of Offsite l'ower Transient Data of KNU #1 Plant Descriptforu This report presents a et>Je assessment study based on a real plant tmnsient that occurrni on June 9,1981 at the KNU #1 (Korea Nudear Unit Number 1). KNU

1. 1 is a two loop West % house PWR plant of 587 Mwe. The loss of offsite powe transient occurred at 7/.5% reactor power with a 0.5% per hour power ramp.

The real plant data were cullected from available on-line plant remrding and computer diagnostics.

The transient was at ated by RELAPS/ MOD 2 Cyde 36.05 and mmpared with the plant data to 6 ess code weaknesses and strengths. Some nodalization studies were performed to assist in developing a guideline for PWR nodallration for the transient analysis.

It was found that the code gives stable steady state results and accurate predictions of the plant behavior for the transient, indicating curilent capability for simulating this type of transient. In particular, the (alculated primary thermal behavior closely follows the plant data validating that the thermal-hydraulle and decay power mooci using previous power history data in REl.APS/ MOD 2 is correctly des <ribing the actual phenomena.

In the nodalization sensitivity study h was found that S/G noding with junctions between bypass plenum and steam dome is preferrtd to simulate the S/G water level decreasing and avoid the spurious icvel peak at turbine trip.

The pressurizer presrure increase is sensitive to the insurge flow, it is believed that the estimate of the interfacial heat transfer in a horizontal stratified flow regime may be on the low side, and the estimatal compression effed due to insurge flow may be high.

Publication Datte April 1990 Prepared by:

Bud-Dong Chung, liho-Jung Kim; Young-Jin Lee Contractor:

Nudear Safety Center, Korea Advanced b.nergy Rescatch Institute, P. O. Box 7, Daeduk-Danji, Daejon, Koren; Departinen of Nudear Engineering, Seoul National University, Seoul, Korea Prepared for:

NRC Office of Nudcar Regulatory Rewarch Keywords:

international cooperation, Korean organizations, transicats, computerized simulation, PWR tvpe reactors, KOP.l.1 reactor, steady-state conditions, steam generators 118

m NUREG/IA-0031 RELAP5/ MOD 2

Title:

ICAP Assessment of RELAPS/ MOD 2, Cycle 36.05 Against LOIT Small Break Experiment L3 7 Descriptioru The LOFT small break (1 inch diameter) (nperiment L3 7 was analyzed using the reactor thermal hydraulic analysis code RELAP5/ MOD 2, Cyde 36.05.

De base calculation (Case A) was completed and compared with the experimental data. Three types of sensitivity studies (Cases B, C, and D) were carried out to investigate the effects of (1) break discharge cuefficient Cd, (2) pump two-phase difference multiplier, and (3) liigh Pressure injection System (llPIB) capacity on major thermal and hydraulle (T/H) parameters. A nodallzation study (Case E) was conducted to assess the phenomena with a simplified nodalization.

The results indicate that Cd of 0.9 and 0.1 fit best among the trini ca.cs to the single discharge flow rate of Test L3 7. The pump two-phase multiplier has httle effect on the T/H parameters because of the low dischar[owever, rate and the e flow early pump coastdown in this smaller sized SBLOCA 1 the primary system pressure, flow, and temperature were very sensitive to llPls charactenstics. It is also shown that a simplified nodalization could accommodate the dominant T/H phenomena with the same degree of code accuracy and effidency.

PubIlcation Date:

April 1990 Prepared by:

Euy-Joon Lee, Bud Dong Chung, Hho-Jung Kim Contracion Nudcar Safety Center, Korea Advanced Energy Research Institute, P. O. Box 7, Daeduk-Danji, Daejon, Korea Prepared fon NRC Office of Nudcar Regulatory Research Keywords:

international cooperation, Korean organizations, computertred simulation, PWR type reactors, LOFT reactor, sensitivity analysis, loss of cuolant, high pressure coolant injection i

119

NUREG/IA 0032 RELAP5/ MOD 2

Title:

Assessment of RELAPS/ MOD 2, Cycle E04 Using LOFT targe Break Ex}vriment L2-5

==

Description:==

N LOFT L2 5 LBLOCA Experimer.r was simulated using the RELA 15/ MOD 2 Cycle E04 code to aswss its arability to predict the phenomena in a large Break Loss of Coolant Acddent (LBIDCA). Experiment L2 5 simulated a 200%

guillotine break at the discharge of a primary coolant pump of a four loop

. commercial pressurital water reactor (PW R). One base case calculation and three caws of different nodaliations were carried out. In the base case, the reactor vessel of the LOIT system was modeled by a split downcomer with crossflow junctions and the single core chantiel. De effects of different nodaliations were studied in the area of the. downcomer and the cure. To determine the effectiveness of the nodallution changes and to quantify their effects on the thermal hydraulle responws, nodalization studies were performed for three different cases of reactor vessel modeling: split downcomer modeling without crossflow junction (Case-A), finer axial modeling of core (Case B), and two core channel modeling (Case C). Another calculation was executed for a sensitivity study using an updated version of RELAPS/ MOD 2 Cycle 104.

A split downcomer with one crossflow Junction and two cure channels was found to be effective in describing the ECC bypass and hot channel imhavior.

The updated version was found to-be effective in overcoming the cede defidendes in interfadal friction and reflood quenching.

Publicatlast Date:

Apnl 1990 Prepared by:-

Young Seok Bang, Sang Yong Lee, Hho Jung Kim Casstractor:

Nudear Safety Center, Korea Advanced Ertergy Rewatch Institute, P. O, Box 7,

- Daeduk Danji, Daejon, Korea Prepared for:

NRC Office of Nudear Regulatory Research Keywords:

- international cooperation, Korean organintions, computertral simulahon, IWR type reactors, LOIT reactor, sensitivity analysis, loss of ecolant 120

.. _ ~ _ _ _.... - _ - _ _ _. _ _ _. _, _ _ _ _ ~ - _...... _. _... _ _. _. ~ _

APPENDIX A: Index by NUREC Report Number Retsort Number j'y NUREC 1150, Vol. I 1

NUREC 1150, Vol. 2 2

NUREC-1266, Vol. 4 3

NUREG-1272, Vol. 4, No.1 4

NUREG-1420 5

NUREG/CP 0105, Vo!.1 6

NUREC/CP41105, Vol. 2 7

NUREG/CP 0105, Vol. 3 8

NUREC/CP-0110 9

NUREC/CP 0113 10 NUREC/CR 2331, Vol. 9, No. 3 11 NUKEC/CR 2331, Vol. 9, No. 4 12 NUREG/CR4214, Rev.1, Part i 13 NUREC/CR-4469, Vol.10 14 NUREC/CR4550, Vol. 3, Rev.1 Part 1 15 NUREC/CR-4550, Vol. 3, Rev.1, Part 3 16 NUREC/CR-4550, Vol. 4, Rev.1, Part 3 17 NUREC/CR 4550, Vol. 5, Rev.1, Part 1 18 NUREC/CR4551c Vol. 2, Acv.1, Part 1 19 NUREC/CR4551, Vol. 2, Rev.1, Part 7 20 NUREC/CR4551, Vol. 3, Rev.1, Part 1 21 NUREC/CR4551, Vol. 3, Rev.1, Part 2 22 NUREC/CR-4551, Vol. 4, Rev.1, Part 1 23 NUREC/CR-1551, Vol. 4, Rev.1, Part 2 24 NUREC/CR4551, Vol. 5, Rev.1, Part 1 25 NUREC/CR-4551, Vol. 5, Rev.1, Part 2 26 NUREC/CR4551, Vol. 6, Rev.1, Part 1 27 NUREC/CR-4551, Vol. 6, Rev.1, Part 2 28 NUREC/CR 4554, Vol. 6 29 NUREC/CR-4554, Vol. 7 30 NUREC/CR4624, Vol. 6 31 NUREC/CR4639, Vol.1, Rev.1 32 NUREC/CR4639, Vol. 4, Rev. I.

33 NUREC/CR4639, Vol. 5, Part 1, Rev. 3 34 NUREC/CR4639, Vol. 5, Part 2, Rev. 3 35 NUREC/CR4639, Vol. 5, Part 3, Rev. 3 36 NUREC/CR4668 37 NUREC/CR-1691, Vol.1 M

A-1

t Report Number Tm NUREG/CR-4691, Vol. 2 39 NUREC/CR4691, Vol. 3 40 NUREC/CR4735, Vol. 6 41 NUREG/CR4753, Vol. 3 42 NUREC/CR4816 43 NUREC/CR4840 44 NUREC/CR-4908 45 NUREC/CR 5111 46 NUREG/CR 5213, Vol.1 47 48 NUREC/CR 5213, Vol. 2 NUREG/CR 5229, Vol. 2 49 NUREG/CR 5253 50 NUREC/CR 5254 51 NUREG/CR 5256 52 NUREC/CR 5262 53 NUREC/CR 5273, Vol. 4 54 NUREG/CR 5316 55 NUREG/CR 5366 56 NUP.EG/CR 5376 57 NUREG/CR 5377 58 NUREG/CR 5393 59 60 NUREC/CR FF,98 NUREG/CR 5404, Vol.1 61 NUREC/CR 5403 62 63 NUREG/CR-5409 NUREC/CR 5119 64 65 NUREC/CR-5121 NUREC/CR 5124 66 NUREG/CP 4138 67 NUREC/CR 5147 68 NUREG/CR-5149 69 NUREC/CR 5153, Vol. 5 70 NUREC/CR 5468 71 NURFG/CR-5173 72 NUREG/CR-M75 73 NUREG/CR-5476 74 NUREC/CR-5177 75 76 NUREG/CR stb 0 77 NUREC/CR-5506 78 NUREC/CR 5510 79 NUREG/CR-5512 A2

Report Number Pace NUREG/CR 5514 80 NUREG/CR-5517 81 NUREG/CR 5521 82 NUREC/CR-5523 83 NUREG/CR 5527 84 NUREC/CR 5528 85 NUREC/CR-5530 86 NUREG/CR-5542 87 NUREG/CR 5545 88 NUREG/CR 5547 89 NUREG/CR 5548 90 NUREG/CR-5553 91 NUREC/CR-5557 92 NUREC/CR-5572 93 NUREG/CR-5573 94 NUREG/CR 5575 95 NUREG/CR 55S4 96 NUREC/CR 55S6 97 NUREG/CR 5588, Vol.1 98 NUREG/CR-5588, Vol. 2 99 NUREG/CR 5588, Vol. 3 llX)

NUREG/CR-5590 101 NUREG/CR-5602 102 NUREG/CR-5605 103 NUREG/CR 5607 104 NUREG/CR-5622 105 NUREC/CR 5644 106 NUREG/CR SM9, Vol. I 107 NUREG 'CR-5649, Vol. 2 108 NUREG/ TR-5653 109 NUREG/Ct5659 110 NUREG/lA.9011 111 NUREG/lA-012 112 NUREC/lA-0013 113 NUREG/IA-0018 114 NUREG/M 0019 115 NUREG/lA-0021 116 NUREG/lA-0022 117 NUREG/lA-0030 118 NUREG/lA-0031 119 NUREG/IA-0032 120 A-3

APPENDIX B: Index by Software Identification l

Software Identificatio_q h

ABAQUS 74 Balance of Plant Database 105 BARRIER 87 i

CA_ TEST 96 CARES 98,99,100 CAUSMO 42 CES 47,48 COMMIX 1C 107,108 COMPBRN 44, 75 CONTAIN 97 CONTAIN DCH 97 Control Room Habitability System 110 DAS 49 DOT 4.3 86 DRFMODX 37 EQ3/6 41 FITEQL 89 CEMINI 106 Cencral 1,2,3,4,5,6,7, 8, 9, 10, 11, 12, 16, 17, 19, 21, 22, 23, 24, 25, 26, 27, 28, 52, 59, 60, 63, 66, 67, 69, 70, 76, 3

77, 79, 90, 91, 95, 102 HECTR 101 HTAS2 56 Human Error Database 84 IMPAC7S-BRC2 81 IRRAS2.0 46 LAPUR 65,103 las Cruces Trench Site Database 104 LEPRICON 86 LHS 15,18,44,82 LOC 42 MACCS 13,20,38,39,40, 57, 58, 85 B-1

Software Identification Pare MARC 62 MARCH 100 MARCH 3 31 MATPRO 54 MELPROG 55 MORECA 80 MRR Databaw 45 NPRDS Databaw 14 NUCLARR 32, 33, 34, 35, 36 OCA.P 72 OPUS 73 ORNL Failure Record Database 61 PANDORA 41 PARTITION 50 PIRR Database 45 PR-EDB 43 PRAAGE IA 64 PRAMis 53 QADS 71 RELAPS 92 RELAPS/ MOD 2 94, 112, 113, 114.

116, 118, 119, 120 SAM 42 SCANS 29,30 SCDAP/RELAP5 68 SETS 15,18 SLIM-MAUD 93 STCP 31 SWIFT 82 TEMAC 15,18,44 TIRCALEX 78 TRAC PF1/ MODI 51, 111, 115, 117 UNSAT-H 83 VICTORIA 88 WGEN 83 B-2

APPENDIX C: Index by Contractor lleport Number Report Number

fag, ANL-90/33, Vol.1 107 ANIA0/33, Vol. 2 108 BMI 2139, Vol. 6 31 BNL-NUREG-51454, Vol. 9, No. 3 11 DNL-NUREG 51454, Vol. 9, No. 4 12 BNL-NUREG52168 51 BNL-NUREG-52212 64 BNL-NUREC-52226 9

BNL-NUREC-52228 84 BNL-NUREG-52236 93 BNL NUREG-52237 94 BNL-NUREG52241, Vol.1 98 BNL-NUREG-52241, Vol. 2 99 BNL NUREG-52241, Vol. 3 100 ECG-2458, Vol.1, Rev.1 32 ECC-2458, Vol. 4, Rev. 2 33 ECG2458, Vol. 5, Part 1, Rev. 3 34 ECC-2458, Vol. 5, Part 2, Rev. 3 35 ECG2458, Vct. 5, Part 3, Rev. 3 36 EGG 2535 46 EGG 2555, Vol. 4 54 ECG2566 57 ECC-2574 68 ECG 2577, Vol. 2 49 ECC 2593 85 i

EGG-2597 87 ECG2599 92 ECC-2602 is EGG 2606 102 LA 11f47 MS ORNL4566, Vol.1 61 ORNL/CSO/TM-267 56 ORNL/CSD/TM 270 71 ORNL/TM 10328 43 ORNL/TM 11267 63 ORNL/Th'.1285 65 ORNL/TM 113%

69 ORNL/TM 11??9 70 C-1

Report Number l'agt 72 ORNL/TM 11450 80 ORNL/TM 11451 86 ORNL/TM-11476 91 ORNL/TM 11505 96 ORNL/TM 11575 103 ORNL/TM 11621 14 PNL 5711, Vol.10 45 PNL-6196 79 PNL-7212 89 PNL-7239 90 PNL-7235 83 PNL 7356 109 PNL-7476

'/tl49/1148 105 78 iAIC-89/1,1 110 5AIC 00/1054 13 1AND85-71B. Rev.1, Part i 37

,, iAND861030 19 AND86-1309. Vol. 2, Riv.1, Part 1 g

AND664309, Vol. 2, Kev.1, Part 7 20 SANDS 6-130), Vol. 3, Riev.1, f' art 1 21 22 SAND 3o.1309, Vd. 3, Rev.1, Par 2 23 SAND 86-1309, Vol. 4, Rev.1, Part 1 24 SAND 861309, Vol. 4, Rev.1, Part 2 25 SAND 861309, Vol. 5, Rev.1, Part 1 26 5AhD861309, Vot 5, Rev.1, Part 2 27 SAND 861309, Vol. 6, Rev.1, Part 1 28 SANDB61309, Vnt 6, Rev.1, P..-t 2 38 SAND 861562, Vol.1 39 WD86-1562, Vol. 2 40

!iAND861562, Vol 3 15 3AND86-2061, Vol 3 Rev.1, Part 1 16 SANDS 6-2084, Vol. 3, Rev.1, Part 3 SAND 86 2084, Vol. 4, Rev.1, Part 3 17 18 SAND 862084, Vol. 5, Rev.1, Part 1 50 SAND 88-2940 52 SAND 88-3020 53 SAND 88-3093 44 SANDS 8 3102 55 SAND 88-3476 SAND 89-1432 59 C-2

Iteport Number fagt SANDS 91557 60 SANDS 9-1650 62 SAND 89-2509, Vol. 5 70 SAND 89-2603 74 SAND 89-XVO 81 SAND 9M127 82 SAND 90-0756 88 SANDWk1102 97 SAND 90 7080 101 UCID-20674, Vol. 6 29 UCID-20674, Vol. 7 30 UCID 21831 77 UCRl ID 1N&ts 106 C-3

ma m Al'I'ENDIX D: Index by Keyword Keyword NUnt'G Report Numt cr AC losses NUREC/CH 5557 axidents NUREG 1150, Vol.1 NUREG 1150, Vol. 2 NUREG 1420 NUREC/CR-4214, Rev.1, Part i NUREC/CR4550, Vol. 3, Rev.1, Part 3 NUREC/CR-4550, Vol. 4, Rev.1, Part 3 NUREC/CR4551, Vol. 2. Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR4551, Vol. 3, Rev.1, Part 2 NUREC/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR-4551, Vol. 4, Rev.1, Part 2 NUltEG/CR-4551, Vol. 5, Rev.1, Part 1 NUREC/CR 4551, Vol. 5, Rev.1. Part 2 NUREG/CR-1551, Vol. 6, Rev.1, Part 1 NUREG/CR-4551, Vol. 6, Hev.1, Part 2 NUREC/CR-4554, Vol. 7 NUREC/CR-4639, Vol. 5, Part 3, Rev. 3 ACT codes NURI:G/CR 5409 ADJEN codes NUREC/CR 5256 ADJMiG codes NUREG/CR 5393 adsorption NUREG/CR 5475 NUREC/CR 5547 after heat NUREC/CR 5545 af ter heat remoyal NUREC/CR-5514 aging NUREC/CP-0105, Vol. 3 NUREG/CR 2331, Vol. 9, No. 3 NUREG/CR 2331, Vol. 9, No. 4 NUREG/CR-54GI, Voi. !

NUREG/CR-5419 NUREC/CR 5510 air flow NUREC/CR 5419 ALARA NUREC/CP-0110 NUREG/CR 2331. Vol. 9, No. 3 NUREC/CR-2331. Vol. 9, No. 4 aquatic ecosystems NUREC/CR-5377 ARANO codes NUREG/CR-5377 artificial intelligence NUREC/CR-5213, Vol.1 NUREC/CR-5213, Vol. 2 D-1

i Keyword NURf'C Reriort Number atmospheric circulation NUREG/CR4691, Vol.1 NUREG/CR4691, Vol. 2 NUREG/CR4691, Vol. 3 at nospheric precipitations NUREG/CR M75 ATWS NUREG/CR-5573 austenitic steels NUREG/CR 4735, Vol. 6 NUREG/CR4908 audliary water systems NUREC/CR 54N, vol. I basalt NUREC/CR-4735, Vol. 6 NUREG/CR 5521 Basic NUREC/CR-5584 behavior NUREC/CR 5572 blackouts NUREC/CR4550, Vol. 3, Rev.1, Part 1 NUREG/CR.4550, Vol. 5, Rev,1, Part 1 NUREG/CR4624, Vol. 6 NUREG/CR-M47 blowdown NUREG/CR 5254 BLT codes NUREG 1266, Vol. 4 NUREG/CR M53, Vol. 5 BNL Engineering Plant Analyzer codes NUREC/CR 2331, Vol. 9, No. 3 NUREC/CR-2331, Vol. 9, No. 4 les NUREG/CR-5424 DOC;&

boro'&

NUREC/CR-5573 borosilli te glass NUREG/CR 4735, Vol. 6 BOSOR codes NUREC/CR-4554, Vol. 6 BR 3 reactor NUREG/CR 5480 bromides NUREG/CR Sm7 building materials NUREC/CR 5545 buildings NUREC/CR 5512 NUREG/CR-5588, Vol.1 NUREG/CR-5588, Vcl. 2 NUREG/CR-5588, Vol. 3 buildup NUREG/CR-5468 BWR type reactors NUREC/CR 2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR4551, Vol. 6, Rev.1, Part 1 NUREC/CR-4551, Vol. 6, Rev.1, Part ?

D2

Keyword d'JJHi?G Report Number DWR type rvactors (continual)

NUREC/CR-4624, Vol. 6 NUREG/CR 5421 NUREG/CR 5449 NUREC/CP ~ *.0 NUPE/CR >a28 NUREG/CR 5573 NUREC/CR 5605 NUREG/CR 5653 BWRLTAS codes NUREG/CR-5528 BWRSAR codes NUREC/CP-0105, Vcl. 2 NUREC/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR 5528 by-prtxiucts NUREC/CR 2331, Vol. 9, No. 3 CAITA codes NUREG/CR 5510 CALACT codes NUREC/CR 5409 Canada NUREC/CR-4753, Vol. 3 carbon 14 NUREC/CR-4735, Vol. 6 CARES codes NUREC/CP-0113 NUREG/CR-2331, Vol. 9, No. 3 cavities NUREG/CR 5514 :

CES codes NUREG 1266, Vol. 4 NUREG/CP-0105, Vol.1 Charpy test NUREC/CR 5409 CHEM codes NUREG/CR 5659 chernical properties NUREC/CR 5273, Vol. 4 chernical reactions NUREG/CR 5545 NUREG/CR-5548 chemical spills NUREC/CR 5659 CHEMIST codes NUREC/CR 5548 CHYMES codes NUREG/CP-0105, Vol. 2 CLAD codes NUREG/CP-0105, Vol.1 cladding NUREG/CR-5254 CLASSI codes NUREC/CP-0105, Vol. 2 NUREC/CR-4550, Vol. 3, Rev.1, Part 3 NUREC/CR-4550, Vol. 4, Rev.1, Part 3 NUREG/CR-4840 clitnate models NUREC/CR-5523 D-3

Keyword NUREG Report Number COURA codes NUREG/CP-0105, Vol. 3 COBRA-NC codes NUREG/CP41105, Vol. 2 combustion NUREG/CR 5586 NUREG/CR 5590 comparative evaluations NUREG/CR 5438 COMPBRN codes NUREG 1150, Vol. 2 NUREG/CR-4550, Vol. 3, Rev.1, Part 3 NUREG/CR-4550, Vol. 4, Rev.1, Part 3 compliance NUREC/CR 5521 compressed air NUREG/CR-5419 computer graphics NUREC/CR 5111 computerized simulation NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-5254 NUREG/CR 5376 -

NUREG/CR 5573 NUREG/lA-0011 NUREG/lA 0012 NUREG/lA 0013 NUREG/lA-0018 NUREG/lA-0019 NUREG/IAUn1 NUREG/lA 0022 NUREC/lA-0030 NUREG/lA-0031 NUREG/lA-0032 concretes NUREC/CR-5542 CONHAB codes NUREG/CR 5659 CONTAIN codes NUREG 1150, Vol.1 NUREG 1150, Vol. 2 NUREG 1266, Vol. 4 NUREG/CP 0105, Vol. 2 NUREC/CR-5528 NUREG/CR 5590 containment NUREG 1150, Vol.1 NUREG-1150, Vol. 2 NUREG/CR 5447 NUREG/CR 5521 NUREG/CR-5528 NUREG/CR-5575 containment buildings NUREG/CR 5405 NUREC/CR 5476 NUREC/CR-5590 D-4

i i

l Keyword NUREG Report Number i

containment systems NUREC/CR 2331, Vol. 9, No. 3 I

NUREC/CR.2331, Vol. 9, No. 4 NUREG/CR 4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 2 NUREC/CR-4551, Vol. 4, Rev.1, Part 1 NUREC/CR-4551, Vol. 4, Rev.1, Part 2 NUREC/CR-4551, Vol. 5, Rev.1, Part 1 NUREC/CR-4551, Vol. 5, Rev.1, Part 2 NUREC/CR-4551, Vol. 6, Rev.1, Part 1 NUREC/CR4551, Vol. 6, Rev.1, Part 2 NUREG/CR 5586 NUREG/CR 5602 contamination NUREG/CR 5512 control rooms NUREC/CR 2331, Vol. 9, No. 4 NUREC/CR 5572 NUREC/CR 5659 convection NUREG/CR 5366 coolants NUREG/CR 5573 COPOX R codes NUREG/CR-4668 copper NUREG/CR-4735, Vol. 6 NUREG/CR 5584 CORCON codes NUREG 1266, Vol. 4 NUREG 1420 NUREC/CP-0113 NUREG/CR-4624, Vol. 6 NL' REC /CR 5528 corium NUREC/CR 2331, Vol. 9, No. 4 CORMLT codes NUREC/CR 5316 CORMLT/PSAAC codes NUREG-1150, Vol. 2 NUREG/CR 4551, Vol. 2, Rev.1, Part I corrosion NUREC/CR-4735, Vol. 6 NUREG/CR-5542 CORSOR codes NUREG/CR-4624, Vol. 6 CORSOR M codes NUREC/CR-5480 cost benefit analysis NUREG/CR-4691, Vol.1 NUREC/CR-4691, Vol. 2 NUREC/CR-4691, Vol. 3 CRAC codes NUREG-1420 NUREC/CR 4214, Rev.1, Part i NUREG/CR 5377 crack propagation NUREC/CR 5473 D-5

Keyword NUHFG Retsort Numlier cracking NUREC/CR 5542 eracks NUREC/CR-4908 NUREC/CR 5409 NUREC/CR 5584 CRAC2 codes NUREG/CP4105. Vol.1 NUREG/CR 5377 NUREC/CR-5528 CREAMS codes NUREC/CR-5523 CREATE codes NUREG-1266, Vol. 4 CRil codes NUREC/CR 5659 CRiiPLCTTC codes NUREC/CR 5659 CRiiPLCTTR codes NUREC/CR,59 criticality NUREG/CR 5622 NUREC/CR 5653 danuge NUREG/CR4668 NUREG/CR 5316 NUREC/CR-5602 NUREG/CR-5653 data baw msmgement NUREC/CR4639, Vol.1, Rev.1 NUREG/CR4639, Vol. 4, Rev. 2 NUREG/CR4639, Vol. 5, Part 1, Rev. 3 NUREC/CR 4639, Vol. 5, Part 2, Rev. 3 NUREG/CR4639, Vol. 5, Part 3, Rev. 3 NUREG/CR-4735. Vol. 6 NUREC/CR-4816 NUREC/CR 5404, Vol.1 NUREC/CR 5438 NUREG/CR 5607 NUREG/CR 5622 data compilation NUREC/CR4639, Vol. 5, Part 3, Rev. 3 NUREG/CR-4816 data pmwssing NUREG/CR4639, Vol.1, Rev.1 NUREC/CR4639, Vol. 4, Rev. 2 NUREG/CR4639, Vol. 5, Part 1, Rev. 3 NUREC/CR4639, Vol. 5, Part 2, Rev. 3 DEBRIS codes NUREC/CP4113 decay NUREC/CR 5421 decision making NUREC/CR 5477 decommissioning NUREC/CR-5512 decomposition NUREG/CR 5542 D-6

Keyword NUREG Hrport Number deformation NUREC/CR 1554, Vol. 6 delayed radiation effects NUREG/CR 4214, Rev.1, Part i NUREC/CR 5253 r

deposition NUREC/CR 4551, Vol. 2, Rev.1, Part 7 depressurization NUREC/CR4624, Vol. 6 NUREG/CR 5447 NUREC/CR 5573 NUREG/CR-5575 design basis accidents NUREC/CR 2331, Vol. 9, No. 4 detonations NUREG/CR-5586 DISPERoE codes NUREC/CR MS3, Vol. 5 displacement rates NUREC/CR M09

. display devices NUREG/CR 2331, Vol. 9, No. 4 NUREG/CR 5572 distribution functions NUREC/CR N24 dose equivalents NUREG/CR 4214, Rev.1, Part I NUREC/CR 5512 dose limits NUREC/CR 2331, Vol. 9, No. 3 NUREC/CR 2331, Vol. 9, No. 4 dose rates NUREC/CP-0110 NUREC/CR 5468 dose-rtsponse relationships NUREC/CR-4214, Rev.1, Part i DOSI ANA codes NUREC/CP 0110 DOSI ECO codes NUREC/CP 0110 dosimetry NUREC/CR-4551, Vol. 2, Rev.1, Part 7 NUREC/CR-4691, Vol.1 NUREG/CR-4691, Vol. 2 NUREG/CR-4691, Vol. 3 NUREG/CR 5409 NUREC/CR M53, Vol. 5 NUREG/CR-5530 DOT-IV codes NUREG/CR 5449 DOTSOR codes NUREC/CR 5449 DOTSYN codes NUREC/CR-M49 drag

- NUREU/lA-0022 drainage NUREC/CR-5523 D-7

Keyword NUREG Report Numb.n DRALIST codes NUREC/CR 5512 ductility NUREC/CR-5473 DYDAS codes NUREC/CP-0110 Dynamic Dose Tracking codes NUREG/CP-0110 dynamics NUREC/CR 5421 DYNA 2D codes NUREC/CR-4554, Vol. 7 carly radiation effects NUREG/CR-4214, Rev.1, Part i NUREC/CR 5253 carth atmosphere NUREC/Ci A551, Vol. 2, Rev.1, Part 7 carthquakes NUREC-1266, Vol. 4 NUREG/CP-0105, Vol. 2 NUREC/CR-4753, Vol. 3 NUREC/CR-4840 NUREC/CR-5477 ECCS NUREC/CR-4624, Vol. 6 NUREG/CR 5254 economics NUREC/CR-4551, Vol. 2, Rev.1, Part 7 NUREC/CR4691, Vol.1 NUREC/CR-4691, Vol. 2 NUREG/CR-4691, Vol. 3 ccosystems NUREC/CR-M75 eddy currents NUREC/CR 5553 electric coils NUREG/CR-5553 electrical faults NUREG/CR4639, Vol. A Part 3, Rev. 3 cmbrittlement NUREC/CR-4816 NUREC/CR 5644 emergency plans NUREC/CR-4551, Vol. 2, Rev.1, Part 7 NUREC/CR 4624, Vol. 6 NUREG/CR4691, Vol.1 NUREC/CR4691, Vol. 2 NUREG/CR-4691, Vol. 3 NIJREG/CR-5213, Vol.1 NUREC/CR-5213, Vol. 2 EMPIRICAL codes NUPEC/CR 5424 engiwered safety systems NLPJ.C 1272, Vo'. 4, No. I environmental exposure pathway NUREC/ cst $51, Vol. 2. Rev.1, Part 7 N'JREC/CPc5377 D-8 I

Keyword NUREG Report.

aber environmental transport NUREG/CR-M53, Vol. 5 NUFEG/CR-5475 NUR2G/CR-5523 EPA codes i4UREC/CP-0113 EQUILIB codes NUREC/CR 5548 EQ3/EQ6 codes NUREG/CR-5548 errors NUREC/CR 5213, Vol.1 NUREG/CR-5212, Vol. 2 NUREC/CR-5438 NUREC/CR 5527 NUREG/CR-!572 ET LOAD codes NUREG/CR 5575 ETA-Il codes NUREG/CR-5575 cvacuation NUREG/CR4551, Vol. 2, Rev.1, Part 7 NUREG/CR4691, Vol.1 NUREC/CR4691, Vol. 2 NUREG/CR4691, Vol. 3 EVNTRE codes NUREG-1150, Vol. 2 NUREG/CR 4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Pev.1, Part 2 NUREG/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR-4551, Vol. 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 6, Rev.1, Part 1 NUREC/CR-5262 NUREG/CR-5575 NUREC/CR-5602 excursions NUREC/CR-5573 NUREG/CR-5653 operimental data NUREG/CR-5607 expert systems NUREG/CR-M24 NUREG/CR-5572 EXTLHS codes NUREG/CR4551, Vol. 3, Rev.1, Part 2 NUREC/CR4551, Vol. 4, Rev.1, Part 2 NUREG/CR4551, Vol. 5 Rev.1, Part 2 NUREC/CR-4551, Vol. 6, Rev.1, Part 2 EXTSEQ6 codes NUREG/CR-4551, Vol. 5, Rev.1, Part 2 failure mode analysis NUREC/CR 4550, Vol. 3, Rev.1, Part 1 NUREG/CR.-1550, Vol. 5, Rev.1, Part 1 NUREG/CR-5111 NUREC/CR-5419 NUREC/CR-5528 NUREG/CR-5575 NUREG/CR-5602 D-9

Keyword NUREG Report Nudg, fauures NUREG-1266, Vol. 4 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR-4554, Vol. 7 NUREC/CR4639, Vol.1, Rev.1 NUREC/CR4639, Vol. 4, Rev. 2 NUREC/CR 4639, Vol. 5, Part L Rev. 3 NUREG/CR 4639, Vol. 5. Part.'. 'ev,3 NUREG/CR 4639, Vol. 5, Part 3, Kev. 3 NUREG/CR-5404, Vol.1 NUREC/CR-5419 NUREG/CR-5506 NUREG/CR 5510 NUREC/CR-5527 NUREG/CR-5572 NUREG/CR-5653 FASTPATH codes NUREC/CR 5548 fault tree analysis NUREG/CR-4840 NUREG/CR-5111 feedback NUKEG/CR-5421 feedwater NUREG/CR-N04, Vol.1 FEF codes NUREG-1272, Vol. 4, No.1 FEMWASTE mdes NUREG-1266, Vol. 4 FEMWATER codes NUREG-1266, Vol. 4 NUREG/CR-5453, Vol. 5 field tests NUREG/CR-5548 NUREG/CR-5607

- finite element method NUREG/CR 4554, Vol. 7 NUREG/CR-5405 NUREG/CR N76 fire hazards NUREG/CR 4550, Vol. 3, Rev.1, Part 3 NUREC/CR 4550, Vol. 4, Rev.1, Part 3 fires NUREG/CR 4840 NUREG/CR-5477 fission product release NUREG-1420 NUREG/CR-4551, Vol. 2, Rev.1, Part 7 NUREC/CR 5480 NUREG/CR-5528 NUREG/CR-5545 fission products NUREG/CR-5316 NUREG/CR 5480 NUREC/CR-5545 flame propagation NUREG/CR-5590 D-10

F 3

Keyword NUREG Report Number

= floods NUREG/CR-4840 NUREC/CR M77

- How models NUREG/CP-0105, Vol. 2 NUREG/lA 0011 NUREG/lA-0019 NUREG/lA-0022 Quid now NUREG/CR W9, Vol.1 NUREG/CR-W9, Vol. 2 fluid-structure interactions NUREC/CR W9, Vol.1 NUREG/CR-W9, Vol. 2 FLXPRO codes NUREC/CR-5409 food chains NUREG/CR-5377 NUREG/CR 5453 Vol. 5 forging NUREG/CR-4816 Fortran NUREG/CR-5376 fracture mechanics NUREC/CR-4469, Vol.10 NUREC/CR 5473 fracture properties-NUREG/CR-5584 fragmentation NUREG/CR-5316 frequer.cy analysis NUREG/CR-4840 fuel pools NUREG/CR-2331, Vol. 9, No. 4 fuel rods NUREG/CR-4668 NUREG/CR-5316

- fuel-cladding interaction NUREG/CR-4668 gamma radiation NUREC/CR-5468 NUREC/CR-5517 gas cooled reactors NUREG/CR-5514 gas flow NUREG/CR 5316 gas spills NUREC/CR-5659 GEMINI codes NUREG/CR-5506 genetic radiation effects NUREC/CR-4214, Rev.1, Part I GENII codes NUREC/CR 5453, Vol.-5 NUREG/CR-5512 '

GEOCHEM codes NUREC/CR-5548

- geochemistry NUREC/CR-5548 D-11

Keyword NUREG Report Number CCSOR codes NUREC/CR4551, Vol. 6, Rev.1, Part 1 NUREC/CR4551, Vol. 6, Rev.1, Part 2 GGUFUN codes NUREG/CR4551, Vol. 6, Rev.1, Part 2 Grand Gulf 1 reactor NUREG/CR-455), Vol. 6, Rev.1, Part 1 NUiEG/CR4551, Vol. 6, Rev.1, Part 2 GRESS codes NUREG/CR 5256 NUREC/CR-5393 ground motion NUREC/CR 5588, Vol.1 NUREG/CR 5588, Vcl. 2 NUREG/CR 5588, Vol. 3 ground water NUREC/CR 5453, Vol. 5 NUREC/CR-5523 NUREC/CR-5607 hazards NUREG/CR4840 NUREG/CR-5622 health NUREG/CR4214, Rev.1, Part I health hazards NUREG/CR4691, Vol.1 NUREC/CR-4691, Vol. 2 NUREG/CR-4691, Vol. 3 NUREG/CR-5253 heat transfer NUREC/CR 2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR4554, Vol. 6 NUREC/CR 5254 NUREG/CR-5366 NUREC/CR-5619, Vol.1 NUREG/CR 5649, Vol. 2 heating NUREG/CR-5447 HEATING codes NUREC/CR-5366 HECTR codes NUREG-1150, Vol. 2 NUREG/CP-0105, Vol. 2 NUREC/CR-5447 HELP codes NUREG/CR-5523 high pressun: coolant injection NUREG/IA 0031 high temperature NUREG/CR-5514 high-level radioactive wastes NUREG/CR-1735, Vol. 6 NUREG/CR-5256 NUREC/CR-5393 NUREG/CR-5398 NUREG/CR-5521 D-12

Eeyword NUREG Reriott Number HIPA codes NUREG-1266, Vol. 4 NUREG/CP-0105, Vol. 3 H'ICR type reactors NUREG/CR-2331, Vol. 9, No. 4 human factors NUREG-1266, Vol. 4 NUREG/CP-0105, Vol.1 NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR 2331, Vol. 9, No. 4 NUREG/CR-4639, Vol.1, Rev.1 NUREG/CR-4639, Vol. 4, Rev. 2 NUREG/CR-4639, Vol. 5, Part 1, Rev. 3 NUREG/CR-4639, Vol. 5, Part 2, Rev. 3 NUREG/CR-4908 NUREC/CR-5213, Vol.1 NUREC/CR-5213, Vol. 2 NUREC/CR-5438 NUREC/CR 5527 NUREG/CR 5572 human factors engineering NUREG/CR-5572 human populations NUREG/CR 5517 hydraulic conductivity NUREG/CR-5607 hydrodynamics NUREC/CR-5421 hydrogen NUREC/CR-5586 NUREG/CR-5590 hydrogen production NUREC/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR-4668 ice condensers NUREC/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR-4551, Vol. 5, Rev.1, Part 1 NUREC/CR 4551, Vol. 5, Rev.1, Part 2 NUREC/CR 5586 NURSG/CR-5602 ICRP 02 codes NUREG/CR-5659 ICRP.30 codes NUREG/CR-5659 impact strength NUREC/CR-4554, Vol. 6 IMPACIS codes NUREG/CR-5517 IMPACTS-BRC codes NUREC/CR-5517 impedance NUREG/CR-5553 induction furnaces NUREC/CR-5480 D-13

s l

l Keyword NUREG R* port Number information retrieval NUREG/CR-4639, Vol.1, Rev.1 NUREG/CR4639, Vol. 4, Rev. 2 NUREG/CR4639, Vol. 5, Part 1, Rev. 3 NUREG/CR-4639, Vol. 5, Part 2, Rev. 3 ingestion NUREC/CR 5517 inhalation NUREG/CR-5517 inspection NUREG/CR-4908 NUREC/CR 5376 NUREG/CR-5404, Vol.1 intergranular corrosion NUREG/CR 4908 intemational cooperation NUREC/CP4)110 NUREC/lA-0011 NUREC/lA-0012 NUREG/lA-0013 NUREG/lA-0018 NUREG/lA-0019 NUREC/lA-0021 NUREG/IA-0022 NUREG/lA-0030 NUREG/lA-0031 NUREG/lA-0032 irradiation NUREG/CR-4735, Vol. 6 NUREG/CR-5584 irradiation procedures NUREC/CR-2331, Vol. 9, No. 3 NUREC/CR-2331, Vol. 9, No. 4 IRRAS codes NUREG 1266, Vol. 4 IRRAS1.0 codes NUREC/CR-5111 IRS codes NUREG-1272. Vol. 4, No.1 ISOE codes NUREG/CP-0110 ISOSHLD codes NUREG/CR-5512 J-TRAC codes NUREG/CP-0105, Vol. 3 knowledge base NUREG/CR-5424 Korean organizations NUREG/lA-0030 NUREC/lA-0031 NUREG/IA-0032 KORI-1 reactor NUREG/lA-0030 LAPUR codes NUREG/CP-0113 leaching NUREG/CR-4735, Vol. 6 NUREG/CR-5542 D-14

Keyword NUREG Report Number LEPRICON codes NUREG/CR M49 LER codes NUREG/CR-5419 LET NUREG/CR-4214, Rev.1, Part i LHS codes NUREG/CR4550, Vol. 4, Rev.1, Part 3 NUREG/CR-4551, Vol. 3, Rev.1, Part 2 NUREG/CR4551, Vol. 4, Rev.1, Part 2 NUREG/CR-4551, Vol. 5, Rev.1, Part 2 NUREG/CR-4551, Vol. 6 Rev.1, Part 2 NUREC/CR-5256 NUREG/CR-5262 NUREC/CR-5393 licensing NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-5398 liners NUREC/CR-5229, Vol. 2 LMFBR type reactors NUREC/CR-2331, Vol. 9, No. 3 NUREC/CR-2331, Vol. 9, No. 4 Load Combination codes NUREG/CR-55%

LOFT reactor NUREC/lA-0011 NUREG/lA-0012 NUREG/IA-0013 NUREG/lA-0018 NUREG/!A-0019 NUREC/lA-0021 NUREC/lA-0022 NUREG/lA-0031 NUREC/IA-0032 LOSP codes NUREG/CR4551, Vol. 5, Rev.1. Pa t 2 NUREG/CR4551, Vol. 6, Rev.1, Part 2 loss of coolant NUREG/CR-4550, Vol. 3, Rev.1, Part 1 NUREG/CR4550, Vol. 5, Rev.1, Part 1 NUREG/CR-4624, Vol. 6 NUREC/CR 5254 NUREG/CR-5447 NUREC/CR-5557 NUREC/IA-0012 NUREC/IA-0013 NUREG/!A-0018 NUREG/lA-0019 NUREG/IA-0021 NUREG/IA-0022 NUREC/IA-0031 NUREG/IA-0032 low-level radioactive wastes NUREG/CR-5229, Vol. 2 NUREG/CR-5453, Vol. 5 NUREG/CR-5517 NUREC/CR-5523 NUREG/CR-5542 i

D-15

Keyword NUREG Report Number low-level radioactive wastes (continued)

NUREG/CR-5547 NUREG/CR-5548 LSL M2 codes NUREC/CR-M09 MAAP codes NUREG-1150, Vol. 2 NUREC/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR-5528 MACCS codes NUREC 1150, Vol.1 NUREG-1150, Vol. 2 NUREG-1266, Vol. 4 NUREG-1420 NUREG/CP-0105, Vol.1 NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR-4551, Vol. 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 6, Rev.1, Part 1 NUREC/CR-5253 NUREG/CR-5262 NUREC/CR 5575 NUREC/CR-5602 MAEROS codes NUREG/CR-4668 NUREG/CR-5393 maintenance NUREG/CR-5510 man-machine systems NUREC/CR-5572 management NUREC/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-5447 MAPPER codes NUREG/CR-5262 MAPPS codes NUREG-1266, Vol. 4 NUREC/CP-0105, Vol.1 MARCH codes NUREG-1150, Vol. 2 NUREG/CP-0105, Vol. 2 NUREG/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR-5528 NUREC/CR-5586 MARCH 2 codes NUREG/CR-4624, Vol. 6 materials testing NUREC/CR-5409 mathematical models NUREC/CR-5393 NUREC/CR-5393 NUREG/CR-5542 maximum credible accident NUREG/CR-2331, Vol. 9, No. 4 MCBETA codes NUREG/CR-5424 D-16

Keyword NUREG Report Number MCDAN codes NUREC/CP-0105, Vol. 2 NUREG/CR-5653 mechanical properties NUREC/CR 4639, Vol. 5, Part 3, Rev. 3 mechanical structures NUREG/CR-5588, Vol.1 NUREG/CR 5588, Vol. 2 NUREG/CR-5588, Vol. 3 MELCOR codes NUREC-1150, Vol.1 NUREG 1150, Vol. 2 NUREG-1266, Vol. 4 NUREG-1420 NUREG/CP-0105, Vol. 2 NUREC/CP-0113 NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR-5377 NUREG/CR-5528 NUREG/CR 5590 MELPROG codes NUREG-1150, Vol.1 NUREG/CR-4551, Vol. 2, Rev.1, Part 1 MELPROG/ TRAC codes NUREG-1266, Vol. 4 meltdown NUREC/CR-2331, Vol. 9, No. 4 NUREC/CR-4668 NUREG/CR-S316 NUREG/CR-5527 MERGE codes NUREG/CR-4624, Vol. 6 MICROMAPPS codes NUREG/CP-0113 MINEQL codes NUREC/CR-5347 NUREC/CR-5548 IVINTEQ codes NUREG/CR-5548 mitigation NUREG/CR-4551, Vol. 2, Rev.1, Part '

NUREG/CR-4691, Vol.1 NUREG/CR-4691, Vol. 2 NURFG/CR-4691, Vol. 3 NUREC/CR-5586 MODEL codes NURcG/CR-4551, Vol. 3, Rev.1, Part 2 NUREC/CR-4551, Vol. 4, Rev.1, Part 2 NUREG/CR-4551, Vol. 5, Rev.1, Part 2 NUREC/CR-4551, Vol. 6, Rev.1, Part 2 monitoring NUREC/CR-5404, Vol. I monoclonal antibodies NUREG/CR-2331, Vol. 9, No. 3 NUREC/CR-2331, Vol. 9, No. 4 D-17 1

Keyword NUREG Report Number Monte Carlo method NUREG/CR 5111 NUREC/CR-5393 NUREG/CR-5523 MORSE codes NUREC/CR 5468 natural convection NUREG/CR-5316 NAUA codes NUREG-1150, Vol. 2 NEA NUREC/CP-0110 NECTAR codes NUREG/CR-5377 neoplasms NUREG/CR-5253 neutron fluence NUREG/CR-M09 NUREG/CR-5530 NUREG/CR-5584 neutron transport theory NUREG/CR-5449 NIKE2D codes NUREC/CR-4554, Vol. 7 NITAWL codes NUREC/CP-0105, Vol. 2 NUREG/CR 5653 nondestructive analysis NUREG/CR-4469, Vol.10 nondestructive testing NUREG/CP-0103, Vol. 3 NUREC/CR44o9, Vol.10 NUREG/CR-4908 notches NUREC/CR-909 NUREG/CR-5584 NPA codes NUPEG/CP-0113 NPRDS codes NUREG/CR-M19 NUREG/CR-5438 NRCPIPE codes NUREG/CP-0113 NUCLARR codes NUREG-1266, Vol. 4

.NUREG/CP-0105, Vol.1 NUREC/CR-5438 nuclear data collections NUREG/CR4639, Vol.1, Rev.1 NUREG/CR-4639, Vol. 4, Rev. 2 NUREC/CR-4639, Vol. 5, Part 1, Rev. 3 NUREG/CR-4639, Vol. 5, Part 2, Rev. 3 nuclear fuel NUREG/CR-4668 NUREC/CR-5273, Vol. 4 NUREG/CR 5480 D-18

Keyword NUREG Reriott Number nuclear power plants NUREG 1150, Vol.1 NUREG-1150, Vol. 2 NUPEG-1266. Vol. 4 NUREG-12*Q, Vol. 4, No.1 NUREG/CP-0105, Vol. 2 NUREC/CP-0105, Vol. 3 NUREG/CP-0110 NUREC/CR4214, Rev.1. Part 1 NUREG/CR-4551, Vol..!, Rev.1, Part 7 NUREC/CR4624, Vol. 6 NUREG/CR-4639, Vol.1, Rev.1 NUREC/CR4639, Vol. 4, Rev. 2 NUREG/CR-4639, Vol. 5, Part 1, Rev. 3 NUREC/CR4639, Vol. 5, Part 2, Rev, 3 NUREG/CR4639, Vol. 5, Part 3, Rev. 3 NUREG/CR-4691, Vol.1 NUREG/CR4691, Vol. 2 NUREC/CR4691, Vol. 3 NUREG/CR4840 NUREG/CR-5213, Vol.1 NUREG/CR-5213, Vol. 2 NUREG/CR-5253 NUREG/CR-5262 NUREC/CR 5419 NUREC/CR-5447 NUREG/CR-5477 NUREG/CR 5506 NUREG/CR-5510 NUREG/CR-5577 NUREG/CR-5588, Vol.1 NUREG/CR-5588, Vol. 2 NUREG/CR-5588, Vol. 3 NUREG/CR-5622 NUCRAC codes NUREC/CR-5377 numerical arulysis NUREG/CR-5649, Vol.1 NUREG/CR-5649, Vol. 2 numerical solution NUREG/CR-5398 OCA-P codes NUREC/CP-0105, Vol. 3 NUREG/CP-0113 occupational exposure NUREC/CP-01')

occupational safety NUREG/CR-2331. Vol. 9, No. 4 OECD NUREG/IA-0011 NUREG/IA-0012 NUREC/lA-0013 NUREG/lA 0018 NUREG/lA-0019 NUREG/lA-0021 NUREG/lA-0022 OKG-3 reactor NUREG/CR-5605 D 19

?

. Keyword --

b'UREG Report Number

- ONSITE/ MAX 11 codes NUREG/CR 5512 ORE codes -

NUREC/CP-0110

' ORIGEN codes NUREG/CR-5468 NUREC/CR-5653 -

ORIGEN2 codes NUREC/CP-0105, Vol.1 NUREG/CR 5480,

ORIGEN2C codes NUREG/CR 5393 oscillations NUREG/CR-5605 oxidation -

NUREG/CR4668 NUREG/CR-5316 NUREG/CR 5480 parametric analysis NUREC/CR-5254 NUREG/CR 5393 L particle resuspension NUREG/CR-5377

. particle size NUREG/CR 5607 PARTITION codes-NUREG-1150, Vol. 2 NUREC/CR4551, Vol. 3, Rev.1, Part 1-NUREG/CR 4551, Vc'. 4, Rev.1, Part 1 NUREG/CR4551, VouS, Rev.1, Part 1 NUREG/CR-4551, Vol. 6, Rev.1, Part 1 NUREG/CR 5262 NUREG/CR 5602 PATHCALC codes NUREC/CR-5548 PATRAN codes

- NUREC/CR-5405 PBSOR codes =

NUREG/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR4551, Vol. 4. Fey! 1. Part 2 PBUFUN codes NUREG/CR4551, Vol. 4, Rev.1,' Part 2 Peach Bottom-2 reactor-NUREG/CR4550, Vol. 4, Rev.1, Part 3 I

NUREG/CR-4551,-Vol. 4, Rev.1, Part 1 NUREG/CR-4551, Vol. 4, Rev.1, Part 2 NUREG/CR-4840

.r performance NUREG-1272, Vol. 4, No.1 NUREG/CP-0105, Vol.1 NUREC/CR4908 NUREG/CR-5438 NUREG/CR 5572 NUREG/CR 5575 NUREG/CR 5586 D-20

. m.

m

Keyword NUREG Reriort Number i

performance testing NUREG/CP-0105, Vol. 3 I

NUREC/CR4469, Vol.10 l

NUREC/CR-5229, Vol. 2 NUREG/CR-5256 NUREC/CR-5393 NUREG/CR-5398 NUREC/CR 5453, Vol. 5 NUREG/CR-5480 NUREC/CR 5514 NUREG/CR 5521 personnel NUREG/CR 5213, Vol. I NUREG/CR-5213, Vol. 2 phase stability NUREC/CR 4735, Vol. 6 PHASE 8 codes NUREC/CR-5512 PHOENICS codes NUREC/CP-0105, Vol. 2 PHREEQE codes NUREC/CR 5548 physical properties NUREC/CR-5273, Vol. 4 pipes NUREC/CP-0105, Vol. 3 NUREG/CR4469, Vol.10 planning NUREC/CP-0105, Vol. 3 plants NUREC/CR-5523 plates NUREC/CR-4554, Vol. 7 NUREG/CR-4816 PLUGM codes NUREG/CP-0105, Vol. 2 PM-ALPHA codes NOREC/CP-0105, Vol. 2 point defects NUREC/CR-5553 point kernels NUREG/CR-5468 Portland cer ent NUREG/CR-5229, Vol. 2 power reactors NUREG/CR4816 NUREC/CR-5273, Vol. 4 PR-EDB codes NUREC/CP-0105, Vol. 3 PRAAGE-IA codes NUREG/CP-0105, Vol. 3 PRAMIS codes NUREC/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR4551, Vol. 4, Rev.1, Part 1 NUREG/CR4551, Vol. 5, Rev.1, Part 1 NUREC/CR4551, Vol. 6, Rev.1, Part 1 pressure NUREC/CR4554, Vol. 6 D-21

Keyword NUREC Rcport Number pressure vessels NUREG/CR4816 NUREG/CR-5409 NUREC/CR-5447 NUREG/CR-M49 NUREG/CR 5473 NUREG/CR-5573 NUREC/CR-SSM NUREG/CR-5644 pressurizers NUREG/CR-5557 pressurizing NUREC/CR-MOS primary coolant circuits NUREG/CP-0105, Vol. 3 probabilistic estimation NUREG-1150, Vol.1 NUREG-1150, Vol. 2 N9 REG-1420 NUREC/C' 0105, Vol. 2 NUREG/CP-0105, Vol. 3 NUREG/CR 2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR4550, Vol. 3, Rev.1, Part 1 NUREC/CR4550, Vol. 3, Rev.1, Part 3 NUREC/CR4550, Vol. 4, Rev.1, Part 3 NUREG/CR-4550, Vol. 5, Rev,1, Part 1 NUREG/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 2 NUREG/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR4551, Vol. 4, Rev.1, Part 2.

NUREC/CR-4551, Vol. 5, Rev.1, Part 1 NUREC/CR4551, Vol. 5, Rev.1, Part 2 NUREG/CR4551, Vol. 6 Rev.1, Part 1 NUREG/CR-4551, Vol. 6, Rev.1, Part ?

NUREC/CR4691, Vol.1 NUREG/CR-4691, Vol. 2 NUREG/CR-4691, Vol. 3 NUREC/CR-4840 NUREG/CR-5111 NUREG/CR-5213 Vol.1 NUREG/CR-5213, Vol. 2 NUREG/CR-5253 NUREG/CR-5262 NUREG/CR-5377 NUREG/CR-M47 NUREG/CR-5477 NUREG/CR-5510 NUREG/CR-5527 NUREC/CR-5572 NUREG/CR-5573 NUREG/CR-5586 probes NUREC/CR 5553 proceedings NUREG/CP-0110 progress report NUREC/CR-2331, Vol. 9, No. 4 D-22

Keyword NUREG Report Number I

PRPOST codes NUREC/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR-4551, Vol. 5, Rev.1, Part 1 NUREC/CR-4551, Vol. 6, Rev.1, Part 1 PSD codes NUREC/CR-2331, Vol. 9, No. 3 PSTEVNT codes NUREG-1150, Vol. 2 NUREC/CR-5602 pumps NUREG/CR 2331. Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-4624, Vol. 6 NUREG/lA-0019 PWR type reactors NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR-4624, Vol. 6 NUREG/CR4668 NUREG/CR-5254 NUREC/CR-5366 NUREG/CR-5405 NUREG/CR 5173 NUREG/CR-5510 NUREG/CR-5557 NUREG/CR-5575 NUREC/CR-5602 NUREG/lA-0011 NUREG/lA-0012 NUREG/IA-0013 NUREC/IA-0018 NUREG/lA-0019 NUREG/IA-0021 NUREG/lA-0022 NUREC/lA-00M NUREG/IA-0031 NUREG/IA-0032 quality assurance NUREG/CR-5376 radiation doses NUREC-1266, Vol. 4 NUREG/CP-0110 NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-5517 NURcG/CR-5659 radiation effects NUREC/CR-5409 NUREG/CR-5530 radiation hazards NUREG/CR-4691, Vol.1 NUREG/CR-4691, Vol. 2 NUREC/CR-4691, Vol. 3 radiation heating NUREC/CR-5366 radiation monitoring NUREC/CR-5449 D-23

Keyword NUREC Report Number radiation protection NUREG/CP 0113 NUREG/CR-4551, Vol. 2, Rev.1, Part 7 radioactive aerosols NUREG/CR-5545 radioactive effluents NUREG/CR 2331, Vol. 9, No. 4 radioactive materials NUREG/CR-4551, Vol. 2, Rev.1, Part 7 radioactive waste disposal NUREG/CR-4735, Vol. 6 NUREG/CR-5229, Vol. 2 NUREG/CR-5517 NUREG/CR 5521 NUREG/CR-5542 NUREG/CR 5547 radioactive waste management NUREG/CR-5393 NUREG/CR-5517 radioactivity NUREG/CR-5512 radioisotopes NUREG/CR-5545 NUREG/CR 5548 radionuclide migration NUREG-1150, Vol. I NUREG-1150, Vol. 2 NUREG-1266, Vol. 4 NUREG/CR-4551, Vol. 2, Rev.1, Part 7 NUREC/CR-4624, Vol. 6 NUREC/CR-5229, Vol. 2 NUREG/CR-5377 NUREG/CR-54'75 NUREG/CR-5512 NUREG/CR-5547 NUREG/CR-5548

. RA17TRAN codes NUREG/CR-5517 RALOC codes NUREG/CP-0105, Vol. 2 RAMONA codes NUREG-1266, Vol. 4 NUREC/CP 0105, Vol. 3 RAMONA-3B codes NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 reactivity NUREC/CR-M21 NUREG/CR-5573 reactivity insertions NUREC/CR-2331, Vol. 9, No. 3 NUREC/CR-2331, Vol. 9, No. 4 reactor accidents NUREG-1272, Vol. 4, No.1 NUREG/CP-0105, Vol.1 NUREG/CP-0105, Vol. 2 NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR-4550, Vol. 3, Rev.1, Part 1 D-24

Keyword NUREG Report Number reactor accidents (continued)

NUREC/CR-4550, Vol. 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 2, Rev.1, Part 7 NUREG/CR-4624, Vol. 6 h" mEC/CR4668 9/CR4691, Vol.1 3/CR-4691, Vol. 2

.J;/CR4691, Vol. 3 e

NUREC/CR4840 NUREC/CR 5253 NUREG/CR-5254 NUREC/CR-5262 NUREC/CR-5273, Vol. 4 NUREC/CR-5316 NUREG/CR 5405 NUREG/CR 5447 NUREC/CR-5477 NUREG/CR-5480 NUREC/CR-5514 NUREC/CR-5527 NUREG/CR-5528 NUREC/CR-5545 NUREC/CR-5557 NUREG/CR-5573 NUREG/CR-5575 NUREC/CR 5586 sq NUREG/CR-5590 NUREC/CR-5653 NUREG/CR 5659 reactor components NUREG/CP-0105, Vol.1

.NUREG/CR 2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREC/CR4639, Vol.1, Re; I NUREG/CR4639, Vol. 4, Rev. 2 NUREC/CR-4639, Vol. 5, Part 1, Rev. 3 NUREG/CR-4639, Vol. 5, Part 2, Rev. 3 NUREG/CR4639, Vol. 5, Part 3, Rev. 3 NUREG/CR 5419 t

NUREG/CR-5476 NUREC/CR-5649, Vol.1 NUREC/CR-W9, Vol. 2 reactor cooling systems NUREG-1266, Vol. 4 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-4951, Vol. 2, Rev.1. l' art 1 NUREC/CR-54,7 NUREC/CR-5500 NUREG/CR-5514 NUREG/CR-5545 NUREC/CR-5557 NUREC/CR-W4 reactor core disruption NUREG/CR-5573 reactor cores NUREG-1150, Vol.1 NUREG-1150, Vol. 2 NUREG-1266. Vol. 4 NUREG/CP-0103, Vol. 3 D-25

,--,r

i K vword NUREG Renort Number f

reactor cores (continued)

NUREC/CR4550, Vol 3, Rev.1, Part 1 NUREG/CR4550, Vol. 3, Rev.1, Part 3 NUREG/CR4550, Vol. 4, Rev.1, Part 3 NUREC/CR4550, Vol. 5, Rev.1, Part 1 NUREG/CR-4840 NUREG/CR-5316 NUREG/CR 5421 NUREG/CR-5510 NUREG/CR 5527 NUREC/CR-5653 NUREG/lA-0018 NUREG/lA-0022 reactcr experimental facilities NUREC/CR4668 rtactor instrumentation NUREG/CR-5119 reactor licensing NUREG/CR-5622 reactor maintenance NUREC/CR-5419 reactor materials NUREG/CR-5273, Vol. 4 reactor noise NUREG/CR-5605 reactor operation NUREG-1272, Vol. 4, No.1 NUREC/CR 5419 reactor operators NUREG/CR-5572 NUREG/CR 5659 reactot protection systems NUREG/CR 5419 reactor safety NUREG-1266, Vol. 4 NUREG-1272, Vol. 4, No.1 NUREG-1420 NUREC/CP-0105, Vol.1 NUREC/CP-0105, Vol 2 NUEEC/CP-0105, Vol. 3 NUREG/CP-0113 -

NUREG/CR-2331, Vol. 9, No. 3 NUREC/CR 2331, Vol. 9, No. 4 NUREC/CR-4469, Vol.10 NUREG/CR4351, Vol. 2, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUREC/CR4551, Vol. 3, Rev.1, Part 2 i

NUREG/CR4551, Vol. 4, Rev.1, Part 1 NUREG/CR4551, Vol. 4, Rev.1, Part 2 NUREC/CR4551, Vol 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 5, Rev.1, Part 2 NUREC/CR4551, Vol. 6, Rev.1, Part t NUREG/CR4551, Vol. 6, Rev.1, Part 2 NUREG/CR-5111 NUREC/CR-5254 r

NUREC/CR-5572 NUREC/CR 5622 NUREG/CR 5644 D-26

Keyword NUREG Report Number I

reactor shutdown NUREG/CR-5622 reactor stability NUREG/CR-M21 NUREG/CR-5605 reactor vessels NUREG/CR-4551, Vol. 2, Rev.1, Part 1 NUREG/CR-5514 NUREG/CR-5530 NUREC/CR-5573 REDEQL code.,

NUREG/CR-5548 REFLA codes NUREG/CP-0105, Vol. 3 regulations NUREC-1266, Vol. 4 NUREG/CR 5398 NUREG/CR-5521 reinforced concrete NUREG/CR-M76 RELAP5 codes NUREC/CP-0105. Vol. 3 RELAPS/ MODI cc. des NUREG/IA-0012 NUREC/lA-0013 NUREC/lA-0021 RELAP5/ MOD 2 codes NUREG/CP-0105, Vol. 2 RELAP5/ MOD 3 codes NUREG/CP-0113 RELAP5/SCDAP codes NUREG-1150, Vol. 2 NUREC/CR-4551, Vol. 2, Rev.1, Part I reliability NUREG/CP-0105, Vol.1 NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331. Vol. 9, No. 4 NUREC/CR-4469, Vol.10 NUREG/CR-4639, Vol.1, Rev.1 NUREG/CR-4639, Vol. 4, Rev. 2 NUREC/CR-4639, Vol. 5, Part 1, Rev. 3 NUREC/CR-4639, Vol. 5, Part 2, Rev. 3 NUREC/CR-4908 NUREG/CR-Sill NUREG/CR-5213, Vol.1 NUREG/CR-5213, Vol. 2 NUREG/CR-M19 NUREC/CR-5438 NUREC/CR-5477 NUREG/CR-5510 relief valves NUREC/CR-M47 research programs NUREG-1266, Vol. 4 NUREG/CP4)105, Vol.1 NUREG/CP-0105, Vol. 2 NUREG/CP-0105, Vol. 3 NUREG/CP-0113 D-27

Keyword NUREC Report Number research programs (continued)

NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 resins NUREG/CR 5229, Vol. 2 reviews NUPEG-1420 NUREG/CR-4735, Vol. 6 NUREG/CR 5398 risk assessment NUREG-1150, Vol.1 NUREG-1150, Vol. 2 NUREC-1266, Vol. 4 NUREG-1420 -

NUREC/CP-0105, Vol.1 NUREG/CP-0105, Vol. 2 NUREG/CP-0105, Vol. 3 NUREG/CR-2331, Vol. 9, No. 3 NUREG/CR-233, Vol. 9, No. 4 NUREG/CR-4550, Vol. 3, Rev.1, Part 1 NUREG/CR-4550, Vol. 3, Rev.1, Part 3 NUREG/CR-4550, Vol. 4, Rev.1, Part 3 NUREC/CR-4550, Vol. 5, Rev.1, Part 1 NUREC/CR-4551, Vol. 2, Rev.1, Part 1 NUREC/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 2 NUREC/CR-4551, Vol. 4, Rev.1, Part 1 NUREG/CR-4551, Vol. 4, Rev.1, Part 2 NUREG/CR-4551, Vol. 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 5, Rev.1, Part 2 NUREG/CR-4551, Vol. 6, Rev.1, Part 1 NUREG/CR4531, Vol. 6, Rev.1, Part 2 NUREG/CR4639, Vol.1, Rev i NUREG/CR4639, Vol. 4, Rev. 2 NUREG/CR4639, Vol. 5, Part 1, Rev. 3 NUREG/CR-4639, Vol. 5, Part 2, Rev. 3 NUREC/CR-4691, Vol.1 NUREG/CR-4691, Vol. 2 NUREG/CR-4691, Vol. 3 NUREC/CR-4MO NUREG/CR-5111 NUREC/CR-5213, Vol.1 NUREC/CR-5213, Vol. 2 NUREG/CR 5253 NUREC/CR-5262 NUREG/CR-5377 NUREC/CR-M19 NUREG/CR 5447 NUREC/CR-M77 NUREC/CR-5510 NUREG/CR-5517 NUREG/CR-5527 NUREG/CR-5528 NUREC/CR-5572 NUREC/CR-5575 NUREG/CR-5586 D-28

Keyword NUREG Report Number RISQUE codes NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUIEG/CR-4551, Vol. 4, Rev.1, Part 1 NUREC/CR-4551, Vol. 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 6, Rev.1, Part i RISQUE /PRAMIS codes NUREG-1150, Vol. 2 Robinson-2 reactor NUREG/CR-5530 RPS codes NUREC/CP-0110 runoff NUREC/CR-5475 NUREC/CR-5523 ruptures NUREG/CR-4551, Vol. 2, Rev.1, Part 1 SAA'IY codes NUREG/CR-M24 SADDE codes NUREG-1266, Vol. 4 safety NUREG/CR-4550, Vol. 3, Rev.1, Part 3 NUREG/CR-4550, Vol. 4, Rev.1, Part 3 NUREG/CR-4639, Vol. 5, Part 3, Rev. 3 SAFT-UT cot NUREG/CP-0113 SAILOR codes NUREC/CR-5449 salt caverns NUREG/CR-4735, Vol. 6 NUREG/CR 5521 sampling NUREC/CR-5262 NUREG/CR-M24 sanitary landfills NUREC/CR-5517 SAP 4 codes NUREG/CR-5506 SCALE codes NUREG/CR-5366 NUREG/CR-5468 SCDAP codes NUREG/CR-4668 NUREG/CR 5316 SCDAP/RELAP codes NUREG/CR-5586 SCDAP/RELAP5 codes NUREG-1266, Vol. 4 NUREG/CP-0105, Vol. 2 NUREC/CR-5273, Vol. 4 scram NUREG/CR-4551, Vol. 2, Rev.1, Part 7 SCSS codes NUREG-1272, Vol. 4, No. I seals NUREG/CR-4624, Vol. 6 sediment-water interfaces NUREG/CR-5547 D-29

Keyword NUREG Report Numbet SEILHS codes NUREG/CR4551, Vol. 4, Rev.1, Part 2 seismic detection NUREG/CR4753, Vol. 3 seismic effects NUREC/CR 5588, Vol.1 NUREG/CR-5538, Vol. 2 NUREC/CR 5588, Vol. 3 seismic events NUREC/CP 0105, Vol. 2 NUREC/CR4550, Vol. 3, Rev.1, Part 3 NUREC/CR-4550, Vol. 4, Rev.1, Part 3 NUREC/CR-4753, Vol. 3 seismicity NURLG/CR-4753, Vol. 3 SENCOF codes NUREC/CR-5393 sensitivity NUREG/CR-5553 sensitivity analysis NUREC/CR4691, Vol.1 NUREC/CR4691, Vol. 2 NUREC/CR-4691, Vol. 3 NUREG/CR 5256 NUREG/CR-5

NUREG/

NUREC,. A-5547 NUREG/CR-5586 NUREG/lA-0018 NUREG/lA-0031 NUREG/lA-0032 SEQSOR codes NUREG/CR4551, Vol. 5, Rev.1, Part 1 NUREG/CR-4551, Vol. 5, Rev.1, Part 2 NUREG/CR-5602 SEQUFUN codes NUREG/CR4551, Vol. 5, Rev.1, Part 2 Sequoyah-1 reactor NUREG/CR4550, Vol. 5, Rev.1, Part 1 NUREG/CR4551, Vol. 5, Rev.1, Part 1 NUREG/CR4551, Vol. 5, Rev.1, Part 2 NUREC/CR-5405 NUREG/CR-5602 SE15 codes NUREG-1150, Vol. 2 NUREC/CR4550, Vol. 3, Rev.1, Part 3 NUREG/CR-4550, Vol. 4, Rev.1, Part 3 NUREG/CR4840 NUREG/CR-5N2 NUREG/CR-5419 SFUEL1W codes NUREC/CP-0105, Vol.1 SHAKE codes NUREG/CP-0105, Vol. 2 NUREC/CR-4550 Vol. 3, Rev.1, Part 3 NUREC/CR4550, Vol. 4, Rev.1, Part 3 NUREG/CR-4340 D-30

.i

Keyword NUREG Report Number shells NUREG/CR-4554, Vol. 6 NUREC/CR-4554, Vol. 7 shielding NUREC/CR-5468 SIAS codes NUREC/CR-5548 SIM codes NUREG/CR-5588, Vol.1 NUREG/CR-5588, Vol. 2 NUREG/CR-5588, Vol. 3 site characterization NUREC/CR-5588, Vol.1 NUREC/CR-5588, Vol. 2 NUREG/CR-5588, Vol. 3 Sizewell-B reactor NUREC/I A 0021 skin NUREG-1266, Vol. 4 SLAVE codes NUREG/CR-5588, Vol.1 NUREC/CR-5588, Vol. 2 NUREG/CR-5588, Vol. 3 SLIM-MAUD codes NUREC-1150, Vol.1 NUREG/CP.0105, Vol.1 SMACS codes NUREG/CR-M77 soil-structure interactions NUREC/CP-0105, Vol. 2 NUREC/CR-4550, Vol. 3, Rev.1, Part 3 NUREG/CR-4550, Vol. 4, Rev.1, Part 3 NUREC/CR-5588, Vol.1 NUREC/CR-5588, Vol. 2 NUREG/CR-5588, Vol. 3 soils NUREG/CR-5512 NUREC/CR-5523 NUREG/CR-5607 SOLGASMIX codes NUREC/CR-M80 SOLMNEQ codes NUREG/CR-5548 solutes NUREG/CR-5607 source terms NUREG-1420 NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUREC/CR-4551, Vol. 3, Rev.1, Part 2 NUREG/CR-4551, Vol. 4, Rev.1, Part 1 NUREC/CR 4551, Vol. 4, Rev.1, Part 2 NUREG/CR 4551, Vol. 5, Rev.1, Part i NUREC/CR-4551, Vol. 5, Rev.1, Part 2 NUREC/CR-4551, Vol. 6, Rev.1, Part 1 NUREC/CR-4551, Vol. 6, Rev.1, Part 2 NUREC/CR-4624, Vol. 6 NUREC/CR-5253 NUREG/CR-5262 NUREC/CR-5528 D-31

Keyword NUREG Report Number spatial distribution NUREG/CR-5607 spmt fuel casks NUREC/CR-4554, Vol. 6 NUREC/CR 4554, Vol. 7 NUREG/CR-5366 spent fuel storage NUREC/CR-2331, Vol. 9, No. 4 spent fuels NUREC/CR-4735, Vol. 6 SQUIRT codes NUREG/CP-0113 SROA codes NUREG/CR-5438 stainless steels NUREC/CR-4tb9, Vol.10 NUREG/CR-5273, Vol. 4 standanfs NUREG/CR-5521 statistics NUREC/CR-5253 NUREC/CR-5262 NUREG/CR-5424 STCP codes NUREG-1150, Vol.1 NUREG-1150, Vol. 2 NUREG/CP-0113 NUREG/CR-5528 NUREG/CR-5633 steady-state conditions NUREC/CR-5421 NUREC/lA-0030 steam generators NUREG/CR-4551, Vol. 2, Rev.1, Part 1 NUREC/CR-5213, Vol. 2 NUREG/CR-5557 NUREC/lA-0018 NUREG/IA-0030 steel-ASTM A516 NUREG/CR-5405 steels NUREC/CR-4816 NUREC/CR-5409 stochastic processes NUREC/CR-5393 strains NUREC/CR-5476 stress corrosion NUREG/CR-4908 stresses NUREG/CR-5476 structural models NUREC/CR-2331, Vol. 9, No. 3 NUREG/CR-2331, Vol. 9, No. 4 NUREG/CR-5476 NUREC/CR-5506 NUREG/CR-5644 submerged arc welding NUREC/CR-5584 D-32 1

Keyword NUREG Report Number supports NUREC/CR-5506 NUREG/CR 5614 SURFACE codes NUREG/CR-5453, Vol. 5 surface waters NUREG/CR-5453, Vol. 5 NUREC/CR-5475 Surry-1 reactor NUREC/CR4550, Vol. 3 Rev.1, Part 1 NUREC/CR 4550, Vol. 3, Rev.1, Part 3 NUREC/CR4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 2 NUREC/CR-4840 NUREG/CR-M47 SURSOR codes NUREG-1150, Vol. 2 NUREG/CR-4551, Vol. 3, Rev.1, Part 1 NUREG/CR-4551, Vol. 3, Rev.1, Part 2 surveillance NUREG/CR-MN, Vol.1 NUREG/CR-5530 SWA'IS codes NUREG/CR-5393 SWENTG codes NUREC/CR-5393 Switzerland NUREG/IA-0018 system failure analysis NUREC/CR-5111 NUREG/CR-M19 NUREC/CR-5527 systems analysis NUREC/CP 0105, Vol.1 NUREC/CR-4550, Vol. 3, Rev.1, Part 1 NUREC/CR-4550, Vol. 5, Rev.1, Part 1 NUREG/CR-5262 taxonomy NUREC/CR-5438 telemetry NUREC/CR4753, Vol. 3 TEMAC codes NUREG-1150, Vol. 2 NUREC/CR-4550, Vol. 4, Rev.1, Part 3 NUREG/CR-5262 temperature dependence NUREG/CR-5405 terrestrial ecosystems NUREG/CR 5377 test facilities NUREC/CR-5254 testing NUREC/CR-4554, Vol. 7

.NUREG/CR-5584 TEXAS codes NUREC/CP-0105, Vol. 2 thermal conduction NUREC/CR-5366 D-33

Keyword NUREG Report Number thennal shock NUREC/CR-5473 thermal stresses NUREC/CR4554, Vol. 6 thermodynamics NUREC/CR-5573 Three Mile Island-2 reactor NUREG/CR 5229, Vol. 2 titration NUREG/CR 5547

~ TORT codes NUREG/CR-5449 toxic materials NUREC/CR-5659 TRAC-BFI/ MOD 1 NUREG/CP-0113 TRAC-BWR codes NUREG/CP-0105, Vol. 3 TRAC PF1/ MOD 2 codes NUREG/CP-0113 TRAC-PWR codes NUREG/CP 0105, Vol. 3 TRAC /MELPROG codes NUREG-1150, Vol. 2 NUREC/CR 4551, Vol. 2, Rev.1, Part 1 TRACR3D codes NUREC/CR-5523 transients NUREG/CR-5473 NUREG/CR 5514 NUREG/CR-5557 NUREG/IA-0030 transition temperature NUREC/CR-5584 TRAP-MELT codes NUREG/CR-4624, Vol. 6 NUREG/CR-5447 tritium NUREG/CR-5607 Trojan reactor NUREC/CR 5506 NUREC/CR 5644 tuff NUREG/CR-4735, Vol. 6 l-turbines NUREG/CR 54G1, Vol. I t-l turbulence NUREG/CR-5649, Vol.1 NUREC/CR-5649, Vol. 2 UFOMOD codes NUREG/CR 5377

[

UFUN codes NUREC/CR-4551, Vol. 3, Rev.1, Part 2 UKAEA NUREG/IA-0011 NUREC/IA-0019 l

NUREC/IA-0022 l

l h*N l

~

Enword NUREG Report Number ultrasonic testing NUREG/CR4469, Vol.10 NUREC/CR-4908 underground disposal NUREC/CR-5256 NUREC/CR-5393 NUREG/CR 5398 NUREG/CR-5523 United Kingdore, organizations NUREG/lA-0012 NUREC/lA-0013 NUREG/lA-0021 UNSAT2 codes NUREC/CR-5523 uranium NUREC/CR-5273, Vol. 4 US EPA NUREG/CR-5521 USRDST codes NUREC/CR-4551, Vol 3, Rev.1, Part 2 NUREG/CR-4551, Vol. 4, Rev.1, Put 2 NUREC/CR4551, Vol. 5, Rev.1, Part 2 USRDSTGG codes NUREG/CR4551, Vol. 6, Rev.1, Part 2 valves NUREC/CP-0105, Vol. I VAM2D codes NUREC/CP-0113 NUREG/CR-5453, Vol. 5 NUREG/CR-5523 VANESA codes NUREG/CP-0113 NUREG/CR4624, Vol. 6 VARSKIN codes NUREC-1266, Vol. 4 verification NUREC/CR-5376 very high temperature NUREG/CR-5480 VICTORIA codes NUREG/CP-0105, Vol. 2 NUREG/CP-0113 NUREG/CR-5480 waste forms NUREG/CR-5229, Vol. 2 waste management NUREC/CR-5453, Vol. 5 WATEQ codes NUREC/CR-5548 water chemistry NUREC/CR-4735, Vol. 6 water cooled reactors NUREG-1272, Vol. 4, No.1 NUREC/CP4105, Vol.1 NUREC/CP-0105, Vol. 2 NUREG/CP-0105, Vol. 3 NUREC/CR4469, Vol.10 NUREG/CR-5419 NUREC/CR-5476 D-35

Keyword NUREG Report Number water cooled reactors (continued)

NUREC/CR-5545 NUREG/CR 5584 water moderated reactors NUREG-1272, Vol. 4, No.1 NUREG/CP 0105, Vol.1 NUREG/CP-0105, Vol. 2 NUREG/CP4)105, Vol. 3 NUREG/CR4469, Vol.10 NUREC/CR 5419 NUREG/CR-5476 NUREG/CR 5545 NUREC/CR-5584 WAVCO codes NUREC/CP-0105, Vol. 2 weather NUREC/CR4691, Vol.1 NUREC/CR4691, Vol. 2 NUREC/CR-4691, Vol. 3 weighting functions NUREC/CR-5424 welded joints NUREG/CRwi69, Vol.10 NUREC/CR-4816 NUREG/CR-5584 WHOCAM codes NUREG/CR-5366 wind NUREC/CR-5477 XSDRNPM codes NUREG/CR-5653 XSDRNPM S codes NUREC/CP-0105, Vol. 2 XSOR codes NUREG-1150, Vol. 2 NUREG 1420 XXSOR codes NUREC/CR-5253 NUREG/CR-5262 Yucca mountain NUREG/CR-4735, Vol. 6 zircaloy NUREC/CR-4735, Vol. 6 ZISOR codes NUREG/CR-5575 D-36

-i

!?059513 W 1 1 13019s US oft - O ta r p 01

' I t, r.

r' U B ! r t.1 1 0 N E "VC' Trt

'-Nunfr, P-?.

s W f S H I N r, T O N DC

t5r l

/

TilIS DOCUMENT WAS PRINTED USING RECYCLED PAPER

- _ _ - _ _ - _ _ - _ _ _ - _