ML20091L841

Computer Codes and Mathematical MODELS.January-December 1994

ML20091L841
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Issue date:	08/31/1995
From:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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References
NUREG-BR-0083, NUREG-BR-0083-V10, NUREG-BR-83, NUREG-BR-83-V10, NUDOCS 9508290356
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Computer Codes and Mathematical Models i

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Energy Science and 'Ibchnology Software Center P.O. Box 1020 Oak Ridge, TN 37831-1020 Au; d 1995 End-User Support Services Branch Omce of Information Resources Management ILS. Nuclear Regulatory Commission Washington, DC 20555-0001

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ABSTRACT l

This report containc citttions of NUREG-series documents issued in calendar year 1994 relating to computer software and mathematical models for scientific, engineering, or technology-related programs l

performed or sponsored by the U.S. Nuclear Regulatory Onmminion (NRC). It is intended as a reference tool to assist the scientific and technical analyst in obtaining information on NRC computer-related activities.

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CITATIONS 1

APPENDE A: Index by NUREG-Series Report Numbes A-1 APPENDE B: Index by Software Identification B-1 APPENDIX C: Index by Contractor Report Number C-1 t

APPENDIX D:Index by Keyword D-1 i

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FOREWORD

'Ibe citations in this h==aat appear in NUREG-series document order. Citations of NRC staff-generated reports designated NUREG-xxxx are listed first, followed by any conference proceedmgs identified as NUREG/CP-xxxx, contractor-generated reports published as NUREG/CR-xxxx docu-meats, grant repons published as NUREG/GR-xxxx documents, and International Agreement repons issued as NUREG/IA-xxxx publications. Each citation contains the following: NUREG series report

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number; software identification (where applicable); contractor report number, report title; a description j

of the report contents; pub!ication date; names of the individuals responsible for preparing, compiling, j

or editing the reporg contractor name and location; sponsoring NRC organiration; and keywords or i

J descriptors. hidexes by NUREG-series report number, software identiScation, contractorreport number, and keyword are included in the Appandivas.

Specific code names and software identification appear in the heading of those citations with primary emphasis on specific mathematical models, computer codes, or databases. The term " General"is used in the heading of those citations which contain significant information on many models, computer codes, or databases.

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l vii NUREG/BR--0083, Vol.10

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NUREG-1502 General

Title:

Assessment of Databases and Modeling Capabilities for the CANDU 3 Design

==

Description:==

The NRC staff has been conducting a prelimmary n: view of the Canadian Deuterium Uranium Model 3 (CANDU 3) reactor design, a new heavy-water design developed by Atomic Energy of Canada Limited through its U.S. affiliate, AECL Technologies. The review has been aimed at identifying key technical areas and policy issues that will have to be addressed for standard design certi5 cation. As part of the research program associated with the preliminary review, the NRC Of5ce of Nuclear Regulatory Research (RES) has completed an assessment of databases and modeling capabilities that might be needed to support the CANDU 3 design To ensure full coverage of the design, a detailed assessment methodology was developed by the RES staff and was implemented with help from research projeets at three nationallaboratories. ' Ibis reponintegrates and summarizes the database and modeling assessments, including major contributions from these laboratories.

Publication Date:

July 1994 Prepared by:

Carlson, D.E.; Meyer, R.O.

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Ke3words:

ARROTTA, CANDU type reactors, CASMO, CATHENA, CERBERUS, CONTAIN, CORCON, CPM-2, CS AU, design, D1F3D, evaluation, fission product release, FRAPCON, FRAP-T, HELIOS, bydraulics, information needs, information systems, MCNP, MELCOR, NESTLE, nuclear fuels, reactor accidents, reactorkinetics, RELAP5, SCDAP, thermodynamics, TRAC-P, WIMS i

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NUREG/CP--0127 NEA/CSNI/R--(93)8: CONF-930157--

General

Title:

Proceedings of the CSNI Spectahsts Meeting on Fuel-Coolant Interactions

==

Description:==

A spectahsts meeting on fuel-coolant interactions was held in Santa Barbara, California, from January 5-7,1993.The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize ongoing work, provide opportunities for mutual check points, seek to focus the technical issues on maners of practical signi5cance, and reevaluate both the objectives as well as path of future research.

Publication Date:

March 1994 Prepared by:

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

BWR type reactors, CHYMES, corium, CULDESAC, ESPROSE, explosions, FEAT, FLOW 3D, fluid flow, fragmentation, fuel-coolant interactions, beat transfer, hydraulics, IDEMO, IFCI, IVA-3, management, meetings, meltdown, mixing, PM-ALPHA, PRECURSOR, PWR type reactors, quenching, reactor safety, RELAP5BdOD2, ROAAM, steam, TEXAS-III, THIRMAL 1, TIOER, TRIO-MC, vapors NUREG/BR--0083, Vol.10 2

CONF-931079--Vol.2 NUREG/CP--0133 Vol.2 General

Title:

Twer,ty-First Water Reaction Safety Information Meeting. Volume 2, Severe Accident Research

==

Description:==

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water keactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27,1993. The papers are printed in the order of theirpresentation in each session and describe progress and results ofprograms in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 2, presents papers on severe accident research.

Publication Date:

April 1994 Prepared by:

Monteleone, S. [ comp.] [Drookhaven National Lab., Upton, NY (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research Keywords:

ADINA, ALPHA, APRIL, BASSIM, BWR type reactors, CLCH model, combustion, COMMIX, COMPACT, CONTAIN, containment, CORCON, corium, CORMLT, CORSOR, DEBRIS, detonations, EVNTRE, FLUTAN, GOTHIC, heat transfer, hydraulics, hydrogen, IFCI, LOWCORV, low-Reynolds number k-e models, MAAP, MAPHY BURN, meetings, MELCOR, MELPROG, NARAL-SM, nuclear power plants, PWR type reactors, RALOC, RASPLAV, reactor accidents, reactor components, reactor safety, research programs, RETRAN, ROAAM, SABRE, SCANAIR, SCDAP/RELAP5, SPARC, STANJAN, TAC-2D, TCE model, TRACG, VICTORIA, WAVCO, ZND model 3

NUREG/BR--0083, Vol.10

NUREG/CP--0138 CONF-9310377--

General'

Title:

Promedags of Workshop I in Advanced Topics in Risk and Reliability Analysis. Model Uncenamty:Its Characterization and Quamification

Description:

The purpose of the workshop series is to provide a forum for in-depth dacussion of key ~

problems in risk and reliability analysis and for the development of solution strategies to.

anack these problems. The workshop topics are selected by an international organmng comminee. Special emphasis is placed on risk and reliability problems that are irapar'=nt timely, dif6 cult to resolve without funber research, and in need of expen input to formulate structured research agendas to assist timely and efficient resolution. The topic of model uncertamty fits all of these criteria. Model uncertamtaes have been acknowledged as being extremely imponantin a wide variety of risk assessment application areas, including nuclear powerplant risk assessments, radioactive waste repository performance assessments, human.

heakh risk assessments, and environmental risk assessments. Clearly, a risk study that neglects to provide a careful treatment of model uncertamties can provide decision makers with a distorted picture of the uncertainnes in the study's results. The papers and working group reports contamed in these proceedings are divided imo three sections.The first section contains papers discussing the appropriate fomialism for dealing with model uspruduty.

'Ibe second section of the proceedings contains papers discussing problems in coping with model uncertainnes and approaches for dealing with these problems.1be third section of the proceedmgs contams the summaries of the three working groups. Group 1 deals with the implications of model uncenamty on decision making (including regulatory applications),

Group 2 deals with the formal definition of model uncenainty, and Group 3 deals with approaches to qnnanfy model uncertainty. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

Publication Date:

October 1994 Prepared by:

Mosleh, A.; Smidts, C. [eds.] [ Maryland Univ., College Park, MD (United States)); Siu, N.

(ed.] [ Idaho National Engineering Lab., Idaho Falls, ID (United States)); I.mi, C. [ed.]

[ Nuclear Regulatory Commission, Washington, DC (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

accuracy, environmental impacts, fires, health hazards, meetings, nuclear facilities, probabilistic estimation, probability, reliability, risk assessment NUREG/BR--0083, Vol.10 4

CONF-9410216--

NUREG/CP--0139 General

Title:

Transactions of the Twenty Second Water Reactor Safety Information Meeting

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Description:==

nis report contains summaries of papers on reactor safety research to be presented at the 22nd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 24-26,1994. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission. Summaries ofinvited papers concerning nuclear safety issues from U.S. government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also induded. *Ihe summaries 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.

Publication Date:

October 1994 Prepared by:

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Researdi Keywords:

aging, ALPHA, APRIL. BWR type reactors, COMETA, CONTAIN, CORCON, engineered safety systems, FAVOR, fuel elements, beat transfer, human factors, hydraulics, IFCI, MAAP, meetings, MELCOR, PWR type reactors, reactor accidents, reactor components, reactor cooling systems, reactor safety, regulations, RELAP/ MOD 3, research programs, risk assessment SAPHIRE, SCDAP/RELAP5, seismic effects 5

NUREG/BR- 0083, Vol.10

NUREG/CP--0145 CONF-9207249-General

Title:

Workshop on Developing Safe Software

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Description:==

The Workshop on Developing Safe Software was held July 22-23,1992, at the Hotel del Coronado, San Diego, Califomia. The purpose of the workshop was to have four world experts discuss among themselves software safety issues that are ofinterest to the U.S.

Nuclear Regulatory Commission.These issues concem the development ofsoftware systems for use in nuclear power plant protection systems. The workshop compnsed four sessions.

Wednesday moming, July 22, consisted of presentations from each of the four panel members. On Wednesday aftemoon, the panel members went through a list of possible software development techniques and commented on them. The Thursday morning, July 23, session consisted of an extended discussion among the panel members and the observers from the NRC. A 6aal session on Thursday aftemoon consisted of a discussion among the NRC observers as to what was leamed from the workshop.

Publication Date:

November 1994 Prepared by:

Lawrence, J.D. [bwrence Livermore National Lab., CA (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Reactor Controls and Human Factors Keywords:

engineered safety systems, failure mode analysis, meetings, nuclear power plants, reactor safety, reliability, security, testing NUREG/BR--0083, Vol.10 6

BNL-NUREG--51934 Vol.5 NUREG/CR--4409-Vol.5 ACE, ACEFAX

Title:

Data Base on Dose Reduction Research Projects for Nuclear Power Plants. Volume 5

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Description:==

' Ibis is the fifth volume in a series of reports that provide information on dose reduction research and health physics technology for nuclear power plants. 'Ibe information is taken from two of several data bases maintained by Brookhaven National Laboratory's ALARA Center for the Nuclear Regulatory Commission. The research section of the repon covers dose reduction projects that are in the experimental or developmental phase. It includes such topics as steam generator degradation, decontammation, robotics, improvements in reactor materials, and inspection techniques. The section on health physics technology discusses dose reduction effons that are in place or in the process of being implemented at nuclear power plants. A total of 105 new or updated projects are described. All project abstracts from this repon are available to nuclear industry professionals with access to a fax machine through the ACEFAX system or with access to a computer with a modem and the proper communications software through the ACE system. Detailed descriptions of how to access all the data bases electronically are in the appendices of the repon.

Publication Date:

May 1994 Prepared by:

Khan, T.A.; Yu, C.K.; Roccklein, A.K. [Brookhaven National Lab., Upton, NY (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

accidents, ACE, ACEFAX, BWR type reactors, data base management, data compilation, nuclear power plants, PWR type reactors, radiation doses, radiation hazards, radiation protection i

7 NUREG/BR--0083, Vol.10

NUREG/CR--4639 Vol.5-Rev.4-Pt.2 EGG--2458-Vol.5-Rev.4-Pt.2 NUCLARR

Title:

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data Manual. Part 2: Human Error Probability (HEP) Data: Volume 5, Revision 4

==

Description:==

This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 data base, which is sponsored by the U.S. Nuclear Regulatory Commission. NUCLARR was designed as a tool for dsk analysis. Many of the nuclear reactors in the United States and several outside the United States are representedin the NUCLARR data base. NUCLARR includes both human error probability estimates for workers at the plants and haniware failure date for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimater for probabilistic risk assessments and human reliability analyses. 'Ibe data manualis organized to permit manual searches of the information if the computerized version is not available.

Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text ofParts 2 and 3. The user can now find introductory material either in the original Part 1, orin Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Pan 3, the hardware component reliability data.

Publication Date:

September 1994 Prepared by:

Reece, W.J.; Gilbert, B.G.; Richards, R.E. [EG and G Idaho, Inc., Idaho Falls, ID (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

compiled data, information systems, NUCLARR, personnel, reactor operators, reactors, reliability, risk assessment, US NRC NUREG/BR--0083, Vol.10 8

EGG-2458-Vol.5-Rev.4-Pt3 NUREG/CR-4639-Vol.5-Rev.4-Pt.3 NUCLARR

Title:

Nuclear Computerized Library for Assessing Reactor Reliability (NUCIARR): Data Manual. Pan 3: Hardware Component Failm Data: Volume 5, Revision 4 i

==

Description:==

This data mannal contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 data base, which is sponsored by the U.S. Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the United States and several outside the United States are represented in the NUCLARR data base. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equiprnent. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. 'Ibe data manual is organized to permit manual searches of the infonnation if the computerized version is not available. Originally, the manual was published in three pans. In this revision the intmduc-tory material located in the original Pan I has been incorporated into the text of Pans 2 and i

3. The user can now find introductory material either in the original Pan 1, orin Parts 2 and 3 as revissi Part 2 contains the human error probability data, and Pan 3, the hardware component reliability data.

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Publication Date:

September 1994 Prepared by:

Reece, W.J.; Gilbert, B.G.; Richards, R.E. [EG and G Idaho, Inc., Idaho Falls, ID (Urited

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States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

compiled data, equipment, failures, information systems, NUCLARR, probabilistic estimation, reactor components, reactors, reliability, risk assessment, US NRC t

9 NUREG/BR--0083, Vol.10

I NUREG/CR--4816.Rev.2 ORNLfrM--10328/R2 PR-EDB

Title:

PR-EDB: Power Reactor Embrittlement Data Bas:, Version 2. Revision 2, Program Description

==

Description:==

Investigations of regulatory issues, such as vessel integrity over plant life, vessel failure, sufficiency of current codes, Standard Review Plans (SRP's), and Guides forlicense renewal, can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers.

The Power Reactor Embrittlement Data Base (PR EDB)is a comprehensive collection of data for U.S. commercial nuclear reactors. The current version of the PR EDB contains the Charpy test data that were irradiated in 252 capsules of % reactors and consist of 207 data points for beat-affected-zone (HAZ) materials (90 different HAZ),227 data points for weld materials (105 different welds),524 data :.ints for base materials (136 different base materials), including 297 plate d.sta pointu85 different plates),119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dB ASE format and can be accessed with any computer using the DOS operating system. User-friendly utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or p rinter, and to fit and plot Charpy irnpact data.

The results of several studies investigated are presented in Apperxiix D.

Publication Date:

January 1994 Prepared by:

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, BJ. [ Oak Ridge National Lab., TN (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering H

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Keywords:

Charpy test, data base management, embrittlement, power reactors, PR-EDB, reactor components, welded joints NUREG/BR--0083. Vol.10 10

SAND--88-1887 NUREG/CR--4838 General

Title:

Microcomputer Applications of, and Modi 6 cations to, the Modular Fault Trees

==

Description:==

The La Salle Probabilistic Risk Assessment was the Srst major application of the modular logic fault trees after the IREP program. In the process of performing the analysis, many errors were discovered in the fault tree modules that led to difficulties in combining the modules to form the final system fault trees.These errors are corrected in the revised modules listed in this report. In addition, the application of the modules in terms of edidng them and forming them into the system fault trees was inefficient. Originally, the editing had to be done lirn by line and no error checking was performed by the computer. ' Ibis led to many typos and logic errors in the construction of the modular fault tree files. Two programs were wrinen to help alleviate this problem: (1) MODEDIT-this program allows an operator to retrieve a file for editing, edit the file for the plant-specific application, perform some general error checking while the file is being modified, and store the file for later use; and (2)INDEX --this program checks that the modules that are supposed to form one fault tree all link up appropriately before the Bles are loaded onto the mainframe computer. Lastly, the modules were not designed for relay-type logic common in Boiling Water Reactor (BWR) designs but for solid state type logic. Some additional modules were defined for modeling relay logic, and an explanation and example of their use are included in this report.

Publication Date:

October 1994 Prepared by:

Zimmennan, T.L.: Graves, N.L.: Payne, A.C., Jr.; Whitehead, D.W. [Sandia National Labs.,

Albuquerque, NM (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution 1

Keywords:

diagrams, fault tree analysis, La Salle County-1 reactor, La Salle County-2 reactor, modifi-cations, reactor safety, risk assessment i

i 11 NUREG/BR-0083, Vol.10

NUREG/CR-5128-Rev.1 BMI--2164-Rev.1 SQUIRT-

Title:

Evaluation and Refinement ofI2ak-Rate Estimation Models. Revision 1 i

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Description:==

Leak-rate estimation models are important elements in developing a leak-before break methodology in piping integrity and safety analyses. Exisung thennal hydraulic and crack-opening-area models used in current leak-rate estirnations have been incorporated into a l

single computer code for leak-rate estimation. The code is called SQUIRT, which stands for Seepage Quanti 6 cation of Upsets In Reactor Tubes. The SQUIRT program has been validated by comparing its thermal bydraulic predictions with the limited experimental data that have been published on two-phase flow through slits and cracks and by comparing its crack-opening-area predictions with data from the Degraded Piping Program. In addition, leak-rate expenments were condacted to obtain validation data for a circumferential fatigue

- crack in a carbon steel pipe girth weld.

- Publication Date:

June 1994 Prepared by:

Paul, D.D.: Ahmad, J.: Scott, P.M.: Flanigan, L.F.; Wilkowski, G.M. [Bauelle, Columbus, OH (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Keywords:

carbon steels, cracks, elasticity, fracture mechanics, leaks, leak testing, nuclear power plants, pipes, plasticity, SQUIRT i

NUREG/BR--0083, Vol.10 12

s EGG--2577-Vol.6 NUREG/CR-5229-Vol.6 EPICOR-II

Title:

Field Lysimeter Investigations: Low Level Waste Data Base Development Program for Fiscal Year 1993. Annual Report: Volume 6

==

Description:==

The March 28, 1979, accident at Three Mile Island Unit 2 released approximately 560.000 gal of contammated water to the auxiliary and fuel handhng buildings. The water was decontaminated using a three-stage demineralization system called EPICOR-Il that contained organic and inorganic ion-exchange media. 'Ihe first stage of the system was designated the prefilter, and the second and third stages were called demineralizers. Research is being conducted at the Idaho National Engineering Laboratory on materials from four of those EPICOR II prefilters. 'Ibe Field Lysimeter Investigations: Low-1.evel Waste Data Base Development Program, funded by the U.S. Nuclear Regulatory Commission. is studying the degradation effects in EPICOR II organic ion-exchange resins caused by radiation, examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified EPICOR-II sesins, obtaining performance information on solidi 6ed EPICOR II ion-exchange resins in a disposal environment, and determining the condition of EPICOR-II liners. Results of the eighth year of data acquisition from the field testing are presented and discussed. During the conunuing field testing, both Portland type 1-11 cement and Dow vinyl i

ester styrene waste forms are being tested in lysimeter arrays located at Argonne National i

Laboratory-East in Illinois and at Oak Ridge National Laboratory. 'Ibe study is designed to provide continuous data on nuclide release and movement, as well as emironment j

conditions, over a 20-year period.

Publication Date:

May 1994 Prepared by:

McConnell, J.W., Jr.; Rogers, R.D.: Jastrow, J.D.; Sanford, W.E.: Sullivan, T.M. [EG and O j

Idaho,Inc., Idaho Falls,ID (United States)]

Prepared for:

Nucint Regulatory Commission, Washington, DC (United States). Div. of Regulatory Appheations Keywords:

antimony 125, calcium, cesium 137, chlorides, cobalt 60. EPICOR-II, ionic composition, low-level waste data base, magnesium, measuring methods, moisture, monitoring, nitrates, organic ion exchangers, performance testing, phosphates, potassium, progress report, radioactive waste disposal, radic!ysis, radionuclide migration, sodium, soil chemistry, soils, strontium 90, sulfates, temperature measurement, waste fonns, weather

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I 13 NUREG/BR--0083, Vol.10 l

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NUREG/CR-5247-Vol.1-Rev.2 ORNL--6820/V1/R2 FM. DOSE, RASCAL, ST DOSE

Title:

RASCAL Version 2.1, User's Guide

==

Description:==

'Ibe Radiological Assessment System for Consequence Analysis, Version 2.1 (RASCAL 2.1) has been developed for use during response to radiological emergencies. The system supplements assessments based on plant conditions and quick estimates based on hand-calculational methods. 'Ibe model is designed to provide a comparison to Environmental Protection Agency (EPA) Protective Action Guidance (PAG) and thresholds for acute health effects. RASCAL will be used by NRC personnel who report to the site of a nuclear accident to conduct an independent evaluation of dose-and-consequence projections and for traming and drills.The model was developed to allow consideration of the dominant aspects of source term, transport, dose, and consequences. The model is a DOS application that can be run under Windows, and the results can be displayed as text or graphics. Revisions to RASCAL 2.0 have requued the release of this new version of the system. Three new soura term calculations have been added to ST-DOSE in RASCA L 2.1. They are (1) a source tenn based f

on the reactor containment monitor reading, (2) a source term for a spent fuel pool accident, and (3) an isotopic concentration source tenn. Field Measurements to Dose (FM-DOSE) calculations have been modi 5ed to include consideration of the effect of delay for re-entry on first-year and second-year dose, to incorporate a variable resuspension rate, and to compute a factor used to estimate first-year dose from R/h measurements on the ground.

'Ibe tabular output of ST-DOSE and FM-DOSE have been changed to include EPA PAGs.

Also, RASCAL 2.1 supports the saving of cases forlater display or modification and the use of a mouse for userinput. References to the technical details are included. A RASCAL 2.1 workbook is available.

Publication Date:

December 1994 Prepared by:

Sjoreen, AA. [ Oak Ridge National Lab.,TN (United States)]; Athey, G.F. [Athey Consult -

ing, Charles Town, WV (United States)); Ramsdell, J.V. [ Pacific Northwest Lab., Richland, WA (United States)); McKenna, T. [ Nuclear Regulatory Commission, Washington, DC (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Office for Analysis and Evaluation of Operational Data l

Keywords:

computerized simulation, computer program documentation, FM-DOSE, radiation transport, RASCAL, reactor accidents, source tenns, ST-DOSE f

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NUREG/BR--0083, Vol10 14

NUREG/CR-5247-Vol.2-Rev.2 l

DECAY, FM-DOSE, RASCAL, ST-DOSE t

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Title:

RASCAL Version 2.1 Workbook. Volume 2, Revision 2 f

==

Description:==

The Radiological Assessment System for Consequence Analysis, Version 2.1 (RASCAL '

2.1) was developed for use by the NRCpersonnel who respond to radiological emergencies.

7 bis workbook complements the RASCAL 2.1 User's guide (NUREG/CR-5247, Vol.1, i

Rev. 2). 7be workbook contams exercises designed to familiarize the user with the computer-based tools of RASCAL through hanac.on problem solving. The workbook contains four

- major sections. The first is a RASCAL familiarization exercise to acquaint the user with the operation of the forms, menus, onhoe help, and documentation. 7be lauer three sections i

contain exercises in using the three tools ofRASCAL Version 2.1: DECAY, FM. DOSE, and i

ST-DOSE. A discussion section describing how the tools could be used to solve the problems i

follows each set ofexercises.

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Publication Date:

December 1994 l

Prepared by:

Athey, G.F. [Athey Consulting, Charles Town, WV (United States)]; Sjoseen, A.L. [ Oak Ridge National Lab., TN (United States)]; McKenna, TJ. [ Nuclear Regulatory Commission, i

Washington, DC (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Office for Analysis and Evaluation of Operational Data Keywords:

biological radiatien effects, computer program documentation, DECAY, emergency plans, environmental exposuse, environmental transport, FM-DOSE, half-life, manuals, nuclear decay, professional personnel, radiation accidents, radiation doses, radioisotopes, RASCAL, reactor accidents, ST-DOSE, traming, US NRC I

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l 15 NUREG/BR--0083, Vol.10 i

NUREG/CR-5344-Rev.1 ANL/EES-TM--364-Rev.1 RECAP

Title:

Replacement Energy Cost Analysis Package (RECAP): User's Guide. Revision 1

==

Description:==

A microcomputerprogram called the Replacement Energy Cost Analysis Package (RECAP) has been developed to assist the U.S. Nuclear Regulatory Commission (NRC)in detennining the replacement energy costs associated with short-term shutdowns or deratings of one or more nuclear reactors. The calculations are based on the seasonal, unit-specific cost estimates for1993-1996previouslypubhsbedinNRCReportNUREG/CR--4012 Vol.3(1992),for all 112 U.S. reactors. Because the RECAP program is menu-driven, the user can define specific case studies in terms of such parameters as the units to be included, the length and timing of the shutdown or derating period, the unit capacity factors, and the reference year for reporting cost results. In addition to simultaneous shutdous cases, more complicated situations, such as overlapping shutdown periods or shutdowns that occur in different years, can be exammed through the use of a present-worth calculation option.

Publication Date:

July 1994 Prepared by:

VanKuiken, J.C.; Willing, D.L. [Argonne National Lab., IL (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

cost, economic analysis, economics, electric utilities, energy expenses, nuclear power plants, reactor shutdown, RECAP i

b NUREG/BR--0083, Vol.10 16

UCRL-ID-118245 NUREG/CR-5403 General

Title:

Predicting the Pressure-Driven Flow of Gases through Microcapillaries and Micro-Orifices

==

Description:==

A large body ofexperunentally measured gas flow rates were obtained from the literature and then compared to the predictions obtamed with constitutive flow equations. This was done to determine whether the equations apply to the predictions of gas flow rates from leaking containment vessels used to transport radioactive materials. The experunents consisted of measuring the volumetric pressure-driven flow of gases through micro-capillaries and micro-orifices. The experimental results were compared to the predictions obtained with the equations given in ANSI N14.5, the American National Standard for Radioactive Materials-t 12akage Tests on Package for Shipment. The equations were applied to both (1) the data set according to the recommendations givenin ANSIN14.5 and (2) globally to the complete data set. It was found that the continuum and molecular flow equation provided good agreement between the eg- - =:2 and e*nhwi flow rates for flow rates less than about I atm cm /s.

3 The choked flow equation resuhedin over-prediction of the flow rates for flow rates less than 8

8 about I atm cm /s. For flow rates higher than 1 atm cm /s, the molecular and continuum flow equation over-predicted the measured flow rates and the predictions obtained with the choked flow equation agreed well with the experimental values. Because the flow rates of interest for packages used to transport radioactive muerials are almost always less than 3

1 atm cm /s,it is suggested that the continuum and molecular flow equation be used for gas flow rate predictions related to these applications.

Publication Date:

November 1994 Prepared by:

Anderson, B.L.: Carlson, R.W.; Fischer, L.E. [ Lawrence Livermore National Lab., CA (United States)]

l Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Industrial and MedicalNuclear Safety Keywords:

capillary flow, comparative evaluations, containers, flow models, flow rate, gas flow, leaks, radioactive materials, transport j

17 NUREG/BR--0083, Vol.10

NUREG/CR-5535-Vol.6 EGG-2596-Vol.6 RELAP5/ MOD 3

)

Title:

RELAP5/ MOD 3 Code Mannat Volume 6: Validation of Numerical Techniques in RELAPS/ MOD 3

==

Description:==

The RELAP code has been developed for best-estimate transient simulation oflight-water reactor coolant systems during large and small break loss-of-coolant accidents and as well as operational transients.The code models the coupled behavior ofthe reactor coolant system and tLe core during a severe accident transiem and models large-and small-break loss-of-coolan' accidents and operational transients, such as anticipated transient without scram, loss of offsit power, loss of feedwater, andloss of flow. A generic modeling approachis used that permits.ts mudi of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. RELAPS/ MOD 3 code docu-mentat;on is divided into five volumes: Volume 1 provides modeling theory and associated -

numreical schemes; Volume 2 contains detailed instmetions for code application and input data preparation: Volume 3 provides the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume 4 presents a detailed discussion of RELAP5 models and conelations; Volume 5 contains guidelines that have evolved over the past several years through use of the RELAPS code; and Volume 6 contams descriptions of numerical modeling of two-phase flow used in RELAP5 and discussions on stability, accuracy, and convergence of the numerical techniques in RELAP5.

Publication Date:

October 1994 I

Prepared by:

Shich, A.S.; Ransom, V.H. [EG and G Idaho, Inc., Idaho Falls, ID (United States));

Krishnamurthy, R. [PaciSc-Nuclear Co., Westmont, IL (United States)]

l Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

comparative evaluations, computer program documentation, flow models, manuals, numerical solution, reactor accidents, reactor control systems, reactor cooling systems, reactor cores, RELAPS/ MOD 3, two-phase flow, water cooled reactors

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l NUREG/BR--0083, Vol.10 18 i

EGG-2596-Vol.7 NUREG/CR-5535 Vol.7 RELAP5/ MOD 3

Title:

RELAP5/ MOD 3 Code Manual

==

Description:==

Summaries of RELAP5/ MOD 3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to fonnulate a compilation of some of the strengths and weaknesses of the code. These results are documemed in the report. Volume 7 was designed to be updated periodically and toinclude the results of the latest code assessments as they become available.

Consequently, users of Volume 7 should ensure that the latest revision is available.

Publication Date:

June 1994

- Prepared by:

Sloan, S.M.: Schultz, R.R.: Wilson, G.E.

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research.

Keywords:

computerized simulation, fluid flow, reactors, RELAPS/ MOD 3, thermal analysis I

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19 NUREG/BR--0083, Vol.10

NUREG/CR-5569-Rev.1 ORNLfrM--12067-Rev.1 HPPOS

Title:

Health Physics Positions Data Base: Revision I l

==

Description:==

'Ibe Health Physics Positions (HPPOS) Data Base of the Nuclear Regulatory Commission (NRC)is a collection of NRC staff positions on a wide range of topics involving ndiation protection (health physics). It consists of 328 documents in the form ofletters, memoranda, and excerpts from technical reports. 'Ibe HPPOS Data Base was developed by the NRC Headquarters and Regional Offices to help ensure uniformity in inspections, enforcement, and licensing actions. Staff members of the Oak Ridge National Laboratory (ORNL) have assisted the NRC staffin summarizing the documents during the preparation of this NUREG repon.These summaries are also being made available as a stand-alone software package for IBM and IBM-compatiMe personal computers. The software package for this report is called HPPOS Version 2.0. A variety ofindexing schemes were used to increase the usefulness of the NUREG report and its associated software. 'Ibe software package and the summaries in the report are written in the context of the new 10 CFR Part 20 (fi20.1001-20.2401). The purpose of this NUREG report is to allow interested individuals to familiarize themselves with the contents of the HPPOS Data Base and with the basis of many NRC decisions and regulations. "Ibe HPPOS summaries and original documents are intended to serve as a source of infonnation for radiation protection programs at nuclear research and power reactors, nuclear medicine, and other industries that either process or use nuclear materials.

Publication Date:

February 1994 Prepared by:

Kerr, G.D.: Borges, T.: Stafford, R.S.: Lu, P.Y. [ Oak Ridge National Lab., TN (United States)] Carter, D. [US NuclearRegulatory Commission, Washington, DC (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

data base management, HPPOS, radiation protection, recommendations, US NRC 1

i NUREG/BR -0083, Vol.10 20 n

BNL-NUREG--52319 NUREG/CR-5850 MELCOR, STCP t

Title:

Analysis of Long-Term Station Blackout Without Automatic Depressurization at Peach Bottom Using MELCOR (Version 1.8)

==

Description:==

This report documents the results from MELCOR calculations of the leng-Term Station Blackout Accident Sequence, with failure to depressunze the reactor vessel, at the Peach Bottom (BWR Mark I) plant, and presents comparimas with Source Term Code Package (STCP) calculations of the same sequence.With STCP the transient has been calculated out 4

to 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after core uncovery. Most of the MELCOR calculations presented have been camed out to between 15 and 16.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after core uncovery. The results include the release of source terms to the environment. 'Ibe results of several sensitivity calculations with MELCOR are also presented, which explore the impact of varying user-input modeling and timestep control parameters on the accident progression and release of source terms to the environment. Most of the calculations documented here were performed in FY1990 using 4

MELCOR Version 1.8BC. However,the appendices also document the results of more recent calculations performed in FY1991 using MELCOR versions 1.8CZ and 1.8DNX.

Publication Date:

May 1994 Prepared by:

Madni, I.K. [Brookhaven National Lab., Uptot. 'iY (United States)]

~

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems I.

Research Keywords:

blackouts, computer calculations, contamment, depressurization, fission product release, beat transfer, hydraulics, MELCOR, Peach Bottom-1 reactor, Peach Bottom-2 reactor, reactor accidents, reactor core disruption, reactor safety, reactor vessels, source terms, STCP, transients 4

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21 NUREG/BR--0083, Vol.10

-n NUREG/CR-5965 General

]

Title:

Modeling Field-Scale Unsaturated Flow and Transport Processes

==

Description:==

The scales of concem in subsurface transport of contaminants from low-level radioactive waste disposal facilities are in the range of I to 1000 m. Natural geologic materials generally show very suustantial spatial variabilityin hydraulic properties over this range ofscales. Such heterogeneity can significantly influence the migration of contammants It is also envisioned that complex earth structures will be constructed to isolate the waste and minumze infiltration of water into the facility. 'Ibe flow of water and gases through such facilities must also be a concern. A stochastic theory describing unsaturated flow and contammation transport in naturally heterogeneous soils has been enhanced by adopting a more realistic charactenza-

' tion of soil variability. The enhanced theory is used to predict field-scale effective properties and variances of tension and moisture content. Applications illustrate the important effects

[

of small-scale heterogeneity on large-scale anisotropy and hysteresis and demonstrate the feasibility of simulating two-dimensional flow systems at time and space scales ofinterest in radioactive waste disposal investigations. Numerical algorithms for predicting field-scrle l

unsaturated flow and contanunant transport have beenimproved by requiring them to respect fundamental physical principles such as mass conservation. These algorithms are able to i

provide realistic simulations of systems with very dry initial conditions and high degrees of heterogeneity Numerical simulation of the movement of water and air in unsaturated soils has demonstrated the importance of air pathways for contanunant transport. The stochastic flow and transport theory has been used to develop a systematic approach to performance assessment and site characterization. Hypothesis-testing techniques have been used to determine whether model predictions are consistent with observed data.

I Publication Date:

August 1994 Prepared by:

Gelhar, L.W.; Celia M.A.: McLaughlin D. [ Massachusetts Inst. ofTech., Cambridge, MA l

(United States). Dept. of Civil and Environmental Engineering]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications i

Keywords:

environmental transport, flow models, gas flow, ground water, liquid flow, low-level radioactive wastes, radioactive waste facilities, Richards equation, site characterization, y

soils, two-dtmensional calculations I

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i NUREG/BR--0083, Vol.10 22 i

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BNL-NUREG. 52353 NUREG/CR-5967 LSARG

Title:

Development and Application of Degradation Modeling To Define Maintenance Practices

==

Description:==

This report presents the development and application of compenent degradation modeling to analyze degradation effects on reliability and toidentify aspects ofmaintenance practices that mitigate degradation and aging effects. Using continuous time Markov approaches, a component degradation model is discussed that includes information about degradation and maintenance. The component model commonly used in probabilistic risk assessments is a simple case of this general m odel.The parameters used in the general modelhave engineering interpretations and can be estimated using data and engineering experience. The generation of equations for specific models, the solution of these equations, and a methodology for estimating the needed parameters are all discussed. Applications in this report show how these models can be used to quantitatively assess the benefits that are expected from maintaining a component, the effects of different maintenance efficiencies, the merits of different maintenance policies, and the interaction of surveillance test intervals with maintenance practices.

Publication Date:

June 1994 Prepared by:

Stock, D.: Samanta, P. [B rookhaven National Lab., Upton, NY (United States)); Vesely, W.

[ Science Applications International Corp., Dublin, OH (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. vf Engineering Keywords:

aging, BWR type reactors, LS ARG, Markov models, Markov process, prediction equations, PWR type reactors, reactor components, reactor maintenance, reliability 23 NUREG/BR--0083, Vol.10 i

NUREG/CR-5990 BNL-NUREG--52359 EMTP

Title:

The Effects of Solar-Geomagnetically Induced Cunents on Electrical Systems in Nuclear l

Power Stations

==

Description:==

This report presents the results of a study to evaluate the potential effects of geomagnetically induced currents (GICs) caused by the solar disturbances on the in-plant electrical distribu-tion system and equipment in nuclear power stations. The plant-specific electrical distribu-tion system for a typical nuclear plant is modeled using the Electromagnetic Transient Program (EMTP). ibe computer model simulates online equipment and loads from the station transformer in the switchyard of the power station to the safety-buses at 120 vohs to which all electronic devices are connected for plant monitoring.1he analytical model of the

- plant's electrical distribution system is studaed to identify the transient effects caused by the half-cy cle saturation of the station transformers due to GlC. This study provides results of the voltage harmonics levels that have been noted at various electrical buses inside the plant. The emergency circuits appear 9 be more susceptible to high harmonics due to the normally light load condations. In addition to steady-state analysis, this model was further analyzed simulating various plant transient conditions (e.g., loss of load or large motor start-up) occurring during GlC events. Detail models of the plant's protective relaying system employed in bus transfer application were included in this model to study the effects of the hannonic distortion of the voltage input Potential harmonic effects.:a the uninterruptible power system (UPS) are qualitatively discussed as well.

Publication Date:

January 1994 Prepared by:

Subudhi, M. [B rookhaven National Lab., Upton, NY (United States)); Carroll, D.P. [ Florida Univ., Gainesville, FL (United States)); Kastun, S. [MOS, Inc., Melville, NY (United States))

Prepared for:

Nuclear Regulatory Comnussion, Washington, DC (United States). Div. of Engineering Keywords:

Bruce site, computerized simulation, electrical equipment, EMTP, frequency analysis, geomagnetic field, harmonics, heating, Hope Creek-1 reactor, Hope Creek-2 reactor, magnetic storms, nuclear power plants, power systems, relays, Salem-1 reactor, Salem-2 reactor, Three Mile Island-l reactor,7hree Mile Island-2 reactor, transformers NUREG/BR--0083, Vol 10 24

SAND--93-1049 NUREG/CR--6044 CONTAIN

Title:

Experiments To Investigate Direct Containment Heating Phenomena with Scaled Models of the Zion Nuclear Power Plant in the Surtsey Test Facility

==

Description:==

'Ibe Surtsey Test Facility at Sandia National Laboratories (SNL)is used to perfonn scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). 'lhese experiments are designed to investigate the effect of specific phenomena associated with direct containment beating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment stmetures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel.

'Ibe reactor pressure vessel (RPV) was modeled with a steel pressure vessel that had a hemispherical bottom head with a 4-cm hole that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron / alumina /

chromium thennite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV.The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entramed through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. 'Ibe results of the IET experiments are given in this report.

Publication Date:

May 1994 Prepared by:

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquer-que, NM (United States)); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

CONTAIN, containment systems, corium, design, experimental data, measuring instruments, meltdown, pressure measurement, scale models, simulation, temperature

)

measurement, thermal analysis, two-cell adiabatic equilibrium model, Zion-1 reactor, Zion-2 reactor l

25 NUREG/BR--0083, Vol.10

NUREG/CR--6053 BNL-NUREG-52380

)

ARANO, CONDOR, COSYMA, LENA, MACCS, OSCAAR j

Title:

Comparison of MACCS Users Calculations for the Intemational Comparison Exercise on Probabilistic Accident Consequence Assessment Code, October 1989-June 1993

==

Description:==

Over the past several years, the Organization for Economic Cooperation and Development /

Nuclear Energy Agency (OECD/NEA) and Commission of the European Community (CEC) sponsored an intemational program comparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland). CONDOR (UK), COSYMA (CEC),

LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed sumlar calculations using the MACCS and COSYMA codes.

Results produced in the MACCS Users Group's (Greece. Italy, Spain, and the United States) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the speci6 cations for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatterin predictions of other consequences was success-fully explained by differences in meteorological files and weather sampling, grids, rain distance imervals, dispersion model options, and population distributions.

Publication Date:

April 1994 Prepared by:

Neymotin, L. [Brookhaven National lab., Upton, NY (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

ARANO, biological radiation effects, computer calculations, computer program documentation, CONDOR, coordinated research programs, COSYMA, crops, economic analysis, fission product release, human populations, LENA, MACCS, nuclear power plants, OSCAAR, plants, public bealth, radiation doses, radiation hazards, radiation transport, reactor accidents, risk assessment, source terms i

NUREG/BR--0083, Vol.10 26

NUREG/CR--6063 General

Title:

INTRAVAL Phase 2: Modeling Testing at the Las Cruces Trench Site

==

Description:==

Several field experiments have been performed by scientists from the University of Anzona and New Mexico State University at the Las Cruces Trench Site to provide data to test deterministic and stochastic models for water flow and solute transpon. These expenments were performed in collaboration with INTRAVAL, an international effort toward validation of geosphere models for the transport of radionuclides. During Phase I ofINTRAVAL, qualitative comparisons between expenmental data and model predictions were made using contour plots of water contents and solute concentrations. Detailed quantitative compansons were not made. A third Las Cruces Trench experiment was designed by scientists fmm the University of Arizona and New Mexico State University to provide data for more rigorous model testing. Modelers from the Center for Nuclear Waste Regulatory Analysis, Massachu-setts lastitute ofTechnology, New Mexico State University, PaciSc Northwest Laboratory, and the University of Texas provided predictions of water flow and tritium transport to New Mexico State University for analysis. The corresponding models assumed soil characteriza-tions ranging from uniform to determunstically beterogeneous to stochastic. ' Ibis report presents detailed quantitative comparisons to Seld data.

Publication Date:

January 1994 Prepared by:

Hills, R.G. [New Mexico State Univ., Las Cruces, NM (United States). Dept. of Mechanical Engineering); Wierenga, P.J. [ Arizona Univ., Tucson, AZ (United States). Dept. of Soil and Water Science); Luis, S.; McLaughlin, D. [ Massachusetts Inst. of Tech., Cambridge, MA (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

BIGFLOW, computer calculations, computer codes, experimental data, Seld tests, fluid flow, leachates, leaching, numerical analysis, PORFLO-3, radionuclide migration, soils, solutes, TRACER 3D, tritium, TRUST + TRUMP, UNSAT2, validation, VAM2D i

27 NUREG/BR--0083, Vol.10

NUREG/CR--6075 SAND--93-1535 General

Title:

'Ibe Probability of Containment Failure by Direct Contamment Heating in Zion

==

Description:==

' Ibis report is the first step in the resolution of the Direct Contamment Heating (DCH) issue for the Zion Nuclear Power Plant using the Risk Oriented Accident Analysis Methodology (ROAAM). ' Ibis reportincludes the definition of a probabilistic framewod; that decomposes the DCH problem into three probabihty density functions that reflect the most uncertain initialconditions(UO mass,23rconiumoxidationfraction,andsteelmass).Uncertaintiesin 2

the initial conditions are significant, but our quanti 6 cation approach is based on establishing reasonable bounds that are not unnemssarily conservative. To this end, we also make use of the ROAAM ideas of enveloping scenarios and splintering. Two causal relations (CR) are used in this frarneworic CRI is a model that calculates the peak pressure in the contamment as a function of the initial conditions, and CR2 is a model that returns the frequency of contamment failure as a function of pressure within the contamment. Uncertainty in CRI is accounted for by the use of two independently developed phenomenological models, the Convection Limited Contamment Heating (CLCH) model and the Two-Cell Equilibrium (TCE) model, and by probabilistically distributing the key parameter in both, which is the ratio of the melt entramment time to the system blowdown time constant. 'Ibe two phenomenological models have been compared with an extensive database including recent integral simulations at two different physical scales. The contamment load distributions do not intersect the containment strength (fragility) curve in any significant way, resulting in contamment failuse probabilities less than 10-8 for all scenarios considered. Sensitivity analyses did not show any areas oflarge sensitivity.

Publication Date:

December 1994 Prepared by:

Pilch, M.M. [Sandia National Labs., Albuquerque,NM (United States)]; Yan,H.:'Ibeofanous, T.G. [ California Univ., Santa Barbara, CA (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

ABAQUS, ALPHA, ANATECH, APRIL, blowdown, CLCH model, COMMDC, contamment, CORCON, CORMLT, DIRHEAT, distribution functions, dynamic loads, failures, HARDCORE, KIVA, MAAP, MELCOR, MELPROG, MELTSPREAD, PARSEC, pressure dependence, pressurir.ation, probability, radiation heating, reactor accidents, reactor safety, risk assessment, ROAAM, SASM, SCDAP/RELAPS, SEhTIN, sensitivity analysis, TCE model, Zion-1 reactor, Zion-2 reactor NUREO/BR--0083, Vol.10 28

f SAND--93-1535-Suppl.1 NUREG/CR-6075-Suppl.1 SCDAP/RELAP5, CONTAIN

Title:

The Probability of Containment Failure by Direct Containment Heating in Zion.

Supplement 1

==

Description:==

Supplement 1 of NUREG/CR-6075 brings to closure the DCH issue for the Zion plant. It includes the documentation of the peer review process for NUREG/CR-6075, the assess-ments of four new splinter scenarios defined in working group meetings, and modeling enhancements recommended by the working groups. In the four new scenarios, consistency of the initial conditions has been implemented by using insights from systems-level codes.

SCDAP/RELAP5 was used to analyze three short term station blackout cases with different lead rates. In all three case, the hot leg or surge line failed well before the lower head and thus the primary system depressurized to a point where DCH was no longer considered a thmat.

However, these calculations were continued tolowerhead failurein order to gaininsights that were usefulin establis'ing the initial and boundary conditions. THE SCDAP/RELAP output was used as input to CONTAIN to assess the containment conditions at vessel breach. The containment-side conditions predicted by CONTAIN are similar to those originally specified in NUREG/CR-6075.

Publication Date:

December 1994 Prepared by:

Pilch, M.M.; Allen, M.D.; Stamps, D.W.;Tadios, E.L [Sandia National Labs., Albuquerque, NM (United States)]; Knudson, D.L. [ldaho National Engineering Lab., Idaho Falls, ID (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Sysems Research Keywords:

blackouts, CONTAIN, containment, documentation, evaluation, failures, meltdown, pressure dependence, pressurization, probability, radiation heating, reactor accidents, reactor safety, SCDAP/RELAP5, Zion-1 reactor, Zion-2 :eactor 29 NUREG/BR--0083, Vol.10

NUREG/CR.-6076 ORNL/TM--12415 i

TR-EDB

Title:

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

==

Description:==

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of resuhs from r

irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data i

Base (PR EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could therefore be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing forlife extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

Publication Date:

January 1994 Prepared by:

Stallmann, F.W.; Wang, J.A.: Kam, F.B.K. [ Oak Ridge National Lab., TN (United States)]

Prepared for:

Nuclear Regulatory Comrmssion. Washington, DC (United States). Div. of Engineering Keywords:

annealing, Charpy test, data base management, embrittlement, materials testing, pressure vessels, radiation effects, reactor materials, tensile properties, test reactors, TR-EDB t

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NUREG/BR--0083 Vol.10 30

ORNL/TM -12460 NUREG/CR-6102 SCALE

Title:

Validation of the SCALE Broad Structure 44-Group ENDF/B-V Cross-Section Library for Usein Criticality Safety Analyses

==

Description:==

This report documents the validation of the recently developed 44-group ENDF/B-V based cross-section library (44GROUPNDF5).1 bis cross-section set has been developed for use in the SCALE code system in the analysis of fresh and spent fuel and radioactive waste systems. Collapsed from a 238-group Bae-structure cross-section library (238GROUPNDF5),

this broad-group library contains approximately 300 nuclides from the ENDF/B-V data files.

Additionally, ENDF/B VI oxygen data have been substituted for ENDF/B V oxygen, because of discrepancies in ENDF/B-V data. The 44 GROUPNDFS library was tested against its parent library using a set of 33 benchmark problems in order to demonstrate that the collapsed set was an acceptable representation of 238GROUPNDF5. Results show that the broad 44-group structure is an acceptable representation ofits parent 238-group library for thermal as well as hard fast spectrum systems. Accurate broad-group analyses of intermediate spectrum systems will require either a more detailed group structure in this l

energy range or a more appropriate collapsing spectrum. Further, validation calculations indicate that the 44-group library is an accurate tool in the prediction of criticality for arrays oflight-water-reactor-type fuel assemblies, as would be encountered in fresh or spent fuel transportation or storage environments. However, a bias caused by inadequate representation of plutonium cross sections was identified. Further, a possible bias exists with respect to uranium enrichment; however, experiments referenced in this report provide an inadequate sampling of uranium enrichments. Additional work will be required to quantify any bias that may be present.

Publication Date:

September 1994 Prepared by:

DeHart, M.D.; Bowman, S 'd [ Oak Ridge National Lab., TN (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

criticality, fuel assemblies, fuel storage pools, nuclear data collections, nuclear fuels, radioactive wastes, SCALE, spent fuel storage, storage, transport, validation, water cooled reactors t

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31 NUREG/BR--0083, Vol.10 w

NUREG/CR-6107 SAND-93-2042 MELCOR, STCP

Title:

Summary of MELCOR 1.8.2 Calculations foribree LOCA Sequences (AG, S2D, and S3D) at the Surry Plant

==

Description:==

Activities involving regulatory implementation of updated source term information were pursued. These activities include the identification of the source term, the identi5 cation of the chemical form ofiodine in the source term, and the timing of the source term's entrance into contamment. These activities are intended to support a more realistic source term for licensing nuclear power plants than the current TID-14844 source term and current licensing assumptions. MELCOR calculations were performed to support the technical basis for the updated source term. This report presents the results from three MELCOR calculations of nuclear power plant accident sequences and presents comparisons with Source Term Code Package (STCP) calculations forthe same sequences.The three low-pressure sequences were analyzed to identify the materials that enter containment (source terms) and are available for release to the environment and to obtain timing of sequence events. The source teuns include Sssion products and other materials such as those generated by core-concrete interactions.

All three calculations, for both MELCOR and STCP, analyzed the Suny plant, a pressunzed water reactor (PWR) with a subatmospheric containment design.

Publication Date:

March 1994 Prepared by:

Kmetyt, L. [Sandia National Labs., Albuquerque, NM (United States)]; Smith, L.

[Geo-Centen Inc., Albuquerque, NM (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

availability, biosphere, comparative evaluations, computer calculations, containment systems, fission product release, loss of coolant, MELCOR, reactor protection systems, source tenns, STCP, Surry-1 reactor, theoredcal data NUREG/BR--0083, Vol.10 32

I

~

i NUREG/CR--6114-Vol.3 SUTRA MAC I

i F

i

Title:

Performance Assessment of a Hypothetical Low-Ievel Waste Facility: Groundwater Flow

[

and Transport Simulation. Volume 3

==

Description:==

Stochastic subsurface hydrologic theory is applied to data for a hypothetical low-level radioactive waste site to demonstrate the features of the hydraulic parameter estimation process, as developed by Gelhar and others. Effective values of hydraulic conductivity, I

l macrodaspersivity, and macrodispersivity enhancement are estimated from the data in this

]

msnner A two-dimensional saturated flow and transport finite-element computer code is used to model the site. Four different isotope inputs and two types ofinput configurations contribute to an evaluation of model sensitivities. These sensitivities of the mean conantra-tions and the uncertamues around the mean are explored using an analytical model as an j

example. Results indicate that the spatial heterogeneity of isotope sorption, through its l

1 contnbution to longitudinal di p.ivity +=L r ---:nt, has a large effeet on the magnitude of concentration predictions, especially forisotopes with short half-lives in companson to their retarded mean travel times. This observation emph== the need for accurate site data measurements that compliment the parameter estimation process. A comparison of simpli-

)

fied analytical screening models with the numerical model predictions shows that the analytical models tend to underesumate concentration levels at low times, potentially as a i

result of oversimplification of the flow field. Future models could address aspects that are l

neglected in this report, such as three-dimensionality or unsaturated flow and transport.

Publication Date:

May 1994 1

Prepared by:

Talbott, M.E.: Gelbar, L.W. [ Massachusetts Inst. ofTech., Cambridge, MA (United States).

j Ralph M. Parsons Iab.)

r 1

I Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory l

Applications t

Keywords:

adsorption, envuonmental transport, experimental data, flow models, geolczy, ground

)

water, hydrology, low-level radioactive wastes, radioactive waste disposal, radioactive waste facilities, sensitivity analysis, site characterization, strontium 90, SITTRA MAC, technetium 99, two-dunensional calculations, underground disposal, uranium 238 i

l t

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5 i

h I

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33 NUREG/BR--0083, Vol.10

, ~,

i NUREG/CR--6116-Vol.1 EGG--2716-Vol.1 IRRAS, SAPHIRE, SARA i

Title:

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE)

Version 5.0, Technical Reference Manual

==

Description:==

Ibe Systems Analysis Programs for Hands-onIntegrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze I

probabilistic risk assessments (PRAs), primarily for nuclear power plants. ' Ibis volume provides information on the principles used in the construction and operation of Version 5.0 of the Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and i

Risk Assessment (SARA) system. It summarizes the fundamental mathematical concepts of r

j sets and logic, fault trees, and probability. This volume then describes the algorithms that these programs use to constmet a fault tree and to obtain the mmunal cut sets. It gives the formulas used to obtain the probability of the top event from the munmal cut sets and the

?

formulas for probabilities that are appropriate under various assumptions conceming 1

repairability and mission time. It defines the measures of basic event importance that these programs can calculate.This volume gives an overview of uncertamty analysis using simple Monte Carlo sampling or Latin Hypercube sampling and states the algorithms used by these programs to generate random basic event probabilities from various distributions. Funber references are given, and a detailed example of the reduction and quantification of a simple fault tree is provided in an appendix.

Publication Date:

July 1994 Prepared by:

Russell, K.D.: Atwood, C.L.: Galyean, WJ.; Sattison, M.B. [EG&G Idaho, Inc., Idaho Falls, ID (United States)); Rasmuson, D.M. [ Nuclear Regulatory Commission, Washington, DC (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

fault tree analysis, IRRAS, Latin Hypercube sampling, Monte Carlo method, nuclear power plants, sisk assessment, safety analysis, SAPHIRE, SARA l

1

^

i NUREG/BR--0083, Vol.10 34 P

EGG--2716-Vol.2 NUREG/CR--6116-Vol.2 IRRAS, SAPHIRE

Title:

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE),

Version 5.0: Integrated Reliability and Risk Analysis System (IRRAS) Reference Manual.

Volume 2

==

Description:==

'Ibe Systems AnalysisPrograms forHands-onIntegra'edReliabilityEvaluations(SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Integrated Reliability and Risk Analysis System (IRRAS)is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the ability to create and analyze fault trees and accident sequences using a microcomputer. ' Ibis program provides functions that range from graphical fault tree constmetion to cut set generation and quantificatiot to report generation. Version 1.0 of the IRRAS program was released in February of 1987. Since then, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system.

This version has been designated IRRAS 5.0 and is the subject of this Reference Manual.

Version 5.0 ofIRRAS provides the same capabilities as earlier versions and adds the ability to perform location transformations and seismic analysis and provides enhancements to the userinterface as well as improved algorithm perfonnance. Additionally, version 5.0 contains new alphanumeric fault tree and event used for event tree rules, recovery rules, and end state partitioning.

Publication Date:

July 1994 Prepared by:

Russell, K.D.; Kvarfordt, KJ.; Skinner, N.L; Wood, S.T. [EG and G Idaho, Inc.. Idaho Falls, ID (United States)); Rasmuson, D.M. [ Nuclear Regulatory Commission, Washington, DC (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div, of Safety Issue Resolution Keywords:

algorithms, computer program documentation, failure mode analysis, fault tree analysis, IRRAS, manuals, nuclear power plants, reactor safety, reliability, risk assessment, SAPHIRE, systems analysis 35 NUREG/BR--0083, Vol.10

r NUREG/CR--6116-Vol3 EGG--2716-Vol3 IRRAS, SAPHIRE

Title:

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE),

Version 5.0: Integrated Reliability and Risk Analysis System (IRRAS) Tutorial Manual.

Volume 3

==

Description:==

The Systems Analysis Programs forHands-onIntegrated Relisbility Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs),primarily for nuclearpowerplants. This volumeis the tutorial manual forthe Integrated Reliability and Risk Analysis System (IRRAS) Version 5.0, astate-of-the-art microcomputer-basedprobabilisticriskassessment(PRA)modeldevelop-ment and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the ability to create and analyze fault trees and accident sequences using a microcomputer. A series oflessons is provided that guides the userthrough basic steps common to most analyses performed with IRRAS. *Ibe tutorialis divided imo two major seetions: basic and additional features.The basic section contmim lessons that lead the student through development of a very simple problem in IRRAS, hipli ehting the program's most basic features. The mMitional features section contains lessons that expand on basic analysis features ofIRRAS 5.0.

Publication Date:

July 1994 Prepared by:

VanHom, R.L; Russell, K.D.; Skmner, N.L [EG and G Idaho,Inc., Idaho Falls, ID (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety issue Resolution Keywords:

automation, computer-aided design, computer program documentation, IRRAS, man-machire systems, manuals, nuclear power plants, risk assessment, SAPHIRE NUREG/BR -0083.Vol.10 36

P EGG-2716-Vol.5 NUREG/CR--6116-Vol.5 SAPHIRE, SARA

Title:

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE),

Version 5.0. Volume 5, Systems Analysis and Risk Assessment (SARA) Tutorial Manual

==

Description:==

The Systems Analysis Programs for Hands-onIntegrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputerprograms that were developed to create and analyze probabilistic risk assessments (PRAs)primarily for nuclearpower plants.This volume is the tutorial manual for the Systems Analysis and Risk Assessment (SARA) System Version 5.0, a microcomputer. based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series oflessons is provided that guides the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis capabilities.

Publication Date:

July 1994 Prepared by:

Sattison, M.B.; Russell, K.D.: Skinner, N.L. [EG and G Idaho,Inc., Idaho Falls,ID (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

manuals, probabilistic estimation, risk assessment, safety engineering, SAPHIRE, SARA i

i i

37 NUREG/BR--0083, Vol.10 m -

+

NUREG/CR--6116-Vol.7 EGG--2716-Vol.7 FEP,IRRAS, SAPHIRE

Title:

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE)

Version 5.0. Fault Tree, Event Tree, and Piping and Instmmentation Diagram (FEP) Editors Reference Manual: Volume 7 4

==

Description:==

The Systems Analysis Programs for Hands-onIntegrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Fault Tree, Event Tree, and Piping and Instrumentation Diagram (FEP) editors allow the user to graphically build and edit fault trees, event trees, and piping and instrumentation diagrams (P and ids). The software is designed to enable the independent use of the graphical-based editors found in the Integrated Reliability and Risk Assessment System (IRRAS). FEP is comprised of three separate editors (Fault Tree, Event Tree, and Piping and Instrumentation Diagram) and a utility module." Ibis reference manual provides a screen-by-screen guide of the entire FEP System.

Publication Date:

July 1994 Prepared by:

McKay, M.K.; Skmner, NL; Wood, S.T. [EG and G Idaho,Inc., Idaho Falls, ID (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

computer graphics, fault tree analysis, FEP, IRRAS, manuals, nuclear power plants, risk assessment, SAPHIRE, text editors NUREG/BR--0083, Vol.10 38

i EGG--2716.Vol 8 NUREG/CR-6116-Vol.8 General

Title:

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE),

Version 5.0: Models and Results Data Base (MAR-D) Reference Manual. Volume 8

==

Description:==

The Systems Analysis Programs for Hands-onIntegrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The primary function of MAR-D is to create a data repository for completed PRAs and Individual Plant Exammations (IPEs) by providing input, conversion, and output capabilities for data used by IRRAS, SARA, SETS, and FRANTIC software. As probabilistic risk assessments and individual plant exanunations are submitted to the NRC for review, MAR-D can be used to convert the models and results from the study for use with IRRAS and SARA. 'Iben, these data can be easily accessed by future studies and will be in a form that will enhance the analysis process. This reference manual provides an overview of the functions available within MAR-D and step-by-step operating instructions, i

Publication Date:

July 1994 Prepared by:

Russell, K.D.; Skinner, N.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States))

J Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

data base management, FRANTIC, IRRAS, manuals, MAR-D, nuclear power plants, reactor safety, reliability, risk assessment, SAPHIRE, SARA, SETS l

t i

39 NUREG/BR--0083, Vol.10

NUREG/CR-6120 General

Title:

Controlled Field Study for Validation of Vadose Zone Transport Models

==

Description:==

Prediction of radionuclide migration through soil and ground water requires models that have been tested under a variety of conditions. Unfortunately, many of the existing models have not been tested in the 6 eld, partly because such testing requires accurate and represen-tative data. This r port provides the design of a large-scale 6 eld experiment representative, in terms of surface area and depth of vadose zone, of an actual dispos I area. Experiments are proposed that will yield documented data of sufficient scale to allow testing of a variety of models including effective media stochastic models and deterministic models. Details of the methodology and procedures to be used in the experiment are presented.

l Publication Date:

August 1994 Prepared by:

Wierenga, PJ.; Warrick, A.W.; Yeh, T.C. [ Arizona Univ., Tucson, AZ (United States));

Hills, R.G. [New Mexico State Univ., Las Cruces, NM (United States). Dept. of Mechanical Engineering]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

geology, ground water, low-level radioactive wastes, model validation, Monte Carlo method, radioactive waste disposal, radionuclide migration, saturation, soils, statistical models l

NUREG/BR--0083 Vol.10 40

BNL-NUREG-523%

NUREG/CR--6128

[

General

Title:

Piping Benchmark Problems for the ABB/CE System 80+ Standardized Plant

==

Description:==

To satisfy the need for verification of the computer programs and modeling techniques that j

will be used to perform the final piping analyses for the ABB/ Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems i

are representative ofpiping systems subjected to representative dynamicloads with solutions developed using the methods being proposed for analysis for the System 80+ standard design.

It will be required that the combined liansees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time-history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System F0+ pi sog benchmark is a time-i history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactoris an advanced PWR type.

Publication Date:

July 1994 Prepared by:

Bezier, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K. [Brookhaven National lab., Upton, NY (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering a

Keywords:

ADLPIPE, auxiliary water systems, compiled data, computer-aided design, NUPIPE, pipes, PIPSYS, PISTAR, PISYS, primary coolant circuits, PSAFE2, PWR type reactors, j

SAPIV, standardization, SUPERPIPE, supports, validation, water hammer, WECAN l

1 4

i J

41 NUREG/BR--0083, Vol.10

NUREG/CR-6133 ANL-93/32 THIRMAL

Title:

Fragmentation and Quench Behavior of Corium Melt Streams in Water

==

Description:==

The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a cnicial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues that arise during the molten core relocation include the thermal attack and possible damage to the reactor pressure vessel (RPV) lower head from the impinging molten fuel stream and/or the debris bed; the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals; the quench rate of the molten fuel through the water in the lower plenum; the steam generation and hydrogen generation during the interaction; the transient pressurization of the primary system: and the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through 6) was perfonned in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam / hydrogen generation, and the posttest debris characteristics.

Publication Date:

February 1994 Prepared by:

Spencer, B.W.; Wang, K.; Blomquist, C.A.; McUmber, L.M. [Argonne National Lab., IL

{

(United States)); Schneider, J.P. [Ilhnois Univ., Urbana, IL (United States). Dept. of Nuclear Engineering]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

cooling.corium,flowmodels.fluidflow. fragmentation,heattransfer hydraulics. hydrogen, interactions, mehdown, mixing, nuclear power plants, quenching, reactor safety, steam, testing.THIRMAL, water t

4 a

NUREG/BR--0083 Vol.10 42

r EGG--2719 NUREG/CR-6138 CEMENT

Title:

User's Guide for Simplified Computer Models for the Estimation of Long-Tenn Performance of Cement-Based Materials t

==

Description:==

This report documents user instructions for several simplified subroutines and driver programs that can be used to estimate various aspects of the long-term performance of cement-based barriers used in low-level radioactive waste disposal facilities. The subrou-tines are prepared in a modular fashion to allow flexibility for a variety of applications. Three levels of codes are provided: the individual subroutines, interactive drivers for each of the subroutines, and an interactive main driver, CEMENT, that calls each of the individual drivers. The individual subroutines for the different models can be taken indepetxiently and used in larger programs, or the driver modules can be used to execute the subroutines separately or as part of the main driver routine. A brief program description is included and user-interface instructions for the individual subroutines are documented in the main report.

These are intended to be used when die subroutines are used as subroutines in a larger computer code.

Publication Date:

February 1994 Prepared by:

Plansky, L.E.: Seitz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

CEMENT, cements, computerized simulation, containment shells, low-level radioactive wastes, radioactive waste storage l

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e 43 NUREG/BR--0083, Vol.10

NUREG/CR-6145 EGG--2713 FEP, IRRAS, MAR-D, SAPHIRE, SARA

Title:

Verification and Validation of the SAPHIRE Version 4.0 PRA Software Package

==

Description:==

A veri 5 cation and validation (V&V) process has been performed for the System Analysis Programs for Hands-on Integrated Reliability Evaluation (SAPHIRE). SAPHIRE is a set of four computer programs that the Nuclear Regulatory Commission (NRC) developed to perform probabilistic risk assessments (PRAs). These programs allow an analyst to create, quantify, and evaluate the risk associated with a facility or process being analyzed. The programs included in this set are Integrated Reliability and IUsk Analysis System (IRRAS),

System Analysis and Risk Assessment (SARA), Models and Results Database (MAR-D),

and Fault Tree / Event Tree / Piping and Instrumentation Diagram (FEP) graphical editor. The V&V steps included a V&V plan to describe the process and criteria by which the V&V would be performed; a software requirements documentation review to determine the correctness, completeness, and traceability of the requirements: a user survey to determine tie usefulness of the user documentation; identification and testing of vital and non-vital features; and documentation of the test results.

Publication Date:

February 1994 Prepared by:

Bolander T.W.: Calley, M.B.: Capps, E.L. [EG and G Idaho, Inc., Idaho Falls,ID (United States)] [and others)

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution Keywords:

documentation, fault tree analysis, FEP, IRRAS, MAR-D, nuclear power plants, performance testing, reactor safety, risk assessment, SAPHIRE, S ARA, systems analysis, validation l

l l

NUREG/BR--0083. Vol.10 44 L.

~

NUREG/CR-6147-Vol.1 General

Title:

Charactenzation of Oass-A Low-level Radioactive Waste 1986-1990. Volume 1: Executive Summary

==

Description:==

Under contract to the U.S. Nuclear Regulatory Commission, oftice of Nuclear Regulatory Research, the firms of S. Cohen & Associates. Inc. (SC&A) and Eastern Research Group (ERG) have compiled a rrpon that describes the physical, chemical, and radiological propenies of Class-A low-level radioactive waste. 'lhe repon also presents infonnation characterizing various methods and facilities used to treat and dispose of non-radioactive waste. A database management program was developed for use in accessing, sorting, analyzing, and displaying the electronic data provided by EG&G. The program was used to present and aggregate data characterizing the radiological, physical, and chemical propenies of the waste from descriptions contained in shipping manifests. The data tiras retrieved are summarized in tables, histograms, and cumulative distribution curves presenting radionu-clide concentration distributions in Qass-A waste as a function of waste streams, by category of waste generators, and by regions of the United States.The repon also provides information characterizing methods and facilities used to treat and dispose of non-radioactive waste, includingindustrial, municipal, and hazardous waste regulated under Subparts C and D of the Resource Conservation and Recovery Act (RCRA). The information includes a list of disposa! options, the geographical locations of the processing and disposal facilities, and a description of the characteristics of such processing and disposal facilities. Volume 1 contains the Executive Summary, Volume 2 presents the Cass-A waste database, Volume 3 presents the information characterizing non-radioactive waste management practices and facilities, and Volumes 4 through 7 contain Appendices A through P with supporting information.

Publication Date:

January 1994 Prepared by:

Dehmel, J.C.; Loomis, D.; Mauro, J. [S. Cohen & Associates, Inc., McLean, VA (United i

States)]; Kaplan, M. [ Eastern Research Group, Inc., lexington, MA (United States)]

Prepared for:

Nuclear Regulatoty Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

chemical propenies, compiled data, low-level radioactive wastes, physical propenies.

radioactive waste management, radiochemical analysis 45 NUREG/BR--0083, Vol.10

i NUREG/CR--6160 NEA/CSNI/R--94-3: EGG--2721 SCDAP/RELAPS

Title:

Summary of Important Results and SCDAP/RELAPS Analysis for OECD LOFT Experi-ment LP-FP-2

==

Description:==

This report summarizes significant technical findings from the LP-FP-2 Experiment spon-sored by the Organization of Economic Cooperation and Development (OECD). It was the second, and final, fission product experiment conducted in the less-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory. The ovemil technical objective of the test was to contnbute to the understandmg of fuel rod behavior, hydrogen generation, and fission product release, transport, and deposition during a V-sequena accident scenario that resultedin severe core damage. An 11 by 1 I test bundle, comprised of100prepressurized fuel rods,11 control rods, and 10 instrumented guide tubes, was surrounded by an insulating shroud and contained in a specially designed central fuel module that was inserted into the LOFT reactor. The simulated transient was a V-sequence loss-of-coolant accident scenario featuring a pipe break in the low-pressure injection system line attached to the bot leg of the LOFT brokenloop piping. The transient was terminated by reflood of the reactor vessel when the outer wall shroud temperature reached 1517 K. With sustained fission power and beat from oxidation and metal-water reactions, elevated temperatures resulted in zircaloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. A description and evaluation of the major phenomena, based upon the response of online instrumentation, analysis of Sssion product data, postirradiation exami-nation of the fuel bundle, and calculanons using the SCDAP/RELAPS computer code, are presented.

Publication Date:

April 1994 Prepared by:

Coryen, E.W. [EG and G Idaho, Inc., Idaho Falls ID (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

computerized simulation, ECCS, loss of coolant, PWR type reactors, reactor accidents, SCDAP/RELAPS, thermal analysis NUREG/BR--0083, Vol.10 46

NUREG/CR--6161 IS--5103 ABAQUS, BOSOR4, BOSOR5

Title:

Buckling Evaluation of System 80+ Containment

==

Description:==

ne System 80+= containment may be subjected to compressive forces that could cause it to become unstable. The stability of the containment shell under prescribed loading combinations was investigated with two analysis levels: axisymmetric and three dimen-sional. An axisymmetric shell model, including addnional mass to account forpenetrations and the spray header system, was analyzed using BOSOR4 and BOSORS finite diffemoce codes. Loading combinations with pressure, temperature, self weight, and seismic data r.tisfied the American Society of Mechanical Engineers (ASME) stress allowables. "Ibe buckling assessment wasperformed using the worst meridian assumption, including material nonlinearities atxt a sinusoidal axisymmetric imperfection. The minimum factor of safety for Service level C was 2.35. A Safe Shutdown Earthquake (SSE) seismic margin of 2.91 was calculated. 'Ibe ABAQUS finite element code was selected for the three-dimensional analysis and tested with classical and BOSOR solutions.The maximum structural response was computed using response spectrum analysis and six potentisl buckling regions were identified. A set of equivalent static loads was determined for each of the six regions to regenerate the maximum SRSS stress resultants. For each region, combined loads were increased until an instability was detected. A minimum factor ofsafety of 1.91 was predicted, which does not satisfy ASME Section NE3222.1 or Regulatory Guide 1.57. Code Case N.

284 is satis 6ed. The analysis is conservative primarily because the SPSS 10% method provides a conservative estimate of model coupling.

Publication Date:

August 1994 Prepared by:

Greimann, L.: Fanous, F.; Safar, S.; Challa, R.; Bluhm, D. [Ames Lab., IA (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Keywords:

ABAQUS, BOSOR4, BOSOR5, containment shells, deformation, pressure dependence, responsefunctions,seismiceffects stability, steels,stressanalysis,temperaturedependence, three-dimensional calculations l

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47 NUREG/BR--0083, Vol.10 1

UILU-ENG--93-2014 NUREG/CR--6162 General

Title:

Numerical Modeling of Ductile Tearing Effects on Oeavage Fracture Toughness j

==

Description:==

Experimental studies demonstrate a significant effect of specunen size, a/W ratio, and prior ductile tearing on cleavage fracture toughness values (J.) measured in the ductile-to-brittle i

I transition region of ferritic materials. In the lower-transition region, cleavage fracture often occurs under conditions oflarge-scale yielding but without prior ductile crack extension. The increased toughness develops when plastic zones formed at the crack tip interact with nearby specimen surfaces which relaxes crack-tip constraint (stress triaxiality). In the mid-to-upper transition region, small amounts of ductile crack extension (often < l-2 mm) routinely precede termmation of the J-A a curve by brittle fracture. Large-scale yielding, coupled with small amounts of ductile tearing, magnifies the impact of small variations in microscale material properties on the macroscopic fracture toughness, which contributes to the large amount scatter observed in measured J,-values. Previous work by the authors described a micromechanics fracture model to correct measured J,-values for the mechanistic effects of large-scale yielding. This new work extends the model to also include the influence ofductile l

crack extension prior to cleavage. The paper explores development of the new model, provides necessary graphs and procedures forits application, and demonstrates the effects of d

the model on fracture data sets for two pressure vessel steels (A533B and A515).

Publication Date:

May 1994 Prepared by:

Dodds, R.H., Jr.; Tang. M. [ Univ. ofIllinois, Urbana (United States)]; Anderson, T.L. [ Texas A&M Univ., College Station, TX (United States)]

Prepared for:

Illinois Univ., Urbana, IL (United States)

Keywords:

ductility, ferritic steels, finite element method, tensile properties NUREG/BR--0083, Vol.10 48

NUREG/CR--6180 LA--12741-M HMS

Title:

Hydrogen Mtung Studies (HMS) User's Manual

==

Description:==

Hydrogen Mtung Studies (HMS)is a best-estimate analysis tool for predicting the transpon, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and other facilities. It can model geometrically complex facilities that have multiple compart-ments and internal structures. The code can simulate the effects of steam condensation, heat transfer to walls and intemal structures, chemical kinetics, and fluid turbulence. 'Ibe gas mixture may consist of components included in a built-in library of 20 species. HMS is a finite-volume computer code that solves the time-dependent, three-dimensional (3-D) compressible Navier-Stokes equations. Both Canesian and cylindrical coonlinate systems are available. Transport equations for the fhad intemal energy and for gas species densities are also solved. HMS was originally developed to run on Cray-type supercomputers with vector-processing units that greatly improve the computational speed, erpecially for large, complex problems. Recently the code has been converted to run on Sun workstations. Both the Cray and Sun versions have the same built-in graphics capabihties that allow 1-D,2-D, 3-D, and time-history plots of all solution variables. Other code features include a restart capability and flexible definitions ofinitial and time-dependent boundary conditions. This manual describes how to use the code. It explains how to set up the model geometry, define walls and obstacles, and spectfy gas species and materialproperties. Definitions ofinitial and boundary conditions are also described. The manual also describes various physical model and numerical procedure options, as well as how to tum them on. The reader also leams how to specify different outputs, especially graphical display of solution variables. Fmally, sample problems are included to illustrate some applications of the code. An input deck that illustrates the minimten required data to run HMS is given at the end of this manual.

Publication Date:

December 1994 Prepared by:

Lam, K.L.: Wilson, T.L.: Travis, J.R.

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

atom transport, combustion, computerized simulation, containment buildings, fission products, HMS, hydrogen, manuals, mixing, Navier-Stokes equations, reactor accidents 49 NUREG/BR--0083, Vol.10

ORNLfrM -12263/V1 NUREG/CR-6182-Vol.1 CSASIN, OFFSCALE, SCALE

Title:

OFFSCALE: A PC Input Processor for the SCALE Code System. The CSASIN Processor for the Criticality Sequences

==

Description:==

OFFSCALE is a suite of personal computer input processor programs developed at Oak Ridge National Laboratory to provide an easy-to-use interface for modules in the SCALE-4 code system. CSASIN (formerly known as OFFSCALE) is a program in the l

OFFSCALE suite that serves as a user-friendly interface for the Crincality Safety Analysis Sequences (CSAS) available in SCALE-4. It is designed to assist a SCALE-4 user in preparing an input file for execution of criticahty safety problems. Output from CSASIN generates an input file that may be used to execute the CSAS control module in SCALE-4 CSASIN features a pulldown menu system that accesses sophisticated data entry screens.

The program allows the user to quickly set up a CSAS input file and perform data checking.

This capabihty increases productivity and decreases the chance of user error.

Publication Date:

November 1994 Prepared by:

Bowman, S.M. [ Oak Ridge National Lab., TN (United States))

Prepared for:

Nuclear Regulatory Commission, Washingtc>u, DC (United States). Div. ofIndustnal and MedicalNuclear Safety Keywords:

computerized simulation, computer program documentation, criticality, CSASIN, OFFSCALE, SCALE, spent fuels NUREG/BR--0083, Vol.10 50

NUREG/CR--6182-Vol.2 ORNL/TM--12263/V2 ORIGEN-S, ORIGNATE, SCALE

Title:

OFFSCALE A PC Input Processor for the SCALE Code System. The ORIGNATE Processor for ORIGEN-S

==

Description:==

OFFSCALE is a suite of personal computer input processor programs developed at l

Oak Ridge National Laboratory to provide an easy-to-use interface for modules in the SCALE-4 code system. ORIGNATE is a program in the OFFSCALE suite that serves as a user-friendly interface for the ORIGEN-S isotopic generation and depletion code. It is designed to assist an ORIGEN-S user in preparing an input 61e for execution oflight water reactor (LWR) fuel depletion and decay cases. ORIGNATE generates an input file that may be used to execute ORIGEN-S in SCALE-4. ORIGNATE features a pulldown menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up an ORIGEN-S input file and perform error checking. This capability increases productivity and decreases the chance of user error.

Publication Date:

November 1994 Prepared by:

Bowman, S.M. [ Oak Ridge National Lab., TN (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of ladustrial and Medical Nuclear Safety Keywords:

computerized simulation, computer program documentation, decay, ORIGEN-S, ORIGNATE, SCALE, spent fuels, transmutation 4

d i

i 51 NUREG/BR--0083, Vol.10 j

BNL-NUREG--52412 NUREG/CR-6200 CSAU

Title:

Uncenainty Analysis of Suppression Pool Heating During an ATWS in a BWR-5 Plant. An Application of the CSAU Methodology Using the BNL Engineering Plant Analyzer

==

Description:==

ne uncenainty of predicting tb-peak temperature in the suppression pool of a BWR power plant that undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS) has been estimated. De ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the La Salle-2 Power Station in March of 1988.

After limit. cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser.

Postulated operator actions to lower the reactor vessel pressure and the level elevationin the downcomer are simulated by a robot model that accounts for operator uncertainty. All balance of plant and control systems modeling uncertaindes were pan of the statistical i

i uncenainty analysis that was patterned after the Code Scaling, Applicability and Uncenainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 'F) has a 95% uncenainty of 14.4 K (26 T), and that the size of this uncenainty bracket is dominated by the experimental uncenamty of measunng safety and relief valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 T) is most likely zero (it is estimated as < 5-104). The squam root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 'F).

Publication Date:

March 1994 Prepared by:

Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States)]; Johnsen, G.W. [ldaho National Engineering Lab., Idaho Falls,ID (United States)];

lellouche, G.S. (Technical Data Services, Chicago,IL (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

ATWS, BWR type reactors, CSAU, evduation, pressure suppression, transients NURECdBR--0083, Vol.10 52

NUREG/CR--6203 NEFTRAN, NEWBALANCE, PHREEQE, VTOUGH

Title:

Validation Studies for Assessing Unsaturated Flow and Transpon ibrough Fractured Rock

==

Description:==

The objectives of this contract are to examme hypotheses and conceptual models concernmg unsaturated flow and transport through heterogeneous fractured rock and to design and execute confirmatory field and la'uoratory experiments to test these hypotheses and concep-tual models. Imponant new infonnation is presented, such as the application and evaluation of procedmes for estimating hydraulic, pneumatic, and solute transpon coefficients for a range of thermal regimes. A field heaterexperiment was designed that focused on identifying the suitability of existing monitoring equipment to obtain required data. A reliable method was developed for conducting and interpreting tests for air permeability using a straddle-packer arrangement. Detailed studies of fracture flow from Queen Ocek into the Magina Copper Company ore haulage tunnel have been initiated. These studies will provide data on travel time for transpon of water and solute in unsaturated tuff. The collection of rainfall runoff and infiltration data at two small watersheds at the Apache Leap Tuff Site allowed evaluation of the quantity and rate of water infiltrating into the subsurface via either fractures or matrix. Characterization methods for hydraulic parameters relevant to weigh-level waste transpon, including fracture apenures, transnussivity, matrix porosity, and fracture wetting front propagation velocities, were developed.

Publication Date:

August 1994

)

Prepared by:

Bassett, R.L; Neuman, S.P.; Rasmussen, T.C.; Guzman, A.; Davidson, G.R.; Lohrstorfer,

{

C.F. [ Arizona Univ., Tucson, AZ(United States). Dept. ofHydrology and Water Resources]

i Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

carbon, fluid flow, fluid mechanics, geologic fractures, high-level radioactive wastes, hydraulics, NEFTRAN. NEWBALANCE, permeability, PHREEQE, porosity, radiation transport, radioactive waste disposal, radionuclide migration, saturation, tuff, vapors, VTOUGH 53 NUREG/BP--0083, Vol.10

ORNLfrM--12693 NUREG/CR--6206 ENDF/B-VI, SAILOR

Title:

Transport Calculations of Radiation Exposure to Vessel Suppon Structures in the Trojan Reactor

==

Description:==

Comparison of transport calculations of the dosimeter activities with the expenmental measurements shows that the values obtained with ENDF/B-VI cross-section data overestimate the measured results for high-energy threshold reactions in the cavity by up to 41 % and thermal reactions by up to a factor of 3.0. The transpon calculations performed with the original SAILOR cross-section library (based on ENDF/B VI data) overestimate mea-sured threshold reactions by only 15% and the thermal reactions by about a factor of 2.50.

"Ibese results are inconsistent with those obtained in earlier studies that compared transport calculations done with SAILOR vs. ENDF/B-VI, which indicate that SAILOR tends to underestimate cavity dosimeter activities for threshold reactions, whereas the ENDF/B-VI values usually agree better with experimental results. One factor that probably contributes to the rather large discrepancy between the computed and measured activities is the core power distribution used in the transport calculations. Because of the unavailabihty of plant-specific j

data, a generic power distribution provided by Westinghouse was used. Since the calculated cavity flux levels appear to be overestimated, the results estimated for the exposure to the support structure should be conservative.

Publication Date:

July 1994 Prepared by:

Asgari, M.; Wilhams, M.L. [ Louisiana State Univ., Baton Rouge, LA (United States).

Nuclear Science Center); Kam, F.B.K. [ Oak Ridge National Lab., TN (United States)];

McGarry, E.D. [Nadonal Inst. of Standards and Technology, Gaithersburg, MD (United States))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Keywords:

calculationmethods, cavities,computercalculations, dosimetry,ENDF/B VI,neutronflux, nuclear data collections, power distribution, radiation transport, SAILOR Trojan reactor NUREG/BR--0083, Vol.10 54

h NUREG/CR--6211 SAND 0406 IFCI

Title:

Integrated Fuel-Coolant Interaction (IFO 6.0) Code. User's Manual

==

Description:==

The lategrated Fuel-Coolant Interaction (IFO) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolantinteraction (FO) problem at large scale using a two-dimensional, four-Beld hydrodynamic framework and physically based models.

IFQ will be capable of treating all major FC processes in an integrated manner. This document is a product of the effort to generste a stand-alone version ofIFC,IFQ 6.0. 'Ibe User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual,provided for the creation of working decks.

Publication Date:

April 1994 Prepared by:

Davis, F.J.: Young, M.F. [Sandia National Labs., Albuquerque, NM (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

computer program documentation, corium, explosions, fuel-coolant interactions, beat transfer, hydraulics, IFO, jets, mixing, nuclear power plants, reactor accidents, reactor safety 4

l 1

55 NUREG/BR--0083, Vol.10

-)

j

CNWRA--93-024 NUREG/CR-6216 UDEC

Title:

Evaluation of Rock Joint Models and Computer Code UDEC Against Experimental Results

==

Description:==

The Mohr-Coulomb,B arton-Bandis, and Continuously Yielding rockjoint models and their numericalimplementation in the UDEC code were evaluated for their ability to simulatejoint behavior under cyclic pseudostatic and dynamic loading conditions. Some de6ciencies of these joint models and their implementation in UDEC were identi6ed. These deficiencies include that the rockjoint models under evaluation may not be able to suf6ciently predict the joint she ar and dilation behavior during reversejoint shearing. Bothjoint forward and reverse shearing are important phenomena of a rockjoint behavior. Reverse sheanng can result from earthquakes or thermal load, both of which are expected to be experienced during the life of a high-level waste repository. These deficiencies could result in an overestimation of the stability ofemplacement drifts and emplacement boretioles and prediction ofincorrect near-6 eld flow pattem (including preferential pathways for water and gas).

Publication Date:

November 1994 Prepared by:

Hsiung,S.M.; Ghosh, A.; Chowdhury, A.H.; Abola,M.P. [ Southwest Researchinstitute, San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses)

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Keywords:

computerized simulation, cracks, fracture mechanics, geologic sssures, geologic fractures, high-level radioactive wastes, radioactive waste disposal, radioactive waste storage, rocks, site characterization, UDEC, underground storage NUREG/BR--0083, Vol.10 56

s i

NUREG/CR--6218 SAND--94-0731 MELCOR

Title:

A Review of the Technical Issues of Air Ingression During Severe Reactor Accidents

==

Description:==

Severe reactor accident scenarios involving airingression into the reactor coolant system are described. Evidence from modem reactor accident analyses and from the accident at Three Mile Island show that residual fuel will be present in the core region when air ingression is possible. This residual fuel can interact with the air. Exploratory calculations with the MELCOR code of station blackout accidents during shutdoum conditions and during operations are used to examme clad oxidation by air and ruthenium release from fuelin air.

Extensive ruthenium release is predicted when air ingression 24tes exceed about 10 mol/s.

Past studies of airinteractions with irradiated reactor fuel are reviewed. Effects airingression may have on fission product release, tr.nsport, deposition and revaporization are discussed.

Perhaps the most important effects of airingression are expected to be the enhanced release of ruthenium from the fuel and the formation of copious amounts of aerosol from uranium oxide vapors. Revaporization ofiodine and tellurium retained in the reactor coolant system might be expected.

1 Publication Date:

September 1994 Prepared by:

Powers, D.A.: Kmetyk, L.N.: Schmidt, R.C.

Prepared for:

Nuclear Regulatory Commission, Washingion, DC (United States). Div. of Systems Research Keywords:

computerized simulation, fission product release, fuel cans, MELCOPs meltdown, oxidation, reactor accidents, reactor cores i

4 i

i 57 NUREG/BR--0083, Vol.10 1

PU.NE-93/1 NUREG/CR-6267

Title:

Air-Water Simulation of Phenomena of Conum Dispersion in Direct Containment Heating

==

Description:==

The presem research at Purdue addresses corium dispersion during the direct containment heatingin reactor severe accidents. The degree ofcorium dispersion has not only the strongest parametric effect on containment pressurization but also has the highest uncertainty in predictingit. In view of this, a separate effect test program on corium dispersion mecharn cms in the reactor cavity and the subcompartment trappmg mecharnsms was initiated in spring of 1992 at Purdue under the direction of the Nuclear Regulatory Commission. Four major objectives of this corium dispersion study are (1) to perform a detailed scaling study using thenewlyproposedstep-by stepintegralscalingmethod,thentoevaluateexistingmodelsfor entratament, particle size, and trapping; (2) to perform carefully designed simulation experiments using water-air and Wood's metal-air in a 1/10 linear scale model; (3) to develop reliable mechanistic models and correlations for corium dispersions that can be used to predict coriumjet disintegration, entramment, drop size, liquid film carry over, and subcompartment trapping; and (4) to use the models to perform stand-along calculations for typical prototypic conditions. The combination of water-air and Wood's metal-air as scorking fluid will give a unique data base over broad parametric ranges that can be used together with the integral test results to develop reliable models and correlations. The results of the experiments that were conducted using air-water are presented.

1 i

Publication Date:

October 1994 Prepared by:

Ishii, M.; Revankar, S.T.; 2 hang, G.; Wu, Q.; O'Brien, P. [Purdue Univ., Lafayette, IN (United States). School of Nuclear Engineering)

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

blowdown, computerized simulation, containment systems, corium, dispersions, distribution, experimental data, beating, meltdown, parametric analysis, pressurization, primary coolant circuits, PWR type reactors, scale models, test facilities 1

NUREG/BR--0083, Vol.10 58 l

1

NUREG/CR--6269 LA--12853-MS CATHENA, TRAC-PF1/ MOD 2

Title:

A Plan for the Modification and Assessment of TRAC-PFI/ MOD 2 for Use in Analyzing CANDU 3 Transient Thermal-Hydraulic Phenomena

==

Description:==

This repon presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-bydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PFI/ MOD 2 (TRAC) computer code needed to adequately predict CANDU 3 transient thermal-bydraulic phenomena, and identify an assessment program to verify the ability of TRAC, when modined, to predict these phenomena. This work builds on analyses and recommendations produced by the Idaho National Engineering Labocatory (INEL).

To identify the thermathydraulic phenomena, a large-break loss-of-coolant accident simu-lation, performed as part of earlier work by INEL with an Atomic Energy of Canada, Ltd.

(AECL) thennal-bydraulic computer code (CATHENA), was analyzed in detail. Other accident scenarios were exammed for additional phenomena. A group of Los Alamos National Laboratory reactor thermal-bydraulics experts ranked the phenomena to produce a prelinunary phenomena identification and rankmg table (PIRT). 'Ibe preliminary nanne of i

the PIRT was a result of alack of direct expertise with the unique processes and phenomena of the CANDU 3. Nonetheless, this PIRT provided an adequate foundation for planning a program of code modiScations. We believe that this PIRT captured the most important phenomena and that tefinements to the PIRT will mainly produce clarification of the relative importance (ranbng) of phenomena. A plan for code modifications was developed based on this PIRT and on infonnation about the modeling methodologies for CANDU-speciSc 1

phenomena used in AECL codes. AECL thermal-bydraulic test facilities and programs were reviewed and the information used in developing an assessment plan to ensure that TRAC-PFl/ MOD 2, when modified, will adequately predict CANDU 3 phenomena.

Publication Date:

November 1994 Prepared by:

Siebe, D.A.; Boyack B.E.; Giguere, P.T.

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research Keywords:

CANDU type reactors, CATHENA, computerized simulation, beat transfer, loss of coolant, performance, thennal analysis, TRAC-PF1/ MOD 2 59 NUREG/BR--0083, Vol.10

UCRL-ID-117524 NUREG/CR-6278 General

Title:

Survey ofIndustry Methods for Producing Highly Reliable Software

==

Description:==

7be Nuclear Reactor Regulation Of6ce of the U.S. Nuclear Regulatory Commission is charged with assessing the safety of new instrument and control designs for nuclear power plants that may use computer-based reactor protection systems. Lawrence Livermore National Laboratory has evaluated the latest techniques in software reliability for measure-ment, estimation, error detection, and prediction that can be used during the software life cycle as a means of risk assessment for reactor protection systems. One aspect of this task has I

been a survey of the software industry to collectinformation to helpidentify the design factors used to improve the reliability and safety of software. The intent was to discover what practices really wort in industry and what design factors are used by industry to achieve highly reliable software. The results of the survey are documented in this report. Three companies participated in the survey: Computer Sciences Corporation, Intemational Busi-ness Machines (Federal Systems Company), and TRW. Discussions were also held with NASA Software Engineering Lab / University of Maryland /CSC and the AIAA Software Reliability Project.

Publication Date:

November 1994 Prepared by:

Lawrence, J.D.; Persons, W.L. [ Lawrence Livermore National Lab., CA (United Stues))

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Reactor Controls and Human Factors Keywords:

computerized control systems, programming, quality assurance, reactor control systems, reactor instrumentation, reactor protection systems, reliability, risk assessment NUREG!BR--0083, Vol.10 60

a NUREG/CR--6289 PNL--10193 CRAS, GENII,ISOSHLD, TOXCHEM

Title:

Reconcentration of Radioactive Material Released to Sanitary Sewers in Accordance with 10 CFR Part 20

==

Description:==

The U.S. Nuclear Regulatory Commission (NRC),in accordance w.i h 10 CFR 20 and state t

regulations, regulates the discharge of radioactive materials into sanitary sewer systems.

A one-year study was conducted by Pacific Northwest Laboratory (PNL) for the NRC to assess whether radioactive materials that are discharged to sanitary sewer systems undergo significant reconcentration within the wastewater treatment plants (WWTP) and to l

determine the physical and/or chemical processes that may result in radionuclide reconcentration within the WWTPs. The study objectives were addressed by collecting information and data on wastewater treatment, relevant geochemical processes, and individual radionuclide behavior in WWTPs from the open literature, NRC reports, EPA surveys, and interviews with NRC licensees and staff of W%TPs that may be affected by these discharges. Radio'2uclide mass balance and removal efficiencies were calculated for WWTPs at Oak Ridge, Tennessee, and Erwin, Tennessee, but were not shown to be reliable since the licensee release data generally underestimated the mass of radionuclide that was ultimately found in the sludge. This disparity may be due, in pan, to the fact that data available for use in this study were collected to address regulatory concems and not to perfonn mass balance calculatiots. A limited modeling study showed some promise for predicting radionuclide behavior in %%TPs, however, the general applicability of these I

empirical models remains uncertain. With the data and models currently available,it is not possible to quantitatively determine the physical and chemical processes that cause reconcentration or to calculate, a priori, reconcentration factors for specific WWTP unit processes or W%TPs in general.

Publication Date:

December 1994 Prepared by:

Ainsworth, C.C.; Hill, R.L.; Cantrell, K.L; Kaplan, D.L; Norton, M.V.; Aaberg, R.L [ Pacific Northwest Lab., Richland, WA (United States)]; Stetar, E.A. [ Performance Technology Group, Inc., Nashville, TN (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications Keywords:

abundance, biogeochemistry, CRAS, GENII, ground disposal, ISOSHLD, radiation monitoring, radioactive effluents, radioecological concentration,TOXCHEM, waste water, water treatment plares 61 NUREG/BR--0083, Vol.10

NUREG/CR--6290 KEY

Title:

KEY Analysis System User's Guide. Version 2.0

==

Description:==

The KEY analysis system is a software program designed to process digital waveform data from the United States National Seismograph Network. The KEY system performs many data processing and scientiSc analysis functions. Detailed operating procedures for the KEY analysis system are provided in this User's Guide.

Publication Date:

November 1994 Prepared by:

Masse, R.P. (Geological Survey, Denver, CO (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Keywords:

data analysis, data processing, KEY. seismic waves, seismopaphs, statistical models l

NUREG/BR--0083, Vol.1C 62

NUREG/CR--6294 General

Title:

Design Factors for Safety-Critical Software

==

Description:==

1 bis report, the fourth of a series ofreports prepared forthe Nuclear Regulatory Comnussion Of6ce ofReactor Regulation, provides the summary and conclusion for this task. It is widely believed in the software engineering community that almost anything can affect the ability of software to reliably perform its tasks, panicularly when safety is at issue. Although this statement is true in both the abstract and spectfic instances, it is not particularly helpful. It remains necessary for auditors and other reviewers to assure themselves and the public that safety cntical software has a suf5ciently low probability of failing in such a way as to cause death orinjury to permit it to be used in safety-critical applications. Achieving this assurance is best done by using a well-planned, methodical approach. A possible approach is to concentrate on those attributes of the software and the development process (design factors) that are most inDuential in achieving dependable software. Seventy-four design factors are identi6ed in this report, divided into nine categories. Seven categones relate to the develop-ment process, and one category relates to the products of that process. The remaining category contains negative factors whose presence should be regarded as cause for intense scrutiny of the development process. Seven of the design factors should be considered mandatory for any organization responsible for developing safety-critical software. An additional nine factors are considered essential to safety but not as important as the first seven. 'Ibe remaining design factors can provide additional important indications of the quality of the development effort and the software resulting from that effort.

Publication Date:

December 1994 Prepared by:

Lawrence, J.D.; Preckshot, G.G. [ Lawrence Livermore National Lab., CA (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Reactor Controls and Human Factors Keywords:

design, nuclear power plants, quality assurance, reactor safety, reliability i

1 l

l l

1 63 NUREG/BR--0083, Vol.10

l UCRL-ID-119239 NUREG/CR--6303 ESFAS

Title:

Method for Perfonning Diversity and Defense-In-Depth Analyses of Reactor Protection Systems

==

Description:==

'Ibe purpose of this NUREG is to describe a method for analyzing computer-based nuclear reactor protection systems that discovers design vulnerabilities to common-mode failure.

7be potential for common-mode failure has become an important issue as the software content of protection systems has increased. This potential was not present in earlier analog protection systems because it could usually be assumed that common-mode failure,ifit did occur, was the result of slow processes such as corrosion or premature wear-out. This assumption is no longer true for systems containing software. Itis the purpose of the analysis method described here to determine points of a design for which credible common-mode failures are uncompensated either by diversity or defense-indepth.

Publication Date:

December 1994 Prepared by:

Prectshot, G.G. [ Lawrence Livermore National Lab., CA (United States)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Div. of Reactor Controls and Human Factors Keywords:

BWR type reactors, computerized control systems, computerized simulation, defects, ESFAS, failure mode analysis, implementation, PWR type reactors, reactor protection systems, recommendations l

t i

NUREG/BR--0083 Vol.10 64 m

I NUREG/IA -0093 i

RELAPS/ MOD 3

Title:

RELAPS/ MOD 3 Assessment for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads. Intemational Agreement Report

==

Description:==

This report presents an assessment study forthe use of the code RELAP 5/ MOD 3/5M5 in the calculation of transient hydrodynamic loads on safety and relief discharge pipes. Its predecessor, RELAP S/ MODI, was found adequate for this kind of calculations by Electric Power Research Institute (EPRI). The hydrodynamic loads are very important for the discharge piping design because of the fast opening of the valves and the presence ofliquid

)

in the upstream loop seals. The code results are compared to expenmentalload measurements performed at the Combustion Engineering Laboratory in Windsor, Connecticut, United States. Those measurements were part of the PWR Valve Test Program undertaken by EPRI after the TMI-2 accident. This particular kind of transient challenges the applicability of the following code models: two-phase choked discharge, interphase drag in conditions withlarge density gradients, beat transfer to metallic structures in fast changing conditions, and two.

phase flow at abrupt expansions. The code applicabihty to this kind of transient is investi-gated. Some sensitivity analyses to different code and model options are perfonned. Finally, the suitability of the code and some modeling guidelines are discussed.

Publication Date:

Febmary 1994 Prepared by:

Stubbe, E.J.; VanHoenacker, L; Otero, R. [TRACTEBEL, Brussels (Belgium)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). Of5ce of Nuclear Regulatory Research Keywords:

computerized simulation, discharge canals, hydrodynamics, modifications, pipes, primary coolant circuits, PWR type reactors, reactor safety, RELAP5/ MOD 3, relief valves, test facilities, theoretical data, transients 65 NUREG/BR--0083, Vol.10

NUREG/IA--0114 RELAP5/ MOD 3

Title:

Assessment of RELAP5/ MOD 3 with the LOFT L9-1/L3-3 Experiment Simulating an Anticipated Transient with Multiple Failures

==

Description:==

De RELAP5/ MOD 3/5M5 code is assessed using the L9-1/L3-3 test carried out in the LOFT facility, a 1/60-scaled experimental reactor, simulating a loss of feedwater accident with multipb failo res and the sequentiallyinduced small-breakloss-of-coolant accident.The code predictabihty is evaluated for the four separated sub-periods with respect to the system response: initial heatup phase, spray and power operated relief valve (PORV) cyQog phase, blowdown phase, and recovery phase. On the basis of the comparisons of the ri. ults from the calculation with the experiment data, it is shown that the overall thermal-hydraulic behavior important to the scenario, such as a heat removal between the primary side and the secondary side and a system depressurization, can be well predicted and that the code could be applied to the full-scale nuclear power plant for an anticipated transient with multiple failures with a reasonable accuracy.He minor discrepancies between the prediction and the experiment are identified in reactorscram time, post-scram behaviorin the initial beatup phase, excessive beatup rate in the cycling phase,insuf5cient energy convected out the PORV under the hot leg stratified condition in the saturated blowdown phase, and void distribution in secondary side in the recovery phase. This may come from the code uncertainties in predicting the spray mass Dow rate, the associated condensation in pressurizer, and junction fluid density under stratified condition.

4 Publication Date:

February 1994 l

Prepared by:

Bang, Y.S.; Seul, K.W.; Kim. HJ. [ Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)]

Prepared for:

Nuclear Regulatory Commission, Washington, DC (United States). OfSce of Nuclear Regulatory Research Keywords:

computerized simulation, loss of coolant, PWR type reactors, reactor accidents, RELAP5/ MOD 3 l

i l

l NUREO/BR--0083, Vol.10 66

APPENDIX A: Index by NUREG-Series Report Number Report Number Pace NUREG-1502.._.... --

.1

=_

NUREG/CP--0127..

._2 NUREG/CP--0133.VoL2 _...

3 NUREG/CP--0138............. _-

4 NUREG/CP-0139..

5 NUREG/CP-0145.-

6 NUREG/CR--4409-Vol.5...... :

.. 7 NUREG/CR--4639-Vol.5 Rev.4-Pt.2 -

..... 8 NUREG/CR--4639-Vol.5-Rev.4-Pt.3.......

9 NUREG/CR--4816-Rev.2 10 NUREG/CR--4838...........

-.11 NUREG/CR-5128-Rev.1 -

. 12 NUREG/CR--5229-VoL6........

13 NUREG/CR-5247-Vol.1.Rev.2...... _-

14 NUREG/CR-5247-Vol.2-Rev.2.....

- 15 NUREG/CR-5344-Rev.1.. --

... 16 NUREG/CR-5403.........._..

.17 NUREG/CR--5535-VoL6.

18 NUREG/CR-5535-Vol.7._..... --

19 NUREG/CR-5569-Rev.1.......

..._. 20 NUREG/CR-5850........ ~.

21 NUREG/CR-5%5..........

22 NUREG/CR--5967

........... 23

... =

NUREG/CR-5990 -

- 24 NUREG/CR--6044

.... 25

=

NUREG/CR--6053......

..._ 26 NUREG/CR--6063...

... 27 NUREG/CR--6075..

. 28 NUREG/CR-6075-Suppl.1.....

... -. 29 NUREG/CR-6076......

-......_.. 30 NUREG/CR--6102

... 31 l

NUREG/CR--6107............

32 NUREG/CR--6114-VoL3

.... _.. 33

. -..... ~.

NUREG/CR--6116-VoLI....

34 NUREG/CR--6116-Vol.2.

.. -. 35 NUREG/CR--6116-VoL3.

... 36 NUREG/CR--6116-Vol.5 -

- 37

. =

NUREG/CR--6116-VoL7.....

38 NUREG/CR--6116-Vol.8.......

.. 39 NUREG/CR--6120...........

40 NUREG/CR--6128..,

41 NUREG/CR--6133......._

... 42

=-

NUREG/CR--6138.. _...........

43 4

NUREG/CR--6145

..............44 NUREG/CR--6147-VoL1....-

45 NUREG/CR--6160..

46 NUREG/CR--6161.....

-.47 NUREG/CR-6162

... 48 A-1 NUREG/BR--0083, Vol.10

Report Number Page NUREG/CR -6180.

49 NUREG/CR--6182-Voll.

- 50

- 51 NUREG/CR--6182-VoL2-NUREG/CR -6200 -

- 52 NUREG/CR--6203 -

..... 53 NUREG/CR--6206 -

- 54 NUREG/CR--6211 ;

. 55 NUREG/CR--6216..

-- 56 NUREG/CR--6218 -

57 NUREG/CR--6267.

- 58 NUREG/CR-6269..

-- 59

- 60 NUREG/CR -6278

= 61 NUREG/CR--6289.

62 l

NUREG/CR--6290...,

i NUREG/CR--6294.....

- 63 l

NUlGG/CR-6303.

. 64

- 65 l

NUREG/1A-0093.

.. 66 NUREG/1A--0114.

NUREG/BR--0083, Vol.10 A-2

I APPENDIX B: Index by Software Identification t

i Software Ident$ cation Page ABAQUS...-

28,47 ACE..

-7 ACEFAX.

7 y

1 ADINA..

.. 3 ADLPIPE 41

.~

ALPHA..

.3,5,.28 ANATECH...~.

m

... 28 y

4 3,5,28 ARANO 26 i

ARROTTA 1

BASSIM.

.3 1

BIGFLOW 27 BOSOR4-47 4.

BOSOR5.......

47 CASMO

.. 1

. ~

... -. ~... _ ~.

CATHENA._.._..

.._. 1, 59 CEMENT

._ 43 -

CERBERUS...

1 CHYMES 2

COMETA

...........5 i.

U a*

b

..... J, 4.0 i

CO ACT....-

. ~......

s C O N D O R.. _....._........

. 26 i

4 CONTAIN.

1,3,5,25,29

,4 CORCON 1,3,5,28 CORMLT..

.....=

.._........... 3, 28 CORSOR...._.......-._._

-3 1

CO >YMA 26 CPM.2 1

CRAS 61 CSASIN 50 CSAU 1,52 CULDESAC--

2

.m DEBRIS

=.

.._.. 3 DECAY.

. 15 DIF3D

..... 1 DIRHEAT _..._ -

28 SCCS.

..... 46 EMTP.._-

24 ENDF/B.V1

... 54 EP1COR.H.....

... 13 4 -

ESFAS........ _.

. 64 ESPROS~..

c n.

EVNTRE..~...

_3 FAVOR.

.5 FEAT

-2

... ~...

FEP..

38,44 FLOW 3D....

2 4 -

B.1 NUREG/BR-0083, Vol.10

Software Meetscation Page FLUTAN

._ _.... 3 FM. DOSE _....

.. 14, 15 FRANTIC..

39 FRAPCON 1

FRAP.T._

... 1 GENII 61 G O T HI C.

......._..._............3 HARDCORE

.........~......................

,8 HELIOS

... ~... I HMS........_....._........_............_..49 HPPOS 20 IDEMO.

.2 IPCI.........

2,3.5,55 IRRAS..

34,35,36,38,39,44 ISOSHLD 61 IVA.3

....._...2 key...

. _. 62 KIVA......

28 1

LENA..

. 26 LOWCORV

......~.....3 LSARG..

.........~....3 i

MAAP..

. _ 3,5,28 5

MACCS..

26 MA. PHY. BURN.

.3 1

..... ~

MAR.D _ _

39,44 t

i MCNP....

.1 i

MELCOR.......

..._...1,3,5,21,28,32,57 t

. 3.

MELPROG.

t MF1TSPREAD

............ 28 NARAL.5M.

._.3 NEFTRAN....

53 N

.1 NEWBALANCE..

53 NUCLARR,..

.9 NUPIPE.

.m

.._ 41 OFFSCALE

...~. _ _.

... ~.. 5 0 ORIGEN-S...

51 ORIGNATE...

51

=

OSCAAR.

.26 PARSEC..

28

__ 53 PHREEQE PIPSYS 41 PISTAR 41 PISYS 41 2

PM. ALPHA PORFLO-3 27 PR.EDB

_... 10 PSAFE2 41 RALOC.

3 RASCAL 14.15 V..

.e..

...e.e...

....e..8..

j NUREG/BR--0083 Vol.10 B.2

= - _ - - -_.

1, i

i SeRware Edication Page

)

.i

. ~

16 RELAP/ MOD 3

..... 5 RELAP5..

1 RELAPS!MODn.

.._......2 P

Ja. LAP 5/ MOD 3 5,18,19,65,66 RETRAN............

3 ROAAM.

2,3,28

. ~.. ~..

S

_.. 3 i

S R....

.. S4 t

SN.

5,34,35,36,37,38,39,44 S APIV

. 41

. SARA......

........... 34,37,39,44 SASM..

....~ _....._ 28 SCALE..._...

31,50,51 4

SCANAIR.

=-

......._........3 r

SCDAP.

1 S

,5,28,39,46 SENKIN.

..... 28 f

t SETS

.39 1

. SPARC

..............3 i

i SQUIRT.

...... _. 12 l

STANJAN -

..........3 STCP 21,32 ST. DOSE 14,15 i

S E.

41 SUTRA MAC..

33

.r m,U e

mvs

_....c.....

J TCE.

28 TEXAS.III 2

THIRMAL.

42 THIRMAL.1.... -

._....._.....9 G

2

,I TOXOIEM

_. 1 TRACER 3D

.. 7 TRACG _

.........3

]

TRAC-P

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

. ~ _....

TRAC-PF1/ MOD 2..

59 TR.EDB..

_.... 30 TRIO.MC............

2

.... _....... ~. -

TRUST + TRUMP...

._.~... 27

.......... ~.

UDEC_.....-.......~..............-.._...._..-.

. 56 UNSAT2 27

....... _...... ~

VAM2D

... 27 VICTORIA

................1,3 VTOUGH..

. 53 WAVQ..

..........3 WECAN 41 WIMS

..............1 B.3 NUREG/BR.-0083, Vol.10 J

APPENDIX C: Index by Contractor Report Number Report Number Page ANL--93/32.

4.,

ANL/EES.TM-364.Rev.1._.._

16 BMI-2164.Rev.1....

... 12 z.

BENUREG--51934-Vol.5.

....7 BENUREG-52319....

...... 21 BENUREG-52353

.. 23 BENUREG-52359 BENUREG-52380 _

--24 26 BENUREG-52396..

41 BENUREG-52412 52 CNWRA-.93-024._......

..... 56 CONF.9207249..

6

=.

CONF.930157 2

i CONF.9310377

- _. 4 CONF.931079-VoL2.

..........3 CCNF.9410216.. _

.._. 5 EGG.-2458-Vol.5-Rev.4-PL2 -

....._.......8

_. ~..

EGG-2458-VoL5.Rev.4-PL3.

-.... 9 EGG-2577-Vol.6 _....... _...

..13 EGG-2596.Vol.6....

. 18 EGG-2596-VoL7............

19 EGG-2713.

44 EGG-2716-Vol.1.

34 EGG-2716-VoL2..

. 35 l

EGG-2716-VoL3

..... 36

. 37 EGG-2716-VoL5 -

EGO-2716.VoL7 38 EGG-2716-Vol8.....

39 EGG-2719...........

3 4

EGG-2721....

.... 46 i

IS-5103....

47 j

LA-12741-M

- 49 l

LA-12853-MS _.

. 59 NEA/CSNI/R-(93)8.

z NEA/CSNI/R-94-3..

.. 46 ORNL-6820/V1/R2 --

. 14 l

ORN1/TM-10328/R2.-...

_..... 10 ORN1/TM-12067-Rev.1

... 20 ORIJ1/TM-12263/V1.

_ -.. 50 ORNI/TM-12263/V2....

._. 51 ORN1/TM-12415 30 i

ORNI/TM-12460

.... 31 ORN1/TM-12693...

..... 54 PNL-10193.....

61 l

PU NE-93/1......

.58 j

SAND-8 8-1887..

... 1 1 SAND 93 1049......~...

.. 25 i

SAND-93-1535...

.......... 2 8 SAND-93-1535. Suppl.1..

-.......... 29 C-1 NUREG/BR--0083. Vol.10

Report Number Page 32 SAND-93 2042....

55 SAND-94-0406..

=

57 SAND.94 0731......................

60 UCRL.ID-117524........

UCRL.ID-.118245.....................

..I7 UCRL.1D-- 1 1923 9..........................-.....

....................64 UILU.ENG-93 2014.

........ 4 8 NUREG/BR-0083, Vol.10 C2

APPENDIX D:Index by Keyword Keyword NUREG Report Number ABAQUS NUREG/CR-6075 NUREG/CR--6161 abundance NUREG/CR--6289 accidents NUREG/CR a409-VoLS accuracy NUREG/CP-0138 ACE NUREG/CR--4409-VoL5 ACEFAX NUREG/CR-4109-VoL5 ADINA NUREG/CP-0133-VoL2 ADLPIPE NUREG/CR--6128 adsorption NUREG/CR--6114-VoL3 aging NUREG/CP--0139 NUREG/CR-5%7 algorithms NUREG/CR--6116-Vot2 ALPHA NUREG/CP-0133 VoL2 NUREG/CP--0139 NUREG/CR--6075 ANATECH NUREG/CR--6075 anucaling NUREG/CR--6076 antimony 125 NUREG/CR-5229-Vol.6 APRIL NUREG/CP-0133 VoL2 NUREG/CP-0139 NUREG/CR-6075 ARANO NUREG/CR--6053 ARROITA NUREG-1502 atom transport NUREG/CR-6180 ATWS N17 REG /CR--6200 automation NUREG/CR--6116-VoL3 D-1 NUREG/BR--0083, Vol.10

Keyword NUREG Report Number auxiliary water systems NUREG/CR--6128 availability NUREG/CR--6107 BASSIM NUREG/CP-0133-VoL2 BIGFLOW NUREG/CR--6063

)

biogeochemistry NUREG/CR--6289 biological radiation effects NUREG/CR-5247-Vol.2-Rev.2 NURFG/CR--6053 biosphere NUREG/CR--6107 blackouts NUREG/CR-5850 l

NUREG/CR-6075-Suppl.1 blowdown NUREG/CR--6075 NUREG/CR--6267 BOSOR4 NUREG/CR--6161 BOSOR5 NUREG/CR--6161 Bruce site NUREG/CR-5990 BWR type reactors NUREG/CP-0127 NUREG/CP-0133-Vol.2 NUREG/CP--0139 NUREG/CR--4409-VoL5 NUREG/CR--5967 NUREG/CR--6200 NUREG/CR -6303 calcium NUREG/CR-5229-VoL6 calculation methods NUREG/CR-6206 CANDU type reactors NUREG-1502 NUREG/CR--6269 capillary flow NUREG/CR-5403 carbon NUREG/CR--6203 carbon steels NUREG/CR-5128-Rev.1 CASMO NUREG-1502 CATHENA NUREG-1502 NUREG/CR--6269 NUREG/BR--0083, Vol.10 D-2 L______...

Keyword NUREG Report Number cavities NUREG/CR--6206 CEMENT NUREG/CR--6138 mments NUREG/CR--6138 CERBERUS NUREG-1502 cesium 137 NUREG/CR-5229-VoL6 Charpy test NUREG/CR--4816-Rev.2 NUREG/CR-6076 chemicalproperties NUREG/CR-6147-VolI chlorides NUREG/CR-5229-VoL6 l

CHYMES NUREG/CP-0127 CLCH model NUREG/CP-0133-Vol.2 NUREG/CR--6075 cobalt 60 NUREG/CR-5229-VoL6 combustion NUREG/CP-0133-VoL2 N'JREG/CR--6180 COMETA NUREG/CP-0139 COMMIX NUREG/CP-0133-Vot2 NUREG/CR-6075 COMPACT NUREG/CP-0133-VoL2 comparative evaluations NUREG/CR-5403 NUREG/CR-5535-Vol.6 NUREG/CR--6107 compiled data NUREG/CR.-4639-VoL5-Rev.4-Pt.2 NUREG/CR--4639-VoL5-Rev.4-Pt.3 NUREG/CR--6128 NUREG/CR--6147-VolI computer-aided design NUREG/CR--6116-VoL3 NUREG/CR--6128 computer calculations NUREG/CR-5850 NUREG/CR-6053 NUREG/CR--6063 NUREG/CR--6107 NUREG/CR--6206 D-3 NUREG/BR--0083, Vol.10

Keyword NUREG Report Number compmer codes NUREG/CR--6063 compmer graphics NUREG/CR--6116-VoL7 compmerized control systems NUREG/CR--6278 NUREG/CR--6303 computerized simulation NUREG/CR-5247-VoL1-Rev.2 NUREG/CR-5535-VoL7 NUREG/CR-5990 NUREG/CR--6138 NUREG/CR--6160 NUREG/CR-6180 NUREG/CR -6182-VoL1 NUREG/CR--6182-VoL2 NUREG/CR--6216 4

NUREG/CR--6218 NUREG/CR--6267 NUREG/CR--6269 NUREG/CR--6303 NUREG/IA--0093 NUREG/IA--0114 i

computer program documentation NUREG/CR-5247-Vol.1-Rev.2 j

NUREG/CR-5247-VoL2-Rev.2 NUREG/CR-5535-VoL6 s

NUREG/CR--6053 NUREG/CR--6116-VoL2 NUREG/CR--6116 VoL3 NUREG/CR--6182 VoL1 l

NUREG/CR--6182-VoL2 i

NUREG/CR--6211 i

CONDOR NUREG/CR--6053 CONTAIN NUREG-1502 NUREG/CP-0133-VoL2 NUREG/CP--0139 NUREG/CR--6044 NUREG/CR--6075-Suppl 1 l

containers NUREG/CR-5403 I

contamment NUREG/CP-0133-VoL2 T

NUREG/CR-5850 t

NUREG/CR--6075 NUREG/CR -6075-Suppl.1 r

contamment buikhngs NUREG/CR--6180 f

i containment shells NUREG/CR--6138 NUREG/CR--6161 i

NUREG/BR--0083, VoL10 D-4 I

Keyword NUREG Report Number containment systems NUREG/CR--6044 NUREG/CR--6107 NUREG/CR--6267 cooling NUREG/CR--6133 coordinated research programs NUREG/CR--6053 CORCON NUREG-1502 NUREG/CP-0133-VoL2 NUREG/CP-0139 NUREG/CR--6075 corium NUREG/CP-0127 NUREG/CP-0133-VoL2 NUREG/CR--6044 NUREG/CR--6133 NUREG/CR--6211 i

NUREG/CR--6267 CORMLT NUREG/CP-0133-VoL2 NUREG/CR--6075 CORSOR NUREG/CP-0133-Vot2 cost NUREG/CR-5344-Rev.1 COSYMA NUREG/CR--6053 CPM-2 NUREG-1502 cracks NUREG/CR-5128-Rev.1 NUREG/CR--6216 CRAS NUREG/CR--6289 criticality NUREG/CR--6102 NUREG/CR--6182-Vol.1 crops NUREG/CR--6053 CSASIN NUREG/CR--6182-VoL1 CSAU NUREG-1502 NUREG/CR--6200 CULDESAC NUREG/CP-0127 data analysis NUREG/CR--6290 data base management NUREG/CR--4409-Vol.5 NUREG/CR--4816-Rev.2 D-5 NUREG/BR--0083, Vol.10

Keyword NUREG Report Number data base management (conunued)

NUREG/CR-5569-Rev.I NUREG/CR--6076 NUREG/CR-6116-Vol.8 data compilation NUREG/CR. 4409-Vol.5 data processing NUREG/CR--6290 DEBRIS NUREG/CP-0133-VoL2 DECAY NUREG/CR-5247-Vol.2-Rev.2 NUREG/CR--6182-Vol.2 defects NUREG/CR--6303 deformation NUREG/CR -6161 depressurization NUREG/CR-5850 design NUREG-1502

)

NUREG/CR--6044 NUREG/CR--6294 detonations NUREG/CP-0133-Vol.2 diagrams NUREG/CR--4838 DIF3D NUREG-1502 f

DIRHEAT NUREG/CR--6075 disdiarge canals NUREG/IA-0093 dispersions NUREG/CR--6267 distribution NUREG/CR.-6267 distribution functions NUREG/CR--6075 documentation NUREG/CR-6075-Suppl.1 NUREG/CR--6145 dosimetry NUREG/CR--6206 l

ductility NUREG/CR--6162 dynamicloads NUREG/CR.-6075 ECCS NUREG/CR--6160 economic analysis NUREG/CR--5344-Rev.1 NUREG/CR--6053 NUREG/BR--0083, Vol.10 D-6

Keyword NUREG Report Number economics NUREG/CR-5344-Rev.1 elasticity NUREG/CR-5128-Rev.1' electricalequipment NUREG/CR-5990 electric utilities NUREG/CR-5344-Rev.1 embrittlement NUREG/CR4816-Rev.2 NUREG/CR--6076 emergency plans NUREG/CR-5247-Vol.2-Rev.2 EMTP NUREG/CR-5990 ENDF/B-VI NUREG/CR-6206 energy expenses NUREG/CR--5344-Rev.1 engineered safety systems NUREG/CP-0139 NUREG/CP-0145 environmental exposure NUREG/CR-5247-Vol.2-Rev.2 environmentalimpacts NUREG/CP-0138 environmental transport NUREG/CR-5247-Vol.2-Rev.2 NUREG/CR-5%5 NUREG/CR--6114-VoL3 EPICOR-II NUREG/CR-5229-Vol.6 equipment NUREG/CR--4639-Vol.5-Rev.4-Pt.3 ESFAS NUREG/CR--6303 ESPROSE NUREG/CP-0127 evaluation NUREG-1502 NUREG/CR-6075-Suppl.1 NUREG/CR--6200 EVNTRE NUREG/CP--0133-VoL2 experimental data NUREG/CR--6044 NUREG/CR--6063 NUREG/CR--6114-Vol.3 NUREG/CR--6267 explosions NUREG/CP-0127 NUREG/CR--6211 D-7 NUREG/BR--0083, Vol.10

Keyword NUREG Report Number failure' mode analysis NUREG/CP--0145 NUREG/CR--6116-VoL2 NUREG/CR--6303 failures NUREG/CR.-4639-VoL5-Rev.4-Pt.3 NUREG/CR--6075 NUREG/CR--6075-Suppl.1

!= ult tree analysis NUREG/CR-4838 NUREG/CR--6116-VoL1 NUREG/CR--6116-VoL2 NUREG/CR--6116-VoL7 NUREG/CR--6145 FAVOR NUREG/CP-0139 FEAT NUREG/CP-0127 FEP NUREG/CR--6116-VoL7 i

NUREG/CR--6145 ferritic steels NUREG/CR--6162 field tests NUREG/CR -6063 Finite Element method NUREG/CR--6162 fires NUREG/CP-0138 i

fission product release NUREG-1502 NUREG/CR-5850 NUREG/CR--6053 NUREG/CR--6107 NUREG/CR--6218 fission products NUREG/CR--6180 flow models NUREG/CR-5403 NUREG/CR-5535-VoL6 NUREG/CR-5965 NUREG/CR--6114-Vol.3 NUREG/CR--6133 flow rate NUREG/CR-5403

[

FLOW 3D NUREG/CP--0127 fluid flow NUREG/CP--0127 NUREG/CR-5535-VoL7 NUREG/CR--6063 NUREG/CR-6133 i

NUREG/CR--6203 NUREG/BR--0083, Vol.10 D-8 i

Keyword NUREG Report Number fluid wha *s NUREG/CR--6203 FLUTAN NUREG/CP--0133-VoL2 FM-DOSE NUREG/CR-5247 Vol.1-Rev.2 NUREG/CR-5247-Vol.2-Rev.2 fracture mechanics NUREG/CR-5128-Rev.1 NUREG/CR--6216 fragmentation NUREG/CP-0127 NUREG/CR--6133 FRANTIC NUREG/CR-6116-Vol.8 FRAPCON NUREG-1502 FRAP-T NUREG-1502 fiqueirf analysis NUREG/CR-5990 l

fuel assemblies NUREG/CR--6102 i

fuelcans NUREG/CR--6218 fuel-coolant interactions NUREG/CP-0127 NUREG/CR--6211 fuel elements NUREG/CP-0139 fuel morage pools NUREG/CR--6102 gas flow NUREG/CR-5403 NUREG/CR-5%5 GENII NUREG/CR--6289 geologic fissures NUREG/CR--6216 geologic fractures NUREG/CR--6203 NUREG/CR--6216 geology NUREG/CR--6114-Vol.3 NUREG/CR--6120 geomagnetic field NUREG/CR-5990 GOTHIC NUREG/CP-0133-VoL2 ground disposal NUREG/CR--6289 ground water NUREG/CR--5%5 D-9 NUREG/BR--0083, Vol.10

Keyword NUREG Report Number ground water (continued)

NUREG/CR-6114-Vol.3 NUREG/CR--6120 half-life NUREG/CR-5247.Vol.2-Rev.2 HARDCORE NUREG/CR--6075 hannonics NUREG/CR-5990 health hazards NUREG/CP-0138 heating NUREG/CR-5990 NUREG/CR--6267 heat transfer NUREG/CP-0127 NUREG/CP-0133-Vol.2 NUREG/CP--0139 NUREG/CR-5850 NUREG/CR--6133 NUREG/CR-6211 NUREG/CR--6269 HELIOS NUREG-1502 high-level radioactive wastes NLPS3/CR--6203 NUREG/CR--6216 HMS NUREG/CR--6180 Hope Creek-1 reactor NUREG/CR-5990 Hope Creek-2 reactor NUREG/CR-5990 HPPOS NUREG/CR-5569-Rev.1 human factors NUREG/CP-0139 human populations NUREG/CR--6053 hydraulics NUREG-1502 NUREG/CP-0127 NUREG/CP-0133-Vol.2 NUREG/CP-0139 NUREG/CR-5850 NUREG/CR--6133 NUREG/CR--6203 NUREG/CR--6211 hydrodynamics NUREG/IA-0093 hydrogen NUREG/CP-0133-VoL2 NUREG/CR--6133 NUREG/CR--6180 NUREG/BR--0083. Vol.10 D-10

Keyword NUREG Report Nunnber bydrology NUREG/CR--6114-Vol.3 IDEMO NUREG/CP-0127 IFCI NUREG/CP-0127 NUREG/CP-0133-VoL2 NUAEG/CP-0139 NUREG/CR-6211 i

implementation NUREG/CR--6303 information needs NUREG-1502 information systems NUREG-1502 NUREG/CR--4639-VoL5-Rev.4-Pt.2 NUREG/CR--4639-VoL5-Rev.4-Pt.3 interactions NUREG/CR--6133 ionic composition NUREG/CR-5229-VoL6 IRRAS NUREG/CR--6116-VoL1 l

NUREG/CR--6116-VoL2 NUREG/CR--6116-VoL3 NUREG/CR--6116-VoL7 NUREG/CR--6116-VoL8 NUREG/CR--6145 ISOSHLD NUREG/CR-6289 IVA-3 NUREG/CP-0127 jets NUREG/CR--6211 KEY NUREG/CR--6290 KIVA NUREG/CR--6075 La Salle County-1 reactor NUREG/CR--4838 La Salle County-2 seactor NUREG/CR--4838 Latin Hypercube sampling NUREG/CR--6116-VoL1 leachates NUREG/CR--6063 leaching NUREG/CR--6063 leaks NUREG/CR-5128-Rev.1 NUREG/CR-5403 leak testieg NUREG/CR-5128-Rev.1 D-11 NUREG/BR--0083, VoL10

Keyword NUREG Report Number LENA NUREG/CR--6053 liquid flow NUREG/CR-5%5 loss of coolant NUREG/CR--6107 NUREG/CR--6160 NUREG/CR--6269 NUREG/1A-0114 LOWCORV NUREG/CP-0133-Vol.2 low-level radioactive wt.stes NUREG/CR-5965 NUREG/CR.-6114-VoL3 NUREG/CR-6120 NUREG/CR -6138 NUREG/CR--6147-VoL1 low-level waste data base NUREG/CR-5229-VoL6 low-Reynolds Number k-e models NUREG/CP-0133-VoL2 LSARG NUREG/CR-5%7 MAAP NUREG/CP--0133-Vol.2 NUREG/CP-0139 NUREG/CR--6075 MACCS NUREG/CR--6053 magnesium NUREG/CR-5229-VoL6 magnetic storms NUREG/CR-5990 management NUREG/CP-0127 man-machine systems NUREG/CR--6116-VoL3 manuals NUREG/CR-5247-Vol.2-Rev.2 NUREG/CR-5535-VoL6 NUREG/CR--6116-Vot2 NUREG/CR--6116-VoL3 NUREG/CR--6116-VoLS NUREG/CR--6116-VoL7 NUREG/CR--6116-VoL8 NUREG/CR--6180 MAPHY-BURN NUREG/CP-0133-Vot2 MAR-D NUREG/CR--6116-VoL8 NUREG/CR--6145 Markov models NUREG/CR-5967 NUREG/BR--0083, Vol.10 D-12

Keyword NUREG Report Nunsber Madtov process NUREG/CR-5967 materials tesung NUREG/CR--6076 MCNP NUREG-1502 measwing instruments NUREG/CR--6044 measuring methods NUREG/CR-5229-VoL6 meetmgs NUREG/CP-0127 NUREG/CP- 0133-VoL2 NUREG/CP-0138 NUREG/CP-0139 NUREG/CP-0145 MELCOR NUREG-1502 NUREG/CP-0133-Vol.2 NUREG/CP--0139 NUREG/CR-5850 NUREG/CR--6075 NUREG/CR-6107 NUREG/CR--6218 MELPROG NUREG/CP-0133-Vol.2 NUREG/CR-6075 meltdown NUREG/CP-0127 NUREG/CR--6044 NUREG/CR-6075-Suppl.1 NUREG/CR--6133 NUREG/CR--6218 NUREG/CR--6267 MELTSPREAD NUREG/CR--6075 mixing NUREG/CP-0127 NUREG/CR--6133 NUREG/CR--6180 NUREG/CR--6211 modelvahdation NUREG/CR-6120 modifications NUREG/CR--4838 NUREG/IA-0093 moisture NUREG/CR-5229-Vol.6 monitoring.

NUREG/CR-5229-Vol.6 Monte Carlo method NUREG/CR--6116-Vol.1 NUREG/CR-6120 D-13 NUREG/BR.-0083, Vol.10

Keyword NUREG Report Number NARAL-SM NUREG/CP--0133-VoL2 Navier-Stokes equations NUREG/CR--6180 NEFTRAN NUREG/CR-6203 NESTLE NUREG-1502 neutron flux NUREG/CR-4206 NEWBALANCE NUREG/CR--6203 nitrates NUREG/CR-5229-VoL6 NUCLARR NUREG/CR--4639-VoL5-Rev.4-Pt.3 nuclear data collections NUREG/CR--6102 l

NUREG/CR--6206 nuclear decay NUREG/CR-5247-Vol.2-Rev.2 nuclear facilities NUREG/CP-0138 nuclear fuels NUREG-1502 NUREG/CR--6102 nuclear power plants NUREG/CP-0133-VoL2 NUREG/CP-0145 NUREG/CR--4409-VoL5 NUREG/CR-5128-Rev.1 NUREG/CR-5344-Rev.1 NUREG/CR-5990 NUREG/CR--6053 NUREG/CR--6116-Vol1 NUREG/CR--6116-VoL2 NUREG/CR--6116-VoL3 NUREG/CR--6116-VoL7 NUREG/CR--6116-VoL8 NUREG/CR--6133 NUREG/CR--6145 NUREG/CR--6211 NUREG/CR--6294 numerical analysis NUREG/CR--6063 numerical solution NUREG/CR-5535-Vol.6 NUPIPE NUREG/CR--6128 OFFSCALE NUREG/CR--6182-VoL1 organicion exchangers NUREG/CR-5229-VoL6 NUREG/BR--0083. Vol.10 D-14

Keyword NUREG Report Number ORIGEN-S NUREG/CR--6182-VoL2 ORIGNATE NUREG/CR--6182-VoL2 OSCAAR NUREG/CR--6053 oxidation NUREG/CR--6218 parametric analysis NUREG/CR--6267 PARSEC NUREG/CR--6075 Peach Bottorn-1 reactor NUREG/CR-5850 Peach Bottom-2 reactor NUREG/CR-5850 performance NUREG/CR.-6269 performana testing NUREG/CR-5229-VoL6 NUREG/CR--6145 penneability NUREG/CR -6203 personnel NUREG/CR--4639-VoL5-Rev.4-Pt.2 phosphates NUREG/CR-5229-VoL6 PHREEQE NUREG/CR--6203 physicalproperties NUREG/CR--6147.VoL1 pipes NUREG/CR-5128-Rev.1 NUREG/CR--6128 NUREG/IA--0093 PIPSYS NUREG/CR -6128 PISTAR NUREG/CR--6128 PISYS NUREG/CR-6128 plants NUREG/CR--6053 plasticity NUREG/CR-5128-Rev.1 PM-ALPHA NUREG/CP-0127 PORFLO-3 NUREG/CR--6063 porosity NUREG/CR--6203 potassium NUREG/CR-5229-VoL6 D-15 NUREG/BR--0083. Vol.10

Keyword NUREG Report Number power distribution NUREG/CR--6206 power reactors NUREG/CR--4816-Rev.2 power systems NUREG/CR-5990 l

PRECURSOR NUREG/CP-0127 PR-EDB NUREG/CR--4816-Rev.2 prediction equations NUREG/CR-5967 pressum dependence NUREG/CR--6075 NUREG/CR-6075-Suppl.1 NUREG/CR--6161

)

pressure measumment NUREG/CR--6044 pressure suppression NUREG/CR--6200 pressure vessels NUREG/CR -6076 pressurization NUREG/CR--6075 NUREG/CR-6075-Suppl.1 NUREG/CR--6267 primary coolarn circuits NUREG/CR--6128 NUREG/CR--6267 NUREG/1A-0093 probabilistic estimation NUREG/CP-0138 NUREG/CR -4639-Vol.5-Rev.4-Pt.3 NUREG/CR--6116-Vol.5 probability NUREG/CP--0138 NUREG/CR--6075 NUREG/CR-6075-Suppl.1 professionalpersonnel NUREG/CR-5247 Vol.2-Rev.2 programndng NUREG/CR--6278 progress report NUREG/CR-5229 Vol.6 PSAFE2 NUREG/CR -6128 public health NUREG/CR--6053 PWR type reactors NUREG/CP-0127 1.UREG/CP-0133-Vol.2 NUREG/CP--0139 NUREG/CR--4409-VoL5 NUREG/BR--0083. Vol.10 D-16

Keyword NUREG Report Number PWR type reactors (continued)

NUREG/CR-5%7 NUREG/CR--6128 NUREG/CR--6160 NUREG/CR--6267 NUREG/CR--6303 NUREG/IA-0093 NUREG/IA-0114 quality assurance NUREG/CR--6278 NUREG/CR--6294 quendring NUREG/CP-0127 NUREG/CR-4133 ndiation accidents NUREG/CR-5247-Vol.2-Rev.2 radiation doses NUREG/CR-4409-Vol.5 NUREG/CR-5247-Vol.2-Rev.2 NUREG/CR-6053 l

radiation effects NUREG/CR-6076 radiation hazards NUREG/CR-4409-VoL5 NUREG/CR -6053 radiation beating NUREG/CR.-6075 NUREG/CR-6075-Suppl.1 radiation monitoring NUREG/CR--6289 radiation protection NUREG/CR--4409-Vol.5 NUREG/CR-3569-Rev.1 radiation transport NUREG/CR-5247-Vol.1-Rev.2 NUREG/CR--6053 NUREG/CR--6203 NUREG/CR--6206 radioactive effluents NUREG/CR -6289 radioactive materials NUREG/CR-5403 radioactive waste disposal NUREG/CR-5229-Vol.6 NUREG/CR--6114-Vol.3 NUREG/CR -6120 NUREG/CR -6203 NUREG/CR--6216 radioactive waste facilities NUREG/CR-5965 NUREG/CR--6114-Vol.3 radioactive waste management NUREG/CR -6147-Vol.1 D-17 NUREG/BR--0083, Vol.10

Keyword NUREG Report Namnber radioactive wastes NUREG/CR--6102 radioactive waste storage NUREG/CR--6138 NUREG/CR--6216 radiochemical analysis NUREG/CR--6147-Vol.1 radioecological concentration NUREG/CR-6289 radioisotopes NUREG/CR-5247-Vol.2 Rev.2 radiolysis NUREG/CR-5229-Vol.6 radionuclide migration NUREG/CR-5229-VoL6 NUREG/CR--6063 NUREG/CR--6120 NUREG/CF--6203 RALOC NUREG/CP-0133-Vol.2 RASCAL NUREG/CR-5247-Vol.1-Rev.2 NUREG/CR-5247-Vol.2-Rev.2 RASPLAV NUREG/CP-0133-VoL2 reactor accidents NUREG-1502 NUREG/CP-0133-Vol.2 NUREG/CP-0139 NUREG/CR-5247-Vol.1-Rev.2 NUREG/CR-5247.Vol.2-Rev.2 NUREG/CR-5535-Vol.6 NUREG/CR-5850 NUREG/CR--6053 NUREG/CR--6075 NUREG/CR-6075-Suppl.1 NUREG/CR--6160 NUREG/CR--6180 NUREG/CR--6211 NUREG/CR--6218 NUREG/IA-0114 reactor components NUREG/CP-0133-VoL2 NUREG/CP-0139 NUREG/CR--4639-VoL5-Rev.4-Pt.3 NUREG/CR -4816-Rev.2 NUREG/CR-5%7 reactor control sysems NUREG/CR-5535-VoL6 NUREG/CR--6278 reactor cooling systems NUREG/CP-0139 NUREG/CR-5535-VoL6 NUREG/BR--0083, Vol.10 D-18

)

Keyword NUREG Report Number reactor core disruption NUREG/CR-5850 reactor cores NUREGUt-5535-Vol.6 NUREG/CR--6218 reactorinstrumentation NUREG/CR--6278 reactor kinetics NUREG-1502 reactor maintenaam NUREG/CR-5%7 reactor materials NUREG/CR--6076 reactor operators NUREG/CR--4639-VoL5-Rev.4-Pt.2 reactor protection systems NUREG/CR--6107 NUREG/CR-6278 NUREG/CR--6303 reactors NUREG/CR--4639-Vol.5-Rev.4-PL2 NUREG/CR--4639-Vol.5-Rev.4-PL3 NUREG/CR-5535 VoL7 reactor safety NUREG/CP-0127 NUREG/CP-0133-Vol.2 NUREG/CP-0139 NUREG/CP-0145 NUREG/CR--4838 NUREG/CR-5850 NUREG/CR--6075 NUREG/CR-6075-Suppl.1 NUREG/CR-6116-VoL2 NUREG/CR-6116-VoL8 NUREG/CR--6133 NUREG/CR--6145 NUREG/CR--6211 NUREG/CR--6294 NUREG/IA-0093 reactor shutdown NUREG/CR-5344-Rev.1 reactor vessels NUREG/CR-5850 RECAP NUREG/CR-5344-Rev.1 recommendations NUREG/CR-5569-Rev.1 NUREG/CR--6303 regulations NUREG/CP-0139 RI1AP/ MOD 3 NUREG/CP-0139 D-19 NUREG/BR -0083, Vol.10

Keyword NUREG Report Number RELAPS NUREG-1502 RELAPS/ MOD 2 NUREG/CP-0127 RELAPS/ MOD 3 NUREG/CR--5535-VoL6 NUREG/CR-5535-VoL7 NUREG/IA-0093 NUREG/IA-0114 relays NUREG/CR--5990 reliability NUREG/CP-0138 NUREG/CP-0145 NUREG/CR--4639-Vol.5-Rev.4-PL2 i

NUREG/CR--4639-Vol.5-Rev.4-Pt.3 NUREG/CR--5%7 l

NUREG/CR--6116-Vol.2 N'JREG/CR--6116-VoL8 NUREG/CR--6278 NUREG/CR--6294 relief valves NUREG/IA-0093 research programs NUREG/CP-0133-VoL2 NUREG/CP-0139 response functions NUREG/CR--6161 RETRAN NUREG/CP-0133-Vol.2 Richards equation NUREG/CR-5965 rist assessment NUREG/CP-0138 NUREG/CP-0139 NUREG/CR--4639-VoL5-Rev.4-Pt.2 NUREG/CR--4639-Vol.5-Rev.4-Pt.3 NUREG/CR--4838 NUREG/CR--6053 NUREG/CR--6075 NUREG/CR--6116-VoL1 NUREG/CR--6116-Vol.2 NUREG/CR--6116-VoL3 NUREG/CR--6116-VoL5 NUREG/CR--6116-Vol.7 NUREG/CR--6116-VoL8 NUREG/CR--6145 NUREG/CR-6278 ROAAM NUREG/CP-0127 NUREG/CP-0133 Vol.2 NUREG/CR-6075 rocks NUREG/CR--6216 NUREG/BR--0083 Vol.10 D-20

1 Keyword NUREG Report Number SABRE NUREG/CP-0133-VoL2 safety analysis NUREG/CR--6116-Vol.1 safety engmeenng NUREG/CR--6116-VoL5 SAILOR NUREG/CR--6206 Salem 1 reactor NUREG/CR-5990 Salem-2 reactor NUREG/CR-5990 SAPHIRE NURFJ3/CP-0139 NUREG/CR-6116-VoL1 NUREG/CR--6116-VoL2 NUREG/CR-6116-VoL3 NUREG/CR--6116-VoL5 NUREG/CR--6116-VoL7 NUREG/CR--6116-Vol.8 NUREG/CR-6145 SAPIV NUREG/CR-6128 SARA NUREG/CR--6116-Vol.1 NUREG/CR--6116-Vol.5 NUREG/CR--6116-Vol.8 NUREG/CR--6145 SASM NUREG/CR--6075 saturation NUREG/CR--6120 NUREG/CR -6203 SCALE NUREG/CR-6102 NUREG/CR -6182-Vol.1 1

NUREG/CR--6182 Vol.2 scale models NUREG/CR-6044 I

NUREG/CR--6267 SCANAIR NUREG/CP-0133-VoL2 SCDAP NUREG-1502 i

SCrud/RELAPS NUREG/CP--0133-VoL2 NUREG/CP--0139 NUREG/CR.-6075 NUREG/CR-6075-Suppl.1 NUREG/CR--6160 security NUREG/CP-0145 D-21 NUREG/BR--0083, Vol.10

Keyword NUREG Report Number seismic effects NUREG/CP-0139 NUREG/CR--6161 seismic waves NUREG/CR--6290 seismographs NUREG/CR-6290 SENKIN NUREG/CR--6075 sensitivity analysis NUREG/CR--6075 NUREG/CR--6114-VoL3 SETS NUREG/CR--6116-VoL8 simulation NUREG/CR--6044 site characterization NUREG/CR-5965 NUREG/CR--6114-Vol.3 NUREG/CR--6216 l

sodium NUREG/CR-5229-Vol.6 soil chemistry NUREG/CR-5229-Vol.6 soils NUREG/CR-5229-VoL6 NUREG/CR-5965 NUREG/CR--6063 NUREG/CR-6120 solutes NUREG/CR--6063 source terms NUREG/CR--5247-Vol.1-Rev.2 NUREG/CR-5850 NUREG/CR--6053 NUREG/CR--6107 SPARC NUREG/CP-0133-Vol.2 spent fucis NUREG/CR--6182 VoL1 NUREG/CR--6182 Vol.2 spent fuel storage NUREG/CR--6102 SQUIRT NUREG/CR-5128-Rev.1 stability NUREG/CR--6161 standardization NUREG/CR--612.8 STANJAN NUREG/CP-0133-VoL2 statistical models NUREG/CR--6120 NUREG/CR--6290 NUREG/BR--0083, Vol.10 D-22

Keyword NUREG Report Number STCP NUREG/CR-5850 NUREG/CR--6107 ST-DOSE NUREG/CR-5247-Vol.1 Rev.2 NUREG/CR-5247-Vol.2-Rev.2 steam NUREG/CP-0127 NUREG/CR--6133 steels NUREG/CR--6161 storage NUREG/CR--6102 stress analysis NUREG/CR.-6161 strontium 90 NUREG/CR-5229-VoL6 l

NUREG/CR--6114-Vol.3 sulfates NUREG/CR-5229-Vol.6 SUPERPIPE NUREG/CR--6128 supports NUREG/CR -6128 Surry-1 reactor NUREG/CR--6107 SUTRA MAC NUREG/CR--6114-VoL3 systems analysis NUREG/CR--6116-Vol.2 NUREG/CR--6145 TAC-2D NUREG/CP--0133-VoL2 TCE model NUREG/CP--0133-VoL2 NUREG/CR--6075 technetium 99 NUREG/CR--6114-VoL3 temperature dependence NUREG/CR--6161 temperature measurement NUREG/CR-5229-VoL6 NUREG/CR--6044 tensile properties NUREG/CR--6076 NUREG/CR--6162 test facilities NUREG/CR--6267 NUREG/lA--0093 testing NUREG/CP-0145 NUREG/CR--6133 Lest reactors NUREG/CR--6076 D-23 NUREG/BR--0083, Vol.10 g,

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Keyword NUREG Report Number TEXAS-III NUREG/CP-0127 text editors NUREG/CR--6116 Vol.7 theoretical data NUREG/CR--6107 NUREG/IA-0093 thermal analysis NUREG/CR-5535-Vol.7 NUREG/CR--6044 NUREG/CR--6160 NUREG/CR-6269 thennodynamics NUREG-1502 THIRMAL NUREG/CR--6133 THIRMAL 1 NUREG/CP-0127 three-dimensional calculations NUREG/CR--6161 Three Mile Island-1 reactor NUREG/CR-5990 l

Three Mile Island-2 reactor NUREG/CR-5990

.p TIGER NUREG/CP--0127 TOXCHEM NUREG/CR--6289 TRACER 3D NUREG/CR--6063 l

TRACG NUREG/CP-0133-VoL2 TRAC-P NUREG-1502 TRAC-PF1/ MOD 2 NUREG/CR--6269 trannng NUREG/CR-5247-Vol.2-Rev.2 transformers NUREG/CR-5990 transients NUREG/CR-5850 NUREG/CR--6200 NUREG/IA--0093 transmutation NUREG/CR.-6182-Vol.2 transport NUREG/CR-5403 NUREG/CR--6102 TR-EDB NUREG/CR-6076 TRIO-MC NUREG/CP-0127 NUREG/BR.0083,Vol.10 D-24

Keyword NUREG Report Nussber tritium NUREG/CR--6063 Trojan reactor NUREG/CR--6206 TRUST + TRUMP NUREG/CR--6063 tuff NUREG/CR--6203 two-cell adiabatic equilibrium model NUREG/CR--6044

- two-dimensional calculations NUREG/CR-5965 NUREG/CR--6114 Vol.3 two-phase flow NUREG/CR-5535-Vol.6 UDEC NUREG/CR -6216 underground disposal NUREG/CR--6114-VoL3 underground storage NUREG/CR--6216 UNSAT2 NUREG/CR--6063 uranium 238 NUREG/CR 5114 Vol.3 l

US NRC NUREG/CR--4639-Vol.5-Rev.4-Pt.2 i

NUREG/CR--4639-Vol.5-Rev.4-Pt.3 NUkEG/CR-5247-Vol.2-Rev.2 NUREG/CR-5569-Rev.1 validation NUREG/CR--6063 NUREG/CR--6102 NUREG/CR--6128 NUREG/CR--6145 VAM2D NUREG/CR--6063 vapors NUREG/CP-0127 NUREG/CR--6203 VICTORIA NUREG-1502 NUREG/CP--0133-VoL2 VTOUGH NUREG/CR--6203 waste forms NUREG/CR-5229-Vol.6 waste water NUREG/CR--6289 water NUREG/CR--6133 water cooled reactors NUREG/CR-5535-Vol.6 NUREG/CR-6102 D-25 NUREG/BR--0083, Vol.10

Keyword NUREG Report Namnber water hammer NUREG/CR--6128 water treatment plants NUREG/CR--6289 WAVCO NUREG/CP-0133-Vol.2 weather NUREG/CR--5229-VoL6 WECAN NUREG/CR--6128 welded joints NUREG/CR-4816-Rev.2 WIMS NUREG-1502 Zion-1 reactor NUREG/CR--6044 NUREG/CR--6075 NUREG/CR-6075-Suppl.1 21on-2 reactor NUREG/CR--6044 NUREG/CR--6075 NUREG/CR-6075-Su;pl.1 ZND model NUREG/CP--0133-Vol.2 t

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