availability of reference materials · availability of reference materials in nrc publications nrc...

I

I

6

AVAILABILITY OF REFERENCE MATERIALSIN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically accessNUREG-serles publications and other NRC records atNRC's Public Electronic Reading Room athttp://www.nrc.aov/readinp-rm.html.Publicly released records Include, to name a few,NUREG-series publications; Federal Register notices;applicant, licensee, and vendor documents andcorrespondence; NRC correspondence and internalmemoranda; bulletins and information notices;inspection and Investigative reports; licensee eventreports; and Commission papers and their attachments.

NRC publications in the NUREG series, NRCregulations, and Title 10, Energy, in the Code of'Federal Regulations may also be purchased from oneof these two sources.1. The Superintendent of Documents

U.S. Government Printing OfficeMail Stop SSOPWashington, DC 20402-0001Internet bookstore.gpo.govTelephone: 202-512-1800Fax: 202-512-2250

2. The National Technical Information ServiceSpringfield, VA 22161-0002www.ntis.gov1-800-553-6847 or, locally, 703-605-6000

A single copy of each NRC draft report for comment Isavailable free, to the extent of supply, upon writtenrequest as follows:Address: Office of the Chief Information Officer,

Reproduction and DistributionServices Section

U.S. Nuclear Regulatory CommissionWashington, DC 20555-0001

E-mail: [email protected]: 301-415-2289

Some publications In the NUREG series that areposted at NRC's Web site addresshttD://www.nrc.pov/readina-rm/doc-collections/nuregsare updated periodically and may differ from the lastprinted version. Although references to material foundon a Web site bear the date the material was accessed,the material available on the date cited maysubsequently be removed from the site.

Non-NRC Reference Material

Documents available from public and special technicallibraries include all open literature items, such asbooks, Journal articles, and transactions, FederalRegister notices, Federal and State legislation, andcongressional reports. Such documents as theses,dissertations, foreign reports and translations, andnon-NRC conference proceedings may be purchasedfrom their sponsoring organization.

Copies of industry codes and standards used in asubstantive manner in the NRC regulatory process aremaintained at-

The NRC Technical LibraryTwo White Flint North11545 Rockville PikeRockville, MD 20852-2738

These standards are available in the library forreference use by the public. Codes and standards areusually copyrighted and may be purchased from theoriginating organization or, if they are AmericanNational Standards, from-

American National Standards Institute11 West 42rd StreetNew York, NY 10036-8002www.ansi.org212-642-4900

Legally binding regulatory requirements are statedonly in laws; NRC regulations; licenses, Includingtechnical specifications; or orders, not InNUREG-series publications. The views expressedin contractor-prepared publications in this series arenot necessarily those of the NRC.

The NUREG. series comprises (1) technical andadministrative reports and books prepared by thestaff (NUREG-XXXX) or agency contractors(NUREGICR-XXXX), (2) proceedings ofconferences (NUREG/CP-XXXX), (3) reportsresulting from International agreements(NUREGIIA-XXXX), (4) brochures(NUREGIBR-XXXX), and (5) compilations of legaldecisions and orders of the Commission and Atomicand Safety Licensing Boards and of Directors'decisions under Section 2.206 of NRC's regulations(NUREG-0750).

NUREG-1635, Vol. 6

Review and Evaluation of theNuclear Regulatory CommissionSafety Research Program

A Report to theU. S. Nuclear Regulatory Commission

Manuscript Completed: March 2004Date Published: March 2004

ADVISORY COMMITTEE ON REACTOR SAFEGUARDSU. S. Nuclear Regulatory CommissionWashington, DC 20555-0001

NUREG-1635, Volume 6, has beenreproduced from the best available copy.

ABSTRACT

This report to the U.S. Nuclear Regulatory Commission (NRC) presents the observations andrecommendations of the Advisory Committee on Reactor Safeguards (ACRS) concerning the NRCSafety Research Program being carried out by the Office of Nuclear Regulatory Research (RES).This report focuses on that portion of the NRC research program dealing with the safety of existingnuclear reactors and advanced light water reactor designs, AP1000 and ESBWR, submitted forcertification. In its review of the NRC research activities, the ACRS considered the programmaticjustification for the research as well as the technical approach and progress of the work. Thisreview attempts to identify research crucial to the NRC mission. It also attempts to identifyresearch activities that have made valuable contributions to the agency mission in the past, but nowhave reached the point where additional research is not needed for efficient and effective safetyregulation. The review also attempts to identify areas where greater international cooperation inresearch useful to the NRC could leverage resources of partners in the research and yield superiortechnical products. The report does not address research on the vulnerability of existing nuclearpower plants to acts of sabotage and terrorism.

.M

TABLE OF CONTENTS

Page

Abstract ..... iii

Abbreviations ......... vii

1 Introduction ........... '.' 1

2 General Observations and Recommendations .......... ................ 3

3 Analysis and Evaluation of Operational Data ........... ................ 9

4 Containment Systems ........................................ ;.13

5 Digital Instrumentation and Control Systems .......... ................ 17

6 Fire Safety Research ........................................ 21

7 Reactor Fuel Research ....................................... 25

8 Neutronics and Criticality Safety .................................... 29

9 Human Factors and Human Reliability Research ........ ............... 31

10 Materials and Metallurgy ....................................... 35

11 Probabilistic Risk Assessment ..................................... 43

12 Radiation Protection . ....................................... 47

13 Seismic Research . ....................................... 49

14 Severe Accident Research ....................................... 51

15 Thermal-Hydraulics Research ..................................... 57

16 References ....................................... 65

V

TABLES

Page

1. Research Activities in Analysis and Evaluation of Operational Data ............. ................. 112. Research Activities in Containment Systems ..................................................... ..... 153. Research Activities in Digital Instrumentation and Control Systems .............. .................. 194. Fire Safety Research Activities .......................................................... 235. Research Activities in Reactor Fuel .......................................................... 286. Research Activities in Neutronics Analysis, Core Physics, and Criticality Safety ......... ...... 307. Research Activities in Human Factors and Human Reliability ..................... .................... 338. Research Activities in Materials and Metallurgy .......................................................... 389. Research Activities in Probabilistic Risk Assessment ..................................................... 4410. Research Activities in Radiation Protection .4811. Research Activities in Seismic Phenomena .5012. Severe Accident Research Activities .5413. Research Activities for Accident Consequence Models .5514. Thermal-Hydraulics Research Activities .61

vi

ABBREVIATIONS

ACRSACNWACR-700AEODANSATHEANAATWSBEIRBMIBWRCAMPCFDCFRCSARPCSNIDOEEPAEPIXEPRIESBWRFYGALLGSIHERAHRAHSSTl&CIAEAIASCCICRPIEEEINPOIPEEEIRISISIISOLERLERFLBLOCALOCALWRMACCSMOVMOXNCRP

Advisory Committee on Reactor SafeguardsAdvisory Committee on Nuclear WasteAdvanced CANDU Reactor-700Office for Analysis and Evaluation of Operational DataAmerican Nuclear SocietyA Technique for Human Event AnalysisAnticipated Transients Without ScramBiological Effects of Ionizing RadiationBare Metal InspectionBoiling Water ReactorCode Applications and Maintenance ProgramComputational Fluid DynamicsCode of Federal RegulationsCooperative Severe Accident Research ProgramCommittee on the Safety of Nuclear InstallationsDepartment of EnergyEnvironmental Protection AgencyEquipment Performance and Information Exchange SystemElectric Power Research InstituteEconomic Simplified Boiling Water ReactorFiscal YearGeneric Aging Lessons Learned (Report)Generic Safety IssueHuman Event Repository and AnalysesHuman Reliability Analysis*Heavy Section Steel TechnologyInstrumentation and ControlInternational Atomic Energy AgencyIrradiation Assisted Stress Corrosion CrackingInternational Commission on Radiological ProtectionInstitute of Electrical and Electronics EngineersInstitute of Nuclear Power OperationsIndividual Plant Examination of External EventsInternational Reactor Innovative and SecureIn-Service InspectionInternational Standard OrganizationsLicensee Event ReportLarge Early Release FrequencyLarge-Break Loss-Of-Coolant AccidentLoss-of-Coolant AccidentLight Water ReactorMELCOR Accident Consequence Code SystemMotor-Operated ValveMixed OxideNational Council on Radiation Protection

Vii

ABBREVIATIONS (Cont'd)

NDE Non-Destructive ExaminationNEA Nuclear Energy AgencyNEI Nuclear Energy InstituteNFPA National Fire Protection AssociationNMSS Office of Nuclear Material Safety and SafeguardsNRC Nuclear Regulatory CommissionNRR Office of Nuclear Reactor RegulationOECD Organization for Economic Cooperation and DevelopmentPARCS Purdue Advanced Reactor Core SimulatorPFM Probabilistic Fracture MechanicsPIRT Phenomena Identification and Ranking TablePRA Probabilistic Risk AssessmentPSHA Probabilistic Seismic Hazard AnalysisPTS Pressurized Thermal ShockPUMA Purdue University Multidimensional Integral Test AssemblyPWR Pressurized Water ReactorRCS Reactor Coolant SystemRES Office of Nuclear Regulatory ResearchRG Regulatory GuideROP Reactor Oversight ProcessRPV Reactor Pressure VesselSC! Secondary Containment IsolationSDP Significance Determination ProcessSNAP Symbolic Nuclear Analysis PackageSPAR Simplified Plant Analysis Risk ModelSSHAC Senior Seismic Hazard Analysis CommitteeTRACE TRAC-RELAP Advanced Computational EngineUCLA University of California, Los AnglesU.S. United StatesUSGS United States Geological SurveyVHP Vessel Head Penetration

viii

1 -INTRODUCTION

The Nuclear Regulatory Commission (NRC)maintains a Safety Research Program to:

* Ensure its regulatory frameworkhas a sound technical basis

* Prepare for the anticipatedchanges in the nuclear industrythat could have safety implications

* Develop improved methods for itsregulatory process

* Maintain an infrastructure ofexpertise, facilities, analyticaltools, and data to supportregulatory decisionmaking

These missions for the research effort weredefined when the NRC was established andthere was limited experience with theoperation of nuclear power plants. The needfor research remains today, despite thegrowth of experience with the plants, as:

* Nuclear power plants age andencounter challenges of materialdegradation not anticipated whenthe plants were designed

* The NRC considers applicationsfor extending licenses andincreasing the operating powerlevels of plants

* Reactor fuels are used to higherlevels of burnup, new claddingsfor the fuels are introduced, andmixed oxide (MOX) fuels areconsidered for the disposal ofexcess weapons-grade plutonium

* The NRC evolves its regulationsfrom a deterministic foundation toa risk-informed basis that makesgreater use of 'best-estimate'analyses

* New technologies such assoftware-based digitalinstrumentation and control (I&C)systems are backfit into theexisting plants

* New light water reactor (LWR)designs making more use ofpassive systems are submitted forcertification

In this report, the Advisory Committee onReactor Safeguards (ACRS) presents itsobservations and recommendationsconcerning that portion of the NRC SafetyResearch Program focused primarily on thesafety of existing nuclear reactors. This reportdoes consider research activities that supportthe certification of the advanced LWRs,AP1000 and ESBWR. It does not addressresearch to prepare the agency to certify theACR-700 design or other advanced reactordesigns such as the GEN IV design beingconsidered by the Department of Energy(DOE) or the IRIS design. The ACRSrecently reported on the agency's plans forresearch on these advanced reactor designs[Ref. 1 ] and the progress in advanced reactorresearch has not been sufficient to warrantreexamination of these research activities atthis time. The present report does notaddress research on nuclear waste and thelicensing of a permanent geologicalrepository for spent reactor fuel. The AdvisoryCommittee on Nuclear Waste (ACNW)monitors this research and will report itsconclusions and recommendations

I NUREG-1 635

lI

separately. This report does not addressresearch on the vulnerability of existingnuclear power plants to acts of sabotage andterrorism. The ACRS will report separatelyon the technical aspects of this research.

In its review of the NRC Safety ResearchProgram, the ACRS considered theprogrammatic justification for the research aswell as the technical approach and progressof the work. The ACRS supports researchthat:

* Provides support to theidentification and resolution ofcurrent safety issues

* Provides the technical basis forresolution of foreseeable safetyissues

* Develops the capability of theagency to independently reviewrisk-significant proposals andsubmittals by the licensees

* Supports initiatives of the agency,including the Reactor OversightProcess (ROP) and the move torisk-informed and performance-based regulation

* Improves the efficiency andeffectiveness of the regulatoryprocess

* Maintains technical expertisewithin the agency and facilities indisciplines crucial to the agencymission and not readily availablefrom other sources

This review of the NRC Safety ResearchProgram, in particular, attempts to identifyresearch activities that have made valuablecontributions to the agency mission in thepast, but now have reached the point whereadditional research is not needed for efficient

and effective safety regulation. This reviewalso attempts to identify areas where greaterinternational cooperation in research useful tothe NRC could leverage resources ofpartners in the research and yield superiortechnical products.

General observations and recommendationsconcerning the NRC research activities arepresented in Chapter 2 of the report.

Other observations and recommendationsconcerning the research activities in thefollowing technical disciplines are discussedin Chapters 3 through 15 of this report:

* Analysis and Evaluation ofOperational Data

* Containment Systems

* Digital InstrumentationControl Systems

and

* Fire Safety Research

* Reactor Fuel Research

* Neutronics and Criticality Safety

* Human Factors and HumanReliability Research

* Materials and Metallurgy

* Probabilistic Risk Assessment

* Radiation Protection

* Seismic Research

* Severe Accident Research

* Thermal-Hydraulics Research

NUREG-1 635 2

2 GENERAL OBSERVATIONS AND RECOMMENDATIONS

Overall, the NRC has a well-focused, well-planned Safety Research Program dealingwith existing reactors and advanced LWRssubmitted for certification. The research effortmay well be near the minimum needed tosupport regulatory activities and agencyinitiatives while maintaining technicalcompetencies crucial to the agency mission.A very large fraction of the research isfocused on immediate user needs.Resources available for exploratory researchto investigate other avenues of improvedsafety and more efficient regulation areminimal and may limit the agency's ability toanticipate future needs.

Some of the research activities are especiallynoteworthy. This includes the research beingdone in probabilistic risk assessment (PRA)to provide computational tools to support theReactor Oversight Process. Confirmatoryresearch resolving safety issues of reactivityinitiated accidents with high-burnup fuel isalso noteworthy. Human factors research atNRC has been re-energized and holds thepromise of significant future contributions tothe regulatory process.

There are research challenges. The importanteffort to provide the agency with aconsolidated, state-of-the-art, best-estimatethermal-hydraulics computer code is such achallenge. Though this effort has progressedwell, the time is approaching when the needfor this code will be critical. It may benecessary to increase the effort and focus'the thermal-hydraulics research on producingthis code and integrating it into the regulatoryprocess. Once this is accomplished, researchresources will be freed to improve thepredictive capabilities in thermal-hydraulics.Further research in -the area of thermal-'hydraulics especially in areas other thanlarge-break loss-of-coolant accidents (LB-LOCAs) might be. facilitated 'by more

extensive international cooperation akin tothat being done in the area of severe accidentresearch.

Some research efforts have producedvaluable technologies, which can beintegrated into the regulatory process.Further research in these areas should bereduced substantially. An example of suchan effort is the research on the realisticstructural capacity of reactor containments.This research has produced, benchmarked,and validated computer codes that can beused to assess containment designs andestimate the' effects of both materialsdegradation and construction errors.

Research on seismic engineering hasreached the point that the most significantagency needs have been met and greaterreliance on engineering consulting firms, asneeded, may be adequate for the regulatoryprocess in the future. There is advantage inmaintaining modest research efforts toaccrue benefits from cooperative research onseismology and seismic engineering.

Other observations and recommendationsconcerning the research efforts 'aresummarized below and also discussed inindividual Chapters.

Analysis and Evaluation of OperationalData

The ACRS supports the research activitiesnow under way in'the general area of analysisand evaluation of operational data. Indeed,these data collection and organizationalactivities, including superior computerizedsearch capabilities, are essential to theagency mission. However, the ACRS isconcerned about the vitality and planning ofcontinued efforts to use the database andespecially the opportunities being made to

3 NUREG-1 635

explore the database independently ofcurrent research needs. Such independentexaminations of the database in the searchfor unexpected interactions amongregulatory activities can always be deferred atlittle cost. Continued deferral will deprive theagency of the opportunity to utilizeoperational data in many ways that may leadto more effective and realistic regulatorypractices.

Containment Systems

Adequate computer models to assess thestructural capabilities of containments ofexisting reactors are now available. Theyhave been benchmarked and validated bycomparison to well-scaled test data. Theeffects of degradation as well as the effectsof construction errors can be evaluatedadequately by analysis without the need forfurther experiments.

Debris accumulation in pressurized waterreactor (PWR) sumps is an important issuethat must be resolved. The planned work tofurther explore the potential for chemicalinteractions that generate suspensions,which can lead to large head losses even withlow fiber loadings, is extremely important.Research in this area needs to be expeditedand promptly reach the point licensees canimplement solutions to this issue.

Digital Instrumentation and ControlSystems

The current reliance on controlling the designand development process of software, ratherthan focusing on the software itself, has thepotential of straining agency resources in theregulatory process. Although a controlleddevelopment process is assumed to lead to ahighly reliable product, this reliability remainsunquantified. Complicating the regulatoryreview of these l&C systems is the pace ofinnovation in the digital electronics industry.Innovations of revolutionary natures are being

made at rates far greater than those theregulatory system can respond to.Furthermore, threats to digital systems aregreatly expanded with the emergence of theso-called "cyber security" threat posed by themalevolent or just the prankster.

In general, the ACRS supports the researchactivities now under way in the Digital l&CSystems. These activities, if successful, willprovide tools to make the review processmore efficient and provide a basis forincluding software reliability into PRAs, thusreducing the need for relying on controllingthe design and development process, andenhancing the ability of the agency to risk-inform its regulations.

Fire Safety Research

The current NRC research efforts in FireSafety seem incongruent with the estimatedrisk significance of fire. The limited fireresearch effort is understandable sinceresources for such research have beendiverted to respond to the events ofSeptember 11, 2001. Now that theseresponses are being completed, the agencyshould revitalize its Fire Safety researchefforts and move the technical capabilities ofthe agency to be more in line with risks beingascribed to fire.

Reactor Fuel Research

The NRC is completing a confirmatoryresearch effort on reactivity insertion withhigh-burnup fuel. A sustained NRC researcheffort in reactor fuels behavior under accidentand off-normal conditions is neededespecially as competitive pressures force fuelvendors to curtail their research efforts.Continuing expertise and even additionalresearch may be needed if, as now expected,licensees make requests to extend fuelburnups up to and beyond 75 GWd/t. It isquite likely that such industry proposals willbe substantiated by minimal experimental

NUREG-1 635 4

research. It is important for the NRC to havea technically justified position on theexperimental research that is necessary tosupport such proposals especially in light ofthe "surprise" that accompanied extendingfuel burnups into the range of 50 to 60GWd/t.

Neutronics and Criticality Safety

The Office of Nuclear Regulatory Research(RES) is doing a commendable, cost-effective job in maintaining the capabilities ofits neutronics and core physics codes and theassociated databases. Upgrades to thesecodes to address both extended fuel burnupand MOX fuels are needed and these needsare addressed in thecurrent researchprojects. The ACRS anticipates that theseextended capabilities will be testedsignificantly in the certification of theadvanced reactor designs.

Human Factors and Human ReliabilityResearch

The attention NRC is paying to HumanFactors research is yielding dividends. TheStandard Review Plan Chapter 18 hasrecently been updated to reflectdevelopments especially in the area of digitall&C systems. A risk-informed method toscreen licensee submittals for human factorsreview has the potential of better focusingagency resources on risk- significant humanfactors issues.

The issues surrounding human performancein teams (such as the team of control roomoperators) and human performance inorganizations, including the question oforganizational reliability, are important andneed to be considered in the ongoing HumanPerformance Research planning.

The NRC is experimenting with an ambitiousmethodology for estimating human reliabilitythat is to assess errors of commission as well

as errors of omission. This methodologycalled ATHEANA focuses on the causes ofhuman performance ("performance shapingfactors"). The ACRS looks forward toreviewing the results of the ongoingATHEANA application efforts and learningwhether the method has become easier toapply, whether it yields insights significantlybetter than those that could be derived fromsimpler methods, and whether it includes aquantification process for human reliability.Of particular interest to the ACRS is the useof expert opinion elicitation in theseapplications to quantify the ATHEANAresults.

Materials and Metallurgy

Research in this area addresses primarilyradiation-induced embrittlement of ferriticpressure vessel steels and austenitic corestructure materials, and the characterizationand monitoring of environmentally-assistedcracking. The quantification of thesephenomena is central to maintaining theintegrity in LWR structures.

Projects dealing with Reactor PressureVessel (RPV) Embrittlement should beconsidered with respect to safety issues ofRPV integrity and with respect to maintainingthe core competency in irradiationembrittlement. The staff needs to evaluatewhat research should be retained in order tomeet these two needs.

The characterization of cracking kineticsshould be continued in the areas ofirradiation-assisted cracking of irradiatedstainless steels used in boiling water reactor(BWR) core structures and of stress-corrosion cracking of nickel base alloys usedin PWR steam generators and in largecomponents in all LWR designs. Thisinformation, in addition to the evaluation ofvarious inspection techniques, is required forthe NRC to maintain an independentcapability to examine integrity issues inherent

5 NUREG-1 635

to licensee plans concerning residual life,inspection, and repair. Many of theseprojects are being conducted withininternational research programs, therebyleveraging NRC resources.

The ACRS supports the "Proactive MaterialsDegradation Assessment" Project recentlyinitiated by the NRC in order to move fromthe economically inefficient reactive mode ofregulating materials degradation issues. Theplan to lead a phenomena identification andranking table (PIRT) effort with the industryover the next two years, with expert elicitationfrom international experts, should identifypotential materials degradations that mayoccur in the future. This knowledge willprovide a better basis for future regulatorydecisions on timely inspection and repaircriteria.

Probabilistic Risk Assessment

The NRC has an impressive PRA researchprogram. The ACRS generally supports theidea that the NRC needs to maintain its PRAcapabilities at or near the state-of-the-artlevel. As the NRC moves toward a risk-informed regulatory system, it will need evergreater amounts of risk information derivedfrom quantitative risk assessments. It willwant to maintain its own independentcapabilities for conducting PRAs at state-of-the-art levels. A continued vital researchprogram in PRA is, then, crucial to theagency mission.

Radiation Protection

The NRC research activities in the area ofRadiation Protection constitute a well-leveraged program that allows the NRC touse quality data and information as thefoundation of its basic radiation protectionregulation, 10 CFR Part 20, "Standards forProtection Against Radiation."

Seismic Research

The NRC has invested heavily in theunderstanding of seismic issues anddevelopment of its rules to ensure thatseismic risks are acceptably low. Licenseesubmittals concerning risks associated withexternal accident initiators such as seismicevents suggest these efforts have beenlargely successful.

Much of the NRC research into seismicissues involves cooperative efforts to addressessential seismic data. The ACRS supportsthese efforts which are highly leveraged bycontributions from other partners.

The ACRS understands that there are needsfor additional work on the engineering ofcomponents and structures to avoid seismicdamage. Unlike many other issues that aredealt with by NRC research, there is in thisCountry a very substantial engineering andscientific community addressing this issue.This community is independent of the NRClicensees. Thus, the NRC-sponsored work isnot needed to maintain a core competency inthis area. If there are open issues associatedwith licensee submittals such as the seismicbehavior of spent fuel casks or buriedstructures, the NRC can rely on engineeringconsulting firms on an as-needed basis todeal with such issues.

Severe Accident Research

The ACRS is very supportive of the NRCstrategy to maintain and update itscapabilities for accident analyses. The ACRSencourages the NRC to give seriousconsideration to joining internationalcooperative research agreements now beingproposed to conduct:

* Prototypic, in-pile tests ofradionuclide releases from, anddegradation of, reactor fuelexposed to air for both reactor

NUREG-1 635 6

accident analyses andspent fuel storage poolaccidents

* Prototypic, in-pile tests ofradionuclide releases from, anddegradation of, high-burnup fuels

* Prototypic, in-pile tests ofradionuclide releases from, anddegradation of, MOX fuel

The ACRS also supports programs tomaintain the MACCS code and looks forwardto reviewing results of the comparison of theGaussian plume model to state-of-the-artdispersion models in the context of riskassessments.

enthusiastically supported by the ACRS. TheACRS understood this to be a challengingundertaking. Indeed, the consolidation hasproved to be every bit as challenging asanticipated. A version of the consolidatedcode, TRACE, is now being tested. Effortsshould be focused on integrating TRACE intothe regulatory process even at the expense ofresearch to further improve the technicalcapability of TRACE.

The ACRS encourages anticipatory researchto couple computational fluid dynamics (CFD)methods with system codes such as TRACE.This has the potential to enhance theaccuracy of thermal-hydraulic analyses.

Thermal-Hydraulics Research

The decision to consolidate NRC's thermal-hydraulic modeling into a single code was

7 NUREG-1 635

3 ANALYSIS AND EVALUATION OF OPERATIONAL DATA

The NRC is moving toward a risk-informedregulatory system that will require morerealistic estimates of vulnerabilities of nuclearpower plants. The ACRS has voiced its belief'that the operational database for nuclearpower plants is a rich source of informationfor improving the quality of quantitative riskassessments and identifying omissions inthese assessments [Refs. 2 and 3]. Forseveral years following the 1979 accident atThree Mile Island, then the Office for Analysisand Evaluation of Operational Data (AEOD)conducted independently motivated, detailedevaluations of the operational database[Refs. 4 and '5]. The products of theseevaluations were of direct use to thequantitative risk analysis community.

In 1999, the function of the AEOD wastransferred to RES. It is of'considerableinterest to the ACRS to see that the intenselyinquisitive, independent evaluation of theoperational database is sustained in thisorganizational structure. Furthermore, theACRS is anxious to see that results of suchindependent investigations are used by otherNRC Offices.

Research Projects in the area of Analysis andEvaluation of Operational Data' are listed inTable 1. These Projects are primarilydevoted to the sustained acquisition andorganization of operational data from licenseeevent reports (LERs), maintenance data andthe like. Currently, the motivation for the datacollection efforts appears to be primarilyanticipation of operational data needs of otherresearch initiatives by the agency to address

m -r .k Electrical Grd StabIlt '7 codjtI '_ 7

she Country :was .-shocked -In '-2003. .by .apopagatilng6ioss"- of .'t'e' elect-rical.' g d 'inventral and Eastern United States that:'leftrI11lon's1 without . electrlcal . power.:;:'.Gridvulneraeiity. be.ame -a topic-of Niumerousiscussions ln' thee. .-news .- media.' -'The

ulnerabity had6een anticipatedbyji'hthe NRCstudy, Vperating Experlenci AssessmentEffeats' 'of -Grid'Events :'onNuclear';Power'Plant 'PerfonnceS, . f ' l., -6: The stud'earned nd events that- affected,`nucleariwer.'plantsf r1994 to 20d1..The eventsere chosen orexaiinmati6n With attentionito

the structural changes that have occurred Inthe electrical energy. market 'as'economcat " . ^f. , : .>!,m.ia, ,,_d1eregiilath'n 'has. progressed. :Thb studyshowed that the number Of r id events nay

d,6S d&e'asing Ahith time, .but theS dura'tons ofevents were increasing -5a finding 'quited bjyAajonetwohrk. Though most events pose no

cxcejptihanIchallengesto nuclearpow`r plantsafetyissues 'are identified in the 'study tha'to have the potential to increase .risk andhallenge -the 'effectiveness ,ofJ the: cuurrentgulations. :.me report docw;enting "the

study pro vides a baseline of grid 'erformance. .I ' . , .- I.:_ ' ' ' .,' ,.; ' ' ' - ! ,- I2 ' .-I.:'! ' ' _ ''! _ '-.. "which can be used to 'gauge the.impact-of+,eregulatidn on, ',and changes in, gridoperations. K -' ¶

rhe report shows the public that the NRC Isemerging'issues sthatdcan affect

owerpiant safety and how bpe'rationalidIacan be< used toanticipat'ea'nd quantifyremerging issues.-.!:,':.-.-'... ;-,-:.''

1 Note that occupational exposuredata and reports are discussed in Chapter12, "Radiation Protection," of this report.

9 NUREG-1 635

regulatory issues. Campaigns for theimmediate future and the longer term toutilize the archives of operational data are nolonger readily apparent from the researchprogram. An undertaking that is apparent andcould be of particular regulatory significanceis the search for safety-significant interactionsamong the ongoing regulatory actions to:

* Extend the license period of nuclearpower plants

* Increase the operating power levels ofthe plants

* Use fuel in the plants to higher levels ofburnup

* Evolve plant operations to do moreonline maintenance

The ACRS is concerned that this search forinteractions among the various operationalchanges has stagnated for lack ofmanagement support.

In summary, the ACRS does not critique theresearch Projects now under way in thegeneral area of Analysis and Evaluation ofOperational Data. Indeed, these datacollection and organizational activitiesincluding superior computerized searchcapabilities are essential to the agencymission. It is evident that useful products arestill being generated in this effort (Seesidebar success story on Electrical GridStability). The ACRS is concerned about thevitality and planning of continued efforts touse the database and especially theopportunities being made to explore thedatabase independently of current researchneeds. Such independent examinations of thedatabase in the search for unexpectedinteractions among regulatory activities canalways be deferred at little cost. Continueddeferral will deprive the agency of theopportunity to utilize operational data in manyways that may lead to more effective andrealistic regulatory practices.

NUREG-1 635 10

Table 1. Research Activities in Analysis and Evaluation of Operational Data2

Project Title Comment

A9134 Sequence Coding Search This is an important Project for analyzing industrySystem trends and establishing a data source for

Database from LERs used for NRC measuring industry performance.studies

G6810 Management of Precursors This is a short-term grant for a two-day workshopStudy for detecting, analyzing, and benefitting from

knowledge of accident precursors.

J8258 International Common-Cause This is an useful collaborative effort to maintainExchange Project cognizance of developments in the realistic

Maintaining awareness of quantification of common-cause failures in PRA.developments in common-causefailures within the internationalreactor safety community

Y6215 Operational Events for This important Project seeks to identify and rankAccident Sequence Precursor those operating events that were most significant in

Program terms of the potential for inadequate core coolingand the potential for core damage.

Y6546 Industry Trends Program These important Projects produce plant-specificand industry-wide estimates, summary tables,graphs, and charts from operational data to support

Y6636 Support for Industry Trends the Industry Trends Program and various dataanalysis activities of the NRC.

Y6626 Access to INPO's EPIX Database on failures of key components; databaseSystem for SPAR models used in the ROP.

Y6214 Integrated Data Coding and Improved database management; only a smallAnalysis Methods effort remaining.

Y6468 Reactor Operating Experience Collects, codes, and maintains operational data forData for Risk Application reactor systems and components, initiating events,

common-cause failures, and fire events for use inregulatory activities. Utilizes methods andprocedures developed under Y6214.

Y6406 Assess and Improve Potential for identification of unanticipated, safety-Regulatory Effectiveness significant interactions among regulatory activities.

(See also Y6522 in Probabilistic RiskAssessment, Table 9) .-

2See also Y6123, Retrospective Risk Analysis of Human Reliability in Operating Events,Table 7.

I1I NUREG-1635

4 CONTAINMENT SYSTEMS

it_-Liz-t-M__It__oCt-_ .- Ace

The defense-in-depth safety philosophy hasled .to encasement of power reactors inrobust containment buildings. Containment isthe final physical barrier to the uncontrolledrelease of radioactive materials to theenvironment in the event of an accident.Comparisons of the environmental releasesof radioactivity during the accidents atWindscale, Three Mile Island, and Chernobyldemonstrate the value of robustcontainments.

The structural capabilities of containments fornuclear power plants in this Country are fargreater than might be inferred from theirdesign bases. To realistically assess thesecapabilities and demonstrate their capacity toprovide protection in the event of severeaccidents, the NRC, along with othercountries and agencies, has established atechnically sophisticated program. Thisprogram involved testing of quarter-scalecontainment models to provide databases for.validating analytical structural models.uRound robin" comparisons of the predictionsof these models to the data obtained in thetests were conducted.

This effort to develop validated technology forthe quantitative characterization ofcontainment capabilities has been bothessential and quite successful. Thetechnology for predicting containmentresponses to loads can now be transferred tothe capable hands of users of theinformation such as line regulatoryorganizations, codes and standardsorganizations, and risk analysts.

Research studies to date have been on Uas-designed" structures. Many containmentshave suffered some degradation to either theconcrete shell or the steel liner and it isimportant to understand how degradation

771

l','' tI Is...

oRf cntainment model dur' in

V5 ~tr 'Ae ; ~

tetdone in cooperation: with Japan ison faseries attie'sts 'that have 'be'en dn

-b becmrk anhdyvalidatestructuralcapacityuter 'codes. 7 yiall, e, 'scle

~ontaineht, mdels~. re foud to Jila

~ha~ twce he design 'pressure.,. Cim- u

coes jvaidted,..S by .w:the' tesig4:: ar used;.'to

o-rak s eictins`o : 'l-Rbbptures of co con ianimoelts undri'cid'ent condtions' ,'

bne'f asend oftest tht hae ben 'don

13 NUREG-1 635

affects the structural margins attributed tocontainments in risk analyses. Because thecomputer codes for structural analyses ofcontainments have been benchmarked andvalidated, the effects of degradation, as wellas the effects of construction errors, can beevaluated adequately by analysis without theneed for further experiments. Projects Y61 64and W6684 (See Table 2) address suchquestions and these efforts should becontinued. However, the value of generalstudy performed under Project W6684 isquestionable.

The aging management of containmentstructures is being addressed in the currentactivities associated with license extension. Iffurther work is needed in this area, it shouldbe done by the licensees in support of theirapplications.

Project W6593, "Effect of Aging andEmerging Issues on MOV performance,"should be concluded. Active componentssuch as motor operated valves (MOVs) areincluded in ongoing maintenance and testingprograms which should be adequate to detectany potential aging issues. Further studiescould improve reliability, but this should bethe responsibility of the licensees or thevendors.

The ACRS has pointed out that thetechnology base is still not complete [Ref. 7].The planned work to further explore thepotential for chemical interactions thatgenerate suspensions, which can lead tolarge head losses even with low fiberloadings, is extremely important. The ProjectY6041 may deserve more managementattention. Research in this area needs to beexpedited and focused to support theregulatory process and provide a basis forlicensees to implement solutions to thisissue.

Assessment of containment structures forentombment, Project Y6331, can besubsumed within decommissioning projects.This does not appear to be a particularlypressing issue.

Terminating or transferring some of theContainment Systems Projects (See Table 2)should free resources available to the agencyto address containment/confinement issuesand security vulnerabilities of reactorcontainments for advanced reactors. TheACRS believes it would be prudent for theNRC to examine the safety issues associatedwith underground siting of future reactors andother new nuclear facilities.

Debris accumulation in PWR sumps is animportant issue that must be resolved.

NUREG-1 635 14

Table 2 Research Activities in Containment Systems


J6043 Inspection of Aged/Degraded Containment This work should be concluded. Ifadditional information is needed, itshould be provided by licenseeswho have the responsibility todemonstrate adequacy.

W6851 Review Guidance for Lightning This effort should be concluded.

Y6757 Containment Capacity Studies Development of technology toestimate capacities of containmentsto sustain pressure loads has beensuccessful and should now bebrought to closure.

W6593 Effects of Aging and Emerging Issues on This work should be concluded.MOV Performance These are active components and

ongoing programs should beadequate to detect any potentialaging issues.

W6684 Assessment of Aged and Degraded Margins associated with real,Structures and Components degraded structures must be

understood for realistic analyses.Such issues are best addressed interms of more specific conditionssuch as in Y6164. The value of ageneral study is questionable.

Y6041 Assess Debris Accumulation on PWR This is an important issue but oneSump Performance that has been slow to resolve;

research to support regulatoryinitiatives should be expedited andfocused to support the regulatoryprocess and provide a basis forlicensees to implement solutions tothis issue.

Y6164 Structural Risk-informed Assessment This work should be continued. ItContainment Degradation provides quantification of loss of

margin as a result of degradation.

Y6331 Long Term Containment Stabilization Studies for entombment of nuclearConcrete Structures facilities as permitted by

regulations. This does not appearto be a particularly pressing issue.

15 NUREG-1 635

5 DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS

The move to software-based, digital l&Csystems is taking place for advanced nuclearpower plants 'and more slowly for existing'nuclear power plants. Analog, electro-mechanical systems in existing plants'aredifficult to maintain as replacement parts ofsuitable quality become less readily available.Furthermore, there is a widespread belief thatsoftware-based digital electronic systemsoffer superior levels of control, informationdensity, and reliability.

Certainly, digital systems have inherentreliability that is potentially much greater thanelectro-mechanical systems. There havebeen, however, notable instances ofcatastrophic failures of digital systems.Usually, these failures have been traced toomissions in the definitions of therequirements and specifications for thesoftware that controls the system. Thechallenge in the regulation of digital l&Csystems is the difficulty in comprehensivetesting of these systems due to their diversecapabilities. At this time, in order to ensureadequate quality and reliability of software-based digital systems, the NRC reliesprimarily on the manpower-intensivemonitoring of the design and developmentprocesses rather than focusing on theproduct. It is assumed that a controlledproduction process will lead to a highlyreliable product. This reliability, however,remains unquantified.

The reliance on process monitoring has thepotential of straining agency resources in theregulatory process if highly capable tools arenot available to the agency. Complicating thedesign and development of such regulatorytools is the pace of innovation in the digitalelectronics industry. Innovations ofrevolutionary natures are taking place at ratesfar greater than those to which the regulatoryframework can respond. Furthermore, threats

to digital systems are greatly expanded withthe emergence of the so-called 'cybersecurity" threat posed by the malevolent orjust the prankster.

There is a vast literature on softwarereliability. This literature surrounds twoschools of thought. One school holds thatsoftware reliability can be quantified andmethods can be developed for doing so. Theother school of thought views softwarereliability as meaningless by itself.Embedded software failures are only onecontribution to the unreliability of the overallsystem. The staff has judged that thesemodels have not reached a level of maturitythat allows them to be used in the regulatoryprocess. The ACRS agrees.

The NRC research activities in Digital l&CSystems are listed in Table 3. Project K6907is intended to provide a much neededassessment of the merits and demerits of themodeling methods of the two schools ofthought. Within this Project, the staff plans todevelop and test methods and models forintegrated assessment of digital systemreliability. Project Y6591 is developing andtesting methods for evaluating softwarereliability. Project Y6332 is an investigation ofdevelopment of PRA models for digitalsystems and Project Y6472 is to investigatethe integration of digital systems into thePRAs for current generation nuclear powerplants. These Projects, if successful, willprovide the basis for including softwarereliability in PRAs, thus reducing the need forrelying on controlling the design anddevelopment process and enhancing theability of the agency to risk-inform itsregulations.

Studies of digital system performance andreliability appear to involve a larger issue thanany one agency can undertake in the face of

17 NUREG-1 635

an innovative digital electronics industry. Thestaff is working in several areas to try to tapinto this wider research, including serving onthe steering committee of the DOE l&C andHuman Machine Interface advisorycommittee. The staff is also in the process ofdeveloping an international collaborativeprogram and a potential research effort inthis area is digital system reliability modeling.The RES staff also chairs the OECD/NEACOMPSYS group, which is charged withdeveloping nuclear power plant specificdigital systems failure data. The ACRSencourages these activities. However, it is notevident that the Project at Halden (Y6349)directly addresses the NRC needs.

Two Projects (Y6371 and Y641 0) addressthe issue of aging of cable systems in existingnuclear power plants. Demonstration thatcables in a power plant have not aged to thepoint that plant safety is threatened is theresponsibility of the licensee. The NRCneeds to understand the issue well enough toensure that adequately realistic conditions foraging rather than unnecessarily conservativeconditions are taken into consideration andthat proposed technical responses to theissue are adequate. Consequently, thesestudies can be brought quickly to a resolutionsufficient for regulatory needs.

Similarly, the study of electromagneticconditions at nuclear power plants (Y6272)has been under way for sometime. Anadequate understanding should be soonavailable to review licensees' arguments thatelectromagnetic conditions do not threatenplant safety. Therefore, this Project shouldbe concluded in FY 2004.

Several Projects address the issue of cybersecurity of digital systems (Y6729, Y6712,Y6733, and Y6734). Again, the issue of cybersecurity seems bigger than what any onegovernment agency can confront, least of allan agency as small as the NRC. RES shouldexplore the possibilities of leveraging theresources it has for cyber security by joiningwith other government agencies in the studyof this issue.

NUREG-1 635 18

Table 3. Research Activities in Digital Instrumentationand Control Systems


K6907 Digital System Performance Program is intended to provide an unbiasedand Reliability. assessment of the merits and demerits of

A survey of the various modeling modeling methods in the two main schools ofmethods for digital systems thought concerning software reliability. The staff

plans to develop within this activity a tool toassess the reliability of an integrated digitalsystem based on fault-tolerant system methods.

Y6475 Wireless Wireless communication may resolve some fireAnticipatory research on the potential issues but is very susceptible to cyber attack.challenges and regulatory issues of There is interest within the nuclear industry. inusing wireless communication using this technology and NRC research is beingtechnologies in nuclear power plants done in anticipation of an industry proposal.

Y6590 Review IEEE Standards for An essential activity since these standards will beEndorsement the bases for regulatory review of the

IEEE Standards are the bases for performance and reliability of digital systems.most Regulatory Guides dealing withissues of digital instrumentation andcontrol

Y6332 Digital Systems Risk Essential to developing the technology to includeInvestigation of digital l&C system digital systems reliability in PRAs.analysis methods

Y6371 Risk Associated With Cable A simple effort that ought to be concluded in FYAging 2004.

Y641 0 Collaborative Research on Progress should be monitored to ensureWire System Aging research does not stray beyond needs of the

Collaborative research on current regulatory process and does not enforceconditions, residual lifetime and fire overly conservative constraints on licensees.risk of cables

Y6272 Characterize Electromagnetic Second year of a two-year effort. Should beConditions at Nuclear Power concluded in FY 2004.

Plants

Y6349 Halden Environmentally Despite the title, this is I&C; it is not evident thatAssisted Cracking the current Halden program in this area directly

addresses NRC needs.

Y6472 Risk-importance of Digital To investigate integration of digital systems inSystems PRAs.

Y6591 Software Reliability Code Development of measures to quantify theMeasurements reliability of safety system software.

. .

19 NUREG-1635

Table 3. Research Activities in Digital Instrumentationand Control Systems3 (Continued)

Project Title CommentY6729 Classification of Digital Vulnerability to cyber attack

System Vulnerabilities

Y6712 Safety System Isolation Study Generically qualified systems vulnerability to cyberattack

Y6733 Protocol Robustness Analyses Anticipatory research on the vulnerability of theindustry to cyber attack

Y6734 Security Tool Vulnerability Adequacy of commercial off-the-shelf tools forCase Study preventing cyber attack

3 See also Y6651, Effects of switchgear aging on energetic faults, Table 4

NUREG-1 635 20

6 FIRE SAFETY RESEARCH

As the NRC moves from a deterministicregulatory process to one based on an evergreater use of information from quantitativerisk assessment, it is inevitable that somesafety issues of the past will wane inimportance and other issues will be found tobe of greater importance than previouslyanticipated. Several industry-sponsored full-scope risk assessments performed about 20years ago, as well as the insights gainedrecentlyfrom the Individual Plant Examinationof External Events (IPEEE) Program [Ref. 8],have shown that fire is a more importantaccident initiator than might have beenconcluded in the past. The risk from fire-initiated accidents is significant at somenuclear power plants in both an absolutesense and in comparison to accident initiatorsinternal to the plant equipment andoperations.

The importance of fire-initiated accidents wasmade profoundly evident by the 1975 fire atthe Browns Ferry plant. Requirements for fireprotection beyond common industrial fireprotection practices were imposed onlicensees as backfits and compliance withthese requirements has been achieved atconsiderable expense. The agency investedheavily in experimental and analytical toolsfor the analysis of fire consequences atnuclear power plants. Since these earlyefforts, progress has not been commensuratewith the major strides made in thequantification of the probabilities of accidentscaused by internal initiators or the predictionof the progression of such accidents. Firerisks today are typically assessed using whatmay be considered at best semiquantitativebounding methods. Fire progression modelsused in safety assessments are decades oldand do not compare favorably with the state-of-the-art methods available in other technical

4

_7_7n r-=5'7 we

f ' * ~b~ftad*, iv ,b s'p f *a lure1. ':ih tii

. _, .. , Tires.

;. .' .. -. . : C s

-, - . -.- in Fires -i n.

At4 -, , '' t ~ ia 'l. ' ' a- ' f''i;}'a_'.''!--I

nv~jestlgate the pro~i6ilities of .tailures, shtks'to6 -the';- -groihd "' ,,aiid '.ihot"f 'i'sh'orts v >ifth~stmumentati6on"cotro,andpowrcircuits of

ru,:,rpower plants during fires. IR7T'6NRCPals6'completed a' 'reiew. of cableo fire testing

in~d operational t' expence i wvith ,senousnucle'arplant fires incduding The Bro'wns F~errf'ire ;ii .1975 ;an'd;several 'fires --'i foreigri'1Luclemar p ower , plants'i < NUREG/CR ,6834M\>f 9)) Si-nce; then,'.',..a ,mbe~thod -for

.incorporating .-circuit, analyses..-into n.rskassessments is being used by NRCand EPRI66 ire quantify" nsk at.two-representatwe US

fr clear, power plants.;~t*'., :i ft-,{s I.. .t' ;-4.eo;................... '' ; -.

21 NUREG-1 635

areas. Fire-induced faults in l&C systems(hallmarks of the Browns Ferry fire) are stillassessed manually and incompletely despitethe long-held belief that such analyses couldbe effectively done by computer. Currently,there is no capability to include fire as aconsequence of accidents initiated by othermeans in PRAs.

The current NRC research efforts in firesafety (See Table 4) seem incongruent withthe estimated risk significance of fire. Firemodeling benchmark and validation studies(Y6688) are intended to show where agencymodels stand relative to the state-of-the-artmodels rather than advancing the modelingcapabilities. The fire risk 'requantification'studies being done cooperatively with theElectric Power Research Institute (EPRI) willhone approximate risk assessment methodsrather than making these methodscommensurate with risk assessment methodsavailable for accidents initiated by internalevents. Lack of adequate risk information forfire comparable to that provided by theNUREG-1 150 study of internal initiators [Ref.10] will continue to hamper the significancedetermination process for deficiencies in thelicensees' fire protection programs especiallyif these programs are guided by the National

Fire Protection Association (NFPA) standard805 [Ref. 11] rather than Appendix R to 10CFR Part 50, "Fire Protection Program forNuclear Power Facilities Operating Prior toJanuary 1, 1979."

The limited fire research effort in comparisonto the estimates of risk posed by fire isunderstandable since resources for suchresearch have been diverted to respond tothe events of September l1, 2001. Now thatthese responses are being completed, theagency should revitalize its fire safetyresearch efforts and move the technicalcapabilities of the agency to be more in linewith the risks being ascribed to fire. Therevitalized efforts should include:

* Automated methodsassociated circuits analysis

* Improved models ofprogression and damage

* Better integration of fireincluding induced fire, inPRAs

for

fire

risk,plant

* Development of methods toestimate the reliability of humanperformance under fire conditions

NUREG-1 635 22

Table 4. Fire Safety Research Activities4

Project

Y6037 Fire Risk Assessment

* Tools for circuit failure mode and likelihoodanalysis

* Tools for fire detection and suppressionanalysis

* IEEE-383 rated cable self-ignited firefrequency analysis: feasibility study

* Fire modeling toolbox: input data andassessment

* Experience from major fires

* Cable failure mode likelihood studies

* Fire risk 'requantification' studies

* Fire significance determination process

* Fire risk assessment tools precursor analysis

* Support for ANS fire PRA standard

A Program with a large number ofimportant elements for fire safety.Activities support the ROP andimproved risk analysis of nuclearpower plants.

Y6688 Fire Model Benchmarking and Validation Assesses the current status; shouldbe used to define whatimprovements agency needs in itsmodeling capabilities.

Y6651 Effects of Switchgear Aging on Energetic Deals with a significant fireFaults protection issue.

4 See also Collaborative Research on Wire System Aging (Y6410) in Table 3 and Low-Power and Shutdown Risk Study - Level 1 (W6904) in Table 9.

23 NUREG-1 635

7 REACTOR FUEL RESEARCH

Reactor Fuel Research at the NRC iscompleting a quintessential example of risk-informed, confirmatory research. By now, thebasis of the work is well known though it stillprovides an object lesson in the need tosustain technical expertise in selected areasof research crucial to the agency's mission.

In the past, the NRC research provided dataand models of fuel behavior under off-normaland accident conditions while fuel vendorsand licensees focused their attention on fuelbehavior during normal operations. Whentypical fuel burnups at the end of life wereless than about 30 GWd/t, a point wasreached in the NRC work that understandingof fuel behavior was thought to be sufficientlysophisticated and there was a confidence thatlicensee proposals to use fuel at higherburnups could be evaluated by extrapolationwith existing fuel models. As a result, theNRC-sponsored research in fuel performancewas greatly curtailed in favor of otherpriorities. End-of-life fuel burnups creptupward even though it was known in thereactor fuel community that microstructural*changes in the fuel occurred .at averageburnups exceeding about 52 GWd/t.

Testing of high-burnup fuels in France andlater in Japan and Russia [Ref. 12] showedthat high-burnup fuels were susceptible todamage and there was the potential for loss-of-fuel coolability during reactivity-initiatedaccidents. In response to these experimentalfindings, the NRC made the regulatorydecision that adequate safety would bepreserved by limiting fuel burnups to 62GWd/t unless requests for higher burnupswere supported by persuasive evidence ofadequate fuel capability. A confirmatoryresearch program was initiated tosubstantiate this regulatory decision.

Eaillooned and ruptured high-bumup'...-fuel 'ro'dfollowing a LOOA test and

'- ..'.ibbseque n 'tteg.

MVC is using these tests of high-bumup fuel,along'' with extensive: 6metallurgica ;and

rna'teriais .;t£stin'g, toe con firm' '.~rgula'toirydecisions bn'the safe use of 'reactor fuels atbumups of 4'to'62 OWdAf.tThese'testsare

art of alargr internatlo'hal effdrt'toexaminethe effects-of highfuel burnup' and newhfuellladings on reactor behaviol: under designbasis. and -.s'evere, i;:accident chditiosFestinig. ould.- )idcusde I Integral ,4ests';. ofegradatfon andfissioh Priduct release.. -I

Though the technical issues presented byhigh-burnup fuel dealt with design basisaccidents, the scope of the confirmatoryresearch was, designed by, using riskinformation. This risk information pointed toPWR reactivity accidents initiated by controlrod withdrawal, :LOCAs and anticipatedtransients without scram (ATWS) in BWRs tobe of risk concern with respect to high-burnupfuel. On the other hand, BWR rod dropreactivity accidents were found not sufficientlyprobable to merit explicit experimentalinvestigation.

25 NUREG-1 635

- :

This risk-significant research has also beenwell organized. Technical issues andconsiderations were identified using aprestigious panel of international experts incomprehensive phenomena identification andranking efforts. Both the vulnerabilities of thefuel and the probabilities of threats have beenconsidered in these efforts. The researchitself was organized in several cooperativeagreements with other Nations, which greatlyleveraged the resources the NRC was able todevote to this effort.

Research on high-burnup fuel behaviorduring reactivity-initiated accidents is nowreaching a conclusion with the definition ofusable criteria for fuel performance. The fuelperformance computer codes, FRAPCONand FRAPTRAN, have been upgraded totreat fuel taken to elevated levels of burnup.

High-Burnup Fuel Research efforts are nowfocusing on fuel behavior under LOCA andATWS conditions. In connection with theseresearch efforts (See Table 5), the ACRSmakes the following recommendations:

* The NRC should withdraw from theCABRI water loop Project (W6832) iftests planned in this area cannot bechanged to involve more appropriatepulse shapes and energies foridentifying cladding failure thresholds.

* Single rod tests of high-burnup fuelbehavior under LOCA conditionsshould continue and should include the"quenching" phase of the hypothesizedscenario. Researchers should considerwhether bounding peak cladtemperature conditions should beimposed on the single rods or whethermore realistic but limiting single rod (orassembly) temperature scenarios aremore meaningful.

* If single rod tests continue to showbehavior that is not extraordinarilydifferent than anticipated based onlow-burnup fuel behavior,considerations should evolve fromissues of fuel embrittlement to issuesof fuel coolability. The NRC shouldconsider joining the internationalefforts being planned for in-pile tests ofmulti-rod fuel bundles under LOCAconditions.

* Research on mechanical propertiesand oxidation kinetics of high-burnupcladding, including Zirlo and M5cladding (Y6367) should continue. Thisresearch will allow amending 10 CFR50.46, "Acceptance Criteria forEmergency Core Cooling Systems forLight-Water Nuclear Power Reactors,"so that alternative fuel claddings thatare important both to the nuclearindustry and to the NRC can be used.Research on El 10 cladding (Y6723)should not be an emphasis for theNRC since there are no indicationsthat the licensees will use thiscladding. Limited efforts to understandwhy this alloy performs poorly incomparison to the compositionallysimilar alloy M5 might be of use, butshould not require continued use ofEl 10 in tests to characterize alloyresponse to accident conditions.

* Knowledge and understanding that hasbeen gained in the resolution of thereactivity-initiated accident issues aswell as in the resolution of the LOCAissues of high-burnup fuel may make itpossible to resolve by analysis theATWS issues of high-burnup fuelwithout explicit testing of ATWSscenarios. This possibility should beexplored by RES.

NUREG-1 635 26

* Plans to upgrade fuel performancecodes (Y6580) to address MOX fuelshould be pursued. The MATPROdatabase documentation should beupdated to reflect modern data actuallybeing used in the fuel performancecodes.

Understanding of reactor fuel is crucial to theNRC mission. It is widely anticipated that thenuclear industry will be proposing furtherchanges in fuel cladding and fuel burnup.Consequently, the agency must maintainexpertise in the reactor fuel area. Expertiseoutside the agency and its contractors areprobably not sufficiently independent of fuelvendors and licensees to be used on an asneeded basis to review fuel proposals bylicensees. A sustained NRC research effort inreactor fuel behavior under accident and off-normal conditions is especially needed ascompetitive pressures force fuel vendors to

curtail their research efforts. Continuingexpertise and even additional research maybe needed if, as now expected, licenseesmake requests for extending fuel burnups up-to and beyond 75 GWd/t. It is quite likely thatsuch industry proposals will be substantiatedby minimal experimental' research. It isimportant for the NRC to have a technicallyjustified position on the experimentalresearch that is necessary to support suchproposals especially in light of the "surprise"that accompanied extending f uel burnups intothe range of 50 to 60 GWd/t.

27 NUREG-1 635

Table 5 Research Activities in Reactor Fuel


W6832 CABRI Water Loop This Project's focus has changedat the behest of partners and no

International program for in-pile testing of high-burnup longer meets regulatory needs infuel behavior during reactivity-initiated accidents reactivity accidents. NRC should

withdraw from this activity if thetests planned for this facilitycannot be changed to involvemore appropriate pulse shapesand energies for identifyingcladding failure thresholds.

Y6195 Dry Cask Storage License for High-Burnup Behavior of high-bumup fuel inFuel dry cask storage.

Y6367 High Burnup Cladding Performance An essential Project to verifyefficacy of the existing fuel

Testing of high-burnup fuel clad properties damage criteria for high-burupfuel operation. Results provideaccess to data from foreignprograms.

Y6580 Fuel Code Applications for High Burnup Fuel FRAPCON and FRAPTRAN areessential regulatory tools and

Upgrade the fuel behavior codes FRAPCON and need to be kept up to date asFRAPTRAN for high burnup fuel licensees propose ever more

aggressive use of fuel.

Y6586 Fuel Code Assessment for MOX Fuel FRAPCON and FRAPTRAN willbe used extensively to assess

Modification of FRAPCON and FRAPTRAN for MOX proposals for using MOX fuel infuel to dispose of excess weapons grade plutonium commercial PWRs.

Y6723 International Agreement on Fuel Behavior Research on boron dilutionand Materials Science Research accidents and on El 10 cladding.

Since there are no indicationsthat the licensees will use theEl1 0 cladding, research on thiscladding should not be anemphasis for the NRC. Limitedefforts to understand the poorperformance of this alloycompared to the similar alloy M5might be of use.

NUREG-1 635 28

NEUTRONICS AND CRITICALITY SAFETY

Neutron transport modeling and criticalitysafety are closely related to fuel safety butare distinguished here.'These technical areasare also crucial to the agency mission. It isessential for public confidence, if nothingelse, that NRC has the capability toindependently ensure neutronic safety ofreactor cores and fissile materials. The toolsthat the NRC uses for these independentevaluations must be maintained at the state-of-the-art level.

The NRC research Projects in NeutronicAnalysis, Core Physics, and Criticality Safetyare listed in Table 6. The PARCS code, ananalytical tool for neutronics analysis, is alsoincluded in Chapter 15, Thermal-HydraulicsResearch, of this report.

maintaining the capabilities of its neutronicsand core physics codes and the associated'database. Upgrades to these codes toaddress both extended fuel bumup and MOXfuels are needed and these needs areaddressed in the research Projects. 'TheACRS anticipates that these extendedcapabilities will be tested significantly in thecertification of the advanced reactor designs.The effort to provide guidance for high-enrichment fuels (Y651 0) reflects a belief thatlicensees will indeed propose eventually toextend fuel burnups beyond 75 GWdIt asdiscussed in Chapter 7, Reactor FuelResearch, of this report. Indeed, DOE issponsoring research to examine fuelsenriched to 7 percent.

The ACRS assessment is that RES is doinga commendable, cost-effective job in

29 NUREG-1 635

Table 6 Research Activities in Neutronics Analysis,Core Physics, and Criticality Safety


Y6320 NEWT Lattice Code Lattice physics cross-sections forMOX and high-bumup fuels.

Y6403 Reactor Core Analysis To compare the PARCS codepredictions for MOX to Frenchdata.

Y6587 Reactor Analysis for High-B urnup Fuel Analyses with PARCS code ofhigh-burnup fuel for transientsand accidents involving rapidreactivity changes.

Y6685 Upgrade Neutronic Code - Nuclear Fuel To upgrade ORIGEN sourceComp/Safety Assessment model for MOX and conventional

fuels.

Y6771 MOX Neutronics To upgrade PARCS code forMOX fuel

Y6510 Extend Fuel Enrichment Anticipatory research for codeand guidance for fuelenrichments of 5-10%.

Y6517 High-Burnup Source Term for Storage To develop chemical andradioassay data for high-burnupfuel to revise guidance on decayheat and shielding for transportand storage casks.

NUREG-1 635 30

9 HUMAN FACTORS AND HUMAN RELIABILITY RESEARCH

The ACRS continues to believe that humanperformance issues will be important for thecontinued safe operation of nuclear powerplants. The quantification of human reliabilityunder accident conditions will continue to beamong the most challenging -aspects ofquantitative risk assessment. Consequently,it is important for the NRC to maintain anactive research program both in the areas ofHuman Factors and Human ReliabilityAnalysis (HRA). The ACRS believes thatissues surrounding human performance inteams (such as the team of control roomoperators) and human performance inorganizations, including the question oforganizational reliability, are important andneed to be considered in ongoing HumanPerformance Research planning. These areboth important to usafety management', aconcept that Chairman Diaz discussedrecently [Ref. 13].

The NRC research activities in the areas ofHuman Factors and Human ReliabilityAnalysis are listed in Table 7. Human Factorsis a broad discipline that extends well beyondthe nuclear arena. The NRC researchstrategy with respect to human factors is toremain aware of pertinent developments inthe field and based on this awareness,provide tools and guidance to line regulatoryorganization to facilitate the regulatoryprocess. The NRC remains aware ofdevelopments in the international nuclearcommunity through its participation in theinternational cooperative Halden Project(B7488) and the ACRS continues to supportparticipation in this Project. The NRC'scapabilities in the area of human factors arelikely to be tested in the near future as plantstaffing requirements are challenged certainlyby advanced reactor designs proposed forcertification and, perhaps, even by existingreactors searching for competitive advantagein a deregulated energy market. The NRC

must have defensible bases for its plantstaffing requirements for nuclear power plantoperations and security. Research isbeginning to address these'staffing issues(Y6630). The attention'NRC is paying toHuman Factors Research is yieldingdividends. The Standard Review PlanChapter 18, "Human Factors Engineering,"has recently been updated to reflectdevelopments especially in the area of digitalsystems. A risk-informed method to screenlicensee submittals for human factors reviewhas the potential of better focusing agencyresources on risk-significant human factorsissues.

The Human Reliability Research at the NRCis undertaking a potentially significant effort toestablish the basis for crediting operatoractions at nuclear power plants (Y6022). Riskassessments now credit only proceduralizedactions for which operators are trained. It is,however, well known that operator actionsoutside of normal procedures can providesignificant safety benefit. This research holds,then, the promise of adding greater realism toquantitative risk assessments.

Quantification of human reliability remains atroublesome feature of risk assessments.Numerous plausible approaches have beendevised that often yield disparate results.The database for the validation of theseapproaches remains distressingly limited.Uncertainties from both parameters and themodels are seldom evaluated. The results ofthe benchmark exercise conducted by theIspra Laboratory of the 'European Union[Ref. 14], although admittedly fairly old bynow, are troubling. These results show thatthe choice of HRA model has a'significantimpact on the results. This is one of the fewareas in Level 1 PRA in which modeluncertainty is significant. The ACRS has notseen a critical review by the NRC of the

31 NUREG-1 635

merits and disadvantages of the existingmodels. Often in the regulatory processsimple, transparent methods such as THERPthat deal only with human errors of omissionor industry-developed models that have notreceived extensive review are used. Theregulatory process seems to accept thesesimple methods in the absence of guidanceon the necessary levels of rigor and accuracyneeded for quantification of human reliability.Acceptance of such simple models withoutsubstantiation by a robust databaseundermines any proposal to develop crediblehuman reliability models.

The ACRS notes that RES is preparing adocument on its assessment of "goodpractices" that will support all aspects ofHRA, including the quantification process(W6994). These descriptions of goodpractices are needed to supplementinformation provided in the ASME PRAstandard. The ACRS understands that thesecond phase of this project (planned for2005) will be a review and evaluation ofexisting HRA approaches for their capabilityto meet the "good practices" when employedto address different regulatory applications.The ACRS anticipates that this review willaddress the important issue of modeluncertainty.

because of both its cumbersome, manpower-intensive character and its lack ofquantification. The NRC is now applying thismethodology to two important issues:

* Human performance during fires

* Human performance issues associatedwith pressurized thermal shock

The ACRS looks-forward to reviewing theresults of these application efforts andlearning whether the method has becomeeasier to apply and whether it yields insightssignificantly better than those that could bederived from simpler methods. Of particularinterest to the ACRS is the use of expertopinion elicitation in these applications toquantify the ATHEANA results.

The staff needs to show how the Davis-Besse incident has affected the set ofperformance shaping factors and whatguidance is being provided to the ROP.

The NRC is developing an events databasecalled Human Event Repository and Analyses( HERA ) (Y6123). This database may be akey step toward much improved analyses ofhuman reliability in nuclear reactors. It shouldprovide an useful basis for evaluating themodel uncertainties mentioned above.

The NRC is experimenting with an ambitiousmethodology for estimating human reliabilityto assess errors of commission as well aserrors of omission. This methodology calledATHEANA (W6994) focuses on the contextthat drives human performance("performance shaping factors"). The ACRShas been critical of this approach in the past

NUREG-1 635 32

Table 7. Research Activities in Human Factorsand Human Reliability


Y6123 Retrospective Risk Analysis of Valuable Project to develop HERA databaseHuman Reliability in Operating on human reliability from past operational

Events experience.

Y6832 Human Reliability Analysis for Identifies human reliability issues for NMSS -Byproduct Material tools of use for quantification of risk.

B7488 HALDEN Reactor Project This is a long-term Project in HumanPerformance that is contributing to NRC workin human factors.

Y6022 Credit for Operator Action This is a small effort to provide guidance forNRC reviewers. It has the potential of addinggreater realism to PRA.

Y6630 Development of Regulatory An important proactive measure in anticipationGuide and Analytical Techniques of trends in current reactors and themes in the

for Assessing Nuclear Power development of more advanced reactors.Plant Staffing

W6994 Review, Application and This is a very ambitious effort to characterizeRefinement of ATHEANA human performance. At issue is whether

ATHEANA can be applied practically to theissue of reactorsafety. This effort will also yieldguidance for performing and reviewing humanreliability analyses and supplementinginformation provided in the ASME PRAstandard.

33 NUREG-1635

10 MATERIALS AND METALLURGY

Research Projects in the area of Materialsand Metallurgy are listed in Table 8. TheseProjects can be grouped into the followingbroad categories:

* Reactor PressureEmbrittlement (W6953,Y6378, W6986, Y6396,Y6737)

VesselIY6533,Y6638,

* Neutron Fluence Determination(Y6391,Y6742)

* Environmentally Assisted Cracking(Y6536,Y6588, K6266, Y6270,Y6388,Y6722)

* Monitoring and Inspection (Y6534,Y6604, Y6649, Y6869, Y6882,Y6909)

* LOCA Frequencies and InternationalStandards (Y6296, Y6538, Y6744)

* Proactive Materials DegradationAssessment (Y6919)

These Projects represent a well-balancedprogram that address current regulatoryneeds, including the need for the Office ofNuclear Reactor Regulation (NRR) tomaintain an independent analytical capability.in various areas of materials and metallurgy.

Capabilities are transitioned to NRR as theyare developed, and new projects areintroduced as their need becomes apparent.The ACRS recommendations on variousProjects in the Materials and Metallurgy areaare given below and summarized in Table8. Many of these Projects are part ofcollaborative programs, involving about 65International Organizations, thereby givinggood leverage of the NRC resources. -

e g o ~ t:'; .4 "

k '; S j > * v ' .; z ~~~~~~~i ,v rLL .> .' il5,. 1 -%i ' '. x AdXell 9

.. :'' -., .,...,,.... '. ''.4..... - ''rSI4.4.44 *''i -" ' ;., d_

-n*ik1r i..ts*$ ,\,m;io-e- i 'v,< -{eRi

C t ,-; %s ._

Exampl ot,. a.. P , .... ^t'!.*red'm.natlAt ,I4. ;'*- Ite-r-ranulaX Stress C roio Crac

-/~~ -'4 a. I; A- ;--4:

Fig~ .J ; ' ' .................................... , . X -0

The Projects dealing with reactor pressurevessel embrittlement should be consideredboth with respect to safety issues of reactorpressure vessel integrity and with respect tomaintaining whatever core competency theagency needs in irradiation embrittlement.The staff needs to evaluate what researchshould be retained in order to meet these twoneeds. The remaining technical issueinvolves high-neutron fluence on the low-copper, high-nickel pressure vessel materialfound in just a few plants (W6953, Y6533).Work in Y6533 will support risk-informing thepressure- temperature limits in 1OCFR Part50, Appendix G, "Fracture ToughnessRequirements." Project Y6638 providessupport for revising Regulatory Guide (RG)1.99, "Radiation Embrittlement of ReactorVessel Materials."

Pr~oject Y6378participation inEnergy Agency

ensures continued NRCthe International Atomic(IAEA) deliberations on

35 NUREG-1635

pressure vessel integrity and will contributeto the preservation of core competency.

Peer review of the technical basis for thereevaluation of the PTS screening criterion(Y6737) is most important and should becontinued.

Work in Project W6986 should beterminated since its objectives have beenmet and fracture mechanics technologydevelopment will be consolidated in Y6533.

The development of analytical capabilitiesfor neutron fluence (Y6391,Y6742),especially in the core regions near thebottom and top of the core, are importantsince cracking of major weldedsubassemblies in BWRs has occurred inthese regions. The ability of the NRC toindependently check the licensees'arguments for disposition and repair of thesecomponents with complex geometries isimportant. Hence, the uncompleted tasksof these Projects should be continued inFY2004.

Corrosion, especially environmentallyassisted cracking, is a complicated technicalissue involving the interplay of chemistry,corrosion science, metallurgy, andmechanics. The current concerns aboutirradiation effects on cracking of corecomponents further complicate this situation.

The management of environmentallyassisted cracking incidents in LWRs is theprime responsibility of the licensees.However, the NRC must have independentcapabilities to assess the licensees'positions regarding the structural integrity ofreactors and their proposed mitigationactions. Indeed, recent events at the IndianPoint and Davis-Besse nuclear plants havedemonstrated that licensees' corrosioncontrol programs merit continued scrutinyand monitoring. Consequently, the NRCshould have active programs in the areas of

materials degradation data collection (inorder to assess the kinetics of degradation),monitoring and inspection (e.g., NDE), andcodification.

A particularly vulnerable element of thereactor coolant system in PWRs is the steamgenerator. Rupture of steam generator tubescan lead to accidents that allow radioactivematerials released from the core to bypassthe reactor containment and enter into theenvironment. Risk analyses show that severeaccidents involving containment bypass canbe risk dominant at some PWRs. Modes ofcorrosion (e.g. cracking, denting) of thesteam generator tubes have changed overthe years as changes have been made inboth the materials used for the tubes and inthe PWR coolant chemistry. The complexityof these phenomena and the risk significanceof tube failures make it important for theNRC to monitor these issues. This need doesnot abate as further changes from alloy 600tubes to alloy 690 tubes are made by theindustry. Although alloy 690 is more resistantto primary water stress corrosion cracking,the complexity of the environments possibleon the secondary side of steam generatorsmakes it impossible to preclude the potentialfor stress corrosion cracking. Therefore, theProjects (Y6588, Y6536) associated with theNRC's Steam Generator Tube IntegrityProgram should be continued.

In recent years, environmentally assistedcracking in LWRs has raised other concernssuch as irradiation-assisted stress corrosioncracking (IASCC) of stainless steel coreinternals of BWRs and stress corrosioncracking /general corrosion of vessel headpenetration (VHP) subassemblies in PWRs.These latter incidents have been the subjectof several Bulletins and an Order.Consequently, it is appropriate for the NRCto sponsor research to independentlyevaluate licensees' plans to mitigate thesephenomena. Projects Y6270 and Y6388should be continued in FY2004 since they

NUREG-1 635 36

provide the essential cracking kinetics datarequired for in-service inspection (ISI)regulations. Project K6266 should also becontinued since it relates to thequantification and understanding of IASCCissues; The value-,of this research isenhanced by collaboration with severalInternational Organizations. Project Y6722was instigated by the Davis-Besse incidentand several of the tasks have beencompleted. This Project should becontinued to complete the remaining tasksand to ensure adequate independentanalysis of the licensees' extensive programspawned by this incident.

Nondestructive examination methods arecrucial for monitoring the integrity of thereactor coolant system. These methods fordetecting corrosion and cracking haveimproved greatly in recent years largely dueto NRC research activities. However, thereliability of these methods in the fieldcontinues to be an issue. Sustained NRCresearch (Y6534, Y6604, Y6649, Y6869,Y6882, Y6909) on the evaluation of thesenondestructive examination methods iscrucial to the agency mission. ProjectsY6649, and Y6869 merit increased attention;these relate to the inspection methodologies(technique and frequency) that should beused for VHP assemblies. In the latterproject, a technical basis for reevaluation ofrequirements related to reactor coolantleakage will be developed.

failure database. This is a further inputneeded for risk-informing 1 OCFR50.46.

Project Y6296 involves NRC participation inthe development of international standardsand should be continued.

Over the years, the NRC has been surprisedby a number of materials degradation issues.The ACRS supports the "Proactive MaterialsDegradation Assessment" Pr6ject (Y6919)recently initiated by the NRC in order to movefrom the economically inefficient reactivemode of regulating materials degradationissues. The plan to lead a PIRT effort withthe industry over the next two years, withexpert elicitation from international experts,should identify potential materialdegradations that may occur in the future.This knowledge will provide a better basis forfuture regulatory decisions on timelyinspection and repair criteria.

A major NRC regulatory initiative is risk-informing 1 OCFR50.46 and redefining large-break LOCA in terms of LOCA frequencydistributions. Part of this work involves thedevelopment a new probabilistic fracturemechanics (PFM) code for piping. Thecurrent PFM code, PRAISE, has been avaluable tool, but an improved code isneeded to better address materialsdegradation modes. Thus, Project Y6538should be continued. Project Y6744maintains NRC access to the CSNI pipe

37 NUREG-1635

Table 8. Research Activities in Materials and Metallurgy

Project Title I Comment

Reactor Pressure Vessel Embrittlement

The remaining technical issue involvesW6953 Heavy-Section Steel Irradiation Program high-neutron fluence on the low-copper,

high-nickel pressure vessel materialfound in just a few plants. Work in Y6533will support risk-informing the pressure-

Y6533 SST-temperature limits in 10CFR Part 50,Y6533 H55T-3 Appendix G.

Y6378 International Pressure Vessel Technology Representation of NRC at IAEA DivisionCooperative Program of Nuclear Power on RPV structural

integrity issues.

W6986 Fracture Mechanics Technology for LWR Should be terminated since its objectivesMaterials have been met and fracture mechanics

technology development will beconsolidated in Y6533.

Y6396 Radiation Embrittlement Damage Analysis Longer term confirmatory research with& Prediction current emphasis on embrittlement at

end-of-life fluences.

Y6638 Statistical Analysis of RPV Steels Longer term support for rev 3 of RG 1.99Y6737 Peer Review of PTS Technical Basis A critical Project at the end of a very

successful program on reevaluation of thePTS screening criterion.

Neutron Fluence Determination

Y6391 Boiling Water Reactor Fluence Support for development of independentNRC analysis of Industry code for neutronfluence. Uncompleted tasks of thisProject should be continued in FY2004.

Y6742 BWR Reactor Vessel Samples Measurement of helium and boronconcentrations in BWR core samples andcorrelation to fluence. Critical to weldrepair decisions. Uncompleted tasksshould be continued in FY2004.

NUREG-1 635 38

Table 8. Research Activities In Materials and Metallurgy(Continued)

Project Title | CommentEnvironmentally Assisted Cracking

Y6536 PWR Primary System Components' Addresses the potential for steamBehavior Under Severe Accident Loads generatorbypass in severe accidents and

should be continued.

Y6588 Steam Generator Integrity Program -3 Addresses potential degradation modesfor new, degraded or repaired tubes andan assessment of their consequences(e.g. leak rates) and control (e.g., ISI)techniques. Should be continued.

K6266 CIR-Il Cooperative Agreement Cooperative program, organized by EPRIto predict the irradiation assisted crackingcharacteristics of new and existing LWRcore materials. Good leverage of NRCfunds. Should be continued.

Y6270 Environmentally Assisted Cracking Evaluation of IASCC and fracturebehavior of stainless steels used BWRcore structures. Relevant to licenserenewal. Should be continued.

Y6388 Environmentally Assisted Cracking in Quantification of a variety of crackingLWRs phenomena (corrosion fatigue, IASCC) to

develop independent life predictioncapability for low alloy steel, stainlesssteel and nickel base alloys in LWRs.Should be continued.

Y6722 Degradation of RPV in Boric Acid Project instigated by Davis-Besseincident. Several tasks have beencompleted. Should be continued tocomplete the remaining tasks.

39 NUREG-1 635

Table 8. Research Activities in Materials and Metallurgy(Continued)

Project |Title I CommentlMonitoring and Inspection

Y6534 Piping - Non-Destructive Examination Cooperative international program to(NDE) Reliability develop NDE methods. Should be

continued.

Y6604 Evaluation of the Reliability of NDE Evaluation of effectiveness, reliability andTechniques adequacy of advanced NDE methods

necessary for e.g. cast stainless steels.Should be continued.

Y6649 Alloy 600 Cracking - Phase II Merits increased attention. Relates toinspection methodologies to be used forVHP assemblies.

Y6869 Barrier Integrity Research Program Merits increased attention. Providestechnical basis for reevaluating therequirements of reactor coolant leakage.

Y6882 Technical Assessment of Bare Metal Evaluation of the inspection techniques toInspection Techniques be used for PWR bottom head

penetrations. Should be continued.

Y6909 Examination of North Anna Unit 2 Head Establish, via inspection of nozzles fromNorth Anna Unit 2 and Davis-Besse, thedevelopment of the leak path ascracking/degradation occurred. Relatesto adequacy of BM_

NUREG-1 635 40

Table 8. Research Activities in Materials and Metallurgy(Continued)

Project Title I CommentLOCA Frequencies and International Standards

Y6296 ISO Participation Ensure NRC participation in thedevelopment of international standards.Should be continued.

Y6538 Technical Development of LOCA Support for risk-informing 1 OCFR 50.46Frequency Distributions and redefining LBLOCA. Should be

continued.

Y6744 CSNI Piping Data Base Cooperative program to maintain NRCaccess to CSNI pipe failure database.Should be continued.

Proactive Materials Degradation Assessment

Y6919 Materials Degradation PIRT Project recently initiated by the NRC. Theplan to lead a PIRT effort with the industryover the next two years, with expertelicitation from international experts,should identify potential materialdegradations that may occur in the future.This knowledge will provide a better basisfor future regulatory decisions on timelyinspection and repair criteria.

U - -

41 NUREG-1 635

11 PROBABILISTIC RISK ASSESSMENT

As the NRC moves toward a risk-informedregulatory system, it will need ever greateramounts of risk information derived fromquantitative risk assessments. It will want tomaintain its own independent capabilities forconducting PRAs at state-of-the-art levels. Acontinued vital research program in PRA is,then, crucial to the agency mission.

The NRC research activities in the area ofPRA are listed in Table 9. A major thrust isthe support for the ROP which is a majorNRC initiative in the use of risk information formonitoring the operations of nuclear powerplants. Support for the development of theSAPHIRE code suite and the SPAR modelsis important and continuing (Y6394, W6355,W6467, Y6153, Y6553). Also under way insupport of the ROP is the development ofobjective, risk-based, performance indicators(J8263, Y6370). A candidate metric calledthe Mitigating System Performance Index isbeing tested. There is also continued supportfor the significance determination process(Y6553) which is an important step in theassessment of findings from the ROP thatcontinues to challenge the agency.

Another major element of the current PRAResearch is the expansion of the method intothe analysis of dry cask storage of spentfuel(Y6423, Y6502, Y6612). Such storage ofspent fuel will become ever more necessaryas the development of a permanent geologicrepository for spent fuel continues to bedelayed.

Research activities of importance are alsounder way to support risk-informed revisionsto 10 CFR 50.46 (W6224, Y6538).

In addition to the many demands onresources for the application of PRAmethods, resources are still made availablefor developing agency capabilities in PRA(K6007). The ACRS generally supports theidea that the NRC needs to maintain its PRAcapabilities at or near the state-of-the-artlevel. A detailed examination of the tasksfunded within this developmenteffort showsthat many overlap with distinct efforts in otherfields. Some items with substantial overlapinclude the study of turbulent mixing incomplex geometries and software reliability.Greater coordination among research withinthe NRC may eliminate the need for suchoverlapping research efforts.

The ACRS would prefer to see work toexpand the scope of the NRC PRAcapabilities. Work to develop capabilities toanalyze risk during low-power and shutdownconditions at nuclear power plants (W6904and Y6103) is very important. The ACRSstrongly supports these efforts that will extendPRA capabilities toward the eventual goal ofhandling all modes of plant operation. Workto better integrate fire-risk assessment,including fires induced by other accidentinitiators, is a need not now being metdespite evidence of risk significance comingfrom the IPEEE Program.

Altogether the NRC has an impressive PRAResearch Program. This Program is all themore impressive since many of the agencyresources in risk analysis are being applied tosecurity issues that are not addressed in thisreport.

43 NUREG-1635

Table 9. Research Activities in Probabilistic Risk Assessment'


Y6423 Dry Cask PRA Expanded application of PRAinto an issue that will be of

Y6502 PRA for Dry Cask Storage - Peer Review growing importance over the nextfew years.

Y661 2 Reliability of a Secondary Containment Radionuclide release paths forIsolation (SCI) accident involving drop of a dry

spent fuel storage cask.

J8263 Reactor Oversight Process Support Essential support for the ReactorOversight Process.

Y6394 Maintain and support SAPHIRE Code andLibrary for PRA

W6355 SPAR Model Development: LowPower/Shutdown

W6467 SPAR Model Development: Level 1 Rev 3

Y6153 SPAR Model Development: Level 2ILERF

Y6553 Significance Determination ProcessFront-end Interface for SPAR Models

Y6370 Development of Risk-based Performance Exploratory activity that hasIndicators yielded a candidate metric that is

now being tested.

Y6522 Synergistic Effects of Power Uprates Important effort to identify issuesnot revealed in the examinationsof individual changes to thelicensing bases of nuclear powerplants.

K6007 Probabilistic Assessment and Applications Program of several activities toimprove PRA capabilities.

* Model and parameter uncertainty* Software reliability Greater coordination among* Thermal-hydraulic uncertainty research within NRC may* Turbulent mixing within complex geometries eliminate the need for* AP1000 LBLOCA uncertainty analysis overlapping research efforts in* Importance analysis this Project.* Causal modeling

5 See also Y6332, Digital Systems Risk, Table 3

NUREG-1 635 44

Table 9. Research Activities in Probabilistic Risk Assessment(Continued)


W6144 Technical Support in Risk Assessment Staff augmentation

W6224 Risk-informing 10 CFR Part 50 Support for the effort to risk-inform 10 CFR Part 50

Y6538 Technical Development of LOCA FrequencyDistributions

W6904 Low-Power and Shutdown Risk Study - Essential study of one of theLevel 1 important elements of risk posed

by nuclear power plants duringlow-power and shutdown

Y6103 Low-Power and Shutdown Risk Study- operation that has beenLevel 2 inadequately studied in the past

W6970 Support to Develop Consensus PRA Necessary NRC participation inStandards standards setting activities

Y6335 Risk-informed Initiatives for Nuclear Safety goals and guidance forMaterials nuclear materials and waste

Y6430 PTS Risk Assessment Activities in support of theagency's reevaluation of the PTSscreening criteria

Y6486 Severe Accident Initiated Steam Generator Essential element of the agencyTube Rupture Sequences strategy to resolve issues of

steam generatorsL

45 NUREG-1 635

12 RADIATION PROTECTION

The NRC research activities in the area ofRadiation Protection are listed in Table 10.These activities constitute a well-leveragedprogram that allows the NRC to use qualitydata and information as the foundation of itsbasic radiation protection regulation, 10 CFRPart 20. A major thrust of these activities isthe collection of occupational exposure data(Y6133 and Y6698). Another important thrustis the support for. commissions (ICRP andNCRP) that establish radiation protectionstandards (G6071, G6251, G6589). Thereare also activities that support doseassessment by the NRC (Y6433, Y6460,Y6470). These are all essential activities forthe agency mission and should be sustained.

A potentially important activity (Y6330) wouldharmonize requirements from multipleagencies for complex site assessments. Thisis' a laudable joint effort with theEnvironmental Protection Agency (EPA) todevelop common databases for use inenvironmental dose assessment calculations.Another cooperative effort (Y6470) is yieldinga modeling platform that gives access to awide range of system and process models aswell as databases used by other governmentagencies. This platform will eventually yielda powerful environmental modeling capabilityfor use in the evaluation of complex,contaminated sites.

47 NUREG-1 635

Table 10. Research Activities in Radiation Protection

Project Title Comments

G6071 BEIR - VIlThe Phase II study deals with low doses and low dose NRC contribution to therates. The BEIR-VIl effort is to conduct a comprehensive development of consensusreassessment of health risks resulting from exposure to standards in radiationionizing radiation since the 1990 BEIR-V [Ref. 15] report protection.

G6251 Radiation Protection Standard Development ICRP support; ICRP isICRP task groups will provide information on radiation- considering the revision ofassociated risks of cancer and severe heredity effects, ICRP-60 [Ref. 16] by 2005 thatapplication of recommendations in terms of secondary simplifies the current system oflimits, and the principles of optimization and radiological protection.management of radiological protection

G6589 NCRP Study on Radiation Protection Issues NCRP support.

Y6133 Collection and Analysis of Occupational Radiation Analysis of required licenseeExposure Data submittals on occupational

exposures.

Y6298 Technical Bases Information Clearance Information to supportCommission decisions onrulemaking.

Y6330 Support for Interagency Cooperative Research Interagency harmonization ofcomplex site evaluations.

Y6407 NUREG-1640 [Ref. 17) Finalization - Clearance Development of conversionfrom dose to concentration formaterials

Y6433 Extremity Code Update the VARSKIN code toThis Project is to correct errors and limitations in the confirm licensee skin doseVARSKIN Mode 2 computer code identified by licensees assessments.

Y6460 Update Codes for Assessing Radiation Doses Update dose coefficients

Y6470 FRAMES Software Development Well-leveraged effort toprovide agency a doseassessment tool for complexsituations. Good example ofinteragency cooperation.

Y6698 Support NEA Information System on Occupational Essential data for agencyExposures mission.

Data are used to compare U.S. and foreign nuclearpower plant initiatives and identify best practices duringmaintenance and operational procedures

NUREG-1 635 48

13 SEISMIC RESEARCH

Seismic events produce a baseline risk fornuclear power plants. The NRC has investedheavily in the understanding of seismic issuesand development of its rules to ensure thatseismic risks are acceptably low. Licenseesubmittals concerning risks associated withexternal accident initiators such as seismicevents suggest these efforts have beenlargely successful. Issues concerning theestimation of seismic hazards of power plantsites in the eastern United States based onmethodologies developed independently byEPRI and NRC have been resolved. TheNRC research activities in the seismic areaare listed in Table 11.

Much of the NRC research into seismicissues involves cooperative efforts to addressessential seismic data. This work is highlyleveraged by contributions from otherpartners. There is advantage in maintainingmodest research efforts to accrue benefitsfrom cooperative research on seismology andseismic engineering.

Projects W6829 and Y6718 are addressingspecific issues that should be done bylicensees. It is not clear that the NRC needsan independent assessment of these issues.Unlike many other issues that are dealt withby the NRC research, there is in this Countrya very substantial engineering and scientificcommunity addressing issues of seismichazards and the engineering of structures toavoid seismic damage.'By'and large, thiscommunity is independent of the NRClicensees. Thus, the NRC-sponsoredresearch is not needed to maintain a corecompetency in this area. If there are openissues associated with licensee submittalssuch as the seismic behavior of spent fuelcasks or buried structures, the NRC couldrely on use of engineering consulting firms onan as-needed basis to deal with such issues.

, ~s.. ...-A Succ~ssStorY , -7 TZT ¢--Probabilistic Seismi'c Hazard Anilysis:..','Guidance.-on' Unce'ifaInty and Use'of-:'

~ariy^ lndustry-sponsored ,~, PRAs, -such-as.thbse'foi:Zion ~and Indian Point, s'howedtheWigniicance of-seismic contrbtutiohs"ror' inuclear ' psowersplants.; 'Later studiesconfirmed-these finding1s,.`'A'idifficulty 'iinfruantifying 'seismticrsk,'especiallyf orplantseast 'of the; ocky Mountains,w th

sparsity'of data. Thisled oextensive use ofexpert~'-judgiment .;-with :,-attendan't iargelunceJtainties,: .in.'.the. ,inIyses. srnportantrobabilis tic ,-, ,teismic :z hazard ',! analyses

sponsored-,by the -NRC-andbji.EP.RI in the9--980s hig' -different.expe.rts andiiffererii

.tS *I' ,r o'!s6. .'., L ' ii

ihe, .controversyK associatedi with .fthes6ivbrgent results was not conducive to public

confidence -lh the regulat or. proess.orear'ess thlssltuation, NRC, the Departmenr

nf "Eergy,- aid. ,EPRI, "flm eJd the "Senior;eismic", Hazard ..Analysls.i.iCommitteeSSHAC) top'r6ide.a procedure for btaining'

heprdd~icible tesults. In a two-volume report".SSHAC'ldentified -keyissues 'egarding theh&fantification :612-; unc'ertainties ',anhd.,' the,btIlization tof expertopbinion." ^;: ,.>ic;k.*X5>

A National Researclh co 'ciuc Panel revie!wed*his work ard concluded tha't. ".-';the SSHAC'epor's disussionsandrecommendations on'

.vncertainty and theuse-'ofexperts are quiterndependent'of PSHA~and-can be applied toother -types- of lrisk..anslysis.' 'The '4Panelbeiieves that theSSHAC report makes a solidcdntribution'. to, the 'methodologyl of :hazardanalysis, -especially -in the, use of expertopinion [Ref v'18],,A,^ '-t.-i''",':^, '

49 NUREG-1 635

Table 11. Research Activities in Seismic Phenomena


W6081 Japanese Collaboration on Seismic Issues Five-year cooperative agreementwith Japan. Highly leveraged andshould be sustained

W6829 Seismic Behavior of Spent Fuel Storage Why should NRC do this workCask Systems rather than licensees?

Y6166 Cooperative Geoscience Research Continuing work on seismichazard of central and easternU.S. Adequate understandingnow available for regulatoryprocesses. The need for thiswork is questionable.

Y6233 Garner Valley Downhole Seismic Array Ongoing data collection of limitedOperation and Analysis of Data potential impact on regulatory

processes.

Y6481 Senior Seismic Hazards Analysis Committee Expert opinion elicitation for 10-(SSHAC) Method year update of a semiquantitative

analysis for 10 CFR Parts 72[Ref. 1 91 and 100 [Ref. 20].

Y671 8 Soil-Structure Interactions for Buried It is not apparent that NRC needsStructures to conduct this research even in

anticipation of advanced reactortechnology. It is only necessarythat NRC have available to it theresources necessary toadequately and independentlyreview licensee submittalsconcerning this issue.

Y6796 IAEA Coordinated Research Project on This is primarily an effort toSeismic Ground Motion acquaint other Nations through

IAEA with NRC data andinformation on ground motionfrom nearby earthquakes ofmoderate intensity.

Y6797 Evaluation of USGS 2002 Seismic Hazards Support for SSHAC method.Assessment Provides seismic hazard

information for central, easternand other regions of U.S.

NUREG-1 635 50

14 SEVERE ACCIDENT RESEARCH

,_.

In the past, the NRC invested very heavily inthe experimental and analyticalcharacterization of severe reactor accidentsthat dominate the risk to the public posed bythe use of nuclear reactors to produceelectrical energy. A very substantialtechnology has been established tounderstand the progression of reactoraccidents and the radiological consequencesof such accidents. Once its immediate needswere met to understand severe reactoraccidents sufficiently well to estimate risks tothe level of confidence needed and to provideassurance of adequate protection, the NRCsubstantially curtailed its investments insevere reactor accident research. The NRCSevere Accident Research activities otherthan those directly associated with advancedreactors are listed in Tables 12 and 13.

Research on Severe Accidents has beencontinuing in other Countries because theyhave more restrictive acceptance criteria thanadequate protection. Substantial programsare under way in both Europe and Japan.The NRC has developed quite an effectivestrategyto maintain the technologyforsevereaccident analyses and to update thistechnology with research results frominternational programs. The body ofknowledge coming from the NRC's past workand ongoing work are systematized inuseable form in the MELCOR accidentanalysis code. At the same time, the NRC isentering into international cooperativeresearch programs to obtain data forvalidating the MELCOR code and improvingits accuracy and realism. Current agreementsinclude the following:

%R.:eblistic Reactor Accident Analyses H

MELCOR Prediction of PHEBUS DataHydrogen Production FPT-1

0.18

{t|- MELCORl* * PHEBUS Data

U0.12

0.06

0.002t

-0.0618000 12000 16000

Time (seconds)

MELCOR Prediction of PHEBUS DataIodine Release Fraction FPTA1

. 1.0 ] z | z . . . I

o 0.8-1tU

U. 0.6-

410

0 0.4

. 2

aMELCOR| * PHEBUS Data I

nnP-. . . I . .. . I

8000 12000 16000 20000

Time (seconds)

RES.':is>sic6eedingg'~iri~achieving,^-greateruracy and;realisi' ii~. its iiseverereactor

a'c cident .£ea naly y~s,. . 7 7w-'. , .xfor-.~1plant' ;risksments;by_.validatfig MELCO?,F model

predictions , with' data from -severaljntemationab . cooperative experimentalprograms. Figures above show campansonsbf MELCOF code predictions with 'datafromthb HEBUS- FPT-'1in-pile testof irradiatedb- a 'c to r f u el d e gra 'd a tio n . A; ' ' *,- *.-l,+~S' -2, d

51 NUREG-1 635

0 MASCA * OECD-MCCI

This is an experimental study underwayin Russia on the behavior of reactor coredebris in the lower plenum of a reactorpressure vessel. Results of these studieswill be needed to evaluate safety claimsrelated to the certification of both theAP1000 and ESBWR designs.

* ARTIST

This is an experimental study under wayin Switzerland to measure the aerosolremoval on the secondary sides ofsteam generators during accidents atPWRs that bypass reactor containments.Such bypass accidents are often riskdominant for PWRs. The high risksassociated with such accidents maystem from conservatism in the aerosoldecontamination assumed in accidentanalysis models for steam generators.Test results are expected to provide thebasis for more realistic analyses of theseaccidents.

* PHEBUS-FP

This Program consists of five large-scale, in-pile integrated tests of fueldegradation, fission product release,radionuclide transport through thereactor coolant system, and aerosolbehavior in the containment. These testshave been designed to validate reactoraccident models. Exceptionallyprototypical results for code validationare being produced by this internationalprogram (see sidebar). Additionalinformation is being provided bysupporting separate-effects experimentalprograms, such as the French ProgramVERCORS, to investigate fissionproduct release from MOX and high-burnup fuels under accident conditions.

This is an experimental study of theviability of using an overlying layer ofwater on reactor core debris that hasescaped the reactor coolant system andis interacting with structural concrete ofreactor containments. This research willprovide data for improved or new modelsof core debris coolability for systems-level accident analysis codes. Improvedmodels will be used in the risk analysisof existing plants and the safetyevaluation of advanced reactor designs.This Program is to be completed inFY2005.

These highly leveraged internationalprograms are providing the bases forvalidating the NRC's accident analysis code,MELCOR. The MELCOR code has proven itsutility to the regulatory process in recentyears in connection with resolving issuessuch as the need for hydrogen igniters in icecondenser and Mark IlIl containments andrisk-informing 10 CFR 50.44, "Standards forCombustible Gas Control System in Light-Water-Cooled Power Reactors." MELCORanalyses are also important for thecertification of the AP1000 and ESBWRadvanced LWR designs. MELCOR analyseswill likely be instrumental in the certification ofother advanced reactor designs.

The ACRS is very supportive of the strategythe NRC has developed to maintain andupdate its capabilities for accident analyses.The ACRS encourages the NRC to giveserious consideration to joining internationalcooperative research agreements now beingproposed to examine:

* Prototypic, in-pile tests of radionuclidereleases from, and degradation of,reactor fuel exposed to air for bothreactor accident analyses and spent fuelstorage pool accidents

NUREG-1 635 52

* Prototypic, in-pile tests of radionuclidereleases from, and degradation of, high-burnup fuels

* Prototypic, in-pile tests of radionuclidereleases from, and degradation of, MOXfuel

The ACRS also supports programs tomaintain the MACCS code' and looks forwardto reviewing results of the comparison of theGaussian plume model to state-of-the-artdispersion models in the context of riskassessments.

o Sound technical basis for evaluatingenergetic fuel-coolant interactions

* Aerosol behavior in steam generatorsecondaries and other pathways thatbypass reactor containments

* Effects of high burnup on core-meltprogression and fission product release

The ACRS foresees the followingchallenges in the area of severe accidentanalysis:

* Assessment of in-vessel retention withrespect to materials interactions

* Spent fuel pool accident progressionassessment especially the effects ofhydrides, effective ignition temperature,extent of pool involvement, and releasesof fission products and actinides, as wellas transport energetics

* Evaluation of severe accidentuncertainties for PRA applications

53 NUREG-1 635

Table 12. Severe Accident Research Activities


Y6312 MASCA Program Supports certification of the AP1000International cooperative research on core and ESBWR designs. The MASCAdebris interactions with the pressure vessel experimental efforts will be completedhead soon.

Y6313 OECD- Melt Coolability and Concrete International cooperative research onInteractions Program core debris coolability during

interactions with concrete.Y6668 Analytical Support for Y6313

Y6321 Benchmark MOX Fuel Release Source Research to define fission productTerm Experiments release from MOX fuel.

Y6377 MELCOR Code Development and Repository for research resultsAssessment generated in cooperative international

programs. Code is being usedY6802 MELCOR Severe Accident Code extensively for revising regulations and

Development and Assessment certifying advanced LWRs.

Y6504 Steam Generator Fission Product Effort to add realism to predictions ofRetention releases during bypass accidents.

Y6607 Support for ARTIST Tests of FissionProduct Retention in Steam Generators

Y6571 CSARP Meeting Useful annual meeting to reviewcooperative research results.

Y6512 Tests of Severe Accident Phenomena Phenomenological tests and analysesin Oxidizing Medium for spent fuel pool accidents.

Y6664 Containment Analysis and Experiments Support for AP1 000 certification.for AP1000

Y6668 Containment Analysis Support Containment analysis of advancedreactors.

Y6848 High Burnup Fission Release Data Receipt and analysis of data fromFrench VERCORS tests.

Y6696 AP 1000 Severe Accident Phenomena Severe accident phenomenaEvaluation assessment of AP1 000.

Y6721 Agreement with IBRAE-RAS on Nuclear Very cost-effective effort to modernizeSafety Analysis Codes the MELCOR code.

NUREG-1 635 54

Table 13. Research Activities for Accident Consequence Models


Y6785 Plume Model Adequacy Evaluation Test the assumption that simpleplume treatments in MACCScode are adequate by comparingwith the state-of-the-artdispersion model.

Y6628 MACCS Uncertainty Assessment for Support for emergency planning.Consequence Models

Y6469 Evaluation of Radionuclide Pathways and Upgrade information on uptakeUptakes pathways.

55 NUREG-1 635

15 THERMAL-HYDRAULICS RESEARCH

The NRC regulatory process makesextensive use of thermal-hydraulic analysesfor certifying advanced reactor designs,processing changes in the licensing bases ofexisting reactors, and investigating thermal-hydraulic safety issues. To meet the manyneeds for.such thermal-hydraulic analyses,the NRC developed and maintained severalthermal-hydraulic transient analysis codes:.Development and assessment of these codes,require separate-effects experiments toprovide data for developing specializedmodels as new operating regimes andphenomena are encountered. Further, asnew reactor concepts are proposed, it isnecessary to obtain data for assessing theapplicability and accuracy of agency codesfor the different nuclear steam supplysystems. The NRC has satisfied these needsthrough the agency staff and contractors, andby developing and maintaining specializedthermal-hydraulic test facilities at universitiesand national laboratories. In addition, theNRC has joined international efforts on thelimited basis of exchanging data whenopportunities have arisen.

The burden to maintain, modernize, and'support several thermal-hydraulic transientanalysis codes was significant. A prudentdecision was made to consolidate the agencyanalysis capabilities into a single code nowcalled TRACE. Consolidation was envisagedas an ambitious effort to provide capabilitiesin one code that could reproduce results'ofthe other NRC codes, some of which hadunique features. There was also the desire tomodernize the consolidated code to someextent and certainly to rationalize the manycorrelations and approximations devised over

The PUMA Thermal-Hydraulics Test- Facility at Purdue Universiy

This' Experimental -Facility provides -data forthe development and Validation of thermal-hydrau'lic models of passive safety systems inadvanced light 'water reactor. designs.

i. - ;. ; ; -;

57 NUREG-1 635

the years to obtain computational resultsconsistent with data, especially in thecomplicated area of two-phase flow.Ultimately, the consolidated code can beused for independent evaluation of licensee'sthermal- hydraulic analyses as well as for usein analysis of safety issues. To avoid the lossof legacy plant models that have beendeveloped for the existing codes, RELAP5,TRAC-P, and TRAC-B, a Symbolic NuclearAnalysis Package (SNAP) is being developedwhich will be able to convert existing models,with little or no modification, to TRACEmodels. This graphical user interface is alsointended to enhance productivity by providinga user-convenient environment for futureinput model development, graphical display ofresults, and input checking.

The consolidation of NRC's thermal-hydraulicmodeling into a single code was a decisionthat was enthusiastically supported by theACRS. The ACRS understood this to be achallenging undertaking. Indeed, theconsolidation has proved to be every bit aschallenging as anticipated. Whileconsolidation is nearing completion, many"loose ends" remain to be resolved throughthe assessment and improvement phases ofthe project. The assessment and refinementefforts constitute an important fraction of theThermal-Hydraulics Research Projects listedin Table 14.

The TRACE code has been developed to thepoint where it has provided useful validationcalculations for the ESBWR and AP1000designs, as well as for the steam generatorhydrodynamic blowdown loads related toGeneric Safety Issue (GSI) 188, "SteamGeneratorTube Leaks/ Ruptures Concurrentwith Containment Bypass." However, it is notyet ready for routine regulatory application.Therefore, the NRC will need to maintain theRELAP5 code (Y6392). RELAP5 is beingused to carry out intensive thermal-hydraulicanalyses in support of the agency's initiativein reevaluating the PTS screening criterion

and in supporting the certification of theAP1000 advanced LWR design. Themaintenance of this older code should includecorrection of deficiencies in RELAP5 whennecessary for satisfactory completion of theseanalysis efforts.

In addition to the work described in the list ofsponsored projects, there have beensignificant contributions from the RES staffitself in the development and improvement ofthe TRACE code. While the system codesprovide important thermal-hydraulic analysiscapability, it is encouraging to see thatcomputational fluid dynamic (CFD) methodsare being used by the RES staff for resolvingcomplex multidimensional thermal-hydraulicissues. Ultimately, the coupling of CFDmethods with the system codes will add anew dimension of confidence to thermal-hydraulic analyses.

The development of new TRACE models andthe refinement of these models to include theoperating regimes of the advanced LWRshave created the need for new empiricaldata. To this end, the NRC is conductingseparate-effects experimental research atuniversities. Electrically heated rod bundlereflood experiments are being conducted atPenn State University (Y6671). These testsuse advanced instrumentation to permitgreater insights into the reflood heat transferprocess so that peak clad temperatures canbe predicted with less uncertainty. Subcooledboiling experiments and associated modeldevelopment are being conducted at UCLA(W6749) to develop more accurate modelsand reduce uncertainty associated with voidgeneration expected in the ESBWR.Entrainment at "Tees" and in the upperplenum of the AP1000 is being investigatedat Oregon State University (Y6507, Y6795) toreduce uncertainty in prediction of theAP1000 accident response. Experiments onlow pressure choked two-phase flow typical ofnatural circulation operation and experimentaland modeling research on interfacial area

NUREG-1 635 58

transport for improving the accuracy of two-phase models are being conducted at PurdueUniversity (W6698, Y6769). The motivationsfor these efforts are the improvement ofbreak flow models at low pressure andimprovement of two-phase flow modelingaccuracy by adding transport equations forthe interfacial area. This is a generic effort toimprove the accuracy of two-phase flowtransient models.

To provide data needed for assessing theapplicability and accuracy of TRACEsimulations for the ESBWR and AP1000designs, experiments are being conducted inintegral system test facilities. The PUMA andAPEX facilities are scaled models of theESBWR and AP1000 designs. The NRC alsoparticipates in international programs such asthe CAMP Program for exchange of codeassessment and user experience and theSETH Program in Germany and the PANDAProgram in Switzerland for exchange ofexperimental data. The SETH Program willcontribute experience and data on boronmixing. The PANDA Program will providenatural circulation data for full-height facilitiesfor code assessment and verification of thescaling methods used in the design of thePUMA facility.

The code consolidation effort has not gone assmoothly as was predicted at the time thework was initiated. The effort is mostchallenging and given the substantialadditional workload for thermal-hydraulicresearch that has distracted from the effort,delay is not a surprise. We note that the codeconsolidation project involves many separateorganizations and is funded from a variety oftask orders having different NRC projectmanagers. One NRC project has about 17separate task orders that address a variety ofthermal-hydraulic issues at separateorganizations. Surely, the coordination ofthese efforts must be challenging.

Another observation is that no research isdirected at development of probabilisticmethods for use with TRACE for establishingmargin of safety at a prescribed confidencelevel. There is a growing need foruncertainty estimates in thermal-hydraulicanalyses used in the regulatory process.

The time is approaching when the TRACEeffort must reach fruition and a consolidatedcode be integrated into the regulatoryprocess. There may be ways to improve theeffectiveness of this effort, such as:

* Creating a TRACE code peer-reviewgroup to provide greater focus ontechnical issues confronting thedevelopment and help achieveconsensus on the approach.

* Combining the code consolidationefforts and thermal-hydraulic modeldevelopment experiments under oneNRC project manager to achievemore effective integration.

* Benefitting from an internationaleffort leveraging NRC resources withsimilar undertakings within theCAMP community.

The NRC bears the substantial burden ofdeveloping and maintaining experimentalfacilities for thermal-hydraulic research. Theburden has been lessened by the use of well-scaled facilities that are less than full heightsuch as the PUMA and APEX facilities. Datafrom these facilities together with results ofthermal-hydraulic simulations have proventhe correspondence between data from thescaled facilities and the results that would beexpected from the full-height systems ofAP1000 and ESBWR. This conclusion isstrengthened by the data from full-heightinternational facilities such as PANDA andROSA-IV. The ACRS believes that thePUMA and APEX facilities are of proven,current utility and will help in the certification

59 NUREG-1 635

of the ESBWRand AP1000 advanced reactordesigns.

In the longer term, the NRC may want toexpand international cooperative researchagreements for maintaining thermal-hydraulicresearch facilities. The successes being hadin such arrangements in fuel behaviorresearch and severe accident research mayserve as examples of the advantages of suchan international approach.

NUREG-1635 60

Table 14. Thermal-Hydraulics Research Activities


W6698 PUMA Integral Test Facility Experimental facility to be usedfor ESBWR design certification.Provides data on LOCA natural

Y6769 PMcirculation instabilities and low-PUMA Test Facility pressure choked flow data forassessment of the TRACE code

Y6673 TRAC-M Development Assessment - Small Consolidated NRC thermal-LOCA Processes hydraulics code.

Y6300 User Support for Consolidated TRAC Code

Y6525 TRAC-M Code Maintenance

Y6583 Advanced Reactor Development of TRAC-MCode - Advanced Reactor Portion of Y6300

Y6830 TRAC-M Development and Assessment -ROSA-IV Tests

Y6666 Advanced Numerical Methods in TRAC-M This work to add a droplet fieldand advanced numericalmethods to TRACE should beaccelerated. The work needsgreater focus and could benefitfrom peer review

Y6507 APEX Experimental Validation Conduct tests for AP1000 reviewand do confirmatory analysis

Y6795 APEX AP1000 Test Program using TRACE. Conduct beyond-design-basis test for AP1 000

Y6667 SNAP Implementation Maintain and modify thegraphical user interface forTRACE to provide ability toimport TRAC-P and RELAP5plant models for TRACE.

61 NUREG-1 635

Table 14. Thermal-Hydraulics Research Activities(Continued)


W6749 Thermal-Hydraulics Research This task is very diverse andmight benefit from prioritization

Task orders to support TRACE consolidation: and additional technical direction* Interfacial area transport perhaps from a peer-review* Modularization of TRACE group.* Development assessment numerical techniques* Phase separation in 'Tees" tests within this* Subcooled boiling at low pressure Project should be closely* SNAP runtime and output visualization coordinated with the TRACE

development code development.* SNAP nodalization knowledge-based expert

system* Improvements to PARCS* BWR model development and assessment* OECD/NRC benchmark for a BWR turbine trip

transient* Multiphase CFD enhancements to nuclear reactor

safety analysis* MELCOR assessment against SCDAP/RELAP5* Automated code assessment program* Two-phase CFD enhancements for NPHASE code* Peach Bottom turbine trip analysis

K6987 Analytical Support for Serena Program Program without any specificobjectives; should be closed ifnot yielding products to crucialthermal-hydraulic activities.

Y6392 Maintain, Apply, Assess, and Develop NRC Consolidate RELAP5 capabilitiesComputer Codes into TRACE. Maintain RELAP5.

Analysis of advanced reactordesigns. TRACE assessments ofPWR steam and feedwater linebreaks. This is a necessaryactivity until TRACE can be usedto do the analyses.

Y6428 Evaluation of Steam Generator Tube Rupture This is a major effort usingPerformance SCDAP/RELAP to analyze

steam generator tube ruptureaccidents.

Y6503 AP1000 analysis TRACE analyses of AP1000large and small break loss of

Y6662 AP1000 Confirmatory Thermal-Hydraulics coolant accidents.Analysis

NUREG-1 635 62

Table 14. Thermal-Hydraulics Research Activities(Continued)

Y6772 AP1000 Confirmatory Research Measure entrainment forAP1 000 certificationNeeds better coordination withTRACE development andassessment.

Y6526 Administer CAMP Meeting Support for users of RELAP5;This is a valuable internationalcollaboration.

Y6571 SETH Program - Test Facilities Boron mixing programs in thePKL facility in Germany and thePANDA facility in Switzerland.

Y6589 Thermal-Hydraulics Research

Five task orders to support various thermal-hydraulics Applicable parts of this workresearch: need to be coordinated and* two-phase CFD enhancement combined with the TRACE code* MOX neutronics development. The longer-term* Interfacial area transport tasks need peer review to* separate effects experiments for model establish relevance.

development* MELCOR assessment and application

Y6598 RELAP5 Analyses for Pressurized Thermal Complete PTS analyses for fourShock plants.

Y6671 Rod Bundle Reflood Experiments - Phase 2 Rod bundle heat transfer tests insupport of TRACE reflood modeldevelopment.

Needs to be more closelycoordinated with the TRACEcode development.

Y6804 ESBWR Containment Support Thermal-hydraulics support forESBWR design certification.

Y6806 Review of GE Scaling Report

63 NUREG-1 635

16 REFERENCES

1. U.S. Nuclear Regulatory Commission,"Review and Evaluation of the NuclearRegulatory Commission SafetyResearch Program; A Report to the U.S.Nuclear Regulatory Commission,"Advisory Committee On ReactorSafeguards (ACRS), NUREG-1635, Vol.5. June 2003.

2. U.S. Nuclear Regulatory Commission,"Review and Evaluation of the NuclearRegulatory Commission SafetyResearch Program, A Report to the U.S.Nuclear Regulatory Commission,"Advisory Committee On ReactorSafeguards (ACRS), NUREG-1 635, Vol.4, May 2001.

3. Report dated May 16, 2003, from MarioV. Bonaca, Chairman, ACRS, to Nils J.Diaz, Chairman, U. S. NuclearRegulatory Commission, Subject:Improvement of the Quality of RiskInformation for RegulatoryDecisionmaking

4. U.S. Nuclear Regulatory Commission,"Evaluation of Air-Operated Valves atU.S. Light-Water Reactors," NUREG-1275, Vol. 13, February 2000.

5. U.S. Nuclear Regulatory Commission,"Operating Experience FeedbackReport, Assessment of Spent FuelCooling," NUREG-1275, Vol. 12,February 1997.

6. U.S. Nuclear Regulatory Commission,"Operating Experience Assessment -Effects of Grid Events on Nuclear PowerPlant Performance," NUREG-1784,December 2003.

7. Report dated September 30, 2003, fromMario V. Bonaca, Chairman, ACRS, toNils J. Diaz, Chairman, U. S. NuclearRegulatory Commission, Subject: DraftFinal Revision 3 to Regulatory Guide1.82, "Water Sources for' Long-TermRecirculation Cooling Following a Loss-of-Coolant Accident."

8. U.S. Nuclear Regulatory Commission,"Perspectives Gained from the IndividualPlant Examination of External Events(IPEEE) Program," NUREG-1742,Volumes 1 and 2, April 2001.

9. LaChance, J.L., et.al., "Circuit Analysis -Failure Mode and Likelihood Analysis,"Sandia National Laboratory,NUREG/CR-6834, September 2003.

10. U.S. Nuclear Regulatory Commission,"Severe Accident Risks: An Assessmentfor Five U.S. Nuclear Power Plants,"NUREG-1150,1990.

11. National Fire Protection Association(NFPA), "Performance-Based Standardfor Fire Protection for Light-WaterReactor Electric Generating Plants,"NFPA Standard 805, January 2001.

12. Meyer, R.O., R.K. McCardell, H.M.Chung, D.J. Diamond, and H.H. Scott,"A Regulatory Assessment of Test Datafor Reactivity-Initiated Accidents,"Nuclear Safety, Special Issue onReactivity- Initiated Accident, Vol. 37,No. 4, pp. 271-288, October-December1996.

13. Diaz, N.J., "To License and Regulate -Sharping the Edges," Presented at theINPO 4t Annual CEO Conference,Atlanta, Georgia, November 6-7, 2003.

65 NUREG-1 635

14. Poucet, A., "The European BenchmarkExercise on Human Reliability Analysis," Presented at the American NuclearSociety, International Topical Meetingon Probability, Reliability, and SafetyAssessment, PSA 89, Pittsburgh, PA,April 2-7, 1989.

15. Committee on the Biological Effects ofIonizing Radiation (BEIR), NationalResearch Council, "Health Effects ofExposure to Low Levels of IonizingRadiation: BEIR V," National AcademyPress, Washington, D.C. 1990.

16. International Commission onRadiological Protection (ICRP),"Recommendations of the InternationalCommission on Radiological Protection,ICRP Publication 60," Ann. ICRP, 1991.

17. U.S. Nuclear Regulatory Commission,"Radiological Assessment for Clearanceof Materials from Nuclear Facilities,"NUREG-1 640, Vols. 1-4, June 2003.

18. Senior Seismic Hazard AnalysisC o m m i t t e e ( S S H A C ) ,"Recommendations for ProbabilisticSeismic Hazard Analysis: Guidance onUncertainty and Use of Experts,"Lawrence Livermore NationalLaboratory, NUREG/CR-6372, April1997.

19. Code of Federal Regulations, Titlelo,Part 72 (10 CFR 72), "LicensingRequirements for the IndependentStorage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste,"U.S. Government Printing Office,Washington, D.C., 2003.

20. Code of Federal Regulations, Title 10,Part 100 (10 CFR 100), "Reactor SiteCriteria," U.S. Government PrintingOffice, Washington, D.C., 2003.

NUREG-1635 66

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER(289) (Assigned by NRC, Add Vol. Supp, Rev.,NRCM 1102. BIBLIOGRAPHIC DATA SHEET and Addendum Numbers, If any.)3201.320ILIGAHI2AA HE

(See Instructions on the reverse)

2. TITLE AND SUTrrLE NUREG-1635, Vol. 6

Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program - AReport to the U.S. Nuclear Regulatory Commission 3. DATE REPORT PUBLISHED

MONTH YEAR

March 20044. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Technical Report7. PERIOD COVERED (Incdusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRG-parodeD sg tOfliceorRegon. U.S. NudearRegdatoryComnvnssisn. andmaing address;fcontractor.pnonide name and mnalkn address.)

Advisory Committee on Reactor SafeguardsU. S. Nuclear Regulatory CommissionWashington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (t NRC type Same as above contraor, provide NRC Dysion. Oft rReof ,n U.S. Nuear Regutatory Cornmitssn.and making address.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 ivtds or less)

This report presents the observations and recommendations of the Advisory Committee on Reactor Safeguards (ACRS)concerning the NRC Safety Research Program being carried out by the Office of Nuclear Regulatory Research (RES). Thisreport focuses on that portion of the NRC research program dealing with the safety of existing nuclear reactors and advancedlight water reactor designs, APIOOO, and ESBWR, submitted for certification. In Its review of the NRC research activities, theACRS considered the programmatic justification for the research as well as the technical approach and progress of the work.This review attempts to identify research crucial to the NRC mission. It also attempts to Identify research activities that havemade valuable contributions to the agency mission In the past, but now have reached the point where additional research Isnot needed for efficient and effective safety regulation. The review also attempts to Identify areas where greater Internationalcooperation in research useful to the NRC could leverage resources of partners In the research and yield superior technicalproducts. This report does not address research on the vulnerability of existing nuclear power plants to acts of sabotage andterrorism.

12. KEY WORDSIDESCRIPToRS (List wordsorphrases thatfwassstresearers hb cg the report.) 13. AVAILABILITY STATEMENT

observations and recommendations unlimitedoperational data radiation protection4. sEcuRf CLSSIFICATIONcontainment systems safety research (MIS Pfte)digital Instrumentation and control systems seismic research unclassifiedfire safety research severe accident research (This Report)

reactor fuel research thermal-hydraulics research unclassifiedneutronics and criticality safetyhuman factors and human reliability research 15. NUMBER OF PAGES

materials and metallurgyprobabilistic risk assessment 16. PRICE

NRC FORM 335 (2-89) This form was electrnically produced by Efte Federal Fo~rms Inc.

Federal Recycling Program