scaling, uncertainty, and 3d coupled code calculations in...

Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2008, Article ID 673587, 3 pagesdoi:10.1155/2008/673587

EditorialScaling, Uncertainty, and 3D Coupled Code Calculations inNuclear Technology

Cesare Frepoli1 and Alessandro Petruzzi2

1 LOCA Integrated Services I, Westinghouse Electric Company, Pittsburgh, PA 15230-0350, USA2 Nuclear Research Group San Piero a Grado (NRGSPG), Department of Mechanical, Nuclear and Production Engineering,University of Pisa, Pisa, 56100, Italy

Correspondence should be addressed to Alessandro Petruzzi, [email protected]

Received 22 May 2008; Accepted 22 May 2008

Copyright © 2008 C. Frepoli and A. Petruzzi. This is an open access article distributed under the Creative Commons AttributionLicense, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properlycited.

System codes such as RELAP, TRACE, CATHARE, orATHLET are currently used by designer/vendors of NPPs, byutilities, licensing authorities, research organizations includ-ing universities, nuclear fuel companies, and by technicalsupporting organizations. The objectives of using the codesmay be quite different, ranging from design or safetyassessment to simply understanding the transient behavior ofa simple system. However, the application of a selected codemust be proven to be adequate to the performed analysis.Thus considerable research efforts have been spent in the lastthree decades, and as a consequence a wide range of activitieshas recently been completed in the area of system thermal-hydraulics. Problems have been addressed, solutions towhich have been at least partly agreed upon on internationalground. These include the need for best-estimate systemcodes, the general code qualification process, the proposalfor nodalization qualification as well as attempts aiming atqualitative and quantitative accuracy evaluations. Moreover,complex uncertainty methods have been proposed, followinga pioneering study which attempted, among other things, toaccount for user effects on code results.

Based on the above considerations, this special issuemostly focuses on the development and application of best-estimate codes emphasizing the role of the scaling, bestestimate and uncertainty, and 3D coupled code calculationsanalyses.

In general terms, scaling indicates the need for theprocess of transferring information from a model to aprototype. In system thermal hydraulics, a scaling process,based upon suitable physical principles, aims at establishinga correlation between phenomena expected in an NPP-transient scenario as (a) phenomena measured in smallerscale facilities, or (b) phenomena predicted by numerical

tools qualified against experiments performed in small-scalefacilities. In connection with this point, owing to limitationsof the equations at the basis of system codes, the scaling issuemay constitute an important source of uncertainties in codeapplications.

By definition, a best-estimate analysis (the term “best-estimate” is usually used as a substitute for “realistic”) isan accident analysis which is free of deliberate pessimismregarding selected acceptance criteria, and is characterized byapplying best-estimate codes along with nominal plant dataand best-estimate initial and boundary conditions. However,notwithstanding the important achievements and progressmade in recent years, the predictions of the best-estimatesystem codes are not exact but remain uncertain because(a) the assessment process depends upon data almost alwaysmeasured in small-scale facilities and not in the full-powerreactors; (b) the models and the solution methods in thecodes are approximate. In some cases, fundamental lawsof physics are not considered. Consequently, the results ofthe code calculations may not be applicable to give exactinformation on the behavior of a nuclear power plant(NPP) during postulated accident scenarios. Therefore, best-estimate predictions of NPP scenarios must be supplementedby proper uncertainty evaluations in order to be meaningful.The term “best-estimate plus uncertainty” was coined forindicating an accident analysis which (1) is free of deliberatepessimism regarding selected acceptance criteria, (2) uses aBE code, and (3) includes uncertainty analysis. Thus, theword “uncertainty” and the need for uncertainty evaluationare strictly connected with the use of BE codes.

Nowadays, advanced 3D coupled neutron-kinetics/thermal-hydraulics computer tools along with powerfulcomputers can perform realistic best-estimate analyses of

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2 Science and Technology of Nuclear Installations

complex power plant transients. The interaction betweenthermal-hydraulics and neutron kinetics is relevant forboth the safety and the design of existing nuclear reactors.The results from the application of coupled computationaltools provide new insights into the conservatisms for thespecification of relevant operational safety margins andcan imply new optimizations of emergency operatingprocedures in existing plants. They also improve knowledgeof the physical phenomena in nuclear water reactortechnology and can specifically shed light on the interactionbetween thermal-hydraulics and neutron kinetics that stillcan challenge the design and the operation of nuclear powerplants.

This special issue collects selected lectures delivered at the3D S.UN.COP (Scaling, Uncertainty, and 3D COuPled codecalculations) seminars-trainings whose aim is to transfercompetence, knowledge, and experience from about 30recognized international experts coming from more than10 different countries and institutions to analysts with asuitable background in nuclear technology. The program ofthe 3D S.UN.COP offers each year about 60 presentationsand 100 hours of parallel code hands-on training subdividedin three weeks and covering the following topics: (a) systemcodes: evaluation, application, modeling and scaling; (b)international standard problems; (c) best-estimate in systemcode applications and uncertainty evaluation; (d) qualifi-cation procedures; (e) methods for sensitivity and uncer-tainty analysis; (f) relevant topics in best-estimate licensingapproach; (g) industrial applications of the best-estimate-plus-uncertainty methodology; (h) coupling methodologiesand applications; (i) computational fluid dynamics codes.From the other side, the parallel hands-on training ses-sions on numerical codes (such as CATHARE, CATHENA,RELAP5, TRACE, and PARCS) allow the participants toachieve the capability to set up, run, and evaluate the resultsof a numerical tool through the application of the proposedqualitative and quantitative accuracy evaluation procedures.Finally, the 3D S.UN.COP seminars provides a forum forexchanges of ideas through scientific presentations anddialogue among representatives of the worlds of academy,research laboratories, industry, regulatory authorities, andinternational institutions.

In the first paper, A. Petruzzi et al. emphasized the roleof the computer code user that represents one of the mainsources of uncertainty influencing the results of system codecalculations. This influence is commonly known as the “usereffect” and stems from the limitations embedded in the codesas well as from the limited capability of the analysts to use thecodes. The paper describes a systematic approach to trainingcode users who, upon completion of the training, should beable to perform calculations making the best possible use ofthe capabilities of best-estimate codes.

In the second paper, A. Petruzzi, and F. D’Auriapresented the commonly used system thermal-hydrauliccodes such as RELAP, TRACE, CATHARE, or ATHLETfor reactor-transient simulations. Whereas the first systemcodes, developed at the beginning of the 1970s, utilizedthe homogenous equilibrium model with three balanceequations to describe the two-phase flow, nowadays the more

advanced system codes are based on the so-called “two-fluidmodel” with separation of the water and vapour phases,resulting in systems with at least six balance equations.However, notwithstanding the huge amounts of financial andhuman resources invested, the results predicted by the codeare still affected by errors whose origins can be attributedto several reasons as model deficiencies, approximations inthe numerical solution, nodalization effects, and imperfectknowledge of boundary and initial conditions. In thiscontext, the existence of qualified procedures for a consistentapplication of qualified thermal-hydraulic system code isnecessary and implies the drawing up of specific criteriathrough which the code-user, the nodalization, and finallythe transient results are qualified.

In “International Standard Problems and Small BreakLoss-Of-Coolant Accident (SBLOCA),” N. Aksan consideredfive small break LOCA-related ISPs since these were usedfor the assessment of the advanced best-estimate codes. Theconsidered ISPs deal with the phenomenon typical of smallbreak LOCAs in Western design PWRs. The experimentsin four integral test facilities, LOBI, SPES, BETHSY, ROSAIV/LSTF, and in the recorded data during a steam generatortube rupture transient in the DOEL-2 PWR (Belgium) werethe basis of ISP calculations. The statistical evaluation of thegeneral data obtained from these ISPs is summarized. Somelessons learned from these small break LOCA ISPs are identi-fied in relation to code deficiencies and capabilities, progressin the code capabilities, possibility of scaling, and variousadditional aspects. ISPs are providing unique material andbenefits for some safety related issues. Some of the technicalfindings and benefits provided by small break LOCA ISPs areprovided as conclusions and recommendations.

In the next paper, A. Petruzzi, and F. D’Auria pre-sented the evaluation of uncertainty methodologies asnecessary supplement of best-estimate calculations per-formed to understand accident scenarios in water-coolednuclear reactors. The needs come from the imperfectionof computational tools, on the one side, and the interestin using such a tool to get more precise evaluation ofsafety margins. The paper reviews the salient features of twoindependent approaches for estimating uncertainties asso-ciated with predictions of complex system codes. Namely,the propagations of code input error and calculation outputerror constitute the keywords for identifying the methodsof current interest for industrial applications. Throughoutthe developed methods, uncertainty bands can be derived(both upper and lower) for any desired quantity of thetransient of interest. For one case, the uncertainty methodis coupled with the thermal-hydraulic code to get the codewith capability of internal assessment of uncertainty, whosefeatures are discussed in more detail.

The task of regulatory body staff reviewing and assessinga realistic large break loss-of-coolant accident evaluationmodel is discussed by R. Galetti in the next paper facingthe actual regulatory licensing environment related to theacceptance of the analysis of emergency core cooling systemperformance. Especially, focus is directed to the question ofhow to fulfill the requirement of quantifying the uncertaintyin the calculated results when they are compared to the

C. Frepoli and A. Petruzzi 3

acceptance criteria for this system. When using a realisticevaluation model to analyze the loss-of-coolant accident,different approaches have been used in the licensing arena.The Brazilian regulatory body has concluded that, in thecurrent environment, the independent regulatory calculationis recognized as a relevant support for the staff decisionwithin the licensing framework of a realistic analysis.

In the sixth paper, H. Glaeser summarized the basictechniques of the GRS uncertainty method together withapplications to a large break loss-of-coolant accident ona reference reactor as well as on an experiment simulat-ing containment behavior. A significant advantage of thismethodology is that no a priori reduction in the numberof uncertain input parameters by expert judgement orscreening calculations is necessary to limit the calculationeffort. All potentially important parameters may be includedand the number of calculations needed is independent ofthe number of uncertain parameters accounted for in theanalysis. A challenge in performing uncertainty analyses withthe GRS methodology is the specification of ranges andprobability distributions of input parameters.

C. Frepoli presented the paper entitled “An Overviewof Westinghouse Realistic Large Break LOCA EvaluationModel.” Since the 1988 amendment of the 10 CFR 50.46rule in 1988, Westinghouse has been developing and apply-ing realistic or best-estimate methods to perform LOCAsafety analyses. Westinghouse methodology is based onthe use of the WCOBRA/TRAC thermal-hydraulic code.The paper starts with an overview of the regulations andits interpretation in the context of realistic analysis. TheCSAU (code scaling, applicability, and uncertainty) roadmapis reviewed in the context of its implementation in theWestinghouse evaluation model. An overview of the code(WCOBRA/TRAC) and methodology is provided. Finally,the recent evolution to nonparametric statistics in the cur-rent edition of the Westinghouse methodology is discussed.Sample results of a typical large break LOCA analysis for aPWR are provided.

The next paper by R. Martin and L. O’Dell illustratesthe development considerations of AREVA NP Inc.’s realisticLBLOCA analysis methodology. The AREVA NP RLBLOCAmethodology is a CSAU-based methodology for performingbest-estimate large-break LOCA analysis. The methodologyaddresses all of the expressed steps of the CSAU process.The key challenge to this process has been the defense ofdeclared engineering judgment and the demonstration of themethodologies’ range of applicability. This was accomplishedby careful characterization of dominant LOCA parametersand emphasis on validation through sensitivity studies andthe statistical nature of the methodology. The generic AREVANP RLBLOCA methodology was approved by the USNRC inApril 2003 and is now being applied to several nuclear powerplants serviced by AREVA NP Inc.

In the next paper, D. Novog and P. Sermer provided anovel and robust methodology for determination of nuclearreactor trip set points, which accounts for uncertainties ininput parameters and models, and for the variations inoperating states that periodically occur. The paper presentsthe general concept used to determine the actuation set

points considering the uncertainties and changes in initialconditions, and allowing for safety system instrumenta-tion redundancy. The results demonstrate unique statisticalbehavior with respect to both fuel and instrumentationuncertainties, which has not previously been investigated.

F. Reventos et al. illustrated the usefulness of computa-tional analysis for operational support in the paper beforethe last. In the first part, he described the specific aspectsof thermal-hydraulic analysis tasks related to operation andcontrol and, in the second part, they briefly presentedthe results of three examples of performed analyses. Allthe presented examples are related to actual situations inwhich the scenarios were studied by analysts using thermal-hydraulic codes and prepared nodalizations. The paper alsoincludes a qualitative evaluation of the benefits obtainedthrough thermal-hydraulic analyses aiming at supportingoperation and plant control.

In the last paper, H. Ikeda et al. reviewed activitiesrelevant to the boiling water reactor (BWR) stability phe-nomenon, which has a coupled neutronic and thermal-hydraulic nature, from the viewpoint of model and codedevelopments. Industrial organizations have developed andimproved the BWR stability analysis using computationaltools specific for the reduced-order frequency-domain andthree-dimensional time-domain codes. The first category iscurrently applied to the BWR stability design analysis, whilethe latter has been exploited to understand the complicatedphenomena related to BWR stability. Proposals to apply best-estimate analysis code with the statistical safety evaluationmethodology are currently under study. This will allow betterevaluation of the stability exclusion region, and will beconsequently applied to the BWR plants with the extendedcore power uprate.

We believe that the collection of papers in this specialissue illustrates the great variety of topics and problems inthe nuclear technology for which advanced tools are availableand applicable.

Finally, we would like to take the opportunity to expressour thanks to all authors who have submitted papers to thisspecial issue and to our colleagues who devoted their valuabletime reviewing these manuscripts.

Cesare FrepoliAlessandro Petruzzi

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