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Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2010, Article ID 425658, 17 pagesdoi:10.1155/2010/425658

Project Report

Role of RELAP/SCDAPSIM in Nuclear Safety

C. M. Allison and J. K. Hohorst

Innovative Systems Software, LLC, 1284 South Woodruff Avenue, Idaho Falls, ID 83404, USA

Correspondence should be addressed to C. M. Allison, [email protected]

Received 17 August 2009; Accepted 10 January 2010

Academic Editor: Michel Giot

Copyright © 2010 C. M. Allison and J. K. Hohorst. This is an open access article distributed under the Creative CommonsAttribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work isproperly cited.

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, isbeing developed as part of the international SCDAP Development and Training Program (SDTP). SDTP consists of nearly 60organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry.The program members and licensed software users include universities, research organizations, regulatory organizations, vendors,and utilities located in Europe, Asia, Latin America, and the United States. Innovative Systems Software (ISS) is the administratorfor the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paperprovides a brief review of some of the more important activities including the analysis of research reactors and Nuclear PowerPlants, design and analysis of experiments, and training.

1. Introduction

The RELAP/SCDAPSIM code, designed to predict thebehaviour of reactor systems during normal and accidentconditions, is being developed as part of the internationalSCDAP Development and Training Program (SDTP) [1, 2].Three main versions of RELAP/SCDAPSIM, as describedin Section 2, are currently used by program membersand licensed users to support a variety of activities.RELAP/SCDAPSIM/MOD3.2, and MOD3.4 are productionversions of the code and are used by licensed users andprogram members for critical applications such as researchreactor and nuclear power plant applications. The mostadvanced production version, MOD3.4, is also used forgeneral user training and for the design and analysis of severeaccident related experiments such as those performed in thePhebus and Quench facilities. In turn, these experimentsare used to improve the detailed fuel behaviour and othersevere accident-related models in MOD3.4 and MOD4.0.MOD4.0 is currently available only to program membersand is used primarily to develop advanced modellingoptions and to support graduate research programs andtraining.

2. RELAP/SCDAPSIM

RELAP/SCDAPSIM uses the publicly available RELAP/MOD3.3 [3] and SCDAP/RELAP5/MOD3.2 [4] modelsdeveloped by the US Nuclear Regulatory Commissionin combination with proprietary (a) advanced program-ming and numerical methods, (b) user options, and(c) models developed by ISS and other members ofthe SDTP. These enhancements allow the code to runfaster and more reliably than the original US NRCcodes. RELAP/SCDAPSIM/MOD3.2 was the first produc-tion version released under SDTP sponsorship. It wasdesigned to duplicate the modeling options available fromthe original US NRC versions of RELAP/MOD3.3 andSCDAP/RELAP5/MOD3.2 but run faster and more reliably.This version used standard RELAP5 and SCDAP/RELAP5input but included enhanced output options such asintegrated 3D and time history plotting options. Thisversion also included a number of coding and numer-ical improvements to improve the performance of thecode. RELAP/SCDAPSIM/MOD3.4 is the current produc-tion version and is designed specifically for “faster-than-real-time” simulations on typical Windows or LINUX PCs.

2 Science and Technology of Nuclear Installations

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Figure 1: SAFARI reactor core configuration.

The designation MOD3.4 is used to indicate that additionalmodeling options have been included relative to the originalUS NRC codes. These modeling options include improvedmodels for a detailed fuel rod, an electrically heated fuelrod simulator, and other SCDAP core components. Othermodeling improvements include new models and correla-tions for air ingression transients and alternative fluids andcladding materials. Continued improvements in the codingand numerics also allow both MOD3.4 and MOD4.0 to run awider variety of transients including low-pressure transientswith the presence of noncondensable gases such as thoseoccurring during mid-loop operations in LWRs, in pool typereactors, or in spent fuel storage facilities.

The most advanced version of the code, RELAP/SCDAPSIM/MOD4.0 [5], is the first version of RELAPor SCDAP/RELAP5 completely rewritten to FORTRAN90/95/2000 standards. This is a significant benefit forthe program members that are using the code for thedevelopment of advanced models and user options such asthe coupling of the code to other analysis packages. Cou-pled 3D reactor kinetics and coupled RELAP/SCDAPSIM-SAMPSON [6] calculations are examples where MOD4.0 isused because of a significant reduction in the code devel-opment effort and expense to link the packages. MOD4.0also includes advanced numerical options such as improvedtime advancement algorithms, improved water propertytables, and improved model coding. As a result the codecan reliably run complex multidimensional problems fasterthan real time on inexpensive personal computers. Plant

simulation and integrated uncertainty analysis are among themost important applications benefiting from the improvedspeed and reliability of MOD4.0. MOD4.0 includes manyenhanced user options to improve the accuracy of thecode or to offer new options for the users. For example,the addition of an alternative material property librarydesigned for Zr-Nb cladding materials is important forVVER and CANDU reactor designs, particularly for severeaccident-related transients. The addition of an advancedwater property formulation is important for many transients,in particular those involving super critical water applica-tions.

3. Review of Representative Applications

RELAP/SCDAPSIM is being used for a variety of appli-cations. As described in Section 3.1, the code is usedfor the analysis of research reactors and nuclear powerplants. The research reactors analysed by the code includeTRIGAs, MTR-plate designs, and other unique designswell as. Nuclear plants analysed include Western designedPWRs and BWRs, Russian designed VVERs and RMBKs,Canadian and Indian designed CANDUs and HWRs, andChinese designed PWRs. The analysis of experiments, asdiscussed in Section 3.2, is also an important applicationof the code, including the design of the experiments, theassessment and development of modelling improvements,and finally for advanced user training. The application of thecode to support the development of improved models andanalytic capabilities is discussed in Section 3.3. Section 3.4presents an overview of the application of the code fortraining.

3.1. Research Reactor and NPP Applications. A combinationof RELAP/SCDAPSIM/MOD3.2 and MOD3.4 is being usedto analyze research reactors. A brief summary of the earlywork by several countries was given in [6]. The researchreactors noted in this paper include (a) the LVR-15 reactorlocated at the Nuclear Research Institute in Rez, CzechRepublic, (b) the CARR reactor being built in Beijing, Chinaby the China Institute of Atomic Energy, and (c) TRIGA reac-tors located at the Atomic Energy Research Establishment inDhaka, Bangladesh, and National Nuclear Energy Agency inBandung Indonesia. LVR-15 is a light-water moderated andcooled pool type reactor with a nominal thermal power of15 MW. The pool operates at atmospheric pressure with anaverage coolant temperature in the core of 320 K. The reactoralso has closed high-pressure/temperature loops suitable fortesting of materials under PWR and BWR conditions. Thereactor core is composed of several fuel assemblies of Al-Ualloy arranged in square concentric tubes. CARR is a tank-in-pool design, cooled and moderated by light water andreflected by heavy water. The rated power is 60 MW. Thecore consists of plate-type fuel assemblies of Al-U alloy. TheIndonesian and Bangladesh TRIGA reactors are pool typereactors with 2 MW and 3 MW thermal power, respectively.The reactor cores are composed of solid U-ZrH fuel rodsarranged in a hexagonal array and are cooled by water in

Science and Technology of Nuclear Installations 3

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Figure 2: MURR RELAP/SCDAPSIM nodalization of the MURR pressurized primary cooling system.

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Figure 3: RELAP/SCDAPSIM nodalization of the MURR bulk reactor pool and pool coolant loop.

either forced or natural circulation, depending upon theconditions.

More recently, the analysis of two additional reactor typeshave been reported in [7–9]. The first is for the SAFARI-1research reactor located in South Africa [7, 8]. The secondis the University of Missouri Research Reactor located inthe United States [9]. The SAFARI-1 research reactor is a

tank-in-pool type reactor operated at a nominal core powerof 20MW. The core is cooled and moderated by forcedcirculation of light water. The reactor core can be operatedin a variety of configurations from 24 to 32 fuel assemblies.Figure 1 shows an example of one such configuration. Thefuel is U-Si-Al plate-type fuel elements. MURR is a 10 MWpool type reactor design with a pressurized primary coolant


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loop to cool the fuel region. The pressurized primary systemis located in a pool allowing direct heat transfer duringnormal operation and transition to natural convection underaccident conditions. The reflector region, control bladeregion, and center test hole are cooled by pool water (naturalconvection).

Because of the unique reactor designs, the RELAP/SCDAPSIM input models were developed separately by eachorganization and include a range of different nodalizationsas presented in the reference papers. However in generalterms, the RELAP/SCDAPSIM input models include all ofthe major components of each reactor system including thereactor tank, the reactor core and associated structures, andthe reactor cooling system including pumps, valves, and heatexchangers. The secondary sides of the heat exchanger(s)are also modelled where appropriate. These input modelswere qualified through comparison with reactor steady statedata, with original vendor safety analysis calculations whereavailable, and with experiments in a limited number of cases.Figures 2 through 4 give examples of the nodalization usedfor MURR. Figures 2 and 3 show the detailed hydrodynamicnodalization for the pressurized primary cooling system andthe bulk pool and pool cooling system, respectively. Figure 4shows the nodalization of the fuel plates. This input model isalso somewhat unique in that all 24 fuel plates were modeledusing RELAP5 heat structures.

Figure 5 shows the nodalization used for the SAFARIresearch reactor. Figure 5 shows the overall system hydro-dynamic nodalization with the upper right corner of thefigure showing the core nodalization. Note from the insert of

the core nodalization diagram that the bypass or unheatedchannels were modelled separately from the heated fuelassembly channels. The core nodalization also included twoseparate hot plate channels located on each side of the hottestplate.

A wide variety of transients have been analyzed using thecode. Examples are included in the references and includereactivity initiated power excursions and loss of flow orcoolant transients. Figure 6 shows one such example forMURR. The figure shows the centerline temperatures for the24 fuel plates during a cold leg LOCA, indicating that thefuel transients remained well below the assumed fuel damagelimits of 900◦F.

All three versions of RELAP/SCDAPSIM have been usedto analyze a variety of nuclear power plant designs. Theapplications have included RELAP5-only input models fornormal operating or transient conditions where core damageis not expected as well as combined RELAP-SCDAP inputmodels that included the possibility of transients with theloss of core geometry. A few representation examples arediscussed in more detail in the remainder of this subsection.

Analysts at the Paul Scherrer Institut (PSI) in Switzerlandhave applied the code to the TMI-2 accident [10], an analysisof a LOCA during cooldown in the Beznau Westinghousetype 2-Loop PWR [11], and an analysis of a station blackouttransient in the Gosgen KWU three-loop 1020 MW PWR[12]. The TMI-2 calculations included comparisons withthe limited data available from the accident as well ascomparisons with the MELCOR [13] and SCDAP/RELAP5[14] codes. The Beznau analysis paper summarized the


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Figure 6: Example of RELAP/SCDAPSIM-calculated fuel platecenterline temperatures for the 24 fuel plates in MURR for a coldleg large break transient.

results of the analyses of postulated LOCAs in the Beznau(KKB) PWR, occurring during hot (HS) and intermediate

(IS) shutdown with emphasis on large break LOCAs duringhot shutdown. The large break LOCA during HS posed thegreatest challenge to the plant safety systems. The analysisof the station blackout transient in the Gosgen NuclearPlant focused on the impact of a potential failure of thedepressurization system. In particular, the analysis focusedon the timing of the heatup and failure of the RCS pipingrelative to the relocation of melt into the lower plenum andfailure of the lower head. MELCOR, RELAP/SCDAPSIM,and SCDAP/RELAP5 were also used in both Beznau andGosgen analyses.

The TMI-2 RELAP/SCDAPSIM and SCDAP/RELAP5nodalization, as shown in Figure 7, used a 2-dimensionalrepresentation of the core region with a detailed SCDAPcomponents being used to describe the behavior of thefuel rods and other core structures within each of the fiverepresentative flow channels in the core. The transitionfrom the initial intact core geometry to a damaged state isautomatically handled by the SCDAP models including theinitial failure of the control rods, liquefaction and relocationof the metallic U-O-Zr fuel rod material, formation and


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growth of a ceramic [U-Zr]-O2 molten pool, and relocationof the molten ceramic into the lower plenum. Figure 8shows one of the set of representative calculations presentedin the paper. The figure shows the variation in predictedsystem pressure for a range of modeling parameters for therelocation of the metallic U-O-Zr fuel rod material for bothRELAP-based codes. It was noted in the paper that both

RELAP-based codes correctly calculated an in-core moltenpool, of which two RELAP/SCDAPSIM cases predictedrelocation to the lower head (via the bypass, as observed),while only one MELCOR case did so. It was further notedthat the RELAP-based codes correctly calculated that lowerhead failure did not occur.

The Beznau RELAP/SCDAPSIM and SCDAP/RELAP5nodalization, as shown in Figure 9, also used a 2-dimensionalrepresentation of the core region with detailed SCDAPcomponents used to describe the behavior of the fuelrods and other core structures within each of the fiverepresentative flow channels in the core. Figure 10 showsone of the set of representative calculations presented inthe paper. The figure shows the peak cladding temperaturescalculated by RELAP/SCDAPSIM and SCDAP/RELAP5 fordifferent assumptions regarding the activation of the SafetyInjection pumps including the number of pumps and delaysin the actuation of the pumps. The paper concludes that allthree codes predict that in the limiting large break case thecore is readily quenched without damage, by the nominaloperation of the system injection system. However, it wasnoted that the more mechanistic RELAP-based calculationsdemonstrated that a larger margin existed (relative to thatpredicted by MELCOR) with recovery being possible even ifonly one pump operates after some delay.

The RELAP/SCDAPSIM and SCDAP/RELAP5 nodaliza-tion used for the Gosgen analysis also included a detailedcore nodalization as described previously. However, thecalculations also looked at the effect of hot leg naturecirculation using a split hot leg model as shown in Figure 11.The split channel model allows the hotter vapor to move


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Figure 10: Example of RELAP/SCDAPSIM and SCDAP/RELAP5Beznau-calculated results for peak cladding temperature for a rangeof safety injection assumptions.

from the vessel to the steam generators along the top ofthe hot leg and cooler vapor to return along the bottom ofthe hot leg. The influence of the split hot leg input modelrelative to a single channel hot leg (which does not allowcountercurrent flow of the vapor within the hot leg) is shownin Figure 12. As shown in the figure the split hot leg modelpredicted a more gradual heatup of the core but both singlechannel and split channel models still predict rupture of thesurge line or hot leg piping before any molten core materialrelocates into the lower head.

The Politehnica University, Institute for NuclearResearch, and National Commission for Nuclear ActivitiesControl in Romania have used RELAP/SCDAPSIM/MOD3.4for a variety of analyses of CANDU reactor designs. Thework in [13] presents the analysis of a reactor inlet headerbreak, looking at (a) the size of the break, (b) the chokedflow model employed, (c) the emergency core cooling (ECC)performance, and (d) the core nodalization. The resultswere compared with the original safety analysis results.The work in [14] presents the analysis of the influence ofthe header manifold modeling for an inlet header breakin a CANDU 6. The paper looked at a 35% inlet headerbreak which was expected to produce the highest peak fuelcladding temperatures among all postulated break sizes.The work in [15] presents the analysis of a reactor outlet


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Figure 12:Example of RELAP/SCDAPSIM and RELAP/SCDAPSIM-calculated results for peak core temperature for single and split hotleg model.

header break in a CANDU-6. The paper focused on a 100%reactor outlet header break which had the highest potentialfor fuel failure and release of radioactivity. The paper alsocompared the results to earlier calculations performed using

the CATHENA code [16]. The work in [17] presents ananalysis of the late phase of a severe accident in CANDU6 where bed of dry solid debris or a molten pool of corematerial had already formed at the bottom of the calandriavessel and was externally cooled by shield-tank water. Thestudy used the SCDAP COUPLE module and includedcomparisons with earlier results performed using the ISAAC[18] and MAAP4-CANDU [19] codes.

The general system thermal hydraulic nodalization forthe CANDU system thermal hydraulic analysis analyses [13–15] is shown in Figure 13. Portions of this nodalizationwere varied somewhat depending on the analyses beingperformed. Figure 14 shows an example of the portion ofthe nodalization that was used in [14] for the study ofthe influence of the inlet header model. Figure 15 showsan example of some of the results from the inlet headerbreak model study. In the figure, the reference curve is theresults from a single average channel circuit model usinga single manifold volume and Cases 1, 2, and 3 representthe break location in the multiple inlet manifold volumes incombination with multiple core channels. The curves shownin the figures are the maximum cladding temperatures in fuelbundles contained within the multiple flow channels.

Figure 16 shows the basic problem analyzed for the latephase of a severe accident in a CANDU-6 where a debrisbed is present in the bottom of the calandria vessel alongwith the RELAP5, SCDAP, and COUPLE nodalization usedin the analysis. The RELAP5 thermal hydraulic volumes on


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Figure 13: RELAP5 thermal hydraulic nodalization for CANDU-6.

From PMP

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OH5

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Break location

Figure 14: RELAP5 thermal hydraulic nodalization of manifold headers for CANDU-6 Inlet manifold break analysis.

the right of the figure represent the pool on the outside ofthe calandria vessel. The RELAP5 and SCDAP volumes aboveand within the COUPLE mesh provide initial and boundaryconditions for the debris bed and calandria vessel wall. (Thepaper also included a more detailed RELAP5 nodalization ofthe outer pool at the elevations associated with the debrisbed. The more detailed nodalization resulted in significantlylower pool containment pressures upon vessel failure due tothe more accurate representation of the external cooling ofthe calandria vessel.)

The papers indicated that RELAP/SCDAPSIM calcula-tions gave comparable results to the CANDU-specific codes,CATHENA for system thermal hydraulics, and ISAAC andMAAP-CANDU for severe accidents. For system thermalhydraulic analysis, RELAP/SCDAPSIM, when using similarinput models, provided similar trends as compared to theoriginal safety analysis reports or comparable CATHENAcalculations. However, the results were sensitive to the levelof detail used in the nodalization; specifically the moredetailed nodalization possible using RELAP/SCDAPSIM had


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Figure 15: Maximum fuel bundle cladding temperatures for a CANDU-6 Inlet manifold break analysis for different break locations.

All heat toshield waterby conduction,radiation andfree convection

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Figure 16: RELAP, SCDAP, and COUPLE module nodalization and conceptual sketch for CANDU-6 calandria vessel analysis.

a noticeable impact on the results for the inlet headermanifold [14], the fuel channels (simulating the effects ofhorizontally stratified flow in the channels) [13], and theouter pool (exterior to the vessel calandria) [17].

RELAP/SCDAPSIM has also been used to analyze VVERreactor designs although the calculations to date have beenproprietary and have not been published in the open

literature. Figures 17 and 18 show a nonproprietary, butrepresentative, input model for a VVER-1000. This repre-sentative VVER-1000 input model and associated detailedinput model engineering handbook was provided by RiskEngineering in Bulgaria [20] as an in-kind contributionfor use in SDTP-sponsored VVER training activities. Thenodalization of the basic components of the reactor (vessel


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and internals) is presented in Figure 17 and includes twochannels in the core, the hot and peripheral channels.This basic vessel nodalization would be replaced by amore detailed multidimensional model comparable to thatused for the PWR calculations described previously forgeneral applications where core damage transients mightbe considered. Four separate main circulation loops withtheir corresponding main coolant pumps and cold and hotcirculation pipelines are also included in the representativeinput model as shown in Figure 17.

Analysts at the Lithuanian Energy Institute (LEI) havepublished a number of papers describing their use ofRELAP/SCDAPSIM/MOD3.2 for the analysis of RBMKs.

Recent references are cited as [21–23]. LEI also providednonproprietary, but representative, input models and asso-ciated engineering handbooks for use in SDTP-sponsoredtraining activities. Figure 19 shows the nodalization diagramused for the representative RBMK RCS input model.

All three versions of RELAP/SCDAPSIM have been usedto analyze BWRs although few results have been publishedin the open literature. The work in [24] describes someof the activities related to the use of the code for plantsimulation and training for the Laguna Verde plants inMexico. Figure 20 shows a representative input model, devel-oped by the Nacional de Seguridad Nuclear y Salvaguardias(CNSNS), the Mexican regulatory authority, for the analysis


x = 1, 2, 3, 4y = 1, 2, 3, 4z = 6, 7, 8, 9

Figure 18: Representative loop input nodalization for VVER-1000.

SDV-A MSVI-III

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Figure 19: Representative RCS nodalization for RBMK.

of the Laguna Verde plants. This model is also used tosupport SDTP-sponsored BWR-specific training activities.See Section 3.4 and [24] for more information on theseactivities in Mexico.

3.2. Experimental Analysis. RELAP/SCDAPSIM/MOD3.4has been used by a number of organizations to help design

experiments, to assess thermal hydraulic and severe accidentmodels, and to support advanced user training. In recentyears, the application of the code to experimental analyseshave focused on European experimental programs includingthe German Quench experiments [25–29], French PhebusFPT experiments [30–33], and most recently RussianPARAMETER experiments [34].


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The most detailed of the calculations have been involvedin the design of new experiments. For example, as describedin detail in [26], the design of new experiments requiresthe development of complex input models to describe theunique features of each experiment and in many cases thedevelopment of specialized new models to treat features ofthe experiments not previously included in the code. In thisexample, the analysts from PSI and experimentalists fromForschungszentrum Karlsruhe (FzK) describe their use ofRELAP/SCDAPSIM/MOD3.4 in conjunction with MELCORand a special FzK-developed version of SCDAP/RELAP5 [35]to design and analyze three different experiments in thequench facility, Quench-10, Quench 11, and Quench 12.Quench 10 was a unique experiment in that it was the firstintegral test to look at the influence of air ingression onbundle heating and reflood. The design and analysis of thisexperiment required PSI to develop and incorporate specialSCDAP models to treat the oxidation of Zircaloy in air/steammixtures. The experimentalist also ran special small separateeffects experiments to help develop the correlations that werethen used in these new models. Quench 11 was a unique testfor the Quench facility in that the test started with the bundlefull of water and then the heat up transient was initiated bythe boil-down of the water. (Previous Quench experimentsused a mixture of steam and argon during the heat up of thebundle prior to reflood.) Although, in this case, it was notnecessary to modify any of the RELAP or SCDAP models, themodeling of the auxiliary heaters, added to the lower plenumof the experimental test train to provide realistic boildown

rates, proved to difficult because of relatively large heat lossesin the lower plenum region. QUENCH-12 was unique inthat it was designed to determine the influence of a VVERbundle configuration and cladding on heat-up, oxidation,and quench response. Previous Quench experiments usedPWR or BWR configurations and cladding materials. TheQuench 12 bundle was significantly modified with changesto cladding material (Zr/1%Nb instead of Zry-4), electricalheating, and geometry. Oxidation correlations for Zr/1%Nbin steam were introduced into SCDAP to support the designand analysis of this experiment. Figure 21 shows a schematicof the Quench facility along with the RELAP/SCDAPSIMnodalization diagram.

The analyses of the German Quench and French Phebusexperiments have also played a pivotal role in the assessmentof RELAP5/SCDAPSIM, the development of new improvedmodels as discussed in Section 3.3, and in advanced usertraining as discussed in Section 3.4. The works in [28, 29, 31–33] are examples of the analysis of these experiments to assessthe accuracy of the code and to identify areas where themodels could be improved. The works in [36, 37] describethe use of these experiments to support advanced usertraining.

3.3. Development of Improved Models and Analytic Capa-bilities. The development of improved models and ana-lytic capabilities is also an important part of the overallSDTP cooperative activities. In addition to the modelling


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Figure 21: Example of experiment design—quench facility layout and input nodalization.

improvements driven by large-scale experimental programsin the Phebus and Quench facilities as discussed in theprevious section, other model and code development activ-ities have been driven by the needs of SDTP members andlicensed software users. INSS (Institute of Nuclear SafetySystem), Japan, one of the original members of SDTP,developed and validated new RELAP/SCDAPSIM models totreat the heat transfer in the gap between a debris bed andthe lower plenum wall [38] and improved correlations forcondensation in the presence of noncondensable gases [39].The applications of the improved correlations are describedin [40]. IAE/NUPEC (Institute of Applied Energy/NuclearPower Engineering) Japan, a long-time member of SDTP, hasbeen working with the code to develop improved analyticcapabilities to support the Japanese nuclear industry. Themerger of RELAP/SCDAPSIM with the IMPACT/SAMPSONpackage [41] is one of the most significant projects. However,

IAE/NUPEC has also been using the code for a variety ofother tasks including the development of enhanced analyt-ical capabilities to analyse corrosive conditions in nuclearpower plants using coupled system thermal hydraulicsand CFD techniques along with corrosion modelling [42].CNSNS and ININ (Instituto Nacional de InvestigacionesNucleares) in Mexico have added integrated BWR con-tainment models to RELAP/SCDAPSIM/MOD4.0 and arenow working on the possible integration of detailed inte-grated subchannel and containment modules developed byIAE/NUPEC [43]. Nuclear plant analyser graphic packagesincluding VISA [44], developed by KAERI (Korean AtomicEnergy Research Institute), and RELSIM, developed byRMA (Risk Management Associates) have been linked toRELAP/SCDAPSIM/MOD3.4 and MOD4.0. Other activitiesby members and licensed users include the coupling of thecode with 3D reactor kinetics packages.


{Steam outlet

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Time dependent volumeTime dependent junctionSingle junctionSingle volumePipe

Figure 22: Example of university applications—natural circulation boiling water reactor applications by IIT-Guwahti.

The development of improved models and code capabili-ties for RELAP/SCDAPSIM/MOD4.0 by university membersof SDTP has also been an important factor in the improve-ment of RELAP/SCDAPSIM/MOD4.0 [5]. The rewriting ofthe code to Fortran 90/95/2000 version of the code hasmade it significantly easier for university faculty and studentsto work with. MOD4.0 also provides a well-characterizedframework for university researchers and students to explorenew modelling approaches since the tedious programmingdetails associated with use of complex fluid/material proper-ties libraries, reactor component models such as pumps andvalves, input/output, and data base management for taskssuch as dynamic data allocation are provided through a stan-dard compile library maintained by ISS. The incorporationof integrated fission product transport models by Honaiser,University of Florida, USA [29], and ongoing work to add anintegrated uncertainty analysis package by Perez, Universityof Catalunya, Spain [45, 46], and CANDU-specific modelsfor fuel channel failure by Mladin, Polectechnic University,Romania [17], are good examples where university studentsare key contributors to the development of the code.

3.4. Training of Analysts and Model/Code Developers.RELAP/SCDAPSIM/MOD3.4 and MOD4.0 are also widelyused to support SDTP-sponsored training activities.MOD3.4 is used for basic user and applications training.This includes (a) 1- to 2-week novice and advanced RELAP5and SCDAP user training workshops and seminars, (b)longer term, 1 to 3 month, user and application trainingunder IAEA and SDTP-sponsored training fellowships, and(c) IAEA-sponsored specialized missions on research reactorapplications, severe accident management, and others. Forexample, novice users will use the code to set up basic

thermal hydraulic problems such as the flow of water in apipe or the boildown and quenching of a representative fuelassembly and then move on to the optimisation or expansionof the input model to a representative full research reactoror NPP. More advanced students or participants in longerterm training sessions will typically use the code to developinput models for their own facilities or more typicallyadapt existing input models to run more reliably or runa much wider variety of possible transients. MOD4.0 andto some extent MOD3.4 are also widely used by the SDTPmember universities to support their graduate and facultyresearch programs. Section 3.3 gave some specific examplesof university students that started out participating in SDTP-sponsored training activities using MOD3.4 and MOD4.0and then going on to make significant contributions toimprovement of MOD4.0. Another good example of theuse of the codes at universities is provided by ProfessorManmohan Pandey and others from the Department ofMechanical Engineering of Indian Institute of TechnologyGuwahati (IIT-Guwahati), India, in a report submitted as anin-kind contribution for their university membership [47].

IIT-Guwahati used RELAP/SCDAPSIM/MOD4.0 andthe Nuclear Plant Analyser RELAP/SCDAPSIM-VISA pack-age (ViSA-RS) for numerical simulations of a natural circula-tion boiling water reactor (NCBWR) and supercritical watercooled reactor (SCWR). Figure 22 shows the example ofthe NCBWR schematic and nodalization. Their applicationsincluded the following areas:

(a) parametric studies of the primary heat transport loopof NCBWR,

(b) stability analysis of NCBWR,

(c) stability analysis of SCWR,

(d) educational use of RELAP5 and VISA-RS.


Acknowledgment

The contributions of the many SDTP members and licensedusers are gratefully acknowledged.

References

[1] http://www.relap.com/.

[2] http://www.sdtp.org/.

[3] RELAP5 Code Development Team, “RELAP5/MOD 3.3 CodeManual, Vol 1–8,” NUREG/CR-5535/Rev1, 2001.

[4] SCDAP/RELAP5 Development Team, “SCDAP/RELAP5/MOD3.2 Code Manual, Vol. 1–5,” NUREG/CR-6150, INEL-96/0422, July 1998.

[5] C. M. Allison, R. J. Wagner, L. J. Siefken, and J. K. Hohorst,“The development of RELAP5/SCDAPSIM/MOD4.0 for reac-tor system analysis and simulation,” in Proceedings of the 7thInternational Conference on Nuclear Option in Countries withSmall and Medium Electricity Grids, Dubrovnik, Croatia, May2008.

[6] A. R. Antariksawan, Md. Q. Huda, T. Liu, J. Zmitkova, C. M.Allison, and J. K. Hohorst, “Validation of RELAP/SCDAPSIM/MOD3.4 for research reactor applications,” in Proceedingsof the 13th International Conference on Nuclear Engineering(ICONE ’05), Beijing, China, May 2005.

[7] A. Sekhri, A. L. Graham, M. Belal, L. E. Moloko, andA. J. D’Arcy, “Thermal hydraulic and safety analyses of aproposed HEU & LEU core for SAFARI-1 research reactor,”in Proceedings of the 12th International Topical Meeting onResearch Reactor Fuel Management (RRFM ’08), Hamburg,Germany, 2008.

[8] A. Sekhri, A. J. D’Arcy, A. L. Graham, and M. Oliver,“Thermal-hydraulic modelling of the SAFARI-1 researchreactor using RELAP/SCDAPSIM/MOD3.4,” in Proceedingsof the International Conference on the Physics of Reactors,Interlaken, Switzerland, September, 2008.

[9] R. C. Nelson, J. C. McKibben, K. Kutikkad, and L. P. Foyto,“Thermal-hydraulic transient analysis of the Missouri Univer-sity Research Reactor (MURR),” in Proceedings of the AnnualMeeting of Test, Research and Training Reactors (TRTR ’07),September 2007.

[10] T. Haste, J. Birchley, E. Cazzoli, and J. Vitazkova, “Analysisof TMI-2 with MELCOR and SCDAPSIM,” in Proceedings ofthe International Congress on Advances in Nuclear Power Plants(ICAPP ’05), vol. 2, pp. 984–993, Seoul, South Korea, May2005.

[11] T. Haste, J. Birchley, and M. Richner, “MELCOR and SCDAPanalyses of loss-of-coolant accidents during cooldown in aWestinghouse 2-loop PWR,” in Proceedings of the InternationalConference, Nuclear Energy for New Europe, Portoroz, Slovenia,September 2007.

[12] T. Haste, J. Birchley, and J.-U. Klugel, “Analysis of stationblackout in the Gosgen nuclear plant,” in Proceedings of theInternational Conference on Advances in Nuclear Power Plants(ICAPP ’08), vol. 2, pp. 1086–1094, Anaheim, Calif, USA, June2008.

[13] I. Prisecaru, D. Dupleac, P. Ghitescu, and L. Biro, “Aparametric study of a large break in reactor inlet header ofCANDU6 reactors using RELAP5 code,” in Proceedings of theInternational Congress on Advances in Nuclear Power Plants(ICAPP ’06), pp. 1116–1125, Reno, Nev, USA, June 2006.

[14] D. Dupleac, I. Prisecaru, M. Mladin, G. Negut, and P.Ghitescu, “Investigation on header manifold model effectat CANDU6 large loss of cooling accident,” in Proceedingsof the 16th International Conference on Nuclear Engineering(ICONE ’08), vol. 4, pp. 673–679, Orlando, Fla, USA, May2008.

[15] D. Dupleac, I. Prisecaru, P. Ghitescu, and G. Negut, “Thermal-hydraulic analysis of CANDU 6 100% reactor outlet headerbreak using RELAP5 code,” in Proceedings of the InternationalCongress on Advances in Nuclear Power Plants (ICAPP ’07), vol.1, pp. 246–255, Nice, France, May 2007.

[16] B. N. Hanna, N. U. Aydemir, D. K. Baxter, et al., “CATHENATheoretical Manual: MOD-3.5/Rev. 0,” AECL RC-982-31,1994.

[17] D. Dupleac, I. Prisecaru, M. Mladin, and G. Negut,“SCDAP/RELAP5 investigation on coolability of severelydegraded CANDU 6 Core—preliminary results,” in Proceed-ings of the International Conference on Advances in NuclearPower Plants (ICAPP ’08), vol. 2, pp. 1095–1101, Anaheim,Calif, USA, June 2008.

[18] S.-Y. Park, Y.-H. Jin, and Y.-M. Song, “An investigationof an in-vessel corium retention strategy for the Wolsongpressurized heavy water reactor plants,” Nuclear Technology,vol. 158, no. 1, pp. 109–115, 2007.

[19] S. M. Petoukhov, B. Awadh, and P. M. Mathew, “CANDU6 severe core damage accident consequence analysis forsteam generator tube rupture scenario using MAAP4-CANDUV4.0.5A: preliminary results,” in Proceedings of the Inter-national Congress on Advances in Nuclear Power Plants(ICAPP ’06), pp. 1404–1414, Reno, Nev, USA, June 2006.

[20] E. Borisov, Personal Communication, Risk Engineering, Ltd,Sofia, Bulgaria, 2008.

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[22] E. Uspuras and A. Kaliatka, “Development of RBMK-1500model for BDBA analysis using RELAP/SCDAPSIM code,”Journal of Power and Energy, vol. 2, 2008.

[23] E. Urbonavicius, E. Uspuras, S. Rimkevicius, and A. Kaliatka,“Application of RELAP/SCOAPSIM and COCOSYS codes forsevere accident analysis in RBMK-1500 reactor,” in Proceedingsof the International Congress on Advances in Nuclear PowerPlants (ICAPP ’06), pp. 1442–1450, Reno, Nev, USA, June2006.

[24] C. Chavez-Mercado, J. K. Hohorst, and C. M. Alli-son, “National Autonomous University of Mexico RELAP/SCDAPSIM-based plant simulation and training applicationsto the Laguna Verde NPP,” in Proceedings of the 6th Interna-tional Conference on Nuclear Thermal Hydraulics, Operationsand Safety (NUTHOS ’04), Nara, Japan, October 2004.

[25] http://hikwww2.fzk.de/quench.[26] J. Birchley, T. Haste, W. Hering, and Ch. Homann, “Pre-test

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[30] http://phebus.jrc.nl.

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[34] http://www.istc.ru/istc/sc.nsf/html/projects.htm?open&id=3690.

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