Proceedings of the 4th International Topical Meeting on High Temperature Reactor Technology HTR2008
September 28-October 1, 2008, Washington, DC USA
ABSTRACT
Simulation of some fluid phenomena associated withGeneration IV reactors requires the capability of modelingmixing in two- or three-dimensional flow. At the same time, theflow condition of interest is often transient and depends uponboundary conditions dictated by the system behavior as awhole.
Computational fluid dynamics (CFD) is an ideal toolfor simulating mixing and three-dimensional flow in systemcomponents, whereas a system analysis tool is ideal formodeling the entire system. This paper presents the reasoningwhich has led to coupled CFD and systems analysis codesoftware to analyze the behavior of advanced reactor fluidsystem behavior. In addition, the kinds of scenarios where thiscapability is important are identified. The important role of acoupled CFD/systems analysis code tool in the overallcalculation scheme for a Very High Temperature Reactor isdescribed.
The manner in which coupled systems analysis andCFD codes will be used to evaluate the mixing behavior in aplenum for transient boundary conditions is described. Thecalculation methodology forms the basis for future coupledcalculations that will examine the behavior of such systems at aspectrum of conditions, including transient accident conditions,that define the operational and accident envelope of the subjectsystem. The methodology and analysis techniquesdemonstrated herein are a key technology that in part forms the
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backbone of the advanced techniques employed in theevaluation of advanced designs and their operationalcharacteristics for the Generation IV advanced reactor systems.
INTRODUCTION
This paper is a description of the process and some ofthe thermal-hydraulic tools that are being readied for theevaluation of plant behavior that will be undertaken for aGeneration IV Very High Temperature Gas-Cooled Reactor(VHTR). The process is similar in some respects to thatfollowed to prepare the thermal-hydraulic tools for performingcalculations for the Generation III+ advanced systems, e.g., theWestinghouse AP600 (Schultz, et al., 1997). The process, as itwill be applied to the Next Generation Nuclear Plant (NGNP)variant of the VHTR, is described in detail in the NGNPMethods Technical Program Plan (Schultz, et al., 2007)However, significant differences and improvements stem fromthe use of single-phase fluids in the VHTR that not only allow,but also dictate, the use of computational fluid dynamics (CFD)modeling. Furthermore, an important new dimension of thetools to be used on the VHTR are coupled thermal-hydraulicsoftware, that is, thermal-hydraulics and neutronics couplingand equally importantly, systems analysis codes1 such as
1 In such a coupling, systems analysis software is used to performcalculations of the overall system behavior considering the interactionsbetween all the parts, e.g., the core, the plena, the hot exit duct, the turbine, andthe remainder of the plant. CFD codes are used to calculate the detailed three-dimensional fluid behavior in a region of the reactor such as a plenum.
Evaluations of fluid behavior in the VHTR, at normaloperational conditions and during abnormal or accidentconditions, are key ingredients in the progression fromspecification of a system, to design of the system, to building ofthe system, and finally to licensing of the system. Becausenuclear reactors, by their nature, consist of nuclear fuel (whichmust be maintained within a safe operational envelope) coupledto the fluid behavior via heat transfer and neutronic interactions,the evaluations of fluid behavior are sometimes complex andhave many facets. To reduce the calculational envelope to amanageable level, various methodologies are used to identifynot only the most important scenarios for consideration, butalso the most important phenomena which must be calculatedwith high resolution and fidelity. Presently, the candidatereactor systems (pebble-bed and prismatic systems) underconsideration for the VHTR use high-temperature helium as theworking fluid. Hence the working fluid remains single-phaseduring during all scenarios of interest. The overall scenario andphenomena identification methodology for the VHTR isillustrated in Figure 1.
Activity 1, the selection of the scenarios andphenomena for analysis, define the kinds of software andanalysis tools, since the important phenomena combined withthe reactor geometry define whether a one-dimensional or amulti-dimensional analysis is required. The capability of thethermal-hydraulic tools requires models that are based on first-principles.
Once the scenarios and phenomena are identified, keyelements of the above process are Activities 2, 6, and 7 sincethese activities ensure the software are validated and shown tobe capable of calculating the important scenarios andphenomena. Activities 2 and 6 usually include comparison ofthe desired calculation to data and additional development ofthe analytical algorithms if the software are demonstrated to beincapable of calculating key phenomena/behavior. Activity 7 isthe ultimate objective of the effort. The remaining activitiesprovide input from sources that may assist in the effort byproviding expert review and collaborations to achieve thedesired objectives.
CALCULATIONAL PROCESS
Activity 7 is accomplished by a series of calculationsthat resolve into extensive calculational work in selected areas.The calculational process is shown in Figure 2 and issubdivided into seven steps as summarized in paragraphs a
2 See Schultz, et al., 2007 for references and further information onsoftware.
a. Material cross section compilation and evaluation.Nuclear interaction probabilities (cross sections) arefundamental to calculating high-fidelity neutronicscalculations. Microscopic cross sections are available on aper atom basis; however, the actual material densities,atomic compositions, and geometry must be used to obtainmacroscopic cross sections data sets in order to calculateneutron interaction rates on a per centimeter basis for thefuel, reactor core and structural materials, as a function oftemperature.
b. Preparation of homogenized cross-sections. The cross-section data are processed into a case-specific form usinglocal cell and assembly modeling codes. The basicphysical data are processed for case-specific resonanceshielding and then weighted with characteristic energy andspatial flux profiles generated from unit cell or super-cellmodels. This step is performed using software thatapproximates the neutron transport equation for the energyflux calculation and a one- or two-dimensional transportcode for the spatial flux. The geometric aspects of thisprocess are significantly different in the prismatic andpebble-bed concepts.
c. Whole-core analysis (diffusion or transport), detailedheating calculations, and safety parameter determination.Nodal diffusion-theory codes, such as PEBBED (PEBBEDis designed specifically for pebble-bed reactor simulation)are used to perform VHTR reactor core analysis. Steady-state eigenvalues, energy and spatial flux profiles, reactionrates, reactivity changes (burnup and control rodmovement), etc., will be calculated with the nodaldiffusion-theory codes. All of these software packages willbe verified against alternate computational models,especially models based on the well known MCNPstochastic simulation (Monte Carlo) code as shown inFigure 3, and various deterministic approaches. Spatialchanges in flux and power level as functions of time duringpostulated transients, predicted by the kinetics module, willprovide the energy source term required for the overallthermal-hydraulics systems code computations at each timestep during each transient. This process permits fullcoupling of thermal and neutronics computations,consistent with modern practice for nuclear systemsanalysis. A time-dependent implementation of thePEBBED code will be used for the pebble-bed concept.
f. Fuel behavior and fission product release. Theperformance of fuel particles under irradiation is modeledto determine whether fuel failure will occur, with thesubsequent release of fission products, and whethersubsequent migration of fission products throughout thesystem must be considered. The INL software designed toperform this function is called PARFUME.
g. Fission product transport. If a loss-of-coolant accidenthas occurred, such that the fission products may migrate orbe impelled into the confinement/containment buildingwith perhaps subsequent release to the environment, thenthe final calculational step is the prediction of the fissionproduct movement into the environment and itsenvironmental distribution. The software tool most likelyto be used to perform these calculations is MELCOR.
The process described in items “a” through “g” is shownin the flow chart of Figure 2. The complete calculation process
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illustrated in Figure 2 is only exercised in its entirety for a fewscenarios. Most scenarios would require the use of only afraction of the calculations represented in Stages a through e.For example, scenarios that do not include a loss of coolant,i.e., a pipe break, usually would not require calculation offission gas transport (Stage g). In addition, if the neutronicshave been thoroughly calculated for the reactor systemoperating condition (Stages a through c), then a multitude ofreactor system calculations can be performed using theevaluated reactor power state at time zero, and hence the Stagea through c calculations may only need to be performed oncefor a desired operating condition. Thereafter, for such scenariosthat assume reactor scram, a multitude of calculations can beperformed using only the software tools developed for Stages dand e. Coupled software, that is, systems analysis codes toneutronic codes and systems analysis codes to CFD codes areneeded for Stages c through g.
KEY PHENOMENA REQUIRING ANALYSIS AND NEEDFOR COUPLED SOFTWARE
To begin the process of defining the fluid behaviorscenarios that will require analysis and also to identify thesoftware that require validation, a preliminary evaluation of thephenomena requiring evaluation has been compiled and is listedin Table 1. The listed phenomena, identified during aphenomena identification and ranking table exercise sponsoredby the U.S. Nuclear Regulatory Commission (Ball, et al., 2007)stem from both normal operational conditions as well as theclassic depressurized conduction cooldown (DCC) andpressurized conduction cooldown (PCC) scenarios. Thefollowing phenomena have been determined to be important:neutronics behavior, core hot channel characterization, bypassanalysis, mixing, laminar-turbulent transition flow and forced-natural mixed convection flow, convective and radiation heattransfer in the reactor cavity cooling system (RCCS), air-wateringress, and fission product transport. The thermal- hydraulicphenomena of interest are described below:
i. Core hot channel characterization. The characteristics ofthe hottest cooling channels at operational conditions areconsidered a key calculational need since the hot channeltemperature distribution defines the hottest initial condition forthe fuel and surrounding materials. Hence preliminarycomputational fluid dynamics (CFD) studies have been initiatedand validation data are sought.
ii. Bypass. The bypass flow passes through the reflectorregions in both pebble-bed and block reactors and, in a block-type reactor, between the blocks. Because the quantity ofbypass flow is a direct function of the bypass area, which in turnis a function of the temperature distribution, fluence, andgraphite properties, the influence of the bypass on the coretemperature distribution may be significant. The influence of
bypass may be assessed in part by performing a series ofparametric calculations that differ in the geometric boundaryconditions—as defined by the various factors that influence thebypass flow passages such as manufacturing tolerances,misalignments, and geometric distortions.
iii. Mixing. Mixing refers to the degree to which coolant ofdiffering temperatures entering a region mixes to produce auniform temperature. Mixing is a three-dimensionalphenomenon in the inlet and outlet plenums and a function of anumber of variables. Thus, for example, for a prismatic designin the inlet plenum, where it is identified as important in thePCC scenario, mixing occurs during natural convection ashelium moves upward through the hottest portion of the corewhile cooler helium moves downward through the bypass andthe cooler regions of the core. In the outlet plenum, mixingoccurs between the bottom of the core and the turbine orimmediate heat exchanger inlet during normal operation. Apreliminary calculation of the temperature variation in the lowerplenum indicates that gas temperature variations could exceed300 C. Although the specification for temperature variation atthe immediate heat exchanger or turbine inlet has not been set,it is thought that the helium temperature variation must be lessthan 20 C. Also, it has been seen that helium has a surprisingresistance to thorough mixing [Ball 2004, based on experienceof Kunitoni, et al., 1986] and that the temperature in the coreoutlet jet can vary over a considerable range, particularly sincethe bypass flow may vary between 10% and 25%. Therefore, itis likely that special design features will be required to ensuregood mixing and minimal thermal streaking from the lowerplenum to the turbine inlet.
iv. Laminar-Turbulent Transition Flow and. Forced-NaturalMixed Convection Flow. During the PCC scenario in the coreregion and during both the PCC and DCC scenarios in thereactor cavity cooling system (RCCS), there is the potential forhaving convective cooling in the transition region. Because theconvective cooling contribution is an important ingredient indescribing the total heat transfer from the core and thus theultimate peak core and vessel temperatures, these heat transferphenomena are potentially important.
v. Air and Water Ingress. For loss-of-coolant scenarios, suchas the DCC, there is the potential for air and water ingress intothe core in perhaps harmful quantities—depending on thescenario assumptions. Air may be present in the reactor cavity(some designs have a cavity filled with inert gas) and may enterthe core by diffusion in a DCC accident. Water is normallypresent in the air in the form of humidity, but it may enter thecore in much greater quantities, with much greater potentialeffect on reactivity, if the shutdown cooling system suffers apipe leak or break. Oxidation of graphite in the prismatic coredesign is also a potential safety issue.
vi. RCCS. Analyses of the natural circulation and radiationheat transfer that will occur in the reactor cavity are crucial to
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determine the peak temperatures of the structural members andthe fuel in particular. It is envisioned that the mixing in thecavity will be done using a CFD code such as FLUENT whilethe boundary conditions to the CFD code will be provided by asystems analysis code such as RELAP5-3D.
A common thread that connects items i through vi is theneed to couple one-dimensional analyses that provide boundaryconditions to multi-dimensional calculations for scenarios thatrequire evaluation as a function of time. Hence results ofanalyses performed for items i and ii lead to boundaryconditions for item iii. Also, a mixture of one-dimensional andmulti-dimensional analysis requirements are needed to satisfythe calculational needs for items iv through vi.
CFD VERSUS SYSTEMS ANALYSIS CODES
Computational Fluid Dynamics (CFD) codes, such asSTAR-CCM+, are commonly used to analyze the flow behaviorin regions of a system where complex flow patterns areexpected or present. Therefore, the CFD codes are usually usedto analyze either two-dimensional or three-dimensional flowbehavior. The great appeal of the CFD codes is their relianceon first principles to describe the fluid behavior and theircapability to calculate the behavior of many complex flowpatterns. On the other hand, the CFD codes also rely on a finemesh discretization to model the region of interest;consequently, even with modern fast computers, the region of asystem that can be modeled is generally limited as defined bypractical computing times. Thus, CFD codes are rarely used tomodel the behavior of an entire system and instead are focusedon the behavior of a region of a system or component.Although today’s CFD codes are well suited to analyze a wide range of single-phase problems that do not undergo phasechanges, their range of applicability to two-phase or multi-phaseproblems is limited.
Although CFD software is rarely used to calculatedsystem-wide behavior, there are experts who advocate thedevelopment of coarse-grain CFD meshes designed to capturethe key phenomena (Class, et al., 2008). The advantages of thisapproach rest with the ability to perform system-widecalculations using a consistent set of field equations throughoutthe calculation. However, the approach for defining and usingcoarse-grain CFD (Hierarchical multi-scale CFD) is a leadingRandD area and a working methodology has not been defined todate.
Thus, today’s CFD codes are tools capable ofanalyzing selected two-phase applications. For example,FLUENT has been demonstrated to model applicationsinvolving film boiling, aerosol deposition in a condenser,nucleate boiling and subcooled nucleate boiling. Systemsanalysis codes, such as RELAP5 on the other hand, are toolsthat can be used to analyze the two-phase phenomena andconditions that will be encountered by the equipment they weredesigned to analyze.
The FLUENT/RELAP5-3D coupled software hasbeen validated and verified by modeling a portion of a simplesystem using FLUENT while the balance of the system wasmodeled using RELAP5-3D. The system is shown in Figure4—and a blowup of the portion modeled using FLUENT isshown in Figure 5.
The integrity of the coupling was validated byinspecting the boundary conditions for both the RELAP5-3Dmodel and the FLUENT at the boundaries (labeled zones 2 and3 in Figures 4 and 5) to establish that the pressures and fluidflow conditions were correct.
ONGOING ANALYSES AND OBSERVATIONS
Since the validation work described in the paragraphsabove, the practices and procedures for using the coupled toolshave been studied and expanded. In a recent paper theFLUENT calculational envelope was studied to determine therange of applicability of the FLUENT segregated solver3
3 FLUENT may be used with coupled or segregated solvers. The coupledsolver is used with higher Mach number flow problems where changes in thefluid properties are important to integrate with the velocity field calculation.The segregated solver is used for lower Mach number flows, and especially forincompressible flows.
A clear need to have a coupled CFD and systems analysistool has been identified. The tool exists and development iscontinuing to accommodate the needs specific to the VHTR.Presently the calculational envelope of the coupledCFD/systems analysis tool is being defined and documented.
REFERENCES
A. G. Class, et al.,“Hierarchical Multi-Scale ThermalHydraulics in Nuclear Applications,” Proceedings of theInternational Workshop on Thermal-Hydraulics of InnovativeReactor and Transmutation Systems - THIRSApril 14-16, 2008, Forschungszentrum Karlsruhe, Germany
S. J. Ball, M.. Corradini, S. E. Fisher, R. Gauntt, G.Geffraye, J. C. Gehin, Y. Hassan, D. L.. Moses, J. P. Renier, R.R. Schultz, and T. Wei, Next-Generation Nuclear Plant (NGNP)Phenomena Identification and Ranking Table (PIRT) forAccident and Thermal Fluids Analysis, NUREG/CR-6944,September, 2007
Schowalter, D.G., N. Basu, A. Walavalkar, and R. R.Schultz, “Discussion on the Calculational Envelope of the FLUENT Computational Fluid Dynamics Code and theRELAP5 Systems Analysis Code when Using Segreagated
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Solvers,” Proceedings of the American Nuclear Society WinterMeeting, November, 2004.
Schultz, R. R., C. M. Kullberg, G. E. McCReery, R. A.Shaw, B. Hanson, N. Newman, C. P. Liou, J. L. Westacott,RELAP5/MOD3 Code Assessment Analyses Based on theROSA-AP600 Program: Small Break LOCAs and the StationBlackout Transient, March, 1997.
Schultz, R. R., A. M. Ougouag, D. W. Nigg, H. D. Gougar,R. W. Johnson, W. K. Terry, C. H. Oh, D. M. McEligot, G. W.Johnsen, G. E. McCreery, W. Y. Yoon, J. W. Sterbentz, J. S.Herring, T. A. Taiwo, T. Y. C. Wei, W. D. Pointer, W. S. Yang,M. T. Farmer, H. S. Khalil, M. A. Feltus, Next GenerationNuclear Plant Methods Technical Program Plan, INL/EXT-06-11804, January, 2007.
Schultz, R. R., A. M. Ougouag, D. W. Nigg, H. D. Gougar,R. W. Johnson, W. K. Terry, C. H. Oh, D. M. McEligot, G. W.Johnsen, G. E. McCreery, W. Y. Yoon, J. W. Sterbentz, J. S.Herring, T. A. Taiwo, T. Y. C. Wei, W. D. Pointer, W. S. Yang,M. T. Farmer, H. S. Khalil, M. A. Feltus, Next GenerationNuclear Plant Methods Technical Program Plan, INL/EXT-06-11804, Revision 1, August, 2008 (to be published).
1. VHTR Project Scenario Selection & Phenomena Identification: PhenomenaIdentification & Ranking Table (PIRT) process used to select the scenarios and to identifythe phenomena of importance.
2. VHTR ProjectSoftwareValidation:Analysis tools areevaluated todetermine whetherimportantphenomena can becalculated.
4. Collaborations with GIF -Partners:Use I -NERIs as medium for internationalrelationships and collaboration projects tovalidate & develop software.
5. Collaborations with UniversitiesUse NERIs as vehicle for R&Drelationships with universities to focus onpertinent VHTR R&D issues (validation &development).
6. Developm ent coordinated by VHTR Project: Ifimportant phenomena cannot be calculated by analysistools, then further development is undertaken.
7. Analysis: The operational and accident scenarios that require study are analyzed.
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Figure 1. Methods RandD process—(see Schultz, et al., 2007)
Table 1. Phenomena identified for analysis during for normal operation, PCC and DCCscenarios.
Scenario InletPlenum
Core RCCS OutletPlenum
Normaloperation
i. Neutronic behaviorii. Bypass flow
iii. Hot channel characteristics
Mixing
DCC i. Thermal radiation and conduction of heatacross the coreii. Axial heat conduction and radiationiii. Natural circulation in the reactorpressure vesseliv. Air and water ingressv. Potential fission product transport
i. Laminar-turbulenttransition flowii. Forced-natural mixedconvection flow
PCC Mixing i. Neutronic behaviorii. Bypassiii. Laminar-turbulent transition flowiv. Forced-natural mixed convection flowv. Hot channel characteristics at operationalconditions
i. Laminar-turbulenttransition flowii. Forced-natural mixedconvection flow
FLUENT Not used Preferred tool Not used Superior for specialized applications—but generallyunable to model phenomena behavior over widethermodynamic ranges and through phase transitions.Fluid properties must be input; FLUENT does not havesteam tables.
Preferred tool for analyses of integral system behavior andapplications that require analyses over wide fluidthermodynamic state ranges and through phase transitions.
