MEETING THE THERMAL-HYDRAULIC ANALYSIS NEEDS FOR ADVANCED GAS REACTOR SYSTEMS

Transcription

1 Proceedings of the 4th International Topical Meeting on High Temperature Reactor Technology HTR2008 September 28-October 1, 2008, Washington, DC USA HTR MEETING THE THERMAL-HYDRAULIC ANALYSIS NEEDS FOR ADVANCED GAS REACTOR SYSTEMS Richard R. Schultz Idaho National Laboratory P.O. Box Fremont Idaho Falls, Idaho, USA (208) ; ABSTRACT Simulation of some fluid phenomena associated with Generation IV reactors requires the capability of modeling mixing in two- or three-dimensional flow. At the same time, the flow condition of interest is often transient and depends upon boundary conditions dictated by the system behavior as a whole. Computational fluid dynamics (CFD) is an ideal tool for simulating mixing and three-dimensional flow in system components, whereas a system analysis tool is ideal for modeling the entire system. This paper presents the reasoning which has led to coupled CFD and systems analysis code software to analyze the behavior of advanced reactor fluid system behavior. In addition, the kinds of scenarios where this capability is important are identified. The important role of a coupled CFD/systems analysis code tool in the overall calculation scheme for a Very High Temperature Reactor is described. The manner in which coupled systems analysis and CFD codes will be used to evaluate the mixing behavior in a plenum for transient boundary conditions is described. The calculation methodology forms the basis for future coupled calculations that will examine the behavior of such systems at a spectrum of conditions, including transient accident conditions, that define the operational and accident envelope of the subject system. The methodology and analysis techniques demonstrated herein are a key technology that in part forms the backbone of the advanced techniques employed in the evaluation of advanced designs and their operational characteristics for the Generation IV advanced reactor systems. INTRODUCTION This paper is a description of the process and some of the thermal-hydraulic tools that are being readied for the evaluation of plant behavior that will be undertaken for a Generation IV Very High Temperature Gas-Cooled Reactor (VHTR). The process is similar in some respects to that followed to prepare the thermal-hydraulic tools for performing calculations for the Generation III+ advanced systems, e.g., the Westinghouse AP600 (Schultz, et al., 1997). The process, as it will be applied to the Next Generation Nuclear Plant (NGNP) variant of the VHTR, is described in detail in the NGNP Methods Technical Program Plan (Schultz, et al., 2007) However, significant differences and improvements stem from the 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 the tools to be used on the VHTR are coupled thermal-hydraulic software, that is, thermal-hydraulics and neutronics coupling and equally importantly, systems analysis codes 1 such as 1 In such a coupling, systems analysis software is used to perform calculations of the overall system behavior considering the interactions between all the parts, e.g., the core, the plena, the hot exit duct, the turbine, and the remainder of the plant. CFD codes are used to calculate the detailed threedimensional fluid behavior in a region of the reactor such as a plenum. 1

2 RELAP5-3D 2 coupled to CFD codes such as FLUENT or STAR-CCM+. Evaluations of fluid behavior in the VHTR, at normal operational conditions and during abnormal or accident conditions, are key ingredients in the progression from specification of a system, to design of the system, to building of the system, and finally to licensing of the system. Because nuclear reactors, by their nature, consist of nuclear fuel (which must be maintained within a safe operational envelope) coupled to the fluid behavior via heat transfer and neutronic interactions, the evaluations of fluid behavior are sometimes complex and have many facets. To reduce the calculational envelope to a manageable level, various methodologies are used to identify not only the most important scenarios for consideration, but also the most important phenomena which must be calculated with high resolution and fidelity. Presently, the candidate reactor systems (pebble-bed and prismatic systems) under consideration for the VHTR use high-temperature helium as the working fluid. Hence the working fluid remains single-phase during during all scenarios of interest. The overall scenario and phenomena identification methodology for the VHTR is illustrated in Figure 1. Activity 1, the selection of the scenarios and phenomena for analysis, define the kinds of software and analysis tools, since the important phenomena combined with the reactor geometry define whether a one-dimensional or a multi-dimensional analysis is required. The capability of the thermal-hydraulic tools requires models that are based on firstprinciples. Once the scenarios and phenomena are identified, key elements of the above process are Activities 2, 6, and 7 since these activities ensure the software are validated and shown to be capable of calculating the important scenarios and phenomena. Activities 2 and 6 usually include comparison of the desired calculation to data and additional development of the analytical algorithms if the software are demonstrated to be incapable of calculating key phenomena/behavior. Activity 7 is the ultimate objective of the effort. The remaining activities provide input from sources that may assist in the effort by providing expert review and collaborations to achieve the desired objectives. CALCULATIONAL PROCESS Activity 7 is accomplished by a series of calculations that resolve into extensive calculational work in selected areas. The calculational process is shown in Figure 2 and is subdivided into seven steps as summarized in paragraphs a software. 2 See Schultz, et al., 2007 for references and further information on through g below. Interactions between thermal-hydraulic tools and other tools such as reactor physics and fuel behavior tools arise in Step c and are important for the remainder of the process (Steps d through g). Figure 3 identifies the software currently associated with each of the steps in Figure 2 for a pebble-bed reactor system. It is noted that a systems analysis code such as RELAP5-3D and a CFD code such as STAR- CCM+ play a central role in the process see Figure 3. Accommodations are made for using other tools if specific needs arise. a. Material cross section compilation and evaluation. Nuclear interaction probabilities (cross sections) are fundamental to calculating high-fidelity neutronics calculations. Microscopic cross sections are available on a per atom basis; however, the actual material densities, atomic compositions, and geometry must be used to obtain macroscopic cross sections data sets in order to calculate neutron interaction rates on a per centimeter basis for the fuel, reactor core and structural materials, as a function of temperature. b. Preparation of homogenized cross-sections. The crosssection data are processed into a case-specific form using local cell and assembly modeling codes. The basic physical data are processed for case-specific resonance shielding and then weighted with characteristic energy and spatial flux profiles generated from unit cell or super-cell models. This step is performed using software that approximates the neutron transport equation for the energy flux calculation and a one- or two-dimensional transport code for the spatial flux. The geometric aspects of this process are significantly different in the prismatic and pebble-bed concepts. c. Whole-core analysis (diffusion or transport), detailed heating calculations, and safety parameter determination. Nodal diffusion-theory codes, such as PEBBED (PEBBED is designed specifically for pebble-bed reactor simulation) are used to perform VHTR reactor core analysis. Steadystate eigenvalues, energy and spatial flux profiles, reaction rates, reactivity changes (burnup and control rod movement), etc., will be calculated with the nodal diffusion-theory codes. All of these software packages will be verified against alternate computational models, especially models based on the well known MCNP stochastic simulation (Monte Carlo) code as shown in Figure 3, and various deterministic approaches. Spatial changes in flux and power level as functions of time during postulated transients, predicted by the kinetics module, will provide the energy source term required for the overall thermal-hydraulics systems code computations at each time step during each transient. This process permits full coupling of thermal and neutronics computations, consistent with modern practice for nuclear systems analysis. A time-dependent implementation of the PEBBED code will be used for the pebble-bed concept. 2

3 d. Thermal-hydraulic and thermal-mechanical evaluations of system behavior. The fluid behavior, and interactions with the neutronics, will be calculated using a systems analysis code, and a coupled systems analysis/ computational fluid dynamics (CFD) code when necessary. Examples of a systems analysis code and a CFD code are RELAP5-3D and FLUENT. In such a coupling, systems analysis software may be used to perform calculations of the overall system behavior considering the interactions between all the parts, e.g., the core, the plenums, the hot exit duct, the turbine, and the remainder of the plant. CFD codes, such as STAR-CCM+, are used to calculate the detailed three-dimensional fluid behavior in a region of the reactor such as a plenum. In addition to analyzing the fluid behavior under a spectrum of operating and accident conditions, the thermal-hydraulic tools also will be used to investigate the significance of material geometric tolerance variations due to manufacturing, thermal responses, and irradiation effects such as graphite swelling. The need to examine factors that affect thermal-mechanical influence on fluid and heat transfer behavior will be included in the tool selection and evaluation process. e. Models for balance of plant electrical generation system and hydrogen production plant. The behavior of the balance-of-plant systems will be modeled using a systems analysis code such as RELAP5-3D. The balance-of-plant models are important to include in the analysis process to account for the important interactions that affect the system efficiency during normal operational conditions, but also to account for the equipment interactions that may lead to undesirable conditions such as turbine over-speed, loss of net positive suction head for auxiliary systems, or oscillatory conditions that may lead to equipment damage. Interactions between the reactor system and its balance-ofplant components lead to boundary conditions that will determine whether fuel-damaging conditions are likely (see item f). f. Fuel behavior and fission product release. The performance of fuel particles under irradiation is modeled to determine whether fuel failure will occur, with the subsequent release of fission products, and whether subsequent migration of fission products throughout the system must be considered. The INL software designed to perform this function is called PARFUME. g. Fission product transport. If a loss-of-coolant accident has occurred, such that the fission products may migrate or be impelled into the confinement/containment building with perhaps subsequent release to the environment, then the final calculational step is the prediction of the fission product movement into the environment and its environmental distribution. The software tool most likely to be used to perform these calculations is MELCOR. The proces described in items a through g is shown in the flow chart of Figure 2. The complete calculation process illustrated in Figure 2 is only exercised in its entirety for a few scenarios. Most scenarios would require the use of only a fraction 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 of fission gas transport (Stage g). In addition, if the neutronics have been thoroughly calculated for the reactor system operating condition (Stages a through c), then a multitude of reactor system calculations can be performed using the evaluated reactor power state at time zero, and hence the Stage a through c calculations may only need to be performed once for a desired operating condition. Thereafter, for such scenarios that assume reactor scram, a multitude of calculations can be performed using only the software tools developed for Stages d and e. Coupled software, that is, systems analysis codes to neutronic codes and systems analysis codes to CFD codes are needed for Stages c through g. KEY PHENOMENA REQUIRING ANALYSIS AND NEED FOR COUPLED SOFTWARE To begin the process of defining the fluid behavior scenarios that will require analysis and also to identify the software that require validation, a preliminary evaluation of the phenomena requiring evaluation has been compiled and is listed in Table 1. The listed phenomena, identified during a phenomena identification and ranking table exercise sponsored by the U.S. Nuclear Regulatory Commission (Ball, et al., 2007) stem from both normal operational conditions as well as the classic depressurized conduction cooldown (DCC) and pressurized conduction cooldown (PCC) scenarios. The following phenomena have been determined to be important: neutronics behavior, core hot channel characterization, bypass analysis, mixing, laminar-turbulent transition flow and forcednatural mixed convection flow, convective and radiation heat transfer in the reactor cavity cooling system (RCCS), air-water ingress, and fission product transport. The thermal- hydraulic phenomena of interest are described below: i. Core hot channel characterization. The characteristics of the hottest cooling channels at operational conditions are considered a key calculational need since the hot channel temperature distribution defines the hottest initial condition for the fuel and surrounding materials. Hence preliminary computational fluid dynamics (CFD) studies have been initiated and validation data are sought. ii. Bypass. The bypass flow passes through the reflector regions in both pebble-bed and block reactors and, in a blocktype reactor, between the blocks. Because the quantity of bypass flow is a direct function of the bypass area, which in turn is a function of the temperature distribution, fluence, and graphite properties, the influence of the bypass on the core temperature distribution may be significant. The influence of 3

4 bypass may be assessed in part by performing a series of parametric calculations that differ in the geometric boundary conditions as defined by the various factors that influence the bypass flow passages such as manufacturing tolerances, misalignments, and geometric distortions. iii. Mixing. Mixing refers to the degree to which coolant of differing temperatures entering a region mixes to produce a uniform temperature. Mixing is a three-dimensional phenomenon in the inlet and outlet plenums and a function of a number of variables. Thus, for example, for a prismatic design in the inlet plenum, where it is identified as important in the PCC scenario, mixing occurs during natural convection as helium moves upward through the hottest portion of the core while cooler helium moves downward through the bypass and the cooler regions of the core. In the outlet plenum, mixing occurs between the bottom of the core and the turbine or immediate heat exchanger inlet during normal operation. A preliminary calculation of the temperature variation in the lower plenum indicates that gas temperature variations could exceed 300 C. Although the specification for temperature variation at the immediate heat exchanger or turbine inlet has not been set, it is thought that the helium temperature variation must be less than 20 C. Also, it has been seen that helium has a surprising resistance to thorough mixing [Ball 2004, based on experience of Kunitoni, et al., 1986] and that the temperature in the core outlet jet can vary over a considerable range, particularly since the bypass flow may vary between 10% and 25%. Therefore, it is likely that special design features will be required to ensure good mixing and minimal thermal streaking from the lower plenum to the turbine inlet. iv. Laminar-Turbulent Transition Flow and. Forced-Natural Mixed Convection Flow. During the PCC scenario in the core region and during both the PCC and DCC scenarios in the reactor cavity cooling system (RCCS), there is the potential for having convective cooling in the transition region. Because the convective cooling contribution is an important ingredient in describing the total heat transfer from the core and thus the ultimate peak core and vessel temperatures, these heat transfer phenomena are potentially important. v. Air and Water Ingress. For loss-of-coolant scenarios, such as the DCC, there is the potential for air and water ingress into the core in perhaps harmful quantities depending on the scenario assumptions. Air may be present in the reactor cavity (some designs have a cavity filled with inert gas) and may enter the core by diffusion in a DCC accident. Water is normally present in the air in the form of humidity, but it may enter the core in much greater quantities, with much greater potential effect on reactivity, if the shutdown cooling system suffers a pipe leak or break. Oxidation of graphite in the prismatic core design is also a potential safety issue. vi. RCCS. Analyses of the natural circulation and radiation heat transfer that will occur in the reactor cavity are crucial to determine the peak temperatures of the structural members and the fuel in particular. It is envisioned that the mixing in the cavity will be done using a CFD code such as FLUENT while the boundary conditions to the CFD code will be provided by a systems analysis code such as RELAP5-3D. A common thread that connects items i through vi is the need to couple one-dimensional analyses that provide boundary conditions to multi-dimensional calculations for scenarios that require evaluation as a function of time. Hence results of analyses performed for items i and ii lead to boundary conditions for item iii. Also, a mixture of one-dimensional and multi-dimensional analysis requirements are needed to satisfy the calculational needs for items iv through vi. CFD VERSUS SYSTEMS ANALYSIS CODES Computational Fluid Dynamics (CFD) codes, such as STAR-CCM+, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. Therefore, the CFD codes are usually used to analyze either two-dimensional or three-dimensional flow behavior. The great appeal of the CFD codes is their reliance on first principles to describe the fluid behavior and their capability to calculate the behavior of many complex flow patterns. On the other hand, the CFD codes also rely on a fine mesh discretization to model the region of interest; consequently, even with modern fast computers, the region of a system that can be modeled is generally limited as defined by practical computing times. Thus, CFD codes are rarely used to model the behavior of an entire system and instead are focused on the behavior of a region of a system or component. Although today s CFD codes are wel suited to analyze a wide range of single-phase problems that do not undergo phase changes, their range of applicability to two-phase or multi-phase problems is limited. Although CFD software is rarely used to calculated system-wide behavior, there are experts who advocate the development of coarse-grain CFD meshes designed to capture the key phenomena (Class, et al., 2008). The advantages of this approach rest with the ability to perform system-wide calculations using a consistent set of field equations throughout the calculation. However, the approach for defining and using coarse-grain CFD (Hierarchical multi-scale CFD) is a leading RandD area and a working methodology has not been defined to date. Systems analysis codes, such as RELAP5-3D, are generally aimed at modeling the behavior of an entire system such as a high temperature gas-cooled reactor including the balance-of-plant. While such codes generally have the capability to model multi-dimensional effects, their capacity to produce widely accepted analyses of multi-dimensional behavior is limited by the assumptions and capabilities that stem 4

5 from their field equation formulations, for example, the viscous stress terms are missing from their field equation formulations. Historically RELAP5-3D was developed first to analyze the behavior of two-phase systems that could be modeled in onedimension. Because of the need to analyze two-phase flow, the assumptions used to define the field equations resulted in a simplification of the viscous stress terms and the use of many empirical relationships that cannot be traced to first-principles, e.g., flow regime transitions and the models describing the interactions between phases. The RELAP5-3D field equation set was later extended to analyze two- and three-dimensions. However, the assumptions inherent to the one-dimensional equation set were retained. The starting point and the needs that led to the development of codes such as STAR-CCM+ and RELAP5-3D, not surprisingly, led to different products with different capabilities, limitations, strengths, and weaknesses. This subject was addressed in Schultz The recent coupling of FLUENT and RELAP5-3D was executed to take advantage of the strengths of each. For example, the systems analysis code can be used to model the balance of a system while the CFD code can be used to model a portion of the same system in great detail. When the two codes are coupled, then the systems analysis code will provide the boundary conditions to the CFD code so that a transient can be modeled with some confidence. Thus interactions between the balance-of-system and the detailed CFD model of a portion of the system can be simulated. In summary, the fundamental strengths and weaknesses of the FLUENT and RELAP5-3D codes, from an analysis perspective, are given in Table 2. The CFD codes are without peer when analyzing the complex flow behavior of single-phase systems in two- or three-dimensions, for either steady-state or transient behavior. The systems analysis codes, such as RELAP5-3D, are without peer for analysis of twophase systems in one-, two-, or three-dimensions. However, as usual, there are exceptions to these statements, e.g., systems analysis codes cannot analyze the behavior of stratified flow systems such as warm water over cold water with a gas above the free surface. While CFD codes can analyze this behavior, commercial CFD codes cannot analyze the behavior of a vapor that can condense on the free surface, such as steam over water. Indeed, commercial CFD codes do not even have the steam tables included in their source coding. Thus, today s CFD codes are tools capable of analyzing selected two-phase applications. For example, FLUENT has been demonstrated to model applications involving film boiling, aerosol deposition in a condenser, nucleate boiling and subcooled nucleate boiling. Systems analysis codes, such as RELAP5 on the other hand, are tools that can be used to analyze the two-phase phenomena and conditions that will be encountered by the equipment they were designed to analyze. COUPLED CFD AND SYSTEMS ANALYSIS CODES FLUENT and RELAP5-3D were linked using an Executive Program (Weaver, et al., 2002) that (a) monitors the calculational progression in each code, (b) determines when each code has converged, (c) governs the information interchanges between the codes, and (d) issues permission to allow each code to progress to the next time step. The Executive Program was interfaced with FLUENT and RELAP5-3D using user-defined functions. User-defined functions were also used to ensure the fluid properties used by FLUENT and RELAP5-3D are equivalent. The Executive Program uses the Oak Ridge National Laboratory Parallel Virtual Machine (PVM) computer software program to control the interactions, data and message passing between the Fluent and RELAP5-3D codes which are running independently. The PVM program execution serves as a "traffic cop" between the codes and gets data from one code at the current time step and passes it to the other code so that the NGNP analysis can be done consistently in parallel. The FLUENT/RELAP5-3D coupled software has been validated and verified by modeling a portion of a simple system using FLUENT while the balance of the system was modeled using RELAP5-3D. The system is shown in Figure 4 and a blowup of the portion modeled using FLUENT is shown in Figure 5. The integrity of the coupling was validated by inspecting the boundary conditions for both the RELAP5-3D model and the FLUENT at the boundaries (labeled zones 2 and 3 in Figures 4 and 5) to establish that the pressures and fluid flow conditions were correct. ONGOING ANALYSES AND OBSERVATIONS Since the validation work described in the paragraphs above, the practices and procedures for using the coupled tools have been studied and expanded. In a recent paper the FLUENT calculational envelope was studied to determine the range of applicability of the FLUENT segregated solver 3 (Schowalter, et al., 2004). And for the first time the coupled FLUENT/RELAP5-3D software are being used to analyze the behavior of the flow in the lower plenum of an advanced gascooled reactor in conjunction with changing boundary conditions in the remainder of the system. 3 FLUENT may be used with coupled or segregated solvers. The coupled solver is used with higher Mach number flow problems where changes in the fluid properties are important to integrate with the velocity field calculation. The segregated solver is used for lower Mach number flows, and especially for incompressible flows. 5

6 The RELAP5-3D model of the reactor vessel itself is illustrated in Figures 6 and 7 (INL model figures courtesy of Korea Atomic Energy Research Institute). The FLUENT model (courtesy of FLUENT), shown in Figure 8, replaces Components 160 and 170 of the RELAP5-3D model. A clear need to have a coupled CFD and systems analysis tool has been identified. The tool exists and development is continuing to accommodate the needs specific to the VHTR. Presently the calculational envelope of the coupled CFD/systems analysis tool is being defined and documented. REFERENCES A. G. Class, et al., Hierarchical Multi-Scale Thermal Hydraulics in Nuclear Applications, Proceedings of the International Workshop on Thermal-Hydraulics of Innovative Reactor and Transmutation Systems - THIRS April 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) for Accident and Thermal Fluids Analysis, NUREG/CR-6944, September, 2007 Schowalter, D.G., N. Basu, A. Walavalkar, and R. R. Schultz, Discusion on the Calculational Envelope of the FLUENT Computational Fluid Dynamics Code and the RELAP5 Systems Analysis Code when Using Segreagated Solvers, Proceedings of the American Nuclear Society Winter Meeting, November, 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 the ROSA-AP600 Program: Small Break LOCAs and the Station Blackout Transient, March, Schultz, R. R., R. A. Riemke, C. B. Davis, and G. Nurnburg, Comparison: RELAP5-3D Systems Analysis Code and FLUENT CFD Code Momentum Equation Formulations, Proceedings of the ICONE-11, Tokyo, Japan, April, 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 Generation Nuclear Plant Methods Technical Program Plan, INL/EXT , January, 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 Generation Nuclear Plant Methods Technical Program Plan, INL/EXT , Revision 1, August, 2008 (to be published). Weaver, W.L., E. T. Tomlinson, and D. L. Aumiller, 2002, A Generic Semi-Implicit Coupling methodology for Use in RELAP5-3D, Nuclear Engineering and Design, 211, pages 13 to 26. 6

7 1. VHTR Project Scenario Selection & Phenom ena Identification: Phenomena Identification & Ranking Table (PIRT) process used to select the scenarios and to identify the phenomena of importance. 3. Validation & Developm ent by Comm unity: Validation performed by analysis community via international standard problems. 2. VH TR Project Softw are Validation: Analysis tools are evaluated to determine whether important phenomena can be calculated. 4. Collaborations w ith GIF -Partners: Use I -NERIs as medium for international relationships and collaboration projects to validate & develop software. 5. Collaborations w ith Universities Use NERIs as vehicle for R&D relationships with universities to focus on pertinent VHTR R& D issues (validation & development). 6. Developm ent coordinated by VH TR Project: If important phenomena cannot be calculated by analysis tools, then further developm ent is undertaken. 7. Analysis: The operational and accident scenarios that require study are analyzed. 8. Peer review: Nuclear community peer review of methods R&D process Figure 1. Methods RandD process (see Schultz, et al., 2007) 7

8 a. Material Cross Section Compilation and Evaluation b. Preparation of Homogenized Cross Sections c. Whole-Core Analysis (Diffusion or Transport), Detailed Heating Calculation, and Safety Parameter Determination e. Models for Balance of Plant Electrical Generation System and Hydrogen Production Plant d. Thermal-Hydraulic and Thermal-Mechanical Evaluation of System Behavior g. Fission g. Fission Product Gas Transport Transport f. f. Fuel Behavior: and Fission Product Gas Release Evaluation Figure 2. Calculation process (see Schultz, et al., 2007) 8

9 Figure 3. Application of process to pebble-bed candidate designs for VHTR with applicable software see Schultz, et al.,

10 Table 1. Phenomena identified for analysis during for normal operation, PCC and DCC scenarios. Scenario Normal operation Inlet Plenum Core RCCS Outlet Plenum i. Neutronic behavior Mixing ii. Bypass flow iii. Hot channel characteristics DCC i. Thermal radiation and conduction of heat across the core ii. Axial heat conduction and radiation iii. Natural circulation in the reactor pressure vessel iv. Air and water ingress v. Potential fission product transport PCC Mixing i. Neutronic behavior ii. Bypass iii. Laminar-turbulent transition flow iv. Forced-natural mixed convection flow v. Hot channel characteristics at operational conditions i. Laminar-turbulent transition flow ii. Forced-natural mixed convection flow i. Laminar-turbulent transition flow ii. Forced-natural mixed convection flow Mixing Table 2 Comparison of FLUENT and RELAP5-3D Capabilities Single-Phase Two-Phase 1-D 2- or 3-D 1-D 2- or 3-D FLUENT Not used Preferred tool Not used Superior for specialized applications but generally unable to model phenomena behavior over wide thermodynamic ranges and through phase transitions. Fluid properties must be input; FLUENT does not have steam tables. RELAP5-3D Preferred tool Preferred tool Input assumptions are required. Preferred tool for analyses of integral system behavior and applications that require analyses over wide fluid thermodynamic state ranges and through phase transitions. 10

11 TMDPVOL Component 210: contain SNGLJUN Component 200: outlet SNGLJUN Component 180: cortoup SNGLVOL Compt 190: upperp PIPE Component 16: upcore SNGLJUN Component 115: upcrin SNGLJUN Component 910: bytoup TMDPVOL Component 15: coretop SNGLJUN Component 115: upcrin Zone 3 of FLUENT model Zone 3 of FLUENT model SNGLJUN Component 105: lcrout FLUENT model Zone 2 of FLUENT model PIPE Component 2: bypass FLUENT model TMDPVOL Component 6: corebtm Zone 2 of FLUENT model PIPE Component 1: lrcore SNGLJUN Component 130: cortolp SNGLVOL Compt 120: upperp TMDPVOL Component 110: contain SNGLJUN Compt 100: inlet Figure 5. Blow-up of FLUENT component in system model Figure 4. System model with FLUENT 3-D component 11

12 130 (Riser & Head) 140 (Upper plenum) Outer Reflecor Outer Core Middle Core Inner Core Inner Reflecor 100 (Inlet) 110 (Inlet ann.) 160 (Lower plenum) 170 (Outlet) 120 (SCS & Lower head) Figure 6. RELAP5-3D nodalization of a VHTR Reactor Vessel Coolant Riser Outer Reflector Outer Bypass 3 Core Inner Bypass Inner Reflector Figure 7. Plan view of RELAP5-3D VHTR nodalization showing heat transfer paths 12

13 Figure 8. FLUENT model of lower plenum: illustrating turbulence intensity; FLUENT model represents Components 160 and 170 of RELAP5-3D VHTR model see Figure 6 13