Micro-channel heat exchangers consist of a number of plates, each containing fluid channels etched in the surface and diffusion bonded together to create a porous core of metal. The primary and secondary sides of the exchanger are formed by connecting the channels on alternating plates to the respective leader pipes. To analyze the thermal response of exchangers during operation, simulation software is used to create a network of numerical models representing the real-life thermal-hydraulics components. The Systems CFD approach uses one-dimensional empirical models for the fluid flow inside the channels and a three-dimensional model for the heat distribution inside the core. Spatial analysis of the geometry gives a connectivity stencil between the one- and three-dimensional models. This stencil implicitly links the equations of the models at matrix level in the numerical solver, with faster convergence in fewer iterations than when the models are coupled explicitly in different software applications. Results presented show the heat flux through an exchanger core and the fluid flow inside the channels.

This CFD study is to simulate a coolant (gas) flow and heat transfer in a PBR core during a normal operation. This study used a pebble array with direct area contacts among the pebbles which is one of the pebbles arrangements for a detailed simulation of PBR core CFD studies. A CFD model is developed to more adequately represent the pebbles randomly stacked in the PBR core. The CFD predictions showed a large variation of the temperature on the pebble surface as well as in the pebble core. The temperature drop in the outer graphite layer is smaller than that in the pebble-core region. This is because the thermal conductivity of graphite is higher than the fuel (UO 2 mixture) conductivity in the pebble core. Higher pebble surface temperature is predicted downstream of the pebble contact due to a reverse flow. Multiple vortices are predicted to occur downstream of the spherical pebbles due to a flow separation. The coolant flow structure and fuel temperature in the PBR core appears to largely depend on the in-core distribution of the pebbles.

Data from a scaled model of a portion of the lower plenum of the helium-cooled very high temperature reactor (VHTR) are under consideration for acceptance as a computational fluid dynamics (CFD) validation data set or standard problem. A CFD analysis will help determine if the scaled model is a suitable geometry for validation data. The present article describes the development of an analysis plan for the CFD model. The plan examines the boundary conditions that should be used, the extent of the computational domain that should be included and which turbulence models need not be examined against the data. Calculations are made for a closely related 2D geometry to address these issues. It was found that a CFD model that includes only the inside of the scaled model in its computational domain is adequate for CFD calculations. The realizable k∼ε model was found not to be suitable for this problem because it did not predict vortex-shedding.

In this study several air and water ingress scenarios for the PBMR [1] were simulated by means of the dynamic reactor code TINTE ( TI me-dependent N eutronics and TE mperatures) [2]. The Power Conversion Unit (PCU) and other sub-systems cannot be modelled with the TINTE code and therefore air ingress rates were obtained from Computational Fluid Dynamics (CFD) analysis performed by utilizing the FLUENT code [3]. The use of the TINTE code was previously validated with simulations of the NACOK corrosion experiments [4], [5], [6]. The validations were performed at Forschungszentrum Juelich, however the results are not yet published. The rates of chemical reactions between graphite and gases like O 2 , CO 2 , H 2 O and H 2 are negligible below 400°C. Air and water ingress into the PBMR core at high temperatures can result in corrosion of the PBMR fuel spheres and a possible increase in the fission product release rate. The air ingress scenarios included in this study are; a break in the core outlet pipe at the turbine inlet location, which results in air ingress from the outlet plenum, and a break in a pipe that is connected to the top of the Reactor Pressure Vessel (RPV), which results in air ingress from top of the core. For both transients it is assumed that a Depressurized Loss of Forced Cooling (DLOFC) event takes place prior to the air ingress. The DLOFC leads to high fuel and reflector temperatures that allow higher oxidation rates. The results show that the oxidation of graphite structures in the core is more severe in the case of the outlet pipe break transient. A break in the Core Conditioning System (CCS) heat exchanger circuit during a maintenance mode or following a reactor trip could result in water ingress of up to 1000 kg into the core (the primary system is depressurized at this stage). During the water ingress the CCS continuously cools down the core. Due to the low water ingress rate and lower fuel temperatures, the water ingress transient is not as severe as the air ingress transients.

This paper presents a model developed for determining fuel particle and fuel pebble temperatures in normal operation and transient conditions based on multi-scale modelling techniques. This model is qualified by comparison with an analytical solution in a one-dimensional linear steady state test problem. Comparison is made with finite element simulations of an idealised “two-dimensional” pebble in transient conditions and with a steady state analytical solution in a spherical pebble geometry. A method is presented for determining the fuel temperatures in the individual batches of a multi-batch recycle refuelling regime. Implementation of the multi-scale and multibatch fuel models in a whole-core CFD model is discussed together with the future intentions of the research programme.

With recent progress in high-temperature pebble-bed reactor programs research focus has started to include more ancillary engineering issues. One very important aspect for the realisability is the mixing of hot and colder helium in the reactor lower plenum. Under nominal operating conditions, depending on core design, the temperature of hot gas leaving the core can locally differ up to 210° C. Due to material limitations, these temperature differences have to be reduced to at least ±15° C. Several reduced-size air experiments have been performed on this problem, but their applicability to modern commercially sized reactors is not certain. With the rise in computing power CFD simulations can be performed in addition, but advanced turbulence modeling is necessary due to the highly swirling and turbulent nature of this flow. The presented work uses the geometry of the German HTR-Modul which consists of an annular mixing channel and radially arranged ribs. Using the commercial CFD code ANSYS CFX, we have made detailed analyses of the complex 3D vortical flow phenomena within this geometry. Several momentum transport turbulence models, e.g. the classical k-e model, advanced two-equation models and Reynolds-Stress Models were compared with respect to their accuracy for this particular flow. In addition, the full set of turbulent scalar flux transport equations was implemented for modeling the three components of turbulent transport of enthalpy seperately and were compared with the standard turbulent Prandtl number approach. As expected from previous work in related fields of turbulence modeling, the differences in predicting the mixing performance between models were significant. Only the full Reynolds-Stress model coupled with the scalar flux equations was able to reproduce the experimentally observed reduction of mixing efficiency with increasing Reynolds number. The correct scaling of mixing efficiencies demonstrates that the utilized turbulence models are able to reproduce the physics of the underlying flow. Hence they could be employed for the scaling and optimization of the lower plenum geometry. The results also showed that the original geometry used for the HTR-Modul is insufficient to provide adequate mixing, and that hence a not sufficiently mixed coolant for future reactor designs might be an issue. Based on this work, an optimization for future lower plenum geometries has become feasible.

TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time.

This paper describes the design process followed by Westinghouse Electric South Africa for the insertion of hot internals into the Reactivity Control System (RCS) and Reserve Shutdown System (RSS) Units Under Test (UUTs) at the Helium Test Facility (HTF) at Pelindaba, South Africa. The aim of the UUTs is to allow the validation of the high temperature operation of the RCS and RSS systems for implementation into the proposed Demonstration Power Plant of the PBMR. The units use electrical heaters to obtain pebble-bed reactor thermal conditions for both the control rods and small absorber spheres (SAS) under a pressurized helium environment. Design challenges include providing for strength under elevated temperatures (900°C maximum); pressure boundary integrity (9MPa maximum); separation of different volumes (representing core barrel, reactor citadel and other Reactor Pressure Vessel (RPV) volumes); thermal protection of carbon steel vessels by using thermal insulation; allowing for diverse thermal expansion coefficients of different materials; allowing for depressurization events within the insulation and internals; having access for temperature, pressure, stress and proximity sensors and electrical wiring through high pressure penetrations; and provision for assembly of the hot internals both on and off-site. Several thermal analyses using Computational Fluid Dynamics (CFD) were performed to evaluate both worst-case and operational conditions of the UUTs. Factors that were considered include thermal insulation properties, heat transfer modes (internal radiation, external radiation and natural convection, forced internal convection for cooling) and operating pressure (ranging from 1 to 9MPa). The thermal design uses elements originally proposed for hot gas duct design. The results obtained show that the proposed design satisfies ASME VIII requirements of the pressure boundary and that all challenges are successfully met.

This paper describes pressure drop and heat transfer coefficient predictions for a typical coolant flow within the core of a pebble bed reactor (PBR) by examining a representative group of pebbles remote from the reflector region. The three-dimensional steady state flow and heat transfer predictions utilized in this work are obtained from a computational fluid dynamics (CFD) model created in the commercial software ANSYS FLUENT™. This work utilizes three RANS turbulence models and the Chilton-Colburn analogy for heat transfer. A methodology is included in this paper for creating a quality unstructured mesh with prismatic surface layers on a random arrangement of touching pebbles. The results of the model are validated by comparing them with the correlations of the German KTA rules for a PBR.

An area that has been identified as important in the development of High Temperature Reactors (HTRs) is the prediction of leakage and bypass flows through such a reactor. It is therefore essential to understand the causes of bypass flows and to determine the effect on the predicted fuel and component temperatures. This paper discusses the identification of leakage flows that are applicable to the PBMR design and the ranking of these leakage flows. The modeling methodology and results are also discussed. Similar to previous HTR’s, it was found that leakage and bypass flows are important to the operation of the PBMR. However, through a focused approach, it is shown that PBMR is able to improve the understanding of this phenomenon, to quantify the flows and the subsequent influence on the operation of the system. This has resulted in a reduction of leakage and bypass from approximately 46% to 20%. The improved understanding of leakage and bypass flows empowers PBMR to address this issue already during the design phase of the project, which subsequently results in a vast improvement over historical HTR designs. This gives PBMR a distinct advantage over previous High Temperature Reactors.

The goal of the work is to investigate the abilities of the realizable k-ε and the k-ω turbulence models to predict phenomena expected in the lower plenum of the prismatic gas cooled reactor by benchmarking them against one another as well as verifying them against available experimental data [6]. Simulations were performed with CD-Adapco’s Star-CCM+ computational fluid dynamics (CFD) code utilizing the steady state approximation of the realizable k-ε turbulence model and the unsteady RANS approximations of the shear stress transport (SST) k-ω turbulence model. The unsteady results were averaged over a period of time corresponding to one full fluctuation cycle of the phenomena present with the lowest frequency. A case with a single jet of Reynolds number 12,700 was simulated as well as a case with dual jets of Reynolds numbers 6,300 and 12,700. Impingement of the jets occurred on the lower plane of the setup as occurs in the lower plenum. A two-layer shear driven [8] y+ wall treatment was used to satisfy the boundary layer profile. A mesh of 8.6 million polyhedral cells was generated to capture critical flow characteristics within the domain. Polyhedral cells were chosen for this application because they provided a better quality mesh and reduced the total number of cells necessary to achieve accurate results [4]. The simulations carried out were defined as having reached a converged solution when all residuals reduced to less than 10 −5 . The simulations of the flow in the rod bundle were successful in providing insight into locations of some key recirculation zones and the dependence they have on inlet conditions. The comparison between numerical and experimental results showed similar key flow patterns as well as aided in possible points of focus for future investigation. The information obtained and conclusions drawn will be critical in future numerical benchmarks in this area of research.

The AVR (Arbeitsgemeinschaft Versuchsreaktor) is a pebble bed type helium cooled graphite moderated high temperature reactor that operated in Germany for 21 years and was closed down in December 1988 [1]. The AVR melt-wire experiments [2], where graphite spheres with melt-wires of different melting temperatures were introduced into the core, indicate that measured pebble temperatures significantly exceeded temperatures calculated with the models used at the time [3]. These discrepancies are often attributed to the special design features of the AVR, in particular the control rod noses protruding into the core, and to inherent features of the pebble bed reactor. In order to reduce the uncertainty in design and safety calculations the PBMR Company is re-evaluating the AVR melt-wire experiments with updated models and tools. 3-D neutronics thermal-hydraulics analyses are performed utilizing a coupled VSOP99-STAR-CD calculation. In the coupled system VSOP99 [4] provides power profiles on a geometrical mesh to STAR-CD [5] while STAR-CD provides the fuel, moderator and solid structure temperatures to VSOP99. The different fuel histories and flow variations can be modelled with VSOP99 (although this is not yet included in the model) while the computational fluid dynamics (CFD) code, STAR-CD, adds higher-order thermal and gas flow modelling capabilities. This coupling therefore ensures that the correct thermal feedback to the neutronics is included. Of the many possible explanations for the higher-than-expected melt-wire temperatures, flow bypassing the pebble core was identified as potentially the largest contributor and was thus selected as the first topic to study. This paper reports the bounding effects of bypass flows on the gas temperatures in the top of the reactor. It also presents preliminary comparisons between measured temperatures above the core ceiling structure and calculated temperatures. Results to date confirm the importance of correctly modelling the bypass flows. Plans on future model improvements and other effects to be studied with the coupled VSOP99-STAR-CD tool are also included.

Very High Temperature Reactors (VHTRs) require high temperature (900–950 °C), high integrity, and high efficiency heat exchangers during normal and off-normal conditions. A class of compact heat exchangers, namely, the Printed Circuit Heat Exchangers (PCHEs), made of high temperature materials, found to have the above characteristics, are being increasingly pursued for heavy duty applications. A high-temperature helium experimental test facility, primarily aimed at investigating the heat transfer and pressure drop characteristics of the PCHEs, was designed and is being built at the Ohio State University. The test facility was designed for a maximum operating temperature and pressure of 900 °C and 3 MPa, respectively. Owing to the high operating conditions, a detailed investigation on various high temperature materials was carried out to aid in the design of the test facility and the heat exchangers. The study showed that alloy 617 is the leading candidate material for high temperature heat exchangers. Two PCHEs, each having 10 hot and 10 cold plates with 12 channels in each plate, are currently being fabricated from alloy 617 plates and will be tested once the test facility is constructed. To supplement the experiments, computational fluid dynamics modeling of a simplified PCHE model is being performed and the results for three flow rate cases of 15, 40, and 90 kg/h and a system pressure of 3 MPa are discussed. In summary, this paper focuses on the study of the high-temperature materials, the design of the helium test facility, the design and fabrication of the PCHEs, and the computational modeling of a simplified PCHE model.

This paper describes the experimental validation of a proposed method that uses a small amount of helium injection to prevent the onset of natural circulation in high temperature gas reactors (HTGR) following a depressurized loss of coolant accident. If this technique can be shown to work, air ingress accidents can be mitigated. A study by Dr. Xing L. Yan et al. (2008) developed an analytical estimate for the minimum injection rate (MIR) of helium required to prevent natural circulation. Yan’s study used a benchmarked CFD model of a prismatic core reactor to show that this method of helium injection would impede natural circulation. The current study involved the design and construction of an experimental apparatus in conjunction with a CFD model to validate Yan’s method. Based on the computational model, a physical experimental model was built and tested to simulate the main coolant pipe rupture of a Pebble Bed Reactor (PBR), a specific type of HTGR. The experimental apparatus consisted of a five foot tall, 2 inch diameter, copper U-tube placed atop a 55-gallon barrel to reduce sensor noise from outside air movement. Hot and cold legs were simulated to reflect the typical natural circulation conditions expected in reactor systems. FLUENT was used to predict the diffusion and circulation phases. Several experimental trials were run with and without helium injection. Results showed that with minimal helium injection, the onset of natural circulation was prevented which suggests that such a method may be useful in the design of high temperature gas reactors to mitigate air ingress accidents.

A low decay heat (implying Spent Fuel (SF) pebbles older than 8–9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks’ vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading / unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell.

The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include the mean velocity field in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow — with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows obtaining improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal developing, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet velocity profiles is also presented.

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.

The nuclear core of High Temperature Gas Reactor (HTGR) with pebble bed type has been investigated intensively due to its benefits in management, but its complicated flow geometry requested the reliable analytical method. Recent studies have been made using the three dimensional computational methods but they need to be evaluated with the experimental data. Due to the complicated and narrow flow channel, the intrusive methods of flow measurement are not proper in the study. In the present study, we developed a wind tunnel for the pebble bed geometry in the structure of Face Centered Cubic (FCC) and measure the flow field using the Particle Image Velocimetry (PIV) directly. Due to the limitation of the image harnessing speed and accessibility of the light for particle identification, the system is scaled up to reduce the mean flow velocity by keeping the same Reynolds number of the HTGR. The velocity fields are successfully determined to identify the stagnation points suspected to produce hot spots on the surface of the pebble. It is expected that the present data is useful to evaluate the three dimensional Computational Fluid Dynamics (CFD) analysis. Furthermore, It would provide an insight of experimental method if the present results are compared by those of scaled down and liquid medium.