Conference Agenda

This study enhances the understanding of thermal effects on energy distribution in the near-wake region of flow over a heated sphere by analyzing time-resolved particle image velocimetry (TR-PIV) experimental data at elevated pressures (3 MPa). The experiment spans a wide range of Reynolds numbers (19,000–29,000) and Richardson numbers (0.5–2.0), conditions characteristic of opposed flow mixed convection. Key parameters, including mean and fluctuating velocities, were calculated from the acquired velocity vector fields, with uncertainties quantified. The unique contribution of this work lies in examining the lateral and streamwise expansion of the recirculation region as heating increases, and in comparing these results with isothermal conditions. Additionally, this study isolates the effects of natural convection by comparing time-resolved turbulent kinetic energy (TKE) at the streamwise center of the recirculating region for both heated and unheated cases. Spectral analysis was conducted on the Reynolds-decomposed streamwise and spanwise velocity components using Power Spectral Density (PSD), providing insights into the turbulence characteristics within and outside the wake region. These findings are particularly relevant to the design and safety of Pebble Bed Gas-Cooled Reactors (PB-GCRs) due to the similarity in geometry and operating conditions. This work contributes to advancing the understanding of mixed convection in nuclear reactor cooling systems, offering insights into thermal-hydraulic performance under elevated pressures and varied thermal conditions.

Pronghorn is a thermal-hydraulics computational tool developed using the Idaho National Laboratory's Multiphysics Object-Oriented Simulation Environment (MOOSE). It is designed to support Computational Fluid Dynamics (CFD) modeling, ranging from subchannel and porous media analysis to Reynolds Averaged Navier-Stokes (RANS) turbulence modeling. As an integral part of the MOOSE-based suite of tools, Pronghorn seamlessly couples with other MOOSE-based applications to simulate a variety of physical phenomena. This article highlights recent significant enhancements to Pronghorn's CFD modeling capabilities and demonstrates their application to advanced nuclear reactor designs.

The recent improvements in Pronghorn primarily focus on modifications to its turbulence modeling capabilities, near-wall corrections and numerical schemes. In terms of turbulence modeling, the two-equation k-ϵ and k-ω-SST models have been implemented and validated with both equilibrium and non-equilibrium wall treatments. Regarding numerical schemes, a second-order hybrid method for gradient computation has been developed, resolving issues related to solution artifacts on skewed computational meshes commonly found in the complex advanced reactor designs. Additionally, corrections for wall roughness, and curvature, and wall-channeling in pebble beds have been introduced in the near-wall modeling.

These developments enable more accurate simulations of advanced nuclear reactors. Two case studies are presented in this work: a pool-type Molten Chloride Reactor and a salt-cooled Pebble-Bed High Temperature Reactor. In both cases, the previous models in Pronghorn are compared with the new implementations, demonstrating the improved accuracy achieved with the updated models.

OECD-NEA has launched thermal hydraulic code validation benchmark for high temperature gas-cooled reactors using the High Temperature Test Facility (HTTF) data. KAERI has joined as a participant to compare the calculated results by GAMMA+ code with the experimental data and code-to-code. Based on full power operation assumption, the steady state temperature profiles by the different codes were compared. During the initial comparison process, it showed that the calculated results of each participant were slightly different. It was thought that it could be from different nodalization and modelling approaches. The feature of the real HTTF test has prismatic blocks with fuel compact holes and coolant holes. But, the computational nodes by the each system code were simplified due to limitation of system codes as equivalent cylindrical domain. Several modelling methods to the each radial node has attempted to get more close data. Steady, PCC and DCC events were analyzed with the best acceptable method in this benchmark.

The impact of optical errors from physical sources was studied to assess their influence on PIV (Particle Image Velocimetry) measurements and 3D photogrammetry reconstructions of pebble beds. This investigation provides guidelines for optimizing these physical parameters to ensure successful optical measurements in "large" pebble beds. In this context, a large pebble bed refers to one containing more than 1,000 pebbles, with a length of at least 10 pebble diameters in each direction. Previously published studies that use similar techniques have pebble beds up to 1000 pebbles in size, but with a depth of around 7 pebbles in the narrowest direction.

Alongside the error analysis results, preliminary PIV measurements are presented, including details on calibration methods and other aspects crucial for generating CFD-grade experimental data. This data is essential for validating CFD tools like NEK-RS, which are progressively improving toward fully resolving the flow dynamics within full-scale PBR (Pebble Bed Reactor) pebble beds. Finally, results from creating a 3D reconstruction of the experimental pebble bed will be discussed. This reconstruction is both as challenging as the optical flow measurements and equally important for generating high-resolution data relevant to simulation validation.

In the reactor core of HTR-PM, due to the structural materials such as graphite blocks and carbon bricks arranged in bulk, the coolant flow paths are complicated. A part of the coolant flows through narrow gaps between the structural materials without cooling the pebble bed, which affects the temperature distribution in the reactor core. Therefore, the accurate simulation of bypass flow is a key issue related to reactor safety. The vertical gaps between side reflectors are the main bypass flow paths.In this paper, the thermal hydraulic system code was employed to simulate flows in the pebble bed and vertical gaps, analyze the flow path of coolant under different bypass flows ratio, and explored the influence on the temperature distribution of the pebble bed and the side reflectors.The numerical results are proven to be in good agreement with experimental data and results by CFD. The model reasonably simulates the bypass flow of core coolant and the temperature distribution in the core. The results also shows that there are some flow paths different from the previous researches and there is the flow direction turning point of bypass flow in vertical gaps.This method solves the problem of high computational cost when using the CFD method to study bypass flows. It is able to calculate accurately while greatly reducing computing costs, which lays a good foundation for further safety analysis under accidents.

This study conducts multiphysics modeling of graphite prismatic blocks used in High Temperature Gas- cooled Reactors (HTGRs) by analyzing the mechanical and thermal property changes induced by neutron irradiation. Graphite conducts a critical role as a moderator, reflector, and structural material in HTGRs, and these properties are significantly dependent on the level of radiation exposure in high temperature and neutron irradiation environments. The radiation-induced creep and dimensional changes have a substantial impact on the structural stability of graphite components, making their assessment essential.

Through 3D structural simulations, insights into the mechanisms of creep stress and dimensional changes occurring at elevated temperatures are provided, enhancing the understanding of how these changes affect the structural stability of graphite. The stress analysis results including this creep phenomenon are expected to be fundamental for evaluating the failure probability of the graphite prismatic block designs.

Accurate prediction and assessment of core bypass flow are vital, as they affect the heat transfer and cooling efficiency of the reactor. To address this, coupled CFD and mechanical studies considering neutron irradiation and thermal expansion have been conducted. The volume expansion with neutron irradiation dose decreases the width of the bypass gap, which increases the pressure drop but increases the heat transfer efficiency by the coolant hole.

This research is expected to contribute to the reliability evaluation of graphite components in HTGRs and provide insights for future reactor core designs and operation, enhancing the stability and efficiency of helium cooling under radiation.