This thesis investigates the flow and heat transport phenomena in a pipe and Pressurized Water Reactor (PWR) rod bundle geometries using high-fidelity Direct Numerical Simulation (DNS). These geometries are essential for nuclear reactor systems, where efficient heat transfer and stable flow patterns are vital to ensure operational safety, reliability, and performance. The study focuses on evaluating the limitations of traditional thermal-hydraulic modeling approaches and advancing the understanding of flow physics in complex geometries.
Since Computational Fluid Dynamics (CFD) simulations are computationally expensive, system codes such as RELAP5, TRACER, and SPECTRA (developed by NRG) are widely used in reactor safety and design analysis. These codes rely on conventional pipe flow correlations to approximate thermal-hydraulic behavior, which are not representative of the intricate flow patterns observed in rod bundle arrangements. Such geometries are characterized by secondary vortices, gap street vortices, and the coupling of these vortices across multiple planes, forming a complex rod bundle vortex network. This study addresses these challenges by comparing the flow and thermal characteristics of pipe and subchannel geometries to evaluate the validity and limitations of pipe-based correlations.
To identify the subchannel geometry that best represents a rod bundle arrangement, square and 2×2 subchannel configurations were selected for detailed investigation. These geometries were modeled with a pitch-to-diameter (P/D) ratio of 1.3263, typical of PWR fuel assemblies, and simulated at a Bulk Reynolds number (Reb) of 5300. The square subchannel geometry represents a simplified cross-sectional domain, while the 2×2 subchannel configuration captures inter-subchannel interactions and the enhanced coupling effects seen in rod bundles. Using the Nek5000 spectral element solver, the simulations employed advanced numerical techniques, including spatial-temporal averaging through flow-through time (FTTs) metrics and further spatial averaging over unit cells, to achieve statistically converged results. Rigorous validation of pipe configuration with reference data ensured the accuracy of the computational framework.
The results highlight significant differences in turbulence structures and heat transfer performance between the pipe, square subchannel, and 2×2 subchannel configurations. While pipe correlations provide a baseline for comparison, they fail to capture the complex flow interactions observed in rod bundles. The 2×2 subchannel geometry emerges as a more accurate representation of rod bundle dynamics due to its ability to simulate enhanced inter-subchannel mixing and vortex coupling. These findings emphasize the importance of geometry-specific modeling in accurately predicting thermal-hydraulic behavior in nuclear reactors.
This work bridges the gap between simplified system codes and detailed physics-based modeling of rod bundle flows. The insights gained from this study provide a foundation for improving thermal-hydraulic predictions and advancing the design and safety of nuclear reactor systems. Future research will extend these findings to explore higher Reynolds numbers, diverse Prandtl numbers, wall effects, and alternative rod bundle configurations, further contributing to the development of advanced engineering tools for nuclear applications.