Thermal-Hydraulic Benchmarking: A Comparative Study between DASSH and CFD

Designing the core of a nuclear reactor requires a continuous interaction between neutronics and thermal-hydraulics due to their intrinsic coupling. In neutronics, the neutron fluence rate determines the distribution of fission power inside the core, which in turn affects the fuel temperature and ultimately the entire temperature field. Conversely, thermodynamic parameters from thermal-hydraulics strongly influence cross sections, impacting the neutron flux distribution. This mutual dependence necessitates iterative coupling between the two analyses to achieve consistent and physically accurate results. Accurate and reliable thermal-hydraulic codes are essential for predicting reactor behavior, ensuring effective heat removal, and maintaining safe operation. These codes provide sufficient fidelity to capture complex thermal phenomena within reactor cores while remaining computationally efficient, thereby supporting iterative reactor core design processes. Among the available tools, the Ducted Assembly Steady-State Heat Transfer Software (DASSH), developed at Argonne National Laboratory, is a finite-volume, first-order subchannel code designed to perform steady-state coolant and fuel pin temperature calculations for full reactor cores composed of hexagonal ducted assemblies. In this study, DASSH is benchmarked against Computational Fluid Dynamics (CFD) simulations. Meaningful benchmarking requires accurate representation of inter-assembly heat transfer, a phenomenon that strongly affects subchannel temperature distributions. While DASSH employs modeling assumptions to reduce computational expense, its predictive capability is carefully evaluated using CFD as a reference to quantify accuracy and uncertainties. The benchmarking highlights the current capabilities and limitations of DASSH, providing insights into its applicability and guiding potential improvements for future development.