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.