Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code

A thermal-hydraulic analysis of the TRIGA Mark II Research Reactor operating at 500 kW under natural convection conditions was conducted using the NCTRIGA code utilizing SRAC2006 neutronics data, and the results were benchmarked against MCNP4C calculations. The study focused on evaluating the key thermal-hydraulic parameters along the axial length of the hottest fuel rod (C4) to ensure reactor safety. The Reynolds number increased consistently with the axial height. Conversely, heat flux, heat transfer coefficient, and fuel centerline temperature demonstrated a similar trend: increasing from the top of the core, peaking near the midpoint, and subsequently decreasing. Notably, fuel centerline temperatures remained significantly below established safety limits. Fuel surface temperatures remained relatively constant, while coolant temperature demonstrated a slow, incremental increase along the axial length from the top. While minor discrepancies were observed between the SRAC2006 and MCNP4C datasets, the peak values and their locations remained consistent across both.