In‐Depth Numerical Elucidation on Thermal‐Hydraulic Phenomena for EBR‐II XX09 Subassembly During SHRT‐45 Experiment by 1D MARS‐LMR Code and 3D CFD Code

In this study, the EBR‐II SHRT‐45 test, simulating a loss‐of‐flow accident without scram, was employed to evaluate the stability and thermal‐hydraulic behavior of the reactor’s fuel assembly cladding. By integrating high‐fidelity computational fluid dynamics (CFD) models with the MARS‐LMR code, this research aims to validate predictive capabilities and assess reactor safety under severe accident conditions. The reliability of the MARS‐LMR code in predicting thermal‐hydraulic behavior for both steady and transient states was demonstrated. The code consistently provided conservative predictions compared to experimental data, ensuring robust safety margins. CFD simulations offered detailed insights into the reactor’s complex heat and flow dynamics, accurately capturing temperature distributions and flow patterns at various locations. Hot channel factors (HCF) were applied to evaluate the cladding temperature under maximum expected conditions. Results indicated that the peak cladding temperature, when considering HCF, remained below the safety limit of 1025°C, thereby ensuring a safety margin of 26%. Furthermore, CFD analysis underscored the importance of accounting for circumferential temperature variations in the cladding, emphasizing the necessity of HCF in comprehensive safety evaluations. This study highlights the importance of integrating MARS‐LMR and CFD simulations for accurate and reliable safety assessments in nuclear reactor operations. The findings provide crucial insights and methodologies to enhance future reactor safety assessments and performance improvements, ensuring that reactors operate within safe limits under various operational and accident scenarios.