Analysis of full-scale thermal-hydraulic characteristics of sodium-cooled fast reactor core under steady-state and accident scenarios

Based on the open-source CFD platform OpenFOAM, CorTAF-SFR has been developed to analyze the 3D thermal-hydraulic characteristics of sodium-cooled fast reactor (SFR) fuel rod assemblies using the finite volume method. The code has been validated against the ORNL-FFM2A and SCARLET-II experiments, demonstrating its accuracy in predicting the thermal-hydraulic behavior of fuel rod assemblies under both steady-state and blockage conditions. The tool was further applied to analyze the thermal-hydraulic performance of the China Experimental Fast Reactor (CEFR) under steady-state operation and accident scenarios. Under steady-state conditions, the average coolant outlet temperature deviation from the design values was within 2.0 K, with significant temperature drops observed at the component interface regions. During an overpower accident, peak temperatures of the coolant, cladding surface, and fuel pellet reached 1028.4 K, 1030.6 K, and 1598.9 K, respectively. In the blockage accident, the temperature of the blocked area increased significantly, and the coolant flow rate at about 150 mm downstream of the blocked area returned to the original level. Detailed analysis revealed the thermal-hydraulic behavior changes during the overpower scenario and the mechanisms of flow and temperature field alterations in blocked regions. These findings are crucial for advancing thermal-hydraulic analysis methods for SFR cores and ensuring reactor safety and performance.