Numerical analysis of thermal-hydraulic characteristics of the whole LFR core under blockage conditions

Blockage accidents are critical scenarios in the design and safety analysis of lead-bismuth cooled fast reactor (LFR) core. Traditional analysis of blockage accidents in LFR focuses on localized, fine-scale computational fluid dynamics (CFD) simulations of single or three assemblies, but the analysis of the whole core scale impact caused by blockage accidents is insufficient. Therefore, this paper uses CorTAF-LBE, a three-dimensional thermal-hydraulic analysis code developed by XJTU-NuTHeL, to analyze the impact of blockage accidents on the whole core of the LFR. The reliability of the code in calculating thermal-hydraulic parameters under blockage accidents was validated based on the KALLA-THEADES and KALLA-IWF experiments. Taking the MYRRHA-FASTEF core as the object, simulations and analyses are conducted for various blockage scenarios with different lengths and positions. The results indicate that blockage accidents lead to an enlarged coolant temperature gradient at the core outlet. Lengthening the blockage results in an elevation of the peak temperature in the cladding. Under 2.06 % blockage at the center of the assembly, blockage in the middle of the heating segment poses the greatest threat to cladding integrity, with the maximum temperature reaching 1336.9K, an increase of 635.4K compared to normal operating conditions. Under 4.59 % blockage at the edge of the assembly, the maximum cladding temperature reaches 1381.8K, and the heat transfer rate of the inter-wrapper flow (IWF) adjacent to the blockage area is 24.2 % higher than under normal operation. Additionally, severe degradation in heat transfer downstream was not observed in several simulated blockage scenarios.