[Background] Lead-cooled fast reactor (LFR) is one of six candidates for Generation IV nuclear systems. Although lead-bismuth eutectic (LBE) has excellent neutron properties, thermal-hydraulic performance and inherent safety features, the structure and cladding materials could be eroded by LBE coolant. The erosion products could accumulate in fuel assembly and grow as a blockage influencing the heat transfer of cladding and coolant. [Purpose] This study aims to investigate the block parameter effect on thermal-hydraulic performance of single fuel assembly in lead-bismuth subcritical reactor under various blockage accidents. [Methods] Based on actual geometry and material properties, modeling of single fuel assembly was established. Five blockage accidents were numerically simulated by the commercial computational fluid dynamics (CFD) software STAR-CCM+ with different block parameters. [Results] Temperature variations along the axial direction of cladding inner face and subchannel center were obtained and compared. Axial velocity distributions around block were analyzed. Influence of block parameters for the coolant backflow and the property of flow fields were achieved. [Conclusion] The maximum temperature rise unanimously occurs inside block. Different block parameters diversely influence flow and heat transfer.%铅铋冷却快堆是第四代核能系统之一,其具有许多运行与安全性优势.但铅铋冷却快堆在运行过程中,堆芯结构材料会受到铅铋合金冷却剂的腐蚀作用,腐蚀产物在堆内堆积可能会引发堵流事故,从而导致包壳传热恶化,并影响冷却剂的流动传热效果.通过对铅铋冷却快堆单盒燃料组件建模,使用商用计算流体力学软件STAR-CCM+对不同堵块参数下的5个堵流事故工况开展了计算分析.通过对事故后包壳内壁面温度、子通道中心温度的轴向发展和堵块周围流场的轴向速度分布进行对比分析,获得了各种堵块参数对堵流事故后传热恶化、流场性质的不同影响规律.
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