The symmetry of coolant flow within rod bundle channels is disrupted by the presence of spacer wires, leading to the generation of secondary flows and vortex structures in the inter-rod gap regions, and consequently, turbulent mixing behavior is significantly altered. More uniform temperature distribution in the flow field is promoted, and the risk of hot spot formation is reduced through the enhanced turbulent mixing, which is crucial for the safe operation of nuclear reactors. Computational fluid dynamics (CFD) was employed to simulate and analyze the turbulent mixing characteristics between different subchannels in a 7-rod bundle unit with 4 rectangular helical wires. The results indicate that the turbulent mixing coefficient at the interfaces of the subchannels exhibits periodic variation along the axial direction, with a cycle length equal to 1/4 of the wire pitch. The minimum value of the coefficient is observed at the wire contact points, and it gradually increases with distance from the wires. Furthermore, the turbulent mixing coefficients at the edge-center and center-center subchannel interfaces are found to be significantly higher than those at the edge-corner subchannel interfaces. Based on the numerical simulation results, an empirical correlation for the turbulent mixing coefficient under different Reynolds numbers and pitch-to-diameter ratios is derived.
核反应堆作为现代能源体系的重要组成部分,其安全性和热工水力性能一直是工程与科研领域的重点研究方向。在中国“热堆-快堆-聚变堆”三步走核能发展路线中,钠冷快堆作为实现核燃料闭式循环、提高铀资源利用率的关键堆型,其燃料组件设计技术已成为当前研发重点。中国首座钠冷快堆——中国实验快堆(China Experimental Fast Reactor,CEFR)已实现满功率运行并积累十余年运行经验,示范快堆CFR600正处于建设阶段。在这些先进堆型中,燃料组件采用绕丝定位结构[1],通过螺旋绕丝诱导产生的二次流强化冷却剂间的湍流交混,以降低堆芯温度梯度、避免局部热点[2]。对于堆芯燃料组件存在的大规模棒束结构的热工水力数值模拟,为简化计算同时兼顾一定的计算精度,常采用子通道[3-5]方法进行数值模拟。子通道方法通过将复杂的棒束流场划分为若干子通道,并基于质量、动量和能量守恒方程计算各子通道间的交混特性,能够高效预测冷却剂的温度与流速分布。然而,子通道分析通常依赖于经验性的交混系数,因此,如何获取绕丝棒束结构的湍流交混参数,成为子通道分析的关键问题之一。
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