Flow boiling in the rod bundle channel is an important phenomenon in nuclear reactors. The local inhomogeneous void fraction distribution in the flow channel will affect the flow and heat transfer in two phase flow. A 5 × 5 rod bundle test section is designed and utilized in the experiments and a conductivity probe is utilized to measure the local void fraction of two phase flow. A remote-control stepping driving assembly is designed to drive the void probe. The local void fraction characteristics at the outlet of a 5 × 5 rod bundle channel during subcooled boiling and saturated boiling are experimentally studied when the system pressure is 2∼3 MPa. The outlet subcooling ranges from 0 °C to 18 °C. The mass flux is 2000 kg/ms and the heat flux ranges from 186.3 kW/m to 320.8 kW/m.The effects of thermal parameters on local phase interfacial parameters in subcooled boiling and saturated boiling are obtained. The research results are important for revealing the two phase flow mechanisms in rod bundle channels.

中文翻译：

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5×5棒束通道内流动沸腾空隙率特性的实验研究

棒束通道内的流动沸腾是核反应堆中的一个重要现象。流道内局部不均匀的空隙率分布会影响两相流的流动和传热。实验中设计并采用了5×5棒束测试段，并利用电导探头测量两相流的局部空隙率。设计了遥控步进驱动组件来驱动空洞探针。实验研究了系统压力为2∼3 MPa时5×5棒束通道出口过冷沸腾和饱和沸腾过程中的局部空隙率特征。出口过冷度范围为0°C至18°C。质量通量为2000 kg/ms，热通量范围为186.3 kW/m~320.8 kW/m。得到了过冷沸腾和饱和沸腾过程中热参数对局部相界面参数的影响。研究成果对于揭示杆束通道内两相流机理具有重要意义。