Abstract
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/m2s and the heat flux ranges from 186.3 kW/m2 to 320.8 kW/m2.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.
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