Axial and radial void fraction measurements in convective boiling flows

- Axial and radial void fraction measurement for vertical forced convective boiling flow.
- γ- ray densitometry has been used for measuring void fraction distribution profiles.
- Drift flux model description used to rationalize the experimental observations.

This work is inspired by the need to have measurements and predictive capability on vertical saturated boiling flows, which are of importance in boiling water nuclear reactors. In these systems, sub-cooled water flows upwards in vertical tubes which contain a multiplicity of nuclear fuel rods, and the heat is taken away from these rods by natural convection boiling. A close analog of this situation exists when the liquid is in forced flow, and the heat flux from the vertical rods is quenched by forced convection boiling of the upward liquid water flux.In this work, we present experimental data has for axial, and radial vapor void fraction distributions in an annular boiling channel for low mass-flux forced the flow of water at high inlet sub-cooling. A single centrally placed electrical rod, designed to mimic the nuclear fuel rod has been used. The void fraction measurement is made using gamma ray densitometry technique. Axial and radial vapor void fractions have been reported, as a function of inlet liquid flux and inlet liquid temperature. The experimental data has been rationalized using a simple one-dimensional drift flux model adapted to the conditions of the experiment.