Evaluation of Pb-Bi-Water Direct Contact Boiling Flow in a Vertical Circular Channel using Drift-Flux Model

Novitrian Novitrian (Nuclear Physics and Biophysics Research Division, Institut Teknologi Bandung)
Minoru Takahashi (Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo 152-8550, JAPAN)

Indonesian Journal of Physics (IJP)
Vol 19 No 4 (2008): Vol. 19 No. 4, October 2008

The characteristics of Pb-Bi-water boiling two-phase flow were investigated in a vertical circular channel. Water was injected into Pb-Bi to circulate Pb-Bi in a loop with an effect of steam lift pump. Pb-Bi flow rates were measured and analytically estimated using balance condition between buoyancy force and pressure losses. The drift-flux model was used to determine void fraction with assumption that bubble sizes were dependent on the fluid surface tension and density ratio of Pb-Bi to steam and water mixture. The result of analysis shows that the deviation from the experimental data was less than 20%.