Flow regime transition criteria for upward two-phase flow in vertical rod bundles

Abstract

The demand of accurate prediction for two-phase flow behavior in a nuclear reactor core composed of fuel rod bundles or a heat exchanger composed of heat transfer tube bundles requires comprehensive understanding of flow regime, void fraction, heat transfer and pressure drop. In comparison with the great success in developing the two-phase flow regime transition criteria for simple geometries such as pipes, annulus and rectangular channels, limited researches have been performed for developing the flow regime transition criteria of upward two-phase flow in vertical rod bundles. In the vertical rod bundles, slug bubbles spanning the bundle casing cannot exist due to their surface instability and the two-phase flow characteristics in the vertical rod bundles are different from those in pipes, annulus or rectangular channels whose channel size are smaller than the length scale of the surface instability. This study has proposed a new flow regime transition criteria model based on the analysis on the underlying physics of the upward two-phase flow behavior in the vertical rod bundles. A reliable drift-flux correlation to predict void fraction in the vertical rod bundle developed recently has been used in modeling the flow regime transition criteria. This study has classified the flow regime into 6 distinct flow regime such as bubbly, finely dispersed bubbly, cap-bubbly, cap-turbulent, churn and annular flows. The newly developed flow regime transition criteria have been compared with existing 4 flow regime maps obtained in vertical rod bundles. The fluid systems include air-water and steam-water. A fairly good agreement with some discrepancies has been obtained between the newly developed transition criteria and the measured flow regime maps.