Full-range drift-flux correlation for boiling and adiabatic co-current downward two-phase flows in vertical pipes

Abstract

Accurately predicting void fraction (VF) for boiling and adiabatic two-phase flows is important in many engineering fields, especially in nuclear thermal-hydraulic analysis. Typical one-dimensional computational codes for nuclear thermal-hydraulic analyses have been developed using the two-fluid model, in which the interfacial drag force is formulated using the drift-flux model. However, because the buoyancy force and the flow direction act in opposite directions, the drift-flux correlations for boiling and adiabatic vertical downward co-current two-phase flows are limited or entirely empirically derived from specific databases. Therefore, this paper developed a “full-range drift-flux correlation” for vertical downward gas-liquid flows, grounded in a scientific philosophy to minimize empiricism. First, the critical area-averaged VF was determined by considering the similarity between the distributions of VF and the mixture volumetric flux. The constitutive equations of the distribution parameter and drift velocity were analytically developed for pure dispersed and beyond-dispersed flows. To validate the new drift-flux correlation, an extensive literature survey was conducted, collecting 506 VF data from 9 sources published from 1962 to present. The mean relative deviation, $m_{\mathrm{rel}}$, and mean absolute relative deviation, $m_{\mathrm{rel},\mathrm{ab}}$, between the experimental and calculated adiabatic and boiling VFs were 3.07% and 8.93%, respectively, validating the accuracy of the proposed full-range drift-flux correlation.