Subchannel analysis is widely used in nuclear reactor core thermal-hydraulic calculation and safety analysis. In subchannel analysis, the axial flow is usually treated as the dominant one-dimensional flow, and the lateral flow is simplified as the intersubchannel interactions and is modeled by introduction of semiempirical source terms or separate models. The accuracy of the subchannel analysis is strongly dependent on the modeling of intersubchannel interactions between adjacent subchannels. The intersubchannel interaction can be decomposed into three components: diversion cross flow that occurs due to imposed transverse pressure gradients, turbulent mixing that occurs due to stochastic pressure and flow fluctuations, and void drift that results from lateral migration of the gaseous phase (void) due to a strong tendency of the two-phase system approaching the equilibrium state of phase distribution. This critical review focuses on void drift research. Both experimental observation of the void drift phenomenon and the proposed void drift models are reviewed. The improvements and corresponding assessments of the void drift models are summarized. Following that, further improvements on the void drift model are proposed.