The thermohydraulics of a nuclear reactor during the reflood phase of a hypothetical loss-of-coolant accident can be represented by moving control volume methodology in which six control volumes are used to represent the downcomer, lower plenum, and reactor core. The one-dimensional, homogeneous, equilibrium constitutive equations for two-phase steam/water flow are solved in each control volume and connecting junctions. One of the three core control volumes represents the quench region; it changes size and position based on the axial location of the clad quench temperature and the condensed liquid level in the flow channel. The lengths of the remaining two core control volumes are determined by the position of the quench region. Simulation of actual reflood experiments demonstrates that the methodology predicts reflood-like flow oscillations and reproduces the correct trends in experimental data. The moving control volume methodology has proven itself as a valid concept for reflood hydrodynamics, but further development of the existing EFLOD code is required for simulation of actual reflood experiments.