Abstract: During the postulated loss of coolant accident of nuclear reactor, the critical flow rate at the break would determine the coolant inventory and energy loss rate, thus dominate the temperature of fuel element in the core and strongly impact the progress of transient. In order to have a better understanding of basic parameter variation law during discharge flow, the general two-fluid model was developed for initially subcooled and saturated water flow in pipe or orifice. In this model, the thermal non-equilibrium between the liquid and vapor bubbles and the interphase relative motion were taken into account. An improved correlation to calculate flashing inception location and surperheat was proposed. The bubble growth equation and governing equations were combined to simulate the bubble diameter change with the quality and the void fraction. The code predicted results agree well with the test data for a wide range of pressure and temperature, and length to diameter ratios, which indicates that the model is a generalized critical flow model.