[en] The COBRA/TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients. The code solves the compressible three-dimensional, two-fluid, three-field equations for two-phase flow in the reactor vessel. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. A five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, and accumulator. In the code modeling of Semiscale Test S-UT-2, the intact and broken loops, and the upper head injection (UHI) systems are represented by one-dimensional components,. The pressure vessel and two steam generators are modeled using the three-dimensional VESSEL component. The results from the COBRA/TRAC calculation give a reasonable match with the measured data