Large-Scale Numerical Simulations of Multiphase Flow

Introduction Although subchannel codes [1--3] are used for the thermal -hydraulic analysis of fuel bundles in nuclear reactors from the former, lots of composition equations and empirical correlations based on experimental results are needed to predict the water-vapor two-phase flow behavior. When there are no experimental data such as the reduced-moderation light water reactor (RMWR) [4, 5] which is currently developed by the Japan Atomic Energy Research Institute, therefore, it is very difficult to obtain highly precise predictions. The RMWR core has remarkably narrow gap spacing between fuel rods (i.e., around 1 mm) which are arranged at a triangular tight-lattice configuration in order to reduce the moderation of the neutron. In such a tight-lattice core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the two-phase fluid flow characteristics. Then, the authors tried to analyze the Japan Atomic Energy Research Institute ha