Evaluation of a critical heat flux is one of the most important issues for design of an advanced water-cooled reactor core. Since it becomes difficult to perform full-scale experiments due to a larger scale of the advanced reactor cores, an analytical approach has been widely noticed in the core design. To predict the critical heat flux in high accuracy, it is required to correctly understand a horizontal distribution of a two-phase flow in the rod bundles. In this study, the two-phase flow characteristics through narrow gaps in the tight-lattice 37-rod bundle experiment at JAERI were investigated using the subchannel analysis code, NASCA. At the center of the bundle, liquid flowed toward the periphery due to the diversion cross-flow at the elevation where boiling started and the turbulent mixing and the void drift were not influential as they can be neglected. On the periphery of the bundle, the flow mixings due to the diversion cross flow, turbulent mixing and void drift were almost the same order. Gas flowed in the same way with the liquid phase due to the diversion cross-flow, and the turbulent mixing and the void drift moved the gas in the opposite way of the liquid phase migration. An amount of the diversion cross-flow for the liquid phase increased in proportion to the square of the mass velocity. The characteristics of cross flow were almost the same in the different local power peaking and in the different gap widths in the present model.

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