Three-Dimensional Large-Scale Bubbly Flow Analysis in a Vertical Minichannel

In light water reactors each fuel rod is arranged in the shape of a square lattice with an interval of around 3 mm. Several spacers are installed on the surface of the fuel rod with arbitrary axial positions. Water flows vertically along fuel rods and is heated by those, and then many bubbles generate. In order to improve the thermal design procedure of the nuclear reactor core, it is needed to clarify velocity, pressure, temperature and void fraction distributions precisely based on the bubbly flow behavior in a vertical minichannel under the water-vapor two-phase flow condition. However, it is not easy to get those three-dimensional distributions by the experimental study. Then, large-scale two-phase flow simulations were performed to predict the three-dimensional bubby flow configurations in the simply simulated nuclear coolant channel. Regarding both mechanisms of the coalescence and fragmentation of bubbles the useful knowledge was obtained.