Numerical analysis of complicated two-phase flow behavior in nuclear reactor cores

Takase, Kazuyuki; Yoshida, Hiroyuki; Tamai, Hidesada; Ose, Yasuo*; Aoki, Takayuki*; Xu, Z.*

Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.2, p.39 - 40, 2006/09

Three-dimensional large-scale numerical simulations were carried out to predict the complicated water-vapor two-phase flow characteristics in a fuel bundle of an advanced light water reactor. Conventional analysis methods with a two-fluid model need composition equations and empirical correlations based on the experimental data. Therefore, it is difficult to obtain high prediction accuracy when experimental data are nothing. Then, a new two-phase flow analysis method was proposed and the TPFIT code was developed. This paper describes the predicted liquid film, bubbly and droplet flow behavior in the simulated fuel channels with the TPFIT code, and the predicted two-phase flow behavior around a curved fuel rod with the FLUENT code which is one of the most famous commercial code. From the present results, the high prospect was acquired on the possibility of development of the thermal design procedure of the advanced nuclear reactors by large-scale simulations.