Nuclear Technology / Volume 152 / Number 1 / October 2005 / Pages 23-37
Technical Paper / Nuclear Reactor Thermal Hydraulics
To improve a void drift model used in a subchannel analysis, new experimental data are obtained for air-water two-phase flows in a vertical 2 × 3 rod channel consisting of six subchannels simulating a square array boiling water reactor fuel rod bundle. The data include the axial redistributions of flow rates of both phases and void fraction in the respective subchannels. By fitting the above data with the Lahey and Moody void settling model, we have determined a void diffusion coefficient in their model. It is found that the void diffusion coefficient for slug, churn, and annular flows could be well correlated in terms of a turbulent Peclet number developed in our previous study. Furthermore, a subchannel analysis code based on a two-fluid model proposed in our previous study is examined against the present data. In the code, the void settling model is incorporated with usual conservation equations of mass and momentum. From the examination, it is found that the subchannel analysis code can predict well the data on subchannel flow and void fraction for the 2 × 3 rod channel if appropriate correlations are adopted to evaluate wall and interfacial friction forces needed in the two-fluid model.