A new approach to two-phase flow analysis in a rod bundle

Highlights: ► A new mathematical and numerical approach to two-phase flow in a rod bundle is presented. ► Drift-flux model for the sub-channel and diversion cross-flow and turbulent mixing between adjacent sub-channels are used. ► Numerical results are compared with GE test data for radially uniform and non-uniform heating. ► Predicted and measured exit mass velocity and quality distributions agree very well. ► Discrepancies observed for the corner sub-channels and for non-uniform heating. -- Abstract: This paper deals with the development of a mathematical and numerical technique for the steady-state subchannel analysis. In view of the importance of the drift-flux model (DFM) in two phase flow analysis, the conservation equations are based on this model. The conservation equations are expressed in terms of five field equations: mixture and momentum continuity; liquid energy equations; gas continuity; and radial cross flow equation for the adjacent subchannels. The numerical algorithm of the single-phase subchannel analysis used in the DIYANA code are extended for the present two phase flow subchannel analysis. The transfer of mass, momentum and energy between adjacent subchannels are split into diversion, turbulent mixing and void drift cross-flow components. The transfer of mass by turbulent mixing is assumed to occur in a volume-for-volume scheme reflecting experimental observations. The phenomenon of lateral vapor drift and turbulent mixing enhancement with flow regime are included in the model. In order to validate the prediction, the GE 3 × 3 rod bundle experiments with both uniform and non-uniform radial power distribution are used. Good agreement has been obtained between the present numerical prediction and the available experimental data. Simulation results showed that the turbulent mixing and void drift phenomena flatten the void fraction and equilibrium vapor quality profile in the rod bundle cross section. The mathematical formulation is considered to be a major step toward a more basic understanding of two-phase flow analysis in fuel rod bundles. Therefore, a fast, efficient and stable numerical technique is presented in this paper for thermal-hydraulic analysis of nuclear reactors