Development of the Subchannel Flow Mixing Model into a Nuclear Thermal-Hydraulic System Code

Subchannel flow mixing is mportant for the accurate prediction of temperature and void fraction distributions in a nuclear reactor core. Dedicated subchannel analysis codes have been used for subchannel flow analysis by adopting either the so-called equal mass exchange (EM) or the equal-volume exchange and void drift (EVVD) model. The former is known to be good for a single-phase flow and, meanwhile, the latter is appropriate for both single- and two-phase flows. The nuclear thermal-hydraulic system code, SPACE code, has been developed for the design and safety analysis of a pressurized water reactor. To extend its applicability to subchannel analysis, the state-of-the art EVVD model has been implemented into the SPACE code in this work. The modified SPACE code with the EVVD model was evaluated using the measured data from two-phase flow rod bundle tests (GE 9-rod test, ISPRA 16-rod test and PSBT 25-rod test). Based on the results, a new void drift coefficient KVD was suggested, which is an important parameter of the EVVD model.