Precedding of ICAPP 2013, Jeju Island, Korea, April 14-18, 2013, on CD-ROM (2013)
The effect of the cross flow is incorporated in the direct whole core transport calculation code nTRACER using a subchannel thermal-hydraulics code MATRA for a realistic representation of the flow field in high-fidelity reactor calculation. To examine the effect of the cross flow modeling in the whole core transport calculation, the nTRACER/MATRA coupled system was applied to the analysis of a checkerboard problem and Yong-Gwang Nuclear Unit 3 Cycle 1 hot full power condition. When compared with the closed-channel T/H model, the cross flow modeling results in a notable change in the coolant temperature distributions, which is
smoothened at the boundaries of each assembly and near the guide tubes. In the pin power distribution, a difference of 1.5% at maximum is observed in the YGU3C1 analysis but with negligible difference in the critical boron concentration.
