Abstract: Research activities are ongoing worldwide to develop nuclear power plants with a supercritical water cooled reactor （SCWR）. However, there is still a big deficiency in understanding and prediction of heat transfer in supercritical fluids. In this paper, thermal-hydraulic behavior of supercritical water in sub-channel of triangular-array rod bundle has been investigated using computational fluid dynamics（CFD） code CFX. Two cases, i.e. constant wall heat flux at cladding surface and constant volume heat density in fuel pellet, are analyzed. Results show that circumferential conduct heat transfer in cladding significantly reduces the non-uniformity of circumferential temperature and heat transfer distributions, but the conduct heat transfer is with weak effect on second flow and velocity fluctuation across the gap. Turbulence mixing at rod gap strongly depends on pitch-to-diameter ratio（P/D）. If P/D＜1.3, the mixing coefficient is in the range of 0.02～0.025. It also shows unusual behavior of mixing coefficient in the vicinity of the pseudo-critical point and needs further investigations.