Subchannel analysis of supercritical water cooled reactor

As the only water cooled reactor among the six generation-Ⅳ reactor, supercritical water cooled reactor (SCWR) has its special characteristics, and takes up attentions extensively. Based on the PWR subchannel analysis code of Shanghai Nuclear Engineering Research and Design Institute, a subchannel code for SCWR was developed. Furthermore, the thermohydralic characteristics of the typical SCWR fuel assembly with the moderator water rod, such as subchannel temperature, fuel rod cladding temperature, heat transfer coefficients and so on, were investigated using the developed subchannel code. Finally, a sensitivity study of different heat transfer correlations was performed on the thermohydralic characteristics. The results show that obvious difference occurs when using different heat transfer correlations.