Steady-State Thermal-Hydraulic Analysis of SCWR Assembly

Among the six GEN-IV reactor concepts recommended by the Gen-IV International Forum（GIF）,supercritical water-cooled reactor（SCWR） is the only reactor type with water as coolant.Due to its high outlet temperature,it achieves a high thermal efficiency and subsequently has economic advantages over the existing reactors.The present paper performs the thermal-hydraulic analysis of the SCWR assembly using the modified COBRA-IV code.Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow or increasing the heat resistant between moderator channel and its adjacent sub-channels,and the latter also can get a better moderation.According to this paper,the heat transfer deterioration phenomenon is inevitable in the SCWR design.Thus,it is necessary to calculate the cladding temperature accurately,to preserve the fuel rod cladding integrity under heat transfer deterioration.