As the only one design using water as coolant in Generation IV Nuclear Energy Systems proposed by Generation IV International Forum (OFF), supercritical water cooled reactor (SCWR) has been widely investigated due to its advantages of simplicity, high thermal efficiency, good economy and safety. The Sino-Euro corporation project Supercritical Water Reactor-Fuel Qualification Test (SCWR-FQT) is proposed to analyze and verify a supercritical water-cooled experiment loop containing a small scale fuel assembly and provide support to this loop's design and license application. To solve the problems above, this paper will carry out the preliminary safety research for the SCWR-FQT facility. Several thermal-hydraulic codes (including CFD code, subchannel code, system code) are selected to perform the safety analysis in the facility. COBRA and ATHLET have been modified by implementing the heat transfer model, turbulent mixing and friction models which are applied to supercritical water cooled reactors into the previous code. To determine the detailed thermal-hydraulic behavior in the bundle region at some transient condition, the multi-scale coupled code has also been developed by transferring the pressure at the core outlet, coolant temperature at the core inlet, coolant mass flow rate at the core inlet, heat generated in the active core and pressure friction coefficient of SCWR-FQT. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test and fuel bundles can be cooled effectively with safety measures in accidental conditions. The CFD results show that wire-wrap has large influence on heat transfer. The subchannel code presents that the cladding temperature will be highest when blockage occurs in the corner subchannel. Meanwhile, the coupled results show that the coupled code can predict the accurate accident procedure and detailed parameters distribution.
