| Chinese Fusion Engineering Test Reactor (CFETR), the next generation of full superconducting fusion reactor of China, is under design. The poloidal field (PF) coils system on CFETR, which provides control of plasma position, shape and current etc.,plays an important role to achieve anticipate mission performance. In the original design scheme, it consists of six central solenoid (CS) coils, six PF coils and two extra superconductive coils (DC1 and DC2). These two DC coils give the CFETR with the ability to generate a series of advanced divertor configurations. Based on the latest physics design of the CFETR, it will be operated in two phases and have a new dimension which has eight CS coils. In phase ?, the fusion power is up to 200 MW; in phase ?, we assume the fusion power Pf of 1 GW, the auxiliary power Paux of 60 MW,considering 20% of Pf is alpha heating. Then, the total heating power is 260 MW. If?30% of the total heating power is radiated in the core plasma, the power goes into the scrape-off layer (SOL) of the CFETR, PSOL is about 182MW. It is a critical value for the divertor of the CFETR,because the wetted area is only ?1 m2 and the heat load removal capacity is ? 10MW/m2. It means some solutions should be taken to enhance radiative power loss of the PSOL and enlarge the wetted-area or even realize a detachment divertor. The experiments in TCV and NSTX showed that a snowflake divertor (SF) configuration could enable larger wetted area, longer connection length,more heat diffusive loss and maybe partial detachment.In this paper, the capacity of PF coils in obtaining the SD and the SF configurations was evaluated with the old design and the new design, respectively. The PF coils must remain their current limits. Although coil field limits, vertical force limits etc. are very important, they will be focused on in future. Instead of doing time-depending discharge evolution, we use a static equilibrium analysis method to calculate the required current in PF coils. There are a couple of fiducial points in the time histories of a discharge,such as start of plasma (SOD), point in ramp (PIR), start of flattop (SOF), middle of flattop (MOF), end of flattop (EOF) and so on. Each fiducial point corresponds the poloidal flux consumption. For any point in the discharge, we can use TEQ equilibrium solver to calculate the equilibrium and the corresponding current in PF coils, then we could know if the PF coils have the proper size. According to the empirical scaling law,the flux consumption for the ramp-up stage could be estimated. We calculate the PF coil currents of the SD and the SF configurations during the flattop phase with a range of li for the 7.5-MA&10-MA H-mode inductive scenario. The results indicate that there is enough volt-seconds of flattop phase for the SD, and the SF configurations.In addition, the Quasi-Snowflake equilibria are also established for EAST. The equilibria were analyzed and an optimal relatively solution was found out. |
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