Blanket design with evaluation on the neutronics and thermal hydraulics performance for CFEDR

China Fusion Engineering Demo Reactor (CFEDR) aims to demonstrate the fusion power output for electricity generation under the condition of tritium self-sufficiency, and it relies on the essential component (i.e. blanket) to achieve this goal. In this present physics design stage, according to the constraints on the geometry and design objectives of CFEDR, both candidate blankets, namely the Water Cooled Ceramic Breeder (WCCB) blanket and the supercritical carbon dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket, are independently designed with the evaluation on the neutronics and thermal hydraulic performance. For the nuclear performance, the tritium breeding capability, neutron irradiation damage as well as the shielding performance on the Toroidal Field Coil (TFC) and Vacuum Vessel (VV) are comprehensively analyzed. In addition, the divertor-blanket is also adopted to study its contribution on the Tritium Breeding Ratio (TBR) increment. As part of the design optimization, the thickness of tungsten is further increased to investigate its effects on the TBR reduction, aiming to find the optimal thickness in conjunction with the plasma corrosion. Furthermore, the thermal hydraulic and MagnetoHydroDynamics (MHD) analysis are performed appropriately for WCCB and COOL blanket, respectively, aiming to verify that the coolant can safely remove the nuclear heat and plasma facing heat flux without the material temperature exceeding the upper limits. The preliminary results will provide effective guidance for the subsequent detailed engineering design of blankets.