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

Abstract The 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 an essential component (i.e. a 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 a water-cooled ceramic breeder (WCCB) blanket and a supercritical carbon dioxide (S-CO 2 ) cooled lithium–lead (COOL) blanket, are independently designed with evaluation of their neutronics and thermal hydraulic performance. For nuclear performance, the tritium breeding capability, neutron irradiation damage as well as the shielding performance on the toroidal field coil and vacuum vessel are comprehensively analyzed. In addition, the divertor blanket is also adopted to study its contribution to the tritium breeding ratio (TBR) increment. As part of the design optimization, the thickness of tungsten is further increased to investigate its effects on reduction of the TBR, with the aim of finding the optimal thickness in conjunction with plasma corrosion. Furthermore, thermal hydraulic and magnetohydrodynamics analyses are performed appropriately for WCCB and COOL blankets, 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 the blanket.