A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods (γLiAlO2) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to γLiAlO2 volume ratio is 4/1. The He inlet and outlet branches are cooling Be and γLiAlO2, respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load = 2 MW/m2), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature = 570 °C; inlet He temperature = 250 °C; total extracted power = 2700 MW; He pumping power percentage = 2%; minimum/maximum γLiAlO2 temperature = 400/900 °C; maximum structural temperature = 475 °C; and maximum Be temperature = 525 °C.
Blanket and first wall conceptual design issues of the spherical tokamak power plant