On the optimal radial build of a normal aspect ratio tokamak fusion system

The optimal radial build and the system parameters of a tokamak fusion system with a normal aspect ratio were found utilizing a tokamak systems analysis coupled with the neutron transport calculation. The optimum build was determined by the requirements on the shielding, the tritium breeding, and the magnetic flux density at the toroidal field (TF) coil. It was shown that for the given fusion power, a thick outboard blanket with a thin inboard blanket allowed for smaller system size. The case with a smaller aspect ratio allowed for a smaller major radius, R ₀ and a smaller system size than the case with a larger aspect ratio. With a confinement enhancement factor, H = 1.3, Q >30 was possible for a fusion power larger than 2200 MW for the case with the aspect ratio, A = 3.0; however, Q > 30 was impossible for the case with A = 4.0. The tritium breeding capabilities of the five blanket concepts to be tested as test blanket modules (TBMs) in the International Thermonuclear Experimental Reactor (ITER) were evaluated by varying the blanket thickness and the lithium-6 (Li-6) enrichment. A helium-cooled solid breeder (HCSB) concept showed the best tritium breeding capability while the shielding effect of Pb was noticeable for a helium-cooled lithium lead (HCLL) concept.