Design and R and D issues for the JT-60 modification to a full superconducting tokamak

This paper describes the modification program of JT-60 towards the tokamak fusion DEMO-reactor in support of ITER. The JT-60 is planned to be modified for the superconducting tokamak, JT-60SC, to establish steady-state operation at high beta and ensure plasma applicability of ferritic steel as reduced activation materials in a reactor-relevant regime with a break-even class plasma. Main parameters of the JT-60SC device are the plasma current of 4 MA, the toroidal field of 3.8 T, the major and minor radii of 2.8 and 0.85 m, respectively, the elongation and triangularity of 1.8 and 0.35, respectively, a current flattop duration of 100 s and the total heating power of 15 MW for 100 s. Nb3Al and Nb3Sn superconducting strands with a high Cu/non-Cu ratio and NbTi strands with ∼10 μm diameter of filaments have been developed for the toroidal and poloidal superconducting coils requiring compactness and low AC losses. Full-scale cable-in-conduit cables were manufactured by automatic welding of SS316 conduits and jacketing the cables. An irradiation test for a mockup of the cost-effectively designed CFC divertor target plate has demonstrated the required high heat handling capability of ∼10 MW/m2 on the JAERI Electron Beam Irradiation Stand