Russian concept for a DEMO-S demonstration fusion power reactor

• Published in: Fusion engineering and design 2008-12, Vol.83 (7), p.870-876
• Main Authors: Kolbasov, B.N., Belyakov, V.A., Bondarchuk, E.N., Borisov, A.A., Kirillov, I.R., Leonov, V.M., Shatalov, G.E., Sokolov, Yu.A., Strebkov, Yu.S., Vasiliev, N.N.
• Format: Article
• Language: English
• Subjects: Demonstration power tokamak reactor; Helium-cooled ceramic tritium breeding blanket; Liquid lithium tritium breeding blanket; Steady-state plasma operation; Vanadium alloy refabrication
• ISSN: 0920-3796; 1873-7196
• DOI: 10.1016/j.fusengdes.2008.07.041

Conceptual design studies for a tokamak-based demonstration fusion reactor have been carried out in Russia since 1991. The preferred concept was a steady-state operating tokamak with superconducting magnets, a single-null divertor configuration and a high contribution of bootstrap current into a plasma current drive. Two blanket concepts were analyzed: (1) a helium-cooled ceramic (Li 4SiO 4) design for tritium breeding, using ferritic steel as the structural material, and (2) a blanket using liquid lithium for the tritium breeding material and coolant and a vanadium–chromium–titanium alloy as the structural material. Conventional-type water/lithium-cooled divertor targets with a maximum heat load of ∼10 MW/m 2 were chosen. Blankets of both designs require beryllium as a neutron multiplier and have to be replaced after the integral fusion neutron load on the first wall reaches 10 MW a/m 2. The results of the analyses show the necessity of additional studies prior to choosing the most promising blanket concept for further development. Aspects of radioactive waste management and scarce material recycling were also considered.