Kenji Tobita

Goals of Japan's fusion demonstration (DEMO) reactor are to demonstrate (1) steady and stable electric power generation in a power plant scale, (2) self-sufficient production of fuel (tritium), and (3) reasonable availability using a remote maintenance scheme anticipated in a commercial plant. Main design parameters of JA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, magnetic field on the plasma axis of 5.94 T. The superconducting coil system of the reactor consists of a central solenoid (CS), 7 poloidal field (PF) coils and 16 toroidal field (TF) coils. Regarding CS and PF coils, superconducting coil technology on DEMO is basically the same as that on the world largest fusion experimental reactor called ITER. In contrast, TF coils have a technology gap on magnetic energy and the resulting stress between ITER and DEMO due to their size and magnetic field. In particular, the necessity of higher design stress is critical for TF coils, requiring the development of high strength cryogenic steels surpassing the existing ones. The fundamental design strategy to mitigate tolerances in TF coil fabrication is also presented.

This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. Japan is currently working on the conceptual study of a steady-state DEMO (JA DEMO) with a major radius Rp of 8.5 m and fusion power Pfus of 1.5 to 2 GW based on water-cooled solid breeding blanket with pressurized water reactor water condition (290ºC to 325ºC, 15.5 MPa). Such a lower Pfus allows to find realistic design solutions for divertor heat removal. Recognizing that divertor heat removal is one of the most challenging issues on DEMO, the divertor design has been carried out in different approaches, including numerical divertor plasma simulation, magnetic configurations, heat sink design, etc. It is noteworthy that the latest divertor simulation led to a design window allowing divertor heat removal of the peak heat flux of 2. The breeding blanket (BB) design has been concentrated on simplification of the internal structure and pressure tightness of the BB casing against the in-box loss-of-coolant accident. Due to a large amount of radioactive waste generated in periodic replacement of in-vessel components, downsizing of waste-related facilities has come to be regarded as a significant design issue. A possible waste management for reducing temporary waste storage was proposed, and its impact on the plant layout was assessed.

The Joint Special Design Team for Fusion DEMO was organized in 2015 to enhance Japan’s DEMO design activity and coordinate relevant research and development (R&D) toward DEMO. This paper presents the fundamental concept of DEMO and its key components with main arguments on DEMO design strategy. Superconducting magnet technology on toroidal field coils is based on the ITER scheme where a cable-in-conduit Nb₃Sn conductor is inserted in the groove of a radial plate. Development of cryogenic steel with higher strength is a major challenge on the magnet. Divertor study has led to a baseline concept based on water-cooled single-null divertor assuming plasma detachment. Regarding breeding blanket, fundamental design study has been continued with focuses on tritium self-sufficiency, pressure tightness in case of in-box LOCA (loss of coolant accident) and material compatibility. An important finding on tritium permeation to the cooling water is also reported, indicating that the permeation to the cooling water is manageable with existing technology.

A new model to evaluate the plant availability of a fusion reactor by considering the required cooling time before starting the scheduled replacement of in-vessel components was developed and incorporated in the tokamak system design code TPC-SCONE. Plant availability strongly depends on the electric output and decreases with increasing electric output. The plant availability does not much improve by advances in the physics or engineering design conditions. Consequently, the effect of the increase in the electric output on reducing the cost of electricity is diminished. Thus, proper selection of the electric output of a fusion reactor (especially for a DEMO reactor) is likely required to achieve the plant availability target.