Optimization of the PF coil system in axisymmetric fusion devices

• Published in: Fusion engineering and design 2018-08, Vol.133, p.163-172
• Main Authors: Albanese, R., Ambrosino, R., Castaldo, A., Loschiavo, V.P.
• Format: Article
• Language: English
• Subjects: Coiling; Coils; Design optimization; Fusion; Nuclear electric power generation; Nuclear energy; Nuclear power plants; Optimization; Optimization problem; Plasma; Plasma fusion; Tokamak; Tokamak devices; Toruses
• ISSN: 0920-3796; 1873-7196
• DOI: 10.1016/j.fusengdes.2018.06.004

•Optimization of position, number and dimension of poloidal field coil system.•Method able to maximize flux swing keeping plasma shape within a certain tolerance.•Reduction of 20% in terms vertical and separation force is shown in DEMO case.•Reduction of 35% in terms of forces is shown in the ITER case. The tokamak is nowadays the most promising fusion device for the plasma confinement and the production of fusion energy. It exploits magnetic fields to confine plasma inside a torus shaped chamber. The magnet system of a tokamak device is mainly composed by Toroidal Field (TF), Central Solenoid (CS) and Poloidal Field (PF) coils. In this paper, a new approach is described for the optimization of the PF coil system. The proposed procedure allows to optimize the number, position and dimension of the PF coils reducing, at the same time, currents and forces on the coils while fulfilling the machine technological constraints. The method exploits the linearized relation between the plasma-wall gaps and the PF coil currents. The procedure effectiveness has been tested on the PF coil system of the ITER tokamak and exploited for the design and optimization of the PF coil system for the next generation tokamak DEMO, as shown in the last section of the paper.