Progress in the Conceptual Design of the Helical Fusion Reactor FFHR-d1
- 103 Downloads
The LHD-type helical fusion reactor FFHR has been studied to realize steady-state fusion power generation without a need for current drive and free from disruption. The conceptual design studies of FFHR are steadfastly progressing based on the presently ongoing experiments in the Large Helical Device (LHD). In order to enhance the attractive features of the base option of FFHR-d1A, which is similar to LHD, configuration optimization is being considered for FFHR-d1C. Slight modification of the helical coil trajectory gives an improved condition both for the plasma confinement and the MHD stability. In order to overcome the difficulty for construction and maintenance associated with the three-dimensional structure, innovative ideas are being explored for the superconducting magnet, divertor, and blanket. For the superconducting helical coils, the joint-winding method confirms a fast manufacturing process. The helical divertor is reexamined and practical feasibility is discussed. The maintenance method of the helical divertor and the helically-segmented breeder blanket is a serious issue and a plausible solution is proposed.
KeywordsHelical fusion reactor FFHR Heliotron Configuration optimization High-temperature superconductor Helical divertor Blanket maintenance Aquarium method
The authors thank the members of the Fusion Engineering Research Project at NIFS, especially, T. Mito, S. Hamaguchi, S. Imagawa, K. Takahata, A. Iwamoto, T. Obana, H. Chikaraishi, N. Ashikawa, M. Tokitani, G. Kawamura, Y. Hamaji, R. Sakamoto, and S. Masuzaki for fruitful discussions. Special thanks are given to T. Muroga and H. Yamada. One of the authors (N. Y.) is grateful for the continuous discussion and encouragement given by A. Komori regarding the feasibility of the helical divertor. The optimized configuration provided by S. Okamura is highly appreciated. The NITA coil configuration was proposed by the late T. Watanabe, who made tremendous contributions to the FFHR design activities. The design of the helical divertor and the helically-segmented blanket made by S. Kinoshita at Hitachi Ltd. are greatly acknowledged. This work was supported in part by the Japan Society for the Promotion of Science (JSPS) Grant-in-Aid for Scientific Research (S) under Grant 26220913.
- 7.Y. Takeiri, Prospect towards steady-state helical fusion reactor based on progress of LHD project entering the deuterium experiment phase. IEEE Trans. Plasma Sci. 99, 1 (2017)Google Scholar
- 8.M. Osakabe, Preparation and commissioning for the LHD deuterium experiment. IEEE Trans. Plasma Sci. 99, 1 (2018)Google Scholar
- 10.C.D. Beidler, Stellarator fusion reactors—an overview. J. Plasma Fusion Res. Ser. 5, 149 (2002)Google Scholar
- 34.S. Ito et al., Mechanical and electrical characteristics of a bridge-type mechanical lap joint of HTS STARS conductors. IEEE Trans. Appl. Supercond. 26, 4201510 (2016)Google Scholar
- 37.H. Tamura et al., Multiscale stress analysis and 3D fitting structure of superconducting coils for the helical fusion reactor. IEEE Trans. Appl. Supercond. 26, 4202405 (2016)Google Scholar
- 54.N. Yanagi, et al., Divertor heat flux reduction by resonant magnetic perturbations in the LHD-type helical demo reactor, in 24th IAEA Fusion Energy Conference, San Diego (2012) FTP/P7-37Google Scholar
- 67.M. Yoda, et al., Underwater laser beam welding for nuclear reactors, in Proceedings on 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference, Anaheim, California, USA, July 30–August 3, 2012 (2012), p. 191Google Scholar