7th Asia-Pacific Conference on Plasma Physics, 12-17 Nov, 2023 at Port Messe Nagoya

Engineering optimization of heliotron configuration for larger blanket space
Nagato Yanagi
National Institute for Fusion Sciencee-mail (yanagi@nifs.ac.jp):

In the conceptual design studies of the heliotron fusion
reactor [1], optimization of the magnetic configuration
has been investigated both from the physical andengineering viewpoints. From the physical viewpoint,
the simultaneous improvement of both plasma
confinement and MHD stability is a critical issue, andthe helical pitch modulation parameter α of thecontinuously-wound helical coils is chosen to be 0.0 inrecent designs in contrast to 0.1 employed in the former
design so that the resultant fusion gain would beenhanced from 10 to 15 [2]. From the engineering
viewpoint, acquiring enough blanket space between thehelical coils and the plasma, especially at the inboard
side of the torus, should be one of the critical issues toensure a high tritium breeding ratio and sufficientradiation shielding capability to the superconducting
coils. For this purpose, various methods have been
investigated so far in addition to having a high current
density in the helical coils. One of them is to apply theNITA coils which are the sub-helical coils locatedoutside the main helical coils having 10-20 % of thecurrent in the main helical coils in the opposite direction[3, 4]. An example of a NITA coil application is depicted
in Fig 1(b) in comparison to the standard case of (a) for
the design with a major radius Rc of 7.8 m [5]. Here it isnoted that the minor radius of the helical coils, ac, should
be enlarged together with the use of the NITA coils,through the helical pitch parameterm ac ,
γc =l Rc
where m (10) and l (2) are the toroidal pitch and poloidal
pole numbers, respectively. The blanket space in (b) isstill not sufficient to have a long (>30 years) lifetime forthe helical coils due to the intense neutron irradiation.
Another difficulty accompanied by the NITA coils is thatthe main helical coil current should be increased to keep
the original toroidal magnetic field. To obtain a larger
blanket space, it is thus required to increase the current
in the NITA coils much more, which should further
increase the helical coil current.
With these conditions, a different set of coils is being
explored to obtain a larger blanket space. One such trialis the toroidal field central (TFC) coils, which are thetoroidal field coils located in the central region of thetorus. An example of magnetic surfaces obtained by theTFC coils, together with a further enlarged γc of 1.35, isdepicted in Fig. 1(c) which shows a significantly larger
blanket space. The TFC coils do not alter the toroidal
magnetic field in the plasma region, and thus, there is noneed to change the main helical coil current. However,there are still many engineering issues associated with
the TFC coils. Presently the current in the TFC coils is aslarge as that in the main helical coils, which give a highelectromagnetic stress in the supporting structure. Thus,a more compact and optimum design of TFC coils isrequired, and the electromagnetic supporting structure
should be optimized, as has been investigated for theformer designs [6].
(a)

(b)

(c)

Figure 1 Vacuum magnetic surfaces of the heliotron
fusion reactor at the toroidal angle where the helical coils
are located on the equatorial plane with (a) the standard
configuration of γc = 1.25, (b) γc = 1.30 and NITA coils,and (c) γc = 1.35 and TFC coils. The right-hand figure ineach row is an enlarged image of the left-hand figure.
References[1] A. Sagara, et al., Nuclear Fusion 57 (2017) 086046.
[2] T. Goto, et al., Plasma and Fusion Research 16 (2021)
1405085.
[3] N. Yanagi, et al., Plasma and Fusion Research 11 (2016)
2405034.
[4] N. Yanagi, et al., Journal of Fusion Energy 38 (2019) 147.
[5] J. Miyazawa, et al., Nuclear Fusion 61 (2021) 126062.
[6] H. Tamura et al., IEEE Transactions on Applied
Superconductivity (2022) 32 4900504.