24D11
Engineering design of helical divertor coils
for the heliotron-type fusion energy reactor
ヘリオトロン型核融合炉における熱流束低減用
ヘリカルダイバータコイルの工学設計
Nagato Yanagi, Takuya Goto, Akio Sagara and the FFHR Design Group柳 長門，後藤拓也，相良明男，FFHR設計グループ
National Institute for Fusion Science322-6, Oroshi-cho, Toki, Gifu 509-5292, Japan
核融合科学研究所 〒509-5292 岐阜県土岐市下石町322-6
To best utilize the built-in helical divertors in the heliotron-type fusion energy reactor, we propose a newdivertor sweeping scheme that reduces both the divertor heat flux and erosion of divertor plates. Thisscheme employs a set of helical divertor coils (HDC) and the current amplitude in the HDC is modulated
by a few percent of that of the main helical coils. The strike point width is effectively increased while
keeping the core plasma unchanged, and rapid sweeping (~0.5 Hz) reduces the time-averaged heat flux to<1 MW/m2. The HDC is fabricated using YBCO high-Tc superconductors joined with half-pitch segments.

1. Introduction
As the result of progress in high-density andhigh-temperature plasma experiments conductedat the Large Helical Device (LHD), conceptual
design studies are being carried out for theheliotron-type fusion energy reactor FFHR [1]. A
remarkable feature of the heliotron magnetic
configuration is that clear divertor legs protrude
from the confining region, and their structures are
not seriously affected by the plasma beta or thetoroidal plasma current. The breeder blankets
effectively shield the divertor plates from direct
irradiation by neutrons emitted from the coreplasma, which assures a significant advantage tosolve the engineering problems.
As is the case with tokamak reactor design, thedivertor is a serious concern in the engineering
design of FFHR. The heat flux through thedivertor legs in FFHR is expected to reach up to~10 MW/m2. Thus, the realization of detachedplasma is considered to be crucial for reducing
the engineering requirements on divertor plates.
As an alternative method, we have proposed a
new strike point sweeping scheme that effectively
reduces the time-averaged heat flux and theerosion of divertor plates.
2. Helical Divertor Coils
The new scheme employs a small set ofhelical coils, which we term helical divertor coils(HDC) on both sides of the main helical coils [2].
Figure 1 shows the vacuum magnetic surfaces of
FFHR-2m2 including the field changes provided
by the HDC, and the divertor legs show

remarkable spreads.

Fig.1. Vacuum magnetic surfaces and divertor legs of
FFHR-2m2 for two toroidal cross-sections at (a) φ = 0°
and (b) 18°. The field changes provided by the HDC areincluded, which spread divertor legs and strike points.

In Fig. 1, the current amplitude in the HDC ismodulated by ±2% of that of the main helicalcoils. The width of the strike points is enhanced
to ~800 mm with a total length of the four
divertor legs of ~900 m along the torus. Although
there is some non-uniform loading of the divertor
flux along the torus, the effective wetted area isestimated to be >600 m2. If a rapid sweeping isrealized, the time-averaged heat flux will then be<1 MW/m2 with a total power flow of ~600 MW
to the divertor regions for a fusion power of 3
GW. Note that there is no assumption ofincreased radiation along the divertor legs.
Erosion of divertor plates will be reduced evenfor slow sweeping allowing the replacement cycle
to be significantly lengthened. Despite themovement of the divertor legs, the magnetic
surfaces are almost unaffected by this scheme.
3. Engineering Design
For the strike point sweeping, we here consider asweeping frequency of 0.5 Hz for the HDC. A
problem is the reduction in the alternating magnetic
field strength due to the skin effect in the nuclear
shield near the HDC. We estimate that theamplitude is reduced to half with a 0.4 m-think
nuclear shield for 0.5 Hz frequency. Thus, thecurrent amplitude in the HDC should be ±4% ofthat of the main helical coils.
3.1 Temperature change of divertor plates
A concern with divertor strike point sweeping isthermal fatigue of divertor plates. For a sweeping
frequency of 0.5 Hz, we crudely model that themean repetition time of heat pulse application at thereversal points on both ends of the divertor plate tobe ~2 s. The irradiation time is modeled to be ~0.1 s
for a heat pulse. This is because the total width ofthe divertor plate is ~800 mm, while the effective
width of the strike point is estimated to be ~80 mm,according to the measured width of ~20 mm in
LHD [3]. As is shown in Fig. 2, no temperature
gradient is observed at the cooled end of a 8-mm
plate for a sweeping frequency of 0.5 Hz.
3.2 Superconducting coils
The maximum current of the HDC is ±1.5 MA in
the present design and this could be supplied by 50
turns of 30 kA conductors. We propose that theHDC be fabricated using high-Tc superconductors(HTS) and be operated at temperature 20 K orhigher. As is considered for the main helical coils,
the HDC could be fabricated by joining half-pitch
conductors with stacks of YBCO tapes [4]. The AC
losses are examined and the maximum temperaturerise is estimated to be ~10 K for a ±3 T change ofmagnetic field at 0.5 Hz sweeping frequency, whichis still acceptable for HTS.

Fig.2. Temperature distribution in a divertor plate madeof tungsten by solving the 1-D heat diffusion equation attimes of t = 0.1 and 2 s with a heat flux of 10 MW/m2
applied to the z = 0 mm surface during t = 0-0.1 s.

Fig.3. Illustration of the YBCO-HTS conductor joint.

4. Summary
To reduce the divertor heat flux and erosion ofdivertor plates in FFHR, we are proposing a newdivertor sweeping scheme using a set of HDC.
Thermal fatigue on divertor plates is expected to beof no serious concern with a rapid sweeping of ~0.5
Hz. The HDC could be fabricated using
YBCO-HTS conductors. To solve the problems
associated with AC operations, we are also
pursuing the configuration modification realized inthe steady-state, like the Super-X divertor
considered for tokamaks [5].
References[1] A. Sagara et al.: Fusion Eng. Des. 83 (2008) 1690.
[2] N. Yanagi et al.: Nuclear Fusion 51 (2011) 103017.
[3] S. Masuzaki et al.: Contributions to Plasma Physics
50 (2010) 629.
[4] N. Yanagi et al: Fusion Science and Technology 60(2011) 648.
[5] P.M. Valanju et al.: Physics of Plasmas 16 (2009)056110.