Plasma and Fusion Research: Regular Articles

Volume 11, 2405034 (2016)

NITA Coil — Innovation for Enlarging the Blanket Space in theHelical Fusion Reactor∗)
Nagato YANAGI, Takuya GOTO, Hitoshi TAMURA, Junichi MIYAZAWA and Akio SAGARA
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan(Received 30 November 2015 / Accepted 2 February 2016)
An innovative idea is proposed for enlarging the blanket space on the inboard side of the torus for the helicalfusion reactor FFHR-d1. A set of sub-helical coils, named NITA coils, with opposite-directed current outside themain helical coils, eﬀectively reduces the helical pitch parameter and enlarges the blanket space. Dependence ofthe blanket space and plasma volume on the eﬀective helical pitch parameter is examined. The obtained magnetic
surfaces and their properties are compared with that of the original configuration.
c 2016 The Japan Society of Plasma Science and Nuclear Fusion ResearchKeywords: helical fusion reactor, FFHR-d1, optimization of heliotron magnetic configuration, blanket space,NITA coil
DOI: 10.1585/pfr.11.2405034

1. Introduction
Conceptual design studies of the helical fusion reactor
FFHR-d1 are steadfastly progressing at National Institutefor Fusion Science (NIFS) taking advantage of the steadystate operation capability of the LHD-type heliotron magnetic confinement, which has no need for current drive andis free from disruption [1]. The present design, FFHRd1, aims at generating 3 GW of fusion power through selfignition. The FFHR-d1 has a pair of helical coils with apoloidal pole number l = 2 and a toroidal pitch number m= 10. The basic magnetic configuration is similar to thatof the presently operating Large Helical Device (LHD) [2].
The major radius R is 15.6 m, which is four times that ofLHD. A multi-path strategy is pursued to secure the design
foundation of FFHR-d1, which presently specifies FFHRd1A as the base option for promoting three-dimensional
engineering design [3]. The FFHR-d1A has the minor radius of the helical coils ac = 3.744 m. Using these values,
the helical pitch parameter γc isγc =m ac
10 3.744
×= 1.20.
=l Rc
215.6(1)Here γc was changed from 1.25 in FFHR-d1 (ac = 3.9 m) toimprove the confinement by reducing the Shafranov shift.
One of the diﬃcult issues for the engineering designof the heliotron-type fusion reactor is that the “blanketspace” Δcp on the inboard side of the torus is very small.
Here Δcp is defined as the distance between the innermost
layer of the helical coil windings and ergodic layer outside the last closed flux surface (LCFS). In FFHR-d1, Δcp
is evaluated at 890 mm (Fig. 1). Within this blanket space,
the following components should be installed [4]: the bottom case of the helical coil, the low-temperature thermal
author’s e-mail: yanagi@LHD.nifs.ac.jp∗) This article is based on the presentation at the 25th International Toki
Conference (ITC25).

Fig. 1 Cross-sectional image of the FFHR-d1 configuration (γc= 1.25) at the toroidal angle of φ = 0◦ .

shield, the radiation-shielding blanket (including the vacuum seal) and the breeding blanket. In the present design, the radiation-shielding and breeding blankets haverelatively thin thicknesses of 550 mm and 150 mm, respectively. We also note that the gap between the coil caseand low-temperature thermal shield only becomes 10 mm2405034-1

c 2016 The Japan Society of Plasma
Science and Nuclear Fusion ResearchPlasma and Fusion Research: Regular Articles

Volume 11, 2405034 (2016)

Fig. 2 Original proposal for enlarging the blanket space on the inboard side of the torus by employing a pair of sub-helical coils (presentlycalled NITA coils). In this drawing (by T. Watanabe; slightly modified and coil names added), the main helical coils are divided
into three thin blocks. The helical pitch parameter is 1.25. The helical coils of LHD are also shown for comparison. The NITA
coils are located at 1.492 times the minor radius of the main helical coils and have an opposite-directed current of −11.59%.

at room temperature during the construction. In the present
base option of FFHR-d1A, due to the decrease of γc , Δcp
increases to 940 mm and the engineering diﬃculty slightlyeases. To further improve the structural integrity, lower theneutron flux and increase the tritium breeding ratio, enlarging Δcp even further is preferable.
Four methods were previously considered for enlarging Δcp . The first was to choose a smaller γc , such as 1.15
for FFHR-2m1 [5], however, this also results in a smaller
LCFS (plasma volume). The second approach was to set
a high current density in the helical coil, which intensifies the engineering challenge. The third was to make an
outward shift of magnetic surfaces and cut the remnant ergodic layer by target plates on X-points [5, 6]. However,
the alpha particle confinement and neoclassical transport
deteriorate; as a result, an inward shifted configuration isemployed for FFHR-d1. The fourth method included splitting the helical coils on the outboard side of the torus toimprove the poloidal symmetry of magnetic surfaces [7],but the outward-shifted configuration is again the problem.
In 2014, another innovative method was proposed by
Tsuguhiro Watanabe. The idea was to vary the current density within the helical coil by subdividing the coil windings
into multiple blocks [8]. However, the proposed method
was later determined technically infeasible due to the extremely high current density required for the helical coilblocks. Then, a modified version based on a similar concept was proposed by the same author to divide the coilblocks in the minor radial direction using a set of subhelical coils [9]. This method is discussed in this paper.

2. FFHR-d1-TW Configuration with
Enlarged Blanket Space by NITA
Coils
By dividing the helical coil into multiple blocks withindividual currents, the helical pitch parameter of the entire

helical coil is given as follows by extending Eq. (1),
1 aci Ii
10 ,×γc =2RcIi(2)

where aci is the minor radius of each block and Ii is the current in that block. Here γc is given by the averaged minor
radius of all blocks, i.e., the current center. We note that inLHD, the helical coil is divided into three blocks: H-I (inner), H-M (middle), and H-O (outer). By uniformly supplying current to the three blocks, γc = 1.25 in the standard
configuration. When the current is only supplied to the H-I
blocks, γc becomes smallest at 1.12. When only the H-O
blocks are energized, γc reaches its maximum value 1.38.
It should be noted that in LHD, Δcp changes by γc since thestructure of the entire helical coil does not change.
The new proposal is an extended version of this idea.
A set of sub-helical coils is placed outside the main helical coils as shown in Fig. 2 (which is the original proposal
by T. Watanabe). By applying an opposite-directed current
to these sub-helical coils, γc is eﬀectively reduced according to Eq. (2) and Δcp is enlarged as a resultant. Here thesub-helical coils are called NITA (Newly Installed TwistAdjustment) coils.
Figure 3 (a) shows a plan view of the coil system of thepresently designed FFHR-d1-TW with a pair of NITA coils
located at aNITA = 2 × aHC . An opposite-directed currentof −7.692% is supplied to give the eﬀective γc of 1.20.
The blanket space becomes 1080 mm in this case as shownin Fig. 4. (Here the original blanket for FFHR-d1 is drawn,
which should be reexamined for its thickness and position.)
Figure 5 shows the dependence of the blanket space on theinboard side of the torus and the averaged minor radius ofLCFS on the helical pitch parameter. Using NITA coils,Δcp is eﬀectively enlarged, but the plasma volume does not
significantly decrease.

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Plasma and Fusion Research: Regular Articles

Volume 11, 2405034 (2016)

Fig. 5 Dependence of the blanket space on the inboard side of
the torus and average minor radius of LCFS on the helicalpitch parameter. Open symbols correspond to the casewithout NITA coils. Closed symbols are with NITA coils.

Fig. 3 (a) Plan view of the coil system of FFHR-d1-TW including the NITA coils. (b) Three-dimensional image of thecoil system (drawn by T. Watanabe; coil names added).
In this case, each NITA coil is divided into two separatecoils.

Fig. 6 Comparison of the two configurations, FFHR-d1A (up)and FFHR-d1-TW (down), with regard to the magnetic
surfaces at two toroidal cross-sections: (a) φ= 0◦ and (b)18◦ .

3. Discussion

Fig. 4 Cross-sectional image of the FFHR-d1-TW configuration
at the toroidal angle of φ = 0◦ .

Since the basic analyses for plasma transport andMHD stability are presently being carried out for theFFHR-d1A configuration, examining the deviation of theFFHR-d1-TW configuration from that of FFHR-d1A is important. A comparison of the magnetic surfaces is shown inFig. 6 for the two toroidal cross-sections, which confirmsthat the core region within the LCFS is almost the same inboth cases. To examine the diﬀerence in detail, the rotational transform and magnetic well depth are compared in2405034-3

Plasma and Fusion Research: Regular Articles

Volume 11, 2405034 (2016)

4. Summary
To enlarge the blanket space on the inboard side of thetorus in the helical fusion reactor FFHR-d1, a new configuration FFHR-d1-TW is proposed, which includes two
sub-helical coils, the NITA coils, with opposite-directedcurrents to eﬀectively reduce the helical pitch parameter
without changing the position of the main helical coils.
The blanket space in the FFHR-d1-TW is enlarged by∼ 200 mm compared to the original design of FFHR-d1 by
supplying −7.692% current of the main helical coils withthe minor radius of twice as that of the main helical coils.
The magnetic surfaces do not significantly change from thebasic option of the presently designed FFHR-d1A.
For the engineering design, the enlarged blanket spaceshould be eﬀectively distributed to the necessary components, which is under discussion. In addition, electromagnetic stress analysis will be carried out to examine the effect of NITA coils. There is also a possibility of promoting
a smaller reactor design utilizing the enhancement of theblanket space.

Acknowledgments
Fig. 7 Comparison of the two configurations, FFHR-d1A andFFHR-d1-TW, with regard to the radial profiles of (a) therotational transform and (b) magnetic well depth.

Fig. 7. The diﬀerence is not significant, which may support that the plasma transport and MHD stability characteristics of FFHR-d1-TW are not very diﬀerent from those
expected for FFHR-d1A. From an engineering viewpoint,
the divertor legs protrude from slightly diﬀerent positions.
With respect to the engineering design, the enlarged
blanket space should be eﬀectively distributed to the necessary components. For example, if the radiation-shielding
blanket is thickened, the neutron flux could be reduced,
which would reduce nuclear heating and enhance the lifetime of the helical coils. Thickening the breeding blanket to increase the tritium breeding ratio is also important.
Electromagnetic stress analysis should be carried out to examine the eﬀect of NITA coils, which come close to theOuter Vertical and Inner Vertical coils.
There is also a possibility of promoting a smaller reactor design, FFHR-c1 [1]. Using NITA coils, a blanket
space of 900 mm (larger than that for FFHR-d1) is still obtained with Rc = 13.0 m.

This paper is dedicated to Tsuguhiro Watanabe, theoriginal proposer of the idea discussed in this paper, whopassed away in February, 2015. The sub-helical coils forenlarging the blanket space are now named “NITA” coils
according to his nickname. The authors thank the members of Fusion Engineering Research Project at NIFS for
fruitful discussions. They would also like to thank Enago(www.enago.jp) for the English language review.
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