Plasma and Fusion Research: Regular Articles

Volume 9, 1405013 (2014)

Feasibility of HTS Magnet Option for Fusion Reactors∗)
Nagato YANAGI, Satoshi ITO1) , Yoshiro TERAZAKI2) , Kyohei NATSUME, Hitoshi TAMURA,
Shinji HAMAGUCHI, Toshiyuki MITO, Hidetoshi HASHIZUME1) , Junji MORIKAWA3) ,Yuichi OGAWA3) , Masataka IWAKUMA4) and Akio SAGARA
1)

National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan
Department of Quantum Science and Energy Engineering, Tohoku University, 6-6-01-2 Aoba, Aramaki, Aoba-ku, Sendai,
Miyagi 980-8579, Japan
2)
The Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan
3)
Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa, Chiba 277-8568, Japan
4)
Research Institute of Superconductor Science and Systems, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka
819-0395, Japan(Received 27 April 2013 / Accepted 22 December 2013)
Conceptual design studies are being carried out on the application of high-temperature superconducting(HTS) conductors and coils to the magnet systems of fusion reactors. A 100-kA-class HTS conductor is required to be applied at high magnetic fields of > 12 T. A simple stack of YBCO tapes embedded in copper andstainless-steel jackets is found to be a practical approach to producing large-scale conductors that exhibit highcryogenic stability and mechanical rigidity. The feasibility of the segmented fabrication method for large complex HTS coils, such as the helical coils in the LHD-type helical fusion reactor FFHR-d1, is being investigatedby developing mechanical bridge-type lap joint technology of HTS conductors.
c 2014 The Japan Society of Plasma Science and Nuclear Fusion ResearchKeywords: high-temperature superconductor (HTS), YBCO, fusion reactor, helical, FFHR-d1
DOI: 10.1585/pfr.9.1405013

1. Introduction
Owing to the significant progress that has been madein wire production technology in recent years, copperoxide based high-temperature superconducting (HTS) materials are now being applied to a number of prototype electrical devices, such as power cables, transformers, energy
storage systems and motors [1]. The production of largescale HTS magnets for use in fusion reactors has also been
considered [2, 3]. Though it is still widely believed thatHTS materials are fragile, it is actually possible to fabricate strong coils using rare-earth barium copper oxide (REBCO) tape. In this type of tape, a substrate composed ofa Ni alloy such as Hastelloy, with a high yield strength of> 1 GPa, is coated with a REBCO material such as YBCO
(where Y represents yttrium) [1].
The application of HTS materials in fusion researchbegan with the floating coil of the Mini-RT device which
has been operational since 2003 at the University of Tokyo[4]. In this coil, Ag-sheathed Bi-2223 tape was used,which was the only available HTS at that time. The
larger RT-1 device was built in 2006 using improved Bi2223 tape, and stable levitation for more than 6 hours wasachieved [5, 6]. Recently, the floating coil of the Mini-RT
was upgraded using the latest GdBCO tape for better perauthor’s e-mail: yanagi@LHD.nifs.ac.jp∗) This article is based on the invited talk at the 29th JSPF Annual Meeting(2012, Fukuoka)

formances [7, 8].
As is well known, the biggest advantage of HTS materials is their high critical current density up to high magnetic fields of > 30 T. Another important advantage is theirhigh cryogenic stability (low quench probability) due tothe increased heat capacity of the conductor composites
at temperatures > 20 K. This is in remarkable contrast to
the situation for Nb-based low-temperature superconducting (LTS) materials, which relies on the heat capacity ofhelium at ∼4 K. Figure 1 shows the temperature dependence of the volumetric specific heat C p ρ (C p is the spe-

Fig. 1 Volumetric specific heat for metals and helium (pressure:
0.6 MPa) as a function of temperature.

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c 2014 The Japan Society of Plasma
Science and Nuclear Fusion Research

Plasma and Fusion Research: Regular Articles

Volume 9, 1405013 (2014)cific heat and ρ is the density) for several metals and forhelium. As can be seen, the C p ρ for stainless-steel andcopper at 30 K are equivalent as that for helium at 4 K. The
stability margin (allowable disturbance per unit volume)
for a conductor can be estimated using w = C p ρΔT , where
ΔT is the temperature margin). Considering that ΔT could
be as much as ∼10 K for HTS conductors (unlike 1-2 K for
LTS), w is found to be ∼2 MJ/m3 (2 J/cc) for a temperature change of 20-30 K under adiabatic conditions. This iseven larger than that measured for the presently used cablein-conduit (CIC) conductors that are internally cooled byforced flow of supercritical helium at ∼4.5 K [9]. The stability margin allows a heat generation w = B jδ due to thedisplacement δ of the conductor caused by an electromagnetic force associated with magnetic field B and current
density j. For B = 12 T and j = 25 A/mm2 , a stability
margin of 2 MJ/m3 allows a disturbance (winding motion)of δ ∼6 mm, which is unlikely to occur in a real magnet.
Considering these advantages, the conceptual design
studies on the use of HTS materials in tokamak reactors
were carried out for ARIES-AT [10] and VECTOR [11]more than 10 years ago. Along with the design of VECTOR, a 10-kA-class CIC conductor using round Bi-2212
HTS strands was successfully tested [12]. Although, thesestudies and conductor development have since been suspended, the HTS option is now being considered for thehelical fusion reactor, and conductor development has been
underway since 2005.
The present paper considers several aspects of HTS
application to the magnet coils of fusion reactors, taking
the LHD-type helical reactor FFHR-d1 as an example.

2. Design of a 100-kA-Class HTS Conductor for DC Magnets of Fusion
Reactors
As an example of the application of HTS conductors
to the magnets of fusion reactors, the conceptual designof the LHD-type helical fusion reactor is considered. The
present design, FFHR-d1, is equipped with a pair of continuously wound helical coils, having a major radius of
15.6 m, a minor radius of 3.744 m and a helical pitch number of 10, that produce a 4.7 T toroidal magnetic field togenerate 3 GW of fusion power [13]. The stored magnetic
energy is 160 GJ and a 100-kA conductor is required to be
used at the maximum magnetic field of ∼12 T. The primary
selection for the helical windings is a CIC conductor using
Nb3 Al strands [14], which is regarded as an extension of
the ITER technology. However, considering the diﬃculties
related to CIC conductors, such as the complicated plumbing required and the degradation of the critical current due
to strain, an indirectly-cooled LTS conductor using Nb3 Sn
strands and an aluminum-alloy jacket has also been considered [15]. The other option for indirect cooling is to use
a HTS conductor.
To fabricate a high-current-capacity HTS conductor,

the general approach has been to employ the same concepts that have been widely applied to LTS conductors. In
order to make the conductor resistant to magnetic instabilities (flux jumps), and to achieve high cryogenic stability by avoiding a non-uniform current distribution, roundmulti-filamentary wires containing thin filaments should
be used as strands, and the strands should be twisted andtransposed. To achieve this, several methods for producing HTS stranded cable have been developed, such as
the “Roebel-assembled” cable [16], “twist-stacked” cable[17] and “Conductor On Round Core (CORC)” cable [18].
However, since these methods are complex and expensive,
we propose a simple stacking of HTS tapes. This should beparticularly eﬀective for DC magnets by allowing a formation of non-uniform current distribution, since a HTS hasa high cryogenic stability. Using this stacking approach,a mechanically strong conductor can be fabricated at low
cost.
Figure 2 shows an example of the 100-kA-class conductor design for FFHR-d1. Forty YBCO tapes, eachwith a width of 15 mm, are simply stacked within a copper jacket. The operation temperature is 20 K. The conductor has a round outer shape produced by a stainlesssteel jacket. The major specifications for the conductor are
listed in Table 1. We should point out that this is actually
an older design and has since been changed [19]. The mostrecent conductor design has internal insulation, and use astainless-steel jacket with a square outer shape, which is
directly cooled by gas helium. However, it is also worthwhile to evaluate the conductor design shown in Fig. 2, notonly with regard to FFHR-d1, but also other applications
such as the toroidal field (TF) coils of tokamak reactors.
Simple stacking of HTS tapes may also be applicable
to TF coils provided the flux changes by vertical field coils
are slow in tokamak reactors. It should be noted that inprevious work on the VECTOR design, simple stacking of
YBCO tapes was considered. If the field changes are too
fast, it is better to reduce both AC losses and the occurrence of non-uniform current distribution, which may bepossible by transposing the HTS tapes [20]. This can be

Fig. 2 Cross-sectional illustration of a 100-kA-class HTS conductor design for FFHR-d1 helical coils.

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Table 1 Major specifications of a 100-kA-class HTS conductor
design for FFHR-d1.

achieved using a Roebel-assembled conductor. To moreeasily produce Roebel conductors, the Roebel-MITO (Meandering for Inter-Transposition Optimization) method has
been proposed, and this ensures an on-demand realization of transposition using soldered joints [21]. A similar
idea was considered for the YBCO conductors of VECTOR [22].

3. Design of HTS Coil Winding Pack
Figure 3 shows a schematic cross-section of the winding pack of the FFHR-d1 helical coil using the type of HTS
conductor seen in Fig. 2. The total current in the helicalcoil is 36.66 MA and the current density in the winding region is 24.45 A/mm2 , which is larger than that for the ITER
TF coils (20 A/mm2 ). The winding pack consists of 390
turns of HTS conductors, each with an operation currentof 94 kA, installed in a stainless-steel “internal plate” withcircular grooves. This is the same concept employed for
the CIC conductor version of the FFHR helical coils [14].
The internal plate is the layer winding version of the “radialplate” used in the ITER TF coils with pancake winding. In
Fig. 3, the internal plate is cooled by gas helium flowing
through cooling pipes imbedded within it. The conductors
are cooled indirectly by the internal plate through thermal
conduction. To enhance the cooling capabilities of the internal plate, innovations such as the use of oscillating heatpipes should be considered [23]. Once again, we note thatthe winding pack shown in Fig. 3 is our former design, anda diﬀerent configuration is employed associated with themore recent square-type conductor design having internal
insulation cooled directly by gas helium [19].
As shown in Fig. 3, the orientation of YBCO tapes isadjusted so that the plane of each tape is parallel to themagnetic field. This arrangement is used in order to mitigate the dependence of the critical current on the magnetic
field orientation, which is still a remaining issue with REBCO tape. This alignment also dramatically reduces AC
losses [24]. This is ananalogous to the reduction in AC

Fig. 3 Schematic cross-section of helical coil windings using
indirectly-cooled HTS conductors for FFHR-d1.

losses that occurred when a flat Rutherford-type cable with
NbTi/Cu strands was wound by twisting the cable being
parallel to the magnetic field [25]. The arrangement shownin Fig. 3 also reduces the production of shielding currents,
and therefore the error magnetic field [26].
When 3 GW of fusion power is being generated by
FFHR-d1, it is required to remove ∼500 W/m3 of nuclear
heating in the winding pack at the inboard side of the torus(∼100 W/m3 at the outboard side) [13]. A one-dimensional
heat conduction analysis using the thermal conductivities
of conductor composites such as stainless-steel, copper andepoxy (insulation), shows that a temperature diﬀerence of
only 1 K develops between the HTS tape and the internal
plate.
Over a full 30 years of operation, the neutron fluenceand the gamma-ray dose are expected to be ∼1 ×1023 /m2
and ∼100 MGy, respectively, near the innermost layer ofthe helical coils at the inboard side of the torus. These
values are both marginal for superconductors and insulators [27]. We are also exploring the possibility of using
inorganic materials such as ceramics as insulators. Furthermore, optimization of the coil configuration is underway in order to increase the amount of space at the inboard
side of the torus, so that the shielding can be increased by∼100 mm, leading to a reduction in the neutron and gamma
ray flux by about 10 times.

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4. Segmented Fabrication of Large
Complex Coils using HTS Conductors
To facilitate the winding of large complex coils, suchas the helical coils of FFHR-d1, one possibility is segmented fabrication using HTS conductors [26, 28–31] instead of continuous winding using LTS conductors. Segmented fabrication was first proposed for helical reactors
in the 1980’s with the so-called “demountable” concept[28], which is illustrated in Fig. 4 (a). A similar concept
was proposed in 2001 using a HTS approach that took advantage of the reduced refrigeration power required dueto the higher temperature operation [29]. However, withpresent-day technology, it is diﬃcult to find a realistic solution to forming simultaneous connections between hundreds of turns of winding conductors. Thus, we proposed
the idea of connecting “conductor segments” as shown in
Fig. 4 (b), instead of “coil segments” [26]. For the bridgetype conductor joint shown in Fig. 5 (a), a mechanical joint
technique developed at Tohoku University can be appliedinstead of the formerly considered soldering method [32].
For a mechanical lap joint, the joint resistivity is estimatedto be ∼10 pΩm2 [33, 34], giving an overall joint resistanceof ∼1 nΩ for a 100-kA conductor (consisting of 2 connections per joint with 40 HTS tapes, each having a 50-mm
joint length). The entire helical coil system has 7,800 joints(390 turns, 10 segments, 2 coils), and requires a ∼5 MW
increase in the electrical power to the cryoplant for a coil
operation temperature of 20 K. This is acceptable given the
power requirements of ∼30 MW for a LTS system operated
at 4 K. To further reduce the refrigeration power, we propose the “finger-type” joint shown in Fig. 5 (b), which has
a 10 times larger contact area and requires only 500 kW ofelectricity for the cryoplant.
A rough outline of the in-situ construction procedure
for the helical coils is as follows:
1. The helical coil cases are assembled by connecting
half-helical-pitch sections along the torus to form thehelical coil structure.
2. To prepare the HTS conductors, the stainless-steel
and copper jackets of the conductor are pre-bent intoa helical shape with each ∼16 m length in a suﬃcient tolerance. YBCO tapes are then embedded into
the copper jacket (applying limited strain) and thestainless-steel jacket is welded along the conductor.
The insulator should be attached to the outer surfaceof the stainless-steel jacket at this stage.
3. One by one, the half-helical-pitch conductors areplaced into the circular grooves in the internal plateof the helical coil case.
4. Joints are then formed between two conductors as described above. To ensure the mechanical strength of
the joints, welding of the stainless-steel jacket is inevitable.
5. After each joint is formed, an inspection should be

Fig. 4 Conceptual illustrations of the segment-fabrication
method for (a) helical coils with coil segments, (b) helical coils with conductor segments and (c) vertical fieldcoils with conductor segments.

Fig. 5 Schematic illustrations of HTS conductor joints: (a)“bridge-type” lap joint and (b) “finger-type” lap joint.

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carried out using a technique such as ultrasound or Xrays. Mechanical testing can also be performed. For
quality assurance, the joint formation and inspection

Plasma and Fusion Research: Regular Articles

Volume 9, 1405013 (2014)

should be done automatically using industrial robots(not by hand). In the inspection stage, the electricalresistance of the joint can be measured while cooling a joint section with liquid nitrogen, should it berequired.
6. Insulation should be separately attached around a
joint after the mechanical jointing process is finished.
7. Using the above procedure, one layer of windings isformed, and the internal plate for the next layer (fabricated in pieces and welded together along the torus
on site) is then placed onto the lower layer. In thisway, all the layers (fifteen in the example of Fig. 3)
are fabricated using 7,800 half-helical-pitch conductors connected one by one to form the continuously
wound helical coils.
8. After the winding process is completed, the top lidsof the helical coil cases are attached and welded.
9. In order to fill the gaps within the winding area, vacuum pressure impregnation (VPI) should ideally be
carried out at a temperature of ∼150 centigrade.
Regarding the final procedure, since performing VPI
for such a large coil is expected to be diﬃcult, we have proposed an alternative to the VPI process [19]. Nevertheless,the use of VPI is worth exploring further.
Segment fabrication could also be applied to the vertical field coils, as shown in Fig. 4 (c). On the other hand,demountable coils could be a future option for the manufacture of commercial helical-type reactors, and also for
producing medium-scale TF coils for tokamak reactors thatallow blanket maintenance [35].

5. Development of 100-kA-Class HTS
Conductor and Joint for FFHR-d1
Large-scale HTS conductor development for FFHR
began with a 10-kA-class conductor in 2005 [36, 37]. In
2012, a 30-kA-class conductor sample was fabricated using the latest GdBCO tape produced by Fujikura Ltd.
Twenty tapes were simply stacked in 10 layers and 2 rows[19]. In the sample, a bridge-type mechanical lap joint
developed at Tohoku University [33, 34, 38] was used sothat the racetrack shaped sample formed a short circuit. A
sample current was induced by changing the bias magnetic
field, and a critical current of 45 kA was measured at 20 K
and 6 T [39]. Stable operation at a current of > 30 kA wassustained for more than 20 min at 30 K. The maximum
current was 70 kA at 4 K and 1 T with a fast ramping rateof 1 kA/s. No premature quench was observed, which confirms the high cryogenic stability of the HTS conductor.
Details concerning this conductor test are described elsewhere [19, 38, 39].

6. Summary
The application of HTS conductors to fusion reactor
magnets seems feasible owing both to the rapid progressin wire production technology and development of highcurrent capacity conductors at National Institute for Fusion
Science. Simple stacking of REBCO coated-conductor
tapes is a promising method for producing cryogenicallystable and mechanically rigid conductors for DC magnets
at a reasonable cost. Segmented fabrication of large complex coils, such as the helical coils for the LHD-type helical reactor FFHR-d1, is expected to be possible using mechanical bridge-type lap joints.

Acknowledgments

The authors are grateful to S. Imagawa, J. Miyazawa,
T. Goto, T. Tanaka and the members in the Fusion Engineering Research Project at NIFS for valuable discussions
and encouragement. One of the authors (N. Y.) thanks G.
Bansal, T. Hemmi, R. Champailler and H. Noguchi who
contributed to starting 10-kA-class HTS conductor development and performed feasibility studies of the HTS magnet option for FFHR.
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