29pB01
核融合原型炉への適用をめざした高温超伝導導体とマグネットの開発
Development of HTS Conductor and Magnet for Fusion DEMO Reactors
柳 長門1、三戸 利行1、寺﨑 義朗2、夏目 恭平1、野口 博基1、田村 仁1、伊藤 悟3、橋爪 秀利3、森川 惇二4、小川 雄一4、岩熊 成卓5、相良 明男1
N. Yanagi1, T. Mito1, Y. Terazaki2, K. Natsume1, H. Noguchi1, H. Tamura1, S. Ito3, H. Hashizume3,J. Morikawa4, Y. Ogawa4, M. Iwakuma5, A. Sagara1
1

核融合研、2総研大、3東北大、4東大、5九大
1NIFS, 2Sokendai, 3Tohoku Univ., 4Univ. of Tokyo, 5Kyushu Univ.
1. Introduction
Due to the significant progress of the wire
production technology, the copper-oxide based
high-temperature superconductor (HTS) is now
being tried for variable electrical applications, suchas power cables, transformers, energy storage andmotors. It is now considered to be the time to apply
HTS also to fusion magnets. One of the advantagesof HTS is the high cryogenic stability of coils (lowquench probability) due to the increased heat
capacity at the operation temperature of >20 K
compared to that of metal-based low-temperature
superconductors (LTS) relying on the heat capacityof helium at 4 K. Reduction of refrigeration poweris also an attractive feature for fusion power plants.
Though it is still widely believed that HTS is fragile,
one can actually make a stronger coil using theREBCO (YBCO or GdBCO) coated conductor
owing to the Ni-alloy substrate. The current statusof the HTS conductor development for the helical
DEMO reactor and the related technology ofapplying HTS for plasma experiments are reviewed.
2. HTS option for the FFHR magnet
Based on the progress of steady-state plasma
experiments in the Large Helical Device (LHD), theconceptual design studies and related engineering
R&D’s are being promoted for the helical fusion
DEMO reactor. The present design, FFHR-d1, has
the major radius of 15.6 m and the toroidal
magnetic field of 5.1 T to generate 3 GW fusion
power [1]. The stored magnetic energy is 160 GJ
and a 100-kA conductor is required to be used at
the maximum magnetic field of 13 T. The
cable-in-conduit (CIC) conductor with forced flowof supercritical helium is the primary selection forthe coils. For the strands of conductors, Nb3Al
should be an attractive choice with better resistanceto strains than the presently used Nb3Sn for theITER magnet. Though the force-cooled coil with

CIC conductors is the established technology andwidely applied to the recent fusion magnets, it is
not easy to handle considering the proliferation offusion reactors. In this connection, the counteroption is to use indirectly-cooled HTS coils basedon the above mentioned attractive features. Figure 1
shows an example of the conductor design. We use
YBCO tapes simply stacked in the copper jacket.
This is a contrast to the activities in EU and US,
where transposed and/or twisted-type conductors
are being developed to be applied also to AC coils,
such as the central solenoids of tokamaks. Since thehelical reactor is DC, we consider that simple
stacking, which is much easier for flat HTS tapes,
should work. The conductor has a round shape andthe orientation of the HTS tapes could be adjusted
depending on the orientation of the magnetic field.
This mitigates the anisotropy of the critical current
on the magnetic field orientation, which is still aremaining issue of the REBCO tape.
Another attractive proposal using HTS is that thesegmented fabrication of the huge and continuous
helical coils be possible. The original proposal wasto have “coil-segments” as illustrated in Fig. 2(a)[2]. Considering the difficulty to join >400 turns ofconductors simultaneously, we are proposing a new
idea of having “conductor-segments” instead, as
shown in Fig. 2(b). A bridge-type joint ofconductors is considered as shown in Fig. 3 [3].

Fig. 1. A 100-kA HTS conductor design for FFHR-d1.

Fig. 2. Comparison between (a) joint of “coil-segments”
and (b) “conductor-segments”.

Fig. 3. Bridge-type joint of the REBCO conductor.

2. Development of 100-kA conductor and joint
We started the HTS conductor development fromthe 10-kA class in 2005 [4]. In 2012, we fabricateda 30-kA conductor sample (Fig. 4) using the latest
GdBCO tape supplied by Fujikura Ltd. (criticalcurrent: >600 A at 77 K, self-field with 10 mm by
0.22 mm cross-section including a laminated copperstabilizer) stacked in 10 layers and 2 rows. A
bridge-type mechanical lap joint developed atTohoku Univ. was used [5] and the sample formeda short-circuit with a race-track shape.

The sample current was induced by changing thebias magnetic field, and the critical current of 40kA was measured at 20 K and 6 T. We also tried toachieve higher currents at 4 K, however, the sample
quenched due to the unexpectedly large joint
resistance in one of the two joints. We are now
planning to improve the joint to achieve 100 kA at
4 K and <2 T in the early 2013. Then, the numberof HTS tapes will be increased so that 100 kA will
be reached at 20 K and >6 T, which may surpass
the performance of the ITER conductor.
3. Application of HTS for plasma experiments
Along with the development of large-current
capacity conductors for fusion DEMO magnets, it isalso important to develop the basic technology forhandling HTS tapes to wind small-scale coils. We
here note that the application of HTS for fusion
research has started with the floating coil of theMini-RT device which has been operational since2003 at Univ. of Tokyo [6]. The Ag-sheathed
Bi-2223 tape was used, which was available at thattime, and the time constant of the persistent current
was unexpectedly short due to a probablemechanical damage of the still fragile tape. Thelarger RT-1 device was later built in 2006 using theupgraded Bi-2223 and the stable levitation of >6hours has been available. Presently, the floating coilof Mini-RT is being upgraded using the same typeof GdBCO tape as mentioned above. It is expectedthat the time constant will be prolonged and themagnetic field will be increased.

Fig. 5. The levitated Mini-RT HTS coil (present).

Fig. 4. The 30-kA class GdBCO conductor sample.

References[1] A. Sagara et al., Fusion Engineering and Design 87(2012) 594.
[2] H. Hashizume et al., Fusion Engineering and Design63 (2002) 449.
[3] N. Yanagi et al., Fusion Science and Technology 60(2011) 648.
[4] G. Bansal et al., IEEE Trans. Appl. Supercond. 18(2008) 1151.
[5] S. Ito et al., IEEE Transactions on Applied
Superconductivity 22 (2012) 6400204.
[6] Y. Ogawa et al., J. Plasma Fusion Res. 79 (2003)
643-644.