The magnetic confinement of plasma is the most promising option to use controlled nuclear fusion as a power source for future generations. A number of different magnetic field con­figurations have been proposed to achieve plasma ignition, all requiring high field strength over a large volume. Most of the experimental machines use conventional, copper windings op­erated in pulsed mode, to investigate the plasma physics. The advanced plasma experiments, as well as the future fusion reactors, call for long confinement time and high magnetic field, which can be reasonably maintained only by supercon­ducting coils.

Unlike other applications of superconductivity, for fusion magnets there is no ‘‘normal conducting” alternative: when­ever a magnetic confinement fusion power plant will operate, it will have superconducting windings. For this reason, fusion magnets are an important, long-term factor in the market of superconducting technology. Today, for NbTi-based conduc­tors, fusion is a nonnegligible share of the market, with over 50 t of strand recently used for the LHD and about 40 t com­mitted for W7-X. For Nb3Sn technology, two large devices, the T-15 tokamak and the ITER model coils, have used most of the conductor ever produced (each about 25 t of strand), being the driving input for the development of high performance Nb3Sn strands.

J. Webster (ed.), Wiley Encyclopedia of Electrical and Electronics Engineering. Copyright © 1999 John Wiley & Sons, Inc.

Strand Weight (t)



Stored Energy (MJ)

Peak Field (T)

Operating Current (kA)

Tokamak T-7






Tokamak T-15






MFTF (all coils)



1 000












Tore Supra


NbTi/pool 1.8 K




LHD-Helical (2 coilsb)


NbTi/pool 4.5(1.8)K

930 (1 650)

6.9 (9.2)

13 (17.3)

LHD-Poloidal (6 coils)



1 980



Wendelstein 7-X






0 FF = forced flow.

b Operation at superfluid helium is planned at a later stage. c Design values, achieved on single coil test.

The first use of superconducting coils in experimental fu­sion devices dates back to the mid-1970s. In the last twenty — five years, six sizable devices for magnetic plasma confine­ment have been built with superconducting coils (see Table 1): T-7 and T-15 in the former Soviet Union, MFTF in the United States, TRIAM and LHD in Japan, and Tore Supra in France. In Germany, Wendelstein 7-X is under construction. Moreover, a number of developmental and prototype coils have been tested in the scope of large international collabora­tions (large coil task, demonstration poloidal coils, ITER model coils).

The operating requirement for fusion magnets may vary over a broad range, depending on the kind of confinement and the size of the device (1), for example, from medium-field, pure dc mode in the helical coils of the stellarators, to the high-field, fast rate in the central solenoid of the tokamaks. There is no general recipe for the magnet design, but a few common issues can be identified. Long-term reliability calls for a conservative component design and generous operating margins. The maintenance by remote handling in a nuclear environment imposes strong restrictions to either repair or replacement of individual parts. Safety regulations are also a major issue for superconducting magnets in a fusion reactor: the design must account for any likely or less likely failure mode of the coil system and provide that it will not turn into a nuclear-grade accident. Last but not least, the cost of the magnets, which is a large fraction of the reactor cost, must be contained to be commercially competitive with other power sources.

Only low-temperature superconductors have been consid­ered to date for use in fusion magnets at field amplitudes up to 13 T. A substantially higher field, which would make at­tractive the use of high-temperature superconductors, is not likely to be proposed as the electromagnetic loads, roughly proportional to the product of field, current and radius, al­ready set a practical limit for structural materials. It may sound surprising that the actual superconducting material cross-section is mostly smaller than 5% of the overall coil cross-section. The choice between NbTi and Nb3Sn conductors is dictated by the operating field. The upper critical field of NbTi conductors is « 10 T at 4.5 K and « 13 T at 1.8 K. According to the design current density and the temperature margin, the operating field is set at least 3 T to 4 T below the upper critical field. In the conservatively designed fusion magnets, the peak field for NbTi conductors is up to « 9 T for coils cooled by a superfluid helium bath (e. g., Tore Supra and LHD helical coils) and up to « 6 T for supercritical helium forced flow (e. g., W7-X and LHD poloidal coils). At a higher operating field, the choice of Nb3Sn conductors is mandatory to obtain adequate temperature margins and high current density. The increasing confidence in Nb3Sn technology, as well as its slowly decreasing cost, tends to move down the field threshold for the NbTi versus Nb3Sn. Conductors based on Nb3Al are in a developmental stage and may become an alternative to Nb3Sn for selected high-field magnets (e. g., the D-shaped toroidal field coils), because of the better tolerance to bending strain.

The winding packs may be either potted in epoxy resin or laid out as a spaced matrix of noninsulated conductors in a liquid helium bath. This last option offers the advantages of constant operating temperature and potential high stability due to the bath-cooled conductor surface. The drawbacks are the poor stiffness of the winding and the limited operating voltages (the insulation relies on the helium as dielectric). The potted coils with forced-flow conductors have superior mechanical performance and may operate at higher voltage: as a rule of thumb, they become a mandatory option for stored energy in excess of 1 GJ to 2 GJ. The cable-in-conduit conduc­tors became, in the last decade, the most popular option for forced-flow conductors because of the potential low ac loss and the good heat exchange due to the large wet surface. To with­stand the mechanical and electromagnetic loads, the coils are fitted in thick-walled steel cases, either welded or bolted. More structural material may be added, if necessary, both in the conductor cross-section and in winding substructures, for example, plates and cowound strips.

The magnetic stored energy is very large, up to 130 GJ for the proposed magnet system of ITER. In case of a quench (local transition from superconducting to normal state), the stored energy must be dumped into an outer resistor to avoid an overheating and damage of the winding. A large operating current is needed to reduce the number of turns, that is, the winding inductance, and extract quickly the stored energy at a moderatly high voltage (up to 10 kV to 20 kV). The op­erating current density in the superconducting cross-section (NbTi or Nb3Sn filaments), J“, is selected according to the specific design criteria to be a fraction of the critical current density, Jc, at the highest operating field. Typically, Jop is in the range of 200 A/mm2 to 700 A/mm2, and Jop/Jc = 0.3

Figure 1. The Yin-Yang coils being assembled at one end the mirror fusion test facility (courtesy of C. H. Henning, Lawrence Livermore National Laboratory).

A/mm2 to 0.6 A/mm2. The current density over the coil cross­section is over one order of magnitude smaller.

In the nonsteady-state tokamak machines, the normal op­erating cycles and the occasional plasma disruption set addi­tional, challenging requirements, in terms of mechanical fa­tigue of the structural materials and pulsed field loads on the superconductors.

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