Superconductors for fusion: Achievements, open issues, roadmap to future

Physica C 470 (2010) 1734–1739

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Superconductors for fusion: Achievements, open issues, roadmap to future P. Bruzzone * Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom – Confédération Suisse, CH-5232 Villigen-PSI, Switzerland

a r t i c l e

i n f o

Article history: Available online 15 May 2010 Keywords: Fusion magnets Superconductors Cable-in-conduit ITER

a b s t r a c t The need of superconducting magnets for fusion reactors has been obvious since over 30 years. In last century, a dozen of fusion devices have been built with superconducting magnets. In the last years the Chinese and Korean tokamaks started operation. Four devices are under construction (SST1, W7-X, ITER, JT60SA). The size, i.e. the energy stored in the magnetic field, has driven the R&D for conductors, from the multifilamentary composite of Tore Supra, to the monolithic conductors of T15 and the Large Helical Device (LHD), to the cable-in-conduit conductors, which dominate the fusion magnets of the last 15 years. The large electro-magnetic forces on the windings also drove the selection of cooling, from the bath cooling of Tore Supra and LHD to the force flow of supercritical helium in all the other devices. The state of the art on conductor design and performance is reviewed and three open issues in the superconducting magnet technology for fusion are highlighted (performance degradation in Nb3Sn, self field limitation in large NbTi cable-in-conduit conductor (CICC), change of length upon heat treatment of ITER conductors). A projection in the future of superconductors for fusion is attempted, including the role of HTS. Ó 2010 Elsevier B.V. All rights reserved.

1. Introduction The magnetic confinement of plasma is the most promising option to use the controlled nuclear fusion as a power source for the 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. The advanced plasma experiments, as well as the future fusion reactors, call for long confinement time and high magnetic fields, which can be reasonably maintained only by superconducting coils. As in other successful applications, the superconductivity plays in fusion the role of ‘‘enabling technology”. Whenever a fusion reactor will produce electricity for our households, it will be with superconducting magnets. The first use of superconducting coils in experimental fusion devices dates back to the mid seventies. Three main kinds of fusion devices have been studied for plasma confinement using superconducting coils: the mirror machines, the stellarators and the tokamaks. The efforts of the last 10 years concentrated on the tokamaks (EAST, KSTAR, SST-1, ITER), with only the W7-X stellarator representing the non-tokamaks options. The stored magnetic energy and hence the size of the superconducting fusion devices grew

* Address: CRPP-Fusion Technology, WMHA/C37, CH-5232 Villigen-PSI, Switzerland. Tel.: +41 56 310 4363; fax: +41 56 310 3729. E-mail address: [email protected] 0921-4534/$ - see front matter Ó 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.physc.2010.05.196

by up to three orders of magnitude in 30 years, from the 20 MJ in the first superconducting tokamak, T-7 to ITER. 2. Technology trends from the devices of last century At large, the size is the main driver for the evolution of the technology in fusion magnets. The main characteristics of the early superconducting devices are gathered in Table 1. The key references for the devices of last century, including the Demo Poloidal Coil (DPC) and Polo coils, are in [1–12]. Three aspects, indirectly related to the size, are reviewed below: the cooling, the superconductors and the operating mode. 2.1. The cooling method In the very first applications, liquid helium was the only option to keep cold a superconducting magnet. Liquid helium can be stocked in large amount independently on the size of the cryoplant. The initial proposal to use force flow coolant dates back to 1965, as Kolm [13] suggested to use the circulation of supercritical helium for reliable cool down and heat transfer in large superconducting magnets. The first milestone is the spark chamber detector built in 1970 at CERN [14], with 50 MJ stored energy and 24 t of NbTi hollow conductor. In the fusion environment, the Russian tokamaks [2,3] were the pioneers in the use of force flow cooling. In the LCT project [6], half of the six coils were cooled by forced flow supercritical helium. All

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P. Bruzzone / Physica C 470 (2010) 1734–1739 Table 1 Summary of relevant superconducting magnet systems for fusion in the last century.

a b

Strand mass in magnet (t)

Conductor/cooling

Stored energya (MJ)

Peak fielda (T)

Operating currenta (kA)

Yearb/place

Baseball II-T

4.5

NbTi/pool

17

7.5

2.4

IMP SUMMA LIN-5

0.41 3.4 0.25

NbTi + Nb3Sn/pool NbTi + Nb3Sn/pool NbZr/pool

2.4 18 <1

8.5 10.3 5.8

0.4–0.8 0.3–0.42 1

LIN-5B

2

NbTi/pool

6.5

8.3

2.8

Tokamak T-7

1

NbTi/forced flow

20

5

6

TESPE MFTF (all coils)

1.2 74

NbTi/pool Nb3Sn + NbTi/pool

8.8 1000

7 12.75

7 1.5–5.9

LCT (six coils)

28

758

8

10–18

Tokamak T-15

25

NbTi–Nb3Sn/pool-forced flow Nb3Sn/forced flow

1968/ Livermore 1968/Oak ridge 1968/NASA 1970/ Kurchatov 1973/ Kurchatov 1974/ Kurchatov 1979/Karlsruhe 1980/ Livermore 1980/Oak Ridge

795

9.3

5.6

TRIAM Tore Supra

2 43

Nb3Sn/pool NbTi/Pool 1.8 K

76 600

11 9

6.2 1.4

DPC (all coils) Polo LHD-Helical (two coils) LHD-Poloidal (six coils)

7 0.5 10

NbTi–Nb3Sn/forced flow NbTi/forced flow NbTi/pool 4.5 K

40 1.9 920

7 1.6 6.9

30 15 13

1981/ Kurchatov 1984/Kyushu 1984/ Cadarache 1989/Naka 1990/Karlsruhe 1994/Toki

43

NbTi/forced flow

846

6.5

21–31

1994/Toki

The data refer to the design point. Approximate, average date of conductor manufacture.

the coils of the DPC [7–9] were force flow cooled. The option of superfluid helium bath cooling at 1.8 K was explored in France in 1984 [5] with the intention to enhance the magnetic field of the tokamak without moving away from the NbTi technology to the more demanding use of Nb3Sn conductors. The last fusion magnet cooled by a bath of helium is the helical coil of the Large Helical Device (LHD) [11], completed in 1997. In a pool-cooled magnet, the liquid helium must be in direct contact with the metallic conductor to take advantage of the large heat transfer coefficient and the large enthalpy inventory of the bath (high thermal stability). To grant liquid helium access to each conductor, the winding pack must be ‘‘helium transparent”, i.e. no fully potting is allowed and the helium has the function of electrical insulation, separating the bare winding sections, which operate with large voltage difference during a current dump. In the large helical coil of LHD, with 1 GJ stored energy, the ‘‘helium transparency”, realized by a number of spacers between conductors put severe limits to the mechanical stiffness. Deformations of the winding pack under operating loads could not be avoided. Mechanical ‘‘instability” was the price of ‘‘thermal stability”. In Tore Supra, the function of dielectricum of the liquid helium separating the pancakes failed during a current dump, destroying one toroidal field (TF) coil (a similar event occurred in 2008 during the commissioning of the LHC, also NbTi bath cooled system at 1.8 K, with even more severe damage [15]). The requirement for mechanical stiffness and reliability of the high voltage electrical insulation has ruled out the pool cooling option from the fusion devices of the present and future generations. In small applications, the pressurized helium for force flow cooling can be obtained straight from the refrigerator: the cooling circuits are laid out allowing a large pressure drop (series cooling) with a limited amount of overall mass flow rate, according the size of the refrigerator turbines. In large devices, with total mass flow rate in the range of several kg/s, the force flow cooling of supercritical helium is assisted by cold ‘‘circulation pumps” and heat exchangers. To limit the ‘‘pumping loss” of the cryogenic closed

loop, the pressure drop over the coils must be kept low, typically 61 bar. The hydraulic characteristics (pressure drop in operation and maximum pressure at quench) become a major driver for the design of fusion conductors. 2.2. The superconductor In the 1970s and 1980s, extensive activities on conductor development led to a large variety of NbTi and Nb3Sn conductor designs. Very large multi-filamentary composite were developed both in NbTi (6  6 mm for the Yin-Yang coils of MFTF [1]) and Nb3Sn (10.5  3.3 mm for TRIAM [4]). In Tore Supra [5] and Polo [10], the large NbTi composites are highly engineered, with multiple barriers and mixed matrix to limit the ac loss. With increasing coil size and stored energy, the operating current must increase to keep low the number of turns and the inductance, i.e. to maintain a manageable voltage level in case of fast current dump. At operating current >5 kA, the single multi-filamentary composite is no longer an option for manufacturing limits and because of the ac loss and stability issues. Cabled conductor, both NbTi and Nb3Sn, replaced the large composites since mid of the 1980s. The copper for stability and protection was added to the conductors by several methods, from the continuous electroplating in the Russian conductors of T7 and T15, to the contact soldering of a perforated sheath in the MFTF NbTi conductor, to the soldering in a copper case for the react&wind bronze composite. The Hitachi conductors for TRIAM and LHD used copper clad pure aluminium stabilizer soldered to the superconductor. In Nb3Sn conductors, the react&wind method was the obvious choice both for cost (smaller furnaces) and engineering reasons (no metal joining going through the heat treatment, better quality of the electrical insulation). With the appearance of the cable-in-conduit conductor (CICC) design for large Nb3Sn conductors (first large realization in the WH-LCT coil [6]), the switch from the react&wind to wind&react method was

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P. Bruzzone / Physica C 470 (2010) 1734–1739

Fig. 1. The mechanism of transverse load/bending strain in a large CICC.

unavoidable. For NbTi, the first large CICC was used for the poloidal coils of LHD [12]. Since mid of the 1990s, the CICC became the unquestioned choice for all the fusion devices, both NbTi and Nb3Sn, despite the limited success in operation (see Section 4 on degradation). The time for R&D activities in fusion conductors stopped in the early 1990s, with the freezing of the ITER conductor design [16]. Only very limited layout variations have been considered since then, e.g. twist pattern, void fraction, jacket alloy. The ability of the CICC to scale over a broad range of size and field using standard, available basic strands and a cabling technology with low accuracy requirements are the keys for the popularity of the CICC in the fusion conductors, where the operating current density is moderately low. 2.3. The operating mode Initially, only the DC coils were considered an application target for superconductors. The T7, T15 and Tore Supra tokamaks had copper coils for plasma start-up and shaping. With Polo [10] it was demonstrated that very large field pulse, up to 1000 T/s, can be sustained by sophisticated superconductors. The appearance of the CICC, which associates moderate ac loss to a high current carrying capability, and the reduction of the field rate requirement for central solenoid and poloidal coils in the large fusion devices, stopped the development of specialized pulsed field conductors for fusion. Today, the conductor layout of the poloidal field coil of the recent tokamaks is substantially identical to the DC conductors of the toroidal field coils. The trend to higher and higher operating currents is also come to an end at the level of about 50–60 kA – the ITER TF being the largest operating current at 68 kA. Higher current would bring only very limited advantages in the magnet design, but would pose serious challenges for the power supplies and switches.

4. Three topical issues 4.1. Irreversible degradation in Nb3Sn CICC Compared to a Nb3Sn monolithic conductor, the transverse load of the whole winding pack does not accumulate on the strands of a CICC. The conduit withstands the load with high stress and transfers to the ‘‘soft” cable a small displacement, i.e. a negligible stress. The Nb3Sn cable of a CICC of radius R sees ‘‘only” the load per unit meter BIop, which scales to a peak stress proportional to BRJop, increasing with the cable size and the operating current density. The transverse load mechanism in a large Nb3Sn by the Lorenz forces is depicted in Fig. 1. The ‘‘load protective” feature of the jacket opened new horizons for the use of Nb3Sn conductors in high field, where at accumulated load above 150 MPa the monolithic conductors come to a severe operation limit. The evaluation of large Nb3Sn CICC results in terms of the strand performance led to discrepancies, see e.g. [17–19]. Later experiments on cyclic load on CICC and tests of strands extracted from loaded cables [20] proved a degradation of the superconductor, with a very broad normal transition, i.e. a drastically reduced index n of the voltage–current characteristic, see Fig. 2. Detailed models are developed, e.g. in [21], to explain the degradation in terms of local bending of the strands in the cable. Eventually, micrographic investigations on bent strands [22] brought evidence that the reason for irreversible degradation is the filament damage (radial cracks). Actually, Rezza [23] observed, about 15 years earlier, hints of this degradation mechanism (another example of a lesson which we refused to learn). An example of transverse load degradation progressing with the number of load cycles is shown in Fig. 3 for a sub-size Nb3Sn CICC with steel jacket [24]. The degradation looks so far as an intrinsic

3. The current fusion devices The last generation of fusion devices, including EAST, KSTAR, SST-1, W7-X, JT60-SA and ITER, share the same concept of conductor design, the CICC. Despite the important differences in the jacket material, from the austenistic steel for ITER, EAST, JT60-SA and SST1, to the Incoloy 908 for KSTAR (with a coefficient of thermal expansion matching the Nb3Sn), to the soft aluminium alloy for the W7-X, the design homogenization reflects a stagnation of the R&D activities on the subject. From the technology point of view, the current generation of fusion magnets belongs to the feasibility demonstration. The cost optimization will be the challenge of next generation, with likely abandon of the CICC for the Nb3Sn conductors.

Fig. 2. Critical current curves of two Nb3Sn strands extracted from a CICC exposed to large and low electromagnetic load.

P. Bruzzone / Physica C 470 (2010) 1734–1739

Fig. 3. Performance degradation of a sub-size Nb3Sn CICC, shown in terms of percentage of prediction from strand as well as n-index.

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the current density over the cable cross section. At large current density and conductor size, the twisted current carrying elements (either filaments in case of composites or strands in case of cable) see a large percentage variation of the local field from one edge of the conductor to the other and hit the critical surface only at a very short section. Even when the average electric field along the conductor is negligibly low, i.e. no macroscopic voltage from current sharing is observed, a very high, local electric peak field may drive a quench, see Fig. 5. These ‘‘self-field induced take-offs” are well known to those who try to measure the critical current of NbTi strands at B < 2 T. This effect is observed also in NbTi CICC with diameter larger than 10 mm. In Nb3Sn CICC the self-field does not lead to sudden take-offs because of the lower dJc/dB and the lower (degraded) n index [26]. At increasing current, the take-off electric field decreases and eventually drops below the Tcs or Ic criterion (10 lV/m), becoming a ‘‘sudden take-off”. The ‘‘sudden take-off”, also known as self-field instability, is perfectly reproducible and is not linked to magnetic forces and strand movements. As long as the current is perfectly balanced among the strands, the CICC performance, given as Iq Tq rather than Ic Tcs, matches the prediction from strand, disregarding the amplitude of the takeoff field. However, in case of initial current unbalance, which is the default case for large conductors, the inter-strand voltage in the region of ‘‘sudden take-off” is too low to drive an effective current redistribution (despite the high peak electric field) and Iq Tq may be lower than expected. Typically, see Fig. 6, the Ic(T) curve matches the strand prediction at background field (the voltage in the current sharing re-distribute the current), but a substantial deviation is observed in the sudden take-offs range. In the case of Fig. 6 (W7-X conductor), the deviation occurs above the nominal Iop. As a safe design rule, the operating current should be selected below the range of current where a sudden take-off occurs.

Fig. 4. Comparison of the performance of two CICC with identical strand and cable layout, one filled with solder for mechanical support (sharp transition), the other with 36% void fraction in the strand bundle (broad transition).

feature of high current, large size Nb3Sn CICC. Only filling the cable space with solder, i.e. restricting any strand bending due to transverse load, prevents the irreversible degradation and fulfils the performance prediction from the strand, see Fig. 4 [25].

4.2. Self-field induced quench in NbTi CICC The local field over a conductor of radius R is the sum of the background field and the self field, proportional to JR, where J is

Fig. 6. DC performance of W7-X NbTi CICC. The deviation from strand prediction is large in the range of sudden take-off (non-measurable Ic).

Fig. 5. Superposition of self field and background field in a large twisted cable.

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P. Bruzzone / Physica C 470 (2010) 1734–1739

Fig. 7. Summary of change of length results upon heat treatment for the 16 CICC [30].

4.3. Change of length upon heat treatment for Nb3Sn CICC The ‘‘wind–react–transfer” technique which is selected for the assembly of the ITER TF coils [16], foresees that the large CICC are precision wound to the required pancake D-shape and then placed in the furnace for the heat treatment. Afterwards, the CICC is wrapped with glass fabric for electrical insulation and lowered in the grooves of the radial plates [27]. The heat-treated CICC and the radial plates must fit with minimum tolerance – the mismatch between the D-shaped conductor and the groove of the radial plates can only be recovered at expenses of the insulation thickness [28]. The change of length of the CICC upon heat treatment must be carefully controlled. During the manufacturing trials of the ITER TF Model Coil in 1998, an elongation of 480 ppm was observed for a straight section of the coil about 1 m [29]. In the ITER TF coil, with about 10 m long straight section, the change of length of the CICC can be in the order of 10 mm and must be accurately predicted and accounted in the geometry of the winding. The change of length upon heat treatment has been measured systematically on the straight short sections of ITER TF CICC and empty steel jacket compacted with the same tools as the CICC [30]. Based on the fact that Nb3Sn forms during the heat treatment at 650 °C and that its coefficient of thermal expansion is much lower than the other CICC components (steel, copper, bronze), it is expected that the CICC elongates upon heat treatment. On the other hand, the steel jacket undergoes some cold work during the jacketing process, where the diameter is reduced by 2–4 mm. At the temperature of the heat treatment, the cold work is relaxed, causing a net shrinkage of the jacket after the heat treatment. The extent of the shrinkage depends mostly on the applied cold work, i.e. on the reduction of diameter and on the kind of tooling applied in the jacketing process (drawing vs. rolling, single pass vs. multiple passes). The results of length change for 16 CICC are summarized in Fig. 7 [30]. The range of the results is impressively large, from +1253 ppm to 730 ppm. Whenever the CICC shrinks, the effect of the jacket cold work dominates over the expected elongation from the Nb3Sn formation at 650 °C. The broad range of change of length does not affect the conductor performance, i.e. the thermal strain in the Nb3Sn filaments, but is a serious issue for the assembly of the TF windings in the radial

plates. A tighter specification on jacket and jacketing procedure is mandatory to drastically reduce the range of length change. As an alternative, the radial plates should be machined based on the actual dimension of the heat-treated winding, introducing substantial delay in the manufacturing plan of the TF coils. 5. An outlook to the future The demand of superconductors for fusion projects is not a constant load for the market. The projects now in construction (ITER and JT60 SA) are a powerful driver for the Nb3Sn strand market, with over one order of magnitude increase of the worldwide production for the next four years (over 100 t/year of strand). However, the demand will abruptly drop after completion of the ITER superconductor manufacture. It will take over one decade before a similar demand shows up on the market. The high temperature superconductors (HTS) have found a niche in the current leads for the high current fusion magnets. The devices either built or under construction in this century use HTS current leads made of Bi-2223 stacks of tape. The demand is limited and cannot be considered a driver for the market of HTS. The use of HTS in the winding of the fusion magnets is unlikely as long as the magnetic field requirements remain in the present range, where the LTS satisfactorily fulfil the needs. The situation may change in future, if the field requirements would exceed the level of 15–16 T. References [1] T.A. Kozman, S.T. Wang, Y. Chang, E.N.C. Dalder, C.L. Hanson, R.E. Hinkle, J.O. Myall, C.R. Montoya, E.W. Owen, R.L. Palasek, D.W. Shimer, J.H. VanSant, IEEE Trans. Magn. (1983) 859. [2] D.P. Ivanov, V.E. Keilin, E.Yu. Klimenko, I.A. Kovalev, S.I. Novikov, B.A. Stavissky, N.A. Chernoplekov, IEEE Trans. Magn. (1979) 550. [3] E.N. Bondarchuk, L.B. Dinaburg, L.I. Doinikov, V.G. Dubasov, M.V. Zhelamsky, V.V. Kalinin, A.B. Kostantinov, A.I. Kostenko, V.V. Makarov, I.F. Malyshev, N.A. Monoszon, V.P. Muratov, V.I. Peregud, I.I. Sabansky, Yu.A. Sokolov, Yu.V. Spirchenko, G.V. Trokhachev, G.F. Churakov, V.A. Alkhinovich, U.O. Anashkin, A.N. Vertiporokh, D.P. Ivanov, S.A. Lelekhov, I.A. Posadsky, B.A. Stavissky, V.S. Strelkov, V.A. Shchepetilov, N.A. Chernoplekov, A.I. Kukshinov, Plasma. Dev. Oper. (1992) 1. [4] Y. Nakamura, A. Nagao, N. Hiraki, S. Itoh, Proc. Magn. Technol. Conf. MT-11 (1990) 767.

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