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Design optimization of the full superconducting coil tokamak JT-60SC is described. Geometries of both toroidal and poloidal field coils have been changed to allow a lower aspect ratio plasma up to A /spl sim/ 2.8. However, the toroidal magnetic field and flux swing capability are kept as close as possible to the original values to sustain a long flattop period under the break-even class plasma condition using the newly-developed superconducting conductors.

Using a 20 kA Nb/sub 3/Al cable-in-conduit (CIC) conductor, a two-turn D-shaped test coil with a height of 2.0 m and a width of 1.8 m was fabricated by a react-and-wind (R&W) method. After heat treatment, the conductor was formed to the D-shaped coil. Bending strain of the conductor is /spl plusmn/0.4% at critical current I/sub c/ test section. The I/sub c/ exceeded 31.4 kA at 7.3 T, 4.2 K. In spite of the stainless steel conduit, a critical current density J/sub c/ of 89% of the strand value was attained. The total strain of the D-shaped test coil was about -0.57%, the same as the triplex-CIC sample fabricated by a W&R method, indicating that bending did not influence the I/sub c/. The I/sub c/ at 7.4 T, 5.6 K including the margin of 1 K exceeds the operational current. Therefore, it was concluded that the R&W method is quite appropriate for the toroidal field coil of JT-60SC.

The modification of JT-60 is programmed to be a superconducting tokamak (JT-60SC) that has superconducting toroidal field (TF) and poloidal field coils. The TF coil system, which consists of 18 D-shaped coils, makes the field of 3.8 T at plasma center (R = 2.8 m). The stored magnetic energy in total is 1.7 GJ. A squared Nb/sub 3/Al cable-in-conduit (CIC) conductor using SS316LN jacket allows us to fabricate the TFC by a react-and-wind (R&W) technique because of low strain sensitivity of Nb/sub 3/Al J/sub c/ characteristic. In order to support TF coils against the centering force, the inboard side of the TF coil case is wedged. Shear panels and keys between the TF coils are placed against the overturning force induced by poloidal fields. Stress analysis of the TF coil support structure was carried out. In the case of the maximum displacement caused by overturning force, the maximum stress was lower than the allowable stress of SS316LN at 4 K. The stress analysis of insulator and conduit at the inboard side of the TF coil with zooming model was also carried out. It is clear from zooming analysis that the stresses of conduit and insulator are within the allowable stress. Regarding a progress of R&D, demonstration of R&W technique using a D-shaped double layer coil with the full-scale CIC conductor is planned in this year in order to confirm its designed performance.

Two types of cable-in-conduit conductors (CICCs) for the JT-60 toroidal field coil (TFC) have been developed. The operational conditions of this conductor are 7.4 T in a maximum magnetic field and 19.4 kA in operational current. Each conductor consists of a stainless steel conduit (17 mm in inner diameter) containing 216 high performance superconducting strand (Nb/sub 3/Al or Nb/sub 3/Sn) and 108 copper wires. The diameter of strand is 0.74 mm including 2-/spl mu/m Cr plating. In order to attain a compact and stable coil, strands with high Cu/non-Cu ratio and high critical current density (Jc) were developed. The Cu/non-Cu ratio of Nb/sub 3/Al and Nb/sub 3/Sn strands are 4.1 and 3.6, respectively. The noncopper J/sub c/ of developed Nb/sub 3/Al and Nb/sub 3/Sn strands were 1914 A/mm/sup 2/ and 1843 A/mm/sup 2/ at 7.4 T, 4.2 K, respectively. Measurements of critical current (I/sub c/) of full size conductors showed that the I/sub c/ of Nb/sub 3/Al and Nb/sub 3/Sn CICCs were 98% and 79% of I/sub c/ estimated from I/sub c/ in strands. These values exceed the target values indicating that the developed conductors are quite appropriate for TFC of JT-60.

The full-size samples of 20-kA Nb/sub 3/Sn and NbTi cable-in-conduit conductors for the superconducting poloidal field coils of the fusion experimental tokamak JT-60 were manufactured. Since these coils operate in pulse mode, reduction of ac losses is one of the key issues. For the Nb/sub 3/Sn conductor, in order to separate the strands sintered with Cr coat after heat treatment, bending strains were applied on it. As a result, ac loss was effectively reduced, and it was found that 0.2% bending strain is enough to realize the design value level of 60 ms. Its critical current was also measured, and it was estimated that the compression strain due to the difference between thermal contractions of the stranded wire and stainless steel jacket was -0.6 %. For the NbTi conductor, comparison between the ac losses of the SnAg- and Cr-coated conductors was carried out. The measured coupling time constant of the Cr-coated conductor satisfied the design requirement of 50 ms, however the SnAg-coated conductor was more than five times that of the Cr-coated one.

At JAERI a plan of remodeling JT-60 tokamak fusion device by using superconducting coil system is being considered to investigate performance of break-even plasma in long pulse operation. The coil system are composed of Toroidal Field (TF), Central Solenoid (CS) and Equilibrium Field (EF) coils. The TF coil is designed to have a maximum field of 7.4 T at winding with a nominal current of 19.4 kA and a magnetic stored energy of 1.7 GJ which is larger than that of every superconducting coil constructed so far. The coil uses a cable-in-conduit conductor that a Nb/sub 3/Al (or Nb/sub 3/Sn) strand cable is inserted into a square stainless steel conduit with a circular hole and is wound with a react-and-wind technique. The design operating temperature is 4.6 K. The CS which consists of 4 modules is operated with a maximum field of 7.4 T and a maximum ramp up rate of 1.0 T/s. For the conductor, a square Nb/sub 3/Sn cable-in-conduit conductor with stainless steel conduit is used. The EF coil consists of 6 coils that the maximum size coil has an outer diameter of 11 in. The EF4 coil which has a maximum field of 7.4 T and a ramp up rate of 2.2 T/s uses the same conductor as CS and other field coils use a NbTi cable-in-conduit conductor.

Some preparations for pulse operation tests of the ITER CS model coil in the JT-60 power supply system have been performed. Since the ITER CS model coil takes a longer current ramp-up time, the pulse duration is extended up to 70 s from 15 s by decreasing the output voltage of the motor-generator. In the protection system a new grounding resistor with higher resistance has been installed and a new protection make-switch has also been developed. An analysis for the local resonances in the coil has been performed, and measurements of the coil impedance is planned to confirm the validity of the analysis results.