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The ITER-TF coil requires a generation of 12.5 T in a size of 12 m x 18 m. For such a high field-large coil, the applicability of [Nb.sub.3]Al conductors was considered with a react-and-wind technique which realizes high reliability and low cost for the coil fabrication. The maximum bending strain on the [Nb.sub.3]Al conductor in use of this technique is 0.39 %. Ic degradation due to the strain is expected to be below 5 %. The limiting current is estimated as 65 kA for 60 kA operation current. Ac loss in the TF coil with [Nb.sub.3]Al conductor is almost the same as with [Nb.sub.3]Sn conductor.

The Engineering Design Activity (EDA) of International Thermonuclear Experimental Reactor (ITER) has been carried out since 1992 among the four Parties which are European Union, Japan, Russian Federation, and United States. The design and the fabrication of the Center Solenoid (CS) model coil is under way in the ITER-EDA. The CS model coil consists of four parts, which are the outer module, the inner module, insert, and the supporting structure. The design of the CS insert coil is reported and discussed. The inner diameter of the CS insert coil is 1.5m. The conductor of the CS insert coil is the same as that of the ITER-CS coil. The CS insert coil will be installed inside the CS model coil. The conductor performance will be demonstrated on the maximum magnetic flux density of 13 T at the test facility of Japan Atomic Energy Research Institute (JAERI).

Japan Atomic Energy Research Institute (JAERI) has been designing and developing the Central Solenoid (CS) model coil for International Thermonuclear Experimental Reactor (ITER). The CS model coil is composed of two modules and they will be manufactured by United States Home Team and Japanese Home Team. The coil winding method is two-conductors-in-hand and the coil which has an inner diameter of 1.6 m and an outer diameter of 3.6 m, produces a maximum field of 13 T. The coil will be tested in the JAERI facility. This paper describes the coil design, manufacturing procedure, stress analysis on the winding pack and AC loss analysis.

The [Nb.sub.3]Sn cable is being fabricated for the Central Solenoid (CS) Model Coil under the ITER Engineering-Design Activity. The cable consists of about 1000 strands whose diameter is 0.81 mm. The design current is 48 kA at a magnetic field of 13T. The 0.6-GJ CS model coil is operated in a pulse mode (0.5T/s). The first trial fabrication of a 100-m dummy cable and a 20-m superconducting cable was completed successfully. The second trial fabrication of a 1000-m dummy cable was performed to establish the stable manufacturing procedure in January, 1995. We measured the AC losses of the full-sized conductor and could determine the cable coupling time constant. We analyzed the heat generation of the CS model coil and calculated the temperature rise of the cable for the model coil.

The design and fabrication techniques for the ITER magnets are supported by the ITER Model Coil Program which focuses on component development and performance verification. The key component of the magnet system is the high current capacity (45 - 60 kA), [Nb.sub.3]Sn, cable-in-conduit conductor, which is now being fabricated under a staged development program. The integrated design features and performance of the 13 T Toroidal Field coils and Central Solenoid will be verified in the Model Coil Program. This paper describes the status of the structural and construction features of the model coils and the test facilities for the model coils at JAERI (Naka) and FZK (Karlsruhe).

The general design and the associated performance analysis of a joint proposed by the European Union for both ITER Central Solenoid and Toroidal Field coils are presented. Results of an extensive testing programme performed on subsize joints (scale 1/8), including DC resistance and pulsed field losses measurements are given and discussed, as well as their interaction on the final joint design. Results of feasibility tests carried out on a full-size piece are also given. This study has finally led to the design of a full-size joint sample to be fabricated in industry and tested in specific laboratory test facilities.

The ITER poloidal field (PF) coil system consists of a central solenoid coil and seven ring coils. The PF coils must provide the magnetic fields to confine the plasma and control its position during the various phase of operation including plasma initiation, ramp-up, burn and shut-down. They contribute the magnetic flux change to ramp up the plasma current and a part of the flux change to maintain it. This paper describes the requirements and the conceptual design for the seven outer PF coils.