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The research and development program and qualification of the toroidal field (TF) coil conductor are discussed. The first test results are considered sufficient to initiate strand procurement actions with the domestic agencies and the conductor layout option II is expected to wider the range of strand suppliers.

The assembly and final installation of the toroidal field (TF) coils in the ITER machine are analyzed. The procedure and tooling required for the TF coils final installation and the feasibility of fulfilling the accuracy needed by the maximum acceptable error fields induced by the TF coils fabrication tolerances and misalignments are analyzed.

The ITER coils are all designed by using superconducting conductors with a high current capability. The electrical design philosophy of the coils has included recovery options in the event of electrical faults internal to the coils.

The design of the ITER Toroidal Field coils and structures, the main critical design and manufacturing issues and the status of the procurement arrangements are described. The qualification and research and development work that are still needed for the preparation of the manufacture of these coils are discussed.

The main design features and drivers for the ITER magnets are discussed. The most critical technical issue, the performance of the [Nb.sub.3]Sn superconductor, is overcome and the fabrication routes allowing the TF coils to be produced within schedule and budget are being identified.

The article presents the THELMA analysis of the effects of current non-uniformity on the DC performance of the Good Joint (GJ) conductor. Results suggest that the reduction of the performance with the chopped joint is an effect of the increasingly unbalanced current distribution.

The predictive analysis of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field Conductor Insert (PFCI) test program is presented. PFCI is the first insert coil based on NbTi and is aimed at bridging the extrapolation gap between its full-size short sample and the ITER Poloidal Field (PF) coils.

A study is conducted to measure the current sharing temperature T(sub cs) in the Toroidal Field Conductor Insert (TFCI) coil. The results indicate that some factors in the fabrication, cool-down or electromagnetic loading are responsible for the reduction of the critical current strands and their n-value.

The tests of the Toroidal Field Model Coil (TFMC) are completed in the TOSKA facility of Forschungszentrun Karlsruhe. The TFMC isolated strand is characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from summers scaling used in the analysis.

Cable-in-conduit conductors (CICC) consisting of about 1000 Nb(sub 3)Sn strands with an outer diameter of bout 0.8 is designed for the TF and CS coils of the ITER. The test results indicate that both joint types are stable and could be used for applications to the full size ITER coils.

The ITER Central Solenoid, which was 12 m high and 4 m in diameter, consists of a stack of 6 electrically independent modules to allow control of plasma shape. The CS model coil and insert coil test results indicate that the conductor design must be modified to achieve an operation margin.

The choice of stability criteria for near steady state operation presents a problem n the design of CIC conductors using NbTi or Nb(sub 3)Sn strands cooled with supercritical He. The main parameters affected by the choice of stability criteria are temperature margin and Cu:nonCu ratio, both of which strongly affect cable cost and overall current density.

Tests of the International Thermonuclear Experimental Reactor (ITER) Toroidal Field (TF) Insert and Central Solenoid Model Coil (CSMC) are evaluated.

Performance of the NbAl Insert and the International Thermonuclear Experimental Reactor (ITER) Central Solenoid Model Coil (CSMC) are discussed.

Possible causes for a premature voltage gradient of the International Thermonuclear Experimental Reactor's (ITER) Central Solenoid (CS) insert coil are examined.

The International Thermonuclear Experimental Reactor's (ITER) Central Solenoid Model Coil (CSMC) project is described and discussed. All of the project's goals were successfully achieved.

Research describing some characteristics of cable-in-conduit conductors is presented. In particular an analytical model is developed based on current distribution and heat flow to study the effects of joined strands and copper.

Research is presented describing the Nb3Sn and NbTi cable-in-conduit conductors used in the International Thermonuclear Experimental Reactor project, and cooled by a supercritical helium forced flow. The design of the conductors, and non-uniform distribution between superconducting strands, are discussed.

The International Thermonuclear Experimental Reactor central solenoid model coil conductors were studied to determine predicted thermohydraulic performance.

The ITER toroidal field model coil is considered in terms of its manufacture, assembly and QA methods.

Performance simulations were conducted on the International Thermonuclear Experimental Reactor-Central Solenoid coil. Operational considerations included the influence of ramp rate limitation. The non-uniform distribution is believed to be caused by the flux loops between the strands or cable substages. The effects of parameters on the coil's stability limit were also measured through simulations.

A copper stabilizer was transferred from its location near the strand surface to the center of the strand's cross-section to enhance the residual resistivity degradation. The approach led to an enhancement of one-sixth of previous value to 1.6x10(super -10) ohmmeter in a laboratory scale strand. The strand exhibited a critical current density of 715 A/mm2 at 12 T. The approach has been applied in the International Thermonuclear Experimental Reactor program.

Results of the research and development effort for sub- and full-size joints for the International Thermonuclear Experimental Reactor conductor are presented. The task, which was launched in 1993, was undertaken by the US, Japan, EU and the Russian Federation. The effort covered layout parameter variations, manufacturing approaches and performance evaluation. Full size joint samples were tested in the SULTAN and PTF facilities under different operating conditions.

The current status of production and test activities concerning Nb3Sn strands for model the coils for the International Thermonuclear Experimental Reactor tokamak is reported. Both activities are currently ongoing to evaluate the technical feasibility of manufacturing 1,500 tonnes of high quality Nb3Sn. The strands, which will be manufactured via the bronze or internal tin process, are designed to have a critical current density of 550-700 A/mm2. at 12 T and 4.2 K.

The results of tests carried out on three Nb3Sn cable-in-conduit conductors are discussed. The tests, which investigated the viability of the superconducting elements for the Next European Torus program, were conducted under conditions exceeding the conductor's rated capacity. The results provide a measure of the temperature margin during dc operation and the pressure drop.