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The ITER toroidal field (TF) strand procurement initiated the largest Nb3Sn superconducting strand production hitherto. The industrial-scale production started in Japan in 2008 and finished in summer 2015. Six ITER partners (so-called Domestic Agencies, or DAs) are in charge of the procurement and involved eight different strand suppliers all over the world, of which four are using the bronze route (BR) process and four the internal-tin (IT) process. In total more than 500 tons have been produced including excess material covering losses during the conductor manufacturing process, in particular the cabling. The procurement is based on a functional specification where the main strand requirements like critical current, hysteresis losses, Cu ratio and residual resistance ratio are specified but not the strand production process or layout. This paper presents the analysis on the data acquired during the quality control (QC) process that was carried out to ensure the same conductor performance requirements are met by the different strand suppliers regardless of strand design. The strand QC is based on 100% billet testing and on applying statistical process control (SPC) limits. Throughout the production, samples adjacent to the strand pieces tested by the suppliers are cross-checked ('verified') by their respective DAs reference labs. The level of verification was lowered from 100% at the beginning of the procurement progressively to approximately 25% during the final phase of production. Based on the complete dataset of the TF strand production, an analysis of the SPC limits of the critical strand parameters is made and the related process capability indices are calculated. In view of the large-scale production and costs, key manufacturing parameters such as billet yield, number of breakages and piece-length distribution are also discussed. The results are compared among all the strand suppliers, focusing on the difference between BR and IT processes. Following the completion of the largest Nb3Sn strand production, our experience gained from monitoring the execution of the QC activities and from auditing the results from the measurements is summarised for future superconducting strand material procurement activities.

The ITER In-Vessel Coils (IVCs) are crucial components of the Tokamak fusion reactor. They are located within the vacuum vessel (VV) behind the blanket shield modules and are responsible for compensating for fast perturbations of the plasma. There are a total of 176 joints used to connect the IVCs, feeders and feedthrough. The joint structure consists of a central copper conductor, a mineral (MgO) insulation layer, and an outer stainless-steel jacket. Given the inaccessibility of the IVCs once the blanket modules are installed, the highest level of quality is required for the joints. One key item is the welding of the copper conductor which has to be done inside the VV. To address the high thermal conductivity of copper conductor and the challenging constraints working inside the VV, a preheating method by the arc of the welding head has been developed. Macro- and micro-structure examination were performed on the welding cross sections, and tensile tests were carried out to check the strength of the copper joint, which demonstrated that arc preheating does not have a significant impact on the heat-affected zone. After more than 40 welding attempts and several design updates, the copper joint has been successfully developed and can be repeatedly and successfully welded.

Four samples of full size ITER toroidal field (TF) conductor are manufactured according to the conductor layout defined by ITER. The manufacturing procedures and the dc test under relevant load conditions are described.

Two European ITER toroidal field (TF) prototype samples are prepared, assembled and tested in SULTAN. The ITER specifications are fulfilled by both samples, with outstanding results for the conductor with longer twist pitches.

High-energy physics (HEP) type [Nb.sub.3]Sn strands with a [J.sub.c] ~ 2400 A/[mm.sup.2] are applied to a steel Cable-in-Conduit Conductors (CICC). The high-field (HF) conductor qualification process, the design modifications and the analysis of the test results are discussed.

An alternative solution is described by using a relatively thin walled rectangular conductor without an additional reinforcing structure. The conductor and the layout of the winding, the magnetic, thermo-hydraulic and mechanical behavior of the ITER toroidal field (TF) coils are analyzed.

A 12.5 T superconducting dipole magnet (European DIPOle, EDIPO) is designed by EFDA, which is procured within the framework of the European Fusion Programme in order to be installed in CRPP-PSI. The magnet parameters are described together with the key supporting electromagnetic, mechanical and thermal analyses.

The Poloidal Field Coil 6 (PF6) coil design based on [Nb.sub.3]Sn superconductor, which has delivered the required temperature margin as well as the higher performance capability, is examined. The design analysis has discussed the performance parameters of the coil including the hot spot temperature and quench propagation.

The test results of the new generation of ITER TF conductor samples in SULTAN are discussed. The homogenization of performance toward the upper range has provided an effective margin for the operation of the ITER coil system.

The effect of strain on the critical current of [Nb.sub.3]Sn cable-in-conduit-conductors (CICC) is examined. Findings reveal the impact of transversal stresses to a current carrying cable exposed to an external magnetic field.

Two conductor short samples were tested for examining the use of high current [Nb.sub.3]Sn strands in International Thermonuclear Experimental Reactor (ITER) cable-in-conduit conductors (CICC). Findings revealed the strong degradation of the CICC performance in comparison to the strand for all conductors.

The study explains various pure bending strain experiments conducted on several jacketed [Nb.sub.3]Sn strands for International Thermonuclear Experimental Reactor (ITER).

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.