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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.

Reporting on the angular and field dependence of the critical transport current densities at various temperatures of Y-123 coated conductors deposited by PLD on a newly developed Ni-alloy with improved mechanical properties is discussed. It shows that the irreversible properties of the specimens are not limited by the pinning structure of the super conducting film, but mainly by substrate impurities.

Studying the transport properties of a-axis oriented YBCO films grown by a self-template layer procedure is strongly improved by using a YCaBCO cap layer is discussed. In contrast, when the magnetic field is oriented perpendicular to the plane of the substrate, strong pinning takes place due to the presence of 90 degree grain boundaries.

The ITER vertical stabilization coil system is designed to provide fast plasma stabilization in all scenarios with elongated plasma cross-section. It consists of two water-cooled coils located in upper and lower portions of the vacuum vessel (VV), with feedthroughs responsible for the transfer of electricity and cooling water across the vacuum boundary of VV. Depending on location, the feedthroughs can be divided into upper port feedthroughs (UPR feedthroughs) and lower port feedthroughs (LWR feedthroughs), each family sharing the same design. Both designs are subject to demanding operation scheme, a severe working environment and the presence of interfacing components, which impose important thermal, electromagnetic and seismic loads on the components. In particular, the LWR feedthrough, object of present paper, is more impacted by seismic loads due to the more flexible structure compared to the UPR feedthrough. During the worst seismic event, it is subjected to a vertical acceleration of 28.8 m s−2, acting on an effective mass of 249.9 kg at its first order modal mode. Furthermore, the current circulating in the conductor of the LWR feedthrough, impose significant linear forces, up to 27.7 kN m−1 during normal operation and 30.3 kN m−1 during categories I & II abnormal plasma events, respectively. Finally, the effect of differential thermal expansions of interfacing structures imposes a relative displacement between the supports up to 32.9 mm in X (radial) direction and up to 14.67 mm in Z (vertical) direction, during baking. Being part of the vacuum boundary, the feedthrough must assure confinement function under all normal and accidental loading conditions, with proper structural integrity assessment by applying analysis approach and design criteria foreseen by the applicable nuclear codes. The paper details the load types and categories, the analysis methodologies, the structural design criteria, and the assessment results in views of static stress and fatigue for the LWR feedthrough. This study could be a comprehensive reference for the structural integrity assessment of other similar components in fusion devices. [ABSTRACT FROM AUTHOR]

The inter- and intra-granular critical current densities were determined for different types of coated conductors as well as mono- and multi-filamentary (BiPb)[sub 2]Sr[sub 2]Ca[sub 2]Cu[sub 3]Ox tapes. Either the inter- or the intra-granular critical current density was changed systematically by bending the tapes or by neutron irradiation, respectively. Thus, we are able to assess these two critical current densities, even in samples with very good grain connectivity. [ABSTRACT FROM AUTHOR]

Inter- and intragranular critical current densities in high temperature super conducting tapes and coated conductors were determined. Two critical current densities could be assessed even in samples with very good grain connectivity as either inter- or intragranular critical current density was changed by bending the tapes or by neutron irradiation.

The superconducting coils of the ITER magnet system have hundreds of electrical lap joints interconnecting superconducting cables. The joints operate in a magnetic field of up to 4 T, field derivatives of 0.5 T/s, and currents up to 70 kA. The acceptance tests for the toroidal field (TF), poloidal field (PF), and correction coil (CC) coils will be performed at 77 K, before they are assembled in the pit. Hence there will be no possibility to measure the resistance of the joints in the superconducting state before the whole magnet system is enclosed in the Tokamak cryostat. In addition, no reliable nondestructivemethod has been found to spot the joints with a failure at room temperature. Therefore, the production of the joints relies on the strict adhesion to established robust manufacturing procedures during the qualification phase. As additional quality monitoring, a periodic test of the joint samples manufactured in parallel with a coil fabrication is foreseen to control the reproducibility of the joint electrical performance. In order to qualify the manufacturing procedures, to establish a series production tools and worker teams, a comprehensive qualification program has been set up for manufacturers of the coils in Russia (Poloidal Coil 1, PF1), China (PF6, feeders, CC), Japan (TF), Europe (TF and PF), and USA (Central Solenoid, CS). This program includes a set of mockups manufactured according to the process to be used for the coils and submitted to different tests. They include mechanical testing of materials, electrical tests of full size joint samples, destructive microscopic examination of the joint mockups, and mechanical testing of the full size joint mockups. All tests are carried out in specialized laboratories qualified for this type of work. This paper describes the main items of the qualification program, the tests performed, and the acceptance criteria. The test results are reported and compared to the criteria. [ABSTRACT FROM AUTHOR]