C. Sborchia

Progress in the Design, Manufacture and Testing of the W7-X Superconducting Magnets

The W7-X machine is a low-shear stellarator of the Wendelstein line being assembled in Greifswald, Germany. The manufacture of its superconducting magnets is in a well advanced phase. After the acceptance tests in the factory, the magnets are tested in cold conditions in a dedicated facility at CEA Saclay. This paper gives an overview of the status of the manufacture of the W7-X coils, including the production of the superconductor, the windings and casings for the magnets, the final assembly and in-factory test procedures. Several design changes and re-work have been implemented in the course of the manufacture due to more detailed engineering analyses or weaknesses and quality problems found in some components. The status of the cold tests of the coils will also be presented

Recent Results of the Cold Tests Performed on the Stellarator W7-X Coils

The plasma fusion experiment Wendelstein 7-X (W7-X) will use a system of superconducting coils for the production of the magnetic field. In the first paragraph, W7-X is described briefly, with emphasis on the coil system. All the superconducting coils are tested thoroughly in a cryogenic test facility. The test program is described in the second paragraph. In the third paragraph, some test results are described in more detail. Focus is here on the measurement of mechanical stresses during cool-down and application of the electrical current to the coils. The measurement of the electrical impedance, performed in order to detect short-circuits, will also be described. Finally, we present the quench tests, which are used to investigate the superconductor quality, and the Paschen-tests used for the check of the electrical insulation

Design Changes and Impact on the Production of the Non-Planar Coils for the W-7X Experiment

The Max-Planck-Institut fur Plasmaphysik is building the stellarator fusion experiment WENDELSTEIN 7-X (W7-X) at the branch institute in Greifswald, Germany. The 50 non-planar superconducting coils are currently manufactured by industry and use a specially developed NbTi cable-in-conduit conductor. The manufacture of the coils includes the conductor, winding of the conductor, vacuum pressure impregnation as well the embedding of the winding pack into a cast steel casing, the final machining and the installation of the helium cooling system on the casing. Design changes became necessary during production in the connection areas of the casings due to more detailed design and structural analysis. In addition, the insulation and the welds in the termination area need to be improved. New quench detection cables were qualified. LINAC inspections allowed to check the presence of casting defects in the thick casings. Additionally high voltage tests in vacuum under Paschen-minimum conditions turned out to be a very sensitive inspection method to check the integrity of the electrical insulation

The ITER-FEAT Toroidal Field Structures

Abstract The magnet system of ITER-FEAT consists of 18 toroidal field (TF) coils, a free standing central solenoid, six poloidal field coils, and a set of correction coils. The TF coils are being designed to provide the magnetic field necessary to maintain plasma in a tokamak configuration with a current up to 17 MA and operate at a maximum field of 11.8 T as reported by Okuno et al. (Key Features of the ITER-FEAT Magnet System, Paper F-26, 21st Symposium on Fusion Technology, Madrid, Spain, 11–15 September 2000). The TF coil cases, which enclose the TF winding packs, are the main structural components of the magnet system. Extensive finite element (FE) analyses have been performed to investigate several design options by using 3D non-linear FE models, representing a 20° symmetry section of the TF coil magnet system. Evaluation of the results was mainly focused on the acceptability of static and cyclic stresses in the TF coil case and loads on keys/pins in the inner intercoil structures and outer intercoil structures.

Installation and Test Programme of the ITER Poloidal Field Conductor Insert (PFCI) in the ITER Test Facility at JAEA Naka

ITER PFCI has been manufactured in the Europe and installed into the ITER Test Facility in Naka, Japan. The conductor is NbTi cable-in-conduit conductor with thick square stainless steel jacket and almost identical with the design of the ITER PF coils. The main objective of this test is the characterization of the conductor and joints at the conditions relevant to the ITER PF coil operation. Intermediate joint is located in the winding at relatively high field to examine its performance. The main items in the PFCI test program are thermo-hydraulic test, DC mode test, cyclic test and pulse mode test. The PFCI and CSMC were successfully cooled down to cryogenic temperature within 450 hours. The test of the PFCI was performed from May to August 2008. The key technology of the installation, the test methods and procedures, and some preliminary results of cool-down are described in this paper.

Manufacture of Full Scale Models of the ITER Toroidal Field Coil Cases

The EFDA Close Support Unit (CSU) has initiated a technological development task to manufacture three full scale models of the austenitic steel cases of the ITER Toroidal Field (TF) coils. The main goals are to verify the feasibility of such a thick construction, assess the achievable material properties, and develop the manufacturing processes and quality control procedures. A first model (called Model 1), reproducing the geometry of the inner curved region of the TF coil, has been manufactured with thick forgings. A second model (Model 2), representing the geometry of one outer intercoil structure, has been produced by casting. A third model of the inner straight leg is being manufactured. The work has included a large campaign of development and qualification of welding processes and inspection procedures for thick case sections. Numerical models for the prediction of the geometrical deformations and residual stresses due to the welding have been developed. The goal of this task is the assessment of the manufacturing and quality procedures for the production of the ITER TF coil cases.

Conceptual Design of Magnet System for JT-60 Super Advanced (JT-60SA)

At Japan Atomic Energy Agency (JAEA), the JT-60 is planned to be modified to a full-superconducting tokamak referred as JT-60 Super Advanced (JT-60SA) as one of the JA-EU broader approach projects. In JT-60SA, the magnet system with 1400 ton consists of 18 toroidal field (TF) coils, 4 stacks of central solenoid (CS) and 7 plasma equilibrium field (EF) coils. The TF coil with using the NbTi cable-in-conduit (CIC) conductor has the maximum magnetic field of 6.4 T in winding, the magnetomotive force of 8.2 Tm and the magnetic stored energy of 1.5 GJ which is 1.5 times larger than present superconducting fusion devices. The conductor for CS which is the Nb3Sn CIC conductor with the maximum magnetic field of 10 T was newly designed to increase the flux swing capability for the ITER plasma simulation and to accommodate the diminution of available space because of the TF coil with NbTi conductor. For CS, the faster maximum varying field of 2.88 T/s is required because of fast plasma initiation scenario of JT-60SA.