[en] 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 deg. 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

[en] This paper reports on a campaign of low cycle/high strain fatigue tests at 300 degrees C which has been carried out on specimens of Inconel 600 to assess the life of critical parts of the JET vacuum vessel. Specimens have been loaded with alternate cycling strain up to ± 1%. The shape of some specimens has been chosen to reproduce the real working condition and the stress distribution of the critical regions of the JET vacuum vessel during plasma vertical instabilities. The results of the tests have been used to evaluate the actual campaigns and to predict the additional damage caused by the expected future operational phases of JET. A global model of the vessel has assessed the critical regions of stresses and detailed elasto-plastic analyses have been performed to evaluate the maximum strain value for reference disruption scenarios

[en] The divertor coil is mounted inside the Jet vacuum vessel and is able to carry 1 MA turns. It is of conventional construction - water cooled copper, epoxy glass insulation -and is contained in a thin stainless steel case. The coil has to be assembled, insulated and encased inside the Jet vacuum vessel. A description of the coil is given, together with technical information (including mechanical effects on the vacuum vessel), an outline of the manufacture process and a time schedule. (author)

[en] The W7-X basic device is presently being assembled at the Greifswald branch of IPP. The specific field configurations of this helical advanced stellarator are realised by a symmetric arrangement of 50 non-planar and 20 planar superconducting coils. In order to sustain the large electromagnetic forces and moments, all coils are bolted to a massive coil support structure and supported against each other by inter-coil support elements. Cooling of superconductor and the casing is provided by supercritical helium. For all coils the same cable-in-conduit conductor is used. This conductor is formed by a NbTi cable which is co-extruded in an aluminium jacket. Low-resistive electrical joints connect the conductor layers within a winding package and potential break provide electrical insulation of the helium pipes. After insulation and vacuum pressure impregnation, the winding packages are embedded in stainless steel casings, which are then finish-machined and equipped with cooling pipes. During a rapid shut-down of the magnet system the windings may experience voltages up to several kilovolts. High voltage tests under degraded vacuum conditions (Paschen tests) provide a sensitive method to detect weak points in the electrical insulation. Manufacture of the magnets is in a well advanced stage. All winding packages are completed, many of them are integrated in the casings and several coils have already been delivered for cold testing. These tests are performed in a cryogenic test facility at CEA Saclay. Tests at nominal operating conditions and quench tests confirmed the electric layout and the specified margin. Design changes have been implemented during fabrication due to more detailed structural analyses. Some manufacturing processes had to be modified and re-qualified to allow repair of weaknesses defects found during tests. The presentation will give an overview of the production status of the superconducting coils, the experiences gained during fabrication of the superconductor, the winding packages, the steel casings and during assembly, as well as of the results of the cryogenic tests. (author)

[en] In February, 1992, JET entered a major shutdown for the installation of a pumped divertor. The new components inside the vessel are listed. These components have been designed to withstand the computed quasistatic forces during normal operation and the dynamic forces expected during upset condition. The higher degree of instability of the divertor plasma necessitated the enhancement of the stabilization system. The assessment of forces during disruption and vertical displacement events is mainly based on the simulation with the 2D equilibrium evolution codes PROTEUS and MAXFEA. The experience of JET operation is taken into account. The circumstances in previous operation are explained. The maximum possible vertical force acting on the vessel/divertor coil assembly during a VDE can be predicted approximately. It is explained. The guideline applied to force estimation and the examples of local force assessment for the poloidal limiter, the saddle coils, the in-vessel pipework and so on are reported. From the measurement and simulation, the largest global forces at the vessel are mainly caused by halo current. (K.I.)

[en] 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

[en] The Swiss contribution to the international Large Coil Task was a D-shaped 2.5 x 3.5 m . m bore, superconducting toroidal field coil of 8 T. The conductor was made from copper stabilized fully transposed NbTi filaments, cabled in three stages and cooled with pressurized supercritical helium. The testing of the six coils was successfully finished last year at Oak Ridge National Laboratory. Results of the single-coil and the full-array Standard-I tests were reported earlier. This paper presents the data of the ac loss measurements taken during the Standard-II tests. The results are compared with predictions made while designing the coil. Next the results of the Extended Condition tests are discussed. Finally the recently analyzed quench data, taken during the whole test period, are presented

[en] As ITER starts the construction phase, the magnets are one of the items on the critical procurement path. The basic design accepted by the ITER participants dates from 2001. The first step in the release of the various procurement packages has been the completion of critical R and D to confirm the design performance as well as a thorough review into the design solutions being proposed (including changes introduced since 2001). The baseline ITER design is by now quite well known, with 18 Toroidal Field (TF) D shaped coils storing 44 GJ of magnetic energy at fields close to 12 T, a Central Solenoid (CS) stack of 6 modules operating up to 13 T and 6 large Poloidal Field (PF) coils at 6 T. Correction of manufacturing and assembly errors in the magnet systems is provided by a set of 18 low field Correction Coils (CC). The coils use cable-in-conduit (CIC) conductor with both Nb3Sn and NbTi superconducting strands cooled by supercritical helium. The low temperature allows the cryogenic strength of structural steels to be exploited with primary stresses approaching 700 MPa in compression. Operating voltages on the coils are in the 10 kV range with insulation designed in the 20-30 kV range to allow a good margin for possible fault conditions. To allow a focused design review, the recent work has concentrated on design areas with controversial or novel features, or on components where performance verification is incomplete. This work is now nearing completion. Areas selected for review, sometimes without resulting in design changes, are (i) the Nb3Sn conductor design and the superconducting performance degradation seen in some recent test samples; (ii) the TF coil windings and the use of a winding configuration to provide insulation redundancy; (iii) the magnet structures, the material requirements compared to the available manufacturing capacity and optimisation to reduce them; (iv) the finalising of the structural design criteria with appropriate design margins against possible failure mechanisms; (v) features that reduce the risk of the machine becoming inoperable through electrical failure. The work, and the resulting final design, will be summarised in the paper (orig.)

[en] Procurement of the ITER magnets is due to start at the end of 2007/early 2008, with the launch of the longest lead time items, the Nb₃Sn conductor and the TF coil windings. The base design for procurement was established in 2001, and the build up of the Cadarache ITER team has been accompanied by a review of the most critical, or controversial, features of the 2001 design. At the same time, an urgent R and D program has been launched to complete the necessary verification of the design solutions that are proposed. In this paper an overview will be presented of the main design features and drivers, and some of the recent issues and R and D results will be summarized. (authors)

[en] 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