R.K. Maix

Progress of the ITER central solenoid model coil programme

The worlds largest pulsed superconducting coil was successfully tested by charging up to 13 T and 46 kA with a stored energy of 640 MJ. The ITER central solenoid (CS) model coil and CS insert coil were developed and fabricated through an international collaboration, and their cooldown and charging tests were successfully carried out by international test and operation teams. In pulsed charging tests, where the original goal was 0.4 T/s up to 13 T, the CS model coil and the CS insert coil achieved ramp rates to 13 T of 0.6 T/s and 1.2 T/s, respectively. In addition, the CS insert coil was charged and discharged 10 003 times in the 13 T background field of the CS model coil and no degradation of the operational temperature margin directly coming from this cyclic operation was observed. These test results fulfilled all the goals of CS model coil development by confirming the validity of the engineering design and demonstrating that the ITER coils can now be constructed with confidence.

First test results for the ITER central solenoid model coil

Abstract The largest pulsed superconducting coils ever built, the Central Solenoid (CS) Model Coil and Central Solenoid Insert Coil were successfully developed and tested by international collaboration under the R&D activity of the International Thermonuclear Experimental Reactor (ITER), demonstrating and validating the engineering design criteria of the ITER Central Solenoid coil. The typical achievement is to charge the coil up to the operation current of 46 kA, and the maximum magnetic field to 13 T with a swift rump rate of 0.6 T/s without quench. The typical stored energy of the coil reached during the tests was 640 MJ that is 21 times larger than any other superconducting pulsed coils ever built. The test have shown that the high current cable in conduit conductor technology is indeed applicable to the ITER coils and could accomplish all the requirements of current sharing temperature, AC losses, ramp rate limitation, quench behavior and 10 000-cycle operation.

Mechanical behavior of the ITER TF model coil ground insulation system after reactor irradiation

Abstract The mechanical properties of glass fiber reinforced plastics (GFRPs) suggested for the turn and ground insulation of the ITER toroidal field (TF) coils are subject to extensive investigations with respect to their design requirements at present. The insulation system used for the ITER TF model coil, manufactured by European industry, consists of a boron-free R-glass fiber reinforced tape, vacuum-pressure impregnated in a DGEBA epoxy system and partly interleaved with polyimide-foils (e.g. Kapton-H-foils). In order to assess the material performance under the actual operating conditions of ITER-FEAT, the system was irradiated in the TRIGA reactor (Vienna, Austria) to neutron fluences of 5×1021 and 1×1022 m−2 (E>0.1 MeV). The composite was screened at 77 K using static tensile, short-beam-shear (SBS) as well as double-lap-shear tests prior to and after irradiation. Furthermore, tension–tension fatigue measurements were done in order to simulate the pulsed ITER-FEAT operation. We observe that the mechanical strength and the fracture behavior of these GFRPs after irradiation are strongly influenced by the three factors: the winding direction of the tape, the quality of fabrication and the delamination process.

Influence of reactor irradiation on the mechanical behavior of ITER TF coil candidate insulation systems

Extensive material tests have to be performed in order to obtain information on the radiation induced change in the mechanical behavior of insulating materials for the ITER Toroidal Field (TF) coil. The investigated insulation systems are R-glass fiber reinforced tapes, vacuum impregnated with a DGEBA epoxy resin and interleafed with Kapton H-foils. According to the actual operating conditions of ITER-FEAT, the systems were irradiated in the TRIGA reactor (Vienna, Austria) to neutron fluences of 5 × 10 21 and 1 x 10 22 m -2 (E > 0.1 MeV). Static tensile, short-beam-shear (SBS) as well as double-lap-shear (DLS) tests were carried out at 77 K prior to and after irradiation. Furthermore, results on swelling and weight loss as well as on the material properties under tension- tension fatigue loading conditions are presented.

Design and manufacture of the Poloidal Field Conductor Insert coil

Abstract The Poloidal Field (PF) coils of ITER (International Thermonuclear Experimental Reactor) will supply the necessary magnetic field to initiate, shape, control and shutdown burning plasmas. The PF coils use NbTi cable-in-conduit superconductors, which operate at maximum currents of the order of 45–60 kA and experience large variations of current and magnetic fields. In order to test full-scale NbTi superconductors at operational conditions similar to ITER, the European Team has been asked to design and manufacture a PF conductor insert (PFCI). The cable has been provided by the Russian Federation. The Insert will be tested in 2004 in the Central Solenoid Model Coil (CSMC) facility at JAERI Naka, Japan.

Fabrication of the first European full-size joint sample for ITER

The European Home Team is responsible for the design, the fabrication and the test of the Toroidal Field Model Coil (TFMC) of the ITER project. Within this task, three full-size samples have to be fabricated in industry and tested. Each sample is made of two parallel straight bars of full-size conductor, connected at the bottom through a joint designed according to the EU proposal for the ITER coils, and having at the top, two connections for the test facility. The first sample uses an ITER type Nb/sub 3/Sn cable-in-conduit embedded in a thick steel square jacket, the lower joint being similar to the inner joints of the TFMC. The second sample uses the TFMC conductor which is an ITER type Nb/sub 3/Sn cable-in-conduit embedded in a thin steel circular jacket, the lower joint being similar to the outer joints of the TFMC. Last, the third sample will use the same cable embedded in a thin incoloy jacket, fully relevant to the ITER TF coils. This paper reports on the fabrication by Ansaldo of the first sample, emphasizing the differences between the original design and the final design produced after preliminary trials, as well as pointing out the main technical issues.

Completion of the ITER Toroidal Field Model Coil

In the scope of the ITER EDA a Toroidal Field Model Coil (TFMC) has been manufactured accompanied by a thorough Quality Assurance (QA) test program. This large superconducting coil has been conceptually designed by the ITER European Home Team (EUHT) and manufactured by European industry. The coil is being completed and will be tested at the Forschungszentrum Karlsruhe in spring 2001. The race track shaped winding is made of a cable-in-conduit conductor in a circular 316LN stainless steel jacket. From this conductor five double pancake (DP) modules were fabricated. Results of conductor and DP manufacture were already presented at previous conferences and are therefore only summarized here. The paper concentrates on the subsequent manufacturing steps, namely the stacking of the DP modules, the insulation and impregnation of the winding pack, the outer joint manufacture by electron beam welding, the assembly of the winding pack with the stainless steel case, the mounting of the helium pipes, the sensors and the busbars. To assemble the coil into the TOSKA facility and to fit it to the EU-LCT coil a heavy Inter-Coil Structure (ICS) has been built, in which the TFMC will rest on four wedges.

The ITER Toroidal Field Model Coil (TFMC)

Abstract The TF coils for ITER (ITER, Final Design Report, December 1997) will use the concept of a circular thin walled Nb3Sn cable in conduit superconductor completely enclosed in an insulated groove in steel plates to form the coil pancakes. These are then stacked together to form the winding pack supported by a steel case. The concept is being demonstrated by the fabrication of a TFMC (E. Salpietro et al., Construction of a Toroidal Field Model Coil (TFMC) for ITER, MT-15 Fifteenth International Conference on Magnet Technology, Oct 20–24, 1997, Beijing, China; R. Maix et al., Manufacture of the ITER TF model coil (TMFC), Poster P1-217) with a bore of 1.4×2.5 m, a peak field of 8.77 T and a total current of 7.8 MAT. The coil will be tested during 1999 in the TOSKA facility at FZK, Karlsruhe, using the EURATOM LCT coil to provide an external field system. The objectives of the TFMC are as follows: 1. to develop and verify the full scale TF coil manufacturing techniques, in particular the following features: ◦ plate manufacturing (forming the grooves); ◦ fitting the conductor in the groove after heat treatment and insulation (i.e. predictable geometry change); ◦ closing the groove with a cover plate and plate insulation; ◦ fitting the winding into the case, gap filling and case closure; 2. to establish realistic manufacturing tolerances; 3. to bench-mark methods for the ITER TF coil acceptance tests, including insulation and impregnation process; 4. to obtain information on the coil’s mechanical behaviour, operating margins and in-service monitoring techniques, particularly for the insulation quality over fatigue cycles.

Design, Production and QA Test Results of the NbTi CIC Conductors for the W7-X Magnet System

The magnet system of the W7-X Stellarator consists of 50 non-planar and 20 planar coils. The superconductors for both types have the same design, which is a Cable-in-Conduit (CIC) conductor of 243 NbTi strands. Namely the non-planar coils asked for a conductor, which can be easily wound to a complex shape, but has high mechanical strength to resist the large magnetic forces acting during magnet operation. This has been achieved by jacketing the cable into an Aluminium conduit by a co-extrusion process. The used 6063 Aluminium alloy is very soft in the extruded state, but gets high strength properties after winding during a precipitation hardening treatment at about 160°C. The production of the needed 390 conductor lengths (including spares) and the related QA tests are nearly completed and a large number of coils are fabricated. Some of them were already subjected to a cold test at CEA Saclay, where the conductor behaved as expected from short sample measurements.

Irradiation and mechanical testing of ITER relevant magnet insulation

In the scope of ITER Engineering Design Activities (EDA) the European Home Team performs an irradiation test program on industrially fabricated specimens of magnet insulation. At the unique test facility available at the research reactor of the Technical University of Munich, the shear compression test samples were irradiated at 5 K and tested without warming up. The samples, provided by Japanese, US and European industry, were fabricated using different types of epoxy resin, but the same boron-free glass fabrics inter-leafed with Kapton foil. The results are presented and assessed. The Munich test facility is no longer available. A complementary test program carried out at the research reactor of the Atomic Institute in Vienna will form the basis of future irradiation work. It is focused on the assessment of intrinsic material parameters and takes advantage of a new fatigue testing machine allowing operation at 77 K. Links to the Munich program were made by irradiating identical samples in both reactors (at 5 K in Munich and at ambient temperature in Vienna). Special sample geometry was developed and scaled versus standards for tensile and inter-laminar shear testing under cyclic load. Results on the static and dynamic properties are presented.