E. Salpietro

Results of a New Generation of ITER TF Conductor Samples in SULTAN

A new generation of ITER TF conductor samples has been assembled and tested in SULTAN in 2007 following a common procedure agreed among the ITER parties. The test results of six SULTAN samples, made of twelve conductor sections manufactured in Europe, Japan, Korea and Russia, are reported here. The conductor layout reflects the ITER TF conductor design, with minor differences for the Nb3Sn strand characteristics, void fraction and twist pitch. The object of the test is a straight comparison with the ITER requirement of 5.7 K current sharing temperature at 68 kA current and 11.3 T field. A broad range of behavior is observed.

Test Results of Two European ITER TF Conductor Samples in SULTAN

Four conductor lengths were prepared according to the ITER TF conductor design and assembled into two SULTAN samples. The four lengths are not fully identical, with variations of the strand supplier, void fraction and twist pitch. Lower void fractions improve the strand support and increased twist pitches also lower the strand contact pressure but both tend to increase the AC loss and the lower void fraction also increases the pressure drop so that the mass flow rate in the strand bundle area of the cable is reduced. The assembly procedure of the two samples is described including the destructive investigation on a short conductor section to assess a possible perturbation of the cable-to-jacket slippage during the termination preparation. Based on the DC performance and AC loss results from the test in SULTAN, the impact of the void fraction and twist pitch variations is discussed in view of freezing the ITER conductor design and large series manufacture. A comparison with the former generation of conductors, using similar strands but based on the ITER Model Coil layout, is also carried out. The ITER specifications, in terms of current sharing temperature, are fulfilled by both samples, with outstanding results for the conductor with longer twist pitches.

Test Results of Two ITER TF Conductor Short Samples Using High Current Density Nb

Two short length samples have been prepared and tested in SULTAN to benchmark the performance of high current density, advanced Nb3Sn strands in the large cable-in-conduit conductors (CICC) for ITER. The cable pattern and jacket layout were identical to the toroidal field model coil conductor (TFMC), tested in 1999. The four conductor sections used strands from OST, EAS, OKSC and OCSI respectively. The Cu:non-Cu ratio was 1 for three of the new strands, compared to 1.5 in the TFMC strand. The conductors with OST and OKSC strands had one Cu wire for two Nb3Sn strands, as in TFMC. In the EAS and OCSI conductors, all the 1080 strands in the cable were Nb3Sn. A dc test under relevant load conditions and a thermal-hydraulic campaign was carried out in SULTAN. The CICC performance was strongly degraded compared to the strand for all the four conductors. The current sharing temperature at the ITER TF operating conditions (jop = 286 A/mm2, B = 11.15 T) was lower than requested by ITER.

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Following the outcome of the conceptual design phase the EFDA dipole magnet will be made of rectangular cable-in-conduit conductors (CICC) jacketed in 316LN. In order to optimize the required amount of superconductor two different conductor types are used: a high-field (HF) conductor consisting of 144 strands and a low-field (LF) conductor with 108 strands. A high strand with a critical current density (at 4.2 K and 12 T) and an effective filament diameter of was selected. The first series of conductor prototype specimens was tested in summer 2006 but the conductor performances were lower than expected from the pre-prototype tests of 2005 and not fulfilling the design criteria. The conductor layouts were modified to increase the strand support inside the cable and the revised HF conductor design was qualified successfully end of 2006. A current sharing temperature 6 K was found at the dipole operating conditions (12.8 T, 17 kA) confirming the required temperature margin of more than 1 K. The HF conductor qualification process including the design modifications, analysis of the test results and comparison to the expectations are discussed.

Sn Strands

A 12.5 T superconducting dipole magnet (European DIPOle, EDIPO) has been designed by EFDA and it is now being procured within the framework of the European Fusion Programme in order to be installed in CRPP-PSI. This saddle-shaped magnet is designed to reach 12.5 T in a 100 times 150 mm rectangular bore over a length of about 1.5 m in order to test full size conductor samples that shall be produced during the ITER magnets procurement. The magnet uses Cable In Conduit Conductor (CICC) technology and the cables are made of high Jc (about 2300 A/mm2 at 4.2 K, 12 T) superconducting strands. In this paper the main magnet parameters are given together with the key supporting electromagnetic, mechanical and thermal analyses. An update on the general status of the procurement of the strand, conductors, dipole magnet and facility is also given together with the key results of the on-going supporting R&D.

Development of the EFDA Dipole High Field Conductor

Within the framework of the European Fusion Programme a FSJS has been designed and manufactured by European Industry using PF coil NbTi superconductor manufactured and supplied by the Russian Federation as part of the R&D for the PF Conductor Insert (PFCI) coil. In addition to the superconductor, this sample contains a number of unique features. In contrast to other samples previously manufactured in Europe, the FSJS has used the Central Solenoid Model Coil (CSMC) joint design with NbTi conductor and a thick square jacket. One leg of the FSJS has had the conductor and sub-petal steel wraps removed before jacketing to assess the difference in the conductor performances between the two configurations. This paper will report on the development and manufacture of the FSJS, in particular the use of the Central Solenoid jacketing and swaging tools for compaction of the conductor and swaging of the joints, the preparation and qualification of the manufacturing route for the joint, the jacketing of a special length without conductor wraps and the high level of instrumentation required for the testing of the joint. The sample has been instrumented with more sensors than any other previous European sample, including temperature sensors, a large number of voltage taps for quench detection and Tcs measurements, quadrupoles to detect uneven voltage distribution, hall arrays for current distribution measurements and saddle coils.

Design and Procurement of the European Dipole (EDIPO) Superconducting Magnet

The CS Insert Coil (CSIC), a well-instrumented 140 m long Nb3Sn solenoid wound one-in-hand and installed in the bore of the CS Model Coil, was tested during the summer of 2000 at JAERI Naka, Japan, within the framework of the International Thermonuclear Experimental Reactor large projects [1]. The maximum transport current in the CSIC was 40 kA and the peak background field was 13 T. The coils were cooled by forced flow HeI nominally at 4.5 K and 0.6 MPa. An inductive heater was used to test stability and quench of the CSIC. In this second of two companion papers we concentrate on the analysis of quench initiation and propagation, based on the study of heater calibration and conductor stability presented in the first paper [2]. The initiation and propagation of an inductively driven quench was tested here for the first time in a two-channel Nb3Sn conductor, for different transport currents, delay times of the dump, and temperature margins, and a selection of the corresponding results will be presented and...

Design and manufacture of a full size joint sample (FSJS) for the qualification of the poloidal field (PF) insert coil

One of the design features which yet offers interesting margins for performance optimization of cable-in-conduit conductors (CICCs), is their geometry. For relatively small size Nb3Sn CICCs, operating at high electromagnetic pressure, such as those for the EDIPO project, it has been experimentally shown that a design based on a rectangular layout with higher aspect ratio leads to the best performance, especially in terms of degradation with electromagnetic loads. To extend this analysis to larger size Nb3Sn CICCs, we manufactured and tested, in the SULTAN facility, an ITER toroidal field (TF) cable, inserted into a thick stainless steel tube and then compacted to a high aspect ratio rectangular shape. Besides establishing a new record in Nb3Sn CICC performances for ITER TF type cables, the very good test results confirmed that the conductor properties improve not only by lowering the void fraction and raising the cable twist pitch, as already shown during the ITER TFPRO and the EDIPO test campaigns, but also by the proper optimization of the conductor shape with respect to the electromagnetic force distribution. The sample manufacturing steps, along with the main test results, are presented here.

INDUCTIVELY DRIVEN TRANSIENTS IN THE CS INSERT COIL (II): QUENCH TESTS AND ANALYSIS

The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in the EU and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, /spl sim/45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of /spl sim/6 T, and it will be cooled by supercritical He at /spl sim/4.5 K and /spl sim/0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure gauges, strain and displacement sensors. The test program will be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermal-hydraulic characterization. Here we give an overview of the preparatory work toward the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far.

Successful performances of the EU-AltTF sample, a large size Nb3Sn cable-in-conduit conductor with rectangular geometry

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.