Correction Coil Conductor Research Articles

Overview of verification tests on AC loss, contact resistance and mechanical properties of ITER conductors with transverse loading up to 30 000 cycles

The ITER magnet system uses cable-in-conduit conductor (CICC) technology with individual strands twisted in several stages resulting in a rope-type cable, which is inserted into a stainless steel conduit. The combination of high current (up to 68 kA) and background magnetic field (up to 13 T) results in large transverse Lorentz forces exerted on the conductors during magnet system operation. The high transverse forces, accompanied with the cyclic nature of the load, have a strong influence on the conductor properties. The Twente Cryogenic Cable Press is used to simulate the effect of the Lorentz forces on a conductor comparable to the ITER magnet operating conditions. An overview is presented of the AC coupling and hysteresis loss, mechanical deformation characteristics and inter-strand contact resistance measurement results obtained on full-size ITER CICCs measured in the Twente Cryogenic Cable Press. The aim of this work is to characterize conductors’ electromagnetic and mechanical properties during cycling of the load up to 30 000 cycles. The evolution of the magnetization (AC coupling loss time constant nτ), mechanical properties and inter-strand resistance Rc between selected strands is presented along with loading history. The Rc between first triplet strands is also measured as a function of applied load. It is shown that transverse load cycling has a strong influence on the CICC properties. An overview of the results for eight toroidal field conductors, two central solenoid conductors, three poloidal field conductors of different types (PF1&6, PF4, PF5), one main bus-bar and one correction coil conductor is presented.

Conductor Performance of CCCN3, CCCN4 and CCCN5 Samples for ITER CC Coils

The conductor for the ITER correction coils (CCs) is manufactured in China and is tested in the SULTAN facility. The third China CC conductor for the qualification SULTAN test sample (CCCN3) was prepared from a 200-m-long CC conductor unit length (Phase II). The fourth and fifth China CC conductor samples for SULTAN testing (CCCN4 and CCCN5) were prepared from the 576-m-long production unit lengths (Phase III). The current sharing temperature $T_{cs}$ test results show that the sections of CCCN3, CCCN4, and CCCN5 have high $T_{cs}$ performance. Using the electrical method, the $T_{cs}$ was 7.13 K for CCCN3, 7.10 K for CCCN4, and 7.09 K for CCCN5 at the ITER reference test conditions, i.e., at ${B}_{\mathrm{sultan}}=3.7 \text{T}$ and current equal to 10 kA $({B}_{\mathrm{peak}}=4.0 \text{T})$ . Even after the 2000 electromagnetic load cycles for CCCN3 conductor and 1000 electromagnetic load cycles for CCCN4 and CCCN5, the $T_{cs}$ performance of all samples did not changed appreciably. Based on the CC conductor sample test results at 10 kA and at different fields, the $T_{cs}$ was extrapolated to the peak field ${B}_{\mathrm{peak}}= 4.1 \text{T}$ and current equal to 10 kA. The extrapolated $T_{cs}$ is 7.07 K for CCCN3, 7.05 K for CCCN4, and 7.03 K for CCCN5. The $T_{cs}$ performance for all conductors is higher than the ITER acceptance criteria of 7.00 K.

Qualification of the ITER Correction Coils and Their Conductor

The ITER correction coils (CCs) are designed to be arranged toroidally around the machine inside the poloidal field coils and mounted on the toroidal field coils in order to compensate field errors arising from misalignment of the coils and winding deviations from the nominal shape as a result of fabrication tolerances, joints, leads, busbars CCs, and assembly tolerances. China is responsible for manufacturing all the CCs and their conductor. The paper presents the qualification progress and results of ITER CCs and CC conductor, including the test results of CC SULTAN samples, double pancakes winding experiments, CC case welding techniques, and manufacturing experiment of CC helium inlet/outlets.

Experimental validation of a lap-type joint AC loss model with an ITER correction coil conductor joint

The ITER correction coils (CC) system features shaking hands lap-type joints to interface the terminations of the conductors. The feasibility of operating plasma scenarios depends on the ability of the magnets to retain sufficient temperature and current margins. In this respect, the joints represent a possible critical region due to the combination of steady-state Joule heating from the resistance of the joint and coupling currents and connected losses in magnetic field ramped operation. Since manufacturing and testing of different joint concepts is demanding from cost and time point of view, a dedicated model has been developed based on the numerical cable model JackPot-ACDC to reproduce the performance of lap-type joints and assess the effect of possible optimization solutions. In this work, the prediction capability of the model is verified through AC loss measurements on a prototype ITER CC joint manufactured at ASIPP. The results of the experiment and model are reported and discussed.

Manufacturing of ITER PF5 and CC Sample Conductors

The design of poloidal field (PF) coils and correction coils (CCs) for the International Thermonuclear Experimental Reactor (ITER) relies on the use of 45-kA NbTi cable-in-conduit conductors. All PF5 and CC conductors are produced in China. Research and development programs are needed to acquire knowledge on the behavior of such conductors. Since the conductors are new, full-size copper dummy conductors have been produced in advance for testing the cabling parameters, the definition of the automatic tungsten-inert-gas welding of a seamless jacket section, the elaboration of cable insertion, compaction, etc. Then, two short qualification conductor samples (the PF5 and the CC) are manufactured at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, China, with the NbTi advanced strands produced by the Western Superconducting Technologies Company, Ltd. In this paper, the manufacturing procedures for the two PF5 and CC conductor samples are described in detail.

Experiment of low resistance joints for the ITER correction coil

A test method was designed and performed to measure joint resistance of the ITER correction coil (CC) in liquid helium (LHe) temperature. A 10 kA superconducting transformer was manufactured to provide the joints current. The transformer consisted of two concentric layer-wound superconducting solenoids. NbTi superconducting wire was wound in the primary coil and the ITER CC conductor was wound in the secondary coil. The primary and the secondary coils were both immersed in liquid helium of a 300 mm useful bore diameter cryostat. Two ITER CC joints were assembled in the secondary loop and tested. The current of the secondary loop was ramped to 9 kA in several steps. The two joint resistances were measured to be 1.2 nΩ and 1.65 nΩ, respectively.

Test Results of ITER Correction Coil Short Samples CCCN1 and CCCN2

This paper presents the first benchmarking test results of ITER correction coil (CC) conductor short samples CCCN1 and CCCN2. The Chinese first CC conductor short sample CCCN1 was prepared and tested at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The second CC conductor short sample CCCN2 was prepared at ASIPP and tested at the SULTAN laboratory. The TCS of the two samples was measured at the peak magnetic field of 4.1 T and at a current of 10 kA. A TCS of 7.02 K was measured at ASIPP and a TCS of 7.20 K was measured at the SULTAN laboratory. The calculated results of the short samples are also presented for a comparison to the measured results in this paper.

Quasi 3D Thermo-Hydraulic Modeling and Conductor Temperatures of the ITER Correction Coils

The ITER magnet system consists of 4 sub-systems: 3 sets of 6 Correction Coils (CC), a Central Solenoid (CS), 6 Poloidal Field coils (PF) and 18 Toroidal Field Coils (TFC). The Correction Coil system aims to correct particular asymmetries and reduce error magnetic fields responsible for locked modes which can induce plasma disruptions. During the plasma scenario, thermal shield radiation, eddy current losses in structures, conduction from the TFC casing and AC losses in the NbTi superconductor will decrease the temperature margin of the CC conductor.

The first DC performance test and analysis of CC conductor short sample at ASIPP conductor test facility

The first DC performance experiments of ITER correction coil (CC) conductor short sample have been carried out in the conductor test facility of Institute of Plasma Physics, CAS (ASIPP) in January this year. Those experiments aim to investigate the DC performance of ITER CC conductor. The tested conductor short sample is bended as a half circle with the diameter of 270mm to meet the background magnetic field shape. The half circle part of sample is longer than the final twist pitch. The current sharing temperature (Tcs) in the 3.86T external magnetic field (Bex), ≤12kA could be measured including the critical current (Ic) run. There is no obvious impact of 1000 cycles on DC performance. Those measured Tcs results are in agreement with the expected results from strand scaling.

Electromagnetic and Stress Analysis of 50 kA Superconducting Transformer for ITER Conductor Test Facility

A superconducting transformer was designed for the correction coils (CC) conductor test facility in the Institute of Plasma Physics, the Chinese Academy of Sciences (ASIPP), for validating the CC conductor of the international thermonuclear experimental reactor (ITER). The maximum current of the superconducting transformer is 50 kA. The conductor samples could be tested at different magnetic fields and cooling conditions similar to the operational condition of ITER. During normal operation, the primary and secondary windings of the superconducting transformer will withstand the high electromagnetic (EM) stress caused by the high current. Integrated analysis of EM and structural fields can ensure these components not to encounter any over-stress problem under normal operational condition. A coupled model is proposed to combine the EM and structural analyses. A detailed finite element (FE) model and EM-structural coupled model are presented and the numerical results show that the stress of the transformer windings is within the allowable limits.

The Design of Test Facility for ITER CC Conductor

A test facility has been designed to measure short sample Ic to verify that the Correction Coil (CC) conductor meets the ITER specifications. This facility features a DC background field up to 10 T at 1.8 K in a bore with diameter of 338 mm. The stored energy is 6 MJ. A 50 kA superconducting transformer is designed as current source. A 500 W at 4.5 K refrigerator is used to provide the supercritical helium in the test. The background magnet and superconducting transformer primary winding are immersed in liquid helium. The transformer secondary winding and the test sample are cooled by supercritical helium.

Manufacturing of 50 kA superconducting transformer for ITER correction coil conductor test

To meet the specifications of International Thermonuclear Experimental Reactor correction coil (CC) conductor, a 50 kA superconducting transformer has been designed and manufactured to provide the short sample of the CC conductor the current. The transformer consists of two concentric layer-wound superconducting solenoids with the primary inside the secondary coil. In order to test the transformer, the two legs of the secondary coil were directly connected by superconducting cables. A 500 W/4.5 K refrigerator was used to provide the supercritical helium. The maximum current of 56.3 kA in the secondary coil loop was obtained.