The SPARC Toroidal Field Model Coil Program

Voltage–temperature characteristics in model coils using a Nb3Sn cable-in-conduit conductor for ITER

Central Solenoid (CS) and Toroidal Field (TF) model coils have been developed in order to establish technologies required to the ITER magnets. All of the development goals have been achieved, including the operation at 13 T, 46 kA and pulsed charging at 1.2 T/s in the CS model coil and the stable operation at 10 T, 80 kA under the ITER relevant stress condition in the TF model coil. Based on these results, further activities are now being performed, in extensive collaboration with industry, to optimize coil designs, to improve performances and to mitigate technical risks during the manufacture of the ITER magnets. One of the activities is to establish production capability of Nb3Sn strands having higher critical current density which was redesigned based on the detailed assessment of the conductor performance in the model coil experiments. Strands fulfilling the ITER requirement have successfully been obtained. Another set of activities are full scale manufacturing demonstrations on jacket material for CS and structural materials for TF coil. So far, large forged blocks of JJ1 and ST316LN and thick hot rolled plates of ST316LN were successfully produced as a trial fabrication for the TF coil structures. Experience in the TF model coil indicated a need to improve manufacturing processes for radial plate and a significant progress has also been obtained in this area. Achievements from these activities will provide a firm technical basis for ITER magnet construction including that for industry

Techniques for high field generation have been advancing. No-insulation (NI) rare-earth barium copper oxide (REBCO) pancake coils enable high field generations with high thermal stability. One of the major challenges in REBCO magnet development is the screening current reduction. The screening current is induced in the REBCO tape by the change of the field perpendicular to the REBCO tape surface. It deteriorates the field, sometimes together with mechanical stability. Since the behaviors of the screening current are clear by simulations, it is important to consider the screening current effect in the design stage to reduce it or control the coil performances. In this paper, we explore the effectiveness of the non-energized coil (hereafter called “dummy coil”) to modify the screening current distribution. The dummy coils are wound employing the NI winding technique. The electromagnetic behaviors of the insert NI REBCO magnet that consists of 12 stacked single pancake coils with a few dummy coils are investigated through the electromagnetic simulation with the finite element method (FEM). The results show that the dummy coils can suppress the variance of the current density distribution, although there is a trade-off with respect to field reduction.

The SPARC Toroidal Field Model Coil (TFMC) is the first large-scale (∼3 m), high-field (∼20 T) superconducting fusion magnet based on Rare Earth Yttrium Barium Copper Oxide (REBCO). Its objective was to retire risk for the toroidal field magnet in the SPARC tokamak, a burning plasma class magnetic confinement fusion energy device. Weighing 10,058 kg and utilizing 270 km of REBCO, the TFMC is a non-insulated, stack-in-plate style superconducting magnet. It has three main components: (1) the winding pack; (2) the structural case; and (3) the case extensions, or plena. The winding pack is composed of sixteen single pancakes with two termination plates top and bottom. The pancakes are Nitronic 40 radial plates machined with spiral channels on one side for the REBCO tape stack and single-pass channels on the opposite side for supercritical helium coolant. After assembly, each pancake undergoes a vacuum-pressure impregnation solder process to provide good mechanical protection of the REBCO tape stack and efficient thermal and electrical connectivity within each pancake. The pancakes are bolted along the inner and outer perimeter to provide mechanical and thermal connectivity while inter-pancake joints provide low resistance current transfer between pancakes. The top and bottom termination plates facilitate electrical connection to a superconducting feeder system. Embedded throughout the winding pack are 211 voltage taps, 34 temperature sensors, 34 helium flow monitors, 4 Hall probes, and 4 resistive surface heaters. The winding pack is contained within a structural case, a “trough and lid” style design composed of two Nitronic 50 forgings machined to shape and bolted together. The case reacts the large electromechanical stresses approaching 1 GPa during operation and serves as a pressure vessel that enables 20 bar supercritical helium flow that cools the winding pack and case. Two case extensions or “plena” are attached to the case with unique high-pressure feedthroughs to provide winding pack access for current, cooling, and instrumentation, completing the magnet assembly.

The no-insulation (NI) winding technique has been attracting attention, because NI Rare-Earth Barium Copper Oxide (REBCO) pancake coils have high thermal stability. It is an indispensable technology to generate an ultra-high magnetic field. In 2017, using the NI winding technique a world-record high magnetic field, 45.5 T, was generated by 12 insert single pancake coils with an outsert magnet, and it showed a high potential to generate 14.4 T inside a background field of 31.1 T. After the experiment, the REBCO tapes were mechanically damaged, so that the critical currents were deteriorated. A large current was induced in NI REBCO coils next to the quenched coil when one of multi-stacked NI REBCO pancake coils quenched, resulting in tape property degradation and irreversible mechanical damage. Therefore, in order to protect NI REBCO pancake coils during quench, it is desired to reduce the amount of induced current. To suppress an induced current, the idea of “magnetic dam” has been proposed previously. The idea is to use a copper pipe installed at the outside of NI REBCO pancake coils. However, the copper pipe just slowed the quench propagation. In this paper, we extended the method to decrease an induced current much more by installing extra NI REBCO windings instead of the copper pipe. The electromagnetic and stress behaviors of 6-stacked NI REBCO coils with extra windings are simulated when one of the stacked NI REBCO coils transitions into a normal state. The ability of mechanical damage protection from a strong stress by an induced current during quench propagation is demonstrated through simulations.

The tests of the toroidal field model coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentrum Karlsruhe, Germany. Operation reached a combined 80 kA in the TFMC and 16 kA in the LCT coil, resulting in a peak electromechanical load very close to that expected in the full-size ITER TF coils (800 kN/m). Here we concentrate on the measurements of the current sharing temperature T/sub cs/ of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature T/sub in/ in double pancake DP1, resulting in an increasing normal voltage V across the DP1.1 and DP1.2 conductors, and were repeated for several combinations of currents in the TFMC and in the LCT coil. The analysis of the V - T/sub in/ characteristic by means of the M&M code allows to self-consistently deriving an estimate of T/sub cs/, as well as an indirect assessment of the average strain state in the conductor. The TFMC isolated strand has also been very recently characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from Summers scaling used in the analysis so far. As a consequence, the performance of the TFMC conductor, as preliminarily re-evaluated here, appears more in line with the strand performance than in previous analysis, although a BI-dependent degradation is still present.

A superconducting Toroidal Field Model Coil (TFMC) was developed by the European Home Team within the framework of the R&D program of the International Thermonuclear Experimental Reactor (ITER) and is presently tested in the TOSKA facility of the Forschungszentrum Karlsruhe. To withstand the Lorentz forces the 98 turns of the ITER TFMC are divided into 10 pancakes, embedded in five stainless steel (SS) radial plates and enclosed in a grounded SS case. To minimize the voltage stress of the conductor insulation the potentials of all five inner pancake joints are connected to their own radial plates. This insulation concept offers the possibility to test three kinds of insulation types: the conductor insulation between the conductor and the surrounding radial plates, the radial plate insulation between two adjacent radial plates, and the ground insulation between all radial plates connected with the conductor and the grounded case. The rated terminal to terminal voltage is 9.8 kV for a symmetrical grounding. Based on the rated voltage, the insulation design and the manufacturers' specifications, a DC and AC test program is executed after each set of defined insulation stresses on every type of insulation at room and cryogenic temperatures. In addition, partial discharge measurements are performed as a nondestructive diagnostic tool for the observation of the status of electrical insulation. It is well known from conventional high-voltage engineering that large electrical systems must be designed to withstand internal overvoltages caused by fast rise times of the applied voltage. Therefore calculations were performed, both in time and frequency domains, to describe the transient electrical behavior of the ITER TFMC. To prove the insulation strength and to verify the calculations, high voltage impulse tests with fast rise times (9.8 kV peak) and low voltage high frequency tests are performed.

No-insulation(NI) Rare-Earth Barium Copper Oxide (REBCO) pancake coils can generate high magnetic fields and have the high thermal stability against normal-state transition due to allowing currents to radially bypass a local normal-state region through the turn-to-turn contacts. Despite these merits, no turn-to-turn insulation causes a charging delay. To solve this charging delay problem, an NI REBCO coil wound with multi-bundled (MB) conductors was proposed. As an experimental result, a charging delay was improved because the inductance per tape of MB coil is smaller than that of single-tape NI coil. Meanwhile, the current behaviors of MB coils are complicated and has not been clarified. We have investigated the current and thermal behaviors of MB REBCO coils during charging and against normal-state transition, comparing with and without turn-to-turn insulation. As the result, it is shown that the MB coil can be operated stably, and no pessimistic temperature rise is observed. However, the operating current of MB coils with turn-to-turn insulation may concentrate on one tape when the normal-state transition occurs. Therefore, MB coils without turn-to-turn insulation seems to be desirable.

ITER toroidal field model coil test: analysis of heat transfer from plates to conductors

The screening current induced in rare-earth barium copper oxide (REBCO) tape generates an unwanted irregular magnetic field. The screening current-induced field (SCIF) is a challenging issue for MRI, NMR, and accelerators magnet composed of REBCO coils. A few FEM-based simulation methods have been proposed to estimate the SCIF; however, they require a long computation time.Recently, we have proposed a simple SCIF computation method based on the self and mutual inductances of REBCO pancake coils and screening current radial paths on the top and bottom of pancake coils. The accuracy of the proposed method is not excellent; however, the computation time is quite short. In this paper, we report an equivalent circuit model that includes the self and mutual inductances of a REBCO pancake coil and screening current radial path. Moreover, with this proposed method, we can compute the SCIF of no-insulation (NI) REBCO pancake coils, which is not the case with the previously proposed FEM-based simulation method. The proposed method has been validated by experimentation. The proposed method is available online.

Calculations of pressure drop and mass flow distribution in the toroidal field model coil of the ITER project

With the rapid advancement of magnetic confinement fusion technology, High- Temperature Superconductors (HTS) have emerged as a cornerstone for compact and efficient tokamak systems due to their exceptional current-carrying capacity under high magnetic fields. Against this backdrop, Energy Singularity Fusion Power Technology (ES Company) initiated the JingTian (JT) magnet project in December 2024 to validate the Toroidal Field (TF) magnet design for its next-generation all-HTS tokamak facility, HH170. Prior to the HH170 project, ES Company developed and constructed the world's first all-HTS tokamak. This facility successfully achieved its first plasma operation in June 2024, marking a significant milestone in fusion technology [1, 2]. This paper details the design, fabrication, and performance testing of the JT magnet, a large-scale D-shaped winding pack comprising 32 modular Rare-earth Barium Copper Oxide (ReBCO)-based pancake coils. Operating at 5 K with supercritical helium cooling, the JT magnet achieved a record-breaking peak magnetic field of 21.7 T, the highest reported for an all-HTS tokamak magnet. Steady-state operation at 24.3 kA confirmed its structural integrity, cryogenic cooling efficiency, and alignment with magnetic field modeling predictions. The success of the JT magnet not only provides critical data for the HH170 design but also establishes a foundational milestone for the engineering application of high-field compact tokamaks, demonstrating significant potential to reduce fusion reactor size and cost.

Electromagnetic characteristics of NI-REBCO coils according to contact surface conditions controlled by the oxidation state and roughness of the REBCO conductor

The superconducting coil systems for the International Thermonuclear Test Reactor (ITER) will consist of 24 toroidal field (TF) coils at 488 tonnes each, a central solenoid (CS) at 964 tonnes, and an external poloidal field coil set of 6 coils, the largest of which has a diameter of 30 m. The TF and CS coils will operate at a maximum field at the windings of 13 T and will use Nb/sub 3/Sn conductor in cable-in-conduit form. The TF coils and the CS are layer wound. The design features of each of these systems are described. >

The HT-7U tokamak is a magnetically-confined full superconducting fusion device, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coils. These coils are wound with cable-in-conductor (CICC) which is based on UNK NbTi wires made in Russian [1]. A single D-shaped toroidal field magnet coil will be tested for large and expensive magnets systems before assembling them in the toroidal configuration. This paper describes the layout of the instrumentation for a superconducting test facility based on the results of a finite element modeling of the single coil of toroidal magnetic field (TF) coils in HT-7U tokamak device. At the same time, the design of coil support structure in the test facility is particularly discussed in some detail.