Toroidal Field Magnet Research Articles

Divertor Tokamak Test facility project: status of design and implementation

Abstract An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, XD divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

An improved analysis method for assessing the nuclear-heating impact on the stability of toroidal field magnets in fusion reactors

An improved analysis method for assessing the nuclear-heating impact on the stability of toroidal field magnets in fusion reactors

The SPARC cryogenic system

The SPARC project at Commonwealth Fusion Systems (CFS) is a tokamak system designed to demonstrate commercially relevant fusion energy and achieve net fusion power output during the first operating campaign, with eventual fusion pulse energies exceeding 1 GJ. The SPARC tokamak includes eight magnet systems, three that use high-temperature superconducting (HTS) tapes. The SPARC cryogenic system (CRYO) consists of three supercritical helium cryogenic loops at nominal temperatures of 8 K, 15 K, and 80 K. CRYO 8 K and 15 K loops cool the HTS magnets to maintain thermal stability and prevent quench, while magnets cooled to 80 K are normally conductive. CRYO provides cold helium using a hybrid system that includes a Brayton-cycle-based cryoplant supporting all CRYO temperature loops, and a fixed volume 8 K blowdown system to remove heat and maintain temperature stability of the toroidal field (TF) magnets during and immediately following fusion pulses. CRYO 4.5 K equivalent peak cooling power is 17 kW for the cryoplant and 2.9 MW for the blowdown system during a 10-second fusion pulse. Primary cryoplant mechanical equipment includes screw compressors, turbo-expanders, heat exchangers, and circulation pumps, while the blowdown system consists of a series of warm and cold helium storage tanks operating at independent temperatures and pressures, with make-up compressors to reset the blowdown system between pulses. CRYO also includes a distribution valve box and multiple vacuum-jacketed (VJ) process lines ranging in nominal diameter from 20 to 600 mm. This paper investigates the various sub-elements which in combination represent SPARC CRYO, and attempts to address some of the technical challenges identified by the team.

DEMO toroidal field coil fast discharge unit integration studies

The Fast Discharge Units (FDUs) of the Superconducting (SC) Toroidal Field (TF) coils in the European demonstration fusion power plant DEMO warrant the machine integrity over its full lifetime against severe failure events, such as SC coil quenches or any other plant events requiring the safe TF magnet system discharge. A low (75 kA) and a high current (105 kA) configuration are under study for the TF coils for DEMO. The FDUs must be extremely reliable for the purpose of commutating in short time (∼1 s) the currents and to discharge the TF magnets safely into resistors outside of the tokamak building. Malfunctions of the FDUs must be avoided. The FDUs are considered as Safety Important Class (SIC) components that need to discharge high amounts of energy of about 161 GJ (@75 kA) resp. 118 GJ (@105 kA) stored in the DEMO TF coils.The TF FDUs Circuit Breakers (CBs) shall be installed in the lower level of the tokamak to minimize the length of the connecting busbars. The FDUs integration is challenging because of the high neutron and gamma radiation and stray magnetic fields of the tokamak.Since in DEMO the neutron fluence over lifetime is much higher than in ITER, the problems of using FDUs with electronic subsystems was expected to be more severe, so that their integration has been considered from the beginning of the DEMO project. Sufficient shielding or possible re-positioning of the whole FDUs or sensitive FDU components compared to ITER are being investigated, to reduce the neutron fluxes and neutron and gamma ray fluences. Alternative concepts, e.g., fully mechanical CBs are studied in the EUROfusion Work Package Plant Electrical System (WPPES) in parallel.This paper presents the CAD integration work on the DEMO TF FDUs supported by neutronics assessments. It is assumed the same FDU technology as in ITER. The magnet feeder´s integration is commenced at the same time.

Design of the cryogenic loop for the superconducting toroidal-field magnets of the Divertor Tokamak Test

The design of the cryogenic circuit for the toroidal field magnets has a relevant role in the overall costs of the DTT cryogenic plant (or cryoplant), hence numerical tools can be useful to investigate and to compare different configurations for the TF cooling circuit in support of the selection of the most suitable one. The system-level cryogenic circuit module of the validated thermal–hydraulic 4C code is adopted in this work to model the DTT TF cooling system. At first, the model of the reference circuit layout is implemented and simulated during the plasma pulsed operation of the DTT machine, highlighting the necessity to reduce the heat load transferred to the refrigerator due to the large power consumption of the cold compressor. In view of the above, different TF circuit layouts and optimization strategies are presented, including mitigation strategies for the smoothing of the peak heat load to the refrigerator, leading to a reduction of the cold compressor power up to the 66% with respect to that computed for the reference TF cryogenic circuit layout.

Preliminary Design Study on Non-Twisted HTS Conductor for Fusion Applications

Full transposition of superconducting wires within a conductor may not be a requirement for high temperature superconductor (HTS) magnet for fusion applications. Already, conduction-cooled 20 T model coil built with non-transposed HTS stacks has been demonstrated by Commonwealth Fusion Systems (CFS). Here, we discuss 3 possible stacked HTS cryogen-cooled conductor concepts for toroidal field (TF) magnet applications. The first one is somewhat like LTS conductors, the second, HTS stacks are capped by copper stabilizer, the last one, only stacks without copper. We simulate a simplified TF magnet model, 12 T, size of KSTAR, using the 3 conceptual HTS conductor designs. A comparative thermo-hydraulic analysis for a fast charging case has been carried out and its implications on HTS conductor design are further discussed.

Design of superconducting tokamak magnet systems and applications to HTS

In this paper we illustrate a self-consistent procedure to design the superconducting Toroidal Field (TF) and Central Solenoid (CS) magnet systems starting from the basic plasma dimensioning parameters. By defining the plasma major radius and minor radii, as well as the inboard TF coil neutron shielding radial built, we proceed by sizing the TF Winding Pack (WP), its case and the CS modules. The ITER magnets design philosophy and criteria are adopted throughout the study. A key outcome of this work is to illustrate the inter-play between machine size, magnetic field and availability of inductive flux for burn. In doing so we investigate the advantages brought by high field tokamak design compared the more conventional (ITER-like) Nb3Sn based magnet design solutions.

Development of a Scalable MMC Pulsed Power Supply through HIL Methodology

Nuclear fusion experiments are becoming one of the most interesting focuses of research, given the hope of generating programmable, safe, and green energy. Among them, ASDEX (axially symmetric divertor experiment) upgrade has been operating at the Max Planck Institute for Plasma Physics (IPP) research center since 1991. To ignite and confine the plasma, several coils must be supplied through controllable high-current pulsed power supplies. The toroidal field magnets are here considered and a modular multilevel converter (MMC)-like system was designed and tested thanks to a small-scale prototype in previous works. The MMC-like topology, consisting of full-bridge submodules (SMs) with IGBTs and supercapacitor and exploitable also for other industrial applications, was chosen because of its modularity, redundancy, fault tolerance, and large amount of stored energy. The prototype, made of four SMs, was necessary to highlight critical key points in the design process. However, its scalability must be further tested before building a full-scale power supply, meant to reach almost 2400 SMs to guarantee the energy required by the load. This paper aims at validating hardware-in-the-loop (a powerful, safe, and relatively inexpensive real-time simulation environment that enables testing with real control boards) as a useful technology for power supply scalability studies and not only for control strategy tests. The results obtained previously from the prototype will allow us to finally increase the number of SMs and test the MMC-like scalability.

Control and Protection Strategy of Toroidal Field Magnet Power Supply System in HL-2M

The HL-2M device is a new generation tokamak at the Southwestern Institute of Physics that is used to develop nuclear fusion research. To confine the plasma, the toroidal field (TF) coil of HL-2M is needed to output 140-kA direct current to generate a 2.2-T magnetic density flux. Therefore, the TF power supply system was developed. It is powered by two flywheel pulsed motor-generator sets, and a cophase counter parallel connection bridge rectifier was used to provide more current. According to the TF power supply system topology, the amplitude of the magnetic flux density of the TF coil can be tuned by adjusting the excitation system of the flywheel pulse generator; the control strategy and mathematical formula are given. Then, the TF power supply protection system is established, and protection strategies under different faults are designed. The control and protection performance of the TF power supply system has been successfully applied and verified in the first plasma discharge of the HL-2M experiment.

Final design of the DTT Toroidal power supply circuit

The Divertor Tokamak Test (DTT) facility is in advanced design and construction phase at the ENEA laboratories in Frascati, Italy, to contribute to the study, the design and assessment of systems for the treatment of heat exhaust. The DTT Toroidal Field (TF) magnet system consists of 18 superconducting coils, fed in series by a single power supply (TFPS), which is a 24-pulses thyristor-based converter with an output DC current up to of 44 kA. In case of a quench in the superconductors, the magnetic energy stored in the coils is extracted by 3 Fast Discharge Units (FDUs) in series with 3 sectors of 6 TF coils. After the respective contracts were awarded, the TFPS and the FDUs are currently in the final design phase which is addressed in this paper. In particular, the analysis is focused on the most innovative aspects of the design, as the use of silicon carbide (SiC) varistors instead of discharge resistors and of a fully electronic and completely redundant high-current switch without any mechanical device.

Development of a MMC demonstrator for nuclear fusion devices power supplies

The modular multilevel converter (MMC) has become one of the most attractive converters for high-power applications such as high voltage DC (HVDC) converters, but also for fusion devices power supplies. The combination of this technology with a high power density energy storages such as supercapacitors (SC) represents a promising alternative to power the toroidal field (TF) magnets of the ASDEX (Axially Symmetric Divertor Experiment) Upgrade experiment operated at the Max-Planck-Institute for Plasma Physics (IPP) research center. After a first feasibility study, a single MMC submodule (SM) composed by four insulated-gate bipolar transistors (IGBT) forming a 4QC (4-quadrant chopper), a SC module and a power stage filter has been developed and successfully tested. Therefore, three additional identical modules were built to test their series and parallel operation in order to prove their capability to be scaled up.This work shows the step-by-step development of the MMC demonstrator, highlighting the results of the SM synchronized operation which is fundamental for the scalability of the system. The outcome of these experiments is relevant not only for the specific application of ASDEX Upgrade TF coils, but also for many other applications due to the flexible four-quadrant operation of the converter.

Neutronic and Shielding Analyses for the DTT Electronics

The radiation effect on electronic devices is a critical issue for nuclear facilities, accelerators, and spatial missions as well as for fusion machines. Phenomena related to particle interactions have significant impact on the normal functioning of such devices: they could arise progressively during machine lifetime due to the cumulated ionization [total ionizing dose (TID)], cumulated atomic displacement (dpa) or it can appear instantaneously after a single highly ionizing particle interaction [single-event effect (SEE)]. The divertor tokamak test (DTT) facility is a tokamak machine, which will produce, in its high-performance phase, up to <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$1.5\times 1017$ </tex-math></inline-formula> n/s at 2.45 MeV from deuterium–deuterium reactions plus about 1% of 14.1-MeV neutrons produced by the triton burn-up inside the plasma. Such significant neutron production implies a severe radiation environment, up to 1010 n/cm2/s and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$5\times 109\,\,\gamma $ </tex-math></inline-formula> /cm2/s [secondary gammas mainly due to (n, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula> ) reactions], inside Torus Hall building (THB) and outside the THB nearby the penetrations. DTT will be equipped with several power and signal electronic devices: 1) critical electronics that play a crucial role for the machine functioning in order to avoid failure, loss of plasma stability, and damages on the structural components (e.g., energy supply to the toroidal field, poloidal field, and central solenoid superconductive magnets); and 2) noncritical electronics, used to power diagnostics and systems whose failures do not compromise the machine operations, but are essential for scientific and engineering exploitation programs. Neutronics and shielding studies of electronics are fundamental for protecting the main components of the tokamak and to avoid the failure of the critical systems, as well as for the evaluation of the positioning and functioning of the noncritical devices and the selection of radiation-hard electronics. In this work, 3-D shielding studies for the DTT electronics devices, carried out with MCNP Monte Carlo code, are presented and discussed as well as the effectiveness of the proposed arrangement and mitigation solutions.

Impact of indentation on the performance of high Jc Nb3Sn strand

To meet the peak magnetic field requirements of the central solenoid (CS) and toroidal field (TF) magnets in the China Fusion Engineering Test Reactor (CFETR), cable-in-conduit conductors (CICC) that are composed of Nb3Sn strands with high critical current density (Jc) have been developed. The Short-Twist-Pitch (STP) design is adopted to avoid performance degradation caused by the electromagnetic cycles. But the STP design and tight cabling will cause severe indentation at the contact point between strands. Previous studies on Nb3Sn CICC show that the indentation has a significant effect on the critical current (Ic) and residual resistivity ratio (RRR) of the superconducting strand. Therefore, the same study was carried out for the high Jc Nb3Sn superconducting strand produced by Oxford Superconducting Technology (OST) and Kis wire Advanced Technology CO., Ltd. (KAT). In this paper, the Ic and RRR measurements of sample at different indentation depths were made, and the results were compared with the previous Nb3Sn indentation strands and flat rolled samples. Also, the hysteresis loss of the samples with different indentation depth was measured, and the change of effective filament diameter with indentation depth was obtained.

Progress of Engineering Design of CFETR Machine Assembly

The major radius of the China Fusion Engineering Test Reactor (CFETR) is 7.2 m, and its minor radius is 2.2 m, which are larger than those of the International Thermonuclear Experimental Reactor (ITER). That makes the assembly of the CFETR machine more intricate and challenging due to the assembly tool design, and their stresses are more complex when the weight of key parts/components increases, especially the assembly of the cryostat vessel, the vacuum vessel (VV), the toroidal field (TF) magnets, the poloidal field (PF) magnets, and the thermal shielding (TS). Based on the characteristics of the CFETR machine, a 45-deg assembly sector (with eight sectors in total), containing one 45-deg VV sector, two TF magnets, and one 45-deg VV TS, was designed as an assembly unit together with its assembly tooling. To improve the assembly efficiency, three working regions along the toroidal direction of the CFETR machine were designed to operate simultaneously. In addition, the assembly tools of the PF magnets and the cryostat were considered, and all of them are capable of supporting and adjusting the large CFETR machine components. Meanwhile, to improve their assembly accuracy and measurement efficiency, a laser tracker, an indoor global positioning system, and a scanner were employed in their assembly process. In addition, a metrology network was built for assembly of the CFETR machine.

AC Losses in JT-60SA TF Magnet During Commissioning: Experimental Analysis and Modeling

During the integrated commissioning of JT-60SA tokamak, the Toroidal Field (TF) magnet has experienced fast current discharges which have induced AC losses in its winding packs (WP) and in its casings. Using the thermo-hydraulic sensors network installed at the inlet and outlet of the TF coils, we have performed enthalpy balances to estimate the transient heat loads generated by AC losses. In parallel, we have modeled the hysteresis and coupling losses produced in the WP and the eddy currents losses produced in the casings by such events. We then assess and discuss the consistency between these experimental and first theoretical analyses from the energy balance point of view.

Homogenization of Winding Pack Properties for the Structural Analysis of Fusion Magnets

As the experimental ITER fusion reactor faces its final construction phase, design activities of the next reactor DEMO that will supply net electrical energy to the grid are being conducted in Europe. DEMO will be significantly larger than ITER, and its magnet system will be key to confine the plasma and control its shape. In particular, the Toroidal Field (TF) Coils are necessary to generate the closed toroidal field lines that confine the plasma in a tokamak fusion reactor. The interaction of the current in these coils with the magnetic field produced by the poloidal magnet system induces out-of-plane forces that make a three-dimensional structural assessment of the TF magnet structure unavoidable. Moreover, Poloidal Field (PF) Coils are supported by the TF Coil casings and are to be included in the analysis of a global model of the magnet system. Nevertheless, due to the fact that Winding Packs (WP) in Fusion magnet coils are heterogeneous and rather complex structures, accounting for a detailed model of the WP with a fine mesh in a 3-D analysis of the magnet system can become extremely costly. Homogenization techniques are therefore commonly used to model the WP by means of a uniform elastic block with orthotropic thermo-mechanical properties. Several techniques exist for this purpose and a variety of approaches are in use in the fusion magnet community yielding different values for the effective properties. This work describes homogenization techniques stemming from an energetic criterion that forces equal stored elastic energy in the heterogeneous structure and in the homogenized volume and presents a comparison between them with the goal to contribute to the founded standardization of the structural analysis procedures in the fusion magnet field, with special emphasis in the challenging DEMO magnet coils. The mentioned standardizations are crucial if the design and construction of Fusion Magnets are to become widespread activities in the future.

Evaluation of AC losses in toroidal field magnets in the China Fusion Engineering Test Reactor

Toroidal field (TF) superconducting magnets are the key components of the China Fusion Engineering Test Reactor (CFETR). While the TF magnets operate in a steady-state, they are located in a time-varying poloidal field environment, which may lead to AC losses. In this paper, the AC losses of a CFETR TF magnet during excitation and plasma disruption were analyzed. According to the experimental data, the relationship between the critical current density and the magnetic field (Jc–B) was obtained, allowing the hysteresis loss of a TF magnet to be calculated. The results show that the hysteresis loss during excitation is high, at about 1.26 MJ. The coupling loss of a TF magnet can be evaluated when the time constant of a conductor is obtained theoretically. The AC losses of TF magnet joints were analyzed by the same method, which showed that they are much less than those of a TF magnet.

Preliminary analysis and design of 4kW@4.5K helium refrigerator for CFETR toroidal field magnet test facility

China Fusion Engineering Test Reactor (CFETR) is the next magnetic confinement fusion device to demonstrate fusion energy. To test the cryogenic performance of CFETR Toroidal Field (TF) superconducting magnet, a large-scale helium cryogenic system was under construction to provide Supercritical Helium (SHe), supported by Comprehensive Research Facility for Fusion Technology (CRAFT). In this paper, heat loads of TF magnet and the dedicated distribution system were analysed to define the scale. Process flow of the helium refrigerator has been designed considering the various operation phases. Preliminary thermodynamic cycle of the helium refrigerator have been calculated with the capacity of 4kW@4.5K+20g/s@50K. This 4kW@4.5K helium refrigerator was not only a test facility for CFETR superconducting magnets, but also a prototype of CFETR helium refrigerator.

Progress in engineering design of CFETR toroidal field superconducting magnet

China Fusion Engineering Test Reactor (CFETR) is a national scientific research project which is aimed to initially build fusion energy production up to 2×108 W and to eventually reach relevant power level 1×109 W of DEMO. The major and minor radius of CFETR are 7.2 m and 2.2 m, respectively. As one of the most important components, a full-sized CFETR Toroidal Field (TF) magnet was designed. The height and width of the TF magnet are 21.3 m and 12.4 m separately. To provide 6.5 T at plasma center, TF magnet operates at a current of 95,600 A/turn. In this paper, engineering design of TF coil case (TFCC) is firstly introduced. And the structural design of internal joints, external joints, conductor, and insulation system of TF winding pack (WP) is presented. Based on the thermal load, the cooling system is also designed. Finally, some auxiliary structures are designed to ensure the normal operation of TF magnet.

Experiments on ST40 at high magnetic field

The combination of the high beta values, which have been achieved in spherical tokamaks, and high toroidal field (TF) that can be produced by high temperature superconductor (HTS) TF magnets, opens a path to lower-volume fusion reactors, in accordance with the fusion power scaling ∼β t 2 B t 4 V. High field spherical tokamak ST40 is the first prototype on this path and is now operating and results of first experiments are presented. Increase in T e, T i and the thermal energy with the TF has been observed for TF up to ∼2 T in OH operations with a sharper increase at TF above 1 T, which suggests transition to a better confinement above this threshold. First results of experiments with the neutral beam injection show expected increase in the T i, plasma density and the thermal energy. Results of the merging-compression plasma formation (m/c) studies show agreement with the previous studies, however at the high TF, formation of run-away electrons has been observed for the first time during m/c start-up. Advances in the HTS development are also presented.