Central Solenoid Research Articles

Development of a high current density, high temperature superconducting cable for pulsed magnets

Abstract A low AC loss Rare Earth Barium Copper Oxide (REBCO) cable, based on the VIPER cable technology has been developed by Commonwealth Fusion Systems for use in high field, REBCO based tokamaks. The new cable is composed of partitioned and transposed copper ‘petals’ shaped to fit together in a circular pattern with each petal containing a REBCO tape stack and insulated from each other to reduce AC losses. A stainless steel jacket adds mechanical robustness—also serving as a vessel for solder impregnation—while a tube runs through the middle for cooling purposes. Additionally, fiber optic sensors are placed under the tape stacks for quench detection. To qualify this design, a series of experiments were conducted as part of the SPARC tokamak Central Solenoid Model Coil program—to retire the risks associated with full scale, fast ramping, high flux HTS Central Solenoid (CS) and Poloidal Field (PF) coils for tokamak fusion power plants and net energy demonstrators. These risk study and risk reduction experiments include (1) AC loss measurement and model validation in the range of ~5 T/s, (2) an IxB electromagnetic loading of over 850 kN/m at the cable level and up to 300 kN/m at the stack level, (3) a transverse compression resilience of over 350 MPa, (4) manufacturability at tokamak relevant speeds and scales, (5) cable to cable joint performance, (6) fiber optic based quench detection speed, accuracy, and feasibility, and (7) overall winding pack integration and magnet assembly. The result is a cable technology, now referred to as PIT VIPER, with AC losses that measure fifteen times lower (at ~5 T/s) than its predecessor technology; a 2% or lower degradation of critical current (Ic) at high IxB electromagnetic loads; no detectable Ic degradation up to 570 MPa of transverse compression on the cable unit cell; end to end magnet manufacturing, consistently producing Ic values within 7% of the model prediction; cable to cable joint resistances at 20 K on the order of ~15 nΩ; and fast, functional quench detection capabilities that do not involve voltage taps. This cable technology will be tested comprehensively in a Central Solenoid Model Coil to prove its readiness for compact, high field tokamak operation.

MANTA: a negative-triangularity NASEM-compliant fusion pilot plant

Abstract The MANTA (Modular Adjustable Negative Triangularity ARC-class) design study investigated how negative-triangularity (NT) may be leveraged in a compact, fusion pilot plant (FPP) to take a ‘power-handling first’ approach. The result is a pulsed, radiative, ELM-free tokamak that satisfies and exceeds the FPP requirements described in the 2021 National Academies of Sciences, Engineering, and Medicine (NASEM) report ‘Bringing Fusion to the U.S. Grid’ (2021 Bringing Fusion to the U.S. Grid). A self-consistent integrated modeling workflow predicts a fusion power of 450 MW and a plasma gain of 11.5 with only 23.5 MW of power to the scrape-off layer (SOL). This low P SOL together with impurity seeding and high density at the separatrix results in a peak heat flux of just 2.8 MW m−2. MANTA’s high aspect ratio provides space for a large central solenoid (CS), resulting in ∼15 minute inductive pulses. In spite of the high B fields on the CS and the other REBCO-based magnets, the electromagnetic stresses remain below structural and critical current density limits. Iterative optimization of neutron shielding and tritium breeding blanket yield tritium self-sufficiency with a breeding ratio of 1.15, a blanket power multiplication factor of 1.11, toroidal field coil lifetimes of 3100 ± 400 MW · yr, and poloidal field coil lifetimes of at least 890 ± 40 MW · yr. Following balance of plant modeling, MANTA is projected to generate 90 MW of net electricity at an electricity gain factor of ∼ 2.4 . Systems-level economic analysis estimates an overnight cost of US$3.4 billion, meeting the NASEM FPP requirement that this first-of-a-kind be less than US$5 billion. The toroidal field coil cost and replacement time are the most critical upfront and lifetime cost drivers, respectively.

Design and development of FPGA based trigger system for automation of metallic tokamak (MT-I)

A field-programmable gate array (FPGA)-based trigger system (FTS) is designed and developed to replace the existing microcontroller-based system for the automation of metallic tokamak (MT-I). MT-I consists of three main systems: MT-I vacuum vessel and gas filling system, a pulsed power supply system, an electronic system based on diagnostics and data acquisition (DAQ) systems. The power supply system provides pulsed power input to the central solenoid (CS), poloidal field coils (PF), toroidal field coils (TF), and microwave systems using thyristors and insulated-gate bipolar transistors (IGBTs). The (DAQ) system acquires experimental data from MT-I diagnostics using national instruments (NI) DAQ cards. A concurrent processing system is required to incorporate time delay in the triggering process of central solenoid (CS), poloidal field coils (PF), toroidal field coils (TF), microwave source and DAQ systems. Therefore, an FTS is designed and developed to complement the processing capability, unlike a microcontroller. The FTS has twenty concurrent triggering channels, adjustable from time reference zero, and control from a simple graphical user interface (GUI) designed on LabVIEW. For current buffering and optical isolation against high voltages in the MT-I power supply system, a peripheral electronic system (PES) and field trigger modules (FTM) have been developed as part of FTS. The designed and developed FTS was tested successfully on MT-I.

Plasma control for the step prototype power plant

In 2019 the UK launched the Spherical Tokamak for Energy Production (STEP) programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: (i) All plasma current during the burn phase should to be generated through non-inductive means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure βN with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. (ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. (iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. (iv) The heat pulses associated with unmitigated ELMs are unlikely to be acceptable imposing ELM free operation or active ELM control. (v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. (vi) High pressure operation is likely to require active resistive wall mode (RWM) stabilisation. (vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the RWM without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.

Selected Topics of Technical Challenges of the ITER Central Solenoid

Selected Topics of Technical Challenges of the ITER Central Solenoid

Quench Simulation of the CFETR Central Solenoid Model Coil

Quench Simulation of the CFETR Central Solenoid Model Coil

Structural Qualification of ITER Central Solenoid Module During Transport From California, USA to ITER Site

Structural Qualification of ITER Central Solenoid Module During Transport From California, USA to ITER Site

Studies of the outer-off-midplane lower hybrid wave launch scenario for plasma start-up on the TST-2 spherical tokamak

Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8 kA was achieved. Coupled ray-tracing and Fokker–Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.

Performance of MT-I spherical tokamak with upgraded power supplies

This paper describes the upgradation of power supply systems for the improved performance of a small spherical tokamak (MT-I) with an aspect ratio of 1.67 in operation at the Pakistan Tokamak Plasma Research Institute (PTPRI), Islamabad. Three mutually independent power supplies are designed to generate current pulses of desired shape and magnitude in central solenoid (CS), toroidal field (TF) and vertical field (VF) coil systems. The main parts of a power supply are capacitor banks as energy storage devices, ignitrons and thyristors as fast switches for controlled power switching. A double capacitor bank topology with enhanced second capacitor bank voltage is found to be suitable for CS to improve the plasma startup conditions. Simulations are performed in MATLAB Simulink for confirmation of the designed power supply system. The Rogowski coil and current transformers are employed to measure plasma current and the coil’s current waveforms during the experiment in the presence of hydrogen as working gas. An ocean spectrometer and a photron high-speed camera (SA-8) are used for optical measurements and fast plasma imaging. An improvement in startup conditions is observed, enhancing the plasma current duration and light emission intensity recorded in the high-speed camera masked with a Hα filter.

DTT toroidal field conductor samples test in Sultan: DC and AC characterization

The superconducting magnet system of the Divertor Tokamak Test (DTT) facility, composed of 18 toroidal field (TF) coils, 6 poloidal field coils and a central solenoid, has been designed and many procurements have been launched. Some manufacturing aspects and some conductor features require characterization under relevant close-to-operative conditions. To confirm the design choices in all details, cryogenic tests in qualified facilities have been foreseen. In this work, the results of the TF samples characterization at the SULTAN facility at the Swiss Plasma Centre (SPC, EPFL) are presented. The 3 week test campaign started on July the 8th, 2022. The DTT TF SULTAN sample was made of two Nb3Sn cable-in-conduit conductor ‘legs’, namely ‘TF-A’ and ‘TF-B’, made with wires produced by Kiswire Advanced Technology, differing for the cabling twist pitch sequence only, and designed to work in DTT at 42.5 kA at 11.9 T peak field. The extensive characterization comprised 3000 electro-magnetic (EM) cycles and two warm-up-cool-down (WUCD) steps, and in detail it included: AC measurements on the virgin conductors, on cyclic loaded conductors and after WUCDs; DC tests at 10.85 T/42.5 kA with intermediate EM cycles at 10.85 T/45 kA before and after WUCDs; DC tests using partial Lorentz force loads, and Minimum Quench Energy tests at 9 T/42.5 kA after cycles and WUCDs. The results of the DC measurement analysis verified the design, in terms of current sharing temperature (T cs) and critical current (I c), as both samples are over the minimum acceptance values. In particular, the ‘TF-A’ sample, characterized by a so-called ‘long twist pitch’ cabling sequence, showed higher performance without any degradation with loading and WUCD cycles, whereas sample ‘TF-B’ presented an initial T cs reduction that afterwards substantially remained unchanged. In terms of strain acting at the Nb3Sn filaments level, this result can be described by a lower effective strain in the ‘TF-A’ sample. AC losses were measured with a calorimetric method as a function of frequency for each series of AC sinusoidal pulsing measurements, and the characteristic coupling time constants were determined.

ENN's roadmap for proton-boron fusion based on spherical torus

ENN Science and Technology Development Co., Ltd. (ENN) is committed to generating fusion energy in an environmentally friendly and cost-effective manner, which requires abundant aneutronic fuel. Proton-boron (p-11B or p-B) fusion is considered an ideal choice for this purpose. Recent studies have suggested that p-B fusion, although challenging, is feasible based on new cross section data, provided that a hot ion mode and high wall reflection can be achieved to reduce electron radiation loss. The high beta and good confinement of the spherical torus (ST) make it an ideal candidate for p-B fusion. By utilizing the new spherical torus energy confinement scaling law, a reactor with a major radius R0=4 m, central magnetic field B0=6 T, central temperature Ti0=150 keV, plasma current Ip=30 MA, and hot ion mode Ti/Te=4 can yield p-B fusion with Q>10. A roadmap for p-B fusion has been developed, with the next-generation device named EHL-2. EHL stands for ENN He-Long, which literally means “peaceful Chinese Loong.” The main target parameters include R0≃1.05 m, A≃1.85, B0≃3 T, Ti0≃30 keV, Ip≃3 MA, and Ti/Te≥2. The existing ST device EXL-50 was simultaneously upgraded to provide experimental support for the new roadmap, involving the installation and upgrading of the central solenoid, vacuum chamber, and magnetic systems. The construction of the upgraded ST fusion device, EXL-50U, was completed at the end of 2023, and it achieved its first plasma in January 2024. The construction of EHL-2 is estimated to be completed by 2026.

Co-simulation of cool down process from 300 to 80 K for ITER Tokamak

Dynamic simulation of Tokamak superconducting magnet system has been conducted to investigate the cool-down process from 300 to 80 K. The simulation focuses on the cool-down speed variations with respect to the global temperature gradients in the different coils; Toroidal Field (TF) coil, TF STructure (TF-ST), Central Solenoid (CS) and Poloidal Field (PF)/Correction Coil (CC) systems. As imposing the maximum temperature gradients dT max ≤ 50K, the speed should be adjusted, ensuring the limited mechanical stresses due to thermal contractions. So far, the process simulation of Tokamak cryogenic system has been concentrated on the Deuterium Tritium (DT) operation phase; therefore, it is necessary to revise the model to extend its capability for the cool-down; for example, the thermo-hydraulic properties of Cable-in-Conduit Conductor (CICC) for each coil, mechanical properties of materials for the magnet system. The cool-down process is implemented at the helium refrigerator by utilizing LN2 heat exchanger up to 80 K. Its speed is set at -0.8 K/hr as a baseline, which can be controlled by the global temperature gradients in the magnet system. The process will be on hold as dT max ≥ 50K, and resumed once dT max < 50K. The paper discusses the cool-down processes of the Tokamak and identifies the impact on the speed, dT i/dt, of each component.

AC Loss Analysis of the Central Solenoid Charge and Discharge

AC Loss Analysis of the Central Solenoid Charge and Discharge

Plasma surrogate modelling using Fourier neural operators

Predicting plasma evolution within a Tokamak reactor is crucial to realizing the goal of sustainable fusion. Capabilities in forecasting the spatio-temporal evolution of plasma rapidly and accurately allow us to quickly iterate over design and control strategies on current Tokamak devices and future reactors. Modelling plasma evolution using numerical solvers is often expensive, consuming many hours on supercomputers, and hence, we need alternative inexpensive surrogate models. We demonstrate accurate predictions of plasma evolution both in simulation and experimental domains using deep learning-based surrogate modelling tools, viz., Fourier neural operators (FNO). We show that FNO has a speedup of six orders of magnitude over traditional solvers in predicting the plasma dynamics simulated from magnetohydrodynamic models, while maintaining a high accuracy (Mean Squared Error in the normalised domain ≈10−5 ). Our modified version of the FNO is capable of solving multi-variable Partial Differential Equations, and can capture the dependence among the different variables in a single model. FNOs can also predict plasma evolution on real-world experimental data observed by the cameras positioned within the MAST Tokamak, i.e. cameras looking across the central solenoid and the divertor in the Tokamak. We show that FNOs are able to accurately forecast the evolution of plasma and have the potential to be deployed for real-time monitoring. We also illustrate their capability in forecasting the plasma shape, the locations of interactions of the plasma with the central solenoid and the divertor for the full (available) duration of the plasma shot within MAST. The FNO offers a viable alternative for surrogate modelling as it is quick to train and infer, and requires fewer data points, while being able to do zero-shot super-resolution and getting high-fidelity solutions.

Simulation of ion cyclotron wave heating in the EXL-50U spherical tokamak based on dispersion relations

This study investigates the single-pass absorption (SPA) of ion cyclotron range of frequency (ICRF) heating in hydrogen plasma of the EXL-50U spherical tokamak, which is an upgraded EXL-50 device with a central solenoid and a stronger magnetic field. The reliability of the kinetic dispersion equation is confirmed by the one-dimensional full-wave code, and the applicability of Porkolab's simplified theoretical SPA model is discussed based on the kinetic dispersion equation. Simulations are conducted to investigate the heating effects of the fundamental and second harmonic frequencies. The results indicate that with the design parameters of the EXL-50U device, the SPA for second harmonic heating is 63%, while the SPA for fundamental heating is 13%. Additionally, the optimal injection frequencies are 23 MHz at 0.9 T and 31 MHz at 1.2 T. The wave vector of the antenna parallel to the magnetic field, with a value of , falls within the optimal heating region. Simulations reveal that the ICRF heating system can play an important role in the ion heating of the EXL-50U.

Exploration of ITER operational space with as-built stiffness of central solenoid modules

The as-built stiffness in the ITER central solenoid (CS) modules (CSM1 thorough to CSM4 are currently manufactured) determines the range of vertical compression forces that can be tolerated by the CS modules during ITER operation. Since the as-built stiffness of the CS modules manufactured (∼32 GPa and ∼34 GPa for CSM1 and CSM2, respectively and similar for the other modules) has been reduced from the design value (53 GPa), the CS axial (vertical) force criteria have been updated assuming a conservative stiffness (25 GPa) with margins for all six CS modules. Initial analysis using the updated CS force criteria has revealed that this reduction affects only the plasma initiation with fully charged CS in the ITER 15 MA Baseline DT scenario, resulting in a slight reduction of poloidal magnetic flux, from 117.5 Wb to 116.2 Wb at initial CS magnetization. Therefore, the 15 MA Baseline scenario has been re-developed with an updated plasma start-up, and then the entire evolution of the CS and poloidal field coil parameters has been validated against all the coil currents, fields and forces criteria. To explore potential risks and opportunities for further optimization of scenarios, the equilibrium operational space (the plasma internal inductance versus the poloidal magnetic flux produced by the coils) at flat-top burn has been analyzed using the CORSICA and DINA codes. The three major ITER reference DT operation scenarios, 15 MA Q = 10 Baseline, 12.5 MA Q > 5 Hybrid and 10 MA Q ∼ 5 Steady-State, satisfy all the coil criteria including the CS force updated reflecting the as-built stiffness. The evolution of the plasma discharge parameters within the equilibrium operational spaces provided a guidance for potential optimization with margins.

Demonstration of transient CHI startup using a floating biased electrode configuration

Results from the successful solenoid-free plasma startup using the method of transient coaxial helicity injection (transient CHI) in the QUEST spherical tokamak (ST) are reported. Unlike previous applications of CHI on HIT-II and on NSTX which required two toroidal insulating breaks to the vacuum vessel, QUEST uses a first of its kind, floating single biased electrode configuration, which does not use such a vacuum break. Instead, the CHI electrode is simply insulated from the outer lower divertor plate support structure. This configuration is much more suitable for implementation in a fusion reactor than the previous configurations. Transient CHI generated toroidal currents of 135 kA were obtained. The toroidal current during the formation of a closed flux configuration was over 50 kA. These results bode well for the application of transient CHI in a new generation of compact high-field STs and tokamaks in which the space for the central solenoid is very restricted.

Real-Time Plasma Magnetic Control System with Equilibrium Reconstruction Algorithm in the Feedback for the Globus-M2 Tokamak

To control the plasma shape during a tokamak discharge, it is necessary to calculate the plasmashape in real-time. The rate requirements for the shape calculations are especially high for tokamaks with asmall radius, such as Globus-M2 (St. Petersburg, Russia). A real-time magnetic plasma control system forthe Globus-M2 tokamak with flux and current distribution identification (FCDI) algorithm for the plasmaequilibrium reconstruction in feedback is presented. The control system contains discrete one-dimensionaland matrix proportional-integral-derivative controllers synthesized by the matrix inequality method usingthe plasma LPV model calculated on experimental data, and carries out the coordinated control of the plasmaposition and shape as well as the compensation for the scattered field of the central solenoid. The FCDI algorithmis improved for the operation in the real-time mode, and makes it possible to reconstruct the plasmashape in 20 μs. The digital control system with a feedback algorithm was simulated on a real-time test bench,consisting of two Speedgoat Performance Real-Time Target Machines (RTTM), and demonstrated the averageTask Execution Time (TET) value in 67 μs.

A simple study for an optimized operation of the ITER Cryodistribution cold rotating machines

The ITER superconducting (SC) magnet system is composed of 4 distinct units: the Central Solenoid (CS) coils, the Toroidal Field (TF) coils, the coil encasing and supporting Structures (ST), and the Poloidal Field and Correction Coils (PF&CC or simply PF). Including another cold component unit, which is the Cryopump (CP) system, each unit is assigned its own Auxiliary Cold Box (ACB) with cold rotating machines (CRMs) to ensure efficient operation from a cryogenic perspective. The CRMs, which are the cold circulator (CCr) and cold compressor (CCp), are respectively in charge of generating the high mass flow rate (2.0–2.7 kg/s) and low temperature (3.7–4.2 K) of supercritical helium (SHe) that is supplied to each magnet unit and CP system. To maintain the proper cryogenic conditions of the magnets and CPs during fusion experiments, the ITER Cryogenic System can produce an average cooling power of 75 kW at 4.5 K, including the need for the 50 K gaseous helium (GHe) cooled high-Tc SC current leads. Depending on the operation mode of the ITER Tokamak, up to 28 kW of that power can be consumed by the CRMs due to compression of helium.In this study, we will explore possibilities for optimized operation of the CRMs by adjusting the mass flow rate and temperature of the SHe to minimize the heat of compression while maintaining the thermal stability of the ITER SC magnets and CP. By doing so, the CRM-heat-of-compression at 4.5 K can be reduced by more than 3 kW, providing additional “cryogenic” margins for higher-heat-load plasma experiments.

Development of prototype power supply for ohmic transformer system of SSST

The conceptual design of Small Scale Spherical Tokamak (SSST) has been completed at Institute for Plasma Research (IPR). It consists of poloidal field (or Equilibrium) field coils, Toroidal field coils and Ohmic Transformer (OT) system. The poloidal magnets are used for the plasma shaping and equilibrium. The plasma current ramp up and sustenance up to 20 ms is achieved by OT system. The OT system comprises of Central Solenoid (CS) and compensating coils TR1, TR2 & TR3. The Preliminary design for power supply for OT system has been finalized and a prototype OT power supply has been developed at IPR. The conceptual design of prototype Ohmic Transformer Power Supply (OTPS), its philosophy of operation, simulation results and experimental results are presented in this paper.