Poloidal field (PF) coil

Overview

Poloidal field (PF) coils are a set of external, typically circular, electromagnets arranged around the toroidal vacuum vessel of a tokamak or related magnetic confinement device. Their primary function is to generate a magnetic field in the poloidal direction—the short way around the torus. This field is essential for initiating, shaping, controlling, and sustaining the plasma. The PF coil system works in concert with the toroidal field (TF) coils, which provide the main confining field in the toroidal direction.

The PF system performs three critical tasks:

Plasma Current Induction: The largest and most central PF coil, the central solenoid (CS), acts as the primary winding of a transformer. By ramping down the current in the CS, a changing magnetic flux is created, which induces a powerful toroidal electric field. This field ionizes the fusion fuel and drives a large current (millions of amperes) through the resulting plasma ring, which is essential for both heating and confinement.
Plasma Shaping: Additional PF coils, often called shaping coils or equilibrium field (EF) coils, are placed above, below, and radially outward from the plasma. By precisely controlling the currents in these coils, a poloidal magnetic field is generated that interacts with the plasma current to produce Lorentz forces. These forces shape the plasma's cross-section, typically into a 'D' shape with specific elongation and triangularity, which is necessary for achieving high plasma pressure and magnetohydrodynamic (MHD) stability.
Plasma Position and Stability Control: The PF coils are used in a dynamic feedback loop to control the plasma's vertical and horizontal position within the vacuum vessel. Magnetic sensors monitor the plasma's location, and a control system adjusts the PF coil currents in real-time (on a millisecond timescale) to counteract drifts and instabilities, particularly the vertical displacement event (VDE) to which elongated plasmas are susceptible.

Without a robust and precisely controlled PF coil system, it would be impossible to achieve the stable, high-performance plasma conditions required to meet the Lawson criterion for net fusion energy gain.

Physics / Mechanism

The operation of the poloidal field coil system is governed by fundamental principles of electromagnetism and plasma physics. The system's functions are distinct but executed by the coordinated action of all its coils.

Inductive Current Drive via the Central Solenoid The primary method for driving the toroidal plasma current (I_p) is inductive, based on Faraday's law of induction. The central solenoid is a large, multi-turn solenoid magnet located in the central bore of the tokamak. Before a plasma discharge, a high current is driven through the CS, storing a large amount of magnetic energy and generating a strong vertical magnetic field, B_z. To initiate and sustain the plasma, this current is rapidly ramped down. The resulting change in magnetic flux (dΦ/dt) through the center of the torus induces a toroidal loop voltage (V_loop = -dΦ/dt). This voltage drives I_p in the plasma, which acts as the single-turn secondary winding of a transformer.

The total flux swing (ΔΦ) available from the CS is a critical design parameter, as it determines the duration of the inductively sustained plasma pulse. For a given plasma current and resistance, a larger flux swing allows for a longer pulse. This is a key reason why steady-state tokamaks require non-inductive current drive methods.

Plasma Equilibrium and Shaping A plasma in a tokamak is in equilibrium when the outward forces—the kinetic pressure of the plasma and the self-repulsion of the plasma current—are balanced by an inward Lorentz force. This inward force is generated by the interaction of the toroidal plasma current (I_p) with a vertical magnetic field (B_z) supplied by the PF coils. The force is described by F_r = I_p × B_z.

The shape of the plasma's magnetic flux surfaces is determined by the solution to the Grad-Shafranov equation, which relates the plasma pressure, current distribution, and the external magnetic field from the PF coils. By adjusting the relative currents in the different PF coils located around the torus, the external field can be tailored to produce a specific plasma shape. Key shaping parameters include:

Elongation (κ): The ratio of the plasma's vertical height to its horizontal width. Elongated plasmas (κ > 1.5) can carry higher current for a given toroidal field, leading to better confinement.
Triangularity (δ): The 'D' shape of the plasma. Positive triangularity improves stability against certain MHD modes and is essential for creating a divertor configuration, where magnetic field lines are guided to strike target plates, managing heat and particle exhaust.

Feedback Control and Stability While elongation is beneficial for performance, it introduces a fundamental vertical instability. A vertically elongated plasma is in an unstable equilibrium, analogous to balancing a pencil on its tip. Any small vertical displacement will be amplified by the interaction with the external shaping field. To counteract this, a fast feedback control system is required. Magnetic pickup coils measure the plasma's vertical position with sub-millisecond resolution. A plasma control system (PCS) processes this data and rapidly adjusts the currents in a subset of the PF coils to generate a corrective radial field (B_r), pushing the plasma back to its desired position. The response time and power handling of the PF coil power supplies are critical for the effectiveness of this control.

Historical Development

The evolution of PF coil systems mirrors the progression of tokamak design from simple, circular devices to complex, high-performance machines.

Early Tokamaks (1960s-1970s): Devices like the Soviet T-3 and the Princeton Large Torus (PLT) used relatively simple PF systems. They consisted of a central air-core solenoid (or an iron core transformer) for inductive current drive and a pair of large-radius equilibrium field coils to provide the basic vertical field for radial position control. The plasmas were circular and did not require complex shaping.
Divertors and Shaping (1980s): The introduction of the divertor concept in machines like ASDEX (Axially Symmetric Divertor Experiment) and DIII-D marked a significant change. To create the magnetic X-point necessary for a divertor, additional PF coils were required inside the TF coil bore, closer to the plasma. These coils allowed for the creation of non-circular, D-shaped plasmas, which were found to have significantly better energy confinement, leading to the discovery of the H-mode (high-confinement mode) on ASDEX in 1982.
Superconducting PF Coils (1990s): The push towards long-pulse and steady-state operation necessitated the move from resistive copper coils to superconducting ones to eliminate ohmic losses. Tore Supra in France was a pioneer, featuring a set of superconducting PF coils. The development of Cable-in-Conduit Conductors (CICC) using Niobium-tin (Nb3Sn) and Niobium-titanium (NbTi) became the standard for large, high-field superconducting magnets. These conductors allowed for the high currents and rapid current changes required for plasma control.
Advanced Shaping and Control (2000s-Present): Modern tokamaks like JET, KSTAR, and EAST feature highly sophisticated PF coil systems with numerous independent coils. This allows for advanced plasma shapes (e.g., high triangularity, snowflake divertors) and precise real-time control. The development of powerful, fast-switching power supplies and advanced digital control algorithms has been crucial to this progress.

Current Status

As of 2026, the state of the art in PF coil technology is defined by the systems being built for large-scale, next-generation fusion devices, primarily ITER. The technology is mature, but pushing the boundaries of scale, field strength, and stored energy.

The dominant technology for large PF coils is Low-Temperature Superconductors (LTS), specifically NbTi and Nb3Sn. These materials are used in CICC designs, where superconducting strands are cabled together inside a steel or Incoloy jacket, with forced-flow supercritical helium for cooling to ~4.5 K.

ITER's PF system is the most powerful ever constructed. It consists of six large-diameter coils, with the largest (PF6) measuring 24 meters in diameter. The central solenoid alone will store 41 GJ of magnetic energy and generate a peak field of 13 T. The manufacturing of these massive coils is a major global undertaking, involving complex winding, heat treatment, and jacketing processes. The successful testing of the first ITER CS module by General Atomics in 2021, which reached its full operating current of 40 kA and a field of 13 T, was a major validation of the design and manufacturing approach.

For smaller, more compact devices, particularly those exploring High-Temperature Superconductors (HTS), PF coil design is also advancing. HTS materials like REBCO (Rare-Earth Barium Copper Oxide) offer the potential for higher magnetic fields at warmer operating temperatures (~20-30 K), which could simplify the cryogenic system. However, challenges remain in manufacturing long, high-quality HTS tapes and managing them in complex coil geometries.

Notable Implementations

ITER: The International Thermonuclear Experimental Reactor has the most advanced and powerful PF coil system ever designed. Its six PF coils and six-module Central Solenoid are all superconducting, using a mix of Nb3Sn (for the high-field CS) and NbTi (for the outer PF coils). The system is designed to provide 55 V·s of flux swing to sustain a 15 MA plasma current for over 400 seconds.
JT-60SA: A joint Japanese-European project, this large superconducting tokamak in Naka, Japan, serves as a satellite experiment for ITER. Its PF system includes a four-module CS and six EF coils, all using NbTi CICC technology. The system is designed for advanced plasma shape control and long-pulse operation.
KSTAR & EAST: The Korean Superconducting Tokamak Advanced Research (KSTAR) and the Experimental Advanced Superconducting Tokamak (EAST) in China are both fully superconducting machines. Their PF coil systems, made entirely of Nb3Sn and NbTi, have been instrumental in achieving and studying long-pulse H-mode discharges lasting over 100 seconds.
SPARC / Commonwealth Fusion Systems: The SPARC experiment, a precursor to the ARC power plant concept from Commonwealth Fusion Systems, plans to use HTS REBCO magnets for all its coil systems, including the PF coils. The goal is to leverage the high field capabilities of HTS to build a more compact, high-performance device. This represents a significant departure from the LTS technology used in most large-scale projects.

Open Challenges

Despite significant progress, several engineering and scientific challenges remain for PF coil systems, particularly for future fusion power plants.

Mechanical Stresses: The immense magnetic fields and currents in PF coils generate enormous Lorentz forces. In the central solenoid, these forces are both hoop stresses (pushing the coil outwards) and large vertical separating forces between modules. Managing these stresses, especially under cyclic loading over the lifetime of a power plant, requires robust structural materials and sophisticated mechanical design. Stress accumulation and fatigue are major concerns for component lifetime.
Quench Detection and Protection: A quench is a sudden loss of superconductivity in a magnet, which can lead to rapid heating and potentially catastrophic damage. For the massive PF coils in a device like ITER, which store gigajoules of energy, detecting a quench early and safely dissipating the stored energy is a critical safety and engineering challenge. The protection systems must be extremely reliable.
Insulation Degradation: The electrical insulation within the coils must withstand high voltages (up to 20 kV during a disruption), cryogenic temperatures, and a lifetime dose of high-energy neutron radiation from the D-T fusion reaction. Finding and qualifying insulation materials (like glass-polyimide composites) that can maintain their mechanical and dielectric properties under these harsh conditions is an ongoing area of materials science research.
Integration and Assembly: The sheer scale of PF coils for a power plant presents significant logistical challenges. The largest coils may be too big to transport and will need to be manufactured on-site. The precise alignment and integration of these massive components with the rest of the tokamak structure is a complex assembly task with micrometer-level tolerances.
HTS Conductor Manufacturing: For HTS-based PF coils, the primary challenge is the industrial-scale production of long-length, high-performance, and cost-effective REBCO tape. Ensuring uniformity and minimizing defects over kilometers of conductor is essential for building reliable, large-scale magnets.

Outlook

The 5-15 year trajectory for poloidal field coil technology is focused on two parallel paths: the successful commissioning and operation of large-scale LTS systems, and the maturation of HTS technology for next-generation devices.

The immediate future will be dominated by the assembly and commissioning of the ITER PF coil system, which is expected to be completed in the late 2020s. The initial plasma operations at ITER in the early 2030s will be the ultimate test of this massive superconducting magnet system, providing invaluable data on its performance, control capabilities, and reliability. Similarly, ongoing experiments at JT-60SA will refine operational scenarios for long-pulse, shaped plasmas using its advanced LTS PF system.

In parallel, the development of HTS PF coils is expected to accelerate. Following the demonstration of HTS TF coils in projects like SPARC, the focus will shift to applying this technology to the more complex geometries and dynamic current requirements of PF coils, particularly the central solenoid. A successful HTS central solenoid could enable more compact and potentially steady-state tokamaks by offering higher fields and larger flux swings. Companies like Commonwealth Fusion Systems and Tokamak Energy are actively pursuing this path.

Looking towards a demonstration power plant (DEMO), R&D will focus on improving the lifetime and reliability of coil components, especially insulation, under high neutron fluence. The development of advanced manufacturing and on-site fabrication techniques will be critical to making the construction of these massive components economically viable. The successful operation of ITER's PF system will provide the foundational experience needed to design and build the even more demanding magnet systems for the first generation of fusion power plants.

References

ITER magnets: Field, force and fusion — ITER Organization (2023)
Design of the ITER Poloidal Field Coils — IEEE Transactions on Applied Superconductivity (2008)
Successful Test of the First ITER Central Solenoid Module — General Atomics (2021)
The physics of the tokamak — Journal of Plasma Physics (2017)
Tokamaks, 4th Edition — Oxford University Press (2015)
Status of the JT-60SA project — Nuclear Fusion (2022)
Overview of the SPARC tokamak — Journal of Plasma Physics (2020)
Plasma control in modern tokamaks — Fusion Engineering and Design (2015)