Poloidal Field Coils Research Articles

Optimization of DC Energy Storage in Tokamak Poloidal Coils

Tokamaks are a very promising option to exploit nuclear fusion as a programmable and safe energy source. A very critical issue for the practical use of tokamaks consists of the power flow required to initiate and sustain the fusion process, in particular in the poloidal field coils. This flow can be managed by introducing a DC energy storage based on supercapacitors. Because such storage may be the most expensive and largest part of the poloidal power supply system, an excessive size would cancel its potential advantages. This paper presents innovative strategies to optimize the DC storage in poloidal power supply systems. The proposed solution involves the sharing of the DC storage between different coil circuits. The study is supported by novel analytical formulas and by a circuital model developed for this application. The obtained results show that this method and the related algorithms can noticeably reduce the overall size of the storage and the power exchange with the grid, providing a practical contribution toward the feasibility and the effectiveness of nuclear fusion systems.

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.

Validation of prediction capability of operating space for plasma initiation in MAST-U

Abstract DYON is a plasma initiation modelling code that solves the differential equation system of the full circuit equations (plasma current, active coil currents and eddy currents in full passive structures) and 0D global energy and particle balance equations[1]. In order to test the capability of the full electromagnetic plasma initiation model to predict individual discharges in experiments and thus the operating space in the device, a dedicated experimental database was built in MAST-U by scanning the prefilled gas pressure p0 and the induced loop voltage Vloop . In the experimental operating space of p0 and Vloop the lower and the upper limits of p0 are determined by the plasma breakdown failure and the plasma burn-through failure, respectively. The lower limit of Vloop is determined by the plasma burn-through failure. By directly reading the control room data used in each discharge (i.e. currents in the solenoid, poloidal field coils, and toroidal field coils, p0 , and gas puffing rate), the full electromagnetic DYON consistently predicted the failed breakdown, failed burn-through, and successful plasma initiation discharges in the experimental database, demonstrating its capability to predict the operating space for inductive plasma initiation. The Paschen curve calculated with the effective connection length in MAST-U indicates a much higher p0 required for plasma breakdown than the experimental data, indicating that individual field line evaluation is necessary to calculate the quantitative requirements for Townsend breakdown. The demonstration in this paper shows that the full electromagnetic DYON could be a useful simulation tool to assess the feasibility of inductive plasma initiation and to optimise operating scenarios in future devices. [1] Hyun-Tae Kim 2022 Nuclear Fusion 62 126012

Energization Result of JT-60SA Poloidal Field Coil

Energization Result of JT-60SA Poloidal Field Coil

ITER materials irradiation within the D–T neutron environment at JET: post-irradiation radioactivity analysis following the DTE2 experimental campaign

Abstract This work presents the results following the first irradiation of ITER materials samples in a tokamak D--T plasma environment operating at significant fusion power. The materials exposed to this nuclear environment at the Joint European Torus during the DTE2 experimental campaign that took place in 2021 include representative ITER samples from various components such as poloidal field (PF) coil jacket samples, toroidal field coil radial closure plate steels, EUROFER 97 steel, W and CuCrZr materials from the divertor, Inconel-718, CuCrZr, and 316L stainless steel for blanket modules, as well as vacuum vessel forging samples.

The experimental results discussed include high-resolution gamma spectrometry measurements and analysis conducted with the post-irradiated samples, of which there were 68 in total. These samples were exposed through different experimental campaigns, including deuterium, deuterium--tritium and tritium phases. Supporting the analysis were 25 dosimetry foil-based neutron diagnostics and two 'VERDI' neutron spectrometry diagnostics. A further 12 samples for positron annihilation spectroscopy (PALS) were also irradiated. The irradiation of all these samples took place in a long-term irradiation assembly located near the JET vacuum vessel.

The post-irradiation analysis of the ITER material samples has yielded valuable insights into their material activation levels and radiation fields. Comparative assessments between experimental measurements and comprehensive neutronics simulations have demonstrated a significant level of agreement in this work, while also revealing some discrepancies in specific material instances. The data and interpretation from this work not only serve as a robust experimental foundation for enhancing the precision and predictability of neutronics simulation approaches for ITER and next-step devices but also present some opportunities for the refinement of simulation methodologies. In light of these findings, a series of recommendations have been proposed, aimed at improving confidence in nuclear predictions associated with materials that have been exposed to fusion nuclear environments and advancing understanding in this important domain.

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.

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.

Delivered Poloidal Field Coils to ITER Fusion Tokamak Facility: Status, Factory Acceptance Test Results and Overview of the Undergone Electrical Tests in Manufacturing Processes

Delivered Poloidal Field Coils to ITER Fusion Tokamak Facility: Status, Factory Acceptance Test Results and Overview of the Undergone Electrical Tests in Manufacturing Processes

End of the Winding of the European Poloidal Field Coils for the ITER Project

End of the Winding of the European Poloidal Field Coils for the ITER Project

Optimisation of the poloidal field system for advanced divertor configurations in STEP

The power exhaust proves to be one of the most challenging and concept–defining aspects in the design of a commercial fusion power plant, while the magnetic coil system, capable of supporting advanced exhaust solutions, emerges as one of the main design and cost drivers. Consequently, much effort should be dedicated to the optimisation of a robust global magnetic configuration, which integrates both the plasma and edge scenarios, while ensuring engineering feasibility and compatibility with the available technology. Here we present a multidisciplinary framework employed to analyse, evaluate, and optimise the Spherical Tokamak for Energy Production (STEP) equilibrium configuration, coupled with a viable divertor solution, and a compatible poloidal field coil system. The complexity of this task leads to a multitude of potentially conflicting requirements and competing constraints. We identify interfaces and conflicts between the aspects of the design that were previously considered independently, and highlight the key benefits, trends, and trade–offs between alternative configurations. We demonstrate that advanced exhaust solutions, simultaneously applied to both inboard and outboard divertors, are accessible with feasible coil sets under conditions relevant for STEP. We show that the most promising inner–X geometry, paired with the outer super–X configuration, can significantly enhance divertor’s power handling capability, allowing access to stable detached regimes. The coil set feasibility is further assessed considering its compatibility with the assumed plasma initiation scenario, and with the most demanding plasma current density profiles utilising alternative heating and current drive schemes.

An innovative coil fabrication and assembly process for the VNS tokamak based on in-situ winding

The paper introduces a new approach to assemble tokamak machines applied to the plasma-based Volumetric Neutron Source (VNS) device, addressing challenges related to machine topology. The VNS, designed as a compact tokamak, serves as a qualification testbed for in-vessel fusion components, especially the breeding blanket. The project aims to bridge the technology gap between ITER and DEMO. The machine size has been minimized to maximize the Neutron Wall Load (NWL). This results in a small, dense plasma and, due to the massive neutron shielding structure, a comparably large distance between the plasma and the equilibrium magnets, challenging the primary stabilization and shaping of the plasma. To overcome this, a magnet system architecture is proposed for VNS in which the poloidal field (PF) coil system is placed inside the TF coils. An assembly procedure is proposed, where the superconducting PF coils are first completed and positioned around the vacuum vessel, followed by the in-situ winding and assembly of the toroidal field coil system. The main magnetic cage components of the machine are based on High-Temperature Superconductors (HTS) to avoid the need for heat treatment. The paper presents the advantages of the magnet system architecture, and the preliminary feasibility study of the machine assembly process.

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.

Modeling of Quench in the ITER Poloidal Field Coils

Modeling of Quench in the ITER Poloidal Field Coils

FreeGSNKE: A Python-based dynamic free-boundary toroidal plasma equilibrium solver

We present a Python-based numerical solver for the two-dimensional dynamic plasma equilibrium problem. We model the time evolution of toroidally symmetric free-boundary tokamak plasma equilibria in the presence of the non-linear magnetohydrodynamic coupling with both currents in the “active” poloidal field coils, with assigned applied voltages, and eddy currents in the tokamak passive structures. FreeGSNKE (FreeGS Newton–Krylov Evolutive) builds and expands on the framework provided by the Python package FreeGS (Free boundary Grad–Shafranov). FreeGS solves the static free-boundary Grad–Shafranov (GS) problem, discretized in space using finite differences, by means of Picard iterations. FreeGSNKE introduces: (i) a solver for the static free-boundary GS problem based on the Newton–Krylov (NK) method, with improved stability and convergence properties; (ii) a solver for the linearized dynamic plasma equilibrium problem; and (iii) a solver for the non-linear dynamic problem, based on the NK method. We propose a novel “staggered” solution strategy for the non-linear problem, in which we make use of a set of equivalent formulations of the non-linear dynamic problem we derive. The alternation of NK solution steps in the currents and in the plasma flux lends this strategy an increased resilience to co-linearity and stagnation problems, resulting in favorable convergence properties. FreeGSNKE can be used for any user-defined tokamak geometry and coil configuration. FreeGSNKE's flexibility and ease of use make it a suitably robust control-oriented simulator of plasma magnetic equilibria. FreeGSNKE is entirely written in Python and easily interfaced with Python libraries, which facilitates machine learning based approaches to plasma control.

Free boundary equilibrium determination of HL-3 discharges

In 2022–2023, HL-3 (previously known as HL-2M) realized the divertor discharge after its first plasma campaign. To analyze equilibrium configurations of HL-3, a free boundary equilibrium code FEEQS, based on the finite element method, has been used in recent experiments. Free boundary equilibrium determinations are carried out for discharges performed in both limiter and divertor configurations. Compared with the experimental diagnostic data from fast camera photos and saturated ion current signals.etc, various time evolution of plasma parameters are well reproduced, such as plasma shape, currents in the poloidal field coils, strike-point displacement. The use of FEEQS successfully supported the operations in HL-3, and physics study as well, which laid a foundation for the application of free boundary equilibrium integration simulation in the next step.

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.

Magnetic field simulations of the GOLEM tokamak via the NICE code

In this paper, the direct and inverse equilibrium computation mode of the NICE code is put into operation for the case of the GOLEM tokamak. This enables, for the first time, simulations of GOLEM's magnetic field. The computation mode put into operation simulates the magnetohydrodynamic equilibrium of a tokamak plasma based on a given plasma position or given currents in poloidal magnetic field coils. In order to set up NICE, a virtual model of GOLEM is introduced that includes axially symmetrical approximations of GOLEM's iron core and primary transformer coils. The results of NICE simulations are compared with experimental measurements, and it is shown that NICE simulations produce plasma equilibrium configurations typical for GOLEM. Nevertheless, a more detailed validation of GOLEM's virtual model and of the results of NICE simulations is to be performed in the future.

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.

Current Limit Avoidance Algorithms for DEMO Operation

Tokamaks are the most promising devices to prove the feasibility of energy production using nuclear fusion on Earth which is foreseen as a possible source of energy for the next centuries. In large tokamaks with superconducting poloidal field (PF) coils, the problem of avoiding saturation of the currents is of paramount importance, especially for a reactor such as the European demonstration fusion power plant DEMO. Indeed, reaching the current limits during plasma operation may cause a loss of control of the plasma shape and/or current, leading to a major disruption. Therefore, a current limit avoidance (CLA) system is essential to assure safe operation. Three different algorithms to be implemented within a CLA system are proposed in this paper: two are based on online solutions of constrained optimization problems, while the third one relies on dynamic allocation. The performance assessment for all the proposed solutions is carried out by considering challenging operation scenarios for the DEMO reactor, such as the case where more than one PF current simultaneously saturates during the discharge. An evaluation of the computational burden needed to solve the allocation problem for the various proposed alternatives is also presented, which shows the compliance of the optimization-based approaches with the envisaged deadlines for real-time implementation of the DEMO plasma magnetic control system.