Toroidal Field Coils Research Articles

Evaluation of localized mechanical properties of modified N50 welded joints at cryogenic temperature through a digital image correlation technique

The modified N50 austenitic stainless steel has been determined to be applied as the superconducting magnet supporting material of the Chinese Fusion Engineering Test Reactor (CFETR), especially the jacket material of the CICCs of both the TF coils and the CS coils and the case material of the TF coils, which inevitably requires welding connection. The welding process generally results in welding defects as well as degradation of cryogenic mechanical properties of metallurgical inhomogeneous zones. However, limited research has been reported to demonstrate the distinct localized performance of these zones at cryogenic temperature. The present paper presents an approach to characterize localized mechanical properties of the modified N50 welds at both room temperature (300 K) and cryogenic temperature (6 K). We identified the welding zone (WZ) through metallographic analysis and then the boundaries of the heat-affected zone (HAZ) were recognized by hardness measurements at room temperature. Finally, the localized optical digital image correlation (DIC) technique was used to obtain the localized yield strength Rp0.2 and percentage extension e of the WZ, the HAZ, and the base metal (BM) at both room and cryogenic temperature, of which the observed trend is verified through comparison with that of hardness measurements at room temperature. The technique proposed provides the localized measurement of cryogenic tensile properties of weld metals and also presents experimental data on different zones of the modified N50 welded joints as a jacket material for the next-generation fusion reactors.

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

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.

Local plastic deformation effects on the critical current of high-Jc Nb3Sn strands under uniaxial strain

Abstract In order to meet the target operating parameters of the toroidal field coils (TFC) for the next-generation Chinese compact burning plasma tokamak, high critical current density (Jc) Nb3Sn strand will be applied to the high-field winding-package (WP) of the TFC. To improve the transverse stiffness of the cable in withstanding the huge Lorentz force and avoiding conductor performance degradation, the Short-Twist-Pitch (STP) and Copper-Wound-Strand (CWS) cable patterns were taken into consideration. In the processes of cabling and compaction of the conductor, the tight cable configurations lead to severe local plastic deformation (LPD) within the strands. The strands in the conductor are subjected to strain caused by thermal contraction and Lorentz force during conductor cooling down and operation. So far it is unknown, whether the LPD could impact the critical current (Ic) versus uniaxial applied strain behaviour of high-Jc Nb3Sn strand. Aiming to investigate the effect of LPD on the Ic of strands under uniaxial strain, three types of high Jc Nb3Sn strand with different indentation depths were tested on a U-shaped bending spring. The axial strain ranges from -0.9% to +0.4% at 14 T and 4.2 K. The three types of strands showed more strain sensitivity and lower tensile irreversible strain limit with increasing LPD, while even irreversible degradation of the Ic could be observed in the compressive strain region. The sample preparation, test process, test results and analysis are reported.

Electromagnetic Analysis of the Demo Magnet System: Electrical Behavior of a Toroidal Field Coil During an Electrical Transient

Electromagnetic Analysis of the Demo Magnet System: Electrical Behavior of a Toroidal Field Coil During an Electrical Transient

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.

Completion of all the ITER toroidal field coil structures

All 19 ITER toroidal field coil structures (TFCSs) were successfully fabricated by the Japan Domestic Agency with solving technical challenges. This study presents the results of the welding-deformation control process, which is a remarkable challenge in an efficient fabrication of the TFCSs.

TF Coil Displacement Distortion FE Analysis Simulating Non-Constant Assembly Gap Between Wedging Interfaces of TF Coils

TF Coil Displacement Distortion FE Analysis Simulating Non-Constant Assembly Gap Between Wedging Interfaces of TF Coils

Simulation of the JT-60SA Supercritical Helium Toroidal Field Coil Loop During Fast Safety Discharge Using Simcryogenics. Comparison With Experimental Data and Extrapolation to Higher Currents

Simulation of the JT-60SA Supercritical Helium Toroidal Field Coil Loop During Fast Safety Discharge Using Simcryogenics. Comparison With Experimental Data and Extrapolation to Higher Currents

Superconductor based, tomographic, neutron diagnostics for fusion power monitoring.

We propose a scalable system of compact, superconducting neutron monitors, which can be embedded in any existing cryogenic infrastructure of a fusion system. The pixel-based nature of the detectors allows them to be placed at intervals following the circumference of a cooled zone, e.g., a field coil, thus allowing for a tomographic measurement of the neutron flux surrounding the plasma. An early stage prototype of the superconducting bolometer is described, and the key results of a previous feasibility study of this prototype performed with cold neutrons are summarized. The bolometer can be adapted for use with fast neutrons by altering the composition and geometry of the neutron-to-heat conversion layer. This paper describes the initial feasibility considerations for implementation in a superconducting tokamak. The sensor is based on a high-temperature superconductor, making it possible to select the operation temperature in the range 1-90K. Neutron flux numbers were found using the ITER MCNP reference model, and these were embedded in a TOPAS model to find the expected signal measured by the bolometer at the position of a toroidal field coil. The results at the coil position indicate suitable operation levels in terms of the magnitude of the measured signal, with a measurable signal of several ohm, which is much smaller than the saturation energy of the detector. Radiation hardness is estimated and found to be on the order of at least 40 years for the relevant radiation levels. The upcoming investigation activities of the project are described for both radiation testing and analytical modeling.

ITER Toroidal Field Coil Delivery and Preparatory Work at the ITER Site

ITER Toroidal Field Coil Delivery and Preparatory Work at the ITER Site

Finite Element Analysis of the Energized ITER TF Coil in Test Conditions

Finite Element Analysis of the Energized ITER TF Coil in Test Conditions

MADE Macro-Block III: Transitioning From the Optimized TF Coil Shape to the PF Coil System

MADE Macro-Block III: Transitioning From the Optimized TF Coil Shape to the PF Coil System

Technical studies about the prestressed insulation cylinder of HL-3 tokamak's center-post

The TF coil of HL-3 consists of 20 D-shaped demountable bundles, and all their L-shaped center sections are bonded together as the center-post. To increase its overall rigidity to resist torsional loads and avoid de-bonding failure, the center-post is wrapped by an insulation cylinder with enough prestress. Additionally, the insulation cylinder needs to be of good dimensional precision because the CS coil is winded around it. This paper describes the results of the technical studies carried out during the development of the insulation cylinder, including material selection, parameter calculation, enlacing tests and curing tests.

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.

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.

Evaluating Conductor Design and Stability Performance for the Conductor of High-Field Winding Package in CFETR TF Coil

Evaluating Conductor Design and Stability Performance for the Conductor of High-Field Winding Package in CFETR TF Coil

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.