Tokamak Operation Research Articles

Experimental investigation on high heat flux plasma parameters of HIT-PSI device in argon discharges

Abstract Researches on plasma-facing materials/components (PFMs/PFCs) have become a focus in magnetic confinement fusion studies, particularly for advanced tokamak operation scenarios. Similarly, spacecraft surface materials must maintain stable performance under relatively high temperatures and other harsh plasma conditions, making studies of their thermal and ablation resistance critical. Recently, a low-cost, low-energy-storage for superconducting magnets, and compact linear device, HIT-PSI, has been designed and constructed at Harbin Institute of Technology (HIT) to investigate the interaction between stable high heat flux plasma and PFMs/PFCs in scrape-off-layer (SOL) and divertor regions, as well as spacecraft surface materials. The parameters of the argon plasma beam of HIT-PSI are diagnosed using a water-cooled planar Langmuir probe and emission spectroscopy. As magnetic field rises to 2 T, the argon plasma beam generated by a cascaded arc source achieves high density exceeding 1.2×1021 m−3 at a distance of 25 cm from the source with electron temperature surpassing 4 eV, where the particle flux reaches 1024 m−2s−1, and the heat flux loaded on the graphite target measured by infrared camera reaches 4 MW/m2. Combined with probe and emission spectroscopy data, the transport characteristics of the argon plasma beam are analyzed.

First application of the island divertor configuration in the J-TEXT tokamak

Abstract For the first time, an island divertor configuration was successfully implemented in the J-TEXT tokamak to improve heat exhaust and impurity control. The magnetic island is generated by applying external resonant magnetic perturbation fields, and the intersection between the edge island and the divertor target is then controlled by adjusting the edge safety factor qa, thereby achieving the island divertor configuration. The overall confinement is maintained in spite of the loss of the edge volume. The island divertor configuration significantly reduces peak heat-load on the divertor target by approximately 50% and improves impurity screening. Additionally, it effectively modulates radiation around the magnetic island's X-point, potentially enhancing the stability and control of radiative divertor operations. These findings highlight the island divertor configuration as a promising strategy for advancing heat exhaust and impurity control in tokamak operations.

Plasma self-driven current in tokamaks with magnetic islands

Abstract Magnetic island perturbations may cause a reduction in plasma self-driven current that is needed for tokamak operation.
 A novel effect on tokamak self-driven current revealed by global gyrokinetic simulations is due to magnetic-island-induced
3D electric potential structures, which have the same dominant mode numbers as that of the magnetic island,
whereas centered at both the inner and outer edge of the island. The non-resonant potential islands
are shown to drive a current through an efficient nonlinear parallel acceleration of electrons.
In large aspect ratio (large-A) tokamak devices, this new effect can result in a significant
global reduction of the electron bootstrap current when the island size is sufficiently large,
in addition to the local current loss across the island region due to the pressure profile flattening.
It is shown that there exists a critical magnetic island width for large-A tokamaks beyond which
the electron bootstrap current loss is global and increases rapidly with the island size.
As such, this process may introduce a size limit for tolerable magnetic islands in large-A tokamak
devices in the context of steady state operation. On the other hand, the current loss caused by magnetic islands in
low-A tokamaks such as spherical tokamak (ST) NSTX/U is minor.
The reduction of the axisymmetric current by magnetic islands scales with the square of island width.
However, the loss of the current is mainly local to the island region, and the pace of current
loss as the island size increases is substantially slower compared to large-A tokamaks.
In particular, the bootstrap current reduction in STs is even smaller in the reactor-relevant
high-$\beta_p$ regime where neoclassical tearing modes are more likely to develop.

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.

Safety Analysis of Metallic Tokamak-I Against Electromagnetic and Structural Loads

During tokamak operation, the structural integrity of the vacuum vessel (VV) of Metallic Tokamak-I (MT-I), a small spherical tokamak, was evaluated. This evaluation involved simulating real experimental data of electromagnetic (EM) and structural loads using the ANSYS platform. Internal heat generation, induced currents, and inertial and pressure loads in the VV were analyzed to determine their effects on the VV. This analysis was conducted on a 180-deg sector model over a 10-ms-event period. To create multiple checkpoint events, the plasma current was assumed to be formed at variable positions of the VV, hence inducing variable current for each event. The events are divided into four cases based on the radial and vertical displacements of plasma. The response of the VV structure was calculated using coupling of EM and structural modules of ANSYS. It is observed from the numerical results that the maximum stress on the VV is in a safe range and that the temperature rise on the vessel can be reduced by natural convection only if the event is ended in 10 ms. A prolonged event can result in permanent deformation in the VV structure. A disruption event on the limiter region is also studied.

SPARC x-ray diagnostics: Technical and functional overview.

An overview is given of SPARC's three main x-ray diagnostics, with a focus on the functions they fulfill with respect to tokamak operation. The first is an in-vessel soft x-ray tomography diagnostic, aimed at providing early campaign information on plasma position, MHD activity, and impurity content. The second is an ex-vessel set of hard x-ray scintillators aimed at detecting the presence of runaway electrons, in particular during plasma startup phases. The third is a set of x-ray Bragg spectrometers, located outside of the tokamak hall, aimed at informing on the ion temperature as an indirect constraint to reduce uncertainties on the fusion power, on providing plasma rotation velocity estimates, and on observing impurity emission. Finally, more technical details are given on the beamlines at the end of which the spectrometers are located. It explains how their design allows us to ensure tritium containment and limit neutron radiation while providing a straight view into the plasma that can also be used for testing new innovative sensors.

Global particle buildup simulations with gas puff scan: application to WEST discharge

This paper deals with the distribution of sources, transport, and exhaust of particles in a tokamak. Knowledge and understanding of all the physical phenomena involved in the global particle buildup are necessary to study and predict density regimes and subsequently to develop optimized scenarios for tokamak operation in order to control heat and particle exhaust. Neutral particles and their interactions with plasma are central in this perspective. This paper discusses the impact of varying the intensity of particle fueling in 2D transport simulations of a WEST discharge. Simulations are performed with an updated version of SOLEDGE-HDG that allows a more realistic transport of neutrals using a self-consistent diffusive model based on charge exchange and ionization processes. New code capabilities allow the entire WEST poloidal cross section to be simulated in a realistic configuration for both geometry and the range of control parameters. A gas puff scan illustrates the main features of the sheath-limited, high-recycling, and detached regimes, such as the buildup of the temperature gradient and the pressure drop in the scrape-off layer (SOL), the target temperature falling to 1 eV, and the ionization source moving away from the targets, as well as the particle flux rollover. A crude estimate of wall erosion is also provided, showing the respective role of each plasma wall component in each of these regimes.

Initial testing of Alfvén eigenmode feedback control with machine-learning observers on DIII-D

A first of its kind fully data-driven system has been developed and implemented into the DIII-D plasma control system to detect and control Alfvén eigenmodes (AE) in real-time. Susceptibility to fast ion-induced AE is a challenge in fully non-inductive tokamak operation, which significantly reduces fast-particle confinement and results in degraded fusion gain. Controlling AEs in real-time to improve fast-ion confinement is, hence, important for future advanced tokamak fusion reactors. The models were implemented and tested in experiments which showed that neural networks (NN) are highly effective in detecting 5 types of AE (BAE, EAE, LFM, RSAE, TAE) using high resolution ECE. To estimate the neutron deficit, a NN has been trained that outputs the classical neutron rate using similar inputs to NUBEAM. Also a preliminary ML-based proportional control has been designed and gone through initial testing in experiment to use feedback-control on the neutral beam power to achieve desired amplitude of AE modes and neutron deficits. The effect of AEs on fast-ion confinement is measured by analysing the gap in classical neutron rate from the proposed NN-based NUBEAM and the measured neutron rate.

Non-linear MHD investigations of high-confinement regimes without type-I ELMs in ASDEX Upgrade and JT-60SA

Large edge localised modes (ELMs) would cause an unacceptable reduction of material lifetime in future large tokamaks due to the significant amount of energy expelled from the magnetically confined region towards the plasma facing components. Thoroughly validated modelling of regimes devoid of large ELMs is crucial as it may then provide predictive insights prior to tokamak operation and design. This paper describes recent efforts pursued with the non-linear extended MHD code JOREK in the modelling of three scenarios without large ELMs: quiescent H-mode (QH-mode), quasi-continuous exhaust regime (QCE regime), and the enhanced D-alpha H-mode (EDA H-mode). For each of these regimes, the non-linear dynamics observed in the simulations are detailed and compared to experimental observations of the underlying instabilities of each regime (edge harmonic oscillation for QH-mode, small ELMs for QCE regime, and quasi-coherent mode for EDA H-mode). For QH-mode, the kink-peeling mode is found to govern the dynamics and a transition to a large ELM is obtained above the same density threshold as in the modelled experiment. For the QCE regime and EDA H-mode, resistive peeling–ballooning modes dominate and pedestal fluctuation frequencies correspond well to experimental observations. The dominant mechanisms for the excitation and suppression of these instabilities are presented and their influence on simulation dynamics is shown. Finally, predictive simulations of edge instabilities at different values of plasma resistivity in a 4.60 MA scenario with low edge safety factor in JT-60SA are presented.

Roles of non-axisymmetric perturbations in free drift vertical displacement events on EAST

The safe operation of most tokamaks, especially the large ones, relies on the feedback control of vertical displacement events (VDEs). However, most of these feedback control systems are based on axisymmetric VDE models. In this study, we use NIMROD simulations to study the role of non-axisymmetric perturbations in free drift vertical displacement events on EAST. The high-n modes in the non-axisymmetric VDE grow first, which drives the formation of high-n magnetic island chains. Subsequently, the magnetic island chains grow and overlap with each other, leading to the destruction of the magnetic flux surface, which induces a minor disruption and accelerates the start of the following major disruption. The magnetic island and the stochastic magnetic field allow the toroidally non-axisymmetric poloidal plasma current to jet towards the hoop force direction, forming finger-like and filamentary structures. Such a plasma current non-axisymmetry strongly depends on the anisotropy in the thermal transport coefficients.

Suppression of the m/n = 2/1 tearing mode by electron cyclotron resonance heating on J-TEXT

Stabilization of tearing modes and neoclassical tearing modes is of great importance for tokamak operation. Electron cyclotron waves (ECWs) have been extensively used to stabilize the tearing modes with the virtue of highly localized power deposition. Complete suppression of the m/n = 2/1 tearing mode (TM) by electron cyclotron resonance heating (ECRH) has been achieved successfully on the J-TEXT tokamak. The effects of ECW deposition location and power amplitude on the 2/1 TM suppression have been investigated. It is found that the suppression is more effective when the ECW power is deposited closer to the rational surface. As the ECW power increases to approximately 230 kW, the 2/1 TM can be completely suppressed. The island rotation frequency is increased when the island width is reduced. The experimental results show that the local heating inside the magnetic island and the resulting temperature perturbation increase at the O-point of the island play dominant roles in TM suppression. As the ECW power increases, the 2/1 island is suppressed to smaller island width, and the flow shear also plays a stabilizing effect on small magnetic islands. With the stabilizing contribution of heating and flow shear, the 2/1 TM can be completely suppressed.

Tokamak divertor plasma emulation with machine learning

Future tokamak devices that aim to create conditions relevant to power plant operations must consider strategies for mitigating damage to plasma facing components in the divertor. One of the goals of MAST-U tokamak operations is to inform these considerations by researching advanced divertor configurations that aid stable plasma detachment. Machine design, scenario planning and detachment control would all greatly benefit from tools that enable rapid calculation of scenario-relevant quantities given some input parameters. This paper presents a method for generating large, simulated scrape-off layer data sets, which was applied to generate a data set of steady-state Hermes-3 simulations of the MAST-U tokamak. A machine learning model was constructed using a Bayesian approach to hyperparameter optimisation to predict diagnosable output quantities given control-relevant input features. The resulting best-performing model, which is based on a feedforward neural network, achieves high accuracy when predicting electron temperature at the divertor target and carbon impurity radiation front position and runs in around 1 ms in inference mode. Techniques for interpreting the predictions made by the model were applied, and a high-resolution parameter scan of upstream conditions was performed to demonstrate the utility of rapidly generating accurate predictions using the emulator. This work represents a step forward in the design of machine learning-driven emulators of tokamak exhaust simulation codes in operational modes relevant to divertor detachment control and plasma scenario design.

Evaluation of cross sections for fast ion reactions with beryllium in helium and hydrogen fusion plasmas

To computationally support hydrogen and helium plasma discharges in the early stages of tokamak operation and to support the commissioning of the neutron detectors during these operational phases, creation of a realistic neutron and gamma ray particle source for Monte Carlo simulations will be needed. One of the most important parts of creating the particle source is calculating the reaction rates of the particle-emitting reactions to determine the emission profile in the plasma and the energy spectra of the emitted particles. In this paper the analysis and evaluation of cross sections for important neutron-emitting reactions, namely, 9Be(p,nγ)9B, 9Be(3He,nγ)11C, and charged-particle emission reactions 9Be(p,d)2α and 9Be(p,α)6Li that cause neutron emission in the next step of interactions are presented. The reaction cross sections were evaluated based on experimental measurements and empirical models describing the interaction of two charged particles. Evaluation of the associated uncertainties was also performed. The main goal of the work is to propose the newly evaluated cross sections for inclusion in the FENDL nuclear data library, thus making the cross section available to other researchers studying the above listed reactions.

Experiments and gyrokinetic simulations of TCV plasmas with negative triangularity in view of DTT operations

Negative triangularity (NT) scenarios in TCV have been compared to positive triangularity (PT) scenarios using the same plasma shapes foreseen for divertor tokamak test tokamak operations. The experiments provided a NT/PT L-mode pair and a PT H-mode with different heating mixes. Regardless of the heating mix, NT L-modes always reached higher values of plasma pressure with respect to PT L-modes with the same power and recovered the central pressure of PT H-mode scenarios heated with up to twice the injected power. The experimental analysis shows that this enhanced performance in NT is due to larger temperature and density gradients close to the edge ( ρtor=[0.8−1.0] ) and higher values of pressure at the separatrix. Local gyrokinetic simulations agree with the experimental results and are able to catch the effect of shaping alone. Integrated modeling performed with ASTRA-TGLF reproduces reasonably well the PT shot but is not able to fully capture the improvements in the NT shot.

Error field detection and correction studies towards ITER operation

In magnetic fusion devices, error field (EF) sources, spurious magnetic field perturbations, need to be identified and corrected for safe and stable (disruption-free) tokamak operation. Within Work Package Tokamak Exploitation RT04, a series of studies have been carried out to test the portability of the novel non-disruptive method, designed and tested in DIII-D (Paz-Soldan et al 2022 Nucl. Fusion 62 126007), and to perform an assessment of model-based EF control strategies towards their applicability in ITER. In this paper, the lessons learned, the physical mechanism behind the magnetic island healing, which relies on enhanced viscous torque that acts against the static electro-magnetic torque, and the main control achievements are reported, together with the first design of the asynchronous EF correction current/density controller for ITER.

Experimental study of core MHD behavior and a novel algorithm for rational surface detection based on profile reflectometry in EAST

Microwave reflectometry is a powerful diagnostic that can measure the density profile and localized turbulence with high spatial and temporal resolution and will be used in ITER, so understanding the influence of plasma perturbations on the reflect signal is important. The characteristics of the reflect signal from profile reflectometry, the time-of-flight (TOF) signal associated with the MHD instabilities, are investigated in EAST. Using a 1D full-wave simulation code by the Finite-DifferenceTime-Domain (FDTD) method, it is well validated that the local density flattening could induce the discontinuity of the simulated TOF signal and an obvious change of reflect amplitude. Experimental TOF signals under different types of MHD instabilities (sawtooth, sawtooth precursors and tearing mode) are studied in detail and show agreement with the simulation. Two new improved algorithms for detecting and localizing the radial positions of the low-order rational surface, the cross-correlation and gradient threshold (CGT) method and the 2D convolutional neural network approach (CNN) are presented for the first time. It is concluded that TOF signal analysis from profile reflectometry can provide a straightforward and localized measurement of the plasma perturbation from the edge to the core simultaneously and may be a complement or correction to the q-profile control, which will be beneficial for the advanced tokamak operation.

Magnetic diagnostics layout design for CFETR plasma equilibrium reconstruction

Plasma equilibrium reconstruction provides essential information for tokamak operation and physical analysis. An extensive and reliable set of magnetic diagnostics is required to obtain accurate plasma equilibrium. This study designs and optimizes the magnetic diagnostics layout for the reconstruction of the equilibrium of the plasma according to the scientific objectives, engineering design parameters, and limitations of the Chinese Fusion Engineering Test Reactor (CFETR). Based on the CFETR discharge simulation, magnetic measurement data are employed to reconstruct consistent plasma equilibrium parameters, and magnetic diagnostics’ number and position are optimized by truncated Singular value decomposition, verifying the redundancy reliability of the magnetic diagnostics layout design. This provides a design solution for the layout of the magnetic diagnostics system required to control the plasma equilibrium of CFETR, and the developed design and optimization method can provide effective support to design magnetic diagnostics systems for future magnetic confinement fusion devices.

Flux pumping studies and Ti peaking by q-profile shaping using ECCD in ASDEX Upgrade

On ASDEX Upgrade a prioritized ‘advanced Tokamak’ program has been run during the last two experimental campaigns, focusing on the effect of non-standard current profiles on the behavior of high beta plasmas. Two lines of non-standard (i.e. non-sawtoothing) q-profiles are followed: (1) plasma self-organization of centrally flat q-profiles (qmin ≈ 1) via ‘flux pumping’, here based on a naturally occurring continuous (1,1) mode and (2) externally shaped q-profiles with qmin > 1, allowing for more experimental freedom of the q-profile to be established. The advantages and requirements of both concepts are discussed. The invited presentation at the workshop focuses on specific effects of Electron Cyclotron Current Drive (ECCD). With respect to flux pumping it could be shown that increasing the plasma beta raises the flux pumping capability. More central coECCD can be redistributed if beta is increased. This is in line with qualitative theory. For the case with qmin > 1, it is shown that small variations of the ECCD profile can modify the peaking of the central ion temperature. For larger radii no significant changes of the kinetic profiles have been observed as the q-profile is changed. Hfactors have not exceeded 1.2, well below the assumptions of several models for steady state tokamak operation. To achieve these changes of q at larger radii we have used ctr-ECCD in the plasma center, increasing the ohmic current globally. The net effect of central ctr-ECCD and increased ohmic current is a strong off-axis current drive, allowing H-modes with Te ≈ Ti ≈ 8 keV and n¯e ≈ 6 · 1019m−3 ≈ 0.5 nGW at q95 = 4 to be run stationary with qmin ≈ 1.4. The design of these discharges was strongly supported by inter-shot model based optimization.

Faraday angle prediction based on 1-D CNN on the J-TEXT tokamak

The Faraday angle is an indispensable physical parameter to evaluate and reconstruct the current density profile which is imperative for advanced tokamak operation and physics study. The POLARIS system has been implemented on J-TEXT since 2012 for the measurement of the Faraday angle and this system occasionally encounters operational failures that result in the inability to provide measurements. So a deep learning model is trained to predict the Faraday angle when the POLARIS is unable to provide the measurement. The model is primarily based on a 1-D convolutional neural network (CNN) and is equipped with multiple convolutional paths that can accommodate various lengths of diagnostic data. The data of 100 shots are used to train the 1-D CNN model, and 30 shots are used to optimize the hyperparameters and select the model which performs best during the training process. The results on a test set of 30 shots show that the model obtains a satisfactory fitting effect.

Non-uniformity of the radial electric field for tokamak plasma on the low and the high toroidal magnetic field side

This paper presents a relativistic mechanism that can create a non-uniformity of a radial electric field on the low side and the high toroidal magnetic field side of tokamak plasmas. This mechanism is associated with the significant current velocity of electrons in the plasma for many experimental modes of tokamak operation. It is important that electric fields arise without the separation of charges in the plasma. The traditional approach to determining radial electric fields in tokamak plasmas gives a uniform radial electric field on magnetic surfaces inside the plasma, while there are no radial electric fields outside the plasma. The presented mechanism leads to the appearance of an electric field outside the plasma with a current, and a non-uniformity of the radial electric field on magnetic surfaces appears inside the plasma. The proposed mechanism can explain the experimentally measured non-uniformity of C+6 carbon ion toroidal rotation velocities in the Tore-Supra tokamak plasma.