Tokamak Operation Research Articles

Introduction of a 3D global non-linear full-f particle-in-cell model for runaway electrons in JOREK

Abstract Disruptions present one of the leading concerns for reliable tokamak operation. The acceleration of electrons from the thermal bulk to relativistic energies, so-called runaway electron (RE) generation, is in particular a problem for future high current machines such as ITER. Accurately predicting the generation and impact of REs is therefore essential for making informed decisions concerning machine design and the use of disruption mitigation systems. This requires high-fidelity modeling also accounting for the large MHD activity observed throughout disruptions, which is made especially difficult by the mutual coupling between REs and the companion plasma. The non-linear 3D extended MHD code JOREK is a powerful tool for studying disruption and RE physics. This work details recent developments in JOREK, introducing a hybrid fluid-kinetic model where the REs are modeled kinetically and coupled to the non-linear MHD equations using a full-f particle-in-cell approach. The model goes beyond the state of the art and can accurately capture phase space distributions and dynamics of REs, drift orbits, and transport and losses caused by stochastic fields. Benchmarks are presented for both 2D and 3D configurations, concerning the impact of REs on the force balance and linear tearing mode growth rates, where a good agreement with analytically derived results is found. In addition, a demonstration of a particularly complicated non-linear application with high relevance to large machines is made, namely a RE benign termination linked to a violent burst of MHD activity.

Design of 3D equilibria and coils for steady-state operation of tokamaks

Abstract Inspired by the respective advantages of tokamaks and stellarators, we propose a quasi-axisymmetric (QA) tokamak-stellarator hybrid concept as a steady-state operation mode of standard tokamaks. Using stellarator optimization tools, we obtain compact 3D equilibria with good QA and sufficient (~20%) rotational transform generated by external coils to replace current drive, with the rest (~80%) generated self-consistently by the bootstrap current. The equilibria can be supported by 3D saddle coils that are compatible with typical tokamak coils and reasonably feasible from engineering perspectives. We consider and compare designs with either high-field-side or low-field-side shaping, and find that the latter is favorable for coil optimization but unfavorable for generating rotational transform and achieving good QA. These results highlight a relatively unexplored region in the space of toroidal fusion configurations.

Strain analysis by neutron diffraction on Nb3Sn strands in ITER central solenoid conductors of short and long twist pitch

Abstract The ITER central solenoid (CS) conductors are composed of Nb3Sn superconducting cables and stainless steel jackets. Approximately 60 000 electromagnetic loading cycles will be applied to the Nb3Sn strands in the CS conductor over the course of ITER tokamak operation, and the CS conductor is required to maintain the current sharing temperature (T cs) for these electromagnetic loading cycles. However, in CS conductor prototypes, degradation of T cs was observed after electromagnetic loading cycles. Visual inspections of the tested CS conductors revealed large bending and buckling of the Nb3Sn strands. These strand deformations were considered to be the cause of T cs degradation because even a small amount of strain markedly affects the critical current of Nb3Sn strands. To prevent the strands from bending and buckling, the twist pitch of the cable was shortened to improve stiffness. The stiffer cables did prevent T cs from degrading after electromagnetic loading cycles owing to the shortened twist pitch of the conductors, but a slight increase of T cs was observed. Visual inspections revealed no significant bent or buckled strands, but small deformations cannot be investigated visually. Thus, internal strain was measured by neutron diffraction and the internal strain states of the prototype CS conductors having long twist pitches and the improved short-twist-pitch (STP) CS conductors were evaluated. The results indicated that, after electromagnetic loading cycles, bending of the strands in the STP CS conductor was limited and compressive strain was reduced. Therefore, we determined that this STP is not only effective to prevent degradation caused by bent strands but it also increases T cs by reducing compressive strain.

Investigation of divertor heat flux characteristics under the influence of resonant magnetic perturbations on EAST

Abstract Controlling the splitting divertor heat flux caused by resonant magnetic perturbations (RMPs) is a topic of concern for fusion devices. As a fundamental prerequisite, it is necessary to understand the characteristics of the heat flux distribution under the applied RMP field, which will be studied in this paper. The nonlinear phenomenon of strike point splitting was found to be strongly dependent on the plasma response under RMP. These splitting heat flux distributions are qualitatively explained by the simulated magnetic footprints. The RMP phase scanning experiment shows that scanning in a certain range of relative phase can maintain good ELM mitigation and simultaneously sweep the striations of heat flux on divertor target. Additionally, even in upper single null (USN) configurations, heat stripes are observed on the lower outer divertor (LO-div), attributed to RMP-induced additional magnetic connections to the LO-div, as confirmed by magnetic topology simulations. A dedicated investigation into the impact of the discrepancy between lower and upper separatrix radii mapped to the low field side mid-plane (〖dR〗_sep) reveals its significant influence on both ELM control and heat flux distribution. Within a certain range of 〖dR〗_sep, the heat flux distribution is improved while ELM suppression is maintained. These findings contribute to divertor heat flux understanding under RMP conditions in tokamak operations.

Experimental study on metallic impurity behavior with boronization wall conditioning in EAST tokamak

Operation of EAST tokamak with full metal wall without any wall conditioning are attempted in 2023 and 2024 experimental campaign to address the issues related to ITER tungsten wall operation. It is found that H-mode plasma could be sustained even with a substantial increase in metallic impurity content caused by strong plasma-wall interaction under uncoated metal wall. Boronization wall conditioning is therefore performed to improve the plasma performance with higher injected power. This study aims to quantitatively assess the impact of boronization wall conditioning on metallic impurity concentration and behavior in the EAST tokamak. It is then proved to be an effective wall conditioning approach for significantly controlling high-Z impurity content. In this work, the impurity spectra at extreme ultraviolet (EUV) wavelength range measured by sets of fast-time-response and space-resolved EUV spectrometers are widely used in the data analysis. The variation in the boron content in plasma after boronization are investigated by monitoring the 2nd order of B V line at 48.59 Å. It is found that the persistence time of boron in EAST device after a boronization with 10 g of carborane (C2B10H12) is about 2000 s (∼150 shots) of discharge duration. The impact of different wall conditions (uncoated metal wall and boron coated wall) on metallic impurity content are then quantitatively studied. After boronization, the concentration of tungsten (CW) and molybdenum (CMo) dropped by 85 %, e. g., from 2.0 × 10-4 to 4.1 × 10-5 and from 4.6 × 10-5 to 6.3 × 10-6, respectively. While the concentration of copper (CCu) and iron (CFe) decreased by approximately 50 % and 65 %, respectively, e. g., from 4.3 × 10-5 to 2.1 × 10-5 and from 2.0 × 10-4 to 6.9 × 10-5. A comparison of the line emission profiles from tungsten ions of W26+ − W32+ and W43+ before and after boronization reveals an overall reduction in the intensity while without obvious change in the profile shape, which suggests a reduction in metallic impurities source after boronization instead of altering impurity transport in core plasma significantly.

Complex for Thomson Scattering Diagnostics on the TRT Tokamak

A diagnostic system for Thomson scattering of the central, edge and divertor plasma regions of a tokamak with reactor technologies is discussed. The rationale and choice of technical solutions are given, the composition of the Thomson scattering diagnostic complex is discussed, as well as an estimate of the accuracy of measuring the electron temperature and plasma density in the central edge and divertor regions of the TRT tokamak. Particular attention is paid to ensuring the functionality of the proposed diagnostics in the reactor mode of the tokamak operation and the results of testing diagnostic equipment in experiments on the Globus-M2 tokamak.

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 q a, 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.

Numerical study of the synergetic current drive by lower hybrid waves and loop voltage

Abstract Thanks to the tail of fast electrons generated by the lower hybrid (LH) waves, the plasma conductivity changes significantly with respect to the ohmic heating status (Fisch 1985 Phys. Fluids 28 245). In this paper, the synergy effects of lower hybrid current drive (LHCD) in the presence of a DC electric field (namely, the loop voltage V loop ≠ 0) on the EAST tokamak are studied numerically using the ray tracing and Fokker–Planck tools of C3PO/LUKE. In addition to the normal case of a positive loop voltage with an asymmetric power spectrum (namely, co-current LHCD), the synergy effects of negative loop voltage with an asymmetric power spectrum (namely, counter LHCD), and positive/negative loop voltage with a symmetric power spectrum are also investigated, which are previously rarely reported. It is found that the total plasma conductivity can be increased by a factor of 1.68–2.18 (depending on the value of V loop) when a 1.1 MW LH power at 2.45 GHz with a symmetric power spectrum is injected. Furthermore, unlike the asymmetric power spectrum, the direction of the sum of the pure LH current (I lh0) and the synergetic current (I ad) resulting from the variation in plasma conductivity is always in line with the ohmic current (I oh) for a symmetric power spectrum, which is favorable for AC tokamak operation. These simulated results are helpful for LHCD applications, especially in current ramp-up and AC operation.

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-β p regime where neoclassical tearing modes are more likely to develop.

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.

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.