Burning Plasma Devices Research Articles

Design of a diamond-based in-vessel soft x-ray detector for the SPARC tokamak.

The in-vessel silicon diode arrays that are used for soft x-ray detection in many tokamaks are sensitive to neutron damage, making them unsuitable for burning plasma devices such as SPARC. In such a device, the silicon diodes would need to be placed far from the plasma-limiting their field of view-or an alternative detector could be used. Here, we present the design of a camera containing an array of chemical vapor deposition single-crystal diamonds, which will be placed in the upper and lower port plugs of the SPARC tokamak with a large enough view of the poloidal cross section to enable tomographic inversion. The camera design presented here is optimized to provide a wide field of view of the poloidal cross section. Simulated plasma conditions are used to estimate the x-ray signal that this detector array will receive and to fine-tune the camera placement within the tokamak.

GaN-based W-band receiver chip development for fusion plasma diagnostics

Millimeter-wave diagnostics have proven effective on various magnetic fusion devices worldwide, yet the formidable challenges posed by the harsh environments of future burning plasma devices, characterized by extreme temperatures, pressures, and radiation levels, remain a significant hurdle. To address these challenges, the utilization of wide bandgap Gallium Nitride (GaN)-based millimeter-wave diagnostics is a most promising solution for fusion reactor safety monitoring and control. A noteworthy W-band GaN-based system-on-chip receiver has been the demonstrated by employing HRL T3 40 nm GaN technology. This receiver chip, compactly designed with dimensions of 3 × 5 mm2, incorporates essential components such as the 75–110 GHz RF Low-Noise Amplifier (LNA), mixer, Intermediate Frequency (IF) amplifier, and Local Oscillator (LO) chain. This receiver chip will be packaged as a millimeter-wave receiver module and applied on the DIII-D National Fusion Facility, for fusion plasma edge shape monitoring for operational safety and dangerous disruption prediction. The laboratory measurement results have demonstrated suitable performance. This advancement is pivotal for accurate analysis of plasma behavior in the extreme conditions of burning plasma devices, driving progress in fusion research and technology.

Core performance predictions in projected SPARC first-campaign plasmas with nonlinear CGYRO

This work characterizes the core transport physics of SPARC early-campaign plasmas using the PORTALS-CGYRO framework. Empirical modeling of SPARC plasmas with L-mode confinement indicates an ample window of breakeven (Q > 1) without the need of H-mode operation. Extensive modeling of multi-channel (electron energy, ion energy, and electron particle) flux-matched conditions with the nonlinear CGYRO code for turbulent transport coupled to the macroscopic plasma evolution using PORTALS reveals that the maximum fusion performance to be attained will be highly dependent on the near-edge pressure. Stiff core transport conditions are found, particularly when fusion gain approaches unity, and predicted density peaking is found to be in line with empirical databases of particle source-free H-modes. Impurity optimization is identified as a potential avenue to increase fusion performance while enabling core-edge integration. Extensive validation of the quasilinear TGLF model builds confidence in reduced-model predictions. The implications of projecting L-mode performance to high-performance and burning-plasma devices is discussed, together with the importance of predicting edge conditions.

Comparison of machine learning systems trained to detect Alfvén eigenmodes using the CO2 interferometer on DIII-D

A Machine-Learning (ML) based detection scheme that automatically detects Alfvén Eigenmodes (AE) in a labelled DIII-D database is presented here. Controlling AEs is important for the success of planned burning plasma devices such as ITER, since resonant fast ions can drive AEs unstable and degrade the performance of the plasma or damage the first walls of the machine vessel. Artificial Intelligence could be useful for real-time detection and control of AEs in steady-state plasma scenarios by implementing ML-based models into control algorithms that drive actuators for mitigation of AE impacts. Thus, the objective is to compare differences in performance between using two different recurrent neural network systems (Reservoir Computing Network and Long Short Term Memory Network) and two different representations of the CO2 phase data (simple and crosspower spectrograms). All CO2 interferometer chords are used to train both models, but only one is processed during each training step. The results from the model and data comparison show higher performance for the RCN model (True Positive Rate = 90% and False Positive Rate = 14%), and that using simple magnitude spectrograms is sufficient to detect AEs. Also, the vertical CO2 interferometer chord passing near the center is better for ML-based detection of AEs.

Optical plasma boundary detection and its reconstruction on EAST tokamak

Plasma boundary detection and reconstruction are important not only for plasma operation but also for plasma facing materials. Traditional methods, for example, EFIT code, which is constrained by electromagnetic measurement, and is very challenging for detecting the plasma boundary in long-pulse burning plasma devices such as ITER. A novel algorithm for the reconstruction of the plasma boundary using one visible camera has been developed on experimental advanced superconducting tokamak (EAST) for fusion reactors. A U-Net convolutional neural network was used to identify the plasma boundary and the pixel coordinates of the boundary points were fitted with EFIT via the XGBoost model. This algorithm can transform the boundary from the image plane to the poloidal plane of the Tokamak based on machine learning without traditional spatial calibration, and then the reconstruction of the plasma configuration shall be realized based on a monocular visible light camera. The reconstruction accuracy of this algorithm is relatively high. The average error on the test set was only 7.36 mm (<1 cm) and satisfied the accuracy requirements of control for EAST tokamak. This result can contribute to the development of the plasma boundary reconstruction and operation based on one visible camera.

Advances in prediction of tokamak experiments with theory-based models

The successful validation of theory-based models of transport, magnetohydrodynamic stability, heating and current drive, with tokamak measurements over the last 20 years, has laid the foundation for a new era where these models can be routinely used in a ‘predict first’ approach to design and predict the outcomes of experiments on tokamaks today. The capability to predict the plasma confinement and core profiles with a quantified uncertainty, based on a multi-machine, international, database of experience, will provide confidence that a proposed discharge will remain within the operational limits of the tokamak. Developing this predictive capability for the first generation of burning plasma devices, beginning with ITER, and progressing to tokamak demonstration reactors, is a critical mission of fusion energy research. Major advances have been made implementing this predict first methodology on today’s tokamaks. An overview of several of these recent advances will be presented, providing the integrated modeling foundations of the experimental successes. The first steps to include boundary plasmas, and tokamak control systems, have been made. A commitment to predicting experiments as part of the planning process is needed in order to collect predictive accuracy data and evolve the models and software into a robust whole discharge pulse design simulator.

Plasma position measurement with collimated neutron flux monitor diagnostics on JET

In future burning plasma devices measuring the plasma position and its shape with great spatial and temporal resolution is a crucial task. Nowadays these information are obtained by means of magnetic coils installed inside the vacuum vessel that in the future devices (like ITER and DEMO), due to the harsh environment caused by the high plasma temperature, will experience degradation over the time. Furthermore, the long plasma discharges will result in large integration drift causing a lack of accuracy in the measured signal. In order to assist the magnetic diagnostics and at the same time provide a novel tool to benchmark them, here the measurement of the plasma magnetic axis position by means of a collimated neutron flux monitor is proposed. Three different methods are here described and applied on JET by means of the neutron camera: a weighted average, the asymmetry method and a neural network. The methods are calibrated on a large database of plasma discharges including NBI and ICRH heated ones, and then compared with the magnetic axis position reconstructed by EFIT. The neural network outperforms the two other methods. In particular, the asymmetry method results to be sensitive to MHD activity, NBI power variation and to neutron emissivity profiles presenting a strong asymmetry, such as in case of peripheral NBI deposition due to high density plasmas or ICRH resonance layer. A possible application to vertical displacement events and disruptions is discussed and envisaged for future applications on DEMO. Finally, the performances of the neural network and of the asymmetry methods are studied in the case of one or two missing channels in the neutron flux monitor, showing how in general the reconstruction of the radial magnetic axis in both methods is more sensitive to the lack of channels than the vertical one. The methods here proposed can be applied and benchmarked on DTT and ITER neutron cameras as part of a future real-time control system.

The role of edge plasma parameters in H-mode density limit on the JET-ILW

A study of a dataset of JET H-mode plasma with the Be/W ITER-like wall (JET-ILW) shows that reaching the edge MHD ballooning limit leads to confinement degradation. However, unlike JET plasmas with a carbon wall (JET-C), the JET-ILW plasmas stay in a marginal dithering phase for a relatively long period, associated with a higher (≈20%) H-mode density limit (HDL) than JET-C equivalents. This suggests that ITER could be operated in H-mode with higher density than the scaling based on carbon wall devices, but likely with a dithering phase plasma with lower confinement. A new, reliable estimator for JET E r, min has been derived by combining HRTS measurements of pedestal gradient and edge-SOL decay lengths. JET radial E r ETB wells are observed in the range of −15 to −60 kV m−1 in high performance H-modes, consistent with previous CXRS results in ASDEX Upgrade. The results imply that a higher positive E × B shear in the near SOL plays a role in sustaining a marginal phase in JET-ILW which leads to a higher HDL than that in JET-C. The results of the JET-ILW dataset show agreement with the Goldston finite collisionality HD model for SOL broadening at high collisionality. A hypothesis for the dithering H-mode phase is proposed: as n e,SOL increases, ν ∗,SOL increases, SOL broadens, E r shear decreases, triggers L-mode; n e drops, ν ∗,SOL decreases, SOL becomes narrower, and E r shear increases, triggering H-mode, resulting in a cycle of H–L–H- oscillations. For burning plasma devices, such as ITER, operating just below the MHD limit for the dithering phase could be a promising regime for maximising core density, and fusion performance while minimising plasma-material interaction. The oscillatory signal during the dithering phase could be used as a precursor of undesirable plasma performance for control purposes.

Follow the power—pathways to steady-state tokamak reactors

Diagramming pathways of dimensionless power is a potent method for extrapolating between operating points on present-day tokamaks and future burning plasma devices. The heat transport power, current drive power, H-mode threshold power and other ‘plasma physics’ powers can be expressed in dimensionally correct (or normalized) form as Pa 3/4, where P is the power and a is the plasma minor radius, with the relative gyroradius (ρ *) dependence ranging from gyro-Bohm-like for transport (), Bohm-like for current drive () and worse than Bohm-like for H-mode threshold (). The D–T fusion power cannot be normalized in the same fashion since it is governed by nuclear physics, but at fixed B T it scales like . Other ‘mixed physics’ powers can be incorporated into the dimensionless power framework by holding B T fixed in the same manner. Diagramming these dimensionless powers vs ρ * shows how the pathway to a steady-state reactor can be optimized relative to various operational boundaries. Using a steady-state hybrid discharge with β N = 3.2 from DIII-D as the starting point, a multi-parameter optimization finds an attractive pathway to steady-state operation on ITER using 76 MW of current drive power (fusion gain of Q fus = 8), along with a pathway to Q fus = 20 in a JET-sized steady-state reactor with B T = 10 T.

New Trends in Microwave Imaging Diagnostics and Application to Burning Plasma

Microwave and millimeter-wave (mm-wave)-based diagnostics have proven invaluable in the understanding of magnetic confinement fusion plasma science and have found wide application at major fusion facilities, including DIII-D, EAST, ASDEX-U, HL-2A, KSTAR, and J-TEXT. This trend is expected to continue unabated in future burning plasma devices such as ITER and DEMO. Attractive features include their nonperturbing nature and relatively modest access requirements due to their short wavelength, as well as recent dramatic advances in the state of the technology. In this tutorial, we describe the basic principles of a number of microwave diagnostics measurements emphasizing imaging, recent technology advances including the so-called “system-on-chip (SoC)” technology, and “synthetic” diagnostics which simulate the response of diagnostics under actual experimental scenarios. The application of microwave diagnostics on reactor plasmas benefits immensely from the development of wide-bandgap electronics and advanced optics designs, both of which ameliorate the effects of the harsh environment.

Fully non-inductive second harmonic electron cyclotron plasma ramp-up in the QUEST spherical tokamak

Fully non-inductive second (2nd) harmonic electron cyclotron (EC) plasma current ramp-up was demonstrated with a newlly developed 28 GHz system in the QUEST spherical tokamak. A high plasma current of 54 kA was non-inductively ramped up and sustained stably for 0.9 s with a 270 kW 28 GHz wave. A higher plasma current of 66 kA was also non-inductively achieved with a slow ramp-up of the vertical field. We have achieved a significantly higher plasma current than those achieved previously with the 2nd harmonic EC waves. This fully non-inductive 2nd harmonic EC plasma ramp-up method might be useful for future burning plasma devices and fusion reactors, in particular for operations at half magnetic field with the same EC heating equipment.

Impact of E × B shear flow on low-n MHD instabilities.

Recently, the stationary high confinement operations with improved pedestal conditions have been achieved in DIII-D [K. H. Burrell et al., Phys. Plasmas 23, 056103 (2016)], accompanying the spontaneous transition from the coherent edge harmonic oscillation (EHO) to the broadband MHD turbulence state by reducing the neutral beam injection torque to zero. It is highly significant for the burning plasma devices such as ITER. Simulations about the effects of E × B shear flow on the quiescent H-mode (QH-mode) are carried out using the three-field two-fluid model in the field-aligned coordinate under the BOUT++ framework. Using the shifted circular cross-section equilibriums including bootstrap current, the results demonstrate that the E × B shear flow strongly destabilizes low-n peeling modes, which are mainly driven by the gradient of parallel current in peeling-dominant cases and are sensitive to the Er shear. Adopting the much more general shape of E × B shear ([Formula: see text]) profiles, the linear and nonlinear BOUT++ simulations show qualitative consistence with the experiments. The stronger shear flow shifts the most unstable mode to lower-n and narrows the mode spectrum. At the meantime, the nonlinear simulations of the QH-mode indicate that the shear flow in both co- and counter directions of diamagnetic flow has some similar effects. The nonlinear mode interaction is enhanced during the mode amplitude saturation phase. These results reveal that the fundamental physics mechanism of the QH-mode may be shear flow and are significant for understanding the mechanism of EHO and QH-mode.

Establishment and Assessment of Plasma Disruption and Warning Databases from EAST**supported by the National Magnetic Confinement Fusion Science Program of China (No. 2014GB103000)

Disruption database and disruption warning database of the EAST tokamak had been established by a disruption research group. The disruption database, based on Structured Query Language (SQL), comprises 41 disruption parameters, which include current quench characteristics, EFIT equilibrium characteristics, kinetic parameters, halo currents, and vertical motion. Presently most disruption databases are based on plasma experiments of non-superconducting tokamak devices. The purposes of the EAST database are to find disruption characteristics and disruption statistics to the fully superconducting tokamak EAST, to elucidate the physics underlying tokamak disruptions, to explore the influence of disruption on superconducting magnets and to extrapolate toward future burning plasma devices. In order to quantitatively assess the usefulness of various plasma parameters for predicting disruptions, a similar SQL database to Alcator C-Mod for EAST has been created by compiling values for a number of proposed disruption-relevant parameters sampled from all plasma discharges in the 2015 campaign. The detailed statistic results and analysis of two databases on the EAST tokamak are presented.

Exploration of the Super H-mode regime on DIII-D and potential advantages for burning plasma devices

A new high pedestal regime (“Super H-mode”) has been predicted and accessed on DIII-D. Super H-mode was first achieved on DIII-D using a quiescent H-mode edge, enabling a smooth trajectory through pedestal parameter space. By exploiting Super H-mode, it has been possible to access high pedestal pressures at high normalized densities. While elimination of Edge localized modes (ELMs) is beneficial for Super H-mode, it may not be a requirement, as recent experiments have maintained high pedestals with ELMs triggered by lithium granule injection. Simulations using TGLF for core transport and the EPED model for the pedestal find that ITER can benefit from the improved performance associated with Super H-mode, with increased values of fusion power and gain possible. Similar studies demonstrate that the Super H-mode pedestal can be advantageous for a steady-state power plant, by providing a path to increasing the bootstrap current while simultaneously reducing the demands on the core physics performance.

Discovery of stationary operation of quiescent H-mode plasmas with net-zero neutral beam injection torque and high energy confinement on DIII-D

Recent experiments in DIII-D [J. L. Luxon et al., in Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] have led to the discovery of a means of modifying edge turbulence to achieve stationary, high confinement operation without Edge Localized Mode (ELM) instabilities and with no net external torque input. Eliminating the ELM-induced heat bursts and controlling plasma stability at low rotation represent two of the great challenges for fusion energy. By exploiting edge turbulence in a novel manner, we achieved excellent tokamak performance, well above the H98y2 international tokamak energy confinement scaling (H98y2 = 1.25), thus meeting an additional confinement challenge that is usually difficult at low torque. The new regime is triggered in double null plasmas by ramping the injected torque to zero and then maintaining it there. This lowers E × B rotation shear in the plasma edge, allowing low-k, broadband, electromagnetic turbulence to increase. In the H-mode edge, a narrow transport barrier usually grows until MHD instability (a peeling ballooning mode) leads to the ELM heat burst. However, the increased turbulence reduces the pressure gradient, allowing the development of a broader and thus higher transport barrier. A 60% increase in pedestal pressure and 40% increase in energy confinement result. An increase in the E × B shearing rate inside of the edge pedestal is a key factor in the confinement increase. Strong double-null plasma shaping raises the threshold for the ELM instability, allowing the plasma to reach a transport-limited state near but below the explosive ELM stability boundary. The resulting plasmas have burning-plasma-relevant βN = 1.6–1.8 and run without the need for extra torque from 3D magnetic fields. To date, stationary conditions have been produced for 2 s or 12 energy confinement times, limited only by external hardware constraints. Stationary operation with improved pedestal conditions is highly significant for future burning plasma devices, since operation without ELMs at low rotation and good confinement is key for fusion energy production.

Observation of Critical-Gradient Behavior in Alfvén-Eigenmode-Induced Fast-Ion Transport

Experiments in the DIII-D tokamak show that fast-ion transport suddenly becomes stiff above a critical threshold in the presence of many overlapping small-amplitude Alfvén eigenmodes (AEs). The threshold is phase-space dependent and occurs when particle orbits become stochastic due to resonances with AEs. Above threshold, equilibrium fast-ion density profiles are unchanged despite increased drive, and intermittent fast-ion losses are observed. Fast-ion Dα spectroscopy indicates radially localized transport of the copassing population at radii that correspond to the location of midcore AEs. The observation of stiff fast-ion transport suggests that reduced models can be used to effectively predict alpha profiles, beam ion profiles, and losses to aid in the design of optimized scenarios for future burning plasma devices.

Measurements of the radial profile of the plasma isotopic composition in JET plasmas using Alfvén eigenmodes

The measurement of the plasma isotopic composition is necessary in future burning plasma devices such as ITER and DEMO as a tool for optimizing the deuterium/tritium (DT) fusion performance. This paper reports on the results of experiments performed on the JET tokamak where toroidal Alfvén eigenmodes (TAEs) with toroidal mode number (N) up to were actively driven with a set of in-vessel antennas, and were then used to infer the value of the plasma isotopic composition at different radial positions. A novel and important result with respect to previous work on JET is that by correctly including the effect of plasma impurities in the calculation of the Alfvén frequency, through its dependence on the plasma mass, it has become now possible to distinguish plasmas with different majority ion species but with the same charge-to-mass ratio, notably majority deuterium and helium4 plasmas. Furthermore, and combined with modelling of AEs in JET discharges, these experimental results indicate that a diagnostic system based on the detection of AEs with different toroidal mode numbers and at different frequencies, could provide profile measurements of the plasma isotopic composition in future burning plasma devices such as ITER and DEMO.

The ITPA disruption database

A multi-device database of disruption characteristics has been developed under the auspices of the International Tokamak Physics Activity magneto-hydrodynamics topical group. The purpose of this ITPA disruption database (IDDB) is to find the commonalities between the disruption and disruption mitigation characteristics in a wide variety of tokamaks in order to elucidate the physics underlying tokamak disruptions and to extrapolate toward much larger devices, such as ITER and future burning plasma devices. In contrast to previous smaller disruption data collation efforts, the IDDB aims to provide significant context for each shot provided, allowing exploration of a wide array of relationships between pre-disruption and disruption parameters. The IDDB presently includes contributions from nine tokamaks, including both conventional aspect ratio and spherical tokamaks. An initial parametric analysis of the available data is presented. This analysis includes current quench rates, halo current fraction and peaking, and the effectiveness of massive impurity injection. The IDDB is publicly available, with instruction for access provided herein.

HELIAS module development for systems codes

In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations.This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work.This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

Summary of ECE Presentations at EC-18

There were nine ECE and one EBE presentation at EC-18. Four of the presentations were on various aspects of ECE on ITER. The ITER ECE diagnostic has entered an important detailed preliminary design phase and faces several design challenges in the next 2-3 years. Most of the other ECE presentations at the workshop were focused on applications of ECE diagnostics to plasma measurements, rather than improvements in technology, although it was apparent that heterodyne receiver technology continues to improve. CECE, ECE imaging and EBE imaging are increasingly providing valuable insights into plasma behavior that is important to understand if future burning plasma devices, such as ITER, FNSF and DEMO, are to be successful.