Robustness of the tokamak error field correction tolerance scaling
Abstract not provided.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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Tokamak Disruption Simulation (Final Technical Report)
This was a multi-institutional project concerned with simulation of tokamak disruption – rapid termination of the plasma and confining magnetic field in a tokamak fusion reactor – with the aim of quantifying the physics associated with disruption mitigation. The PI’s role is as one of a group of computational scientists concerned with development of effective solution algorithms for the underlying models of magnetohydrodynamics and transport that represent the physical processes, and to develop techniques of uncertainty quantification for treating unknown aspects of the model such as boundary conditions and effects of impurities in plasmas.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
The dual-electron cyclotron emission based measurement of 3D structures on DIII-D tokamak
The measurements of the locked mode and other 3D magnetohydrodynamic modes are essential for real-time control of plasma stability in future fusion reactors. Current magnetic measurements present non-negligible challenges for the application on fusion pilot plants (FPPs) due to neutron radiation and pose challenging design issues. In this paper, we discuss an alternative solution with microwave-based measurements. The dual-electron cyclotron emission (ECE) method has been developed and applied on DIII-D. The dual-ECE measurement method demonstrates clear identification of magnetic islands compared to single ECE, which is highly beneficial for locked mode feedback control on the FPPs. Toroidally separated ECE measurements provide accurate three-dimensional structure of magnetic islands and tearing modes. This technique also offers deeper physical insights into the penetration of resonant magnetic perturbations. Numerical calculations indicate that small locked modes (magnetic island width > 2 cm) can be detected using the dual-ECE measurement method in ITER plasma. This will enhance measurement capabilities for the safe operation of FPP plasmas.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Net versus gross erosion of silicon carbide in DIII-D divertor
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Physics - Plasma physics↗
Theory for transport in magnetized plasmas
We present a two fluid model for anomalous thermal, particle and momentum transport which has been derived from kinetic theory using a small parameter ε~10-2. This small parameter can come from either the ratio of wave frequency to cyclotron frequency or the ratio of potential to thermal energies. This fluid model is strongly nonlinear and has an exact closure within the accuracy ε. It can be used to simulate the full radius of a tokamak including internal and edge transport barriers at the same time without introducing any local assumptions for physics model or grid size. It involves a full transport matrix, thus giving possibility for both density and heat pinches. The model can be derived from the Vlasov equation, adding also close collisions and includes particle pinches of the type found in the magnetosphere and in the Massachusetts Institute of Technology elevated dipole machine. Pinches are also in agreement for the heat pinch experiments on Doublet III D including simultaneous particle and heat pinches. For tokamaks it gives the right scaling of confinement time versus heating power. Finally, we have also recovered the Turbulent Equipartition result that the density scales as inverse safety factor q in typical simulations. Thus this model is both fundamental and of high practical use.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Using convolutional neural networks to detect edge localized modes in DIII-D from Doppler backscattering measurements
In H-mode tokamak plasmas, the plasma is sometimes ejected beyond the edge transport barrier. These events are known as edge localized modes (ELMs). ELMs cause a loss of energy and damage the vessel walls. Understanding the physics of ELMs, and by extension, how to detect and mitigate them, is an important challenge. In this paper, we focus on two diagnostic methods—deuterium-alpha (D α ) spectroscopy and Doppler backscattering (DBS). The former detects ELMs by measuring Balmer alpha emission, while the latter uses microwave radiation to probe the plasma. DBS has the advantages of having a higher temporal resolution and robustness to damage. These advantages of DBS diagnostic may be beneficial for future operational tokamaks, and thus, data processing techniques for DBS should be developed in preparation. In sight of this, we explore the training of neural networks to detect ELMs from DBS data, using D α data as the ground truth. With shots found in the DIII-D database, the model is trained to classify each time step based on the occurrence of an ELM event. The results are promising. When tested on shots similar to those used for training, the model is capable of consistently achieving a high f1-score of 0.93. Furthermore, this score is a performance metric for imbalanced datasets that ranges between 0 and 1. We evaluate the performance of our neural network on a variety of ELMs in different high confinement regimes (grassy ELM, RMP mitigated, and wide-pedestal), finding broad applicability. Beyond ELMs, our work demonstrates the wider feasibility of applying neural networks to data from DBS diagnostic.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Advances in physics and applications of 3D magnetic perturbations on the J-TEXT tokamak
In the last two years, three major technical improvements have been made on J-TEXT in supporting of the expanded operation regions and diagnostic capabilities. (1) The successful commission of the 105 GHz/500 kW/1 s electron cyclotron resonance heating (ECRH) system increasing the core electron temperature from 0.9 keV up to around 1.5 keV. (2) The poloidal divertor configuration with an X-point in the high-field side has been achieved. In particular, the 400 kW electron cyclotron wave has also been successfully injected into the diverted plasma. (3) A 256-channel electron cyclotron emission imaging diagnostic system and two sets of four-channel Doppler backscattering diagnostics have been successfully developed on J-TEXT, allowing detailed measurement of the electron temperature and density fluctuations for turbulence and MHD research. The locked mode (LM), especially the 2/1 LM, is one of the biggest threats to the plasma operation. Both the thresholds of 2/1 and 3/1 LM are observed to vary non-monotonically on electron density. The electrode biasing was applied successfully to unlock the LM from either a rotating or static resonant magnetic perturbation (RMP) field. In the presence of 2/1 LM, three kinds of standing wave (SW) structures have been observed to share a similar connection to the island structure, i.e. the nodes of the SWs locate around the O- or X-points of the 2/1 island. The control and mitigation of disruption is essential to the safe operation of ITER, and it has been systematically studied by applying a RMP field, massive gas injection (MGI) and shattered pellet injection on J-TEXT. When the RMP-induced 2/1 LM is larger than a critical width, the MGI shutdown process can be significantly influenced. If the phase difference between the O-point of LM and the MGI valve is +90° (or –90°), the penetration depth and the assimilation of impurities can be enhanced (or suppressed) during the pre-thermal quench (TQ) phase and result in a faster (or slower) TQ. A secondary MGI can also suppress the runaway electron (RE) generation, if the additional high-Z impurity gas arrives at the plasma edge before TQ. When the secondary MGI has been applied after the formation of the RE current plateau, the RE current can be dissipated, and the dissipation rate increases with the injected impurity quantity but saturates with a maximum of 28 MA s –1 . The non-local transport is experimentally observed in the ion transport channel. The electron thermal diffusivity significantly increases with the ECRH power. Theoretical work shows that significant intrinsic current can be driven by electromagnetic turbulence, and the robust formation mechanism of the E × B staircase is identified from the Hasegawa–Wakatani system.
divertor↗
Understanding NBI heating and fueling in LTX-β tokamak
My work on the contract DE-SC=0023274 (July, 2022 - July, 2025) was focused on understanding the plasma fueling by NBI. On Dec. 1998 with Sergei Krasheninnikov (UCSD, San Diego, CA) we initiated the Li Wall Fusion (LiWFusion) as a new concept of magnetic fusion [1]. The new 1 2 plasma 3 4 5 6 2.1 Overview of rate coefficients, mean-free paths λH0 , and diffusions coefficients 6 2. GSV code for ⟨σv⟩ analysis of NBI fueling concept was a reaction of the lack of luck of tokamak fusion with QDT = 1 on TFTR and JET. We recognized the edge plasma cooling by recycling to be the route reason of the tokamak problems on the way to burning plasma. LiWFusion relies on plasma pumping by a lithium layer on the inner walls of the plasma chamber, combined with the plasma heating and fueling by the Neutral Beam Injection (NBI). These two innovations eliminate the route problem of the tokmak fusion. The concept became theoretically mature in 2006. In 2012, the technology of continuously Flowing Liquid Lithium (24/7-FLiLi) was invented by me for the future implementations of the LWFusion concept. At the same time, my every presentation on LiWFusion to International Symposium on Lithium Applications since 2010 was objected by some person with words “Everybody since the 1970s knows that plasma fueling by NBI is impossible”. I dropped the name of the author of this objection and ignored his views. My work on the current grant on NBI fueling of LTX-β not only clarified the issue but resolved it in an astonishing way. The NBI fueling was fully understood (thus humiliating the current dogmatic fusion community). The new future of LTX-β dedicated for decades to Li in tokamaks, as well as of the entire magnetic fusion program was envisioned, in sharp contrast with the fallure of OFES in the post-TFTR era of 21st century.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Similarity for downscaled kinetic simulations of electrostatic plasmas: Reconciling the large system size with small Debye length
A simple similarity has been proposed for kinetic (e.g., particle-in-cell) simulations of plasma transport that can effectively address the long-standing challenge of reconciling the tiny Debye length with the vast system size. This applies to both transport in unmagnetized plasma and parallel transport in magnetized plasmas, where the characteristics length scales are given by the Debye length, collisional mean free paths, and the system or gradient lengths. The controlled scaled variables are the configuration space, x/L, and an artificial Coulomb Logarithm, L ln Λ, for collisions, while the scaled time, t/L, and electric field, LE, are automatic outcomes. The similarity properties are examined, demonstrating that the macroscopic transport physics is preserved through a similarity transformation while keeping the microscopic physics at its original scale of Debye length. To showcase the utility of this approach, two examples of 1D plasma transport problems were simulated using the VPIC code: the plasma thermal quench in tokamaks [Li et al., Nuclear Fusion 63, 066030 (2023)] and the plasma sheath in the high-recycling regime [Li et al., Physics of Plasmas 30, 063505 (2023)].
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Variable magnetic field electron spectrometer to measure hot electrons in the range of 50–460 keV
Resonance absorption (RA) occurs when a p-polarized electromagnetic wave, obliquely incident on an inhomogeneous plasma, tunnels past its turning point and resonantly excites an electron plasma wave (EPW) at the critical density. This phenomenon is important, for instance, in the direct drive approach to inertial fusion energy and is a particular example of a wider phenomenon in plasma physics known as mode conversion, which is crucial for heating magnetic fusion devices, such as tokamaks, via RF heating. Direct measurement of these RA-generated EPW accelerated hot electrons, with energy in the range of a few tens to a few hundreds of keV, is a challenging task due to the relatively low deflecting magnetic fields needed. The solution described here is a magnetic electron spectrometer (MES) with a continually changing magnetic field, lower at the entrance of the MES and gradually increasing toward the end, that enables the measurement of a wide spectral range of electrons with energies between 50 and 460 keV. Electron spectra taken in a LaserNetUS RA experiment were acquired from plasmas generated by irradiating polymer targets with the combination of an ~300 ps pulse followed by a series of ten high intensity 50–200 fs duration laser pulses from the ALEPH laser at Colorado State University. In conclusion, the high intensity beam is designed as spike trains of uneven duration and delay pulses in order to modify the RA phenomenon.
47 OTHER INSTRUMENTATION↗
Prospects of negative triangularity tokamak for advanced steady-state confinement of fusion plasmas in MHD stability consideration
Not Available
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Critical role of current-driven instabilities for ELMs in NSTX
The impact of different extended-magnetohydrodynamic (MHD) contributions on the stability thresholds of peeling-ballooning modes in ELMing and ELM-free plasmas in the spherical tokamak NSTX is investigated with the initial value code M3D-C1. We show that ELMing discharges in NSTX are limited by resistive current-driven peeling modes, whereas non-ELMing wide-pedestal H-mode discharges are located near the ideal pressure-driven ballooning threshold. It is demonstrated that extended-MHD can lead to more reliable edge stability predictions than existing ideal-MHD models. Resistive peeling-ballooning modes are found to exist well before the ideal stability threshold is met, and kink-peeling modes exhibit considerable sensitivity to plasma resistivity. Other effects not considered in ideal-MHD models affect PB modes in NSTX in a weaker way. Gyroviscous stress appears stabilizing such that the stability boundary lies closer to the experimental point. Equilibrium rotation can suppress ideal core modes and thus isolate edge modes. These results are important for the development of a predictive pedestal model for low-aspect ratio tokamaks.
pedestal↗
Long plasma duration operation analyses with an international multi-machine (tokamaks and stellarators) database
Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international Coordination on International Challenges on Long duration OPeration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Overview of recent progress in 3D field physics in KSTAR
Various 3D field physics challenges of magnetically confined plasmas arise when the driving source comes from either externally applied non-axisymmetric 3D magnetic perturbations or plasma instabilities inside the plasma. Recently, several key outstanding topics of 3D field physics have been extensively studied in the Korean Superconducting Tokamak Advanced Research (KSTAR), such as edge-localized-mode (ELM) control by resonant magnetic perturbation (RMP), error field (EF) control, 3D field effects on rotation and transport, and RMP-induced alteration of divertor heat flux and detachment. KSTAR has a few physically unique features (i.e., high rotation and long-pulse plasmas with a low intrinsic EF) and machine/diagnostic capabilities (i.e., 3-row in-vessel control coil and state-of-the-art 2D/3D imaging diagnostics), which have been taken advantage of until now to address critical 3D field physics issues relevant to ITER and K-DEMO. Among many remarkable achievements are the robust access to and control of n = 1 RMP ELM suppression, along with a development of its physics basis tools, parameter expansion, optimization, and long-pulse control techniques. Nonetheless, a series of unresolved 3D physics themes, as well as limited coverage of 3D field operating regimes, have also been identified as future works for the 3D field research in KSTAR. In this paper, we provide an overview about the recent progress of KSTAR 3D field physics and present future plans of KSTAR 3D research toward a future fusion reactor.
3D field physics↗
Gyrokinetic simulations of electrostatic microturbulence in ADITYA-U tokamak
Global gyrokinetic simulations of the electrostatic microturbulence driven by the pressure gradients of thermal ions and electrons are carried out for the ADITYA-U tokamak geometry using its experimental plasma profiles and with collisional effects. The dominant instability is trapped electron mode (TEM) based on the linear eigenmode structure and its propagation in the electron diamagnetic direction. Collisional effects suppress turbulence and transport to a certain extent. Zonal flow is not playing a critical role in the TEM saturation, which is dominated by the inverse cascade. The frequency spectrum of the electrostatic fluctuations is in broad agreement with the experimentally recorded spectrum in the ADITYA-U, with a bandwidth ranging from ~0 to 50 kHz.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Basic Physical Processes Involving Dust in Fusion Plasmas
This report presents the main outcomes of our research focused on understanding the behavior and effects of dust particles in fusion plasmas, particularly in the edge regions of tokamaks and stellarators. We conducted computer modeling studies of burst injections of carbon and tungsten dust particles in DIII-D like divertor plasmas. The studies investigated effects of transient influx of the low- and high-Z material plasma contaminants in form of dust on the edge plasma dynamics in a modern mid-size tokamak. We also performed theoretical and computational studies of the forces acting on non-spherical dust grains in magnetized and nonmagnetized plasmas. In addition, we cooperated with General Atomics slag management group on development of DIII-D dust injection experiments to measure trajectories of dust grains in the divertor plasma and compare them with DUSTT code predictions for validation of the dust modeling capabilities. We collaborated with JET experimentalists on evaluation of radiative losses induced by injection of mixed neon-deuterium ice pellets for disruption and run-away electron generation mitigation during thermal and current quench phases. We also cooperated with experimentalists at LHD fusion device (Japan) on modeling support of experiments on injection of boron granules in fusion plasma discharges to assess their dynamics and effectiveness for in situ dynamic wall conditioning.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Drift reduced Landau fluid model for magnetized plasma turbulence simulations in BOUT++ framework
Recently the drift-reduced Landau fluid six-field turbulence model within the BOUT++ framework has been upgraded. In particular, this new model employs a new normalization, adds a volumetric flux-driven source option, the Landau fluid closure for parallel heat flux and a Laplacian inversion solver which is able to capture $n$ = 0 axisymmetric mode evolution in realistic tokamak configurations. As we report here, these improvements substantially extended model's capability to study a wider range of tokamak edge phenomena, and are essential to build a fully self-consistent edge turbulence model capable of both transient (e.g., ELM, disruption) and transport time-scale simulations.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Concept Design and Analysis of the Magnet System of the Sustained High Power Density Tokamak Facility
A recent National Academy study recommended a next-step sustained high power density (SHPD) facility for the United States that can be a bridge to a compact fusion pilot plant. A design has been initiated to investigate a possible SHPD non-deuterium-tritium device that builds upon recent low aspect ratio tokamak studies. Here, a 1.2-m device with a 2.4 aspect ratio has been chosen to evaluate physics performance goals within a double-null plasma machine arrangement centered on three component features: high current density toroidal field (TF) and central ohmic heating/poloidal field (PF) coils, liquid metal divertor/first wall systems, and the integration of a limited set of outboard dual-coolant lead lithium test blankets that can be maintained within a vertical maintenance approach.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗