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DOE OSTI2024

Experimental evidence of enhanced radial transport in small ELM regimes at DIII-D

Small/type-II edge-localized-modes (ELMs), carrying 1% of the plasma stored energy, are found to deposit only 45 ± 5% of the ELM power near the strike point, and the remaining 55 ± 5% to the far scrape-off-layer (SOL). Small ELMs spread their power over a larger area compared to type-I ELMs, where such a ratio is about 60% and 40% for near- and far-SOL regions, respectively. The larger spread is reflected in the heat flux width (λq) in the SOL for the intra-small ELMs profile of 6.0 mm, almost a factor 2 larger than that of type-I ELMs of 3.15 mm, for similar plasma conditions and magnetic configuration. At the ELM peak, the small ELMs λq is found to be up to 4 times larger than for the type-I ELMs, going from 2 to 7.9 mm, indicating enhanced radial transport in the neon-seeded small ELM scenario. Inter-ELM λqs have been also calculated at the secondary outer divertor in quasi-double-null (QDN) discharges. It is found that, on average, λq is 2.2 times larger in the high-separatrix-density small ELM regime, compared to a reference type-I ELM one. These findings are supported by small ELMs radial velocity profiles, measured at the outer midplane with a fast reciprocating probe, showing a decay length (λvr) in the SOL of 12.8 cm, which is 3.3 times larger than that for the type-I ELMs of 3.9 cm. This analysis shows that small ELMs, although attractive for future machines due to low peak heat flux and large λq, might be of concern for the larger flux to the outer wall.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Measurement of Stark-split beam and carbon charge exchange emissions for simultaneous B-field and temperature/rotation analysis at DIII-D

A set of two newly designed, single-channel Czerny–Turner spectrometers has been deployed at the DIII-D tokamak for measurements of the motional Stark effect (MSE) split beam emission and the C6+ (CVI) carbon charge exchange recombination (CER) emission at high spectral (δλ = 0.13 nm) and temporal (1–5 kHz) resolution. High throughput optics (f/# = 2.8) allow for good signal-to-noise at high time resolution using fast EMCCD detectors. The MSE emission allows for spectral fitting of the magnitude and direction of the local B-field, while the carbon emission yields local ion temperature and toroidal rotation information. To reduce so-called Doppler broadening of the MSE emission, a new channel-specific variable lens-masking approach has been developed. Experimental data collected from the 2023 DIII-D experimental campaign demonstrate the signal quality and instrument fidelity for both diagnostic measurements. Moreover, initial CER data analysis shows a clear evolution of the toroidal rotation during edge localized modes. Initial progress on the advanced MSE model, including a new validated ray-trace model of the DIII-D collection optics, is shown via sensitivity analysis.

Instruments & Instrumentation↗

DOE OSTI2025

First-wall fluxes from large and small ELMs

Experimental analysis at DIII-D shows that small edge localized modes (ELMs) deposit a larger fraction of their energy to the first wall, compared to type-I ELMs in similar magnetic configuration and input power. The energy (λQ) and particle (λΓ) flux decay lengths in the scrape-off layer (SOL) are up to 3 and 5 times larger, respectively, for small ELMs than for larger type-I ELMs. Transport dynamics of ELM filaments in the SOL are found to be related to divertor conditions, where high divertor collisionality, typical for partially detached plasmas, is associated with increased cross field ELM radial fluxes. Results show that a sufficiently large outer wall gap and/or limiters might be needed in future scenarios to protect the first wall, if operating with small ELMs and a cold divertor. This might also have implications for RF heating in future devices, where the coupling efficiency is dependent to some degree to the outer-wall gap.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Evolution of intermittent filaments in the scrape-off layer of NSTX

Filamentary structures naturally arise from background turbulence in the scrape-off layer (SOL) of plasmas, leading to significant particle and heat transport that can degrade overall plasma confinement. This enhanced transport can contribute to unacceptably high heat loads on plasma-facing components. As such, understanding the physics of SOL plasma filaments is critical for predicting and mitigating their effects in future fusion devices. On the National Spherical Torus Experiment (NSTX), plasma filaments—commonly referred to as “blobs”—were investigated using the gas-puff imaging (GPI) diagnostic in the edge and SOL regions. The analysis involved identifying, segmenting, and tracking the characteristic contours of the blobs in each frame of the GPI video sequences. Their evolution was characterized through shape descriptors, velocity, and angular velocity derived from their contour coordinates. The results indicate that as the blob area increases, their shapes become more concave and less circular, suggesting reduced structural stability in larger blobs. This result aligns with previous theoretical results where it was shown that larger blobs are more susceptible to instabilities [Krasheninnikov et al., J. Plasma Phys. 74, 679–717 (2008) and D'Ippolito et al., Phys. Plasmas 18, 060501 (2011)]. A positive correlation was observed between radial velocity and radial position, suggesting radially outward acceleration of the filaments, potentially driven by decreasing viscous drag toward the far SOL. Interestingly, blobs in background SOL turbulence exhibited minimal spinning in contrast to filaments originating from edge localized modes, which show substantial rotation during their paths [Lampert et al., Phys. Plasmas 29, 102502 (2022)]. Statistical analysis of the solidity and total curvature shape descriptors, along with their temporal evolution, revealed relatively broad, near-Gaussian distributions. This suggests that blob morphology is strongly influenced by stochastic turbulent processes in the surrounding plasma environment. Blob parameters were also compared with bulk plasma and radial profile measurements. Notable trends were found between blob rotation and poloidal velocity with collisionality and line-integrated density. These findings contribute to a deeper understanding of blob dynamics and provide valuable insights for refining SOL turbulence models.

Covariance and correlation↗

DOE OSTI2019

Surface Erosion of Plasma-Facing Materials Using an Electrothermal Plasma Source and Ion Beam Micro-Trenches

Erosion characteristics of tungsten-alternative plasma-facing materials (PFMs) were tested under high heat flux conditions in the electrothermal plasma source facility at Oak Ridge National Laboratory. The PFMs of interest are high-purity β-3C chemical vapor deposition silicon carbide (SiC) and the MAX phases Ti 3 SiC 2 and Ti 2 AlC [MAX = chemical formula M n+1 AX n , where M is an early transition metal (such as Ti or Ta), A is an A-group element (such as Si or Al), and X is carbon or nitrogen]. Additionally, an erosion analysis method was developed using a combination of focused ion beam microscopy and scanning electron microscopy, carving micro-trench geometries into polished sample surfaces. Samples of SiC, Ti 3 SiC 2 , and Ti 2 AlC were exposed to the electrothermal plasma source alongside tungsten and monocrystalline silicon. Samples were exposed to a Lexan polycarbonate (C 16 H 14 O 3 ) electrothermal plasma stream in a He environment, at a specified impact angle, with infrared camera diagnostics. Edge localized mode–relevant heat fluxes of 0.9 to 1 GW/m 2 over 1-ms discharges were generated on the target surfaces. Tungsten samples exhibited pronounced melt-layer formation and deformation, with measured molten pits 2 to 10 μm in diameter and melt-layer depths of up to 7 μm deep. Surface erosion rates for Ti 3 SiC 2 and Ti 2 AlC ranged from 80 to 775 μm/s and 85 to 470 μm/s, respectively. Both MAX phases exhibited extreme surface fracture and material ejection, with damage depths past 4 μm for Ti 2 AlC and 11 μm for Ti 3 SiC 2 . Lastly, SiC displayed the best performance, in one case surviving 15 consecutive electrothermal plasma exposures with an average erosion rate of about 29 μm/s and no surface fracturing. SiC erosion rates ranged from 23 to 128 μm/s.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Reconfiguration of an Electrothermal-Arc Plasma Source for In Situ PMI Studies

An electrothermal-arc plasma source (ET-Arc) has been developed to produce transient plasma heat and particle fluxes similar to those produced by edge localized modes onto divertor plasma-facing components in tokamaks. The ET-Arc utilizes a capacitive discharge to send current through a 4-mm-diameter, 9-cm-long capillary source liner. The liner material is ablated to form a high-velocity plasma jet that impacts the target downstream. With the current discharge circuit configuration, pulse lengths are 1 to 2 ms in duration and deliver heat fluxes of 0.25 to 2.1 GW m-2. The plasma was previously characterized with optical emission spectroscopy (OES) on helium emission lines. The He I line ratios were interpreted with collisional radiative analysis to calculate ne and Te. The electron temperature and electron density ranged from Te = 1 to 5 eV and ne = 1022 to 1028 electrons/m3, respectively.

Lindquist, Elizabeth G.↗

DOE OSTI2025

Doppler Backscattering Data Analysis and Integrated Modeling with OMFIT

One Modeling Framework for Integrated Tasks (OMFIT) is a widely used software tool in the magnetic fusion research community. OMFIT provides magnetic fusion energy researchers with a framework for the development of special-purpose physics modules. This paper describes an OMFIT physics module pertaining to the Doppler Backscattering (DBS) fusion plasma diagnostic. DBS measures density fluctuations and flow velocity through plasma scattering of electromagnetic waves. The OMFIT DBS module was developed to analyze experimental DBS data and facilitate modeling of DBS systems installed on multiple tokamak devices. The OMFIT DBS module is designed to support several analysis workflows: detailed analysis of experimental data, experimental planning, and theory-based synthetic diagnostic modeling. The DBS module uses integrated modeling by leveraging other OMFIT physics modules to perform tasks related to DBS, e.g. ray/beam–tracing simulations, edge-localized mode–synchronized data analysis, magnetic equilibrium reconstruction, and fitting kinetic profile data. Furthermore, this paper describes several supported workflows and serves a reference for the OMFIT DBS module.

Doppler backscattering↗

DOE OSTI2025

Design and Analysis of an Integrated Additively Manufactured Test Article for Plasma-Facing Components

Current plasma-facing components (PFCs) used in helium-cooled divertor modules are complex structures with tungsten tile, steel sleeve components, and cartridges, all assembled in a helium-cooled multiple jet (HEMJ) structure. The goal of this project is to simplify the complex PFC design using additive manufacturing techniques to create a single integrated tungsten test article. Apart from the flexibility this opens up in exploring a wide array of geometries for the article, having a single integrated article significantly reduces the number of joints and parts in the article, thus reducing chances of leaks. A process called electron beam melting has shown to produce very high-density samples and unique geometries, enabling HEMJ or similar designs. To validate and optimize this novel design, the model underwent a series of computational fluid dynamics and finite element analysis simulations to replicate steady-state heat flux in the divertors. The simulations presented in this study consider a steady-state base heat flux of 5 MW/m 2 , with water serving as the coolant. Future research will explore the use of helium as a coolant, simulate edge-localized-mode conditions, and include experimental validation. Since 3D-printed tungsten is anisotropic, the build direction versus build plane of the article are taken into consideration for the test article strength. Because of the high operating temperatures and low ductility of tungsten, thermal creep and brittle fracture are important failure mechanisms to consider. In conclusion, the cap is evaluated with various flow velocities and nozzle diameters, and an optimal design choice is made for which this cap will survive the divertor conditions with a conservative safety margin.

Divertor↗

DOE OSTI2025

Wide-Bandgap Semiconductor Amplifiers for Fusion Plasma Heating and Control

This paper discusses power electronics developed under the ARPA-E GAMOW program to support nuclear fusion power production. The goal of this project was to develop and assess the potential for wide-bandgap (WBG) semiconductor devices in power electronics to enable high-efficiency and high-voltage solid-state systems for fusion plasma generation, heating, and control. The power electronics use an architecture in which multiple high-power boards can be combined to produce megawatt-level power, where using multiple boards provides high reliability. Two main areas of power electronics boards are developed in this project for fusion plasma heating and control applications: (1) pulse generation and control and (2) radiofrequency generation. The first area is for boards capable of driving high-voltage millisecond pulses at high duty cycles. The envisioned application of these pulses is in plasma control of magnetohydrodynamic instabilities, plasma position, and edge-localized modes. Pulse-width modulation allows for the implementation of a wide variety of linear and nonlinear control systems. The boards developed for this project could actuate control coils based on digital input signals and can be parallelized to provide megawatts of output power. The design of the pulse generator is a low-side load switch. A load switch was designed and constructed that utilized 2-kV-rated field-effect transistor (FET)-based cascodes developed by Qorvo under this project to perform initial testing of these cascodes. The second area is being implemented using class E amplifiers with WBG devices and a reactance steering network to handle inductive or capacitive plasma loads. Applications include ion cyclotron resonance heating (ICRH) and high-harmonic fast-wave (HHFW) heating. A class E reactance steering network is demonstrated in modeling and experiment with a resistive-inductive load that models an inductively-coupled plasma. Power combining of boards with class E reactance steering networks is also simulated and demonstrated experimentally, to enable scaling up to high power. Modeling of high-power-density cooling and remaining useful life is conducted to enable reliable, effectively cooled high-power electronics for fusion applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Stabilizing effect of enhanced resistivity on peeling-ballooning instabilities on EAST

Previous stability analysis of NSTX equilibrium with lithium-conditioning demonstrates that enhanced resistivity due to the increased effective charge number Z eff (i.e. increased impurity level) can provide a stabilizing effect on low-n edge localized modes. This paper extends the resistivity stabilizing effect to the intermediate-n peeling-ballooning (PB) instabilities with the linear stability analysis of EAST high-confinement mode equilibria in NIMROD two-fluid calculations. However, in this work, the resistivity stabilizing effect on PB instabilities in the EAST tokamak appears weaker than that found in NSTX. This work may provide better insight into the physical mechanism behind the beneficial effects of impurities on pedestal stability.

EAST tokamak↗

DOE OSTI2023

Drift kinetic effects on plasma response to resonant magnetic perturbation for EU DEMO design

Abstract A systematic investigation of the plasma response to an externally applied resonant magnetic perturbation (RMP) field, for the purpose of controlling edge localized modes, is carried out for an EU DEMO reference plasma. Particular emphasis is placed on the role of kinetic effects associated with both thermal particles and fusion-born alphas. The single fluid, resistive model predicts a large peak amplification of the n = 1 ( n is the toroidal mode number) plasma response to the target equilibrium, which is found to be close to the Troyon no-wall limit. A more advanced response model, including kinetic resonances between the RMP perturbation and drift motions of thermal and energetic particles, on the other hand, finds a strong suppression of the n = 1 field amplification. A major role is played by the precessional drift resonance of fusion-born alphas. A strong parallel sound wave damping model is found to well reproduce the full kinetic response results for the DEMO plasma, in terms of both the resonant field response amplitude and the plasma displacement. Finally, both fluid and kinetic models produce similar responses for the n = 2 and 3 RMP fields for the considered DEMO plasma, whilst kinetic effects again become important for the n = 4 RMP due to proximity of the reference plasma to the no-wall limit for the n = 4 ideal kink instability.

Physics↗

DOE OSTI2024

Assessment of vertical stability for negative triangularity pilot plants

Abstract Negative triangularity (NT) tokamak configurations may be more susceptible to magneto-hydrodynamic instability, posing challenges for recent reactor designs centered around their favorable properties, such as improved confinement and operation free of edge-localized modes. In this work, we assess the vertical stability of plasmas with NT shaping and develop potential reactor solutions. When coupled with a conformal wall, NT equilibria are confirmed to be less vertically stable than equivalent positive triangularity (PT) configurations. Unlike PT, their vertical stability is degraded at higher poloidal beta. Furthermore, improvements in vertical stability at low aspect ratio do not translate to the NT geometry. NT equilibria are stabilized in PT vacuum vessels due to the increased proximity of the plasma and the wall on the outboard side, but this scenario is found to be undesirable due to reduced vertical gaps which give less spatial margin for control recovery. Instead, we demonstrate that informed positioning of passively conducting plates can lead to improved vertical stability in NT configurations on par with stability metrics expected in PT scenarios. An optimal setup for passive plates in highly elongated NT devices is presented, where plates on the outboard side of the device reduce vertical instability growth rates to 16% of their baseline value. For lower target elongations, integration of passive stabilizers with divertor concepts can lead to significant improvements in vertical stability. Plates on the inboard side of the device are also uniquely enabled in NT geometries, providing opportunity for spatial separation of vertical stability coils and passive stabilizers.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Chasing the multi-modal plasma response in MAST-U

Abstract Achieving edge localized modes (ELMs) suppression in spherical tokamaks by applying resonant magnetic perturbations (RMPs) has proven challenging. The poloidal spectrum of the applied RMP is a key parameter that has an impact on the capability to mitigate and eventually suppress ELMs. In this work the resistive magnetohydrodynamic code MARS-F (Liu et al 2000 Phys. Plasmas 7 3681) is used to evaluate the possibility of directly measuring the plasma response in MAST-U, and particularly its variation as function of the applied poloidal spectrum, in order to guide the experimental validation of the predicted best RMP configuration for ELM suppression. Toroidal mode number n = 2 RMP is considered to minimize the presence of sidebands, and to avoid the deleterious core coupling of n = 1. Singular Value Decomposition is used to highlight linearly independent structures in the simulated magnetic 3D fields and how those structures can be measured at the wall where the magnetic sensors are located. Alternative ways to measure the multimodal plasma response and how they can be used to infer the best RMP configuration to achieve ELM suppression are also presented, including the plasma displacement and the 3D footprints at the divertor plates.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

First observation of RMP ELM mitigation on MAST Upgrade

Abstract The first experimental attempts at controlling edge localised modes (ELMs) via the application of resonant magnetic perturbations on the MAST Upgrade tokamak are reported. Using the linear MHD model MARS-F, the phase shift between the upper and lower coil rows ΔΦ was optimised for toroidal mode number n = 1 and n = 2 fields, to provide forward guidance to experiments. In low β N discharges, the application of n = 1 3D fields caused the ELM frequency f E L M to increase by over a factor 20 relative to the reference, and also induced a locked mode, which did not cause a plasma termination nor an H-L back transition. However when β N was raised, this induced locked mode caused plasma termination which precluded mitigation access. Initially, applying a numerically optimised n = 2 field had no effect. However applying a rigid toroidal shift to this field caused a locked mode disruption, demonstrating the presence of a substantial n = 2 error field. Coil current ramps were conducted with ΔΦ set at 6 different values, resulting in either locked mode disruptions or no effect, but mitigation with n = 2 fields was not established.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Achievement of highly radiating plasma in negative triangularity and effect of reactor-relevant seeded impurities on confinement and transport

The first achievement of highly radiating plasmas in negative triangularity (NT) is shown with an operational space featuring high core radiation at high Greenwald fraction obtained with the injection of reactor-relevant seeded gases. These NT shape diverted discharges reach high values of normalized plasma pressure ( β N > 2) at high radiation fraction with no edge localized modes (ELMs) and a simultaneous reduction of the divertor heat flux. Here, we demonstrate that as long as the impurity level in the core is kept low to avoid excessive fuel dilution and impurity accumulation, integration of NT configuration with high radiation fraction not only is achievable but it can lead to confinement improvement with stabilization effects originating from collisionality, E × B shear and profiles changes due to impurity radiation cooling. The underlying physics mechanism is robust and reproducible and holds for a variety of impurity species. The absence of the requirement to stay in H-mode translates in a higher core radiation fraction potentially allowed in NT shape effectively mitigating the power exhaust issue. The results presented here demonstrate a path to high performance, ELM free and highly radiative regime with reactor-relevant seeding gases making this regime a potential new scenario for reactor design.

confinement improvement↗

DOE OSTI2025

A review of collaborative studies between the NSTX/-U and MAST/-U spherical tokamaks

The National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory in the United States, and the mega ampere spherical tokamak (MAST) at the United Kingdom Atomic Energy Authority in the United Kingdom, and their respective upgrades (NSTX-U and MAST-U) are two MAST fusion devices that have operated roughly over the past two decades. Both devices have made significant contributions to understanding spherical tokamak (ST) plasma physics, and fusion plasmas in general, and both have contributed data to multi-machine database studies. Several diagnostics have been physically moved from one machine to the other by diagnostic teams working on both devices. Collaboration has benefited both research teams in the areas of operational expertise, scenario development, and equilibrium reconstruction techniques. More focused comparative studies between the two devices have been pursued over the years in many areas as well, including stability calculations, disruption characterization, pedestal and edge localized mode stability, confinement and transport, energetic particles, and heating and current drive modelling. Together NSTX/-U and MAST/-U set the stage for the future of STs, which is entering the phase of design of demonstration power plant devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

High-field side scrape-off layer density profile measurements and implications for high-field side LHCD coupling in DIII-D

The high-field side (HFS) scrape-off layer (SOL) is an often under-diagnosed region of tokamak plasmas. Situated in a region with favorable curvature, the HFS SOL has minimal turbulence-induced radial transport compared to the low-field side. Using profile reflectometry, the HFS SOL density profile is measured with high temporal resolution for a wide range of DIII-D plasma discharges for the first time. The magnetic configuration, particularly the location of the secondary separatrix (SS) largely determines the HFS SOL density profile. Additionally, density perturbations induced by edge-localized modes (ELMs) are observed in the HFS SOL, and the impact of ELMs on the SOL density profile is characterized for a range of magnetic configurations. It is found that ELM-induced HFS SOL perturbations are generally localized to within the SS and ELM-induced changes to the HFS wall density are minimized in near double-null configurations. The characterization of the HFS SOL density profile on DIII-D is crucial for the effective coupling of the planned HFS lower hybrid current drive (LHCD) launcher. Experimental HFS SOL measurements are used as input to full-wave simulations of LHCD coupling. Furthermore, this work simulates the effect of the magnetic configuration on LHCD coupling. It also simulates LHCD coupling in ELMs and high q min discharges. During such discharges, the LHCD coupling is predicted to be resilient to ELM-induced SOL density perturbations.

high-field side↗

DOE OSTI2025

Editorial for the 28th Workshop on Magnetohydrodynamic (MHD) Stability Control

This special issue contains four papers which were presented at the 28th Workshop on Magnetohydrodymanic (MHD) Stability Control. The workshop was held at Auburn University and on Zoom from 3–5 October 2024. Furthermore, this annual workshop covers the active and passive control of MHD instabilities such as kink, tearing and edge-localized modes in tokamaks, reversed field pinches, spherical tori or other magnetic confinement devices.

Yang, J. [Princeton Plasma Physics Laboratory (PPP↗