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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

TorbeamNN: machine learning-based steering of ECH mirrors on KSTAR

We have developed TorbeamNN: a machine learning surrogate model for the TORBEAM ray tracing code to predict electron cyclotron heating (ECH) and current drive locations in tokamak plasmas. TorbeamNN provides more than a 100 times speed-up compared to the highly optimized and simplified real-time implementation of TORBEAM without any reduction in accuracy compared to the offline, full fidelity TORBEAM code. The model was trained using KSTAR ECH mirror geometries and works for both O-mode and X-mode absorption. The TorbeamNN predictions have been validated both offline and real-time in experiment. TorbeamNN has been utilized to track an ECH absorption vertical position target in dynamic KSTAR plasmas as well as under varying toroidal mirror angles and with a minimal average tracking error of 0.5 cm.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Measurements of radial neutral density profiles from Balmer- α emission in Wendelstein 7-X

Radial neutral density profiles are estimated from measurements of passive H α emission in the Wendelstein 7-X stellarator. To parametrize the generally three-dimensional distribution with a low number of degrees of freedom, the neutral density is reduced to a flux surface quantity. Accounting for emission from excitation and recombination processes, neutral density profiles are derived independently for each of the available lines of sight. Density profiles obtained from the different viewing geometries are found to vary within one order of magnitude. Toroidally oriented lines of sight predict systematically lower neutral densities when compared to poloidally oriented ones. This discrepancy is attributed to the simplifications inherent in the imposed model and significant differences in integration volumes across the viewing geometries. In line with expectations, obtained neutral densities are found to decrease with increasing plasma density. Key restrictions of the model include the reduction of the neutral density to a flux surface quantity, uncertainties in the plasma profiles and instrument function, and line integration effects outside the last closed flux surface.

Wendelstein 7-X↗
DOE OSTI2025

The new ITER baseline, research plan and open R&D issues

A new baseline (NB) has been proposed by the ITER Project to ensure a robust achievement of the Projects’ goals, in view of past challenges including delays incurred due to the Covid-19 pandemic, technical challenges in completing first-of-a-kind components and in nuclear licensing. The NB includes modifications to the configuration of the ITER device and its ancillaries (e.g. change from beryllium to tungsten as first wall material, modification of the heating and current drive mix, etc.) as well as additional testing of components (e.g. toroidal field coils) or phased installation (start with inertially cooled first wall before later installation of the final actively water-cooled components) to minimise operational risks. In the NB, the ITER research plan (IRP) will be divided into three main phases: (a) start of research operation, with 40 MW of ECH and 10 MW of ICH, which will focus on the demonstration of 15 MA operation in L-mode, commissioning of all required systems, including disruption mitigation, and the demonstration of H-mode plasma operation in deuterium; (b) DT-1, with 60–67 MW of ECH, 33 MW of neutral beam injection (NBI) and 10–20 MW of ICH, which will demonstrate robust operation in high confinement H-mode plasmas in DT up to Q ⩾ 10 and for burn durations of 300–500 s within an accumulated neutron fluence of ∼1% of the ITER machine’s lifetime total, and; (c) DT-2, with up to 67 MW of ECH, up to 49.5 MW of NBI and up to 20 MW of ICH, with the ITER tokamak and ancillaries in their final configuration to demonstrate routine operation in DT plasmas at high Q and the Q ⩾ 5 long-pulse and steady-state scenarios to the final neutron fluence and to perform R&D on nuclear fusion reactor issues. The logic, physics basis, modelling and experimental evaluations carried out to support the NB and the associated IRP are described. These include the impact of the tungsten wall on plasma scenarios and associated risk mitigation measures, as well as the optimisation of the tokamak components and ancillaries to minimise Project risks. Open R&D issues related to these evaluations and mitigation measures are also described together with experimental, modelling and validation activities required to address them.

ITER↗
DOE OSTI2025

Error field identification through torque balance on a saturated island in DIII-D

Measurement of the electromagnetic torque on a magnetic island could be an attractive method for error field identification in the early phase of ITER operation. Previous DIII-D experiments (Strait 2014 Nucl. Fusion 54 073004; Shiraki et al 2015 Plasma Phys. Control. Fusion 57 025016) have demonstrated the principle of this approach using a stationary or slowly rotating island, while recent developments in magnetic data analysis (Sweeney and Strait 2019 Phys. Plasmas 26 012509) allow the field of a rapidly rotating island to be readily distinguished from that of the wall currents induced by its rotation. In a recent experiment, a rotating n = 1 magnetic perturbation forced a saturated magnetic island to rotate, thus sampling all toroidal phases periodically in a single discharge. The phase and amplitude of the error field are inferred from analysis of the time-dependent torque balance on the island, including torques from the error field, the applied magnetic perturbation, and the wall currents induced by rotation of the applied perturbation and the island. Furthermore, results agree well with those from more conventional methods.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Divertor characterization and access to dissipative divertor conditions in negative triangularity discharges in DIII-D

Reduced radial transport, short midplane-to-target parallel connection lengths, and a strong effect of cross-field drifts were responsible for the high densities required for detachment in strong negative triangularity (NT) configurations in DIII-D. Dissipative divertor conditions were achieved in NT discharges at different triangularity, injected power, plasma current, and toroidal field direction. Differences between negative and positive triangularity (PT) discharges are analyzed in this paper to understand the requirements for access to detached divertor conditions: power balance, geometry, radial transport and effect of cross-field particle drifts. Parametric dependencies of access to detachment on plasma current and power flowing into the scrape-off layer remained similar in negative and PT and impurity seeding was observed to reduce the density needed to detach by up to 30% at the expense of core impurity dilution. The impact of triangularity on core-edge integration was tested varying bottom triangularity at fixed top triangularity. The high density needed to detach was not intrinsic to the NT edge as shapes with positive lower triangularity and negative upper triangularity were able to detach at lower upstream densities while maintaining an ELM-free NT edge. Confinement degradation at deeper detachment levels was however observed in all NT shapes, often associated with radiation instabilities.

detachment↗
DOE OSTI2025

A study of resistive peeling–ballooning modes across spherical tokamaks

We investigate how non-ideal-magnetohydrodynamics (MHD) effects, in particular plasma resistivity, impact the peeling–ballooning stability thresholds in spherical tokamaks. This analysis follows the discovery of resistive kink-peeling modes in ELMing National Spherical Torus Experiment (NSTX) discharges. In the present study we extend this modeling to ELMing pulses in the Mega Ampere Spherical Tokamak (MAST) and MAST—Upgrade (MAST-U), where we find a clear resistive scaling for peeling–ballooning modes. While in NSTX ideal-MHD predicts stability for ELMing discharges, in MAST-U we find that the plasma is slightly unstable to peeling–ballooning modes, but is fully stabilized once diamagnetic effects are considered in terms of a growth rate normalization. A resistive power law scaling is calculated for these modes on MAST-U, which lies in between that of tearing modes and resistive interchange modes. A comparison between M3D-C1 and NIMROD shows reasonable agreement for this scaling. Resistivity destabilizes the modes and the peeling–ballooning unstable domain is considerably expanded in both, MAST and MAST-U. In addition to the MAST/-U pulses we also analyze resistive PB stability in a NSTX-similarity discharge on DIII-D. While having a different aspect ratio from NSTX, this discharge uses NSTX-like shaping parameters, toroidal field and plasma current. By considering these discharges alongside NSTX cases, we identify conditions influencing the onset of resistive peeling–ballooning modes. Furthermore, our findings indicate that magnetic shear in the pedestal region is closely linked to the emergence of resistive edge modes.

MAST-U↗
DOE OSTI2025

Effects of equilibrium pressure on plasma response to RMPs in a spherical tokamak

This study presents a comprehensive analysis of the equilibrium pressure on the plasma response to resonant magnetic perturbations (RMPs) in the spherical tokamak (ST) MAST-U, employing both single-fluid and MHD-kinetic hybrid models (implemented via the MARS-F/K codes). As a key finding, the study identifies two different pressure-driven eigenmodes, exhibiting Sturmian property, that affect the Troyon no-wall limits for the onset of the n = 1 and n = 2 ( n is the toroidal mode number) ideal external kink instabilities as well as the corresponding plasma response to the applied RMP. With increasing equilibrium pressure, the plasma response to RMPs is significantly enhanced in the ST plasma, particularly in the high-pressure regime where kinetic effects strongly stabilize the external kink instability. The Troyon no-wall limit divides the plasma response into two regions: well below the limit, the response amplitudes and trends (versus pressure) are similar between the fluid and kinetic models; as the equilibrium pressure approaches the Troyon limit, the kinetic model predicts significant amplification of the RMP field, up to 30 times for cases considered. A relatively weak dependence of the optimal coil phasing on the equilibrium pressure is computed in this ST plasma, similar to the trend obtained for the conventional aspect ratio devices. These findings underscore the importance of incorporating kinetic effects in accurate prediction of the plasma response to RMPs in high-pressure ST tokamak plasmas and provide a theoretical basis for optimizing RMP-based control of the edge-localized modes in future ST devices.

RMP↗
DOE OSTI2025

Effects beyond ideal MHD on stability of wide and enhanced pedestal regimes in NSTX

Stability of edge-localized modes (ELMs) in spherical tokamaks is explored using the extended MHD model. Linear NIMROD simulations have been performed for three NSTX discharges 132543, 132588, and 141133, to investigate the role of resistivity, diffusivity, and shear flows on the onset of ELMs. The first discharge represents the wide pedestal regime and the later two discharges represent ELM-free enhanced pedestal H-mode. We first present the effect of toroidal rotation shear and find a flow shear destabilizing effect in these NSTX discharges. Simulations are also extended to include the two-fluid and ion gyroviscosity effects. Simulations show that the flow shear can shift the mode spectrum and alter the critical condition of ELM onset. We also uncover that ELM onset prediction in spherical tokamaks requires effects beyond MHD, in particular gyroviscosity and diamagnetic terms could stabilize Peeling-Ballooning modes consistent with the experimental observation of ELM-free regimes in NSTX. The findings give new insight into the nature of the interplay between resistivity, flows, and diamagnetic stabilization in ELM suppression and have potential applications to ELM control schemes in NSTX-U and next-generation spherical tokamaks. This study identifies the essential physical effects that must be included in future predictive and validation simulations.

NSTX↗
DOE OSTI2025

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↗
DOE OSTI2025

Error field measurements and correction on MUSE permanent magnet stellarator

The magnetic field topology and error fields are measured in the MUSE permanent magnet stellarator with an electron beam and fluorescent rod. The effective magnetization of the magnets is determined by measuring rational flux surface locations. Electron magnetic drift effects must be considered self-consistently due to the small magnetic shear on MUSE. Error fields are characterized by flux surface shape deviation and magnetic island chain width and phase. Error fields are then corrected by adjusting the permanent magnet (PM) holders as rigid bodies. Resonant error fields are on the order of or smaller than 3 x 10 -6 of the toroidal field after correction and surface shape matches well with the design. Resonant components of the quasi-symmetry deviation are estimated, using a novel magnetic drift measurement, to be in the 10 −4 range. The PM approach is validated as a low-cost and precise way to build optimized stellarators.

electron beam mapping↗
DOE OSTI2025

Omnigenous stellarators with improved ideal and kinetic ballooning stability

Omnigenity is a property of a magnetic field which ensures confinement of trapped particles. It is a necessary requirement for any high-performance stellarator. After creating an omnigenous equilibrium, one must also ensure reduced transport resulting from kinetic and magnetohydrodynamic (MHD) instabilities. To this end, we leverage the GPU-accelerated DESC optimization suite, which is used to design stable, finite-β omnigenous equilibria with poloidal, toroidal, and helical symmetry, achieving Mercier, ideal ballooning, and as a consequence, improved kinetic ballooning stability. We discover stellarators with second stability, a regime of large pressure gradient where an equilibrium becomes ideal ballooning stable, and demonstrate and explore both using theory and gyrokinetic simulations the connection between ideal and kinetic ballooning stability.

optimization↗
DOE OSTI2026

Analysis and modeling of tungsten emission and net erosion in the DIII-D divertor using updated atomic data

Tungsten (W) is one of the leading candidate materials for plasma-facing components. However, its main drawback is its high radiative efficiency; if W penetrates the plasma, it can lead to core degradation or even collapse. Since eroded tungsten tends to ionize in the sheath and redeposit promptly, the net erosion flux that escapes prompt redeposition can differ significantly from the gross erosion. This work presents a modeling framework to estimate net erosion and photon emission from W coatings exposed to the lower divertor of DIII-D using the DiMES material exposure probe. The approach couples RustBCA for sputtering yields with a Monte Carlo transport code (LPTMC) that models redeposition and W emission. Computation is carried out with new atomic data, based on R-matrix and Mons calculations, leading to lower ionization probabilities and a twofold increase in net erosion estimates compared to calculations done with OPEN-ADAS atomic data. The model results are benchmarked against experimental measurements, showing quantitative agreement for erosion, although the trends in W emission are reproduced only qualitatively. The model is also used to assess whether W II emission can serve as a direct measurement of the net erosion of W in the lower divertor of DIII-D. Simulations show that this is not valid if the electron pressure is above ~120 Pa or if the toroidal length of the eroded material is smaller than the parallel-to-B distance traveled by impurity ions before steady-state conditions are reached. Finally, simulations suggest that when W is sputtered by carbon ions with high impact energies (≳300 eV) in DIII-D, W net erosion scales with W gross erosion and can be numerically approximated using W I flux alone as input.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

MHD-induced SOL filaments and divertor heat flux striations in NSTX

Magnetohydrodynamic (MHD) activity in NSTX has been found to induce striated heat flux patterns at the divertor. These striated heat fluxes were previously interpreted as strike-point splitting caused by stochastic magnetic lobes near the X-point. Recent analysis indicates that MHD can directly generate filaments in the scrape-off layer (SOL), and the MHD-related filaments are consistent with the rotation and the frequency of the MHD. Magnetic field line tracing from the mid-plane of the MHD-related filaments to the divertor is consistent with the location of striated heat fluxes on the divertor. Additionally, gas puff imaging (GPI) diagnostic data shows the generation of regular blobs following each edge harmonic oscillation (EHO) toroidal mode number n = 1 perturbation, which indicates that the MHD perturbation directly drives the blobs. These blobs are responsible for the divertor heat flux striations. Here, this finding provides a new framework for understanding MHD-induced edge transport and has important implications for heat-load control in NSTX-U and future spherical tokamaks.

MHD↗
DOE OSTI2026

Predicting core transport in ITER baseline discharges with neon injections

Achieving self-consistent performance predictions for ITER requires integrated modeling of core transport and divertor power exhaust under realistic impurity conditions. We present results from a systematic power-flow and impurity-content study for the ITER 15 MA baseline scenario constrained directly by existing SOLPS-ITER neon-seeded divertor solutions. Using the OMFIT STEP workflow, stationary temperature and density profiles are predicted with TGYRO for $1.5 \unicode{x2A7D} Z_\textrm{eff} \unicode{x2A7D} 2.5$, and the corresponding power crossing the separatrix $P_\textrm{sep}$ is evaluated. We find that $P_\textrm{sep}$ varies by more than a factor of 1.7 across this scan and matches the ${\sim}100$ MW SOLPS-ITER prediction when $Z_\textrm{eff} \simeq 1.6$ or when auxiliary heating is reduced to ${\sim}75\%$ of nominal. Rotation-sensitivity studies show that plausible variations in toroidal flow magnitude modify $P_\textrm{sep}$ by $\lesssim 20\%$, while AURORA modeling confirms that charge-exchange radiation inside the separatrix is dynamically negligible under predicted ITER neutral densities. These results identify a restricted compatibility window, $Z_\textrm{eff} \approx 1.6$ –1.75 and $0.75 \lesssim f_{P_\textrm{aux}} \unicode{x2A7D} 1.0$, in which core transport predictions remain aligned with neon-seeded divertor protection targets. This self-consistent, model-constrained framework provides actionable guidance for impurity control and auxiliary-heating scheduling in early ITER operation and supports future whole-device scenario optimization.

ITER↗
DOE OSTI2026

Multimachine analysis of intermittent fluctuations in the scrape-off layer of magnetically confined fusion plasmas

Far scrape-off layer fluctuations, also known as blobs/filaments, are compared for the first time across six tokamaks: Alcator C-Mod, DIII-D, TCV, KSTAR, MAST, and MAST-U, under the context of stochastic modeling in Ohmic and low confinement mode scenarios. In this model, which well describes these fluctuations, measurements from Langmuir probes are modeled as a superposition of uncorrelated, two-sided exponential temporal pulses. The key parameters of this model are the mean waiting times between consecutive pulses, the pulse duration times and the mean of the pulse amplitudes. The study examines the influence of line-averaged density, plasma current, magnetic field and aspect ratio. The mean amplitudes and pulse waiting times increase with the line-averaged density across devices, while the pulse duration times remain constant. Whereas the variation in waiting times is gradual across the density scan, the mean amplitudes increase abruptly as the density profile broadens and flattens for Alcator C-Mod. The blob waiting times decrease with increasing plasma current and the toroidal magnetic field, while they increase with aspect ratio for conventional tokamaks. The duration times follow the same trends as the waiting times in these parameters but vary more slowly, resulting in the trends in the waiting times dominating the intermittency level of the fluctuations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Access and limits of RMP ELM suppression with n = 1 fields in DIII-D

This work reports on DIII-D experiments aimed at extending resonant magnetic perturbation (RMP) suppression of edge localized modes (ELMs) to n = 1 fields, where n is the toroidal mode number. Modeling of the 3D ideal MHD plasma response to the RMPs using the GPEC code is used to quantify edge and core resonant fluxes, guiding experimental strategies to increase plasma resilience against core error field penetration, optimize multicoil phasing, and explore higher q 95 operation. In DIII-D, ELM mitigation is regularly observed across a wide range of n = 1 RMP scenarios. A ∼100 ms phase of complete ELM suppression was achieved at q 95 ∼ 3.9 using an odd-parity coil configuration. The suppressed phase exhibited clear signatures of RMP ELM suppression, including the elimination of Dα spikes, increased pedestal rotation, enhanced magnetic response, and elevated broadband density turbulence. An optimized coil configuration for edge-to-core resonant flux did show increased edge resonance indicated by increased density pumpout, but did not yield RMP ELM suppression. At q 95 ∼ 5.1, a bifurcation to a grassy-like ELM regime occurred, while large type-I ELMs persisted. These results demonstrate progress in experimental access to n = 1 RMP ELM suppression in DIII-D, motivating further study for robust access. This work also highlights the potential role of 3D edge stability as well as rational surface alignment in RMP ELM suppression access, which has important implications for the use of low-n RMPs in future reactor-scale devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Visible spectroscopy of highly charged tungsten ions with the JET charge exchange diagnostic

Since the installation of the ITER-like wall on the JET tokamak, visible emission from a broad range of tungsten charge states has complicated plasma ion temperature and toroidal rotation measurements. A plethora of charge exchange transitions, from ions up to W 56+ , and 21 suspected magnetic dipole transitions have been observed. In particular W 39+ and W 46+ charge exchange emission contaminates the long-established carbon measurement at 529 nm. Fortunately the wavelength and relative intensity of the tungsten charge exchange lines is predictable and their influence can be mitigated when they are included in the spectral fit. Neon based charge exchange measurements at 525 nm are now preferred on JET since the ITER-like wall, however in this case an unidentified tungsten magnetic dipole transition contaminates the spectrum when the electron temperature is low.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗