Search NASA⌕ Search

SEARCH · Search NASA

Results for “plasma shaping”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Impact of shape on pedestal characteristics in the wide pedestal quiescent H-mode in the DIII-D tokamak

Experiments on the DIII-D tokamak have advanced the operational limits of wide pedestal quiescent H-mode (WPQH) plasmas towards increased ITER relevance by simultaneously demonstrating well-matched plasma shape and net zero injected torque. Wide pedestal QHmodes are a compelling candidate regime for a future power producing device because they maintain a stationary pedestal without ELMs via additional edge transport. The pedestal is wider than what would be predicted from kinetic ballooning mode physics due to enhanced edge transport generated by broadband turbulence, a limit cycle oscillation, or some combination thereof. Here, compared to the double null shape, the lower single null shape is observed to have a lower density, narrower pedestal width, larger density fluctuations over a broad range of wavenumber, and operates closer to the peeling-ballooning instability boundary calculated from the simple pedestal scaling, Δ ψΝ , which is still observed to be wider than the EPED prediction.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Future of plasma etching for microelectronics: Challenges and opportunities

Plasma etching is an essential semiconductor manufacturing technology required to enable the current microelectronics industry. Along with lithographic patterning, thin-film formation methods, and others, plasma etching has dynamically evolved to meet the exponentially growing demands of the microelectronics industry that enables modern society. At this time, plasma etching faces a period of unprecedented changes owing to numerous factors, including aggressive transition to three-dimensional (3D) device architectures, process precision approaching atomic-scale critical dimensions, introduction of new materials, fundamental silicon device limits, and parallel evolution of post-CMOS approaches. The vast growth of the microelectronics industry has emphasized its role in addressing major societal challenges, including questions on the sustainability of the associated energy use, semiconductor manufacturing related emissions of greenhouse gases, and others. The goal of this article is to help both define the challenges for plasma etching and point out effective plasma etching technology options that may play essential roles in defining microelectronics manufacturing in the future. The challenges are accompanied by significant new opportunities, including integrating experiments with various computational approaches such as machine learning/artificial intelligence and progress in computational approaches, including the realization of digital twins of physical etch chambers through hybrid/coupled models. These prospects can enable innovative solutions to problems that were not available during the past 50 years of plasma etch development in the microelectronics industry. To elaborate on these perspectives, the present article brings together the views of various experts on the different topics that will shape plasma etching for microelectronics manufacturing of the future.

Engineering↗
DOE OSTI2021

X-Ray Spectroscopic Studies of X-Pinch Plasmas with 3-5 Picosecond Resolution: A Quest for Clear Experimental Evidence for Radiative Collapse in the X-ray Spectra (Final Report)

Dense Z-pinch plasmas produced from current-carrying exploding wires frequently produce very brief (less than 0.1 nanosecond), intense, bursts of soft X-rays from tiny (about 1 micrometer), very hot (10-30 million K) regions of the plasma. This project addresses the physical processes that lead to those tiny, high-energy density plasmas and develops means to measure the conditions just before and during the X-ray bursts. One possible relevant process is radiative collapse, which starts if the power radiated by a dense Z-pinch, proportional to the density-squared, exceeds the resistive heating rate from the current flow, thereby cooling the plasma and enabling the magnetic pinching force to exceed the outward plasma thermal pressure. As the resulting radius reduction further increases the radiation rate, this process produces a runaway magnetic implosion. A competing mechanism is a sausage-like instability that can develop in a Z-pinch plasma at a lower current than is needed to induce radiative collapse. This mechanism can also produce tiny hot spots in the neck regions of the sausage-shaped plasma column. It is also possible that these two mechanisms work together to produce the X-ray emitting hot spots. If radiative collapse does contribute to hot spot formation, it would be terminated when the plasma becomes so dense that the radiation cannot escape, enabling the plasma kinetic pressure to build up and stop the current-driven implosion. This project aims to study the development of the hot spots by means of high temporal and spatial resolution X-ray spectroscopy to determine if there is evidence for the presence of the radiative collapse mechanism. Progress toward this goal is presented. To summarize, during the four-year period 09/15/2017 - 9/14/2021, we have collected necessary spectroscopic data to determine the plasma conditions and continue to do so. We are also now in the process of calibrating diagnostic systems (especially the x-ray streak camera system) and analyzing data to determine plasma conditions as a function of time before, during and after the X-ray burst.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1986

The structure of a cometary type I tail - Ground-based and ICE observations of P/Giacobini-Zinner

Comparison of ground-based and in situ observations of P/Giacobini-Zinner are used to investigate the morphology of a type I cometary tail. ICE magnetic field and plasma measurements show a well-defined cometary magnetotail composed of two magnetic lobes in pressure equilibrium with a central plasma sheet. A dependence of ion tail width on IMF direction is found which strongly suggests that the classical type I ion tails observed on the ground consist predominantly of emissions from the slab-shaped plasma sheet separating the magnetic lobes. The width of the G-Z magnetotail is determined to be 9.8 (+ or - 0.5) x 10 to the 3rd km with a quasi-circular cross section. The results of this study also indicate that some of the dynamical thinnings and thickenings observed in long type I tails may be caused by IMF variations changing the angle with which the plasma sheet is viewed at earth.

Slavin, J. A.↗
DOE OSTI2026

Demonstration of tokamak vertical stability control based on non-inductive Faraday-effect polarimetry measurements

Long-pulse or steady-state fusion reactors are envisioned to control vertical stability based on non-inductive measurements, i.e. that do not rely on temporal change of magnetic field. For the first time, vertical stability control using non-inductive Faraday-effect polarimetry measurements has been demonstrated. The Radial Interferometer-Polarimeter system on DIII-D is capable of microsecond resolution and was used to absolutely determine the vertical position of the plasma magnetic axis Z 0 . A vertical stability controller was developed to robustly stabilize diverted plasmas using Faraday-based measurements. The system was able to stabilize against vertical displacement events with growth rates up to 350 s -1 in elongated and elliptical plasma shapes, and instabilities with even higher growth rates are likely to be controllable with further improvements to controller tuning. Tests show that the Faraday-based controller remains effective and is capable of recovering from loss of control even when the plasma vertical position is far from the region where the linear model used to calculate Z 0 is most valid. Faraday control has also been activated during plasma ramp-up, demonstrating the robustness of the technique to larger systematic diagnostic uncertainty at low electron density.

Faraday-effect polarimetry↗
DOE OSTI2023

Development of high-performance long-pulse discharge in KSTAR

High-performance long-pulse plasma operation is essential for producing economically viable fusion energy in tokamak devices. To achieve such discharges in KSTAR, firstly, the rapid increase in the temperature of plasma-facing components was mitigated. The temperature increase of the poloidal limiter, especially, was associated with beam-driven fast ion orbit loss and the discrepancy of the equilibrium reconstructed with heated magnetic probes of signal drift. The fast ions lost to the poloidal limiter were reduced by optimizing the plasma shape and the composition of neutral beam injection (NBI). This nonlinear signal drift was successfully reduced by a new thermal shielding protector on the magnetic probes. Secondly, a lower loop voltage approach was implemented to reduce a poloidal flux consumption rate. A plasma current of 400 kA and a line-averaged electron density of ~2.0 × 10 19 m –3 were chosen by considering the L–H power threshold, fast ion orbit loss, and beam shine-through power loss for low loop voltage in KSTAR. In addition, the application of electron cyclotron heating also helped maintain the plasma with low loop voltage (~25 mV) by enhancing the NBI-driven current and achieving a high poloidal beta (β P ) state. KSTAR has achieved a long pulse (~90 s) operation with the high performance of β P $\leqslant$ 2.7, thermal energy confinement enhancement factor (H 98y2 ) ~ 1.1, and fraction of non-inductive current (f NI ) $\leqslant$ 0.96. Still, gradual degradation of the plasma performance has been observed over time in the discharges. In one of the long-pulse discharges, β P reduced by ~18% over the time of ~8τ R (current relaxation time, τ R ~5 s) and ~1067τ E,th (thermal energy confinement time, τ E,th ~45 ms). The degradation may be closely associated with weak, yet growing, and persistent toroidal Alfvén eigenmodes and their effect on fast ion confinement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Effect of negative triangularity on tearing mode stability in tokamak plasmas

Abstract The influence of negative triangularity (NT) of the plasma shape on the n = 1 (n is the toroidal mode number) tearing mode (TM) stability has been numerically investigated, with results compared to that of the positive triangularity (PT) counterpart. By matching the safety factor profile for a series of toroidal equilibria, several important plasma parameters, including the triangularity, the plasma equilibrium pressure, the plasma resistivity as well as the toroidal rotation, have been varied. The TM localized near the plasma edge is found to be more unstable in the NT plasmas as compared to the PT counterpart. The fundamental reason for this difference is the lack of favorable average curvature stabilization in NT configurations. Direct comparison of the Mercier index corroborates this conclusion. For the core-localized mode, where the difference in the local triangularity between NT and PT becomes small and the curvature stabilization is significantly reduced, larger Shafranov shift in the plasma core associated with the NT configuration results in more stable TM. The plasma toroidal flow generally stabilizes the TM in plasmas with both NTs and PTs. The flow stabilization is however weaker in the case of negative triangularity with finite plasma pressure.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

MMMnet: A Neural Network Surrogate for Real-Time Transport Prediction Based on the Updated Multi-Mode Model

The Multi-Mode Model (MMM) is a physics-based anomalous transport model integrated into TRANSP for predicting electron and ion thermal transport, electron and impurity particle transport, and toroidal and poloidal momentum transport. While MMM provides valuable predictive capabilities, its computational cost, although manageable for standard simulations, is too high for real-time control applications. MMMnet, a neural network-based surrogate model, is developed to address this challenge by significantly reducing computation time while maintaining high accuracy. Trained on TRANSP simulations of DIII-D discharges, MMMnet incorporates an updated version of MMM (9.0.10) with enhanced physics, including isotopic effects, plasma shaping via effective magnetic shear, unified correlation lengths for ion-scale modes, and a new physics-based model for the electromagnetic electron temperature gradient mode. A key advancement is MMMnet’s ability to predict all six transport coefficients, providing a comprehensive representation of plasma transport dynamics. MMMnet achieves a two-order-of-magnitude speed improvement while maintaining strong correlation with MMM diffusivities, making it well-suited for real-time tokamak control and scenario optimization.

DIII-D↗
DOE OSTI2024

Error field predictability and consequences for ITER

Abstract ITER coil tolerances are re-evaluated using the modern understanding of coupling to least-stable plasma modes and an updated center-line-traced model of ITER’s coil windings. This reassessment finds the tolerances to be conservative through a statistical, linear study of n = 1 error fields (EFs) due to tilted, shifted misplacements and nominal windings of central solenoid and poloidal field coils within tolerance. We also show that a model-based correction scheme remains effective even when metrology quality is sub-optimal, and compare this to projected empirical correction schemes. We begin with an analysis of the necessity of error field correction (EFC) for daily operation in ITER using scalign laws for the EF penetration threshold. We then consider the predictability of EF dominant mode overlap across early planned ITER scenarios and, as measuring EFs in high power scenarios can pose risks to the device, the potential for extrapolation to the ITER Baseline Scenario (IBS). We find that carefully designing a scenario matching currents proportionally to those of the IBS is far more important than plasma shape or profiles in accurately measuring an optimal correction current set.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Magnetic-Confinement Fusion—Plasma Theory: Tokamak Magnetohydrodynamic Equilibrium and Stability

Magnetohydrodynamics (MHD) provides a useful model to describe the crucial plasma macroscopic equilibrium and stability behaviors in toroidal tokamak devices by considering the plasma as a conducting fluid interacting with a surrounding confining electromagnetic field. MHD is the most basic plasma model, incorporating most large-scale phenomena, including plasma equilibrium and all major instabilities. MHD equations are obtained by taking moments of the Boltzmann equations for different plasma species. They provide a set of comprehensive physics constrains to compute and optimize the equilibrium plasma shape and pressure and current profiles that are critical to its stability and performance. In the ideal case, the equations have special properties that lead to efficient numerical calculation schemes, the most important of which is the ideal MHD energy principle for linear stability against small departures from equilibrium. In a tokamak plasma, equilibrium pressure is mostly destabilizing for MHD modes, whereas equilibrium current is also often a major driving force. Plasma resistivity creates new freedom for a MHD instability to grow, but there are also cases where the plasma resistivity plays a stabilizing role. Equilibrium toroidal flow and/or flow shear can affect MHD instabilities. Principal MHD instabilities include the internal kink mode, sawtooth, fishbone, external kink, resistive wall mode, resistive interchange, tearing and neoclassical tearing modes (NTMs), locked modes, toroidal Alfven eigenmodes (TAEs), and edge localized modes (ELMs). Fast-growing MHD instabilities can lead to an abrupt plasma disruption and termination that can potentially damage the device plasma facing components (PFCs) and in-vessel structures. Furthermore, an important MHD application is to develop robust techniques to mitigate and control MHD instabilities.

Dispersive pellet injection↗
DOE OSTI2024

Enhancing predictive capabilities in fusion burning plasmas through surrogate-based optimization in core transport solvers

Abstract This work presents the PORTALS framework (Rodriguez-Fernandez et al 2022 Nucl. Fusion 62 076036), which leverages surrogate modeling and optimization techniques to enable the prediction of core plasma profiles and performance with nonlinear gyrokinetic simulations at significantly reduced cost, with no loss of accuracy. The efficiency of PORTALS is benchmarked against standard methods, and its full potential is demonstrated on a unique, simultaneous 5-channel (electron temperature, ion temperature, electron density, impurity density and angular rotation) prediction of steady-state profiles in a DIII-D ITER Similar Shape plasma with GPU-accelerated, nonlinear CGYRO (Candy et al 2016 J. Comput. Phys. 324 73–93). This paper also provides general guidelines for accurate performance predictions in burning plasmas and the impact of transport modeling in fusion pilot plants studies.

Physics↗
DOE OSTI2024

Observation of a new pedestal stability regime in MAST Upgrade H-mode plasmas

Abstract The first pedestal stability and structure analysis on the new MAST Upgrade (MAST-U) spherical tokamak H-mode plasmas is presented. Our results indicate that MAST-U pedestals are close to the low toroidal mode number ( n ) peeling branch of the peeling-ballooning instability, in contrast with MAST H-mode pedestals which were deeply in the high- n ballooning branch. This offers the possibility of reaching the ELM-free quiescent H-mode (Burrell et al 2005 Plasma Phys. Control. Fusion 47 B37–B52) or high-performance super H-mode (Snyder et al 2015 Nucl. Fusion 55 083026; Snyder et al 2019 Nucl. Fusion 59 086017) regimes. In addition, the coupling between the peeling and ballooning branches is weak in MAST-U, suggesting that a path to very high pedestal pressure gradient at high density may exist with sufficient heating power. A possible explanation for the differences between MAST and MAST-U pedestal stability is given in terms of plasma shaping parameters, in particular squareness and elongation, as well as the pedestal top temperature and collisionality.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Piecewise Omnigenous Stellarators

In omnigenous magnetic fields, charged particles are perfectly confined in the absence of collisions and turbulence. For this reason, the magnetic configuration is optimized to be close to omnigenity in any candidate for a stellarator fusion reactor. However, approaching omnigenity imposes severe constraints on the spatial variation of the magnetic field. In particular, the topology of the contours of constant magnetic field strength on each magnetic surface must be such that there are no particles transitioning between different types of wells. This, in turn, usually leads to complicated plasma shapes and coils. This Letter presents a new family of optimized fields that display tokamak-like collisional energy transport while having transitioning particles. This result radically broadens the space of accessible reactor-relevant configurations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Ultra-short pulse generation from mid-IR to THz range using plasma wakes and relativistic ionization fronts

In this work, we discuss numerical and experimental results on frequency downshifting and upshifting of a 10 μ m infrared (IR) laser to cover the entire wavelength (frequency) range from λ = 1 to 150 μ m ( ν = 300–2 THz) using two different plasma techniques. The first plasma technique utilizes frequency downshifting of the drive laser pulse in a nonlinear plasma wake. Based on this technique, we have proposed and demonstrated that in a tailored plasma structure, multi-millijoule energy, single-cycle, long-wavelength IR (3–20 μ m) pulses can be generated by using an 810 nm Ti:sapphire drive laser. Here, we extend this idea to the THz frequency regime. We show that sub-joule, terawatts, single-cycle terahertz (2–12 THz or 150–25 μ m) pulses can be generated by replacing the drive laser with a picosecond 10 μ m CO 2 laser and a different shaped plasma structure. The second plasma technique employs frequency upshifting by colliding a CO 2 laser with a rather sharp relativistic ionization front created by ionization of a gas in less than half cycle (17 fs) of the CO 2 laser. Even though the electrons in the ionization front carry no energy, the frequency of the CO 2 laser can be upshifted due to the relativistic Doppler effect as the CO 2 laser pulse enters the front. The wavelength can be tuned from 1 to 10 μ m by simply changing the electron density of the front. While the upshifted light with 5 < λ ( μ m ) < 10 propagates in the forward direction, that with 1 < λ ( μ m ) < 5 is back-reflected. These two plasma techniques seem extremely promising for covering the entire molecular fingerprint region.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Upgrade of the Lyman-alpha diagnostic system on DIII-D for main chamber edge neutral studies

The LLAMA (Lyman Alpha Measurement Apparatus) pinhole camera diagnostic had previously been deployed on DIII-D to measure radial profiles of the Lyman-α (Ly-α) deuterium neutral line brightness across the plasma boundary in the lower chamber to infer neutral deuterium density and ionization rate profiles. This system has recently been upgraded with a new diagnostic head, named ALPACA, that also encloses two pinhole cameras and duplicates the LLAMA views in the upper chamber. Similar to LLAMA, ALPACA provides two times 20 lines of sight, viewing the plasma edge on the inboard and outboard sides with a radial resolution of ~2.5 cm (FWHM) and an effective time resolution of ~1 ms that allows for the investigation of inter-ELM dynamics. The extended Ly-α system provides better coverage to study neutrals in experiments with various plasma shapes utilizing both the upper and lower divertors. Furthermore, post-campaign calibration of the LLAMA diagnostic has successfully been demonstrated for the first time. This was facilitated by various upgrades to the calibration set-up and detailed measurements of the emissivity distribution of the Ly-α calibration source using a pinhole collimator. It was found that the sensitivity of the inboard LLAMA pinhole camera was reduced by a factor of 2.0 ± 0.2 over the course of six months of plasma operation in 2021. In conclusion, the upgraded Ly-α system, equipped with improved absolute calibration, will provide key input for neutral fueling and pedestal particle transport studies and for 2D edge transport code validation on the DIII-D tokamak.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Bus Design for the Poloidal Field Coils of the NSTX-Upgrade Fusion Device

The National Spherical Torus eXperiment (NSTX) has undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). NSTX upgrade (NSTX-U) will double the toroidal field, plasma current, and neutral beam injection heating power, as well as significantly increase the pulse duration. NSTX-U uses three poloidal field (PF) coils at the vessel top and three at the bottom near the divertor areas to control the local plasma shape there. These shaping coils operate at a maximum terminal voltage of 2 kV, corresponding to a maximum current about 20 kA. To supply the power to the PF coils, hard copper bus bars are typically used near the vacuum vessel, while water cooled flexible cables are used away from the vessel. The detailed design of the hard copper bus bars will be covered in this article. During operation, these hard bus bars are subject to high electromagnetic (EM) forces, thermal displacement loads, as well as plasma halo loads due to disruption. The EM, thermal, and structural analysis were performed, and the results revealed that, with the worst loads combined, the new design meets the NSTX-U thermal, structural, and fatigue cycle requirements. Finally, the manufacturing and installation process of the bus work will be discussed at the end of this article.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Plasma performance and operational space with an RMP-ELM suppressed edge

Abstract The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over 3 × 10 19 m −3 . This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Scenario optimization for the tokamak ramp-down phase in RAPTOR: Part A. Analysis and model validation on ASDEX Upgrade

We discuss how the combination of experimental observations and rapid modeling has enabled to improve understanding of the tokamak ramp-down phase in ASDEX Upgrade. A series of dedicated experiments has been performed, to disentangle the effect of individual actuators like plasma current, auxiliary heating and plasma shaping. Optimized discharge termination strategies with increased margin with respect to radiative and vertical stability limits are proposed and tested in experiment. Radiative collapse of the edge $T$ $e$ profile after the HL back-transition is avoided by initially maintaining auxiliary heating during L-mode, showing beneficial effects even after the auxiliary heating is turned off. The capability of the RAPTOR code to model the time evolution of the internal inductance $\ell$ $i3$ has been validated, including the effect of a change in the $I$ $p$ ramp-down rate and the HL transition timing. The reduction of $\ell$ $i3$ caused by rapid compression of the plasma cross-section has been quantitatively recovered in simulations. Successful modeling of the $\ell$ $i3$ time evolution is essential to optimize ramp-down scenarios for future fusion reactors, for which vertical stability and power balance control will be more challenging.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗