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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Numerical minimization of neoclassical poloidal viscosity for supersonic equilibria in tokamak geometry

Extensive experimental evidence has shown that the presence of poloidal flow in tokamaks can dramatically improve transport properties. However, theory indicates that poloidal flows are damped by poloidal viscosity, thus necessitating external drivers, such as neutral beam injection or radio frequency heating. In this work, ideal magnetohydrodynamic equilibria are calculated via the FORTRAN code FLOW [Guazzotto et al., Phys. Plasmas 11, 604 (2004)] and a postprocessor is used to estimate the neoclassical poloidal viscosity. The equilibrium inputs, which correspond to intuitive physical quantities, are then numerically optimized to reduce a viscosity figure of merit. We present supersonic equilibria in tokamak geometry with minimized neoclassical poloidal viscosities for various velocity free function inputs, plasma aspect ratios, and collisionality regimes. Benchmarks are made against an analytic theory as well as a classical expression of poloidal viscosity. Numerical confirmation of the analytic theory is obtained in the high aspect ratio and high collisionality limit. Good agreement is also seen near the plasma core and edge, with discrepancies arising in the intermediate region. Outside of these limits, rotation input function profiles are found that provide ∼order of magnitude improvements over the analytic theory, with additional progress being made toward predictions for tokamak-relevant equilibria.

Physics↗
DOE OSTI2021

A laser inverse Compton scattering diagnostic to study runaway electron dynamics during tokamak disruptions

Both technical and simulation progress has been made toward implementing a laser inverse Compton scattering diagnostic on the DIII-D tokamak. This is a novel diagnostic which uses a short (sub-ns) pulse laser together with a fast (sub-ns) gated x-ray detector to measure relativistic electrons inside a tokamak plasma. The study of these relativistic electrons is important to make sure they can be removed from any tokamak-based power reactor, as they could damage vessel walls and cooling lines. This diagnostic is completely new for magnetic fusion research and has never been implemented before on a tokamak. It is found that good signals and radial profiles could be obtained with an optimized high-energy (7 Joule) laser. Using a standard off-the-shelf Nd:YAG (0.7 Joule) laser would give some signals above the noise but no ability to measure the relativistic electron radial profile.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

An ideal MHD δW stability analysis that bypasses the Newcomb equation

In the following work, we demonstrate the efficacy of a Riccati δW ideal MHD stability analysis that bypasses the numerically intractable integration of the Newcomb equation. By transforming the linear Newcomb equation into a quadratic Riccati equation, an accurate and equivalent δW analysis is performed that is shown to enjoy some numerical advantages. Additionally, we demonstrate that the Riccati approach is better conditioned than its Newcomb counterpart at the magnetic axis, and we apply dynamical systems insights to examine its behavior at singular points. We further discuss the constraints involved in solving for the extremal admissible perturbations of a δW analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Results from a new flowing liquid Li limiter with TZM substrate during high confinement plasmas in the EAST device

A third generation flowing liquid lithium (FLiLi) limiter with a substrate made of TZM, an alloy with >99% Mo, was fabricated by conventional manufacturing techniques. TZM has a high corrosion resistance, a high sputtering threshold, and a good wettability to Li, as compared to stainless steel (SS), which had been used as an FLiLi substrate surface in 2014 and 2016. The third generation FLiLi was inserted into the edge in EAST H-mode plasmas in an upper single-null configuration with an ion grad-B drift toward the upper divertor with a limiter temperature of 330–380 °C and an auxiliary heating power of about 2–8 MW. Analysis has shown that by using TZM FLiLi, fuel particle recycling continuously decreased and near-complete edge localized mode elimination was achieved in H-mode plasmas with RF-only heating. The main impurities during the initial FLiLi discharges were Mo, Fe, and W resulting from strong plasma interaction at the Mo plate side, the SS collector, and the upper W divertor. Plasma stored energy increased by about 10 kJ in subsequent FLiLi discharges due to decreased impurity radiation. Engineering analysis shows a uniform lithium flow with an ~80% Li coverage ratio on the limiter surface, similar to the second FLiLi, even though only one of two J × B pumps was functioning. Despite technical difficulties, the FLiLi gen.3 improved the overall plasma performance, providing support for flowing liquid Li plasma facing component applications in present and future devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Solving the problem of overdetermination of quasisymmetric equilibrium solutions by near-axis expansions. II. Circular axis stellarator solutions

Here we apply the near-axis expansion method for quasisymmetric magnetic fields with anisotropic pressure to construct numerical solutions to circular axis stellarators. The solutions are found to second order in the distance from the axis, not possible in the standard Garren–Boozer construction, which assumes magnetostatic equilibria with isotropic pressure. In the limit of zero anisotropy, it is shown that a subset of coefficients can be chosen to avoid the overdetermination problem.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Correlation between the relative blob fraction and plasma parameters in NSTX

Here, we present a simple analysis of the fraction of blobs observed in the scrape-off layer of NSTX and calculates the correlation between this blob fraction and various plasma parameters. The measurements were made using the gas puff imaging (GPI) diagnostic on a database of 103 shots from 2010 with neutral beam power P NBI = 0–6 MW. The blob fraction ranged from f b = 0.1% to 4.8%, where f b is defined as the fraction of time the GPI signal was larger than three times its average value within 0–7.5 cm outside the separatrix. The blob fraction generally decreased with increasing neutral beam injection power and was lower on average in H-mode plasmas than Ohmic and L-mode plasmas. The blob fraction had its highest correlation with the measured poloidal turbulence velocity and its radial gradient just inside the separatrix. This is in part consistent with a model in which the blob generation rate increases with the poloidal velocity shear.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

A Pockels cell enabled heterodyne phase contrast imaging diagnostic for detection of ion cyclotron emission

This work describes a novel optical heterodyne detection scheme that significantly extends the frequency response of the phase contrast imaging method to detect electron density fluctuations in tens of megahertz frequency range. The system employs a variable frequency electro-optic modulator to allow operation at any frequency in the range 10–40 MHz without the need to realign the system. The frequency coverage of the system makes it suitable to measure the radial structure of the electron density component of ion cyclotron emission on devices having confining magnetic field of a few tesla, thus extending the purely temporal measurements provided so far by magnetic probes.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

High-power helicon system commissioning and conditioning progress on DIII-D

The high-power helicon, aka fast wave in the lower hybrid range of frequencies (LHRF), program has completed initial experiments at DIII-D. A 30-element comb-line traveling wave antenna successfully coupled over 0.3 MW of rf power at 476 MHz into both L-mode and ELMing H-mode plasmas for approximately 0.2 s. Here, we describe the helicon systems that enabled high-power operations, and the progress of conditioning in June 2021. Upgraded vacuum feedthroughs are described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Validation of MARS-F modeling of plasma response to RMPs using internal measurements on DIII-D

The linear resistive plasma response model is validated against the plasma internal measurement data from DIII-D edge-localized mode (ELM) control experiments with applied resonant magnetic perturbation (RMP). Considered are DIII-D discharges where the n = 1, 2, and 3 (n is the toroidal mode number) RMP field was applied. Experimental data for the plasma boundary displacement, as well as the three-dimensional (3D) pressure perturbation in the edge pedestal region, are deduced from the vertical Thomson scattering (TS) system and the horizontal charge exchange recombination (CER) system on DIII-D. The linear response model produces results that are in reasonable quantitative agreement with the DIII-D internal measurements. The plasma boundary displacement of up to 15 mm is modeled, with the pedestal pressure perturbation reaching 3 kPa. As an important insight, the larger plasma displacement measured by the vertical TS system, as compared to that measured by the horizontal CER system, is due to the contribution from the tangential component of the plasma displacement to the former. This mixing of displacement components is also found to influence the sensitivity of the CER measurement comparisons. The results of this study provide further confidence in the linear resistive plasma response model for analyzing ELM control experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Experimental inference of flux tunneling between magnetic island chains in tokamaks

We report confinement bifurcations due to flux tunneling in magnetic islands in toroidal plasmas. This occurs in 2/1 islands, triggered by decoupling of non-overlapping adjacent 5/2 islands. While subject to electron cyclotron current drive (ECCD), the temperature at the 2/1 O-point is flat before but peaks after decoupling. Here, this is shown to be due to a bifurcation from stochastic to nested magnetic structure, caused by the removal of flux tunneling through intersecting manifolds of overlapping heteroclinic tangles. 2/1 island stability correlates with coupling events in an ITER baseline scenario plasma where ECCD can not stabilize the coupled islands, showing the critical impact of flux tunneling on disruption avoidance in tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Disruptive neoclassical tearing mode seeding in DIII-D with implications for ITER

New studies identify the critical parameters and physics governing disruptive neoclassical tearing mode (NTM) onset. An m/n = 2/1 mode in DIII-D that begins to grow robustly after a seeding event (edge localized mode ELM or sawtooth precursor and crash) causes the mode rotation to drop close to the plasma's Er = 0 rest frame; this condition opens the stabilizing ion-polarization current 'gate' and destabilizes an otherwise marginally stable NTM. Our new experimental and theoretical insights and novel toroidal theory-based modeling are benchmarked and scalable to ITER and other future experiments. Here, the nominal ITER rotation at q = 2 is found to be stabilizing ('gate closed') except for MHD-induced transients that could 'open the gate'. Extrapolating from the DIII-D ITER baseline scenario (IBS) discharges, MHD transients are much more likely to destabilize problematic robustly growing 2/1 NTMs in ITER; this makes predictions of seeding and control of both ELMs and sawteeth imperative for more than just minimizing divertor pulsed-heat loading.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Direct preemptive stabilization of $m,n$ = 2,1 neoclassical tearing modes by electron cyclotron current drive in the DIII-D low-torque ITER baseline scenario

We report the first preemptive direct stabilization of $m,n$ = 2,1 neoclassical tearing modes (NTMs) in DIII-D low-torque ITER baseline scenario plasmas by electron cyclotron waves. These experiments strongly support that the observed stabilization is achieved by replacing the missing bootstrap current in the island by electron cyclotron current drive (ECCD). These plasmas (with and without ECCD) are stable to 2,1 NTMs when the differential rotation between the q = 1 and q = 2 surfaces (Δf 1,2 ) is sustained above a critical value (Δf CRIT ≈ 1 kHz), but those evolving to low differential rotation (Δf 1,2 < Δf CRIT ) routinely develop a disruptive 2,1 island at the time and frequency of a sawtooth precursor. Preemptive, local and direct stabilization by ECCD is tested by scanning the current drive amplitude and location inside and outside of the q = 2 rational surface shot by shot. Analysis of only low differential rotation time windows, i.e. when stabilization from Δf 1,2 is absent, shows focused ECCD at q = 2 prevents the onset of 2,1 NTMs. These results are important, as they give the first demonstration of disruptive NTM control by ECCD in the plasma scenario planned for ITER.

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 OSTI2025

Paradigm for global gyrokinetic turbulence

Using a novel wave-number-advection algorithm, we describe a method to add global profile curvature terms to the local gyrokinetic equations. This approach enables a high-precision global simulation capability without sacrificing the efficiency and spectral accuracy of local simulations. Preliminary numerical results suggest that there are two distinct types of global correction to the local result: (1) a trivial shift of the local result in radius, (2) a non-trivial decorrelation that is roughly equivalent to the well-known E×B shear effect. This interpretation provides a new, intuitive framework for understanding global gyrokinetic simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Experimental Observation of Magnetic Island Heteroclinic Bifurcation in Tokamaks

We report empirical observations of magnetic island heteroclinic bifurcation for the first time. This behavior is observed in interacting coupled 2/1 tearing modes in the core of a DIII-D tokamak plasma. Poincaré maps constrained by measured magnetic amplitudes and phasing show bifurcation from heteroclinic to homoclinic topology in the 2/1 island as the 4/2 relative amplitude (R 4/2 ) decreases. Initially, the local electron temperature peak in the 2/1 island splits, consistent with two O points. As R 4/2 decreases a single peak forms, consistent with one O point. Lastly, these call for developing tearing stability theory and control solutions for heteroclinic islands in tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

NSTX-U liquid metal core-edge facility (LMCE)

NSTX-U/LMCE will provide a unique and world-leading research facility to address the primary challenge to delivering economic and timely magnetic fusion energy, namely the need to develop a power and particle exhaust and first-wall system that can withstand very high edge heat fluxes, maximize energy confinement, and avoid the production of large masses of solid eroded first-wall material. The NSTX-U/LMCE facility will assess the ability of liquid metals (LMs) – especially liquid lithium – to provide a new boundary condition for magnetic fusion systems, to extend the lifetime of the plasma facing components (PFCs) and improve core plasma confinement. Such capability is needed to establish the basis for next-step fusion facilities including fusion pilot plants, and to maintain U.S. world leadership in core-edge integration research. NSTX-U/LMCE will leverage the ability to generate very high divertor perpendicular heat flux q⊥ ~ 100MW/m 2 , extensive diagnostics, and liquid-metal-applicable infrastructure of NSTX-U. NSTX-U/LMCE will provide access to a high-confinement plasma core with majority self-driven plasma current, the flexibility to test a range of liquid metal divertor concepts, access to a range of separatrix collisionalities (from high to very low), and the ability to controllably vary the first-wall temperature to vary the plasma- wall interaction physics on liquid lithium components. Further, NSTX-U/LMCE will utilize more reactor-relevant high-Z refractory-metal PFC substrates. With these capabilities the NSTX-U/LMCE facility will explore the full continuum of core-edge solutions ranging from high core radiated power, to conditions with radiative losses concentrated in the scrape-off layer (SOL), and ultimately low recycling conditions. The low collisionality SOL that may be accessible in the low recycling regime is relatively unexplored and will require a kinetic treatment of the edge, which can be addressed theoretically, and with experiments in LTX-β. Additional smaller-scale preparatory R&D facilities will be required to reduce the risk of premature technical/engineering failure of liquid metal systems implemented in NSTX-U. The NSTX-U/LMCE facility aligns very well with recommendations in the FESAC Long-Range Plan and NASEM Pilot Plant reports and the Bold Decadal Vision, will be unique in the world program throughout the next decade, and is garnering private company interest in utilizing NSTX-U/LMCE for development of LM PFCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗