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Greenwald, M.

Publications and source records attributed to Greenwald, M..

14 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

The role of ion and electron-scale turbulence in setting heat and particle transport in the DIII-D ITER baseline scenario

In this work, the heat and particle transport in a DIII-D ITER Baseline Scenario (IBS) discharge has been investigated using both linear and nonlinear gyrokinetic simulations performed with the CGYRO code [J. Candy et al: Journal Comp. Phys. (2016)].These simulations were used to investigate the role that ion-scale ($k_θp_s$ < 1:0) and electron-scale ($k_θp_s$ > 1:0) turbulence play in determining heat and particle transport in the core of conditions believed to directly extrapolate to ITER operation. This investigation spans over nearly half of the plasma minor radius, from ρ = 0:45 - 0:85. To probe the nature of the transport and turbulence in these conditions and to validate the model against experimental fluxes, scans of $a/L_{T_i}$, $a/L_{T_e}$, and $a/L_n$ were performed at all radial locations. Long wavelength turbulence is found to be dominated by ITG modes with strong non-adiabatic electron effects and appears unable to reproduce ion and electron heat fluxes and electron particle fluxes simultaneously at most radial positions when single parameter scans are performed. To investigate the potential role of short wavelength turbulence, a series of electron-scale simulations are performed that indicate that experimentally relevant levels of electron heat flux could arise at sub-ion scales. Quantitative comparison of the simulated fluxes obtained from ion and electron-scale simulations with experimental levels along with the sensitivity of simulations results to changes in inputs is presented. Through extensive sensitivity scans and comparison with multiple transport channels, this work demonstrates a clear need for self-consistent modification of multiple inputs and suggests multi-scale interactions play a significant role at many of the radial locations studied. The results of this analysis help shape our understanding of the model fidelity needed to predict turbulence and transport reactor-relevant conditions and have important implications for the prediction of future fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Quasi-Coherent Fluctuations Limiting the Pedestal Growth on Alcator C-Mod: Experiment and Modeling

Performance predictions for future fusion devices rely on an accurate model of the pedestal structure. The leading candidate for predictive pedestal structure is EPED, and it is imperative to test the underlying hypotheses to further gain confidence for ITER projections. Here, we present experimental work testing one of the EPED hypotheses, namely the existence of a soft limit set by microinstabilities such as the kinetic ballooning mode (KBM). This work extends recent work on Alcator C-Mod [Diallo, et al., Phys. Rev. Lett., , (2014), 115001], to include detailed measurements of the edge fluctuations and comparisons of edge simulation codes and experimental observations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Alcator C-Mod: Research in support of ITER and steps beyond

This paper presents an overview of recent highlights from research on Alcator C-Mod. Significant progress has been made across all research areas over the last two years, with particular emphasis on divertor physics and power handling, plasma-material-interaction studies, ELM-suppressed pedestal dynamics, core transport and turbulence, and RF heating and current drive utilizing Ion Cyclotron and Lower Hybrid tools. Specific results of particular relevance to ITER include: inner wall SOL transport studies that have led, together with results from other experiments, to the change of the detailed shape of the inner wall in ITER; runaway electron studies showing that the critical electric field required for runaway generation is much higher than predicted from collisional theory; core tungsten impurity transport studies reveal that tungsten accumulation is naturally avoided in typical C-Mod conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

MHD stability and disruptions in the SPARC tokamak

SPARC is being designed to operate with a normalized beta of $β_N$ = 1.0, a normalized density of $n_G$ = 0.37 and a safety factor of $q_{95}$ ≈ 3.4, providing a comfortable margin to their respective disruption limits. Further, a low beta poloidal $β_p$ = 0.19 at the safety factor $\textit{q}$ = 2 surface reduces the drive for neoclassical tearing modes, which together with a frozen-in classically stable current profile might allow access to a robustly tearing-free operating space. Although the inherent stability is expected to reduce the frequency of disruptions, the disruption loading is comparable to and in some cases higher than that of ITER. The machine is being designed to withstand the predicted unmitigated axisymmetric halo current forces up to 50 MN and similarly large loads from eddy currents forced to flow poloidally in the vacuum vessel. Runaway electron (RE) simulations using GO+CODE show high flattop-to-RE current conversions in the absence of seed losses, although NIMROD modelling predicts losses of ~80 %; self-consistent modelling is ongoing. A passive RE mitigation coil designed to drive stochastic RE losses is being considered and COMSOL modelling predicts peak normalized fields at the plasma of order 10 –2 that rises linearly with a change in the plasma current. Massive material injection is planned to reduce the disruption loading. A data-driven approach to predict an oncoming disruption and trigger mitigation is discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Projections of H-mode access and edge pedestal in the SPARC tokamak

In order to inform core performance projections and divertor design, the baseline SPARC tokamak plasma discharge is evaluated for its expected H-mode access, pedestal pressure and edge-localized mode (ELM) characteristics. A clear window for H-mode access is predicted for full field DT plasmas, with the available 25 MW of design auxiliary power. Additional alpha heating is likely needed for H-mode sustainment. Pressure pedestal predictions in the developed H-mode are surveyed using the EPED model. The projected SPARC pedestal would be limited dominantly by peeling modes and may achieve pressures in excess of 0.3 MPa at a density of approximately 3 × 10 20 m -3 . High pedestal pressure is partially enabled by strong equilibrium shaping, which has been increased as part of recent design iterations. Edge-localized modes (ELMs) with >1 MJ of energy are projected, and approaches for reducing the ELM size, and thus the peak energy fluence to divertor surfaces, are under consideration. The high pedestal predicted for SPARC provides ample margin to satisfy its high fusion gain (Q) mission, so that even if ELM mitigation techniques result in a 2x reduction of the pedestal pressure, Q > 2 is still predicted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Access to pedestal pressure relevant to burning plasmas on the high magnetic field tokamak Alcator C-Mod

Experiments on the Alcator C-Mod tokamak have utilized reactor-relevant magnetic fields to sustain substantially higher pedestal pressure than in other devices and allow close approach to the ITER H-mode baseline target pedestal pressure of 90 kPa. The EPED model, which couples the physics of transport driven by kinetic ballooning modes and MHD instabilities arising from peeling-ballooning modes, predicts the pressure profile at the onset of edge-localized modes (ELMs), and yields to lowest order a critical-βN like behavior for the pedestal: p∝Bt×Bp ( ∝Bt^2 for fixed edge q). C-Mod routinely accesses edge plasma pressure in excess of 30 kPa, often by using a high-density (ne>3×10^20 m^-3) approach to high confinement, taking advantage of a regime known as enhanced D-alpha (EDA) H-mode. In the EDA H-mode, plasma transport regulates both the pedestal profiles and the core impurity content, thus holding the pedestal stationary at just below the peeling-ballooning stability boundary. This stationary ELM-suppressed regime has approached the maximum pedestal predicted by EPED at these densities: 60 kPa. This in turn gives rise to volume-averaged core plasma pressure in excess of 0.2MPa, a world record value for a magnetic fusion device. Another approach to achieving high pressure utilizes a pedestal limited by current-driven modes at low collisionality, in which pressure increases with density and which allows access to a higher EPED solution, termed “super-H”. C-Mod experiments at reduced density (ne<2×10^20 m^-3) and strong plasma shaping (δ>0.5) accessed this regime, producing pedestals with pressures up to 80kPa (approximately 90% of the ITER target) and temperatures of nearly 2 keV. In a number of these hot H-modes, we observe strong edge instabilities at low toroidal mode number (n=1) when pedestal pressure approaches predicted values from EPED, showing that current-driven MHD modes can serve as a limit on the pedestal in a metal-walled tokamak at high pressure and low collisionality.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Access to pedestal pressure relevant to burning plasmas on the high magnetic field tokamak Alcator C-Mod

Here, experiments on the Alcator C-Mod tokamak have utilized reactor-relevant magnetic fields to sustain substantially higher pedestal pressure than in other devices and allow close approach to the ITER H-mode baseline target pedestal pressure of 90 kPa. The EPED model, which couples the physics of transport driven by kinetic ballooning modes and MHD instabilities arising from peeling-ballooning modes, predicts the pressure profile at the onset of edge-localized modes (ELMs), and yields to lowest order a critical-β N like behavior for the pedestal: p α B t x B p (α B 2 t for fixed edge q). C-Mod routinely accesses edge plasma pressure in excess of 30 kPa, often by using a high-density (n¯ e > 3 x 10 20 m –3 ) approach to high confinement, taking advantage of a regime known as enhanced D-alpha (EDA) H-mode.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

A fast low-to-high confinement mode bifurcation dynamics in the boundary-plasma gyrokinetic code XGC1

Here, a fast edge turbulence suppression event has been simulated in the electrostatic version of the gyrokinetic particle-in-cell code XGC1 in a realistic diverted tokamak edge geometry under neutral particle recycling. The results show that the sequence of turbulent Reynolds stress followed by neoclassical ion orbit-loss driven together conspire to form the sustaining radial electric field shear and to quench turbulent transport just inside the last closed magnetic flux surface. As a result, the main suppression action is located in a thin radial layer around ψ N ≃0.96–0.98, where ψ N is the normalized poloidal flux, with the time scale ~0.1 ms.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Electron critical gradient scale length measurements of ICRF heated L-mode plasmas at Alcator C-Mod tokamak

A profile for the critical gradient scale length ( L c ) has been measured in L-mode discharges at the Alcator C-Mod tokamak, where electrons were heated by an ion cyclotron range of frequency through minority heating with the intention of simultaneously varying the heat flux and changing the local gradient. The electron temperature gradient scale length ( L Te -1 = |∇ T e |/ T e ) profile was measured via the B T -jog technique and it was compared with electron heat flux from power balance (TRANSP) analysis. The T e profiles were found to be very stiff and already above the critical values, however, the stiffness was found to be reduced near the q = 3/2 surface. The measured L c profile is in agreement with electron temperature gradient (ETG) models which predict the dependence of L c -1 on local , Z eff , T e /T i , and the ratio of the magnetic shear to the safety factor. The results from linear Gene gyrokinetic simulations suggest ETG to be the dominant mode of turbulence in the electron scale ( k ⊥ ρ s > 1), and ion temperature gradient/trapped electron mode modes in the ion scale ( k ⊥ ρ s > 1). Finally, the measured L c profile is in agreement with the profile of ETG critical gradients deduced from GENE simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Multi-scale gyrokinetic simulations of an Alcator C-Mod, ELM-y H-mode plasma

High fidelity, multi-scale gyrokinetic simulations capable of capturing both ion (${k}_{\theta }{\rho }_{s}\sim { \mathcal O }(1.0)$) and electron-scale (${k}_{\theta }{\rho }_{e}\sim { \mathcal O }(1.0)$) turbulence were performed in the core of an Alcator C-Mod ELM-y H-mode discharge which exhibits reactor-relevant characteristics. These simulations, performed with all experimental inputs and realistic ion to electron mass ratio (${({m}_{i}/{m}_{e})}^{1/2}=60.0$) provide insight into the physics fidelity that may be needed for accurate simulation of the core of fusion reactor discharges. Three multi-scale simulations and series of separate ion and electron-scale simulations performed using the GYRO code are presented. As with earlier multi-scale results in L-mode conditions, both ion and multi-scale simulations results are compared with experimentally inferred ion and electron heat fluxes, as well as the measured values of electron incremental thermal diffusivities—indicative of the experimental electron temperature profile stiffness. Consistent with the L-mode results, cross-scale coupling is found to play an important role in the simulation of these H-mode conditions. Extremely stiff ion-scale transport is observed in these high-performance conditions which is shown to likely play and important role in the reproduction of measurements of perturbative transport. In conclusion, these results provide important insight into the role of multi-scale plasma turbulence in the core of reactor-relevant plasmas and establish important constraints on the fidelity of models needed for predictive simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Fast low-to-high confinement mode bifurcation dynamics in a tokamak edge plasma gyrokinetic simulation

Transport barrier formation and its relation to sheared flows in fluids and plasmas are of fundamental interest in various natural and laboratory observations and of critical importance in achieving an economical energy production in a magnetic fusion device. Here we report the first observation of an edge transport barrier formation event in an electrostatic gyrokinetic simulation carried out in a realistic diverted tokamak edge geometry under strong forcing by a high rate of heat deposition. Here, the results show that turbulent Reynolds-stress-driven sheared E x B flows act in concert with neoclassical orbit loss to quench turbulent transport and form a transport barrier just inside the last closed magnetic flux surface.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗