Integration of full divertor detachment with improved core confinement for tokamak fusion plasmas

Divertor detachment offers a promising solution to the challenge of plasma-wall interactions for steady-state operation of fusion reactors. Here, we demonstrate the excellent compatibility of actively controlled full divertor detachment with a high-performance (βN ~ 3, H98 ~ 1.5) core plasma, using high-βp (poloidal beta, βp > 2) scenario characterized by a sustained core internal transport barrier (ITB) and a modest edge transport barrier (ETB) in DIII-D tokamak. The high-βp high-confinement scenario facilitates divertor detachment which, in turn, promotes the development of an even stronger ITB at large radius with a weaker ETB. This self-organized synergy between ITB and ETB, leads to a net gain in energy confinement, in contrast to the net confinement loss caused by divertor detachment in standard H-modes. These results show the potential of integrating excellent core plasma performance with an efficient divertor solution, an essential step towards steady-state operation of reactor-grade plasmas.

Tomographic analysis of tangential viewing cameras (invited)

Many tokamaks now use visible light cameras to observe plasma-wall interactions and integrated line emission. The DIII-D coherence imaging spectroscopy diagnostic cameras image interferograms that encode line integrated velocity. By modeling the 2D camera image pixels as line of sight integrals through an axisymmetric discrete grid, it is possible to do tomographic analysis to determine the local plasma line emissivity and parallel velocity. Methods to solve the inverse problem posed by these tangential viewing cameras are presented. The inversion begins with calculation of the sparse response matrix that encompasses all the geometry and diagnostic information and reduces the process of image formation to a sparse matrix-vector multiply. This work includes techniques for determining the detailed geometry of the camera views and methods for handling physical quantities that vary spatially. Additionally, the size of the response matrix has driven the development of capability to distribute the coarse parallel calculation across a heterogeneous cluster of computers on the Energy Sciences Network. Iterative techniques are then used to solve the sparse matrix-vector linear system.

Imaging divertor strike point splitting in RMP ELM suppression experiments in the DIII-D tokamak

Fast visible imaging of the lower divertor from above is used to study the structure and dynamics of lobes induced by resonant magnetic perturbations (RMPs) in Edge-Localized Mode (ELM) suppression experiments in DIII-D. The best compromise between the amount of light and sharp imaging was obtained using emission at 601 nm from Fulcher band molecular deuterium. Multiple spatially resolved peaks in the D₂ emission, taken as a proxy for the particle flux, are readily resolved during RMPs, in contrast to the heat flux measured by infrared cameras, which shows little spatial structure in ITER-like conditions. The 25 mm objective lens provides high spatial resolution (2-4 mm/pixel) from the centerpost to the outer shelf over 40° toroidally that overlaps the field of view of the IRTV that measures the divertor heat flux, allowing direct comparison in non-axisymmetric discharges. The image is coupled to a Phantom 7.3 camera using a Schott wound fiber bundle, providing high temporal resolution that allows the lobe dynamics to be resolved between ELMs and across ELM suppression onset. These measurements are used to study the heat and particle flux in 3D magnetic fields and to validate models for the plasma response to RMPs.

E×B Flux Driven Detachment Bifurcation in the DIII-D Tokamak

A bifurcative step transition from low-density, high-temperature, attached divertor conditions to high-density, low-temperature, detached divertor conditions is experimentally observed in DIII-D tokamak plasmas as density is increased. The step transition is only observed in the high confinement mode and only when the B×∇B drift is directed towards the divertor. This work reports for the first time a theoretical explanation and numerical simulations that qualitatively reproduce this bifurcation and its dependence on the toroidal field direction. According to the model, the bifurcation is primarily driven by the interdependence of the E×B-drift fluxes, divertor electric potential structure, and divertor conditions. In the attached conditions, strong potential gradients in the low field side (LFS) divertor drive E×B-drift flux towards the high field side divertor, reinforcing low density, high temperature conditions in the LFS divertor leg. At the onset of detachment, reduction in the potential gradients in the LFS divertor leg reduce the E×B-drift flux as well, such that the divertor plasma evolves nonlinearly to high density, strongly detached conditions. Experimental estimates of the E×B-drift fluxes, based on divertor Thomson scattering measurements, and their dependence on the divertor conditions are qualitatively consistent with the numerical predictions. The implications for divertor power exhaust and detachment control in the next step fusion devices are discussed.

Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

Access to a new plasma edge state with high density and pressures using the quiescent H mode

A path to a new high performance regime has been discovered in tokamaks that could improve the attractiveness of a fusion reactor. Experiments on DIII-D using a quiescent H-mode edge have navigated a valley of improved edge peeling-ballooning stability that opens up with strong plasma shaping at high density, leading to a doubling of the edge pressure over the standard H mode with edge localized modes at these parameters. The thermal energy confinement time increases as a result of both the increased pedestal height and improvements in the core transport and reduced low-k turbulence. Calculations of the pedestal height and width as a function of density using constraints imposed by peeling-ballooning and kinetic-ballooning theory are in quantitative agreement with the measurements.

Reduction of edge-localized mode intensity using high-repetition-rate pellet injection in tokamak H-mode plasmas

High repetition rate injection of deuterium pellets from the low-field side (LFS) of the DIII-D tokamak is shown to trigger high-frequency edge-localized modes (ELMs) at up to 12× the low natural ELM frequency in H-mode deuterium plasmas designed to match the ITER baseline configuration in shape, normalized beta, and input power just above the H-mode threshold. The pellet size, velocity, and injection location were chosen to limit penetration to the outer 10% of the plasma. The resulting perturbations to the plasma density and energy confinement time are thus minimal (<10%). The triggered ELMs occur at much lower normalized pedestal pressure than the natural ELMs, suggesting that the pellet injection excites a localized high-n instability. Triggered ELMs produce up to 12× lower energy and particle fluxes to the divertor, and result in a strong decrease in plasma core impurity density. These results show for the first time that shallow, LFS pellet injection can dramatically accelerate the ELM cycle and reduce ELM energy fluxes on plasma facing components, and is a viable technique for real-time control of ELMs in ITER.

Pedestal magnetic field measurements using a motional Stark effect polarimeter

Temperature-controlled, 0.15 nm interference filters were installed on an edge-viewing system of the motional Stark effect (MSE) polarimeter on the DIII-D tokamak. The upgraded system provides a factor of two reduction in the bandpass compared to the previous design, and linear control of the bandpass, which is unaltered by wavelength tuning. With the new system, there is a reduced dependence of the inferred polarization angle on the filter wavelength calibration. Recent measurements from the calibrated edge-viewing system show increased agreement with other MSE arrays.

Doppler coherence imaging and tomography of flows in tokamak plasmas (invited)

This article describes the results of spatial heterodyne Doppler "coherence imaging" of carbon ion flows in the divertor region of the DIII-D tokamak. Spatially encoded interferometric projections of doubly ionized carbon emission at 465 nm have been demodulated and tomographically inverted to obtain the spatial distribution of the carbon ion parallel flow and emissivity. The operating principles of the new instruments are described, and the link between measured properties and line integrals of the flow field are established. An iterative simultaneous arithmetic reconstruction procedure is applied to invert the interferometric phase shift projections, and the reconstructed parallel flow field amplitudes are found to be in reasonable agreement with UEDGE modeling.

Resonant pedestal pressure reduction induced by a thermal transport enhancement due to stochastic magnetic boundary layers in high temperature plasmas

Good alignment of the magnetic field line pitch angle with the mode structure of an external resonant magnetic perturbation (RMP) field is shown to induce modulation of the pedestal electron pressure p(e) in high confinement high rotation plasmas at the DIII-D tokamak with a shape similar to ITER, the next step tokamak experiment. This is caused by an edge safety factor q95 resonant enhancement of the thermal transport, while in contrast, the RMP induced particle pump out does not show a significant resonance. The measured p(e) reduction correlates to an increase in the modeled stochastic layer width during pitch angle variations matching results from resistive low rotation plasmas at the TEXTOR tokamak. These findings suggest a field line pitch angle resonant formation of a stochastic magnetic edge layer as an explanation for the q95 resonant character of type-I edge localized mode suppression by RMPs.

Quiescent H-mode plasmas with strong edge rotation in the cocurrent direction

For the first time in any tokamak, quiescent H-mode (QH-mode) plasmas have been created with strong edge rotation in the direction of the plasma current. This confirms the theoretical prediction that the QH mode should exist with either sign of the edge rotation provided the magnitude of the shear in the edge rotation is sufficiently large and demonstrates that counterinjection and counteredge rotation are not essential for the QH mode. Accordingly, the present work demonstrates a substantial broadening of the QH-mode operating space and represents a significant confirmation of the theory.

Measurements of spatial line emission profiles in the main scrape-off layer of the DIII-D tokamak

A video camera system is described as that measures the spatial distribution of visible line emission emitted from the main scrape-off layer (SOL) of plasmas in the DIII-D tokamak. A wide-angle lens installed on an equatorial port and an in-vessel mirror, which intercepts part of the lens' view, provide simultaneous tangential views of the SOL on the low-field and high-field sides of the plasma's equatorial plane. Tomographic reconstruction techniques are used to calculate the two-dimensional (2D) poloidal profiles from the raw data, and one-dimensional (1D) poloidal profiles simulating chordal views of other optical diagnostics from the 2D profiles. The 2D profiles can be compared with SOL plasma simulations; the 1D profiles with measurements from spectroscopic diagnostics. Sample results are presented, which elucidate carbon transport in plasmas with toroidally uniform injection of methane and argon transport in disruption mitigation experiments with massive gas jet injection.

Dust measurements in tokamaks (invited)

Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 microm in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

Suppression of large edge-localized modes in high-confinement DIII-D plasmas with a stochastic magnetic boundary

A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.