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.

Demonstration of Tokamak Discharge Shutdown with Shell Pellet Payload Impurity Dispersal

The first rapid tokamak discharge shutdown using dispersive core payload deposition with shell pellets has been achieved in the DIII-D tokamak. Shell pellets are being investigated as a possible new path toward achieving tokamak disruption mitigation with both low conducted wall heat loads and slow current quench. Conventional disruption mitigation injects radiating impurities into the outer edge of the tokamak plasma, which tends to result in poor impurity assimilation and creates a strong edge cooling and outward heat flow, thus requiring undesirable high-Z impurities to achieve low conducted heat loads. The shell pellet technique aims to produce a hollow temperature profile by using a thin, low-ablation shell surrounding a dispersive payload, giving a greatly increased impurity ablation (and radiation) rate when the payload is released in the plasma core. This principle was demonstrated successfully using 3.6 mm outer diameter, 40  μm thickness diamond shells holding boron powder. The pellets caused rapid (<10  ms) discharge shutdown with low conducted divertor heat fluence (∼0.1  MJ/m^{2}). Confirmation of massive release of the boron powder payload into the plasma core was obtained spectroscopically. Some evidence for the formation of a hollow temperature profile during the shutdown was observed. These first results open a new avenue for disruption mitigation research, hopefully enabling development of highly effective methods of avoiding disruption wall damage in future reactor-scale tokamaks.

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.

Verification of Doppler coherence imaging for 2D ion velocity measurements on DIII-D

Coherence Imaging Spectroscopy (CIS) has emerged as a powerful tool for investigating complex ion phenomena in the boundary of magnetically confined plasma devices. The combination of Fourier-transform interferometry and high-resolution fast-framing cameras has made it possible to make sensitive velocity measurements that are also spatially resolved. However, this sensitivity makes the diagnostic vulnerable to environmental effects including thermal drifts, vibration, and magnetic fields that can influence the velocity measurement. Additionally, the ability to provide an absolute calibration for these geometries can be impacted by differences in the light-collection geometry between the plasma and reference light source, spectral impurities, and the presence of thin-films on in-vessel optics. This paper discusses the mitigation of these effects and demonstration that environmental effects result in less than 0.5 km/s error on the DIII-D CIS systems. A diagnostic comparison is used to demonstrate agreement between CIS and traditional spectroscopy once tomographic artifacts are accounted for.

Upper wide-angle viewing system for ITER

The Upper Wide Angle Viewing System (UWAVS) will be installed on five upper ports of ITER. This paper shows major requirements, gives an overview of the preliminary design with reasons for some design choices, examines self-emitted IR light from UWAVS optics and its effect on accuracy, and shows calculations of signal-to-noise ratios for the two-color temperature output as a function of integration time and divertor temperature. Accurate temperature output requires correction for vacuum window absorption vs. wavelength and for self-emitted IR, which requires good measurement of the temperature of the optical components. The anticipated signal-to-noise ratio using presently available IR cameras is adequate for the required 500 Hz frame rate.

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.

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.

Numerical modeling of edge-localized-mode filaments on divertor plates based on thermoelectric currents

Edge localized modes (ELMs) are qualitatively and quantitatively modeled in tokamaks using current bursts which have been observed in the scrape-off-layer (SOL) during an ELM crash. During the initial phase of an ELM, a heat pulse causes thermoelectric currents. They first flow in short connection length flux tubes which are initially established by error fields or other nonaxisymmetric magnetic perturbations. The currents change the magnetic field topology in such a way that larger areas of short connection length flux tubes emerge. Then currents predominantly flow in short SOL-like flux tubes and scale with the area of the flux tube assuming a constant current density. Quantitative predictions of flux tube patterns for a given current are in excellent agreement with measurements of the heat load and current flow at the DIII-D target plates during an ELM cycle.

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.

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.

Mitigation of tokamak disruptions using high-pressure gas injection

High-pressure gas-jet injection of neon and argon is shown to be a simple and robust method to mitigate the deleterious effects of disruptions on the DIII-D tokamak. The gas jet penetrates to the central plasma at its sonic velocity. The deposited species dissipates >95% of the plasma by radiation and substantially reduces mechanical stresses on the vessel caused by poloidal halo currents. The gas-jet species-charge distribution can include >50% fraction neutral species which inhibits runaway electrons. The favorable scaling of this technique to burning fusion plasmas is discussed.