[en] We present fluid model simulation results for the edge plasma in the DIII-D tokamak with unbalanced double-null magnetic configurations, including cross field drifts. Input parameters are typical of low-power operation in DIII-D. For high-recycling the plasma tends to be detached from all divertor plates. Midplane plasma and electric field profiles are relatively insensitive to the magnetic imbalance. Divertor heat flux profiles exhibit sharp peaks due to cross-field drifts when the ion grad-B drift direction is away from the x-point toward the magnetic axis

[en] Measurements in DIII-D show that the carbon chemical sputtering sources along the inner divertor and center post are toroidally periodic and highest at the upstream tile edge. Imaging with a tangentially viewing camera and visible spectroscopy were used to monitor the emission from molecular hydrocarbons (CH/CD) at 430.8 nm and deuterium neutrals in attached and partially detached divertors of low-confinement mode plasmas. In contrast to the toroidally periodic CD distribution, emission from deuterium neutrals was observed to be toroidally symmetric along the inner strike zone. The toroidal distribution of the measured tile surface temperature in the inner divertor correlates with that of the CD emission, suggesting larger parallel particle and heat fluxes to the upstream tile edge, either due to toroidal tile gaps or height steps between adjacent tiles

[en] Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBI ⩽ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.

[en] In recent DIII-D experiments, we concentrated on extending the operating range and improving the overall performance of quiescent H-mode (QH) plasmas. The QH-mode offers an attractive, high-performance operating mode for burning plasmas due to the absence of pulsed edge-localized-mode-driven losses to the divertor (ELMs). Using counter neutral-beam injection (NBI), we achieve steady plasma conditions with the presence of an edge harmonic oscillation (EHO) replacing the ELMs and providing control of the edge pedestal density. These conditions have been maintained for greater than 4s (∼30 energy confinement times, τ_E, and 2 current relaxation times, τ_R [1]), and often limited only by the duration of auxiliary heating. We discuss results of these recent experiments where we use triangularity ramping to increase the density, neutral beam power ramps to increase the stored energy, injection of rf power at the electron cyclotron (EC) frequency to control density profile peaking in the core, and control of startup conditions to completely eliminate the transient ELMing phase

[en] Quiescent double barrier (QDB) conditions often form when an internal transport barrier is created with high-power neutral-beam injection into a quiescent H-mode (QH) plasma. These QH-modes offer an attractive, high-performance operating scenario for burning plasma experiments due to their quasi-stationarity and lack of edge localized modes (ELMs). Our initial experiments and modeling using ECH/ECCD in QDB shots were designed to control the current profile and, indeed, we have observed a strong dependence on the q-profile when EC-power is used inside the core transport barrier region. While strong electron heating is observed with EC power injection, we also observe a drop in the other core parameters; ion temperature and rotation, electron density and impurity concentration. These dynamically changing conditions provide a rapid evolution of T_e T_i profiles accessible with 0.3 < (T_e T_i)_axis < 0.8 observed in QDB discharges. We are exploring the correlation and effects of observed density profile changes with respect to these time-dependent variations in the temperature ratio. Thermal and particle diffusivity calculations over this temperature ratio range indicate a consistency between the rise in temperature ratio and an increase in transport corresponding to the observed change in density

[en] One of the key challenges for the conventional divertor tokamak is the plasma wall interaction interface and its materials. High divertor heat loads and material erosion in the spherical torus (ST) are of particular concern because of the compact divertor, and as a result, small plasma-wetted surfaces. The implications of the toroidal plasma physics at low aspect ratio and high β for edge energy and particle transport, properties of the scrape-off layer (SOL) and divertor are being studied on the National Spherical Torus Experiment (NSTX)--a medium size ST (R = 0.85 m, a = 0.67 m, A ≅ 1.27, β_t < 32 %, β_N < 5 %). NSTX operates routinely with stationary outer target plate peak heat loads up to 6 MW/m² for up to 1 s in the 6 MW NBI heated H-mode regime with type I, III, V ELMs , with the largest peak heat flux measured to date q_out = 10 MW/m² A detached divertor is an effective heat flux mitigation technique which has been developed in large aspect ratio tokamaks. Heat flux at the plate is reduced in the detached divertor through volumetric momentum and energy dissipative processes--the ion-neutral elastic collisions, recombination and radiative cooling

[en] The quiescent H (QH) mode, an edge localized mode (ELM)-free, high-confinement mode, combines well with an internal transport barrier to form quiescent double barrier (QDB) stationary state, high performance plasmas. The QH-mode edge pedestal pressure is similar to that seen in ELMing phases of the same discharge, with similar global energy confinement. The pedestal density in early ELMing phases of strongly pumped counter injection discharges drops and a transition to QH-mode occurs, leading to lower calculated edge bootstrap current. Plasmas current ramp experiment and ELITE code modeling of edge stability suggest that QH-modes lie near an edge current stability boundary. At high triangularity, QH-mode discharges operate at higher pedestal density and pressure, and have achieved ITER level values of β_PED and ν*. The QDB achieves performance of α_NH₈₉ ∼ 7 in quasi-stationary conditions for a duration of 10 tE, limited by hardware. Recently we demonstrated stationary state QDB discharges with little change in kinetic and q profiles (q₀ > 1) for 2 s, comparable to ELMing ''hybrid scenarios'', yet without the debilitating effects of ELMs. Plasma profile control tools, including electron cyclotron heating and current drive and neutral beam heating, have been demonstrated to control simultaneously the q profile development, the density peaking, impurity accumulation and plasma beta

[en] 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.

[en] We continue to explore Quiescent Double Barrier (QDB) operation on DIII-D to address issues of critical importance to internal transport barrier (ITB) plasmas. QDB plasmas exhibit both a core transport barrier and a quiescent, H-mode edge barrier. Both experiments and modeling of these plasmas are leading to an increased understanding of this regime and it's potential advantages for advanced-tokamak (AT) burning-plasma operation. These near steady plasma conditions have been maintained on DIII-D for up to 4s, times greater than 35τ_E, and exhibit high performance with β_N > 2.5 and neutron production rates S_n ∼ 1 x 10¹⁶s^-1. Recent experiments have been directed at exploring both the current profile modification effects of electron cyclotron current drive (ECCD) and electron cyclotron (ECH) heating-induced changes in temperature, density and impurity profiles. We use model-based analysis to determine the effects of both heating and current drive on the q-profile in these QDB plasmas. Experiments based on predictive modeling achieved a significant modification to the q-profile evolution [1] resulting from the non-inductive current drive effects due to direct ECCD and changes in the bootstrap and neutral beam current drive components. We observe that the injection of EC power inside the barrier region changes the density peaking from n_e/< n_e> = 2.1 to 1.5 accompanied by a significant reduction in the core carbon and high-Z impurities, nickel and copper

[en] 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.