Session CO6: DIII-D

Overview of Recent DIII-D Experimental Results

Recent DIII-D experiments have added to the ITER physics basis and to physics understanding for extrapolation to future devices. ELMs were suppressed by RMPs in He plasmas consistent with ITER non-nuclear phase conditions, and in steady state hybrid plasmas. Characteristics of the EHO during both standard high torque, and low torque enhanced pedestal QH-mode with edge broadband fluctuations were measured, including edge localized density fluctuations with a microwave imaging reflectometer. The path to Super H-mode was verified at high beta with a QH-mode edge, and in plasmas with ELMs triggered by Li granules. ITER acceptable TQ mitigation was obtained with low Ne fraction Shattered Pellet Injection. Divertor ne and Te data from Thomson Scattering confirm predicted drift-driven asymmetries in electron pressure, and X-divertor heat flux reduction and detachment were characterized. The crucial mechanisms for ExB shear control of turbulence were clarified. In collaboration with EAST, high beta-p scenarios were obtained with 80$\%$ bootstrap fraction, high H-factor and stability limits, and large radius ITBs leading to low AE activity.   

Type-I ELM Suppression by Resonant Fields in Steady State Hybrid Plasmas on DIII-D

``Steady-state'' hybrid plasmas in DIII-D with zero surface loop voltage have been maintained for $\le$2 current relaxation times with complete suppression of Type-I ELMs using n=3 Resonant Magnetic Perturbations (RMPs). The edge $q_{95} \approx$ 6 and q(0) $>$1 due to anomalous poloidal magnetic flux pumping in the hybrid scenario. The sustained hybrid performance without type-I ELMs indicates an alternative mechanism for flux pumping, possibly due to the interaction of the m/n=3/2 mode with an m=1/n=1 kink displacement. The plasma confinement degradation during ELM suppression is less than 10$\%$, with H$_{98y2} \approx$ 1.2 sustained for the duration of the discharge. After ELM suppression is obtained, $\sim$1 kHz D-alpha bursts are observed. The underlying instability may be beneficial for particle and impurity control.   

Rotational Shear Effects on Edge Harmonic Oscillations in DIII-D Quiescent H-mode Discharges

In quiescent H-mode (QH) regime, the edge harmonic oscillations (EHO) play an important role in avoiding the transient ELM power fluxes by providing benign and continuous edge particle transport. A detailed theoretical, experimental and modeling comparison has been made of low-n (n=5) EHO in DIII-D QH-mode plasmas. The calculated linear eigenmode structure from the extended MHD code M3D-C1 matches closely the coherent EHO properties from external magnetics data and internal measurements using the ECE, BES, ECE-I and MIR diagnostics, as well as the kink/peeling mode properties of the ideal MHD code ELITE. The numerical investigations indicate that the low-n EHO-like solutions from M3D-C1 are destabilized by the toroidal rotational shear while high-n modes are stabilized. This effect is independent of the rotation direction, suggesting that the low-n EHO can be destabilized in principle with rotation in both directions. These modeling results are consistent with experimental observations of the EHO and support the proposed theory of the EHO as a rotational shear driven kink/peeling mode [1][2].   

Enhanced H-mode pedestals with lithium injection in DIII-D

ELM-free H-mode periods with increased pedestal pressure and width were observed on DIII-D when density fluctuations localized near the separatrix were present. Lithium powder injection increased the duration of these enhanced pedestal phases, and also the likelihood of a transition to this regime. The fluctuations, $\tilde{n}/n\sim0.1$, f $\sim$ 80 kHz, occur in bursts every $\sim$ 1 ms, with frequency varying within each burst. The mode propagates in the electron diamagnetic drift direction with $k_\theta \rho_s \sim0.1-0.2$, consistent with a trapped electron or micro-tearing instability. The radial structure of the mode indicates outward radial propagation, and its presence correlates with flattening of the pressure profile near the separatrix. This flattening moves the pedestal high pressure gradient region inward, allowing higher pedestal pressure at the peeling-ballooning stability limit. Lithium injection at a level sufficient for triggering the extended enhanced phases resulted in significant lithium in the plasma core, but carbon and other higher Z impurities as well as radiated power levels were reduced, while recycling of the working deuterium gas appeared to be unaffected.   

Shattered Pellet Injection System applied to thermal quench and runaway mitigation

Shattered pellet injection (SPI) has demonstrated excellent performance in terms of response time and global plasma thermal energy dissipation. The ITER SPI system has to achieve heat dissipation while keeping radiation asymmetries at an acceptable level, and efficient mitigation of runaway electrons. These capabilities have been studied for the first time on DIII-D. Experiments focused on the characterization of the radiation asymmetries during an SPI shutdown and its relation to the large MHD activity observed during massive gas injection (MGI) cases to be a dominant effect in radiation asymmetries. The observed higher core radiation during for SPI could indicate differences in the energy dissipation: MHD may have a smaller effect since the particles deposition is deeper than for MGI. SPI was also applied to a developed runaway beam. The consequences on the runaway losses and current decay have been studied to determine its potential as a runaway suppression tool.   

Effect of ECH on Turbulent Fluctuations During ITER Baseline Discharges on DIII D

Recent experiments on the DIII-D tokamak simulating the ITER Baseline Scenario show that torque-free and spatially localized Electron Cyclotron Heating (ECH), compared to neutral beam heating, modifies flow shear and density fluctuations, resulting in a slightly worse confinement degradation than that given by the IPB98(y,2) scaling[1]. After turning off ECH, the Phase Contrast Imaging diagnostic measures, on confinement time scales, a decrease in the intensity of fluctuations at frequencies lower than 200 kHz, consistent with the mean flow shear exceeding the maximum linear growth rate of ITG modes. In contrast, at higher frequencies the intensity of fluctuations increases promptly, due to ETG modes enhanced by the prompt increase of the electron temperature inverse scale length in the outer third of the minor radius. The latter effect is seen in preliminary non-linear gyro-kinetic simulations to generate a larger transient heat flux and an inward particle pinch.\par \vskip6pt \noindent [1] R.I. Pinsker, EPJ Web of conferences 87 (2015) 02003.   

Energy Released During the H-L Back Transition

Prompt energy loss ($\Delta$W) at the H-L transition, as a fraction of total stored energy before the transition, is about 30$\%$ and is insensitive to density in ITER-similar DIII-D plasmas. Occasionally, some ELMs will appear before the transition and reduce total energy, thus reducing $\Delta$W across the following transition. Other results (not in the ITER-similar shape) have shown that ELMs can be triggered in low powered H-modes, prior to H-L transitions, when the plasma is stable to ideal P-B modes (these are not typical type-I ELMs, despite superficial similarities) and $E\times B$ shear is strong. These are indeed ELMs occurring in H-mode and not part of a dithering transition. Finally, ELM $\Delta$W is sensitive to edge toroidal rotation and becomes smaller than uncertainty ($<$ 5 kJ) at low rotation ($\omega_{tor}<$ 5 krad/s). These results point to a strategy where $\Delta$W for the H-L transition may be reduced by the presence of (not type-I) ELMs before the transition, and $\Delta$W for the ELMs may be reduced by controlling rotation.   

Scaling of Intrinsic Torque with Normalized Gyroradius in DIII-D

Experiments at DIII-D have investigated the scaling of intrinsic torque with the normalized gyroradius, $\rho $*, by performing a dimensionless parameter scan. High- and low-$\rho $* versions of a plasma have been created such that the scan can be combined with similar scans on other tokamaks. Intrinsic torque and momentum transport coefficients are measured by applying repeated torque steps at constant power with co- and counter-current neutral beam injection. The intrinsic torque was also measured by determining the neutral beam torque required to null the toroidal rotation profile. These results show that the density of intrinsic torque in the core of these plasmas more than doubles while $\rho $* decreases by approximately 23{\%}. Independent measurements of intrinsic torque are consistent with each other and show that the dependence of intrinsic torque on the Mach number is minimal. The scaling of the Prandtl number and momentum pinch with $\rho $* is also investigated.   

Joint DIII-D/EAST Experiments Toward Steady State AT Demonstration

Joint DIII-D/EAST experiments on fully noninductive operation at high poloidal beta have demonstrated several attractive features of this regime for a steady-state fusion reactor. Very large bootstrap fraction ($>$80$\%$) is desirable because it reduces the demands on external noninductive current drive. High bootstrap fraction with an H-mode edge results in a broad current profile and internal transport barriers (ITBs) at large minor radius, leading to high normalized energy confinement and high MHD stability limits. The ITB radius expands with higher normalized beta, further improving both stability and confinement. Electron density ITB and large Shafranov shift lead to low AE activity in the plasma core and low anomalous fast ion losses. Both the ITB and the current profile show remarkable robustness against perturbations, without external control.   

Alfven Eigenmode Stability and Fast Ion Transport in High $q_{min}$ Steady State Discharges on DIII-D

A wide range of Alfven eigenmode (AE) activity and beam ion loss is observed in high $q_{min}$ steady state target plasmas on DIII-D. Modeling the losses with the NOVA-k code and the Critical Gradient Model indicates that the observed reduction in the neutron signal, usually up to $20\%$, can be attributed to the AEs. In those high qmin reversed shear discharges both normal shear and reversed shear AEs are excited. The normal shear AEs can be suppressed by increasing the pressure gradient or increasing the q(0). However, the reversed shear AEs emerge from the Alfven continuum above a critical pressure gradient. It will be shown that both the normal and reversed shear AEs can be suppressed or their effects strongly mitigated by raising q(0) and moving qmin to larger radius. This prediction is consistent with observations in DIII-D EAST SS plasmas.   

Measurement and Simulation of First-Orbit Fast-Ion D-Alpha Emission and the Application to Fast-Ion Loss Detection in the DIII-D Tokamak

A new fast-ion diagnostic method uses passive emission of D-alpha radiation to determine fast-ion losses quantitatively. The passive fast-ion D-alpha simulation (P-FIDAsim) forward models the Doppler-shifted spectra of first-orbit fast ions that charge exchange with edge neutrals. Simulated spectra are up to 80$\%$ correlated with experimental spectra. Calibrated spectra are used to estimate the 2D neutral density profile by inverting simulated spectra. The inferred neutral density shows the expected increase toward each x-point and an average value of $8\times 10$$^9$ cm$^{-3}$ at the plasma boundary and $1\times 10$$^{11}$ cm$^{-3}$ near the wall. Measuring and simulating first-orbit spectra effectively ``calibrates'' the system, allowing for the quantification of more general fast-ion losses. Sawtooth crashes are estimated to eject 1.2$\%$ of the fast-ion inventory, in good agreement with a 1.7$\%$ loss estimate made by TRANSP. Sightlines sensitive to passing ions observe larger sawtooth losses than sightlines sensitive to trapped ions.   

Scaling of midplane separatrix density with power at divertor detachment onset

The midplane separatrix density at divertor detachment onset is found to increase with higher parallel heat flux, $q_{||}$, flowing into the divertor, but at a slower rate than expected from simple scaling models. The separatrix density will be an important parameter in determining the compatibility of divertor heat flux control with robust pedestal operation and high core confinement in future devices. The parallel heat flux is examined by separately varying several parameters, including injected power, plasma current, toroidal field and injected impurities. Several methods are employed to locate the separatrix in this critical region of steep density gradients, including magnetic equilibrium reconstruction, power balance assumptions and spatial fiducials from other diagnostics. All methods exhibit a slower than the $q_{||}^{5/7}$ scaling predicted by a simple two point model. The nonlinear dependence of divertor radiation with power and density is one of several factors leading to this difference.   

Modeling Detached Divertor Plasma Characteristics in the DIII-D Tokamak

Detached divertor-plasma operation, where a large fraction of the core exhaust power is radiated before striking the target plates, is attractive for limiting the peak target heat flux. Such plasmas have electron temperature $\sim$ 1 eV near the target. Changing the position of the separatrix strike points on the geometrically varied DIII-D target plates is allowing a systematic study of how plate shape impacts accessibility to detached operation. Reported here are 2D plasma/neutral transport simulations of these configurations using the UEDGE code including cross-field drifts and impurities. Results are given on how the onset of detachment scales with strike-point location, wall pumping of neutrals, separatrix density, and core power. Different initial conditions sometimes yield different steady-state solutions for identical input parameters, one being an attached plasma and the other detached. Comparisons are made of simulation results and experimental measurements, especially divertor Thomson scattering data.   

Reduction in resonant magnetic field induced heat flux splitting caused by detachment of the divertor

Measurements in DIII-D show that in high-density detached divertor conditions, the inter-ELM non-axisymmetric heat flux striations generated by resonant magnetic perturbations (RMPs) are eliminated. Non-axisymmetric heat loads caused by the RMP fields used to mitigate ELMs could reduce the lifetime of divertor components in ITER and future devices. It is shown that for RMPs with an n=3 toroidal mode number low levels of gas puffing can cause an increase in the heat flux splitting, but at high densities where the divertor becomes detached this splitting is eliminated. VUV imaging and 2D divertor Thomson scattering are used to measure RMP induced perturbations to the plasma conditions above the target plates. Modeling performed with the 3D fluid transport code EMC3-EIRENE both with and without the plasma response calculated by M3D-C1 is compared to the measured divertor conditions.   

Modeling of high-Z materials erosion and its suppression in DIII-D

Erosion of plasma facing components is a key issue for high-power, long pulse operation. The 3D Monte Carlo code ERO has been used to simulate the erosion/redeposition of Mo and W samples exposed to DIII-D divertor plasma using the DiMES. The net erosion rate is significantly reduced due to the high local re-deposition ratio of eroded materials, which is mainly controlled by the electric field and plasma density within the Chodura sheath as indicated by ERO modeling. Similar re-deposition ratios were obtained from the modeling using three sheath models for small inclined magnetic field angle, all being close to the measured value. ERO modeling shows that local CH$_4$ injection can create a carbon coating on the Mo sample to mitigate Mo erosion; the local decrease of electron temperature due to gas injection also suppresses net erosion, consistent with experimental observation.