DIII-D research advancing the physics basis for optimizing the tokamak approach to fusion energy

DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I (p) steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L-H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at similar to 8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I (p) beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate beta (N) in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation

Gyrokinetic analysis and simulation of pedestals, to identify the culprits for energy losses using fingerprints

Fusion performance in tokamaks hinges critically on the efficacy of the Edge Transport Barrier (ETB) at suppressing energy losses. The new concept of fingerprints is introduced to identify the instabilities that cause the transport losses in the ETB of many of today's experiments, from widely posited candidates. Analysis of the Gyrokinetic-Maxwell equations, and gyrokinetic simulations of experiments, find that each mode type produces characteristic ratios of transport in the various channels: density, heat and impurities. This, together with experimental observations of transport in some channel, or, of the relative size of the driving sources of channels, can identify or determine the dominant modes causing energy transport. In multiple ELMy H-mode cases that are examined, these fingerprints indicate that MHD-like modes are apparently not the dominant agent of energy transport; rather, this role is played by Micro-Tearing Modes (MTM) and Electron Temperature Gradient (ETG) modes, and in addition, possibly Ion Temperature Gradient (ITG)/Trapped Electron Modes (ITG/TEM) on JET. MHD-like modes may dominate the electron particle losses. Fluctuation frequency can also be an important means of identification, and is often closely related to the transport fingerprint. The analytical arguments unify and explain previously disparate experimental observations on multiple devices, including DIII-D, JET and ASDEX-U, and detailed simulations of two DIII-D ETBs also demonstrate and corroborate this

Mitigation of plasma-wall interactions with low-Z powders in DIII-D high confinement plasmas

Experiments with low-Z powder injection in DIII-D high confinement discharges demonstrated increased divertor dissipation and detachment while maintaining good core energy confinement. Lithium (Li), boron (B), and boron nitride (BN) powders were injected in high-confinement mode plasmas ($I_p=$1 MA, $B_t=$2 T, $P_{NB}=$6 MW, $\langle n_e\rangle=3.6-5.0\cdot10^{19}$ m$^{-3}$) into the upper small-angle slot (SAS) divertor for 2-s intervals at constant rates of 3-204 mg/s. The multi-species BN powders at a rate of 54 mg/s showed the most substantial increase in divertor neutral compression by more than an order of magnitude and lasting detachment with minor degradation of the stored magnetic energy $W_{mhd}$ by 5%. Rates of 204 mg/s of boron nitride powder further reduce ELM-fluxes on the divertor but also cause a drop in confinement performance by 24% due to the onset of an $n=2$ tearing mode. The application of powders also showed a substantial improvement of wall conditions manifesting in reduced wall fueling source and intrinsic carbon and oxygen content in response to the cumulative injection of non-recycling materials. The results suggest that low-Z powder injection, including mixed element compounds, is a promising new core-edge compatible technique that simultaneously enables divertor detachment and improves wall conditions during high confinement operation

In-situ coating of silicon-rich films on tokamak plasma-facing components with real-time Si material injection

Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from $3.9-7.5\times10^{19}$ m$^{-3}$ and input powers ranging from 5.5-9 MW. The small Si pellets were delivered with the impurity granule injector (IGI) at frequencies ranging from 4-16 Hz corresponding to mass flow rates of 5-19 mg/s ($1-4.2\times10^{20}$ Si/s) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system (DiMES) to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings of varying morphology mainly containing silicon oxides, with SiO$_2$ being the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The Si-rich coating growth rates were found to be at least $0.4-0.7$ nm/s, and the erosion rate was $0.1-0.3$ nm/s. The technique is estimated to coat a surface area of at least 0.94 m$^2$ on the outer divertor. These results demonstrate the potential of using real-time material injection to grow silicon-rich layers on divertor PFCs during reactor operation

Divertor characterization experiments

Recent DIII-D experiments with enhanced Scrape-off Layer (SOL) diagnostics permit detailed characterization of the SOL and divertor plasma under various operating conditions. We observe two distinct plasma modes: attached and detached divertor plasmas. Detached plasmas are characterized by plate temperatures of only 1 to 2 eV. Simulation of detached plasmas using the UEDGE code indicate that volume recombination and charge exchange play an important role in achieving detachment. When the power delivered to the plate is reduced by enhanced radiation to the point that recycled neutrals can no longer be efficiently ionized, the plate temperature drops from around 10 eV to 1-2 eV. The low temperature region extends further off the plate as the power continues to be reduced, and charge exchange processes remove momentum, reducing the plasma flow. Volume recombination becomes important when the plasma flow is reduced sufficiently to permit recombination to compete with flow to the plate

SUSTAINED STABILIZATION OF THE RESISTIVE WALL MODE BY PLASMA ROTATION IN THE DIII-D TOKAMAK

OAK-B135 A path to sustained stable operation, at plasma pressure up to twice the ideal magnetohydrodynamic (MHD) n = 1 free-boundary pressure limit, has been discovered in the DIII-D tokamak. Tuning the correction of the intrinsic magnetic field asymmetries so as to minimize plasma rotation decay during the high beta phase and increasing the angular momentum injection, have allowed maintaining the plasma rotation above that needed for stabilization of the resistive wall mode (RWM). A new method to determine the improved magnetic field correction uses feedback to sense and minimize the resonant plasma response to the non-axisymmetric field. At twice the free-boundary pressure limit, a disruption precursor is observed, which is consistent with having reached the ''ideal wall'' pressure limit predicted by stability calculations