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

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

Current ramps in tokamaks: from present experiments to ITER scenarios

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H(96-L) = 0.6 or H(IPB98) = 0.4) has been validated on a multi-machine experimental dataset for predicting the l(i) dynamics within +/- 0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than +/- 0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of I(p) = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down.</p

RESEARCH PROGRESS AND HARDWARE SYSTEMS AT DIII-D

OAK-B135 During the last two years significant progress has been made in the scientific understanding of DIII-D plasmas. Much of this progress has been enabled by the addition of new hardware systems. The electron cyclotron (EC) system has been upgraded from 3 MW to 6 MW, by adding three 1 MW gyrotrons with diamond windows and three steerable launchers (PPPL). The new gyrotrons have been tested to 1.0 MW for 5 s. The system has been used to control the 3/2 and 2/1 neoclassical tearing modes and to locally heat the plasma and thereby indirectly control the current density. Electron cyclotron current drive ECCD has been used to directly affect the current density. A Li-beam diagnostic has been brought on-line for measuring the edge current density using Zeeman splitting. A set of 12 coils (1-coils), consisting of six picture frame coils each above and below the midplane, with a capability of 7 kA for 10 s has been installed inside the DIII-D vessel. These coils, along with the existing six C-coils, are used to apply non-axisymmetric fields to the plasma for both exciting and controlling plasma instabilities. The DIII-D digital plasma control system is now used to not just control the shape and location of the plasma but also the electron temperature, density, the NTMs, RWMs, plasma beta and disruption mitigation. Plasma disruption experiments are extended to mitigation of real time detected disruptions on DIII-D