27 research outputs found
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 (1 MA, 2 T,
6 MW, m) 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 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 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 m
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 ( 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 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
nm/s, and the erosion rate was nm/s. The technique is
estimated to coat a surface area of at least 0.94 m 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
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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
Scrape-off layer ion acceleration during fast wave injection in the DIII-D tokamak
Fast wave injection is employed on the DIII-D tokamak as a current drive and electron heating method. Bursts of energetic ions with energy E o>20keV are observed immediately following fast wave injection in experiments featuring the 8th ion cyclotron harmonic near the antenna. Using the energy and pitch angle of the energetic ion burst as measured by a fast-ion loss detector, it is possible to trace the origin of these ions to a particular antenna. The ion trajectories exist entirely within the scrape-off layer. These observations are consistent with the presence of parametric decay instabilities near the antenna strap. It is suggested that the phase space capabilities of the loss detector diagnostic can improve studies of wave injection coupling and efficiency in tokamaks by directly measuring the effects of parametric decay thresholds. © 2012 IAEA, Vienna
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
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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