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 ∼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 β 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.
The ability to measure rf driven waves in the edge of the plasma can help to elucidate the role that surface waves and parametric decay instabilities (PDIs) play in rf power losses on NSTX. A microwave reflectometer has recently been modified to monitor rf plasma waves in the scrape-off layer in front of the 30MHz high harmonic fast wave antenna array on NSTX. In rf heated plasmas, the plasma-reflected microwave signal exhibits 30MHz sidebands, due primarily to the modulation of the cutoff layer by the electrostatic component of the heating wave. Similarly, electrostatic parametric decay waves (when present) are detected at frequencies below the heating frequency, near 28, 26,…MHz, separated from the heating frequency by harmonics of the local ion cyclotron frequency of about 2MHz. In addition, a corresponding frequency matched set of decay waves is also detected near the ion cyclotron harmonics at 2, 4,…MHz. The rf plasma-wave sensing capability is useful for determination of the PDI power threshold as a function of antenna array phasing (including toroidal wavelength), outer gap spacing, and various plasma parameters such as the magnetic field and the plasma current.
An upgrade to the National Spherical Torus Experiment (NSTX) is currently in progress. The NSTX Center Stack Upgrade (NSTX-U) experimental device has an operating space that is both larger and more complex than that of the original NSTX. The mechanical integrity of some machine components can be compromised both by the instantaneous values of combinations of magnet currents and as a result of the time histories thereof. An upgrade to the existing protection systems and methodology is required to allow for both safe and effective use of the expanded operating space. A Digital Coil Protection System (DCPS) is planned as a major component of an upgraded Coil Protection System (CPS).
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