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 PEGASUS-III experiment is a solenoid-free, low aspect ratio spherical tokamak that will serve as a dedicated U.S. platform for comparative nonsolenoidal tokamak plasma startup studies. Approximately 175 megavolt-ampere (MVA) of reconfigured and expanded programmable power systems, 7 MJ of new stored energy, and new digital control and protection systems for the facility are being commissioned to support PEGASUS-III upgrades. These include: increased toroidal field (0.15-0.6 T); new divertor and poloidal field coils; increased pulse length; local and coaxial helicity injectors for solenoid-free plasma initiation; radio frequency (RF) systems for heating and current drive; and a diagnostic neutral beam (DNB). A new realtime digital control system implements 16 proportional-integral differential (PID) feedback controllers with 25 kHz loop rates to control the electromagnets and helicity injectors. The poloidal field coils, helicity injector arc currents, and toroidal field are driven by 36 3.6 MVA (4 kA, 900 V) insulated-gate bipolar transistor (IGBT) buck converters. Helicity injector bias voltage and current will be provided by a set of four 10.8 MVA multilevel buck converters (MLBCs). Each is comprised of an 1800 V integrated gate-commutated thyristor (IGCT) stage and a ±900 V IGBT stage in series, providing controllable I inj ≤ 4 kA at V inj ≤ 2.7 kV. A field programmable gate array (FPGA)-based digital fault protection system multiplexes controller commands to individual power semiconductors in these supplies, monitors their operational status, and executes shutdown sequences within 10 µs of fault detection. An 80 kV, 4 A zero-voltage-switching (ZVS) resonant converter with <1% output ripple is under development for the DNB and is being evaluated as a topology to drive RF sources.

Initiating current without using magnetic induction from a central solenoid is a critical scientific and technical challenge facing the spherical tokamak (ST). One such technique that has shown promise on several devices is coaxial helicity injection (CHI). In CHI, a dc voltage applied to large area coaxial electrodes injects current and helicity into the vacuum vessel for plasma startup and plasma current (I p ) sustainment. Major outstanding issues for CHI include eliminating the need for a vacuum vessel break; the scaling of I p with injector and/or flux footprint shape and separation; mitigating plasma material interaction (PMI) and minimizing impurity injection; and the degree of axisymmetry required to achieve high I p . Thus, a first of its kind, CHI system is being installed on PEGASUS-III. It utilizes two coaxial, segmented, floating electrodes located entirely within the vacuum vessel in the upper divertor region. The design enables I p scaling studies as the electrode shape, coupled with a new 480 kA/288 kA/244 kA-turn divertor coil triplet, and allows for variation of the flux footprint shape and location simply through manipulation of coil currents. Together, they are projected to allow over 50 mWb connected flux and I p > 300 kA. The segmented electrodes facilitate simple changes to their shape, position, and plasma-facing material. This flexibility may be critical for mitigating PMI or impurity sourcing from the electrodes. Independent current feeds to each segment enable tests of the impact of axisymmetric drive on I p .

No abstract

No abstract