The small-angle slot (SAS) divertor of the DIII-D tokamak, and its upcoming upgrade, the V-shape small-angle slot (SAS-V) divertor, are numerically investigated using the SOLPS-ITER code package, including the effect of particle drifts, for a range of plasma density, heating power, strike point position in the slot, and for both magnetic field directions. The simulations show that the electron temperature near the strike point is reduced in SAS-V compared to SAS, for both magnetic field directions, such that SAS-V achieves divertor detachment at a lower value of the outboard mid-plane separatrix electron density. The detachment threshold is lower because the V-shape focuses recycling neutrals on the V-end, densifying and cooling the plasma in the slot. At sufficiently high density, the V-shape also reduces the radial gradient of the temperature profile at the target, which in turns reduces the radial electric field and the E × B drift velocities, further densifying and cooling the plasma in the slot and leading to detachment. The V-shape effect, however, is reduced for higher heating power. With more heating power, the detachment density increases, reducing the ionization mean free path of recycled neutrals, which therefore become less sensitive to target shape changes. This suggests that in a fusion reactor, where the heating power is high, optimization of the divertor target shape needs to be combined with other strategies to lower the detachment density, such as in-slot injection of low-Z impurities.

It is estimated that pilot plants and reactors will experience rates of net erosion and deposition of solid PFC, Plasma Facing Component, material of 103 – 105 kg/year. Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘UFOs’ and unsafe dust levels. The potential implications appear to be as serious as for plasma contact with the divertor target: a dust explosion or a major UFO-disruption could be as damaging for an actively-cooled DT tokamak as target failure. It will therefore be necessary to manage material deposits to prevent their fouling operation. This situation appears to require a fundamental paradigm shift with regard to meeting the challenge of taming the plasma-material interface: it appears that any acceptable solid PFC material will in effect be flow-through, like liquid-metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable like brake pads in cars. It will be as essential to be able to in situ remove from the vessel the deposited PFC material as it will be to introduce new material to the vessel for refurbishing the claddings of the PFCs. ITER will use high-Z (tungsten) armor on the divertor targets and low-Z (beryllium) on the main walls. The ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding of the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc.) used as in situ replenishable, relatively thin - of order mm - claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing risk of degrading the confined plasma. Separately, wall conditioning has proven essential for achieving high performance [1], even more so today with the change from C to W PFCs in many tokamaks. For DT devices, however, standard methods appear to be unworkable, but recently powder droppers injecting low-Z material ~continuously into discharges have been quite effective [2-5] and may be usable in DT devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current tokamaks, independent of whether their protection tiles are low-Z or high-Z.

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10-15 µm of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a "V" shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned "V" shape predict a substantial reduction in W sourcing and SOL accumulation in either B×∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. This coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

The advancement of fusion reactor engineering is currently inhibited by the lack of knowledge surrounding the stability of plasma facing components (PFCs) in a tokamak environment. During normal operation, events of high heat loading occur periodically where large amounts of energy are imparted onto the PFC surface. Concurrently, irradiation by low-energy helium ions present in the fusion plasma can result in the synthesis of a fibre form nanostructure on the PFC surface, called ‘fuzz’. In order to understand how this heterogeneous structure evolves and deforms in response to transient heat loading, a pulsed Nd:YAG millisecond laser is used to simulate these events on a fuzz form molybdenum (Mo) surface. Performance was analysed by three metrics: nanostructure evolution, particle emission, and improvement in optical properties. Experiments performed at the upper end of the expected range for type-I edge-localized modes (ELMs) found that the helium-induced nanostructure completely disappears after 200 pulses of the laser at 1.5 MJ m−2. In situ mass loss measurements found that the amount of particles leaving the surface increases as energy density increases and the rate of emission increases with pulse count. Finally, optical properties assisted in providing a qualitative indication of fuzz density on the Mo surface; after 400 pulses at 1.5 MJ m−2, the optical reflectivity of the damaged surface is ~90% of that of a mirror polished Mo sample. These findings provide different results than previous studies done with tungsten (W), and further help illustrate the complicated nature of how transient events of high heat loading in a tokamak environment might impact the performance and lifetime of PFCs in ITER and future DEMO devices (Ueda et al 2014 Fusion Eng. Des. 89 901–6).

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.