The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant beta phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T-e and radiation which need control. High-beta(P) scenarios with q(min) > 2, q(95)-7.9, beta(N) > 4, beta(T)-3.3% and H-98y2 > 1.5 are sustained with high density ((n) over bar = 7E19 m(-3), f(G)-1) for 6 tau(E), improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

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.

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (similar to 30 kPa) and stored energy (3.2 MJ) with H-98y2 approximate to 1.6-2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric beta(TE) independent of current between q(95) = 2.8-3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses approximate to 0 injected torque and the operating space is more ITER-relevant. Finally, the high-beta(N) (<= 3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving similar to 40% divertor heat flux reduction using either argon or neon with P-tot up to 15 MW.

Collector probes have long been used to measure impurity fluxes in the scrape off layer (SOL) of tokamaks. In this study, collector probes were inserted in the main SOL of DIII-D during the tungsten Metal Rings Campaign, and the W deposits on the probes were analyzed ex situ using Rutherford backscattering spectrometry analysis to obtain radial profiles of W deposition. A simple picture is hypothesized for how the W transports through the SOL from the target to the two sides of the probe, based on a long-theorized impurity accumulation at the crown (i.e. the top) of the plasma. The patterns observed in the deposition profiles along the probes support the hypothesized picture; however, 3D modeling and further experimental studies are needed to support more definitive conclusions.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (ne/nGW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

A cross machine analysis of heat flux profile measurements has produced a particularly simple scaling law for the scrape-off-layer (SOL) width, going as Bp,mpe with e∼−1.0 where Bp,mp is the poloidal magnetic field at the outer midplane separatrix. This result is in very good agreement with a model based on drift-induced transport. We are also developing a model that extends the kinetic ballooning mode pedestal paradigm to the separatrix and into the SOL. We have calculated the critical pressure gradient at the separatrix using the BALOO code and found that it scales proportionately with the measured pressure gradient. Using an improved high rep-rate and higher edge resolution Thomson scattering system on DIII-D, it is now possible to make detailed comparisons of SOL profile characteristics with the heat flux profile. We find that a two-point flux-limited model is in better agreement with measurements than a conduction-limited model.

A model is developed which constrains heat width, λr based on global power balance, momentum conservation, pedestal stability and sheath heat transmission. The model relies on measurements of the ratio of separatrix to pedestal pressure; a ratio ∼5% is found to be expected for ITER. Applying this model indicates a constraint that the allowed λr∼10–30mm for ITER if the divertor is in the high-recycling regime as expected (T<20eV) while a λr∼1–3mm requires a separatrix pressure approximately equal to the top pedestal pressure in violation of physical reasoning and the concept of a pedestal. A weaker constraint is applied in the model that upstream separatrix temperature simultaneously satisfies power balance. The constrained model cannot satisfy power balance with λr<3mm, and in order to obtain λr∼5mm requires divertor plasma temperature >100eV, a condition which would have very negative consequences for the divertor, but has never been observed experimentally.

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while a Spitzer resistivity model agrees at higher collisionality where it is more valid. The SOL profile measurements and divertor heat flux scaling are consistent with a heuristic drift based model as well as a critical gradient model.

A cross machine analysis of heat flux profile measurements has produced a particularly simple scaling law for the scrape-off-layer (SOL) width, going as Bp,mp e with e ~ -1.0 where Bp,mp is the poloidal magnetic field at the midplane. This result is in very good agreement with a model based on drift-induced transport. We are also developing a model that extends the kinetic ballooning mode pedestal paradigm to the separatrix and into the SOL. We have calculated the critical pressure gradient at the separatrix using the BALOO code and found that it scales proportionately with the measured pressure gradient. Using an improved high rep-rate and higher edge resolution Thomson scattering system on DIII-D, it is now possible to make detailed comparisons of scrape-off-layer (SOL) profile characteristics with the heat flux profile. We find that a two-point flux-limited model is in better agreement with measurements than a conduction-limited model.

We report new results from recent particle control and transport experiments in the DIII-D tokamak. We find that dynamic particle balance calculations (particle sources and sinks calculated vs time) yield similar results to a shot-averaged “static” (calculated by pressure rise) particle balance. The dynamic particle balance measurements show very low wall retention in both NBIand ECheated H-modes, compared with large retention during plasma startup. Two oxygen bakes at 1.3 kPa oxygen pressure for 2 hours at 350°C were completed and advanced inductive plasma operation was quickly recovered. No damage to tokamak internal components was observed. The removal of the hydrogenic species in DIII-D by thermo-oxidation compares well with laboratory experiments and we obtained data on thermo-oxidation of fresh C layers in DIII-D. Particle balance in discharges with resonant magnetic perturbation (RMP) ELM suppression show that the wall retention rate and inventory is dependent on pedestal density and divertor conditions. C injection experiments have shown that most of the carbon is deposited at the inner strike point of a SN divertor in L-mode, additional deposition in the private flux zone is present in H-mode. With an unbalanced DN plasma shape, there is more localized C deposition near the injection point in the non-active divertor.

The ITER divertor will be more collisional than divertors in existing tokamaks and the neutral modeling tools used for predictive ITER modeling need to be validated for similar divertor conditions in order to increase confidence in the results. The EIRENE neutral code has been applied to the Alcator C-Mod divertor, where collisional effects such as Lyman series photon trapping and neutral viscosity have been observed. The plasma solution used as input to EIRENE agrees well with the available diagnostic data for the divertor and was generated using an empirical Onion-Skin Method interpretive code that requires a large amount of experimental data as input. The calculated divertor pressure is 11±3 mTorr, a factor ∼2 lower than the measured value of 25±3 mTorr. The neutral solution is sensitive to photon trapping, viscosity, elastic collisions between plasma ions and molecules, and the amount of leakage through the outer divertor substructure.

Electron density, ne, and temperature, Te, in the MkIIa divertor have been measured using neutral helium line ratios. Experimental evidence indicates that the spectroscopic signal for helium injected from the vertical target is dominated by recycling. When helium is injected from the horizontal target it is possible to differentiate between injected and recycled spectral sources. To determine the spatial location of the neutral helium emission, and hence of the ne, Te measurements, an interpretative model has been developed which uses the impurity/neutral codes, DIVIMP/NIMBUS, the atomic database ADAS and the diagnostic lines of sight. Measurements and simulations have been performed using the various injection points over a range of target conditions. Comparisons with other divertor diagnostics, such as the divertor interferometer and target Langmuir probes are presented.