The SPARC tokamak is a high-field, B-t0 similar to 12 T, medium-sized, R-0 = 1.85 m, tokamak that is presently under construction in Devens, MA, led by Commonwealth Fusion Systems. It will be used to de-risk the high-field tokamak path to a fusion power plant and demonstrate the commercial viability of fusion energy. SPARC's first campaign plan is to achieve Q(fus) > 1 using an ICRF-heated, <10 MW, high current, I-p similar to 8.5 MA, L-mode fueled by D-T gas injection, and its second campaign will investigate H-mode operations in D-D. To facilitate plasma control and scientific learning, a targeted set of similar to 50 plasma diagnostics are being designed and built for operation during these campaigns. While nearly all diagnostics are based on established techniques, the pace of deployment, relative to the first plasma, and the harshness of the thermal, electromagnetic, and radiation environment are unprecedented for medium-sized tokamaks. An overview of the SPARC diagnostic set is given, providing context to further details communicated by the SPARC team in companion publications that are system-specific. The system engineering philosophy for SPARC diagnostics is outlined, and the design and engineering verification process for components inside and outside the primary vacuum boundary are described. Diagnostics are mounted directly to the vacuum vessel as well as housed within a series of eight midplane and 24 off-midplane replaceable port plugs. With limited exceptions, signal conditioning, digitization electronics and cameras as well as lasers and microwave sources are localized to a series of five Diagnostic Lab spaces, totaling similar to 350 m(2), located >15 m from the center of the tokamak, on the other side of a 2.4 m concrete shielding wall. A series of 31 large-scale penetrations have been included in the SPARC Tokamak Hall to facilitate integration of early campaign diagnostics and to provide upgradability.

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 DIII-D small angle slot (SAS) divertor is designed for divertor physics studies with enhanced neutral confinement and special target geometries in a closed divertor. The closed nature of the SAS makes optical diagnostic measurements difficult, so a specially designed, multipurpose array of Langmuir probes has been implemented to study the plasma conditions in and around the slot. The probes are spaced to provide at least 2 mm resolution (shorter than the energy decay length) of the near scrape-off layer when mapped to the outer mid-plane. Due to space limitations at the bottom of the slot, a novel spring-loaded probe and tile design was developed to clamp several short rooftop probe tips and insulators to the cooled baseplate. Initial probe measurements revealed tile to tile edge shadowing, especially where magnetic field line surface angles were less than 1°. Additionally, it was found, using three Langmuir probes (at 90°, 180°, and 270°), that the strike point variation of ±5 mm radially around the torus was not well aligned with the circular slot geometry [Watkins et al., Nucl. Mater. Energy 18, 46 (2019)]. These issues were resolved by (1) designing tiles with all probes mounted near the tile center instead of near the edges and (2) aligning these new custom tiles to the measured strike point toroidal surface with a very accurate laser scanning alignment tool. Post-alignment Langmuir probe measurements and plasma behavior demonstrated close agreement at two separate toroidal locations that were 45° apart.

A. Bortolon 1 , E. Gilson 1 , R. Lunsford 1 , R. Maingi 1 , A. Nagy 1 , C. Chrobak 2 , A. Hyatt 2 , T. Wilks 3 , M. Fenstermacher 4 , T. Rognlien 4 , C. Samuell 4 , M. Umansky 4 , J. Boedo 5 , I. Bykov 5 , D. Rudakov 5 , R. Smirnov 5 , M. Shafer 6 , D. Curreli 7 , J. Drobny 7 , D. Donovan 8 , J. Duran 8 , J. Ren 8 , A. Drenik 9 , R. Dux 9 , A. Herrmann 9 , A. Kallenbach 9 , R. McDermott 9 , R. Neu 9 , V. Rohde 9 , E. Wolfrum 9 and the DIII-D and ASDEX-Upgrade teams

Nanostructure cone growth on chromium during helium plasma exposure in the PISCES-A linear plasma device is found to increase with the deposition of heavy tantalum atoms. Tantalum is found to be especially effective at increasing cone growth if the tantalum is incident on the surface at a grazing angle (∼5° or less). Tantalum atoms at grazing incidence are observed to be dominantly deposited at the cone tips, consistent with reduced sputtering of the cone tips and increased cone growth. Monte-Carlo simulations are found to be qualitatively consistent with the experimental observation that grazing-incidence heavy atom deposition on a lighter substrate will tend to increase cone growth.

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.

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as the leading candidate. However, tokamak experiments with tungsten divertors have seen significant detrimental effects on plasma performance. The DIII-D tokamak presently has carbon as the plasma facing surface but to study the effect of tungsten on the plasma and its migration around the vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Z metal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated two week experimental campaign was run with the high-Z metal inserts. One row was coated with tungsten containing naturally occurring levels of isotopes. The second row was coated with tungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than 90%. The different isotopic concentrations enabled the experiment to differentiate between the two different sources of metal migration from the divertor. Various coating methods were explored for the deposition of the tungsten coating, including chemical vapor deposition, electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatings were tested to see if they were robust enough to act as a divertor target for the experiment. Tests included cyclic thermal heating using a high power laser and high-fluence deuterium plasma bombardment. The issues associate with the design of the inserts (tile installation, thermal stress, arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used to select the coating method and preliminary experimental observations are presented.

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Diver tor Material Evaluation System (DiMES) is presented. The EWES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices. (C) 2017 Elsevier B.V. All rights reserved.

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.

Impurity transport in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] is investigated in stationary high confinement (H-mode) regimes without edge localized modes (ELMs). In plasmas maintained by resonant magnetic perturbation (RMP), ELM-suppression, and QH-mode, the confinement time of fluorine (Z = 9) is equivalent to that in ELMing discharges with 40 Hz ELMs. For selected discharges with impurity injection, the impurity particle confinement time compared to the energy confinement time is in the range of tau(p)/tau(e) approximate to 2 - 3. In QH-mode operation, the impurity confinement time is shown to be smaller for intense, coherent magnetic, and density fluctuations of the edge harmonic oscillation than weaker fluctuations. Transport coefficients are derived from the time evolution of the impurity density profile and compared to neoclassical and turbulent transport models NEO and TGLF. Neoclassical transport of fluorine is found to be small compared to the experimental values. In the ELMing and RMP ELM-suppressed plasma, the impurity transport is affected by the presence of tearing modes. For radii larger than the mode radius, the TGLF diffusion coefficient is smaller than the experimental value by a factor of 2-3, while the convective velocity is within error estimates. Low levels of diffusion are observed for radii smaller than the tearing mode radius. In the QH-mode plasma investigated, the TGLF diffusion coefficient is higher inside of rho = 0.4 and lower outside of 0.4 than the experiment, and the TGLF convective velocity is more negative by a factor of approximately 1.7. (C) 2015 AIP Publishing LLC.

Aluminum (Al) is a convenient proxy for beryllium (Be) plasma material interaction studies since they have a number of physical and chemical similarities. Al samples were exposed at the lower outer strike point of an L-mode divertor plasma in DIII-D (conditions 7–11 × 1018 D-ions cm−2 s−1, Te = 12–47 eV). The gross erosion rate was directly measured using post-mortem ion beam analysis of small 1 mm-sized samples where local re-deposition was determined to be negligible. The gross erosion rate was also calculated using spectroscopic methods, but these rates greatly underestimate the direct (i.e. non-spectroscopic) measurement. The direct measured erosion yields were within the range of published D+ → Al ion beam sputtering yields. The ionizations per photon (S/XB) coefficients used in the spectroscopic analysis were determined in separate experiments using He plasmas at the PISCES-B linear plasma facility at UCSD. The measured S/XB coefficients were on average ∼6× higher than the theoretically calculated values.