ITER Baseline Scenario plasmas were studied in DIII-D using krypton and xenon gases as a proxy for the tungsten that will be present in ITER. These impurities were chosen for having the same radiative loss rate Lz as tungsten would exhibit in the hotter ITER core. Results show that the scenario with these core radiators spans the range of impurity concentration and W radiated fraction expected for ITER, and up to 50% higher values, explored at zero injected torque, as well as 1 Nm and full co-torque injection with T similar to 3 Nm. Stationary discharges with duration >2-4 tau(R) are achieved with f(rad)>= 30% leading to a reduction in confinement of similar to 10%, and a comparison with real metal radiators in the same range of f(rad) shows that the higher Lz at the lower temperatures in these plasmas yields too pessimistic results on the survivability and performance of this scenario in ITER. Simulations of ITER power balance including W radiation show that with concentration up to three times higher than in the DIII-D plasmas the scenario can be stationary, remaining at acceptable core radiated fraction values.

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.

Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q 95 ∼ 3 and P IN / P LH ∼ 1–2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation ( P LH ) was determined experimentally without added impurities and found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20%–25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations C W = 10 −5 . Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at C W = 1 × 10 −5 , moreover Q = 8–10 would still be achieved for C W up to 3 × 10 −5 . In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities, indicating that hydrogen H-modes in ITER may not inform the D-T phase with respect to W accumulation and discharge survival. Compared to deuterium, the pedestal pressure in hydrogen is ∼25% lower, with much higher ELM frequency of 150 Hz, decreasing with input power. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2–3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3–0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous International Tokamak Physics Activity database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H 98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation. The new DIII-D results support a more attractive option for the ITER Research Plan with a short hydrogen phase for system commissioning, transitioning to deuterium operations as soon as possible to provide relevant conditions for deuterium-tritium operations.

The US ITER Research Program Basic Research Needs Workshop, held over the course of several months in 2022 with over 400 participants, sought to identify steps to be taken to both maximize the return of the US investment in ITER construction and operation and to ensure US research priorities on ITER strengthen the domestic program aimed at the development of a fusion pilot plant (FPP).

The path to fusion in the United States requires partnership between public and private sector. While the private sector provides the vigor to take some of the major steps necessary, there is a depth of expertise and capability in the public sector that is vital to resolving feasible approaches. As an open national user facility, DIII-D provides a crucial testbed to develop the required new technologies and approaches in relevant conditions. It has unparalleled potential to meet this challenge, thanks to its extreme flexibility and world leading diagnostics. This provides a basis to rapidly develop solutions that project to future reactors with confidence. The program has thus been redeveloped to enable public and private sector engagement and testing of new concepts. A new technology program has been launched to resolve plasma interacting technologies. With modest heating upgrades, the facility can confront the crucial "Integrated Tokamak Exhaust and Performance" gap, to resolve core, exhaust and technology solutions together. The device is also being redeveloped as a training facility, with dedicated student run time, a mentorship program, and open access to all opportunity roles, part of wider efforts to diversify and open pathways through inclusion, access, and equity. This exciting agenda is enabling scientists and technology researchers to pioneer the solutions needed for a Fusion Pilot Plant (FPP) and ITER this decade. As a national user facility, DIII D has singular potential to provide the tools, teams, and insight necessary, to do its part in moving the United States rapidly toward the commercialization of fusion energy.

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 U.S. Fusion Energy Sciences Advisory Committee was charged "to identify the most promising transformative enabling capabilities (TEC) for the U.S. to pursue that could promote efficient advance toward fusion energy, building on burning plasma science and technology." A subcommittee of U.S. technical experts was formed and received community input in the form of white papers and presentations on the charge questions. The subcommittee identified four "most promising transformative enabling capabilities": 1. advanced algorithms 2. high critical temperature superconductors 3. advanced materials and manufacturing 4. novel technologies for tritium fuel cycle control. In addition, one second-tier TEC, defined as a "promising transformative enabling capability," was identified: fast-flowing liquid-metal plasma-facing components. Each of these TECs presents a tremendous opportunity to accelerate fusion science and technology toward power production. Dedicated investment in these TECs for fusion systems is needed to capitalize on the rapid advances being made for a variety of nonfusion applications to fully realize their transformative potential for fusion energy.

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.

The Trapped Gyro-Landau Fluid (TGLF) quasi-linear model [G. M. Staebler, et al., Phys. Plasmas 12, 102508 (2005)], which is calibrated to nonlinear gyrokinetic turbulence simulations, is now able to predict the electron density, electron and ion temperatures, and ion toroidal rotation simultaneously for internal transport barrier (ITB) discharges. This is a strong validation of gyrokinetic theory of ITBs, requiring multiple instabilities responsible for transport in different channels at different scales. The mystery of transport inside the ITB is that momentum and particle transport is far above the predicted neoclassical levels in apparent contradiction with the expectation from the theory of suppression of turbulence by E×B velocity shear. The success of TGLF in predicting ITB transport is due to the inclusion of ion gyro-radius scale modes that become dominant at high E×B velocity shear and to improvements to TGLF that allow momentum transport from gyrokinetic turbulence to be faithfully modeled.

Author(s): Van Zeeland, MA; Heidbrink, WW; Park, JM; Prater, R; Holcomb, CT; Austin, ME; Ferron, JR; Greenfield, CM; Grierson, BA; Hong, RM; Luce, TC; McKee, GR; Moyer, RA; Murakami, M; Murphy, CJ; Muscatello, CM; Pace, DC; Petty, CC; Rauch, J; Scoville, JT; Solomon, WM; Tobias, BJ

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L–H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δ v / v ∼ 60% via non-resonant braking. Changes to global Δ n / n , Δβ/β and ΔH 98 /H 98 were ∼3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

DIII-D is making significant contributions to a scientific basis for sustained burning plasma operation. These include explorations of increasingly reactor-relevant scenarios, studies of key issues for projecting performance, development of techniques for handling heat and particle efflux, and assessment of key issues for the ITER research plan. Advanced scenarios are being optimized in DIII-D via experiments to empirically determine the relationship between transport and the current profile, which in turn can provide essential input to inform improvement of the theory-based models that do not currently capture the observed behaviour. Joint DIII-D/JET ρ* scans in the hybrid regime imply Bohm-like confinement scaling. Startup and shutdown techniques were developed for the restrictive environment of future devices while retaining compatibility with advanced scenarios. Towards the goal of a fully predictive capability, the DIII-D program emphasizes validation of physics-based models, facilitated by a number of new and upgraded diagnostics. Specific areas include transport, rotation, energetic particles and the H-mode pedestal, but this approach permeates the entire research programme. Concerns for heat and particle efflux in future devices are addressed through studies of ELM control, disruption avoidance and mitigation, and hydrogenic retention in DIII-D's carbon wall. DIII-D continues to respond to specific needs for ITER. Recent studies have compared H-mode access in several different ion species, identifying not only isotopic, but density, rotation and geometrical dependences that may guide access to H-mode during ITER's non-activated early operation. DIII-D used an insertable module to simulate the magnetic perturbations introduced by one of ITER's three test blanket module sets, demonstrating that little impact on performance is seen at ITER equivalent levels of magnetic perturbation.

Important progress has been made in the correction of 3D fields, based on the improved understanding of plasma response using the Ideal Perturbed Equilibrium Code (IPEC). The key to error field correction is to reduce the dominant distribution of 3D fields that is stronger often by an order of magnitude than any other distribution in breaking magnetic surfaces. The important validation is achieved in presently the most extreme case, the DIII-D mock-up experiments for the ITER Test Blanket Modules (TBMs). Although the TBM 3D fields are highly localized and cannot be controlled by typical error field correction coils, the optimal operation could be restored using I-coils by minimizing the dominant part in the TBM 3D fields as IPEC prediction. Including TBM experiments, various error field correction results in tokamaks such as NSTX, DIIID, and CMOD, have been successfully understood and quantified based on the dominant external fields, as can be summarized in the robust parametric scaling of the locking threshold. The implications are favourable for ITER, since the highly reliable 3D field compensation can be provided for a wide range of different plasmas if the correction coil is designed based on the robust patterns of the dominant external fields. Present Error Field Correction Coil (EFCC) in ITER is under active investigations using IPEC to assess their capabilities.

The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization of fusion reactors and enable development of fusion's energy applications. A concept of FDF based on the tokamak approach with conservative expressions of advanced physics and nonsuper-conducting magnet technology is presented. It is envisioned to nominally provide 2 MW/m(2) of neutron wall loading and operate continuously for up to 2 weeks as required for fusion nuclear component research and development. FDF will have tritium breeding capability with a goal of addressing the tritium self-sufficiency issue for fusion energy. A zero-dimensional system study using extrapolations of current physics and technology is used to optimize FDF for reasonable power consumption and moderate size. It projects a device that is between the DIII-D tokamak (major radius 1.8 m) [J. L. Luxon, Nucl. Fusion, Vol. 42, p. 614 (2002)] and the Joint European Torus (major radius 3 m) [P. H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion, Vol. 25, p. 1011 (1985)] in size, with an aspect ratio A of 3.5 and a fusion gain Q of 2 to 5. Theory-based stability and transport modeling is used to complement the system study and to address physics issues related to specific design points. It is demonstrated that the FDF magnetohydrodynamic stability limits can be readily met with conservative stabilizing conducting wall placement. Transport analysis using a drift-wave-based model with an edge boundary condition consistent with the pedestal stability limit indicates that the FDF confinement requirement can also be readily satisfied. A surprising finding is that the toroidal Alfven eigenmodes are stabilized by strong ion Landau damping. Analysis of vertical stability control indicates that the basis configuration with an elongation kappa(X) similar to 2.35 can be controlled using a power supply technology similar to that used in DIII-D. Peak heat fluxes to the divertor are somewhat lower than those of ITER (R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Control. Fusion, Vol. 44, p. 519 (2002)), but FDF will operate with a higher duty factor.

by J.A. SNIPES, M.J. SCHAFFER, P. GOHIL, P. de VRIES, M.E. FENSTERMACHER, T.E. EVANS, X. GAO, A.M. GAROFALO, D.A. GATES, C.M. GREENFIELD, W.W. HEIDBRINK, G.J. KRAMER, S. LIU, A. LOARTE, M.F.F. NAVE, N. OYAMA, J-K. PARK, N. RAMASBRAMANIAN, H. REIMERDES, G. SAIBENE, A. SALMI, K. SHINOHARA, D.A. SPONG, W.M. SOLOMON, T. TALA, J.A. BOEDO, R. BUDNY, V. CHUYANOV, E.J. DOYLE, M. JAKUBOWSKI, H. JHANG, R.M. NAZIKIAN, V.D. PUSTOVITOV, O. SCHMITZ, T.H. OSBORNE, R. SRINIVASAN, T.S. TAYLOR, M.R. WADE, K-I. YOU, L. ZENG, and the DIII-D TEAM

High performance plasmas and daily reference shots (DRSs) with both L-mode and H-mode phases were used to demonstrate the maintenance of good wall conditions over similar to 7000 s of plasma operation in DIII-D with no intervening boronizations or high temperature bakes during each of the 2006 and 2007 campaigns. High performance discharges with high normalized beta and confinement factor and good density control over the duration of the high-power beam injection period were very repeatable over the course of these campaigns. High performance operation was also demonstrated after a six week entry vent followed by the standard high temperature bake at 350 degrees C and plasma conditioning, but prior to a boronization. Over the 2006 and 2007 campaigns, the DRS database indicated little to no secular increase in impurity content. Oxygen content and nickel line emission were higher after the entry vent, but were still minor contributors to plasma contamination compared with carbon. Because DIII-D has a plasma facing surface that is >95% graphite, we take this as a demonstration that erosion of boronization films used for wall conditioning will not be a limitation to establishing long-pulse high performance discharges in the new generation of superconducting tokamaks if graphite is used as the primary plasma facing material.