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DIII-D experiments have achieved promising core–edge integrated plasma scenarios which combine a high-temperature low-collisionality pedestal (pedestal top temperature T _e _,ped > 0.8 keV and collisionality ν * _ped < 1) with a partially detached divertor by leveraging the benefits of a low-density-gradient pedestal in a closed divertor. It is found that with a closed divertor and high heating power, strong gas puffing to achieve detachment moves the peak density gradient outward with respect to the maximum gradient of electron temperature and reduces the density gradient at the pedestal region, which correlates with shallow pedestal fuelling due to the closed divertor geometry. In high-current plasmas in particular, the pedestal top density is found to change little with gas puffing while the separatrix, density increases to allow access for divertor detachment. The separation between density and temperature pedestals results in a high- η _e well above the electron-temperature-gradient stability threshold. Electron turbulence is found to be enhanced in the pedestal and correlated with high η _e resulting from the pedestal shift. The pedestal is wider than the EPED scaling. A revised empirical width scaling is derived based on the combination of EPED scaling with η _e and highlights the important role of additional turbulence on the pedestal structure. The wide temperature pedestal facilitates the achievement of a high-temperature, low-collisionality pedestal and high global performance. Simultaneously, the outward shift of the density pedestal facilitates access to detached divertor conditions with low temperature and heat flux towards the target plate. This approach may be promising for closing the core–edge integration gap for future fusion reactors, which may have a weak-gradient density pedestal due to the highly opaque boundary plasmas.

The transition from small edge-localized modes (ELMs) to large ELMs has been repetitively observed in minute-scale long-pulse high-confinement mode (H-mode) discharges during the 2017 EAST campaign. The appearance of large ELMs is found to be strongly correlated with the decrease in separatrix density due to the gradual decrease in fuel recycling during long-pulse H-mode operations (LPHOs). By the numerical scanning of separatrix density with a fixed temperature profile, it has been found that the dependence of ELM instability on separatrix density is related to the competition between the ion diamagnetic stabilizing effect and destabilizing effect of pressure gradient and current density in the pedestal region. This sheds light on a comprehensive understanding of the different roles of separatrix density in ELM instability observed during EAST experiments. With a high separatrix density, the ideal ballooning mode can be destabilized near the separatrix, which is thought to help achieve small ELMs in EAST LPHOs. During the 2021 EAST campaign, an experiment of large ELM control was performed through actively changing fuel recycling by moving the strike point location on the lower tungsten divertor target plate. It has been demonstrated that the mitigation of large ELMs is strongly correlated with the significant increase in separatrix density, which is thought to be attributed to a higher ionization source in the scrape-off layer (SOL) region by SOLPS-ITER simulation. The high ionization source in the SOL region is believed to provide a strong fueling effect near the separatrix and thus raise the local density, which is considered an important reason for triggering ballooning instabilities near the separatrix and achieving small ELMs.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β _N ∼ 4.2, toroidal beta β _T ∼ 3.3% with q _min > 2, q _95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H _98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f _Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β _N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

Integration of transient and steady-state divertor heat fluxes control with a high-performance core is necessary for future fusion reactors. In recent DIII-D high- ${\beta _{\text{p}}}$ experiments, divertor detachment and simultaneous edge localized mode (ELM) suppression are demonstrated while the plasma confinement quality is maintained high in ITER-similar shape. By optimizing the neon injection in high- ${\beta _{\text{p}}}$ scenario with ITER-similar shape, deep detachment and ELM suppression are achieved with a high-performance core ( ${\beta _{\text{N}}}$ ∼ 2.8, ${\beta _{\text{p}}}$ ∼ 2.3) at ${q_{95}}$ ∼ 7.5. Partial divertor detachment and suppression of large ELMs are achieved at ${q_{95}}$ ∼ 6. The stability analyses suggest that with low neon injection, the density pedestal becomes higher and steeper and the ${T_{\text{i}}}$ profile also increases, therefore the increased edge pressure and higher current density destabilize the Peeling-Ballooning mode (PBM), which would lead to a large ELM collapse. With strong neon gas puffing, the significantly reduced pedestal pressure and current density, due to the degraded ${T_{\text{e}}}$ pedestal, lead to the stabilization of PBM and ELMs are suppressed. For both cases, the coupling between the large radius internal transport barrier (ITB) and edge pedestal is the key reason for maintaining high global performance. The formation of large radius ITB compensates for pedestal degradation. Such results could provide an attractive scenario to well control the transient and steady-state heat flux onto the divertor plates while maintaining good plasma performance, which is an important step toward the steady-state operation of future fusion reactors.

Experiments performed during strongly-shaped high-power diverted negative triangularity (NT) experiments in DIII-D achieved detached divertor conditions and a transient-free edge, showcasing the potential for application of NT to a core-edge integrated reactor-like scenario and providing the first characterization of the parametric dependencies for detachment onset. Detached divertor conditions will be required in future devices to mitigate divertor heat fluxes. Access to dissipative divertor conditions was investigated via an increase in upstream density. Detachment onset at the outer strike point was achieved with H-mode level confinement $H_{98-y2}\sim 1$ and reactor-relevant normalized pressures $\beta_\mathrm N \sim 2$ . Confinement degradation was observed with deeper detachment, associated with the loss of an electron temperature pedestal. Differences in geometry, radial transport, impact of cross field drifts are discussed to explain differences in access to detachment in NT discharges. Higher normalized densities, with respect to equivalent discharges in positive triangularity, were necessary to achieve detachment, partially explained by the shorter parallel connection length to the targets. The effect of cross-field particle drifts ( E × B , B $\times \nabla$ B ) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B $\times \nabla$ B drift directed outside of the active divertor (Greenwald fraction $f_{\mathrm{Gw}}\sim$ 0.9–1.0 vs $f_{\mathrm{Gw}}\sim$ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B $\times \nabla$ B drift into the divertor, consistent with the observed narrowing of the scrape-off layer heat flux width λ _q . Edge fluid simulations capture separatrix densities needed to achieve detachment in NT plasma and their dependence on drift direction. The ability to reproduce detachment dynamics in NT plasma increases the confidence in future design studies for NT divertors.

Experiments in DIII-D demonstrate that the upstream plasma density to detach an un-pumped slot divertor is similar for a V-shaped and a flat-end slot, despite significantly higher neutral pressure in the V-shaped slot and in contrast to SOLPS-ITER predictions. The detachment threshold can be reduced by using in-slot instead of main-chamber gas fuelling or by placing the strike point on the inner slanted slot baffle instead of the slot end, as described by simulations with full drift physics. When increasing the plasma line-averaged density (without extrinsic impurities), the transition to detachment in DIII-D slot divertor is sharp and requires a high value of plasma density with the ion $\overrightarrow {\mathbf{B}} {{ \times }}\nabla B$ drift into the slot, whereas it is smooth and requires a lower value of plasma density with the opposite drift direction, in accord with detachment experiments in the DIII-D open lower divertor. Unique experiments on DIII-D and comparison to advanced simulations expand the scientific understanding of slot-shaped divertors, considered highly desirable for next step fusion devices.