Qiming Hu

Recent DIII-D experiments have demonstrated the compatibility of complete divertor detachment with a high-beta core using the high beta hybrid scenario plasmas with an ITER-similar shape and nitrogen impurity seeding. With complete divertor detachment (near-zero divertor particle flux and temperature), the radiation peaks inside the separatrix, near the X -point indicating the achievement of the X -Point Radiator (XPR) regime, which is a highly dissipative divertor operation scenario that may be attractive for future reactors. SOLPS-ITER modelling with full drifts is able to qualitatively reproduce the experimentally measured boundary plasma conditions and radiation patterns from attached to XPR detached divertor states for these high-beta hybrid plasmas. Experiments found that when the radiation peak is inside but close to the X -point, as noted ‘shallow XPR’, complete divertor detachment and a high-beta, high-confinement core ( β _N ∼ 3.0, H ₉₈ ∼ 1.25) could be simultaneously achieved. However, this plasma remains ELMing with giant ELMs ( W / W ∼ 3% and 4%). With stronger N ₂ impurity injection, the plasma enters a ‘deep XPR’ regime where the radiation peak is close to the inboard side of the pedestal and core radiation is about a factor of 2 higher. With deep XPR, the ELMs are strongly mitigated. However, the confinement is significantly reduced to H ₉₈ < 1.0, and the β _N is reduced to ∼2.2, while the power is comparable to the attached divertor discharges, but the feedback controlled power is reduced during the discharges which may lead to the loss of the advanced core scenario. The reduced global performance is attributed to the 50% lower pedestal pressure and 30% colder pedestal temperature. SOLPS-ITER simulations highlight the key role of impurity radiation, neutrals and divertor closure in the formation of an XPR and the effects of drifts on the distribution of plasma and radiation near the X -point, all of which are important for the divertor design and operation in future reactors. In particular, SOLPS-ITER simulations predict that in an open divertor configuration without baffling structures, with XPR operation, complete detachment is not achieved at the target plate. The modelling also exhibits qualitatively good agreement with experimental observations on the pedestal performance responding to the radiation dynamics, which provides physics insight on core-edge integration between divertor dissipation and a high-performance core that is critically important for future tokamak fusion reactors.

Recent DIII-D tokamak experiments have demonstrated the integration of intrinsic low-collisionality grassy edge-localized mode (ELM) regime with high-performance hybrid core scenario, possibly offering a core-edge compatible solution for ITER and future fusion reactors. This regime features grassy ELMs (with ELM energy loss over pedestal stored energy ν e ∗ ∼ 0.1), ITER similar shape, while maintaining high core confinement ( H 98 y 2 ∼ 1.5) in non-inductive hybrid scenarios. A small ELM-focused database is constructed to investigate the parametric dependence of the small/grassy ELM regimes. Access and sustainment of this regime appear to be favored by high poloidal beta ( β p > 1.5), higher ratio of separatrix-to-pedestal density ( n e,sep / n e,ped > 0.4) and low pedestal top collisionality ( ν e,ped ∗ = 0.1–0.4). In this scenario, the pedestal width exceeds the prediction from EPED–KBM scaling, consistent with the expectation of a turbulence-limited pedestal. Linear modeling using ELITE indicates that this grassy ELM regime is along the peeling boundary. Relative to a large-ELM phase, the grassy-ELM phase exhibits a broader inner-target heat-flux width and a substantially reduced ELM-induced transient heat-flux increment. These results motivate further evaluation of low-collisionality grassy ELMs as a potentially reactor-relevant operating regime, while full-duration sustainment and compatibility with divertor detachment remain unresolved.

Experiments on DIII-D have demonstrated a density-confinement synergy that enables sustainment of high performance in a previously unattained parameter regime of simultaneous very high energy confinement quality (H _98y2 ⩾ 1.5), very high line-average density Greenwald fraction ( ƒ _Gr = πa ² < n >/ I _P ⩾ 1.4), and high toroidal beta ( β T ⩾ 3%). Tokamak operation in this regime is essential for a compact steady-state FPP, as well as for Q = 10 with 500 MW of fusion power in ITER at I _P ≪ 15 MA. These experiments leveraged the knowledge that, in the high-poloidal-beta ( β P ) regime, density gradients can enhance turbulence stabilization caused by high α MHD ( α MHD ∼ d β P / d r ). This was described by theoretical predictions and gyrokinetic transport simulations (Kotschenreuther et al 2024 Nucl. Fusion 64 076033), and later confirmed by experiments on DIII-D (Ding et al 2024 Nature 629 555). To increase both β P and β T , the new experiments increased the ideal-wall stability β N -limit by using a smaller plasma-outer wall distance and higher triangularity in the plasma cross section (top/bottom average δ ∼ 0.9), enabled by the recent ‘shape & volume rise’ (SVR) modification to the DIII-D divertor. The higher triangularity also contributed to achieving higher ƒ _Gr by enabling higher pedestal density. At high density, the pedestal is ballooning limited and exhibits small and frequent ELMs, while the divertor is near detachment even without any impurity seeding. Highplasma performance was attained and sustained reproducibly, with the eventual terminations brought about by an MHD mode destabilized as the current profile slowly continued to evolve. A path to stationary fully noninductive operation might include ECH injection to reduce both core impurity accumulation and the electron collisionality, thus increasing the bootstrap current. These experiments provide the first experimental demonstration of the ƒ _Gr , H _{98 y 2} , and β T values required simultaneously for ITER Q = 10 at I _P < 10 MA, pointing to practical ways to improve the energy confinement in a fusion reactor.

High-pressure, peeling-limited pedestals with pedestal normalized beta β _N,ped > 2 and pedestal top density n _ped at or above the Greenwald density n _G have been achieved in DIII-D high poloidal beta plasmas, together with high global normalized beta β _N > 3 and energy confinement H ₉₈ ∼ 1.2–1.7. A higher β _N allows higher pedestal density above the Greenwald value and higher pedestal pressure, even with low injected torque. MHD modeling confirms that the experimental profiles lie near the peeling-mode instability boundary with high normalized pressure gradient α and high edge current density. Experimental analysis and stability calculations indicate that the high poloidal beta with strong Shafranov shift, high α , high edge q, and weak/negative magnetic shear improves the pedestal stability by decoupling the peeling and ballooning modes and stabilizing the ballooning modes, thus facilitating access to the second stability region of peeling–ballooning mode. Access to this second stability regime opens Super-H-like channels without requiring extremely strong shaping or strong torque injection. The high-pressure peeling pedestal allows the pedestal density to go beyond the Greenwald limit with strong E × B shear maintained: the pedestal pressure increases with pedestal density even when n _ped > n _G , until the ideal MHD instability boundary is reached, where giant ELMs occur. The giant ELMs are dominated by a strong n = 1 component and cause a large reduction of the edge pressure, but a negligible change of the core pressure, consistent with kink/peeling-mode induced instability. The pedestal recovers from the collapse and typically sustains a high baseline density, around the Greenwald limit, during the whole discharge duration. Experiments also found that internal transport barriers and peeling-limited pedestal with n _ped ∼ 0.9 n _G can be simultaneously achieved in high β _N plasmas, revealing a feedback interaction between ITB strength and pedestal performance.

Systems driven far from equilibrium are often unstable and prone to violent bursts, from hurricanes and turbulent flows to fusion plasmas. Although stabilisation can mitigate such events, it often requires substantial energy and is not always practical. Yet guiding such systems into less destructive states without suppressing their underlying drive remains a central challenge in nonequilibrium science [1,2]. Tokamak fusion plasmas provide an important example because controlling energy release is essential for fusion power production [3, 4]. Their eruptions can impose heat loads exceeding 1 GW m ^-2 on reactor walls [5], yet can be suppressed by magnetic perturbations of order 0.1% of the confining field strength [6–8]. For over two decades, tokamak experiments have raised a specific question within this broader problem: can such minute perturbations change the macroscopic state without primarily reducing the drive? [8–10] Here, we show that a small spatially patterned magnetic perturbation drives state switching between violent eruptions and gradual dissipation in a fusion plasma. By coordinating the motion of interacting plasma structures, the perturbation strengthens nonlinear interaction and promotes transfer into broadband turbulence. We directly resolve this transfer pathway in a narrow spatial region, and nonlinear simulations reproduce the same structure while showing that the plasma remains unstable. These results support redistribution of energy release rather than simple stabilisation. More broadly, our results suggest a general control principle for far-from-equilibrium dynamics: rather than overpowering the instability, an appropriately structured perturbation can redirect energy release by synchronising interacting structures and strengthening their nonlinear interaction.

Long-pulse, high-performance operation with tungsten plasma-facing components is a crucial challenge for achieving steady-state goals in ITER and future fusion devices. This paper presents a joint international effort between the DIII-D and KSTAR tokamaks aimed at demonstrating the compatibility of the high q95~6 hybrid scenario with a tungsten divertor and superconducting coils. A DIII-D hybrid scenario recipe is successfully tailored to KSTAR’s operational constraints, which facilitates cross-device studies. The results show successful 30-second long-pulse operation in KSTAR at a normalized beta (~2.4) with a benign m/n=4/3 mode and effective mitigation of tungsten accumulation. Despite this success, the KSTAR scenario exhibits a greater than 25\% reduction in thermal confinement compared to the DIII-D reference. To investigate this discrepancy, we performed a predictive TGYRO validation study, which suggests that the difference is likely caused by a weaker temperature pedestal in the KSTAR discharge due to a higher and narrower density pedestal. In addition, such lower thermal confinement leads to higher impurity accumulation, possibly due to a change in neoclassical transport with increased inward pinch, which further degrades global confinement. We also report that stationary, long-pulse operation with sustained thermal confinement requires consistent fueling and wall conditioning. These findings highlight the importance of optimized fueling and real-time wall conditioning in achieving a high temperature pedestal and strong global confinement during stable long-pulse operation. We propose potential approaches toward fully steady-state, long-pulse operation, including real-time wall conditioning using an impurity dropper in KSTAR and future tokamaks.

As contemporary experimental tokamaks are pushed toward reactor-relevant operation, they provide essential testbeds for demonstrating and exploring ELM control strategies for deployment in future fusion pilot plants. Accurate predictions of full plasma responses are essential to guide and optimize these demonstrations. This paper introduces and validates an integrated modeling framework over a historical range of DIII-D operational space. The integrated modeling framework uses only scalar plasma parameters and optional reference boundary to self-consistently and flexibly scan through tokamak operational space and estimate the corresponding ELM-suppression relevant plasma response over 3D coil phase space. The framework generates tightly converged equilibria that satisfy a target set of plasma parameters (I _p , β _N , l _i ), with kinetic profiles constrained by an EPED(NN)-computed pedestal and empirical core model. The plasma response of these modeled equilibria is calculated with GPEC. Validated against 55 distinct DIII-D equilibria, the framework consistently reproduces experimentally constrained equilibrium pressure, q, and other representative profiles using the time-varying information from only evolving scalar plasma parameters; plasma response validation is performed for the n=3 perturbation. In particular, the plasma response of a DIII-D discharge scanning q ₉₅ in search of ELM suppression windows is quantitatively reproduced. By accurately reproducing experimental equilibria and plasma responses across wide parameter variation, the framework supports potential for synthetic parameter scans in key operational, stability, and plasma response dimensions to investigate RMP ELM-suppression experiments and inform predictive RMP scenario optimization.

Permanent magnets can provide static magnetic fields without power supplies or feed lines, offering a simpler alternative to conventional electromagnets in tokamaks. This work examines permanent magnet arrays for edge localized mode (ELM) control in double-null configurations, where traditional RMP coils exhibit limited effectiveness. Linear plasma response calculations using IPEC assess high-field-side (HFS) permanent magnet placement, benchmarking performance against DIII-D low-field-side (LFS) internal coils (I-coils) while evaluating engineering constraints. Modeling results demonstrate that HFS permanent magnet configurations generate HFS plasma response amplitudes more than 5 times larger than conventional I-coil systems. While other resonant response metrics show comparable or reduced response relative to I- coils, the combination of conventional RMP systems and permanent magnets is expected to provide enhanced performance. Operational impact assessments of persistent magnetic fields, including sideband field effects and startup/ramp- up compatibility, reveal acceptable performance within established operational boundaries. This analysis establishes permanent magnet arrays as a technically viable approach for double-null ELM control with reductions in system complexity, motivating further experimental validation on existing tokamak facilities.

An international team from several laboratories and universities has made key advances over the last few years in the control of plasma transport and edge instabilities with applied 3D fields in the KSTAR tokamak to optimize long pulse operation scenarios. This overview begins with the optimization of both core and edge resonant magnetic perturbations (RMPs) to improve fast ion confinement to avoid excessive limiter heat loads due to fast ion losses and successful modeling of the experimental results. Integrated and advanced plasma control techniques with machine learning (ML) and adaptive control were then used to optimize the 3D field spectrum in real-time to control edge localized modes (ELMs) while avoiding core locked modes that could disrupt the plasma. Accelerating the offline model of 3D fields with a surrogate ML model can optimize ELM suppression in the edge while limiting the impact of the applied RMP fields deeper in the plasma core in real-time. In addition, the impact of the 3D fields on the divertor heat load has been modeled and compared with experimental measurements. An analysis of a multi-machine database including KSTAR has been performed to better understand the metrics for the observed RMP thresholds for ELM suppression and the resulting plasma performance. Predictive modeling of the operational space for ELM suppression and density pumpout due to RMP has shown the importance of magnetic islands in the plasma edge and their impact on plasma turbulence. This research has culminated in the development of successful long pulse operational scenarios on KSTAR while attempting to overcome challenges of the new tungsten divertor.

We report the experimental results of controlling the pedestal-top electron density by applying resonant magnetic perturbation (RMP) with in-vessel control coils and the main gas puff in the 2024-2025 KSTAR experimental campaign. The density is reconstructed using a parameterized ψ N grid and five channels of line-averaged density measured by the two-colored interferometer (TCI). The reconstruction procedure is accelerated by deploying a multi-layer perceptron to run in approximately 120 µ s and is sufficiently fast for real-time control. A proportional-integral controller was adopted, with the controller gains estimated from the system identification procedure. The experimental results demonstrate that the developed controller can follow a dynamic target while exclusively using both actuators. The absolute percentage errors between the electron density at ψ N = 0.89 and the target were approximately 1.5% median and a 2.5% average, respectively. The developed controller can even lower the density by using the pump-out mechanism under RMP, and it can follow a more dynamic range of density targets than a single actuator controller. The developed controller will enable experimental scenario exploration within a shot by dynamically setting the density target or maintaining a constant electron density within a discharge.