Recent Publications

Multi-diagnostic characterization of neutrals in the confined region of DIII-D using interpretive DEGAS2 simulations

Quinn Pratt, Shaun R Haskey, George Wilkie, Laszlo Horvath, Raúl Gerrú, Gilson Ronchi, Mathias Groth

Measurements from multiple diagnostics are combined to constrain the density of neutrals in the confined plasma and improve our understanding of edge particle sources. Passive D α emission spectrum measurements are obtained along tangential views at the plasma midplane and near the X-point. Spectral D α measurements provide a strong constraint on the neutral population through energy information in the wavelength distribution of emission. The two-dimensional distribution of neutrals is calculated using interpretive DEGAS2 neutral transport simulations with a plasma background based largely on 1D profiles and magnetic equilibrium reconstruction. DEGAS2 is used as a forward model to predict the emission measured along various lines of sight. We demonstrate two approaches for calibrating DEGAS2 simulations to match spectral emission measurements: (1) fitting the strength of neutral sources at the simulation boundary, and (2) optimizing the plasma background in the pedestal/SOL. Traditional filter-based measurements of D α and Ly α emission are used to validate the calibrated DEGAS2 case at multiple poloidal locations. Once calibrated, DEGAS2 is able to match the measurements generally within a factor of 2, garnering confidence in our diagnostic models and the physics included in DEGAS2. The experimentally constrained 2D neutral distribution is used to quantitatively study particle transport. We report the flux surface averaged neutral density and (main ion) particle source for a standard DIII-D H-mode plasma. In the pedestal, the particle source is found to be primarily driven by divertor neutral sources (recycling). However, neutrals originating in the main chamber play a significant role further inside the plasma. We find the global (main ion) particle confinement time to be τ p,D+ ≈ 130 ms (τ p,D+ ≈ τ E /2). Finally, we present evidence for poloidal asymmetries in the plasma, including decreased main ion temperature above the X-point, and a high density region above the inner target.

Optical diagnostic design for measuring the radiation front with a mid-leg pumped divertor on DIII-D

Gilson Ronchi, Robert S. Wilcox, Jonathan H. Yu, Andreas Holm, Filippo Scotti, Adam McLean, Morgan Shafer, Joshua Hicok, Tyler Elsey

Oak Ridge National Laboratory, General Atomics, Lawrence Livermore National Laboratory

Radial phase variation and energy flow of Alfvén gap modes

Xinran Xu, Jian Bao, Wenlu Zhang, Chao Dong, Jintao Cao, Ding Li

Chinese Academy of Sciences Institute of Physics

Radially curved mode structures of Alfvén eigenmodes are commonly observed in fusion experiments associated with energy transport, which indicate the radial phase variation arising from non-ideal magnetohydrodynamics (MHD) and global effects. In this work, based on MAS global eigenvalue simulations with Landau-fluid bulk plasmas and non-perturbative gyrokinetic energetic ions, we investigate the physical mechanisms responsible for the curved poloidal mode structure tail of RSAE and the rapid phase change of radially coupled RSAE and TAE in DIII-D plasmas. The former one is due to kinetic interaction between RSAE and Alfvén continuum with enhanced mode conversion to kinetic Alfvén waves, and the latter one can be explained by global effects of higher-order radial eigenstate or RSAE-TAE hybrid modes with multiple poloidal harmonics. An improved energy transport model based on global mode structures is formulated and implemented in the MAS framework, which clearly demonstrates the radial phase variation and EI non-perturbative effects on the radial Poynting vector.

Fast-ion enhanced modeling of neoclassical tearing modes at NSTX and DIII-D

James Yang, Eric Fredrickson, John W Berkery, Robert John La Haye, Mario Podesta

Princeton University Plasma Physics Laboratory, General Atomics, Ecole Polytechnique Federale de Lausanne

A new framework for the solution of modified Rutherford equation including fast ions is successfully applied to interpret the tearing mode stability of two discharges NSTX #134020 and DIII-D #135861. The simulated island width growth rates are in better agreement with the measured island growth rate when the fast ions are included. While constants are multiplied to the polarization current contribution terms for NSTX #134020, no constants are necessary to match the simulated and measured island width growth rates for DIII-D #135861. The estimated island frequencies appear to provide an explanation of the different constants used in the two discharges. The gradient scale lengths suggest that the fast ion contribution can become significant in plasmas with flat thermal ion density profile and steep fast ion density profile.

Non-dimensional confinement scaling in similar negative triangularity plasmas on the DIII-D and TCV tokamaks

Alessandro Marinoni, Colin Chrystal, Stefano Coda, Reinart Coosemans, Claudio Marini, Mario Podesta, Olivier Sauter, Matteo Agostini, Max E Austin, Emily A Belli, et al.

General Atomics, Ecole Polytechnique Federale de Lausanne, Columbia University, University of California San Diego, Consorzio RFX

Similarity experiments were performed on the DIII-D and TCV tokamaks to explore the scaling of energy confinement in negative triangularity plasmas using non-dimensional variables. Near up-down symmetric plasmas with large top-bottom averaged negative triangularity were created in a lower single null configuration, with the shape of the separatrix being closely matched between the two devices. The normalized energy confinement is found to weakly improve at increasing collisionality and, between the two devices, shows a machine size scaling behavior between Bohm and gyro-Bohm. Engineering scaling on a large DIII-D dataset is in agreement with the non-dimensional experiment.

Integration of X-point radiator divertor operation with high beta hybrid core plasmas in DIII-D

H Q Wang, Xinxing Ma, Zeyu Li, Roberto Maurizio, Anthony W Leonard, Filippo Scotti, Qiming Hu, Dan M Thomas, Andrea M Garofalo, Siye Ding, et al.

General Atomics, Peking University, Lawrence Livermore National Laboratory, Princeton Plasma Physics Laboratory, University of Tennessee Knoxville

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 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 experimental measured boundary plasma conditions and radiation patterns from attached to XPR detached divertor state 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 high-beta high-confinement core (βN ~3.0, H98~1.25) could be simultaneously achieved. However, this plasma remains ELMing with giant ELMs (W/W ~ 3-4%). With stronger N2 impurity injection, the plasma enters a ‘deep XPR’ regime where the radiation peak is close to 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 H98<1.0, which 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. The modelling also exhibits qualitatively good agreement with experimental observations on the pedestal performance responding to the radiation dynamics, which provides physics insight on the core-edge integration between divertor dissipation and high-performance core that is critically important for future tokamak fusion reactors.

SOLPS-ITER modeling of a dedicated divertor for negative triangularity operation on DIII-D

J.D. Lore, F. Scotti, A. Marinoni, K.E. Thome, C. Lasnier, M. Shafer, D. Truong, H.Q. Wang

Lawrence Livermore National Laboratory, Oak Ridge National Laboratory, General Atomics, University of California San Diego

Achievement of a high-density, high-confinement, and high-beta tokamak plasma regime in DIII-D, and implications for a lower-current path for ITER and FPP

Andrea M Garofalo, Bart Van Compernolle, Siye Ding, Jeremy M Hanson, Christopher Thomas Holcomb, Tomas Odstrcil, Nathan Jordan Richner, Shengyu Shi, H Q Wang, Torrin Bechtel, et al.

General Atomics, Columbia University, Lawrence Livermore National Laboratory, Oak Ridge Associated Universities, Princeton Plasma Physics Laboratory

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, impurity and density gradients can enhance turbulence stabilization caused by high α MHD (α MHD ~(dβ P )⁄dr). This was described by theoretical predictions and gyrokinetic transport simulations [M.T. Kotschenreuther et al, 2024 Nucl. Fusion, 64 076033], and later confirmed by experiments on DIII-D [S. 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. High plasma 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 98y2 , 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.

Understanding carbon sourcing and transport originating from the helicon antenna surfaces during high-power helicon discharge in DIII-D Tokamak

Atul Kumar, Dhyanjyoti Nath, Wouter Tierens, Jeremy D Lore, R S Wilcox, Gilson Ronchi, Morgan W Shafer, Aditya Y Joshi, Onkar Sahni, Mark Shephard, et al.

Oak Ridge National Laboratory, Rensselaer Polytechnic Institute, General Atomics, University of Wisconsin-Madison

The high-power helicon wave system in the DIII-D tokamak could potentially introduce new plasma--material interaction (PMI) challenges owing to rectified RF sheath potentials that develop near the antenna and surrounding plasma-facing components. We present the first application of the STRIPE (Simulated Transport of RF Impurity Production and Emission) framework to helicon-induced PMIs, extending previous STRIPE studies of ICRH antennas by incorporating net erosion, local re-deposition, and three-dimensional global impurity transport. The integrated workflow couples SOLPS, COMSOL, RustBCA, GITR, and GITRm to simulate carbon erosion, re-deposition, and impurity transport for two experimentally constrained DIII-D H-mode helicon operating scenarios with different antenna--plasma gaps, coupled RF powers, and edge plasma conditions. COMSOL predicts rectified RF sheath potentials of 1--5~kV localized near the lower portion of the antenna, where the magnetic field intersects the surface at grazing incidence. Carbon self-sputtering dominates the erosion source, whereas RF-accelerated D$^+$ ions contribute approximately 1\% of the total gross erosion. The smaller-gap operating scenario exhibits substantially stronger gross erosion, enhanced local re-deposition ($\sim$12\%), and a larger confined carbon inventory owing to increased plasma accessibility and broader RF sheath coverage. Comparison with available DIII-D measurements shows no distinct change in the global carbon signal that correlates with the helicon RF pulse, consistent with the simulations indicating that the helicon-generated carbon source remains small compared with the existing background carbon inventory under the present graphite-wall operating conditions. These results demonstrate the capability of STRIPE to integrate RF sheath modeling, plasma transport, surface interaction physics, and three-dimensional impurity transport for the interpretation of helicon-induced PMIs. The study further identifies the principal sources of modeling uncertainty, including grazing-angle RF sheath physics, slow-wave resolution, plasma-background extrapolation, and trace-impurity transport assumptions, providing a framework for future validation and model development.

Multi-field turbulence and transport barrier measurements and validation of predictive codes for high-performance, negative triangularity ELM-free DIII-D plasmas

Guiding Wang, Terry L Rhodes, Julius Damba, Rongjie Hong, William A Peebles, Quinn Pratt, L Zeng, Max E Austin, Kathreen E Thome

General Atomics, University of California - Los Angeles, Princeton Plasma Physics Laboratory

A recent DIII-D campaign demonstrated high confinement, ELM-free plasmas in strongly shaped, diverted negative triangularity (NT) configurations. This paper presents new multi-field turbulence and flow measurements near the edge to understand the physics of edge transport of these plasmas. Correlation electron cyclotron emission and Doppler backscattering measurements provide electron-temperature fluctuation levels, density-fluctuation levels, radial correlation lengths, and poloidal turbulence velocity profiles. The measurements show a modest edge electron temperature/pressure pedestal, a localized poloidal-velocity well near the pedestal-top region, and the peaking of both temperature and density fluctuation levels near the region of minimum poloidal velocity shear, consistent with edge-transport-barrier formation. Linear TGLF calculations indicate a transition from predominantly ITG-like modes in the inner region to TEM-like modes near the pedestal top, approximately coincident with the observed increase in fluctuation levels. These results provide new quantitative constraints for testing predictive models of turbulence and transport in high-performance NT tokamak plasmas.

Comparative qualification of advanced plasma-facing materials for fusion pilot plants through public- and private-sector experiments in DIII-D

Florian Effenberg, Jonathan D. Coburn, Luca Cappelli, Simon D. Corah, Amoolya Grandhi, Jerome Guterl, Charlie Hirst, Mike Jackson, Dylan A. Kohler, Rob Kolasinski, et al.

A coordinated DIII-D campaign exposed and comparatively assessed 44 advanced plasma-facing materials from 12 institutions, including four public-private fusion partnerships, to support fusion pilot plant wall and divertor material down-selection. Samples were exposed using the Divertor Materials Evaluation System (DiMES) under Ohmic, L-mode, and H-mode conditions with edge-localized modes, at 0.2-2.5 MW m$^{-2}$ on flush geometries and 10-15 MW m$^{-2}$ on 10$^{\circ}$ angled geometries. Engineered tungsten architectures retained integrity; long-fiber Wf/W showed the clearest crack-arrest behavior. W-Re and K-doped W showed near-ITER-W-like responses, while additively manufactured W-Ta showed heat-flux-sensitive mass losses of 0.64 mg for the flat sample and 2.19-2.87 mg for angled samples. After irradiation to 0.3 dpa at 550$^{\circ}$C, neutron-irradiated ITER-grade W retained 2.8 times more deuterium than pristine W, while TiB$_2$ showed the lowest D$_2$ release in the Ohmic set. VTaHfMo was the most stable refractory multi-principal-element alloy. NbC and (Nb$_{0.5}$Ta$_{0.5}$)C retained integrity with 0.02-0.03 mg mass loss, whereas ZrC lost 7 mg. CVD SiC retained macroscopic integrity but exhibited an effective Si erosion yield of 0.5, about 5-10 times above prior DIII-D trends. Renewable boron pebble rods underwent controlled recession; 13% of released boron was ionized near the outer strike point and up to 50% was recovered locally. Initial in-situ chromium gross-erosion measurements yielded values of order $10^{-2}$. Together, these results provide cross-material benchmarks for fusion pilot plant down-selection and future AI/ML-assisted plasma-facing-material development.

Spacecraft heat shield study in the DIII-D tokamak

Dmitri M. Orlov, Evdokiya G. Kostadinova, Igor Bykov, Dmitri L. Rudakov, Roman Smirnov, Jayson Barr, Gabrielle Bladon, Alessandro Bortolon, Justin Burzachiello, Lane Carlsson, et al.

We report a new experimental platform developed at the DIII-D National Fusion Facility to investigate carbon ablation and spallation under extreme heat fluxes relevant to fusion plasma-facing components and high-enthalpy atmospheric entry. Carbon samples were exposed to parallel heat fluxes of $30$--$40~\mathrm{MW\,m^{-2}}$ in the scrape-off layer using two complementary approaches: stationary carbon rods inserted near the divertor strike point and slow-launch carbon pellets injected vertically into the edge and core plasma. Pellets penetrating the core experienced heat fluxes approximately an order of magnitude higher. The conditions reproduce key aspects of the shock-layer environment encountered by the Galileo probe during entry into Jupiter's atmosphere. Fast visible imaging, divertor spectroscopy, infrared thermography, CO$_2$ interferometry, and post-exposure profilometry provided measurements of ablation rates, surface recession, and temperature evolution. Measured mass-loss rates of $(1$--$3)\times10^{-2}~\mathrm{g\,cm^{-2}\,s^{-1}}$ agree with semi-empirical aerospace ablation models, while wedge-shaped rods exhibited greater ablation than cylindrical and concave samples. UEDGE-DUSTT simulations incorporating parallel plasma flows, ${\bf j}\times{\bf B}$ forces, and ablation-cloud shielding reproduce the measured pellet trajectories and ablation timescales. These results establish tokamak plasma as a high-heat-flux environment for validating carbon ablation models and studying material response and impurity dynamics in reactor-relevant divertor plasmas.

Assessment of 0-D L-H Power Threshold Scaling and Regression Stability in DIII-D with Applied 3D Magnetic Fields

Michael O Hanson, George R Tynan, Dmitri M Orlov

A database of 192 L-H transitions in DIII-D is used to assess the effects of applied three-dimensional (3D) magnetic fields on the H-mode power threshold and the stability of zero-dimensional (0-D) empirical regressions. The dataset includes nominally axisymmetric discharges and discharges with resonant or non-resonant magnetic perturbations. Filtering criteria reduce uncertainties associated with absorbed power, neutral-beam modulation, and fast-ion losses, while applied-field components are quantified using equilibrium reconstruction and spectral analysis. Measured threshold powers show substantial scatter and systematic deviations from the 2008 ITPA multi-machine scaling, including for discharges without applied perturbations. TRANSP modeling indicates that empirical estimates can substantially underpredict fast-ion losses, particularly at low plasma current and with non-axisymmetric fields. Adding global 3D-field metrics does not robustly isolate the effects of applied perturbations. A fully unconstrained regression retains a 70% residual root-mean-square error and produces nonphysical parameter dependencies, including a plasma surface-area exponent of 2.79. Extrapolations to ITER-relevant conditions consequently have broad confidence intervals. These results show that hidden-variable dependencies can strongly affect empirical threshold parameterizations even in a restricted single-machine dataset. Machine-specific conditions, local edge physics, and power-accounting uncertainties limit the predictive capability of purely 0-D scalings. Improved predictions for ITER and future devices will require better fast-ion-loss treatment and physics-based, edge-localized quantities.

Effect of outer divertor leg detachment on the high field side scrape-off layer in DIII-D and ASDEX Upgrade

R. Gerru, D. Hachmeister, M.G. Burke, L. Horvath, T.M. Wilks, A. Bortolon, J.W. Hughes, Q. Pratt, F. Scotti, C. Tsui, et al.

MIT, Princeton Plasma Physics Laboratory, Max-Planck-Institut für Plasmaphysik, Lawrence Livermore National Laboratory, Sandia National Laboratories

In this work, evidence is presented that detachment of the outer divertor leg leads to a reduction of electron density and neutral pressure in the high-field side (HFS) scrape-off layer (SOL) of the ASDEX Upgrade (AUG) and DIII-D tokamaks with ion B × ∇ B drift directed toward the X-point (favorable configuration). These results are observed across multiple diagnostics and without the use of impurity seeding to reach detachment. In AUG, outer divertor leg detachment correlates with a decrease in electron density near the separatrix at the inner midplane, measured with HFS reflectometry. A concurrent reduction in inner divertor density and neutral pressure at the inner target is observed using divertor Thomson scattering and neutral pressure gauges. These effects are present in both L- and H-mode plasmas. In DIII-D, a similar reduction is detected through analysis of line-integrated hydrogenic emission measured by multiple diagnostics in the HFS SOL close to the separatrix. The consistent trends in both devices indicate that high electron density and strong hydrogenic emissivity in the HFS SOL are common features of H-mode plasmas in the favorable configuration, independent of wall material and divertor geometry. In L-mode plasmas, the reduction in electron density and neutral pressure is not observed in DIII-D, possibly due to differences in wall material. These results emphasize the importance of the divertor state in determining the two-dimensional neutral distribution and edge fueling.

Dependence of Momentum Transport on the Dominant Turbulence Regime in the DIII-D Tokamak

C. F. B. Zimmermann, C. Chrystal, E. Perez, T. Tala, C. Angioni, S. Haskey, F. Khabanov, R. M. McDermott, G. McKee, A. Salmi, et al.

Accurate prediction of toroidal plasma rotation is essential for optimizing confinement and stability in future fusion devices. This work investigates turbulent core momentum transport in the DIII-D tokamak across a transition from ion-temperature-gradient (ITG)- to trapped-electron-mode (TEM)-dominated turbulence. A momentum transport framework previously developed for ASDEX Upgrade is applied to modulated neutral beam injection experiments, separating diffusive, convective, and residual-stress contributions via Fourier analysis of the rotation response. The dataset spans low-rotation conditions, dominant electron heating, and background ExB shearing rates below turbulence growth rates, accessing more reactor-relevant conditions. Gyrokinetic CGYRO and gyrofluid TGLF calculations confirm the scan covers an ITG-to-TEM transition. The analysis yields Prandtl numbers near unity. The pinch number shows no explicit dependence on the transition, instead ordering roughly with the logarithmic density gradient. The normalized residual stress, in contrast, exhibits a non-monotonic, V-shaped dependence across the transition: co-current in deep ITG and deep TEM regimes, near-zero or counter-current in the intermediate mixed-mode regime. This trend collapses onto an approximately linear dependence against electron kinetic profile gradients, suggesting residual stress generation by profile-shearing effects. Weaker background ExB shearing further shifts residual stress toward counter-current values. Linear CGYRO simulations for representative ITG and TEM discharges yield Prandtl and pinch numbers in good agreement with experiment, supporting gyrokinetic momentum-transport predictions in TEM-dominated regimes. These results indicate residual stress plays an important role in core rotation prediction for low-torque plasmas and should be included in predictive models of future reactor scenarios.

First reduced model for integrated computations of helicon wave heating and current drive in magnetic fusion plasmas

Zi-Chen Kan, Lei Chang, Zhen-Yu Wang, Hua-Sheng Xie, Ping-Wei Zheng, Lai Wei, Qi-Bin Luan, Xue-Mei Zhai, Zhao-Qing Hu, Zheng-Xiong Wang, et al.

Fast predictive modelling of radio-frequency heating and current drive is important for integrated tokamak scenario design, yet kinetic calculations of helicon-wave absorption remain too computationally expensive for large-scale parameter scans. We present a reduced model for helicon-wave heating and current drive that retains the dominant parallel electron Landau-damping channel. The wave response is evaluated on the cold-plasma dispersion root, and a single-Landau-pole correction is introduced to obtain compact expressions for the local damping rate and current-drive efficiency. The model is benchmarked against the Chiu-Chan heating model using approximately 1.6 million samples covering representative conditions of EAST, HL-3, DIII-D and KSTAR. The reduction error is found to be governed primarily by the electron Landau parameter and electron beta. Within an identified sub-lower-hybrid-frequency validity window, results from different devices collapse onto a common error curve, which enables an empirical correction that is further tested using ITER-like and BEST-like extrapolation cases. Near and above the lower-hybrid frequency, the agreement deteriorates rapidly owing to changes in the cold-dispersion root structure and the breakdown of the single-branch WKB description. When coupled to a reduced current-drive source, the corrected heating model gives a median deviation of 10.8 percent from the Landau-channel Ehst-Karney reference and reproduces published CFETR current-density profiles. The resulting model provides a computationally efficient reduced closure for helicon-wave heating and current-drive calculations, together with physically interpretable limits on its range of validity.

Experimental Demonstration of Reactor-Relevant Adaptive Density Control on the DIII-D Tokamak with Coordinated Gas and Pellet Fueling

Hassan Al Khawaldeh, Andres Pajares, Sai Tej Paruchuri, Vincent Graber, Tariq Rafiq, Eugenio Schuster, June-Woo Juhn, T Pederson, Daisuke Shiraki, Francesca Turco

Lehigh University, General Atomics, Korea Institute of Fusion Energy, Oak Ridge National Laboratory, Columbia University

Precise regulation of plasma density is crucial for achieving and maintaining fusion-relevant conditions in reactor-grade tokamaks. Tokamak reactors will utilize both gas puffing and pellet injection as standard fueling mechanisms. However, the coordinated use of gas puffing and pellet injection within feedback control frameworks is largely unexplored, underscoring the necessity to develop and validate dedicated strategies on existing machines. These strategies must address the various challenges associated with both actuators, including actuation delays, unknown fueling efficiencies, and the coexistence of continuous-time and discrete-time dynamics. To overcome these challenges, which surpass the capabilities of traditional empirically tuned Proportional-Integral-Derivative (PID) control, an indirect adaptive control algorithm is proposed in this work for the regulation of the line-averaged electron density through coordinated gas puffing and pellet injection. After initial validation in simulations with a multiple-reservoir global particle model, the controller was successfully implemented and tested on the DIII-D tokamak. Experimental results demonstrate robust density tracking under reactor-relevant scenarios, showcasing the controller's ability to seamlessly coordinate actuators while handling disturbances and evolving actuator constraints. This work provides a critical step forward in the control of future fusion reactors.

Access and Limits of RMP ELM Suppression with n =1 Fields in DIII-D

Priyansh Lunia, Nils Leuthold, Nikolas Logan, Carlos Paz-Soldan, Daniel Alexander Burgess, Evan Maxwell Bursch, Richard J Buttery, Qiming Hu, SangKyeun Kim, Jong-Kyu Park

Columbia University, Princeton Plasma Physics Laboratory, General Atomics, Seoul National University

This work reports on DIII-D experiments aimed at extending resonant magnetic perturbation (RMP) suppression of edge localized modes (ELMs) to n =1 fields, where n is the toroidal mode number. Modeling of the 3D ideal MHD plasma response to the RMPs using the GPEC code is used to quantify edge and core resonant fluxes, guiding experimental strategies to increase plasma resilience against core error field penetration, optimize multi-coil phasing, and explore higher q 95 operation. In DIII-D, ELM mitigation is regularly observed across a wide range of n =1 RMP scenarios. A ~100 ms phase of complete ELM suppression was achieved at q 95 ~3.9 using an odd-parity coil configuration. The suppressed phase exhibited clear signatures of RMP ELM suppression, including the elimination of D α spikes, increased pedestal rotation, enhanced magnetic response, and elevated broadband density turbulence. An optimized coil configuration for edge-to-core resonant flux did show increased edge resonance indicated by increased density pumpout, but did not yield RMP ELM suppression. At q 95 ~5.1, a bifurcation to a grassy-like ELM regime occurred, while large type-I ELMs persisted. These results demonstrate progress in experimental access to n =1 RMP ELM suppression in DIII-D, motivating further study for robust access. This work also highlights the potential role of 3D edge stability as well as rational surface alignment in RMP ELM suppression access, which has important implications for the use of low- n RMPs in future reactor-scale devices.

Toroidal phase of post-disruption runaway electron loss to wall in presence of applied 3D fields in DIII-D

Xue Bai, Eric M. Hollmann, Yueqiang Liu, Claudio Marini

University of California, General Atomics

The toroidal distribution of runaway electrons (REs) striking the centre post in DIII-D post-disruption RE final loss events is simulated utilizing the MARS-F code combined with the REORBIT module, by tracing the guiding-centre drift orbits of test REs in the presence of external (applied) plus internal (intrinsic instability) 3D fields. To better recover the experimental results, three different equilibria with different safety factor profiles are adopted. Dominant resistive magnetohydrodynamic instabilities are found to be m / n = 2/1, 3/2 and 1/1 modes (where m and n are the poloidal and toroidal mode numbers). Externally applied n = 1 resonant magnetic perturbations (RMPs), with both even and odd parity configuration, are considered in each equilibrium. Locking of MHD instability to the applied RMPs is simulated by minimizing the total perturbed magnetic energy in the plasma. The simulated toroidal locations, where most REs strike in the presence of an edge m / n = 2/1 instability and the n = 1 RMPs in even parity, are found to agree well with experiments. Modelling also captures the measured n = 2 RE loss pattern on the wall for experiments where an edge m / n = 3/2 instability is dominant instead of the m / n = 2/1 mode. The key overall findings include ( a ) the measured global RE toroidal impact pattern in experiments can be explained by a locked resistive kink instability, ( b ) the toroidal phase of the locked mode is well predicted by minimizing the total perturbed magnetic energy, and ( c ) this edge locked mode determines the toroidal phase of the peak RE wall impact.

Investigating long-duration plasma operation with the international multi-machine CICLOP database

Xavier L Litaudon, Ernesto A Lerche, Olaf Grulke, Christopher Thomas Holcomb, Juan Huang, Marcin Jakubowski, Hyun-Seok Kim, Pierre Manas, Tomohiro Morisaki, Francesca Turco, et al.

IRFM, Max-Planck-Institut für Plasmaphysik, Chinese Academy of Sciences, General Atomics, Culham Science Centre

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Since the previous 2023 IAEA Fusion Energy Conference, significant progresses have been made in tokamaks and stellarators including very recent achievement in duration and/or performance. These progresses are reviewed by analyzing the experimental data provided by 10 tokamaks and two stellarators. The published database [Litaudon X. et al 2024 Nucl. Fusion 64 015001], which initially included data up to January 2022, has been significantly updated for the 2025 IAEA Fusion Energy Conference to incorporate the latest 2023-2025 experiments (up to May 2025) including recent records performance with new entries provided by DIII-D, EAST, JET, KSTAR, WEST, and W7-X. The update dataset has been gathered and coordination have been provided by the IEA-IAEA international CICLOP group (Coordination on International Challenges on Long duration OPeration). An overview of the recent progress toward long pulse operation analysing the CICLOP database is provided in this publication.