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Aug 17

Identifying non-performing or Dud Plasmas for Burning Plasma Control: Insights from JET and TFTR Deuterium-Tritium Campaigns

4 days ago

Lidia Piron, Nicolò Ferron, Eric Fredrickson, Morten Lennholm, Alessandro Pau, Timo Ravensbergen, Olivier Sauter, Fulvio Auriemma, Matteo Baruzzo, Krassimir K. K Kirov, et al.

Consorzio RFX, Culham Centre for Fusion Energy, United Kingdom Atomic Energy Authority, Max-Planck-Institut fuer Plasmaphysik, PPPL

Among the burning plasma controllers for research fusion reactors, the dud detector will be of primary importance as it determines whether the plasma is performing well or if it is a dud. In the latter case, the discharge needs to be terminated to remain within tritium and neutron activation limits. To this scope, monitors which track the plasma performance will be integrated in the plasma control system. In this work, we present a novel dud detector that has been empirically identified based on Deuterium-Tritium campaigns carried out at JET and TFTR. This controller estimates a proxy of the neutron rate using a combination of the diamagnetic energy and the density peakedness. If the predicted neutron rate deviates from the real-time measurement, then the dud detector will trigger an alarm leading to a safe plasma termination if plasma recovery is not expected or, to actuator requests when adjustment is possible. This monitoring function can also be coupled with equilibrium solver and control-oriented models, such as RAPDENS, as proposed in the 15 MA plasma current, 5.3 T toroidal magnetic field baseline Deuterium-Tritium ITER scenario.

United Kingdom Atomic Energy AuthorityMax Planck Institute for Plasma PhysicsCulham Centre for Fusion EnergyConsorzio RFXPrinceton Plasma Physics Laboratory

4 days ago

Aug 9

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

Aug 9, 2026

Plasma Physics and Controlled Fusion

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.

Princeton Plasma Physics LaboratoryGeneral AtomicsSwiss Federal Institute of Technology Lausanne

Aug 9, 2026

Aug 7

Deuterium retention of boron powder from deuterium gas or ion exposure

Aug 7, 2026

Nuclear Fusion

Shota Abe, Adam Q. Kuang, Christopher P Chrobak, Alessandro Bortolon, Camilo Jaramillo-Correa, Bruce E Koel

Princeton Plasma Physics Laboratory, Princeton University, Commonwealth Fusion Systems LLC

We report retention of deuterium (D), a proxy for tritium (T), from D2 neutral gas and D+ ion exposures on commercial boron (B) powder, a proxy for B dust potentially formed in fusion reactors. D removal behavior from B powder is reported as a means of estimating a potential T inventory in B dust in advanced fusion reactors, such as SPARC and ITER, the latter currently plans to employ B wall conditioning by glow-discharge boronization. B wall conditioning forms surface coatings of chemical compounds on tungsten (W) plasma-facing components (PFCs), thus suppressing plasma contamination by impurities such as oxygen, carbon, and tungsten. However, B-based slag or dust particles are expected to form. Such B dust particles can retain hydrogen isotope species, thereby causing a T inventory issue. In this work, the commercial B powder was exposed to D2 neutral gas or D+ ions and analyzed by temperature-programmed desorption to quantify D retention and the desorption temperature. The experiment confirmed D retention from D2 neutral gas exposure. A strong D2 desorption peak at 700 K, corresponding to B-D bonding, was observed for all D2 gas and D+ ion exposure cases. For D2 gas exposures, D retention was significantly enhanced at a B powder temperature of 550 K. B powder bakeout under vacuum at 600 K for 22 hours after D2 exposure exhibited efficient D removal. In contrast, the B powder bakeout at 420 K, even for 1 day, did not remove D efficiently. D retention yields from D+ ion exposure, emulating charge-exchange involving D atoms, were determined. Experiments confirmed that D retention was suppressed by oxidation of the B powder surface, which naturally occurs in reactor environments.

Princeton Plasma Physics LaboratoryCommonwealth Fusion SystemsPrinceton University

Aug 7, 2026

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

Aug 7, 2026

Nuclear Fusion

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.

Princeton Plasma Physics LaboratoryLawrence Livermore National LaboratoryGeneral Atomics

Aug 7, 2026

Aug 4

WEST long-pulse achievements in support of next-step fusion devices

Aug 4, 2026

Nuclear Fusion

Remi Dumont, Theo Fonghetti, Patrick Maget, Pierre Manas, Jean-Francois Artaud, Tullio Barbui, Clarisse Bourdelle, Laurent Colas, Guido Ciraolo, Yann Corre, et al.

French Alternative Energies and Atomic Energy Commission, Chinese Academy of Sciences, Princeton University Plasma Physics Laboratory, IUSTI, Koninklijke Militaire School

The WEST tokamak is equipped with a superconducting toroidal magnetic field system, a multi-megawatt radiofrequency auxiliary power system, and an actively cooled ITER-grade tungsten divertor. As such, it is well adapted to explore experimental aspects related to the long pulse operation of next-step devices. Supported by predict-first integrated modeling, bespoke scenario development has allowed zero-loop voltage pulses to be achieved. The resulting discharges, with plasma currents in the range I p ∼0.22-0.28 MA exclusively sustained by the Lower Hybrid Current Drive (LHCD) system as an auxiliary power source, have achieved durations in excess of 22 min and injected/extracted energies up to 2.61 GJ. Plasma performance is characterized by ranges of poloidal beta β p ∼1.6-2.0, normalized toroidal β N ∼0.6-0.9 and confinement factor H 96L ∼1.0-1.3. Mild MHD activity, identified as resulting from the interaction of 3/1 and 4/1 tearing modes, is occasionally present, depending on the LHCD antenna combination used. This article describes the predict-first approach that has been employed in the context of this long-pulse scenario development endeavor. The main achievements and the physics analyses performed are reviewed, including post-experiment integrated modeling aspects. Prospects for further long-pulse developments are drawn.

French Alternative Energies and Atomic Energy CommissionPrinceton Plasma Physics LaboratoryEcole Royale Militaire-Koninklijke Militaire SchoolChinese Academy of Sciences

Aug 4, 2026

Aug 3

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

Aug 3, 2026

Nuclear Fusion

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.

Lawrence Livermore National LaboratoryPrinceton Plasma Physics LaboratoryGeneral Atomics

Aug 3, 2026

Jul 31

On the transition to large fluxes and access to second stability in gyrokinetic simulations of electromagnetic turbulence in STEP

Jul 31, 2026

Nuclear Fusion

Daniel Kennedy, Yujia Zhang, Toby Adkins, Plamen Ivanov, Francis Casson, Harry Dudding, Bhavin Patel, Colin Roach, Howard Wilson

United Kingdom Atomic Energy Authority, PPPL, UK Industrial Fusion Solutions Ltd

This work investigates the nonlinear transition to large heat fluxes observed in local gyrokinetic simulations of electromagnetic turbulence in STEP. Using the stress-balance framework of Zhang et al. (arXiv:2606.04616, arXiv:2607.11789), we confirm that the onset of extreme transport correlates with a critical value of $q^{2}β_{e}$, where $q$ is the safety factor and $β_{e}$ is the ratio of electron thermal pressure to magnetic pressure, and relate this to a limit on the poloidal beta $β_{\mathrm{pol}}$. Crucially, this critical value lies below any relevant linear stability limit in the ($q$, $β_{e}$) space (e.g., the onset of ideal or kinetic ballooning modes). Using an extensive set of nonlinear gyrokinetic simulations, we demonstrate that the transition to large fluxes in STEP is governed by a balance between the electrostatic and magnetic-flutter stresses. We argue, and also show numerically, that larger-major-radius tokamaks reach the electromagnetic non-zonal regime at lower $β_{e}$, making this MHD-controlled saturation limit more accessible in reactor-scale devices than in small spherical tokamaks. We also demonstrate that access to a second-stable regime enables re-saturation at larger values of $β^{\prime}$. We further show that the ideal ballooning mode (IBM) threshold serves as a useful proxy for delineating this second-stable region and also as a qualitative guide for the onset of large fluxes. These results provide a predictive framework for identifying no-go zone predictions from local gyrokinetics and offer new insight into the electromagnetic saturation physics relevant to STEP and other high-$β_{e}$ devices.

Princeton Plasma Physics LaboratoryUK Industrial Fusion SolutionsUnited Kingdom Atomic Energy Authority

Jul 31, 2026

Jul 30

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

Jul 30, 2026

Nuclear Fusion

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.

Princeton Plasma Physics LaboratoryGeneral AtomicsUniversity of California Los Angeles

Jul 30, 2026

Jul 27

Commissioning and operation of a real-time multi-point Thomson scattering evaluation system at Wendelstein 7-X

Jul 27, 2026

Plasma Physics and Controlled Fusion

Ameer Mohammed, Jürgen Baldzuhn, Torsten Bluhm, Kai Jakob Brunner, Golo Fuchert, Kenneth Hammond, Florian Krafft, Erik Piehler, Mark de Haas, G L Schmidt, et al.

Max-Planck-Institut für Plasmaphysik, Princeton Plasma Physics Laboratory

A real-time multi-point Thomson scattering (rtMPTS) evaluation system has recently been installed and commissioned at the Wendelstein 7-X (W7-X) stellarator during the 2024 -2025 operational phases. This system has demonstrated successful evaluation of Thomson scattered spectra on millisecond timescales. We discuss initial results from this system, consisting of a comparison of different evaluation techniques, assessment of system evaluation times, and proofof-principle control experiments carried out with the W7-X poloidal correlation reflectometer (PCR) system. This diagnostic represents the first step towards the development of a density profile controller at W7-X, which can be leveraged to support steady-state operation.

Princeton Plasma Physics LaboratoryMax Planck Institute for Plasma Physics

Jul 27, 2026

Jul 24

Real-time observation of toroidal current redistributions induced by three-dimensional MHD phenomena triggering vertical displacement events in tokamak plasmas

Jul 24, 2026

Plasma Physics and Controlled Fusion

Matthew Tobin, Steve A Sabbagh, Veronika Zamkovska, Guillermo Bustos Ramirez, Hankyu Lee, Joseph R Jepson, Juan Riquezes, Frederick C Sheehan, Grant A Tillinghast, Keith Erickson, et al.

Columbia University, Korea Institute of Fusion Energy, Culham Centre for Fusion Energy, Princeton Plasma Physics Laboratory, UKAEA

Three-dimensional MHD instabilities, including edge-localized modes (ELMs) and internal reconnection events (IREs), have been observed to precipitate loss of vertical stability in tokamak plasmas, resulting in vertical displacement events (VDEs). This vertical destabilization can occur due to toroidal current redistributions and/or shape changes resulting from these phenomena. Using a recently introduced method for rapidly reconstructing the two-dimensional toroidal plasma current density profile in real-time, results are presented that demonstrate the specific current distribution changes that occur during ELMs (on KSTAR) and IREs (on MAST-U) that lead to loss of vertical control. The method most efficiently reconstructs the toroidal current density profile by doing so on a basis of principal components of historical profiles. These principal components isolate dominant current profile dynamics, improving interpretability, increasing speed, and reducing dimensionality of the profile computation. On KSTAR, this computation is executed in the real-time plasma control system at a rate of 10 kHz (limited by available CPU cycle times), allowing the current profile evolution to be assessed at several times over the course of each ELM event. Further, by incorporating the reconstructions into a novel vertical stability metric, the contribution of specific current profile dynamics to the loss of vertical stability can be assessed in real-time for VDE avoidance and improved understanding of the causal relationship between three-dimensional MHD phenomena and VDEs. The success of this method in approximating toroidal current density profiles from kinetic equilibrium reconstructions is also presented ($R^2=0.990$), along with its capability to produce other equilibrium quantities of interest in real-time at high time resolution.

Culham Centre for Fusion EnergyPrinceton Plasma Physics LaboratoryUnited Kingdom Atomic Energy Authority

Jul 24, 2026

Experimental investigation of a closed vapour box module for a divertor-like configuration in Magnum-PSI

Jul 24, 2026

Nuclear Fusion

Fabio Romano, Victor Tanke, Jacob Schwartz, Robert James Goldston, Serge Brons, T W Morgan

DIFFER, Princeton University Plasma Physics Laboratory, Dutch Institute For Fundamental Energy Research

Efficient management of extreme heat fluxes in the divertor region to extend the lifetime of the components remains a critical challenge for the realization of nuclear fusion-based power plants. Among the alternative concepts explored for the divertor region, the use of liquid metals, particularly lithium, is of interest due its ability to dissipate the incoming plasma heat flux through the vapour shielding effect (VS). In this work, we experimentally investigated a "closed" configuration of a dedicated Vapour Box Module (VBM) in the linear plasma device Magnum-PSI. The goal of the experiments is to simulate the vapour box divertor environment conditions and assess its performance in terms of power mitigation and redistribution and lithium confinement. Initial testing without Li demonstrated the efficacy of a closed VBM structure in inducing detachment via neutral gas accumulation. Apertures which enabled non-condensing gas to be effectively pumped while ensuring lithium condensed on the inner surfaces were therefore added. With a lithium capillary porous structure target used, lithium is directly vaporized by the plasma, forming a dense lithium vapour cloud that interacts with the incoming plasma. This resulted in a significant reduction of the target temperature of at least 48 %, together with a temperature locking effect, a phenomenon typically observed in the VS regime. Lithium vapour confinement within the VBM was strongly correlated with the wall temperature. Relatively cold walls promoted Li re-condensation and therefore improved Li confinement, although with the expected trade-off of increased hydrogenic retention on lithium-wetted surfaces. As the wall temperature increased, the confinement efficiency decreased, consistent with reduced Li re-condensation and thermally activated Li--H chemistry and remobilization at the walls. Diagnostic measurements through embedded thermocouples and calorimetry revealed that lithium vaporization and re-condensation processes also played significant roles in plasma power dissipation. The results advance the case for a closed divertor chamber with direct lithium evaporation from the strike-points as a viable method to manage divertor heat fluxes in future fusion reactors.

Dutch Institute for Fundamental Energy ResearchPrinceton Plasma Physics Laboratory

Jul 24, 2026

Jul 20

Strong gradient neoclassical transport in the plateau regime

Jul 20, 2026

Journal of Plasma Physics

Silvia Trinczek, Felix I. Parra, Peter J. Catto, Iván Calvo

Princeton Plasma Physics Laboratory, Massachusetts Institute of Technology, CIEMAT

Strong gradient regions in tokamaks such as the pedestal or internal transport barriers are regions of reduced turbulence where neoclassical transport can play a dominant role. In pedestals, gradient lengths comparable to the ion poloidal gyroradius have been measured. Standard neoclassical theory can miss important strong gradient effects in these regions because it assumes that the gradient length scales of density, temperature and potential are larger than the ion poloidal gyroradius. We extend plateau regime neoclassical theory into regions of gradients of the order of the ion poloidal gyroradius to capture strong gradient effects on transport processes in the pedestal and internal transport barriers. The fundamental idea behind our new framework is to keep a scale separation between the orbit widths and the gradient length scales by performing a large aspect ratio expansion. In the plateau regime, strong gradients cause poloidal variation that is in–out as well as up–down asymmetric. We study two different test cases assuming either radial force balance or the absence of turbulence and show that strong gradient effects can enhance or reduce standard neoclassical theory predictions in the plateau regime in strong gradient regions.

Massachusetts Institute of TechnologyCentre for Energy, Environmental and Technological ResearchPrinceton Plasma Physics Laboratory

Jul 20, 2026

Jul 17

Interpretation of high-harmonic fast-wave propagation in the scrape-off layer of NSTX-U as a geometrically bounded-waveguide mode

Jul 17, 2026

Journal of Plasma Physics

Seung-Gyou Baek, Nicola Bertelli, Ricardo Antonio De Levante Rodriguez, Paul Thaddeus Bonoli, Syun’ichi Shiraiwa

MIT Plasma Science and Fusion Center, Princeton Plasma Physics Laboratory

This paper revisits the onset of high-harmonic fast-wave (HHFW) propagation in the scrape-off layer (SOL) plasma of the NSTX/NSTX-U spherical tokamak, motivated by past HHFW heating and current drive experiments and modelling. Previously, the fast-wave propagation in the SOL was correlated with the opening (suppression) of the fast-wave right-hand cutoff layer in front of the antenna. In this work, the SOL propagation is interpreted as a geometric waveguide mode guided by the SOL geometry, whose radial width can be comparable to the wave’s perpendicular wavelength in the HHFW regime. A two-dimensional circular model is first employed using full-wave solvers to characterise the poloidal eigenmode structures and to clarify their relationship to annulus resonance. By progressively adding a tokamak-like magnetic-field configuration, starting from a uniform axial field, the effects of magnetic-field gradients and pitch on the SOL waveguide mode are characterised. The poloidal mode numbers supported by the SOL plasma agree well with analytic estimates, indicating that an anisotropic plasma in a bounded geometry selectively supports and amplifies the resonant poloidal mode number. Additionally, two-dimensional axisymmetric NSTX-U simulations demonstrate that the SOL eigenmode features identified in the circular model persist in the experimentally relevant configurations. A control approach based on lengthening the wave perpendicular wavelength relative to the SOL width, including a higher magnetic (B)-field operation, is discussed. The analysis here shows the key role of a bounded geometry in interpreting HHFW eigenmode coupling and propagation in the NSTX-U SOL plasma.

MIT Plasma Science and Fusion CenterPrinceton Plasma Physics Laboratory

Jul 17, 2026

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

Jul 17, 2026

Nuclear Fusion

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.

Max Planck Institute for Plasma PhysicsMIT Plasma Science and Fusion CenterPrinceton Plasma Physics LaboratoryLawrence Livermore National LaboratorySandia National Laboratory

Jul 17, 2026

Jul 13

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

Jul 13, 2026

Plasma Physics and Controlled Fusion

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.

General AtomicsPrinceton Plasma Physics Laboratory

Jul 13, 2026

Nonlinear saturation of ballooning modes in stellarators

Jul 13, 2026

Nuclear Fusion

X. Chu, S.C. Cowley, N. Ferraro, Y. Zhou, F.I. Parra

Princeton Plasma Physics Laboratory, Princeton University, Shanghai Jiao Tong University

Ballooning mode saturation is investigated in realistic stellarator configurations using the flux tube approach of Ham et al (2018 Plasma Phys. Control. Fusion 60 075017), Ham et al (2016 Phys. Rev. Lett. 116 235001). The method is adapted to account for the lack of exact force balance in stellarator equilibrium solvers that assume existence of nested flux surfaces. A variational approach for calculating flux tube energy is developed to overcome this force error problem in stellarator numerical equilibria. Saturated (equilibrium) flux tube states that cross 10%–20% of the plasma minor radius are shown to exist for linearly ballooning unstable profiles. It is shown that several features of the displaced flux tube structure in a full nonlinear MHD simulation of Wendelstein 7X are reproduced by our model. Saturated states are found in a compact stellarator equilibrium close but below the marginal ballooning linear instability, i.e. the unperturbed equilibrium is metastable. This suggests that edge-localized-mode-like explosive MHD behavior may be possible in stellarators.

Princeton UniversityPrinceton Plasma Physics Laboratory

Jul 13, 2026

Jul 10

Maintenance strategy, structural design, and site layout of the ST-E1 fusion power plant

Jul 10, 2026

Nuclear Fusion

J. Willis, K. Chandrasekhar, P. Cheema, J. England, V. Godhani, E. Guise, S.M. Levine, R. Pocock, A. Scott, A. Shone, et al.

Tokamak Energy Ltd, Remote Applications in Challenging Environments (RACE), Tokamak Energy Inc, Princeton Plasma Physics Laboratory

An effective fusion reactor maintenance scheme enables safe operations and short downtimes. This in turn leads to high availability, which is critical to the commercial viability of a power-producing plant. In tokamak-based fusion power plants, the chosen maintenance approach has a significant impact on the spatial design of the tokamak, as well as the surrounding infrastructure, and therefore needs to be considered from the outset. Tokamak Energy has developed a pre-concept design of a fusion power plant, ST-E1. This work describes the major drivers and constraints that have been considered, presents the tokamak architecture and chosen maintenance regime, and discusses how this enables the plant’s two-phased approach to demonstrating commercial operations. It also shows the implications for the design of other systems areas, in particular the machine structural arrangement and bioshield and hot cell layout. The reactor core segmentation and removal scheme replaces entire toroidal segments radially through a large vacuum port, along a single axis only. The result is a change-tolerant machine and plant layout that can accommodate the evolving designs of the tokamak.

Tokamak EnergyPrinceton Plasma Physics Laboratory

Jul 10, 2026

Auxiliary heating and current drive physics for the ST-E1 fusion power plant

Jul 10, 2026

Nuclear Fusion

N.A. Lopez, A.F.P. McAdam, N. Bertelli, S. Shiraiwa, M. Ono, Y. Takase, X. Zhang, M. Borscz, J. Stirling, A. Alieva, et al.

Tokamak Energy Ltd, Princeton Plasma Physics Laboratory, University of Edinburgh, University of New South Wales

This work describes the physics basis for the proposed auxiliary heating and current drive system on the ST-E1 fusion power plant. The ST-E1 flattop plasma considered here is fully non-inductive with a bootstrap fraction of 0.9 and the remaining current driven by EC waves. Using the recently published physics-based optimization method for EC launchers (Lopez et al 2025 Plasma Phys. Control. Fusion 67 055012), we show that the target flattop ECCD can be achieved with a net efficiency of 52 kA MW −1 using fundamental O-mode (O1) with frequency range 160–200 GHz launched from the low-field side top half of the vacuum vessel (LFS top-launch). From considering two candidate rampup scenarios, we conclude that LFS top-launch O1 ECCD can be equally effective during the early stages of plasma operation, although poloidal steering might be needed. X-mode waves injected from the LFS midplane are also shown to be effective for rampup even when T e < 1 keV. We also present modeling results for the pre-conceptual design of an ICRH system proposed for ST-E1. Using TORIC, we find that an ICRH system aiming for 42–48 MHz and toroidal mode number n ϕ ∼ 10 robustly achieves dominant ion damping via Helium-3 minority heating transitioning to second-harmonic Tritium heating. We then show that such waves can be efficiently generated by a 5-strap traveling-wave antenna (TWA) using the Petra-M code. The TWA has a 40–45 MHz passband within which ∼ 60 % of the power entering the TWA is coupled to the plasma with the remaining ∼ 40 % of the power being transmitted through the TWA and possibly recirculated; the power reflected back into the transmission lines is negligible. This passband structure persists even when the evanescent distance is increased by a factor of two, or when the magnetic-field angle is increased by 30 ∘ , demonstrating inherent load resilience that will be crucial for effective ICRH on ST-E1.

Tokamak EnergyUniversity of EdinburghPrinceton Plasma Physics Laboratory

Jul 10, 2026

Time-dependent scenario modeling for the ST-E1 fusion power plant

Jul 10, 2026

Nuclear Fusion

X. Zhang, N.A. Lopez, M. Borscz, J. Kang, Y. Takase, M. Scarpari, C. Marsden, M. Ono, S.A.M. McNamara, E.N.J. Maartensson, et al.

Tokamak Energy Ltd, Princeton Plasma Physics Laboratory

ST-E1 is a low aspect ratio fusion power plant being designed by Tokamak Energy targeting 1.5 GW of fusion power. Characterization of the ST-E1 flat-top scenario is described elsewhere McNamara et al (2026 Nucl. Fusion 66 086008); here we focus on addressing the question of how to ramp-up the ST-E1 plasma from an initial state following breakdown and flux-surface formation to the target flat-top state. Being low-aspect ratio, the available solenoid flux of ST-E1 is limited. Therefore, particular consideration is placed on developing ramp-up scenarios that predominantly use inductive flux provided by external vertical field coils. Through time-dependent modeling with METIS, we show that this is possible when the ramp-up is performed at relatively high plasma density: although auxiliary current drive efficiency is reduced, this is significantly outweighed by (1) higher electron-ion collisional equilibration, (2) higher fusion power once ions become sufficiently hot, (3) higher poloidal beta for increased vertical-field flux, and (4) potentially favorable exhaust compatibilities. Ultimately, we show the target ST-E1 flat-top performance can be reached after a ramp-up period lasting 150 s using less than 40 Vs of solenoid flux (with vertical field providing ∼ 90 Vs of flux). The sensitivity to model assumptions are presented, with the general observation that deleterious effects can be mitigated through minor alterations of the auxiliary power temporal waveform and/or total auxiliary power level. The impact of a solenoid and the auxiliary power mix (electron cyclotron heating only versus electron and ion cyclotron heating) on the ST-E1 ramp-up success are also discussed in appendices. On this latter topic, we show that the effect of direct-ion heating during ramp-up is obscured by the uncertainty in the pedestal dynamics, identifying a clear line of future work required to make a definite decision on the ST-E1 auxiliary power mix.

Tokamak EnergyPrinceton Plasma Physics Laboratory

Jul 10, 2026

Power and particle exhaust in the ST-E1 fusion power plant

Jul 10, 2026

Nuclear Fusion

M. Robinson, A. Scarabosio, E. Vekshina, J.H. Nichols, J.D. Lore, K. Borowiec, J. Varje, S.M. Levine, M. Scarpari, E.N.J. Maartensson, et al.

Tokamak Energy Ltd, Oak Ridge National Laboratory, Fondazione LINKS, Princeton Plasma Physics Laboratory

Power exhaust challenges and potential solutions for a 5 m major radius, low-aspect ratio burning tokamak have been explored. 1D edge plasma models have been used to screen for access to detachment using short and long outer divertor legs in double and single null configurations, using Ar as the primary impurity and assuming tungsten plasma-facing components (PFCs). These show that detachment access can be accessed for all but the most conservative assumptions on scrape-off layer (SOL) width and power, but that trade-offs will be required between magnet engineering and the size of the acceptable window of as-yet uncertain plasma parameters. SOLPS-ITER was used to further model selected plasma scenarios, confirming that Ar seeding can be used to achieve dissipative divertor scenarios with peak deposited heat fluxes below 15 MWm − 2 . Initial scoping of first wall loads and positioning of limiters has been carried out, showing the feasibility of protecting the breeding blanket wall during steady state without impeding tritium breeding. Initial PFC technology selection is also presented, identifying this as a critical area where further work is needed to find an attractive solution for helium-cooled PFCs that can handle high heat fluxes without excessive power requirements. Key questions and trade-offs for concept development have been identified, including: how to achieve high radiation for reduction of SOL power without core performance degradation; whether power exhaust can be well-controlled in a double null plasma; mechanical design and materials challenges of high-heat flux PFCs; and control of material erosion, redeposition and tritium retention.

Tokamak EnergyPrinceton Plasma Physics LaboratoryOak Ridge National Laboratory

Jul 10, 2026