Recent Publications

Oct 12

Integrating Gyrokinetic Flux Predictions with Ideal MHD Stability Boundaries

J. McClenaghan, K. E. Thome, T. F. Neiser, J. Candy, F. D. Halpern, T. H. Osborne, S. Saarelma, the MAST-U team

Plasma Physics and Controlled Fusion·Oct 12, 2026Plasma & ConfinementAI, Modeling & Simulation

Accurate prediction of pedestal height and width in tokamaks remains a critical issue as it strongly influences the predicted plasma performance of all future reactors. We present an integrated pedestal-stability workflow that combines equilibrium scans with magnetohydrodynamic (MHD) stability analysis using ELITE and GATO and gyrokinetic transport predictions using CGYRO/QLGYRO. The workflow reproduces the characteristic KBM first- and second-stability structure previously identified in gyrokinetic pedestal studies. Applied to spherical tokamaks (STs), the workflow shows good agreement with past studies when low-n peeling stability is included, emphasizing the importance of resolving low-n physics in ST pedestals. Extending the analysis to SPARC-like, high toroidal field plasmas, kinetic ballooning mode (KBM) and microtearing mode (MTM) heat fluxes are found to increase strongly with toroidal field, suggesting that access to the KBM second-stability region may become significantly more difficult in high toroidal field devices. Comparison with global ELITE finite-n analysis at SPARC parameters suggests that the H-mode pedestal lies in an intermediate regime bounded by local KBM second stability on one side and global finite-n ballooning instability on the other, consistent with the EPED picture once global effects are included

Swiss Federal Institute of Technology Lausanne·Culham Centre for Fusion Energy·General Atomics

Oct 9

System identification simulations using the Advanced Fluid Neutral model and Wide Grid formalism in SOLPS-ITER.

Stijn Peter Kobussen, Sander Van den Kerkhof, Federico Cursi, Wouter de Keyser, Gijs Lukas Derks, Bob Kool, David Moulton, Kevin Verhaegh, M van Berkel, Sven Wiesen

Nuclear Fusion·Oct 9, 2026Plasma & ConfinementControl & DiagnosticsTritium & Fuel CycleAI, Modeling & Simulation

This work presents a first step towards using the SOLPS-ITER code package for synthetic system identification using multisine perturbations. Numerical modelling is used to replicate system identification experiments performed to establish exhaust control done on MAST-U. These experiments involve perturbing the gas injection rate and observing the SOL response in the frequency domain. The model used is the Advanced Fluid Neutral (AFN) model of SOLPS-ITER, implemented in the Wide Grids (WG) version of the code. The grid was generated with the Grid Optimization and Adaptation Toolbox (GOAT) and extends to the vessel wall, to enable calculating fluid neutral parameters across the entire domain. New boundary condition types were implemented in the code to allow for multisine gas injection perturbations, as well as reservoir models for remote vessel regions without plasma. In steady state, it is possible to simulate plasmas ranging from attached to deeply detached using different gas injection rates. This allows for comparison between simulation and experimental results at similar levels of detachment. Comparison between simulation and experimental data in the frequency domain shows that the AFN SOLPS-ITER simulations predict faster response times to divertor fuelling perturbations than the experiments, likely in part due to the fluid approximation used for the neutrals. When including the reservoir models for remote vessel regions the phase shifts increase, but are still too low compared to experiments.

Dutch Institute for Fundamental Energy Research·Culham Centre for Fusion Energy·Forschungszentrum Jülich

Oct 7

Multi-beamlet operation of the ITER negative ion source prototype SPIDER towards full beam extraction

Emanuele Sartori, Riccardo Casagrande, Isabella Mario, Antonio Pimazzoni, Basile Pouradier-Duteil, Alastair Shepherd, Riccardo Agnello, Kamran Ahmad, Piero Agostinetti, Matteo Agostini, et al.

Nuclear Fusion·Oct 7, 2026Heating & Current Drive

The neutral beam heating systems for ITER will be based on radio-frequency driven sources, accelerating a negative ion beam current of tens of amperes for up to one hour. In its second operation phase SO-2, the full-scale prototype source SPIDER was operated for the first time in multi-beamlet configuration at ⁓70 kW/driver, extracting a H– current density of 215 A/m2 over ¼ of the extraction area at 0.3 Pa source filling pressure, providing a total accelerated beam power of ⁓700 kW. Scaling of the results confirms the specifications of the on-going RF generators upgrade, indicating that target current density of 330 A/m2 in hydrogen is achievable by increasing the RF power to 100 kW/driver. Beam uniformity was assessed at the beamlet-group scale under moderate underperveance conditions: at high power row-averaged beamlet currents were within ±10% for fifteen out of sixteen rows. A lower homogeneity was observed at the single-beamlet level. The influence of the magnetic filter field on beam uniformity was also characterised, confirming the beneficial effect of multi-driver operation on homogeneity at the extraction region. A caesium conditioning procedure was developed and validated for the partial-source configuration. An anomalous amplification of the acceleration current peculiar of the vacuum-insulated source design specific to ITER sources, attributed to stray backstreaming charges and surface emission effects, was identified and mitigated by electrostatic shielding. Together with the ongoing installation of new solid-state RF generators and an improved pumping system, the results presented in this work open the way to full-source operation in the next experimental campaign and yield findings of direct relevance to the ITER injector design.

National Research Council·Culham Centre for Fusion Energy·ITER Organization

Sep 25

On the relationship between the H-mode separatrix density and engineering parameters across multiple tokamaks: physics-based models, regression, and extrapolations to next-step devices

Davide Silvagni, Ondřej Grover, Adriano Stagni, Jerry W Hughes, Marco Andrés Miller, Bartosz Lomanowski, L. Balbinot, Guido Ciraolo, Wouter Dekeyser, Michael G Dunne, et al.

Nuclear Fusion·Sep 25, 2026Plasma & ConfinementAI, Modeling & Simulation

The electron density at the separatrix (n e,sep ) plays a central role in balancing energy confinement, detachment achievement, and ELM suppression in tokamaks, thereby influencing core-edge integration. To study what determines this key parameter, a database of H-mode separatrix density measurements from the Alcator C-Mod, ASDEX Upgrade, and JET tokamaks has been assembled using a consistent analysis method across all devices. This dataset is used to assess the validity of a physics-based predictive model and to derive a regression scaling expression for n e,sep , both requiring only engineering parameters as input. The theory-based expression is obtained by coupling two-point model equations with simple geometrical relations, and successfully reproduces experimental measurements across all three devices, with the exception of a common multiplicative constant. The regression confirms similar parameter dependencies, revealing a positive dependence on divertor neutral pressure and the ratio of the power entering the scrape-off layer to the major radius, a negative dependence on the toroidal magnetic field and minor radius, and no significant dependence on the plasma current. Both the resulting scaling and theory-based expressions predict n e,sep within a factor of 1.5 across the three machines, and provide projections to next-step devices (ITER, SPARC, DTT, JT-60SA and COMPASS-U) that are in agreement with available SOLPS simulations.

Consorzio RFX·Culham Centre for Fusion Energy·Massachusetts Institute of Technology·Oak Ridge National Laboratory·Max Planck Institute for Plasma Physics

Sep 22

Global Gyrokinetic Simulations of Electromagnetic Turbulence in STEP

Daniel Kennedy, Facundo Sheffield, Tobias Görler, Colin Roach, Maurizio Giacomin, Arka Bokshi, David Dickinson, Harry Dudding, Bhavin Patel

Nuclear Fusion·Sep 22, 2026Plasma & ConfinementAI, Modeling & Simulation

This paper presents gradient-driven global electromagnetic gyrokinetic simulations for a conceptual burning flat-top operating point of STEP [1], STEP-EC-HD, and investigates how non-local effects influence the nonlinear saturation and transport of the electromagnetic turbulence at finite $β$. Local gyrokinetic simulations have shown that including $δB_{\parallel}$ is essential for the dominant hybrid kinetic ballooning modes, or hKBMs, to be unstable in STEP [2]. Using the long-wavelength $δB_{\parallel}$ solver [3] implemented in GENE [4], this work demonstrates that the linear mode spectrum can be accurately captured in global geometry, which results in good agreement with an ensemble of local flux-tube simulations. The global framework reproduces the hKBMs identified in [5], while microtearing modes remain challenging to resolve due to their shorter radial scales. Nonlinear simulations reveal clear evidence of an electromagnetic transition to states with extremely large heat fluxes, consistent with local simulations and with the predicted loss of zonal-flow regulation for this proposed operating point [6]. These findings establish the capability of global gyrokinetics to capture finite-$β$ dynamics in STEP-like plasmas and motivate future work to identify the conditions governing this transition.

United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·Max Planck Institute for Plasma Physics

Sep 17

Determining neutral fueling response to RMPs in MAST using a multi-reservoir particle balance model and EMC3‑EIRENE

K Flesch, James R Harrison, Andrew Kirk, Ian Waters, Heinke Frerichs, Oliver Schmitz, Livia Casali

Nuclear Fusion·Sep 17, 2026Plasma & ConfinementAI, Modeling & Simulation

The application of certain configurations of resonant magnetic perturbations (RMPs) has been shown to cause a plasma pump-out in both L- and H-mode discharges at MAST. In this paper we discuss the impact of neutral fueling on this density reduction. The neutral fueling and average particle confinement time $\tau_p$ of the main ion species were calculated using a 0-D particle balance analysis. When the RMPs were applied, it was found that there was an increase in ionizations and a 15$\%$ reduction in $\tau_p$ for L-mode discharges and a similar result for inter-ELM (edge localized modes) periods of H-mode discharges. A time-dependent global multi-reservoir particle balance (MRPB) was developed, which included atomic and molecular reservoirs, to further investigate the role neutrals had on the density change. We discuss how this model was able to accurately reproduce the experimentally measured density reduction and ionization increase due to either a reduction in $\tau_p$ or a reduction in particle fueling efficiency. Results from EMC3-EIRENE modeling indicate this change could be attributed to neutral particle fueling occurring in locations with now-opened fieldlines due to the chaotic edge-region from RMP applications.

University of Wisconsin Madison·Culham Centre for Fusion Energy

Sep 10

BLUEMIRA: a modular, open-source framework for designing tokamak fusion reactors

Matti Coleman, James Cook, Fabrizio Franza, Ivan Maione, Simon Mcintosh, Hudson Baker, Alexander Ian Blair, Shail Desai, Oliver Funk, Georgina Graham, et al.

Nuclear Fusion·Sep 10, 2026Blankets & NeutronicsMagnets & SuperconductorsFusion Plant EngineeringAI, Modeling & Simulation

As efforts across the world to deliver fusion power enter their respective conceptual design stages, many begin to encounter the difficulties of generating complete designs, evaluating them, performing trade-off studies, and comparing different alternatives. Many organisations successfully use fusion reactor systems codes with low-fidelity formulations of the critical physics and engineering aspects to inform the initial design stage and steer them towards a region of the design space they consider attractive, but struggle to explore the design space in more depth. This work is an overview of the open-source tokamak reactor design framework, BLUEMIRA, built from the BLUEPRINT [1, 2] and MIRA [3, 4] codes, which goes beyond the “classical” systems code paradigm and enables the user to parameterise their reactor designs at higher fidelity. Reactor design workflows ranging from 0-D “radial build” design, equilibrium design, to 3-D CAD and automated neutronics analyses can be run within minutes. We present the functionality implemented in the BLUEMIRA code and demonstrate its application to the conceptual design of a conventional aspect ratio fusion reactor, performing an indicative design study over aspect ratio and number of toroidal field coils.

United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·ITER Organization

Sep 7

Neoclassical tearing mode seeding by Alfvén eigenmode coupling in MAST-U

Kitt Cameron Medley Thomas, Laszlo Bardoczi, Kieran Gibson, Juan Ruiz Ruiz, Mykola Dreval, David Anthony Ryan, Kenneth G McClements, Clive A Michael

Nuclear Fusion·Sep 7, 2026Plasma & ConfinementControl & Diagnostics

We present evidence of tearing mode seeding by nonlinear three-wave coupling of Alfvén eigenmodes in a tokamak plasma. Here, a m,n=2,1 magnetic island appears to be seeded in a MAST-U discharge by coupling of n = 2 and n = 3 Alfvén eigenmodes, which is demonstrated by increased squared normalised bicoherence calculated from magnetic probes preceding tearing mode onset. This observation offers a new explanation for tearing modes that would be otherwise addressed as spontaneous, and indicates three-wave coupling can seed 2,1 magnetic islands in plasmas without pre-existing n ≥ 2 tearing modes. Further, this mechanism has the potential to be more prevalent in burning plasmas expected in future fusion devices.

Kharkiv Institute of Physics and Technology·Culham Centre for Fusion Energy·General Atomics

Aug 25

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

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

Plasma Physics and Controlled Fusion·Aug 25, 2026Plasma & ConfinementControl & Diagnostics

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.

Princeton Plasma Physics Laboratory·Consorzio RFX·United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·Max Planck Institute for Plasma Physics

Validation of Integrated Modelling of KSTAR Plasmas with Carbon Divertor using JINTRAC and Prediction for Tungsten Divertor

Beomsu Kim, Sun Hee Kim, Florian Koechl, Francis J Casson, Romain Futtersack, Chan-Young Lee, SeulChan Hong, JAEGON LEE, Boseong Kim, Yong-Su Na

Plasma Physics and Controlled Fusion·Aug 25, 2026Plasma & ConfinementMaterials & Plasma-Facing ComponentsAI, Modeling & Simulation

The JINTRAC integrated modelling suite has been adapted for the KSTAR tokamak, with its primary goal being the establishment of a core-edge integrated modelling framework for predictive analysis of KSTAR plasmas. This framework was validated by benchmarking JINTRAC's physics modules against both established codes and experimental data from a KSTAR carbon (C) divertor discharge (#25458), demonstrating high fidelity. Based on the validation, predictive core-edge coupled simulations were performed to investigate the effects of changing the divertor material from C to tungsten (W). The simulations predicted a significant increase in core radiation and a corresponding decrease in plasma temperatures, consistent with the behavior of high-Z materials. Specifically, the high tungsten influx causes radiative cooling in the pedestal and core regions, leading to a reduction in edge bootstrap current and a compensatory increase in core inductive current. This redistribution of current density drops the central safety factor (q 0 ) below unity, triggering sawtooth instability which serves to suppress core impurity accumulation. Sensitivity analysis indicated that while toroidal rotation exhibits a non-monotonic correlation with tungsten profile peaking, the total tungsten inventory and radiated power (P rad ) increased with higher rotation. Additionally, increasing the separatrix electron density effectively suppressed the tungsten source at the divertor target, yet the reduction in core tungsten inventory was limited compared to the extent of source reduction. These findings imply that achieving high-performance operation requires a comprehensive control strategy that minimizes the absolute impurity influx across the separatrix through combined source suppression and transport control.

Culham Centre for Fusion Energy·ITER Organization

Aug 24

Experimental Identification of the n=2 Error Field and its Interaction with MHD Activity and Plasma Rotation in MAST-U

Lidia Piron, David Anthony Ryan, Andrew Kirk, Alessandra Tonel, Paolo Zanca, Geoffrey Cunningham, Matteo Baruzzo, Sam Blackmore, Christopher J Ham, Scott Alan Silburn, et al.

Plasma Physics and Controlled Fusion·Aug 24, 2026Plasma & ConfinementControl & Diagnostics

Spurious magnetic field perturbations, known as error fields (EFs), with toroidal mode number n = 2 can have deleterious effects similar to those associated with n = 1 error fields. An n = 2 error field source is expected in MAST-U because the poloidal field coils P4 and P5, retained from the previous device MAST, exhibit n = 1 and n = 2 deformations due to coil manufacturing imperfections. This work presents the n = 2 error field identification studies carried out in MAST-U. The n = 2 compass scan indicates that, for 750 kA plasma current, double-null divertor H-mode plasmas, the n = 2 error field is relatively small when assessed using locked mode onset and rotation braking as metrics, suggesting that the n = 2 EF is effectively screened by the plasma in this scenario. Furthermore, during the n = 2 EF identification studies, an interesting interplay between the onset times of n = 1 and n = 2 rotating modes and plasma rotation was observed. This observation enables the identification of control strategies aimed at delaying the onset of n = 1 and n = 2 MHD modes while sustaining plasma rotation.

Consorzio RFX·United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy

Aug 21

Mitigation of sawtooth transient heat flux in the MAST-U Super-X divertor with deuterium and nitrogen

Rory Scannell, Jack Flanagan, Zhouji Huang, James R Harrison, Kevin Verhaegh, Scott Alan Silburn, Peter Ryan, Stuart Scott Henderson, Nicola Lonigro, Sidney Leigh, et al.

Nuclear Fusion·Aug 21, 2026Plasma & ConfinementMaterials & Plasma-Facing Components

The MAST Upgrade Super-X divertor protects plasma-facing components from heat fluxes during transient events. This paper examines heat loads from sawtooth crashes with energies ΔWsawtooth ≈ 2–9kJ and investigates the impact of deuterium and nitrogen gas pressure on their mitigation. In a D2 gas scan for sawteeth with energies up to 6kJ, the peak heat flux (q⊥,max) scales with the transient energy and inversely with the neutral pressure, with variations in ΔW/Pgas accounting for 70% of the observed variation in q⊥,max. Larger sawtooth transients (≈ 6–9 kJ) fall outside this trend and exhibit much higher heat fluxes. With D2 puffing in the divertor, the Super-X configuration exhibits substantially lower peak q⊥,max for comparable transient energy loads than observed in conventional divertor configurations. However, this reduction is broadly consistent with geometric expectations especially at higher pressures. Nitrogen gas seeding was applied to transients with energies below ≈ 6 kJ. Increasing the nitrogen pressure reduces q⊥,max, by 4.5MW/m2/Pa in Super-X while nitrogen seeding is found to have much less impact on transients in conventional divertor configuration. In the case of N2 seeding, the Super-X divertor outperforms conventional divertor configuration beyond geometric expectations. Direct Super-X divertor Te and ne measurements during transients were obtained from Thomson scattering. These indicate quiescent inter-transient temperatures of < 1 eV, corresponding to strong detachment. The temperatures rise to 6–8 eV during typical sawtooth transients and 10 eV where divertor heat loads exceed q⊥,max > 2MWm−2. The experimental data are compared to modelling results using a 1D exhaust code in the ReMKiT1D framework. The modelling highlights the importance of recycling on understanding the buffering of transients.

United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·University of Liverpool

Development of a high-current, high fusion performance scenario on JET: physics insights and operational challenges

Luca Garzotti, Domenico Frigione, Peter J Lomas, Fernanda Rimini, Dirk Van Eester, Vito Konrad Zotta, Spyridon Aleiferis, Edoardo Alessi, Fulvio Auriemma, Rennan Bianchetti Morales, et al.

Plasma Physics and Controlled Fusion·Aug 21, 2026Plasma & ConfinementMaterials & Plasma-Facing ComponentsAI, Modeling & Simulation

Since the installation of a beryllium-tungsten(Be/W) first wall on JET an intense scenario development programme has taken place to realize a high-current (3.5-4 MA) scenario, denominated JET baseline scenario, with $q_{95} \sim 3$, $\beta_N \sim 1.8$ and capable of delivering high fusion performance ($P_{fusion} > 10$ MW) for 5 s in the D-T campaigns conducted in 2021 and 2023. The JET baseline scenario was realized in D-T at 3.5 MA in 2021 (DTE2 campaign), but, despite achieving transitorily $P_{fus} \sim 8$ MW, it could not be sustained for more than 3 s. In this paper we will revisit and expand the analysis of the baseline scenario at 3.5 MA and present new results obtained in D-T at 3.0 MA in DTE3, where we were able to sustain the scenario for 5 s. We will also present recent modelling results showing how available semi-empirical and first-principles transport models can explain a variety of phenomena ranging from fusion performance, impurity transport and control of the fuelling mix. Many of these points are potentially relevant for bigger machines such as ITER. Finally, we will discuss open questions and improvements to our present modelling capabilities required to extrapolate the results to power plant devices.

National Research Council·United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·Institute of Plasma Physics and Laser Microfusion

First wall particle fluxes and Be erosion in high-performance JET-ITER baseline plasmas

Eduardo de la Cal, Carine Giroud, Henri Aaron Kumpulainen, Juri Romazanov, Itziar Balboa, Scott Alan Silburn, Pedro Carvalho, Juuso Karhunen, Beth Thomas, Alex Tookey, et al.

Nuclear Fusion·Aug 21, 2026Plasma & ConfinementControl & DiagnosticsMaterials & Plasma-Facing Components

In next step large fusion devices with a tungsten (W) first wall such as ITER, impurity control will be a challenge to avoid strong core radiation from highly ionized metals. Here we analyse the deuterium (D) fluxes impacting on the beryllium (Be) first wall limiters and the resulting surface erosion in high-performance neon (Ne) seeded ITER-baseline plasmas with up to 35 MW heating power in JET tokamak. Visible cameras are used to quantify the fluxes using the spectroscopic S/XB method at the regions of strongest plasma-wall interaction: the Outer Midplane (OMP) and the Upper Dump Plates (UDP). We first describe how Ne seeding, which significantly improves core plasma performance and power exhaust control, modifies the plasma flux dynamics and ELM properties at the walls. It is then shown, that the global Be erosion by sputtering is not significantly affected by Ne seeding. This is because the time-averaged fluxes are dominated by the inter-ELM phase, where the main erosion precursor is deduced to be D+. Furthermore, we observe a beneficial strong decrease of the fluxes at the OMP when slightly increasing the Separatrix-limiter clearance, indicating short far scrape-off layer (SOL) decay lengths. We also describe, how the fluxes change with average plasma density and toroidal magnetic field and plasma current. Regarding the fluxes at the UDP, we show the critical effect of the magnetic topology at the top of the chamber. The formation of a secondary Separatrix in this region enhances the parallel plasma fluxes to the UDP, which eventually becomes the main plasma-wall interaction area. Finally, we compare the experimentally estimated Be effective sputtering yields at the OMP for the inter- and intra-ELM phases with the values calculated using the SDTrim code to infer information on the mean ion impinging energies Ei and to show that the main global erosion precursor is D+. The obtained Ei values are in the range or at least compatible with those expected for the plasmas analysed: approximately 30 eV for the inter- and 1 keV for the intra-ELM periods respectively.

Centre for Energy, Environmental and Technological Research·Consorzio RFX·Culham Centre for Fusion Energy·Forschungszentrum Jülich

Measurements and analysis of short-term activation of ITER samples exposed in irradiation-end during DT operations at JET

Ewa Łaszyńska, Jakub Piotr Włodarczyk, Sabahattin Akbas, Barbara Bieńkowska, Zamir Ghani, Callum Grove, Xavier L Litaudon, Lee William Packer, Chantal Shand, Rosaria Villari, et al.

Plasma Physics and Controlled Fusion·Aug 21, 2026Materials & Plasma-Facing ComponentsBlankets & NeutronicsAI, Modeling & Simulation

Short-lived activation products in fusion materials are critical for assessing radiological safety and maintenance planning in future fusion reactors like ITER. These products influence shutdown dose rates, impact maintenance procedures, and provide essential validation for nuclear data and activation models in the short term. Understanding their formation and decay behavior under deuterium-tritium (DT) plasma conditions allows for more accurate predictions of material performance and safety constraints. This study analyzes the short-term activation of ITER materials exposed at the KN2 6U irradiation end in the JET (Joint European Torus) tokamak during the DTE2 and DTE3 campaigns. CuCrZr, IWS A286, and SS316L(N) samples were activated during DTE2, while tungsten monoblock, a key ITER divertor material, was irradiated during DTE3. Activated samples were promptly transported via a pneumatic post system for gamma spectrometry measurements using UKAEA’s 190% relative efficiency HPGe (High-Purity Germanium) detector. The efficiency calibration was performed with the MCNP 6.1 code, employing an MCNP detector model validated by a certified multi-gamma calibration source. The experimental data were compared with FISPACT-II activation calculations, enabling the determination of calculation-to-experiment (C/E) ratios. This study provides key validation data for nuclear activation models and improves the accuracy of computational predictions for material behavior in fusion reactors.

French Alternative Energies and Atomic Energy Commission·United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·Institute of Plasma Physics and Laser Microfusion

Aug 19

Determining neutral fueling response to RMPs in MAST using a multi-reservoir particle balance model and EMC3 EIRENE

Kurt Flesch, James Harrison, Andrew Kirk, Ian Waters, Heinke Frerichs, Oliver Schmitz, Livia Casali

arXiv (physics.plasm-ph)·Aug 19, 2026Plasma & ConfinementAI, Modeling & Simulation

The application of certain configurations of resonant magnetic perturbations (RMPs) has been shown to cause a plasma pump-out in both L- and H-mode discharges at MAST. In this paper we discuss the impact of neutral fueling on this density reduction. The neutral fueling and average particle confinement time $τ_p$ of the main ion species were calculated using a 0-D particle balance analysis. When the RMPs were applied, it was found that there was an increase in ionizations and a 15$\%$ reduction in $τ_p$ for L-mode discharges and a similar result for inter-ELM (edge localized modes) periods of H-mode discharges. A time-dependent global multi-reservoir particle balance (MRPB) was developed, which included atomic and molecular reservoirs, to further investigate the role neutrals had on the density change. We discuss how this model was able to accurately reproduce the experimentally measured density reduction and ionization increase due to either a reduction in $τ_p$ or a reduction in particle fueling efficiency. Results from EMC3-EIRENE modeling indicate this change could be attributed to neutral particle fueling occurring in locations with now-opened field lines due to the chaotic edge-region from RMP applications.

University of Wisconsin Madison·Culham Centre for Fusion Energy

Aug 14

The physics of ballooning-limited ELM-free regimes in EUROfusion tokamaks

Mike G Dunne, Michael Faitsch, Olivier Sauter, Eleonora Viezzer, Benoit Labit, Athina Kappatou, David Keeling, Branka Vanovac, Itziar Balboa, Petra Bilkova, et al.

Nuclear Fusion·Aug 14, 2026Plasma & Confinement

The development of operational scenarios without large Type-I ELMs is of utmost importance for the stable operation and longevity of future tokamaks. The EUROfusion tokamak exploitation program has therefore made the understanding of ELM-free regimes a major topic of exploration across all its contributing devices (ASDEX Upgrade, JET, MAST-Upgrade, TCV, and WEST). An integrated program to investigate a range of Type-I ELM-free regimes has been developed covering the enhanced D-alpha (EDA), magnetic perturbations (MP), negative triangularity (NT), quasi-continuous exhaust (QCE), quiescent H-mode (QH), the baseline small ELMs (SE), I-mode, and X-point radiator (XPR) regimes. This contribution focuses on the development and understanding of the NT and QCE regimes on ASDEX Upgrade, JET, and TCV. The importance of transport via ballooning modes in both regimes is highlighted, as well as the progress in developing access models based on ideal-MHD. In the case of the QCE, this can also be expressed as a minimum separatrix density, which corresponds well to experimentally measured separatrix densities. Particular focus is paid to the performance of the QCE in terms of the achieved pedestal top values, which, when appropriately normalised, do not differ significantly from ELMy H-mode plasmas. This, combined with the predicted minimum separatrix density for the 15~MA ITER baseline plasma, highlight the relevance of the QCE as a potential operational scenario for both ITER and future reactors.

Czech Academy of Sciences·Swiss Federal Institute of Technology Lausanne·Culham Centre for Fusion Energy·Max Planck Institute for Plasma Physics

Aug 7

Detailed modelling of alpha transport due to ELM control fields in ITER: implications for PFCs and diagnostic design

Fabio Camilo de Souza, Kenneth G McClements, Alexander Philip Kofi Prokopyszyn, Antti Snicker, Alex Reyner-Vinolas, Javier Gonzalez Martin, Lucia Sanchis, Rafael Marqués Gómez

Nuclear Fusion·Aug 7, 2026Plasma & ConfinementControl & DiagnosticsMaterials & Plasma-Facing ComponentsAI, Modeling & Simulation

This paper presents simulations of alpha-particle transport in ITER driven by static 3D magnetic field perturbations in the high-performance 15 MA Q = 10 baseline scenario, specifically resonant magnetic perturbations (RMPs) arising from edge-localized mode (ELM) control coils, combined with toroidal field ripple (TFR) and effects from ferromagnetic materials. We employ the Lorentz-Orbit Code for Use in Stellarators and Tokamaks (LOCUST), which tracks fast-ion orbits under the Lorentz force and Monte Carlo collisions with the bulk plasma, taking into account the detailed geometry of ITER plasma-facing components (PFCs). LOCUST uses GPU cards to enable the high-resolution modelling required to accurately resolve power fluxes across surfaces with complex morphologies, including unprotected cooling pipes beneath the dome divertor, and the generation of reliable synthetic diagnostics for the ITER Fast Ion Loss Detector (FILD) to support its design. The simulations include a range of ELM control coil current profiles with toroidal mode number n=3. The results indicate that the total alpha-particle energy loss has a negligible impact on plasma performance, remaining below 1% of the alpha energy produced in D–T reactions. Furthermore, the power flux density on the divertor structures and the first wall remains well below design limits and is comparable to thermal and radiative loads. The simulated alpha flux on the FILD scintillator plate is well above the noise threshold and can be significantly higher than conservative estimates.

Tokamak Energy·Culham Centre for Fusion Energy·University of Sevilla

Aug 4

Initial observations in X-point target divertor discharges on MAST-U

Nicola Lonigro, Kevin Verhaegh, James R Harrison, Bruce Lipschultz, Christopher Bowman, Fabio Federici, Jack Flanagan, Daniel Greenhouse, David Moulton, Peter Ryan, et al.

Nuclear Fusion·Aug 4, 2026Plasma & ConfinementMaterials & Plasma-Facing Components

The first high-power (≥ 3 MW) H-mode experiments using a double-null X-point-target (XPT) divertor configuration have been performed on MAST-U. The XPT geometry is obtained by combining a large strike point radius, similar to the Super-X divertor (SXD), with an additional X-point near the separatrix in the baffled outer divertor chambers and leads to additional exhaust benefits over the SXD. The broader electron density profile near the secondary X-point leads to additional plasma-neutral interactions, evidenced by a broader hydrogenic emission profile, and resulting in larger power and ion sinks. The increase in plasma-neutral interactions also leads to lower target electron temperatures and heat fluxes. These benefits appear to extend to transients, and preliminary evidence of improved ELM buffering in the XPT is presented. These results showcase how multiple alternative divertor configuration strategies can be combined to improve momentum, power, and particle losses, which may be required for the challenging exhaust conditions of future reactors.

United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy·Oak Ridge National Laboratory

Jul 24

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

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.

Plasma Physics and Controlled Fusion·Jul 24, 2026Plasma & ConfinementControl & DiagnosticsAI, Modeling & Simulation

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.

Princeton Plasma Physics Laboratory·United Kingdom Atomic Energy Authority·Culham Centre for Fusion Energy