Recent Publications

Analysis of background plasma behavior under external fields in the low energy beam transport section of LIPAC

Tomonobu Itagaki, Tomoya Akagi, Kai Masuda, Nicolas Chauvin, Luca Bellan, Fabio Cismondi, Hervé Dzitko, Yann Carin

QST, Fusion for Energy, CEA, INFN

Background plasma behaviour in the low energy transport section of the LIPAc is analysed with 3-dimensional particle-in-cell simulation. The simulation showed some characteristic effect under external field elements in the LEBT: a positive biased chopper absorbed the electron plasma in the surrounding drift region. Such absorption by the chopper is limited spatially by the solenoid lenses. However, some extent of electron plasma flows through the solenoid. Analytically estimated flux of the electron leakages through solenoid is compared to the simulation results. Results of experiments still show mysterious transient motion of beam pulse shape that is not shown in the simulations, while saturated state is roughly reproduced in it. Additionally, we developed a one-dimensional electron fluid simulation to investigate effects by thermalization of the background electron plasma, but this also could not reproduce the transient motion observed in the experiment. As a secondary outcome, the fluid simulation showed qualitatively close results to that of the particle in cell simulation in saturated states with much smaller computational resource.

Progress of LHCD experiment at 4.6 GHz towards long-pulse and high-power operation on EAST

Miaohui Li, Weiwei Zhang, Junlin Chen, Chenbin Wu, Bojiang Ding, Mao Wang, Liang Liu, Lianmin Zhao, Wendong Ma, Yong Yang, et al.

Chinese Academy of Sciences - Hefei Institutes of Physical Sciences, Institute of Plasma Physics Chinese Academy of Sciences, CEA

Recent achievements of lower hybrid current drive (LHCD) experiment at 4.6 GHz towards long-pulse and high-power operation on EAST tokamak are reported. The duration of long-pulse plasmas has been extended to 1056 s with 1.1 MW LH power in I-mode and 1066 s with 0.92 MW in H-mode. The operational domains in plasma current and line-averaged density for fully non-inductive discharges with LH alone and with the combined LH and electron cyclotron (EC) waves are presented. The dependence of LHCD efficiency on plasma density in both L- and H-mode discharges characterized by residual loop voltage Vloop = 0 is quantified. It is found that the LHCD efficiency is improved significantly by EC heating due to the increase of electron temperature. The dominant issues in long-pulse and high-power operation are summarized and discussed, including the power coupling imbalance, the hot spot and arc events in front of the antenna, and the deteriorated plasma heating effect with high LH power. Finally, prospects with a new 4 MW LHCD system at 4.6 GHz which is under development are given.

Direct comparison of 3D non-linear JOREK simulations of shattered pellet injection with ASDEX Upgrade experiments

Weikang Tang, Matthias Hoelzl, Paul Heinrich, Di Hu, F J Artola, Pascal de Marne, Mathias Dibon, Mike G Dunne, Ondřej Ficker, Peter Halldestam, et al.

Max-Planck-Institut für Plasmaphysik, ITER Organization, Beihang University, Institute of Plasma Physics Czech Academy of Sciences, CEA

Shattered pellet injection (SPI) as primary mitigation method for major disruptions in ITER has a large parameter space available for optimization including the total amount of injected material, the size of the individual pellet fragments, the material composition, and the timing of multiple injections. This flexibility needs to be exploited to simultaneously minimize thermal heat loads, electromagnetic vessel forces, and formation of relativistic electrons and their impacts on plasma facing components. In this article, we apply 3D non-linear magnetohydrodynamic modelling to SPI experiments in the ASDEX Upgrade tokamak, going beyond our previous work [Tang et al Nucl. Fusion 65 116003 (2025)] by resolving some discrepancies between simulations and experiment and carrying out direct qualitative and quantitative comparisons to experimental measurements. The key element that enables the transition is the incorporation of the parallel heat-flux limit, which is done here in a simplified form. The work increases the confidence of reproducing key processes of disruption mitigation in direct 3D non-linear simulations in view of future predictive studies for ITER.

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

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.

Development of tungsten actively cooled divertor target plasma facing components for W7-X and JT-60SA fusion devices

Marianne Richou, Mehdi Firdaouss, Joris Fellinger, Thierry BAFFIE, Bernd Böswirth, Mariano Di Bartolomeo, Diogo Dias Alexio, Daniel Dickes, Daniel Dorow-Gerspach, Martin Draksler, et al.

CEA, Max-Planck-Institut fuer Plasmaphysik, Institut Jozef Stefan, Univ. Grenoble Alpes, Laboratoire Georges Friedel

The present study describes the ongoing developments for plasma-facing components (PFCs) intended for future fusion devices, requiring PFCs to handle high heat loads of at least 10 MW/m² in the divertor region. tungsten (W) is chosen as the armor material due to its resistance to plasma-wall interactions and high heat fluxes (HHF). Significant efforts are underway on various fusion devices (WEST, EAST, KSTAR…) to test and optimize water-cooled W-based PFCs. For W7-X, new PFCs are being developed, using W armor material placed at the divertor target region. The design aims to simplify manufacturing, inspection, and installation processes compared to the currently operated actively cooled carbon fiber composite based divertor, while meeting thermal and mechanical constraints. The developed concept and related manufacturing choices are promising since, we show that it is able to sustain the required heat loads. Also, the JT-60SA tokamak plans a transition to metallic PFCs after 2029. The W monoblock concept, successfully used in WEST, EAST and KSTAR and planned for ITER, is being adapted for JT-60SA, and is currently the reference concept. However, some advanced designs exploring cost-saving measures and enhanced heat transfer capabilities are also under development. Manufacturing processes, such as laser powder bed fusion (LPBF) and hot isostatic pressing (HIP), are optimized to ensure reliable cooling channels for the development of the proposed enhanced designs. For these developments, thermo-hydraulic analyses and HHF tests have shown promising results, with surface temperatures being in agreement with models and material temperature limits. These developments lead to show a successful use of combined advanced manufacturing processes (additive manufacturing…), design hypotheses and adapted modelling tools to propose components with relevant performances.

Evolution of SPI-induced disruptions in ASDEX Upgrade

Paul Heinrich, Gergely Papp, Stefan Jachmich, F J Artola, Matthias Bernert, Pascal de Marne, Mathias Dibon, Ralph Dux, Thomas Eberl, Ondřej Ficker, et al.

Max-Planck-Institut für Plasmaphysik, ITER Organization, Czech Academy of Sciences - Institute of Plasma Physics, TU Wien, Commissariat `a l’ Energie Atomique (CEA)

Disruptions are a major concern for future fusion reactors based on the tokamak principle. To ensure machine protection, the thermal loads and vessel forces that arise during disruptions have to be mitigated reliably. For the ITER disruption mitigation system (DMS), the shattered pellet injection (SPI) technology has been selected. It can provide a prompt delivery of the injection material into the plasma core, with the mitigation efficiency depending on fragment size and velocity. A highly flexible SPI system was built and installed at the tokamak ASDEX Upgrade (AUG) to aid the finalization process of the ITER DMS and provide crucial input for modeling. The SPI-induced disruptions in the 2022 AUG experiments follow a typical chain of events, which are discussed in this paper: The first light (FL), main fragment arrival (MFA), plasma movement event (PME), MARFE, thermal quench (TQ)/plasma current spike (IP-spike), current quench (CQ), and vertical displacement event (VDE) phase. Depending on the injection parameters, these phases may vary significantly or some might not be present at all. In this paper, we will focus on the characterization of these disruption phases and figures of merit for the mitigation efficiency, depending on the SPI configuration. With increasing amount of assimilated neon in the plasma - primarily influenced by the neon content in the pellet but also the shattering parameters - the disruptions exhibit different behaviors. This disruption evolution seems to be a continuous process, with the most prominent feature being the changing disruption time scales and plasma current time trace shape during the CQ from convex (poorly or unmitigated) → concave (well mitigated/radiation dominated). Depending on the injection, pre-TQ durations between 15 - 0.5 ms and early CQ durations (Δt CQ 100 → 80 ) between 13.3 - 8.2 ms had been achieved at AUG.

Development of advanced technologies for lithium purification

Aryles Slimani, Leïla Gicquel, Nicolas Auriac, Jean-Louis Courouau, Julien Aubert

CEA

Activation analysis on diagnostic windows and coating materials for HCPB and WCLL blankets in EU DEMO

Sabahattin Akbas, Barbara Bieńkowska, Ewa Łaszyńska, Jakub Piotr Włodarczyk, Matthew Lukacs, Sandrine Rosanvallon, Joelle Elbez-Uzan

Institute of Plasma Physics and Laser Microfusion, EUROfusion, CEA, AGH University of Krakow, UKAEA

Plasma diagnostics and reactor control systems in future fusion power plants, such as DEMO, will rely on optical windows to monitor key plasma characteristics under intense irradiation conditions. Selecting suitable window and coating materials such as quartz, fused silica, sapphire, HfO₂, and MgF₂ is therefore critical to ensuring component reliability and safety. This study investigates these candidate materials’ activation calculations in the DEMO environment. Using the MCNP code, neutron spectra have been calculated at window-relevant locations in the outboard equatorial port and upper port limiters, considering both Helium-Cooled Pebble Bed and Water-Cooled Lithium Lead breeding blankets. Activation analyses have been subsequently performed with the FISPACT-II inventory code to evaluate the activity, decay heat, contact dose rates, and dominant contributing nuclides under the initial DEMO irradiation scenario. The results provide insights into neutron irradiation effects relevant to the activation performance of window materials and coatings, and contribute valuable input to component design and safety evaluations within the EUROfusion Safety & Environment Work Package framework.

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.

United Kingdom Atomic Energy Authority, Institute of Plasma Physics and Laser Microfusion, AGH University of Krakow, CEA, ENEA

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.

Impact of ICRH resonant layer position on core plasma performances in WEST H-minority plasmas

Samuele Mazzi, J Morales, Ernesto Lerche, Laurent Colas, Julien Hillairet, Hugo Corvoysier, Remi Dumont, Nicolas Fedorczak, Jeronimo Garcia, Philippe Huynh, et al.

CEA, LPP, ERM/KMS, Technical University of Denmark, Institut Jean Lamour

A systematic experimental study of hydrogen-minority ion cyclotron resonance heating (ICRH) has been performed in deuterium plasmas in WEST to optimize the power deposition. Dedicated scans of the radial position of the ion cyclotron resonant layer and of the minority concentration were carried out while keeping global plasma parameters and antenna coupling nearly constant. The maximum performance is obtained when the resonant layer is shifted towards the high-field side (HFS) by around 7 cm with a hydrogen minority concentration between 5 and 8%. A pronounced degradation occurs for larger inward displacements, whereas the performances are almost halved when the resonance is located in the low-field side. No clear dependence of turbulence characteristics on the resonance position is observed, indicating that the confinement variations are primarily governed by wave absorption and fast-ion confinement modifications. Infrared thermography and calorimetry measurements show that fast-ion ripple-induced losses are minimized for resonance positions beyond a certain value on the HFS, revealing a trade-off between optimal absorption and wall load mitigation. These results demonstrate that coordinated control of the resonant layer position and minority concentration allows simultaneous optimization of ICRH efficiency and fast-ion wall loads in a fully metallic reactor-relevant environment.

European test blanket modules: An overview of fabrication technologies development and EUROFER97 steel supply

Milan Zmitko, Yves Poitevin, Laurent Forest, Antonella Li Puma, Rémi Boullon, Jérôme Tosi, Denis Sornin, Philippe Emonot, Emmanuel Rigal, Laurence Cogneau, et al.

CEA, Université Paris-Saclay, Univ. Grenoble Alpes, APAVE Exploitation France, ALSYMEX

Neutronics, radiolysis and OSCAR-Fusion analysis for the design of a flow loop for Activated Corrosion Products studies

S. Noce, G. Bailey, L. Butt, Y.-L. Chiu, A. Colangeli, F. Dacquait, M. Damiano, D. Flammini, N. Fonnesu, M.R. Gilbert, et al.

ENEA, UKAEA, The University of Birmingham, CEA, University of Rome “Tor Vergata”

Overview of material migration and erosion experiments in the full-tungsten WEST tokamak during Phase 1 and Phase 2 operations

A. Hakola, M. Diez, N. Fedorczak, J. Gaspar, E. Tsitrone, M. Balden, Y. Corre, S. Di Genova, A. Huart, C. Martin, et al.

CEA, National Centre for Scientific Research ‘Demokritos’, VTT Technical Research Centre of Finland Ltd, Max-Planck-Institut für Plasmaphysik, Ruđer Bošković Institute

This paper gives an overview of erosion and migration studies of tungsten (W) in the WEST tokamak during its Phase 1 (2016–2021) and Phase 2 (from 2022) experimental campaigns with a focus on plasma-facing components (PFCs) at the divertor. In Phase 1, gross erosion of PFCs is in line with observations from other major fusion devices and attributed to low- Z impurities in the plasma. In addition, a strong asymmetry is observed between the high- (inner) and low-field (outer) side divertor targets, in favour of the inner side. Net erosion at rates of <0.5 nm s −1 is measured around the strike points while the remaining areas are dominated by net deposition. The thickest deposited layers (up to 50 μ m) with the most complex structures result from a cumulated plasma exposure of ∼7 h. The overall erosion-deposition pattern is further influenced by the strong magnetic ripple of WEST, which can result in almost an order of magnitude difference between the maxima and minima of the ripple. In Phase 2, increasing plasma fluence leads to the deposits growing to hundreds of micrometres in thickness. At the same time, erosion proceeds at a constant rate and can reach values up to 30 µ m in ∼18 h of plasma time. In the main chamber, erosion is weaker than at the divertor but especially at low densities it can result in notable transport of W into the core. In addition, upon switching on the ICRF antennas, W sputtering on the close-by limiter structures can increase by a factor of more than 10. Modelling is able to catch many of the observed phenomena in Phase 1, with the exception of the inner–outer asymmetry and the formation of the thick deposits. In contrast, the patterns during the high-fluence operations in Phase 2 require more work to be reproduced.

Validation of the GBS code against COMPASS experimental data

P. Macha, J. Seidl, D. Oliveira, D. Galassi, K. Lim, D. Tskhakaya, J. Horacek, J. Adamek, J. Cavalier, P. Bilkova, et al.

Institute of Plasma Physics of the CAS, EPFL, Czech Technical University in Prague, CEA, Nanyang Technological University

This work presents the first full-size turbulence simulation of the COMPASS tokamak edge with the three-dimensional, self-consistent, flux-driven, two-fluid Global Braginskii Solver (GBS) code. The simulation is validated against experimental data from a standard L-mode COMPASS discharge, covering the scrape-off layer (SOL) from the outer midplane to the divertor region. The dataset combines four diagnostic systems and provides 43 observable profiles, including electron temperature and plasma potential measurements that enable the first direct comparison of their fluctuations and higher-order statistics in COMPASS. The validation results show good agreement between simulation and experiment in profile shape, magnitude, and fluctuation levels for most outer midplane observables. The parallel heat-flux fall-off length is also reproduced. In the divertor, the mean profiles agree reasonably well in magnitude, but the simulated fluctuation levels remain significantly underestimated. Advanced blob analysis shows that both simulation and experiment lie mainly in the sheath-connected regime, with partial presence in the connected ideal-interchange regime. The remaining discrepancies in the COMPASS simulation are probably a consequence of simplified parameters and boundary conditions, reflecting limitations that are common to fluid turbulence models. Overall, the validation demonstrates that GBS captures the dominant upstream SOL turbulence dynamics, while identifying divertor physics as the main target for further improvement.

Boronization-enabled I-mode on EAST tokamak with an expanded density window and favorable-configuration access

X. M. Zhong, X. L. Zou, A. D. Liu, L. Q. Xu, B. Zhang, C. Zhou, J. P. Qian, X. Z. Gong, Y. T. Song, G. Zhuang, et al.

Institute of Plasma Physics, Chinese Academy of Sciences, University of Science and Technology of China, CEA

I-mode is a promising confinement regime for future fusion reactors because it combines enhanced energy confinement with L-mode-like particle transport and naturally ELM-free operation. Previous EAST I-mode studies were performed exclusively under lithium-conditioned wall conditions. Here we report the first systematic experimental investigation of I-mode under boronized wall conditions on EAST and compare it with an existing lithium-conditioned I-mode database at the same toroidal field, $B_t = 2.47$\,T. The boronized-wall dataset exhibits a substantially broader accessible density range, with the Greenwald fraction extending from $f_{\mathrm{GW}} = 0.26 - 0.77$ , compared with $f_{\mathrm{GW}} = 0.35 - 0.54$ under lithiation. A higher normalized $\mathrm{D}_α$ emission suggests that enhanced edge recycling may contribute to this density extension. A striking increase in favorable-configuration I-mode is also observed: $51\%$ boronized-wall discharges are obtained in favorable-configuration, compared with only $8\%$ lithium-conditioned discharges. These favorable-configuration cases are concentrated at high density and exhibit a deeper radial electric-field($E_r$) well and stronger $\mathbf{E_r}\times\mathbf{B}$ velocity shear. When ETRO is present, the associated transition between electron and ion turbulence is similar under the two wall conditions, although ETRO occurs less frequently ($15\%$) under boronization. An empirical EAST I-mode energy confinement scaling at fixed $B_t$ is obtained, $τ_E = 3.29 I_p^{0.51 \pm 0.10} P_{\mathrm{loss}}^{-0.53 \pm 0.05} \bar{n}_e^{0.08 \pm 0.07}$, indicating weaker power degradation than IPB98(y,2) H-mode scaling and a weak density dependence. These results show that boronization can broaden the operational space of EAST I-mode and support the development of reactor-relevant ELM-free scenarios.

Activation analysis and waste assessment of the electron cyclotron in EU-DEMO

S. Akbas, B. Bienkowska, E. Laszynska, J. Wlodarczyk, S. Rosanvallon, J. Elbez-Uzan

Institute of Plasma Physics and Laser Microfusion, EUROfusion, CEA, AGH University of Krakow

Analysis of the alpha particle transport induced by Alfven eigenmodes and fish-bones in JET DT discharges: multiple energetic particle simulations

J. Varela, J. Garcia, D. Zarzoso, S. Mazzi, Y. Kazakov, Z. Stancar, M. Baruzzo, J. Ongena, D.A. Spong, L. Garcia, et al.

United Kingdom Atomic Energy Authority, University of Texas at Austin, Universidad Carlos III de Madrid, CEA, Oak Ridge National Laboratory

The present study is dedicated to analyze the extensive fast ion (FI) induced perturbations in ion cyclotron resonance heating (ICRH) dominated plasmas in JET deuterium–tritium (D–T) discharges, the closest experiment to reactor-like operation. A set of nonlinear simulations are performed with the gyro-fluid FAR3d code to reproduce the FI induced activity and energetic particle (EP) transport observed in the discharge 99896 by multiple EP populations. Well validated simulations with JET D–T diagnostics show that Toroidal Alfvén eigenmodes (TAEs) and fish-bone are destabilized by ICRH accelerated passing D and trapped H ions, respectively. Multiple EP simulations show important nonlinear couplings between EP species, leading to a strong effect on the alpha particle confinement even in the alpha particles do not destabilize EP modes. Simulations including trapped H and alpha particles indicate fish-bones may induce alpha particle losses up to 3 % . The situation is different with respect to TAEs, which have a smaller impact on the alpha particle transport, leading to negligible losses. In addition, the generation of shear flows during the saturation phase of TAE and fish-bones is observed in the simulations. Shear flows induced by TAEs in the inner-middle plasma region are almost two times larger compared to the shear flows caused by the fish-bone in the inner plasma region. Shear flow generation by AE/fish-bones may have an important impact in the thermal plasma and alpha particle confinement of devices as ITER and JT60SA. Consequently, nonlinear effects in multiple EP populations must be considered in the analysis of EP transport. It is shown that this is a fundamental result that has to be considered to be properly evaluate fusion power in reactor relevant plasmas.

Towards digital twins of fusion systems

F. Jenko, J. Ball, D. Borodin, C. Bourdelle, G. Ciraolo, J.E. Cook, J.M. García-Regaña, T. Görler, M. Hoelzl, F. Imbeaux, et al.

Max Planck Institute for Plasma Physics, CEA, Ecole Polytechnique Fédérale de Lausanne (EPFL), Forschungszentrum Jülich GmbH, United Kingdom Atomic Energy Authority

Bridging the gap to next-step devices—while saving valuable time and resources—requires more than semi-empirical models, which struggle to predict plasma behavior in unexplored parameter regimes. Instead, validated simulation tools are essential, leveraging high-fidelity exascale computing and multi-fidelity models, including AI-based surrogates. To address these challenges, the ‘EUROfusion Theory and Advanced Simulation Coordination (E-TASC)’ initiative was launched in 2021. It includes 15 TSVV (Theory, Simulation, Verification, and Validation) projects supported by five Advanced Computing Hubs. This ‘team of teams’ has made substantial progress toward developing digital twins of fusion systems, with key scientific achievements to be presented in the paper. This includes the following topical areas: core performance in burning plasmas, magnetohydrodynamic transients, L–H transitions and ELM-free regimes, plasma exhaust, and plasma–wall interactions—with applications to both tokamaks and stellarators.

Comparative analysis between particle tracing model and thermographic data for a MeV class beam for fusion application

Antonio Pimazzoni, Piero Agostinetti, Giuseppe Chitarin, Giulia Emma, Mieko Kashiwagi, Kisaki Masashi, Nicolo Marconato, Isabella Mario, Basile Pouradier-Duteil, Beatrice Segalini, et al.

Università degli Studi di Padova, Consiglio Nazionale delle Ricerche, Consorzio RFX, ENEA, INFN

For the ITER experimental fusion reactor up to 3 heating neutral beams (HNBs) are foreseen. Each HNB will generate 1280 individual beamlets of H-/D- ions which will be accelerated up to 870 keV/1 MeV, neutralized and focused into the tokamak plasma, to provide up to 16.5 MW of heating power. To this purpose, divergence and aiming of each beamlet are required to be ≤7 mrad and within ± 2 mrad, respectively. An accurate compensation is thus essential, both for the magnetic deflection induced by the magnets embedded in the extraction grid (necessary to immediately deflect the co-accelerated electrons) and for the Coulomb repulsion among the beamlets. The Asymmetric Deflection Compensation Magnets (ADCM), the solution developed for the full-scale ITER HNB prototype called MITICA, was recently tested, for the first time on a MeV class beam, at the MegaVolt Test Facility (MTF) at QST (Naka, Japan). In these experiments, a MITICA-like extraction grid was built and installed on MTF. In this work, the beam pattern measured on a calorimeter made of unidirectional carbon fiber composite (1D-CFC) is compared with the predictions from simulations by the numerical code IBSimu, in order to develop methodologies and identify limitations for its application for MITICA exploitation and improvements. The model accuracy varied with experimental conditions; in all the cases however the proposed methodology proved very valuable in providing trends and dependencies.

HL-3 research towards high-performance plasma and power exhaust solution

Wulyu Zhong, Xiao Quan Ji, Wei Chen, Xingyu Bai, Alain Becoulet, Jerome B Bucalossi, Zhe Chen, Shaoyong Chen, Zhipeng Chen, Yihang Chen, et al.

Southwestern Institute of Physics, CEA, UKAEA, Huazhong University of Science and Technology, National Institute for Fusion Science

The HL-3 tokamak program addresses critical challenges in developing integrated high-performance scenarios compatible with power exhaust demands for ITER and future reactors. Through systematic facility enhancements including auxiliary heating upgrades to 19.5 MW and AI-enabled control systems achieving 95.5% disruption avoidance, HL-3 finished a new round of exploration and validation for high-performance operation and power exhaust solution. Experiments at mega-ampere plasma currents demonstrated a reactor-relevant hot-ion regime with core ion temperatures exceeding 10 keV and a fusion triple product of . High normalized beta scenarios featuring internal and double transport barriers were successfully established. To address the compatibility of the plasma boundary, various small- or no-edge localized mode (ELM) regimes were achieved, including the enhanced D-alpha (EDA) H-mode, quiescent H-mode (QH-mode), and the quasi continuous exhaust (QCE) regime. Investigations into power exhaust integration highlighted the efficacy of advanced divertor configurations, specifically snowflake and tripod geometries, in significantly reducing peak heat fluxes. Active ELM control was demonstrated via resonant magnetic perturbations (RMPs), lower hybrid waves (LHW), and impurity supersonic molecular beam injection (SMBI), complemented by real-time feedback control of divertor detachment. These developments establish the fundamental physics basis and technical foundations necessary for demonstrating and extrapolating high-performance operations to ITER and next-step devices.