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Woong-Ryol Lee, Taehyun Tak, Luca Zabeo, Timo Ravensbergen, Anze Zagar, Piotr Perek, Walter Van Herck, Peter De Vries, Seung-Ju Lee, Lana Abadie, et al.
T W Morgan, Martin Balden, Jos Scholte, Selanna Roccella, J H You
ITER is anticipated to operate with occasional slow transients, resulting in an increase in expected heat loads to the divertor from 10 to 20 MW m −2 and thus in the surface temperature increasing to > 2000 ◦C. At the same time mitigated ELMs may be expected to strike the surface. To investigate the consequences of this, two sets of experiments were carried out in Magnum-PSI, representing the first experiments to explore this regime. In the first set, seven ITER-like tungsten monoblocks were exposed to either hydrogen plasma, or impurity-seeded hydrogen plasma at a surface temperature of 2050 ◦C. Simultaneously five of the seven monoblocks were exposed to 10 5 ELM-like pulses using a 1 ms duration laser at an energy density of 0.1 − 0.19 MJ m −2 . Very strongly roughened and cracked laser exposed areas with localized melted regions were observed. These areas protruded several hundreds of micrometers above the original surface. In the second experiment, three tungsten plates, with either small, large or no castellations cut into the surface, were exposed to a hydrogen plasma at 2100 ◦C with a wider range of ELM-like pulse numbers (10 2 − 10 5 ) and energies (0.13 − 0.31 MJ m −2 ). The surface was found to evolve by roughening and pre-crack formation towards similar strongly protruding roughened structures as pulse number increased, while increasing the transient energy resulted in increasing the fraction of surface melting observed in the loaded region. Mass loss measurements of the plates indicated that erosion was significant, between 0.05-0.13 nm per pulse for the different plates. Extrapolating these results to ITER suggests that this mass loss rate would be relatively high and would constitute a significant tungsten source.
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.
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.
Gwang-Geun Seo, Jong-Kyu Park, S. J. Han, M. Dubrov
Reliable correction of nonaxisymmetric error fields (EFs) is essential to the safety and performance of tokamak operation. Although resonant error field correction (EFC) is well established, supported by an improved understanding of 3D plasma response, the residual fields left after resonant EFC remain an open question: they are expected to be predominantly nonresonant, yet can still degrade both confinement and stability. Here we introduce a systematic EFC scheme that minimizes resonant and nonresonant EF effects simultaneously, based on neoclassical torque response computed from self-consistent perturbed equilibria. The scheme extends the method developed for designing quasi-symmetric magnetic perturbations to the case where actual error fields are present. Application to standard ITER target plasmas with a range of intrinsic EF scenarios demonstrates the advantages of this quasi-symmetric (QS) EFC scheme for controlling residual EFs. QS EFC consistently yields low-torque solutions while strongly suppressing resonant response, outperforming single-mode resonant overlap EFC in most cases and approaching the performance of multimodal resonant EFC. We also show that the QS EFC solution varies only tolerably between half- and full-$I_p$ ITER scenarios, despite the greater sensitivity expected from its higher-order nature.
Jiming Chen, Jianbao Wang, Qixiang Cao, Shen Qu, Zongyu Yang
This paper provides an overview of the recent progress in reactor-core fusion technologies within the vacuum vessel presented at the 30th International Atomic Energy Agency (IAEA) Fusion Energy Conference held in October 2025 in Chengdu, China, including plasma control, neutronics and nuclear data, tritium technologies, design and development of in-vessel components, related materials and intense neutron sources for material irradiation. Plasma control focusing on achievements for ITER and major fusion research devices are covered and the application of digital twin and Artificial Intelligence are highlighted. Research and Development (R&D) results from the International Thermonuclear Experimental Reactor (ITER) and current devices to future Demonstration Fusion Power Plant (DEMO) reactors are summarized, particularly new findings, data validation, design and analysis tool improvement, properties assessment and technical solutions for critical issues.
Anna Wójcik-Gargula, Grzegorz Tracz, Rojin Mehrara, Urszula Wiącek, Giovanni Mariano, Bruno Coriton, Ryszard Kantor, Dariusz Morawski, Marek Scholz
Youpeng Wang, Artem M. Dmitriev, Laurent Marot, Paul Hiret, Maitane Amarika, Gorka Beaskoetxea, Aitor Marco, Jordi Puig, Laura Sanchez Garcia, Ernst Meyer
The First Mirror Unit (FMU) is a critical front-end component of the collection optics in the ITER Core Plasma Thomson Scattering (CPTS) diagnostic, responsible for monitoring electron density and temperature. Large-sized metallic first (M1) and second (M2) mirrors are integrated into the CPTS FMU to collect and reflect optical signals from the fusion plasma to downstream optics. Due to its proximity to the plasma, M1 requires periodic in-situ cleaning to prevent severe optical degradation caused primarily by deposition. A simplified full-scale 1:1 mock-up of the FMU was manufactured to experimentally assess the radio-frequency (RF) plasma parameters over a wide frequency range of 13.56-80MHz using a retarding field energy analyzer. With increasing pressure and frequency, the magnitude of the direct current (DC) self-bias, mean ion energy, and energy splitting decreased, while the ion flux increased. In addition, particular attention was given to the DC grounding effect by introducing quarter-wavelength shorted stubs as notch filters (NF). In this preliminary RF circuit design, one or both mirrors could be RF-powered, with the remaining surfaces acting as grounded electrodes. Based on the selected parameters, M1 cleaning experiments with DC grounding or self-bias were conducted, followed by elemental composition and reflectivity analysis via X-ray photoelectron spectroscopy and spectrophotometry. The complete removal of 20 nm alumina on M1 was achieved at 13.56 MHz for both configurations, and substantial heating of RF components and the chamber was observed at all frequencies. Moreover, compared with the self-bias case, structural material could be deposited on both mirrors after cleaning with NF, owing to enhanced wall sputtering and weaker mirror cleaning. The feasibility of powering both mirrors simultaneously with self-bias was also demonstrated. Finally, significant power reflection from the load was confirmed using an Octiv V/I probe, highlighting the need for improved power transfer in the CPTS mirror cleaning system.
Daniela Kropáčková, Francesco Pastore, Ondrej Kudlacek, Timo Ravensbergen, Olivier Sauter, Guillermo Suarez-Lopez, Andrei Pshenov, Emiliano Fable, Davide Silvagni, Maximilian Reisner
The plasma density is a key parameter in tokamak operation, as it influences plasma performance and stability. Hence, accurate real-time estimation and control are essential. Real-time estimation of the plasma electron density profile can be maintained even in the presence of corrupted or unavailable diagnostics by employing model-based electron density observers. This work extends one of these observers, RAPDENS [1], by incorporating a heuristic scrape-off layer model capable of predicting the separatrix electron density, thereby building upon previous work that implemented a non-zero boundary condition at the separatrix [2]. Implementations for both ASDEX Upgrade (AUG) and ITER are presented. For AUG, the RAPDENS adaptation is tested using offline Extended Kalman Filter simulations of existing AUG discharges. These simulations show that RAPDENS is capable of providing a reasonable reconstruction of the electron density profile, supporting its future real-time implementation and routine use at AUG. For ITER, the implementation is based on and compared against results of a high-fidelity simulation performed with JINTRAC, the integrated modeling suite, for a start of research operation ramp-up scenario, concluding that the improved version of RAPDENS can be used for ITER density controller design.
Huasheng Xie, Ruohan Zhang, Xingyu Li, Feng Zhang, Zhengxiong Wang
Tokamak equilibria are commonly exchanged as gridded G-EQDSK files whose conventions, resolutions, and machine-specific formats impede cross-device comparisons and data-driven modeling. Here, we present VEQDB, an open, compact, and reconstructible fixed-boundary equilibrium database built on continuous MXH--Chebyshev geometry and independent physical-profile roots. By decoupling authoritative equilibrium physics from rectangular meshes, VEQDB enables continuous evaluation and metric differentiation at arbitrary application-demanded resolutions. Backed by an automated numerical validation pipeline, VEQDB is structured as an extensible repository for ongoing community expansion. Its inaugural release provides 13,291 accepted equilibria across 267 conventional and spherical tokamaks, encompassing parameter-sampled Grad--Shafranov solutions, G-EQDSK projections spanning EAST, MAST-U, and ITER scales, and controlled variation families with explicit provenance. Benchmark projections reproduce normalized flux maps with RMS errors between $1.09 \times 10^{-3}$ and $1.45 \times 10^{-3}$, while compact JSON representations achieve an 89--96-fold size reduction relative to standard $129 \times 129$ G-EQDSK files. The complete initial release occupies 41~MB in raw JSON and 18~MB in compressed archives, and all records successfully passed independent reload and evaluation tests. VEQDB establishes an extensible, provenance-preserving foundation for equilibrium studies, reduced-order surrogate modeling, and cross-machine workflows.
Jinwoo Gwak, Yeongsun Lee, Jeongwon Lee, Won Ik Jeong, Hyun-Tae Kim, Yong-Seok Hwang, Min-Gu Yoo, Yong-Su Na
Microwave breakdown driven by electron cyclotron (EC) waves provides a non-inductive route to plasma initiation in reactor-scale tokamaks. We introduce a three-dimensional Monte Carlo simulation that, for the first time, self-consistently treats nonlinear wave-particle interactions, atomic collisions, and guiding-center transport. The Monte Carlo simulation unveils the key role of parallel Brownian motion in the ionization avalanche mechanism. The predicted breakdown boundary is validated against KSTAR experiments. This work concludes that microwave gas breakdown will be successful under ITER-relevant conditions at a D$_2$ prefill pressure near 2 mPa with 1 MW of injected EC power.
Roman Ivanov, Alessandro Biancalani, Alberto Bottino, Didier Gossard, Thomas Hayward-Schneider, Alexey Mishchenko, Ruoyuan Wu
We investigate the interaction between energetic particles (EPs) and ion temperature gradient (ITG) modes using the global gyrokinetic particle-in-cell ORB5 code. This work expands the parameter space to a broader range of EP temperatures, explicitly focusing on the burning plasma regime and introducing a wider variety of EP distribution functions. While we give confirmation that direct dispersion relation modification (DDRM) stabilizes ITG modes at intermediate EP temperatures, our results demonstrate that the dilution effect (DE) independent of temperature becomes dominant over DDRM in the burning plasma regime (T f > 50\,T i ). Furthermore, we extend our studies from Maxwellian distribution functions of EPs to more experimentally relevant distributions like slowing-down. To validate these findings for future reactor operations, we analyze an ITER pre-fusion operation scenario, comparing EP stabilization against electromagnetic and kinetic electron effects. In this context, EP stabilization is found to be weaker than electron Landau damping and β-stabilization. Overall, these results provide better understanding of EP-ITG interactions over a wider range of EP parameters relevant to burning plasma regime which is important for predicting turbulence and confinement in future devices such as ITER.
Tomone Suwa, Takaaki Isono, Keiya Takebayashi, Yasuhiro Uno, Tsutomu Kawasaki, Masaru Kawabe, Tsutomu Hemmi
In developing the ITER central solenoid (CS) conductors, performance degradation due to electromagnetic loading cycles was identified as a critical issue. This issue was addressed by developing conductors having a short twist pitch, resulting in stable Tcs under electromagnetic loading cycles. While this degradation issue has been resolved, the contribution of the strain state of the Nb₃Sn strands within the conductor to its performance remains unclear, and the mechanism by which the short twist pitch mitigates Tcs degradation has not yet been clarified. Since the electromagnetic loading in DEMO will be larger than that in ITER, it is important to mitigate the strain in the conductor under electromagnetic loading in order to develop conductor whose performance can withstand the electromagnetic loading. To support the development of future conductors for DEMO, strain analysis based on the results of SULTAN test, CS insert test with internal strain measurement by neutron diffraction to investigate both the mechanism by which the short twist pitch mitigates performance degradation and the strain state under actual operating conditions. By integrating these results, it was clarified that the mitigation of Tcs degradation achieved by the short twist pitch cabling originates from strain suppression, especially through the reduction of bending strain. As a result, the short twist pitch design is expected to maintain its performance due to the reduction of the bending strain, and relaxation of axial compressive thermal strain is induced by hoop strain during the operation.
O. Kudláček, S. Stanek, W. Treutterer, I. Gomez Ortiz, P. Lang, R. Nouailletas, L. Pangione, T. Ravensbergen, M. Reich, J. Schmalz, et al.
Tomohiko Ushiki, Tatsuo Sugie, Shunsuke Tainaka, Shintaro Kimura, Kosuke Shimizu, Yuzi Katayanagi, Naoto Kasano, Butch Buenavidez, Hiroyuki Tachibana, Sin-iti Kitazawa, et al.
Atul Kumar, Dhyanjyoti Nath, Wouter Tierens, Jeremy D. Lore, Andrei Pshenov, Tom Wauters, Andrea Galvan, Davide Curreli, Syun'ichi Shiraiwa, Nicola Bertelli, et al.
Ion cyclotron radio-frequency heating (ICRH) is a key auxiliary heating system in ITER, but high-power RF operation can enhance plasma-material interactions through rectified RF sheath potentials on antenna structures and nearby plasma-facing components. We present the first predictive application of the STRIPE (Simulated Transport of RF Impurity Production and Emission) framework to assess RF sheath-driven tungsten (W) erosion and global impurity transport from the ITER ICRH antenna under ITER-relevant neon-seeded conditions. STRIPE couples SOLPS-ITER plasma backgrounds, full-wave RF sheath calculations, geometry-specific ion energy-angle distributions, sputtering physics, and three-dimensional impurity transport. Simulations predict RF sheath potentials of 1 to 3 kV on antenna limiter sidewalls, increasing gross W erosion by about a factor of 64 relative to thermal sheath conditions and producing a gross source of 3.34e18 W atoms per second. Erosion is governed by RF-modified ion energy-angle distributions together with local plasma flux rather than sheath voltage alone. About 10 percent of sputtered W is locally redeposited, giving a net source of 3.01e18 W atoms per second. The RF-induced antenna source remains about three orders of magnitude smaller than the thermal divertor source and more than two orders of magnitude smaller than the integrated thermal main-chamber source. After 100 ms, about 22 percent of the mobile W inventory resides within the SOLPS-covered confined-plasma region, corresponding to an annular W concentration of 1.70e-6. These results indicate that the ITER ICRH antenna is unlikely to dominate the total W source budget under the conditions considered and demonstrate the need for coupled modeling of RF waves, sheaths, sputtering, redeposition, and global impurity transport.
P. Platania, A. Simonetto, J.M. Arroyo, E. Carbonell, N. Casal, M. Choe, D. Farina, G. Garre, M. Henderson, M. Jimenez, et al.
Louis Puel, Eric Nardon, F J Artola, Di Hu
Runaway electron (RE) generation represents a major concern for future tokamaks such as ITER, where high plasma temperatures and large plasma currents significantly enhance the risk of RE formation. In particular, the hot tail mechanism, triggered during the thermal quench (TQ), occurs when a rapid drop in plasma temperature prevents high-energy electrons from thermalizing, allowing them to become REs under the accelerating effect of a strong toroidal electric field. In ITER, this mechanism is expected to be the dominant primary generation before avalanching in the case of an unmitigated/poorly mitigated disruption. The recently developed hot tail tracker framework in JOREK is used here to perform the first 3D estimates of hot tail generation during disruptions of an ITER 15 MA D–T H-mode scenario mitigated by Shattered Pellet Injection (SPI). Two situations are considered, with and without pre-disruption thermal energy degradation. The non-degraded case, more pessimistic, results in strong RE generation driven by weak stochastic losses limited by the trapped population, strong helical cooling, and the migration of hot electrons from the core toward regions of high parallel electric field. In contrast, the degraded case shows successful mitigation within the assumptions of this simulation, with suppression of the core hot tail generation and the removal of the edge seed through stochastic transport. The difference between a critical and a non critical scenario appears to be linked to localized overdensities of injected material produced by magneto-hydrodynamic (MHD) activity, which, counter-intuitively, significantly enhance local generation. This phenomenon is confirmed by a 0D two-fluid model of the bulk plasma during the TQ, coupled with a new 0D hot tail estimator based on backward tracking of a single particle in velocity space.
Shota Sugiyama, Takeo Nishitani, Hideaki Matsuura, Shuhei Sumida, Kouji Shinohara, Vitaly Krasilnikov, Bruno Coriton
We have investigated neutron emission anisotropy caused by deuterium beam injection in ITER deuterium plasmas. We evaluate the double-differential emission spectrum and emission anisotropy of neutrons produced by the D(d,n)3He reaction, using the energetic deuteron velocity distribution function obtained by following guiding-centre orbits of test particles. We clarify the dependence of neutron emission anisotropy on the electron density and temperature. Anisotropy increases with decreasing electron density and temperature. We examine the effect of neutron emission anisotropy on the neutron incident flux distribution on the first wall. We show that neutron emission anisotropy can affect the measurements of the neutron emission rate and emission profile based on the incident flux distribution, the distribution of neutron emission anisotropy, and the correspondence relationship between the poloidal angular positions and the neutron detectors that are planned to be installed in ITER. It is inevitable to consider neutron emission anisotropy for plasma diagnostics and monitoring of the neutron generation rate in beam-injected deuterium plasmas.
Jie Zhang, Florian Koechl, Alexei R Polevoi, Clarisse Bourdelle, Sunhee Kim, Alberto Loarte, Simon D Pinches, Ge Zhuang
The enhanced confinement of tokamak plasmas (H-mode) makes it a preferred regime for achieving fusion power production goals in future devices such as ITER. Nevertheless, low confinement mode (L-mode) remains worthy to investigate in reactor relevant conditions, primarily due to no/reduced requirements for ELM and divertor heat load control. In this regard, this study aims at exploring a new potential approach to maximise the achievable fusion gain Qfus. This approach attempts to increase the core density with enhanced pellet fuelling and then investigates the feasibility of high Qfus L-mode operation in ITER. The JINTRAC integrated modelling suite has been employed for core-edge transport and source modelling, using the HPI2 module for pellet fuelling. In some of the scenarios considered, the core density reaches up to ∼185% of the Greenwald density, nGW, with edge densities approaching nGW, motivated by recent re-evaluations of the density limit that suggest a power-dependent threshold. We compare core transport modelling results obtained by applying the semi-empirical Bohm-gyro-Bohm (BgB) or the quasi-linear gyrokinetic TGLF-SAT2 anomalous transport models, with interpretive vs. predictive impurity transport modelling, and pellet fuelling describing continuous vs. discrete particle sources. The core plasma confinement of high-density L-mode operation in ITER predicted by the TGLF-SAT2 model is significantly better than that predicted by the BgB model, resulting in a significantly improved Qfus. Fusion performance metrics, including Pfus and Qfus, exhibit only minor changes when switching from the interpretive impurity model to the predictive SANCO model, and/or from the continuous ad-hoc pellet model to the discrete HPI2 pellet model. The highest Qfus value predicted in the ITER high-density L-mode simulations is ~ 4, with indications that further improvement may be limited by increased transport associated with electro-magnetic turbulence at elevated plasma beta. This integrated modelling prediction demonstrates the potential of improved Qfus L-mode operation in ITER and future fusion devices, while exploring its boundary.
Lorenzo Votta, F J Artola, Eric Nardon, Oskar Vallhagen, Mathias Hoppe
We assess runaway-electron (RE) generation in ITER disruptions mitigated by shattered pellet injection (SPI) using improved physics modelling in the 1D disruption simulation framework DREAM. To this end, we extend DREAM with four ITER-relevant physics models: (i) a reduced model for RE scrape-off associated with the vertical plasma motion, (ii) a semi-analytical plasmoid-drift model for material deposition, (iii) an adaptive hyper-resistive transport model to suppress unphysical thin-current channels during the current quench (CQ), and (iv) an updated Compton RE generation seed calculated for the new ITER tungsten first-wall design. We simulate full-current 15 MA L-mode (H26, non-nuclear) and H-mode (DTHmode24, nuclear) scenarios, and an intermediate-current 7.5 MA H-mode non-nuclear case, from realistic ITER inputs. Within the adopted reduced-model framework, complete avoidance of a multi-MA RE beam is found to require a long pre-thermal quench (TQ) duration to thermalize the hot-tail electrons, high deuterium assimilation with limited neon, and a representative seed current comparable to a single RE in ITER. As previously found with lower fidelity setups [Vallhagen et al, Nucl. Fusion 64 (2024)], these conditions are met by staggered or low-Ne injections in H26, but are typically violated in DT H-mode when nuclear seeds are present. In addition to analyzing the effect of the new models, we investigate the role of the current spike associated with the TQ and the importance of radial transport of runaways in the CQ. After incorporating these additional physical effects into a comprehensive disruption model and analyzing their impact, we present a representative ITER DT H-mode SPI scenario which , within the adopted reduced-model framework, yields a substantially mitigated representative RE current. This case illustrates a possible route toward reduced RE-current levels in ITER DT operation, although its quantitative tolerability remains dependent on scenario-specific impact and termination assumptions.
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