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F. Sabourin, L. Depradeux, J. Delcroix, A. Gravouil, L. Giraudon, Y. Le Nôtre, R. Cluzel, P. Clerc, M. Privalov, A. Lee, et al.
I. Ricapito, M.Y. Ahn, A. Aiello, B. Bellin, S. Cho, S. D’Amico, M. Fabbri, F. Fantini, E. Garcia, L. Giancarli, et al.
A. Perez Pecharroman, A.T. Vu, P.C. de Vries, A. Bustos, I.S. Carvalho, W. Helou, F. Kazarian, E. Lerche, M. Schneider
Tarundeep Kaur Lamba, L. Moser, Martin Kocan, Martin Vincent, Yan Rong, Ding Rui, Zhang Wei, Tom Wauter, Richard A. Pitts, Walsh Michael, et al.
Emanuele Sartori, Riccardo Casagrande, Isabella Mario, Antonio Pimazzoni, Basile Pouradier-Duteil, Alastair Shepherd, Riccardo Agnello, Kamran Ahmad, Piero Agostinetti, Matteo Agostini, et al.
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.
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.
A. Zoppoli, E. Kolokotronis, N. A. Pintos Cabrera, S. Esque, H. Saarinen, E. Ruiz Morales, D. Hamilton, G. Di Gironimo
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.
Francesco Pastore, Olivier Sauter, Federico Felici, Daniela Kropáčková, Alessandro Balestri, Cristian Galperti, Ondrej Kudlacek, Kenneth Lee, Adriano Mele, Alessandro Pau, et al.
Real-time estimation and control of the electron plasma density profile is fundamental for monitoring and regulating particle confinement, fusion power, heating efficiency, exhaust performance, impurity concentration, and proximity to the density limit [1] in present and future tokamaks. This work presents the experimental applications of a multi-rate electron density observer [2] based on RAPDENS [3], integrated into the TCV Plasma Control System, to control the density profile in a range of plasma scenarios relevant to future fusion power plants. Three main experimental contributions are reported. First, the observer enables control of the line-averaged electron density within the last-closed flux surface (NEL LCFS ) in ohmic, nitrogen-seeded plasmas with alternative divertor geometries, decoupling upstream density from Scrape-Off Layer pickup in the interferometer signal – a limitation of the traditional control scheme. Second, local control of the central electron density below the ECH cutoff is demonstrated in L-mode plasmas with mixed ECH and NBI heating, where auxiliary-heating-induced profile peaking is treated as a disturbance to the control task. Real-time estimation and adjustment of the electron pinch velocity-to-diffusivity ratio ν/D improves the spatial accuracy of the profile reconstruction, especially in the core region. The underlying turbulent transport in this scenario is characterised via linear and non-linear gyrokinetic simulations with GENE, confirming quantitatively a TEM-dominated pumpout regime during ECH, which affects the amount of gas flux needed to sustain the target density reference. Last, simultaneous control of the edge-normalised density fraction and toroidal beta is demonstrated in high-performance H-mode plasmas (β N ≈ 2.15, f GW ≈ 0.80), yielding reproducible scenarios with a device-agnostic density metric and robust handling of interferometer fringe jumps.
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.
Mauricio Rodriguez Ramos, Pablo Alvarez-Frau, Javier Gonzalez Martin, Micaela Videla-Trevin, Alex Reyner-Viñolas, Martin Kocan, Javier Garcia-Lopez
Accurate characterization of suprathermal particles, known as fast ions, is essential for the secure and efficient operation of nuclear fusion plasmas, as energetic particle losses can degrade plasma performance and damage reactor components. Scintillator-based detectors are widely employed to monitor fast ions; however, existing studies often assume isotropic light emission from the scintillator, neglecting potential angular dependencies that can compromise the determination of ion fluxes using this type of diagnostic. In this work, we investigate the angular emission properties of two commercial scintillators, TG-Green and β-SiAlON, under irradiation with 3.5 MeV He ++ and 1 MeV D⁺ beams, representative of conditions in future fusion devices such as ITER. A novel experimental setup, combining precise optical alignment, angular scanning, and calibration of the acquisition system, was developed to measure the relative efficiency as a function of the observation angle. Prior to characterization, emission stability tests demonstrated negligible radiation-induced degradation under the applied fluences, and transmission losses due to optical fiber bending, one of the mechanisms of light transmission loss, were found to be below 1.5%. The results reveal a pronounced angular anisotropy in scintillation emission for both materials, with relative efficiency decreasing as the detection angle increases, well described by an empirical cosine-based model. Additionally, the normalized response shows minimal dependence on ion species or energy. The observed emission anisotropy has been incorporated into the modelling of the optical system for the ITER Fast Ion Loss Detector. This results in an approximately 8% increase in the predicted signal compared to the isotropic emission assumption, enabling more accurate measurements of lost fast-ion fluxes in fusion plasmas.
Xuemin Wu, Youwen Sun, Qun Ma, Shuai Gu, Manni Jia, Yueqiang Liu, Yifeng Wang, Cheng Ye, Pengcheng Xie, Alberto Loarte, et al.
Existence of operational window in both edge safety factor and line averaged plasma density for suppression of ELMs using n=4 Resonant Magnetic Perturbations in low input torque plasmas has been observed in EAST experiment, in which q95 and plasma normalized beta (βN) close to that required in ITER high-Q operation. Here, n is toroidal mode number of the magnetic perturbation. In contrast to previous reports from other tokamaks, there is not only an upper density limit but also a lower one for accessing ELM suppression. Modelling results using the MARS-F code show that the RMP with linear plasma response has a peak at an intermediate density and decays as the density increases or decreases, which results in a minimal RMP field penetration threshold at the intermediate density. In this experiment, the observed lower density limit operationally manifests a sensitivity of the q-profile: different densities alter the edge current profile, which change the alignment of the eigenmode structure with the RMP coil configuration, causing a reduction of the resonant field in both low- and high-density cases, and hence making field penetration more difficult. The modelled window of the strongest resonant plasma response in terms of [⟨ne⟩, q95] agrees well with the observed ELM suppression in EAST. Peeling-ballooning modes stability analysis using the ELITE code shows that plasmas gradually approach peeling instability boundary caused by increase of edge bootstrap current as the plasma density decreases, which is consistent with the observation that ELMs come back again in lower density plasmas for fixed q95. These results indicate that linear modelling with full toroidal geometry can well predict the optimized RMP configuration for ELM suppression and reveal the important roles of pedestal plasma current, which need to be carefully considered in the application of high n RMPs for ELM suppression in future ITER.
Matti Coleman, James Cook, Fabrizio Franza, Ivan Maione, Simon Mcintosh, Hudson Baker, Alexander Ian Blair, Shail Desai, Oliver Funk, Georgina Graham, et al.
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.
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.
Jose Ramon Martin Solis, Jose Angel Mier, F J Artola, Alberto Loarte
A simplified model analysis is used for the investigation of the survival of runaway electrons following the break-up of the magnetic field lines and the reformation of the flux surfaces during the current-quench phase of tokamak disruptions in vertically unstable plasmas. It is found that even low levels of the runaway current (~ a few tens of kAs) when the plasma touches the wall might lead to runaway damage due to the strong runaway avalanche during scraping-off. Low enough runaway deconfinement times (τd < 1 ms) in a sufficiently long stochastic phase (τ/τd > 5) would be required to avoid a large energy transfer to the runaway electrons and damage of the plasma facing components for low temperatures (a few eVs) of the residual ohmic plasma. The role played by the primary runaway generation mechanisms in regenerating the runaway beam after the magnetic surfaces have healed is also discussed.
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.
Matthew Pharr, Evan Bursch, Nikolas Logan, Priyansh Lunia, Jong-Kyu Park, Carlos Paz-Soldan
The Fourier spectra of resonant quantities in tokamaks depend on the choice of magnetic coordinates, and an area weighting of the Fourier integrand preserves the resonant coefficients on rational surfaces. That result constrains only the resonant interior; the coordinate dependence of the external Fourier spectrum, which determines the coupling to Resonant Magnetic Perturbation (RMP) coils and error-field penetration, was left untreated. This paper shows that pairing a square-root-area weighted vacuum field perturbation with a full-area-weighted resonant field yields a coupling matrix C whose singular values are invariant under coordinate transformations and whose right singular vectors reconstruct to a consistent real-space field pattern across coordinate systems, completing the coordinate-invariance picture for the plasma-3D-field coupling paradigm. GPEC calculations confirm the analytic result and show that improperly weighted coupling matrices can produce dominant modes whose overlap with the vacuum field perturbation differs by a factor of $2--3$ between coordinate systems for strongly shaped, low aspect ratio equilibria, with the discrepancy growing with inverse aspect ratio. The same coordinate dependence afflicts alternative formulations such as the three-mode metric or zeroing the $q=2$ resonant field without proper weighting. The result applies to any tool computing Fourier spectra of resonant or external quantities on flux surfaces.
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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