Recent Publications

Error field penetration threshold from two-fluids compressible linear drift magneto-hydrodynamics

Paolo Zanca

The linear plasma response to a resonant error field (EF) is formalized by the so-called 'delta-prime' parameter Δ', representing the amplitude and phase of the current sheet induced at the resonant surface mostly by the electron fluid rotation. In combination with momentum transport, Δ' determines the critical EF amplitude above which a wall-locked magnetic island is formed (penetration threshold). In the context of two-fluids drift linear magneto-hydrodynamic (MHD) a semi-analytical computation of Δ', carried out with a low-beta ordering assumption, and the estimated penetration threshold for the ohmic tokamak, have been recently published [P. Zanca 2025 Plasma Phys. Control. Fusion 67 105033]. Here we present a necessary completion of that work brought by the inclusion of compressible flow, density perturbation, parallel energy transport and electron viscosity. In particular, electron viscosity has been discovered to be an important term in Ohm's law in a previous study which adopted the high-poloidal beta ordering [J. C. Waybright, J.-K. Park, Phys. Plasmas 31, 022502 (2024)]. Here, we confirm the relevance of electron viscosity within the scope of a different model. On the contrary, compressible flow, density perturbation and parallel energy transport do not modify significantly the EF penetration threshold, at least within the present linear theory. After reconsidering the ohmic tokamak with the refined model, we present a new analysis of the additionally heated tokamak, showing a strong positive dependence of the EF threshold on the electron fluid perpendicular rotation frequency ωe. All the scaling laws of the EF penetration threshold here derived present a strong density dependence, in general agreement with the experiments. Moreover, the model prediction is compatible with EF penetration experiments performed in JET NBI heated discharges. Finally, we provide a sensitivity study of the model prediction from the electron viscosity strength.

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.

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.

Investigation of the magnetic flux pumping effect in MAST Upgrade

Sam Blackmore, Christopher J Ham, Daniele Brunetti, Clive A Michael, Bhavin S Patel, Koki Imada, Lucy Kogan, cameron Olde, Fulvio Auriemma, Chiara Piron, et al.

Stationary safety factor q profiles with qmin ≈ 1 and without sawtooth activity have been measured for the first time in MAST Upgrade (MAST-U) plasmas, indicative of the magnetic flux pumping mechanism. This anomalous current re-distribution phenomenon is observed in MAST-U plasmas with 2/1 tearing MHD instabilities. Simulations using the ideal MHD code MISHKA indicate that q profiles, measured using the motional Stark effect (MSE) diagnostic, with zero magnetic shear are also ideally unstable to n = 1, m = 1 MHD modes. Database analysis of MAST-U pulses demonstrate overlap in βN ≈ 2 in pulses which exhibit magnetic flux pumping and sawtooth activity. The operational space for robust access to the flux pumping regime on MAST-U is characterised by operating at q 95 > 6 and β p ≈ 0.6.

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.

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.

Non-dimensional confinement scaling in similar negative triangularity plasmas on the DIII-D and TCV tokamaks

Alessandro Marinoni, Colin Chrystal, Stefano Coda, Reinart Coosemans, Claudio Marini, Mario Podesta, Olivier Sauter, Matteo Agostini, Max E Austin, Emily A Belli, et al.

Similarity experiments were performed on the DIII-D and TCV tokamaks to explore the scaling of energy confinement in negative triangularity plasmas using non-dimensional variables. Near up-down symmetric plasmas with large top-bottom averaged negative triangularity were created in a lower single null configuration, with the shape of the separatrix being closely matched between the two devices. The normalized energy confinement is found to weakly improve at increasing collisionality and, between the two devices, shows a machine size scaling behavior between Bohm and gyro-Bohm. Engineering scaling on a large DIII-D dataset is in agreement with the non-dimensional experiment.

Diagnostics for large tokamaks: from JET to JT-60SA 1

Carlo Sozzi, A Jokinen, G Phillips, K Tanaka, Juan Ayllon-Guerola, Andrea Belpane, Attila Buzás, Santiago Cabrera, mario - cavinato, Daniel Carralero, et al.

The main scientific purpose of JT-60SA is complementing ITER in the preparation of the operation of a DEMOnstration fusion reactor, in particular investigating the conditions for a controllable high beta steady-state regime able to optimize the fusion gain. In order to accomplish this task, a sequence of operation and machine enhancement periods in the next few years are planned to reach the target performance of the machine before a transition to a full tungsten wall. EUROfusion and Fusion for Energy are jointly contributing to the enhancement plan of JT-60SA, in particular, for what concerns the present contribution, to provide JT-60SA with state-of-art diagnostics in support of its scientific and technical objectives. This paper reports the status of the projects being implemented in view of the next scientific campaigns or under consideration through the various stages from feasibility to detailed design.

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.

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.

First step toward multi machine ELM energy scalings and extrapolations to SPARC and ITER

R. Perillo, A. Redl, T. Eich, C. J. Lasnier, A. Nelson, R. Rizkallah, D. Silvagni, A. Stagni, J. A. Boedo, A. McLean, et al.

It is shown that the ELM energy loss normalized by the plasma stored energy (ΔEELM/Wplasma) for high-density small/QCE ELM regimes scales inversely with the separatrix turbulence parameter a_t. In contrast, the neoclassical electron collisionality at the pedestal top, nu*e,neo, expected to regulate ΔEELM/Wplasma according to the Loarte scaling (Plasma Phys. Control. Fusion 2003 45 1549), does not adequately capture ΔEELM/Wplasma data for peeling-ballooning-limited type-I ELMs and ballooning-limited small/QCE ELMs, limiting its applicability for extrapolation to one scenario window. A multi-machine database including seven tokamaks and with ΔEELM/Wplasma ranging from 0.5% to 14%, has been analyzed. A regression analysis on only type-I ELMs yields ((ΔE_ELM)/W_plasma )_(Type-I ) [%]=6.8*T_(e,ped)^0.03 n_(e,ped)^(-0.4) \k{appa}^(-0.4) R_major^0.4, corresponding to ΔEELM/Wplasma =4.5% for nominal SPARC pedestal parameters and 12% for the ITER D-T Q=10 scenario. For the small/QCE ELM class, however, as a_t increases, the pedestal moves toward a ballooning-limited boundary, the toroidal mode number increases, the ELM frequency rises following the scaling f_ELM=46e^((2.25*a_t)), and ΔEELM/Wplasma decreases via the relation (ΔE_ELM)/W_plasma [%]=1.6e^(-(α_t/2)). For SPARC QCE-relevant a_t=0.86 and ITER high-fueling scenario a_t = 0.64, the scaling favorably predicts ΔEELM/Wplasma of 1.0% and 1.2%, respectively, with values below 1% if the small/QCE ELM regime is pushed beyond a_t >1. The small/QCE ELM-fitted results represent an initial step toward future analysis on broader datasets, which will be necessary to improve the accuracy of projections for future reactor-relevant scenarios.

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.

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.

The status and design challenges of the heating and current drive systems for DTT

Gustavo Granucci, Silvio Ceccuzzi, Afra Romano, Andrea Murari, Gian Luca Ravera, Piero Agostinetti, Sofia Bertolami, Falk Braunmüller, Alessandro Bruschi, Daniele Busi, et al.

This paper reports the main design keys and the challenging issues of the Heating and Current Drive System (HCD) of the Divert Tokamak Test (DTT) facility that is under construction at ENEA site in Frascati with the aim to perform studies on the power exhaust in a flexible and easily modifiable environment. The selected HCD systems for DTT are those with the most consolidated technology and expected to be relevant for the future reactor. The status of each system is reported, both in terms of design and procurement, which are well advanced for the system required in the first phase of the DTT exploitation: electron and ion cyclotron resonant heating. The third system is neutral beam injector, based on negative ion acceleration, which will be installed in DTT in a second phase, after the first five years of operation. The full heating power will be reached with the doubling of the radiofrequency power to reach the 45 MW at plasma needed to test the divertor with a power density at reactor level.

Beam optics analysis by visible cameras applied to the multi-beamlet configuration in the ITER NBI prototype source

Giulia Emma, Margherita Ugoletti, Matteo Agostini, Riccardo Agnello, Riccardo Casagrande, Isabella Mario, Roberto Pasqualotto, Antonio Pimazzoni, Carlo Poggi, Basile Pouradier-Duteil, et al.

The production of high energy neutral hydrogen/deuterium beams (0.87 MeV/1 MeV) with more than 90% uniformity and extremely low divergence (< 7 mrad) is highly challenging for the realization of the ITER Neutral Beam Injection system. The ITER prototype negative ion source, SPIDER, is in operation at the Neutral Beam Test Facility in Padua and it is equipped with a full set of diagnostics to study and investigate the beam properties. Among them, a system of 15 2D visible cameras, surrounding the vacuum vessel, is employed to measure the beam uniformity and divergence by detecting the light emitted from the beam-background gas interaction. In this paper, the beam divergence measured by these cameras is evaluated for the first time in the multi-beamlet configuration explored during the most recent SPIDER operation. The analysis is performed using the top-view and the side-view cameras, studying both the horizontal and vertical beam profiles through a 1D Gaussian fitting procedure, to estimate the width of the beamlet column or row aligned with the camera's lines of sight. The divergence is then retrieved from the linear fit of the width measured at different positions along the beam propagation direction. The method is here presented and applied to the experimental data acquired with an open beam segment. The dependence of the divergence on the main source and accelerator parameters, namely the RF power, the source filling pressure, and the voltage of the accelerator grids, is investigated. The results are then compared with measurements of the divergence given by the other beam diagnostics available in SPIDER, showing good agreement and validating the use of the visible camera system for the beam optics characterisation in the multi-beamlet operation.

Machine learning aided neutron yield for dud detection based on JET and TFTR Deuterium-Tritium plasmas

Lidia Piron, Alessandro Pau, Nicolò Ferron, Eric Fredrickson, Olivier Sauter, Matteo Baruzzo, Clive D Challis, Remi Dumont, Dirk Van Eester, Michael Fitzgerald, et al.

as it indicates fusion performance. To optimize Tritium consumption and limit neutron activation, a support function included in the plasma control system called a dud detector will trigger an alarm if the plasma fails to achieve expected fusion performance. This function has been developed and routinely employed at JET during DT campaigns. This study presents machine-learning methods based on a surrogate model of the neutron rate, which can be used as an advanced dud detector. In preparation for DT operations in BEST, HL-3, ITER, and SPARC, we investigate the portability and inherent limitations of these ML methods by analysing similar DT experiments conducted at TFTR.

Integrated core-SOL modelling of JET baseline plasmas in D/D-T/T: impact of ELMs and isotopic composition on Tungsten sputtering

Rachele Cicioni, Luca Garzotti, Lidia Piron, Vito Konrad Zotta, Vassili Parail, Agata Chomiczewska, Domenico Frigione, Alexander Huber, Florian Koechl, Henri Aaron Kumpulainen, et al.

The JET DTE2 experimental campaign revealed an isotopic dependence in the baseline scenario, where the target stationary conditions achieved in deuterium plasmas could not be maintained in tritium and deuterium-tritium mixtures. This loss of stationarity was associated with a gradual increase in plasma density, reduced ELM activity, and rising core radiation. This work investigates whether isotope-dependent tungsten sputtering contributes to the loss of stationarity observed in JET DTE2 baseline plasmas. The analysis focuses on the tungsten sputtering source and on tungsten transport during inter- and intra-ELM phases, as a function of the isotopic species of the main plasma. The study is carried out using the integrated core-edge-SOL framework COCONUT/JINTRAC, which was used to evolve the plasma self-consistently across the entire domain. The simulations show a clear isotopic dependence of the tungsten sputtering source, which increases with the mass of the main plasma ion and reaches the highest levels in tritium plasmas. Despite the enhanced tungsten source, most of the sputtered tungsten remains confined to the SOL, and no substantial net increase of the tungsten content is observed in the plasma core over the ELM time window. Pedestal onditions affect tungsten transport across the separatrix and the transient tungsten response during the ELM cycle. The results also indicate that ELM dynamics, in particular the ELM frequency, can modify the transient tungsten content in the core. Overall, the simulations suggest that isotope composition and ELM activity affect the tungsten sputtering source and its short-time redistribution, but that ELM-induced sputtering alone does not explain the loss of stationarity observed experimentally.

Two-fluid boundary turbulence simulations in reversed field pinch plasmas

M. Giacomin, B. Momo, I. Predebon, N. Vianello, M. Zuin

Turbulent transport in magnetic confinement fusion devices governs the overall plasma confinement properties and regulates the plasma-material interaction at the first wall. In the plasma boundary, turbulence is typically investigated through three-dimensional two-fluid flux-driven turbulence simulations. In this work, the GBS boundary turbulence code is extended to enable turbulence simulations in reversed field pinch configurations, encompassing the reversal surface and an arbitrary level of magnetic chaos. The differential operators implemented in the code are modified to avoid the approximations of large-aspect ratio and weak poloidal magnetic field. Three-dimensional Poisson and Ampere solvers are implemented to allow for turbulence simulations in conditions of partially or fully disrupted magnetic flux surfaces. This modified version of the GBS code is then applied to simulate turbulence in the boundary of RFX-mod reversed field pinch plasmas. Turbulent eddies across the reversal surface show properties similar to those typically found in tokamak boundary turbulence simulations. Despite the good agreement found with experimental measurements, these simulations reveal a significant limitation of the fluid-based turbulence modeling of the edge region in reversed field pinch plasmas, which arises from the intrinsically short parallel connection length. This conclusion is also supported by a linear gyrokinetic analysis that identifies trapped electron modes as the dominant microinstability in this region.

Selling a dream? Fusion experts’ perspectives on the challenges of fusion deployment: Evidence from conference workshops

Piotr Stankiewicz, Gaston Meskens, Christian Oltra, Vanja Čok, Alessio Giacometti, Chiara Bustreo

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.

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.

Effect of neutral and ion temperatures in the 2D fluid model for the RF heated ion source SPIDER

Roman Zagorski, Iacopo Regoli, Daniel Lopez-Bruna, Isabella Mario, Antonio Pimazzoni, E Sartori, Alastair Shepherd, Gianluigi Serianni

This paper presents the basic physical and numerical principles of a fluid model implemented into numerical code FSFS2D (Fluid Solver For SPIDER in 2D) used for simulating the SPIDER negative ion source. It gives self-consistent 2D description of the source, including the neutral gas flow, plasma chemistry, RF coupling in the source driver and plasma transport through the magnetic filter. This paper focuses on the recent developments of the plasma model and the first simulations showing the role of the neutrals and ions temperature equations on the source performance are presented. The upgraded model is validated against SPIDER Langmuir-probe measurements and shows reasonable agreement for the axial plasma profiles. The new heavy-species temperature equations reveal strong differences between atoms, molecules, and ions, with D- temperatures increasing in low-density regions near the plasma grid. The simulations provide insight into the energy exchange mechanisms between plasma species and show that only about one third of the RF power transferred to electrons is consumed by plasma-chemical reactions. Atomic deuterium and positive ions are heated mainly through chemical processes, whereas molecules and negative ions are heated predominantly by elastic collisions. Parametric studies demonstrate the sensitivity of heavy-species temperatures to RF power, gas pressure, and magnetic-filter strength.

The Divertor Tokamak Test Facility Research Plan

Gerardo Giruzzi, P Martin, Clemente Angioni, Sebastijan Brezinsek, Flavio Crisanti, Giacomo Dose, Matteo Valerio Falessi, Paolo Innocente, Paola Mantica, Eric Nardon, et al.

The Divertor Tokamak Test facility (DTT) is a device presently under construction at the ENEA site in Frascati (Italy) in the framework of a joint public/private partnership. It has been designed as a superconducting tokamak with breakeven class performance, with the main objective of developing credible solutions for heat and particle exhaust, a key challenge in view of future fusion reactors. This needs to be addressed in a core-edge integrated approach, to assess the compatibility of exhaust solutions with reactor relevant core performance. In this paper, an overview is provided of the DTT research plan, recently developed by an international team. It covers the DTT programmatic objectives, research strategy and expected scientific contributions connected with the device characteristics, not only in the key area of heat exhaust and edge plasma physics, but also on other subjects of high fusion relevance, such as MHD stability in high performance scenarios, transport and turbulence, energetic particle physics, validation of advanced theoretical developments, as well as tests of technological solutions for reactor relevant components.

Conceptual design of the Halo Machine

V.V. Yanovskiy, M. Alessio, G. Becatti, E. Cattaruzza, R. Cavazzana, E. Martines, B. Momo, R. Paccagnella, D. Tskhakaya, F. Villone, et al.

In 1996, when the danger of sideways forces was first recognised, the operational window of the JET tokamak was significantly restricted. Limits were placed on the maximum magnetic field and plasma current to prevent excessive stresses on the vacuum vessel. Despite three decades of research, predictions for the sideways force on the ITER wall remain highly uncertain, ranging from 2 to 60 MN with the upper estimate exceeding the design value of 48 MN, which would result in operational restrictions (operation with I p < 15 MA potentially affecting ITER’s ultimate fusion performance). To address this issue, the ITPA community launched the dedicated joint experiment MDC-25 in 2018. While substantial progress has been made, recent findings highlight the need for complementary studies on a specialised linear device to investigate plasma–wall currents, key contributors to sideways forces. Here, we propose the first linear device specifically designed to investigate plasma–wall interactions in the presence of kink modes, with the aim of resolving uncertainties related to plasma–wall currents and associated sideways forces. It is expected that the findings obtained with the new device will guide the design of vacuum vessels and plasma facing components for future tokamaks.

Non-inductive high-performance discharges on TCV on the path to steady state

S. Coda, C. Piron, I. Voitsekhovitch, M. Agostini, F. Auriemma, L. Cordaro, A. Mele, M. Podestà, S. Garavaglia, A. Jardin, et al.

An extended experimental effort is underway on the TCV tokamak to develop scenarios compatible with long-pulse operation, featuring mostly non-inductively driven current—preferably with a large fraction of bootstrap current. A closely related goal is to achieve good plasma performance, typically measured through the normalized beta β N . This work is part of a broader endeavor involving several European tokamaks, under the auspices of the Tokamak Exploitation Work Package (WPTE) of EUROfusion, and aimed in part at preparing advanced scenarios for the new JT-60SA tokamak, which is the largest such device ever operated and has these scenarios at the core of its mission. This paper reports on the encouraging progress achieved in the last campaign, featuring an extensive set of discharges sustained over multiple current relaxation times ( T pulse ≫ τ R ) with zero flux contribution from the central solenoid (CS), and approaching stationary conditions with β N ∼ 2 and ion temperature ( T i ) rising towards the same order of magnitude as the electron temperature ( T e ). Auxiliary power currently appears to be a key limitation. With increased heating being added in 2027, there is now a realistic prospect of TCV reaching a fully stationary, high- β N , fully non-inductive neutral-beam injection-heated scenario. In the process of exploring the boundaries of this scenario, the hot-electron ( T e ≫ T i ) internal-transport-barrier (ITB) regime was also revisited and temperatures of 12 keV were recorded (a record for TCV non-inductive discharges). Additionally, a fully CS-free current ramp-up, starting only 30 ms after breakdown and displaying robustly negative central magnetic shear, has also been demonstrated, with a smooth continuation into the flat-top, non-inductive, advanced-scenario phase. This scenario, which is accompanied by an electron ITB of varying strength, is also a promising step towards a possible spherical-tokamak power plant. Finally, central-solenoid-free operation in the X -point target divertor configuration has been achieved, with a view to attempting detachment in a closed divertor geometry.