Hollow toroidal rotation profiles in strongly electron heated H-mode plasmas in the ASDEX Upgrade tokamak

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After a 26-month vent ASDEX Upgrade (AUG) went back in operation with a newly designed upper W-divertor suitable for alternative divertor configurations (featuring in-vessel coils and cryo-pump). Parameter scans and an extensive set of measurements were obtained and their interpretation is ongoing. Prompted by the ITER wall change, dedicated experiments on non-boronized plasma startup were contrasted to that employing asymmetric and more symmetric boronizations. The asymmetric boronization proved to be as b

In JET-ILW, beam-target reactions contribute to a large fraction of the fusion power generated in Deuterium – Tritium (D-T) plasmas, with core ion temperatures of 10-12keV and large neutral-beam injection (NBI) power. Previous modelling done in preparation for the recent D-T campaigns in JET have shown that injecting D beam ions with energies of ~120keV in T-rich plasmas produces larger 14MeV fusion yield than in 50:50 D:T plasmas, but such scenario had never been tested in past D-T experiments. In addition

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

Plasma state reconstruction methods combining measurements and physics modeling improve the estimation of physics quantities in real-time and in post-discharge analysis. We present a workflow to reconstruct the dynamic evolution of a set of internal tokamak plasma profiles with consistent equilibria, as applied in this paper for experimental data from TCV L and H-mode discharges. Plasma profile estimates for electron temperature Te, electron density ne and parallel current density jpar are obtained by data

This Roadmap article addresses the critical and multifaceted challenge of plasma-facing component (PFC) damage caused by runaway electrons (REs) in tokamaks, a phenomenon that poses a significant threat to the viability and longevity of future fusion reactors such as ITER and DEMO. The dramatically increased RE production expected in future high-current tokamaks makes it very difficult to avoid or mitigate REs in such devices when a plasma discharge terminates abnormally. Preventing damage from the intense

The recent decision by ITER to move from a Be to a W first wall has significant implications for material transport and deposition, and can be expected to lead to increased W layer growth rate in deposition regions, which may flake leading to plasma disruption. In this work a novel sputtering set-up was implemented in the Magnum-PSI linear plasma device to create W layers under ITER relevant conditions. A W plate was inserted into the plasma edge and biased to cause sputtering of W in an Ar plasma. This spu

The TCV tokamak contributes to the development of nuclear fusion energy with proof-of-principle experiments and by validating models that are used to predict reactor performance. In the next upgrade of the TCV divertor it is planned to test a tightly baffled, long-legged divertor (TBLLD), a novel concept designed to enhance power exhaust handling with minimal modification to the magnetic configuration. The project is guided by simulations using the SOLPS-ITER code that indicate that a TBLLD can improve TCV’

Advanced tokamak regimes, featuring extended regions of low magnetic shear, are promising candidates for future fusion reactors but are also more prone to specific kinds of MHD instabilities. The proximity to a rational surface in a very low shear region weakens field line bending stabilisation and amplifies the effects of toroidal coupling between modes, leading to the emergence of long-wavelength resistive infernal modes. These modes can grow collectively as a discrete spectrum, leading to a cascade of di

Energetic particles (EP) generated by the neutral beam injectors (NBI) in Large Helical Device (LHD) destabilize Alfvén Eigenmodes (AE) and energetic particle modes (EPM), leading to a reduction of the device performance. In particular, AE/EPM induce EP losses before thermalization that causes a lower plasma heating efficiency and damages to plasma facing components. Pressure gradient driven modes (PGDM) as interchange and ballooning modes also hamper the capability of LHD to confine the thermal plasma, lim

Proper comparison of simulated turbulent fluctuations with experimental observations is neces- sary for code validation, and for establishing a link between experimental fluctuation measure- ments and the underlying physics. In this contribution we report on the progress of modelling localised measurements of turbulent electron density fluctuations, obtained with the tangential phase contrast imaging (TPCI) diagnostic installed on the TCV tokamak. The modelling is done in two steps. First, nonlinear flux-tu

Simulating the most external plasma region of a tokamak, the scrape-off layer (SOL), is of crucial importance on the path towards a fusion reactor as heat load on the vessel wall, impurity generation, and overall plasma confinement all depend on the plasma dynamics in this region. The GBS code has been developed to solve the drift-reduced Braginskii equations which describe turbulent behaviour of the plasma in the SOL. Until recently, only limited scenarios were considered in GBS and it was not possible to

Optimization of the plasma discharge can be defined as determination of an optimal time evolution of the plasma parameters to lead a plasma to a desired state keeping it within the specific limits: physical ones (like the Greenwald density limit, low normalized beta and internal inductance values) and technical ones (like the vertical stability limit). The parameters, time-trajectories of which have to be optimized, are the ones significantly changing the plasma state, and, depending on the optimization goa

Plasma turbulence in the tokamak scrape-off layer (SOL) region, where magnetic field lines are open and intersect the reactor inner walls, determines the heat loads on the limiter or divertor targets. This is one of the most crucial issues on the way towards a fusion reactor. Since SOL plasma is colder compared to the tokamak core, it is reasonable to use a fluid approximation to describe its dynamics. In particular the drift-reduced Braginskii equations are chosen to study the SOL plasma turbulence. To fur

Simulating the most external plasma region of a tokamak, the scrape-off layer (SOL), is of crucial importance in the way toward a fusion reactor as heat load on the vessel wall, impurity generation, overall plasma confinement, all depend on the plasma dynamics in this region. In the last few years a numerical code, GBS, has been developed for solving the drift-reduced Braginskii equations (DRBE), which describe turbulence in the tokamak SOL. In the present work, we re-formulate the DRBE in a general form th

Earthing of electrical installations is mainly governed by safety rules. Electromagnetic compatibility also deals with earthing, among other circuit characteristics. Tokamaks are large-scale electrical installations that are known to generate large and low frequency magnetic fields as well as large and high frequency electric fields. Four European Tokamak installations have been investigated, from the earthing point of view, to identify appropriate techniques to earth the electrical equipment and to provide

To ensure sufficient divertor target lifetime, the loss in plasma stored energy due to ELMs in ITER should be restricted to Delta W-ELM

Four of the 16 ITER upper port plugs will be devoted to electron cyclotron resonance heating (ECRH) in order to control the magneto-hydrodynamic (MHD) instabilities. In order to achieve the stabilisation of the neoclassical tearing modes (NTM) and sawtooth oscillation, a deposition of a very localized and peaked current density profile over a broad poloidal steering range is required. In the present optical configuration eight 2MW mm-wave beams enter each of the four upper launchers (UL) through waveguides

Magnetic reconnection in collisionless magnetohydrodynamic plasmas is studied under controlled conditions in toroidal X-line configurations formed in the center of the Versatile Toroidal Facility [J. Egedal , Rev. Sci. Instrum. 71, 3351 (2000)]. Profiles of key parameters are measured during driven reconnection pulses in the reproducible plasma discharges. The experimental results prove to be highly sensitive to the boundary conditions applied. In the configuration with "open" boundaries reconnection is fas

Recent results from the ASDEX Upgrade experimental campaigns 2001 and 2002 are presented. An improved understanding of energy and particle transport emerges in terms of a 'critical gradient' model for the temperature gradients. Coupling this to particle diffusion explains most of the observed behaviour of the density profiles, in particular, the finding that strong central heating reduces the tendency for density profile peaking. Internal transport barriers (ITBs) with electron and ion temperatures in exces

In current tokamaks and, in particular, in future larger devices such as ITER, the control of neo-classical tearing modes (NTM) is essential for achieving high performance in terms of the beta limit. A commonly used scheme for NTM stabilization consists in driving a helical current at the resonance surface of interest with electron-cyclotron-current-drive. Depending on the ratio between the magnetic island size and the RF beam width, complete stabilization of the NTM will only be achieved with deep RF power

Hollow toroidal rotation profiles in strongly electron heated H-mode plasmas in the ASDEX Upgrade tokamak

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Overview of the ASDEX Upgrade results

Heating D ions to optimal D-T fusion energies in JET-ILW

Towards Digital Twins of Fusion Systems

Model-based Estimation of Tokamak Plasma Profiles and Physics Parameters: Integration with Improved Equilibrium Reconstruction and Experimental Data

The Roadmap for runaway electron-induced plasma facing component damage in tokamaks

Growth, Characterization and Adhesion of Thick Deposited Tungsten Layers Under Transient Loading in Magnum-PSI

Implementation of a Tightly-Baffled, Long-Legged Divertor (TBLLD) in TCV

The Properties of Resistive Mhd Modes and Unstable Spectra in Advanced Tokamak Regimes

Analysis of energetic particle driven modes and interchange modes in LHD plasma using the gyro-fluid code FAR3d

Modelling of turbulent fluctuations measured in the TCV tokamak with gyrokinetic simulations and a synthetic phase contrast imaging diagnostic

Implementation of Diverted Equilibria in the GBS simulation code

Numerical optimization of ramp-down phases for TCV and AUG plasmas

The impact of the Boussinesq approximation on plasma turbulence in the scrape-off layer

A flexible numerical scheme for simulating plasma turbulence in the tokamak scrape-off layer

ITER-Earthing

The impact of large ELMs on JET

Design and manufacturing of the ITER ECRH Upper launcher mirrors

Experimental evidence of fast reconnection via trapped electron motion

Overview of ASDEX upgrade results

Power modulation capabilities of the 140 GHz/1 MW gyrotron for the stellarator Wendelstein 7-X