The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

Massive material injections in the JET tokamak have been observed to substantially affect resistive bolometer measurements, resulting in a spurious radiated power signal proportional to the quantity injected and reaching up to 8 MW. These bolometers are calibrated and designed to operate in near vacuum but certain scenarios requiring large gas injections can push the neutral pressure past nominal values. This study demonstrates that the bolometry measurement can be affected at neutral pressures above 0.1 Pa following injections with standard gas valves, shattered pellet injections, and particularly massive gas injections. The power measurement of resistive bolometers is based on the temperature difference between a measurement sensor exposed to radiation and a shielded reference sensor. We employ a thermal conductivity model to demonstrate that the conduction through the gas and the distinct geometries between the sensors can affect their cooling efficiency. This additional cooling pathway, coupled with the Joule heating from the applied voltage causes the equilibrium temperatures of the sensors to diverge. Being the very basis of the measure, this temperature difference induces a signal that is erroneously interpreted as radiated power. Experiments show large discrepancies in the response to neutral pressure among bolometer channels, attributed to variations in channel physical parameters. Nonetheless, the modeled total radiated power reproduces the experimental measurements within an order of magnitude, affirming the sensitivity of resistive bolometers to neutral pressure and gas species.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

We present global linear and nonlinear simulations of ion temperature gradient instabilities based on a fluid formulation, with an adapted version of the JOREK code. These simulations are performed in realistic global tokamak equilibria based on the solution of the Grad–Shafranov equation. Benchmarking of linear growth rates was successfully completed with respect to previously published data. We find two distinct types of eigenstructures, depending on the magnetic shear. For high shear, when the coupling of poloidal harmonics is strong, ballooning-type eigenmodes are formed, which are up-down asymmetric with a finite ballooning angle, θ0. The poloidal harmonics which form the global eigenmode are found to demonstrate a radial shift, being centered well outside of their corresponding rational surface. Stronger diamagnetic effects increase both θ0 and proportionately shift the m harmonics to larger radii (by as much as two rational surfaces). In the low shear regime, the unstable eigenmodes become narrowly localized between neighboring pairs of rational surfaces, and exhibit no up-down asymmetry. Our simulations also show the generation of finite Reynolds stress due to nonlocal/global profile effects. This stress possesses both poloidally symmetric (n=m=0) and asymmetric (finite-m) components. Turbulent saturation in nonlinear simulations is demonstrated for both shear regimes.

Disruptions are a major threat to future tokamaks including ITER. They generate excessive electromagnetic forces, heat loads and multi-MeV runaway electrons. The runaway electron beam carries the risk of in-vessel component damage and even the structures beyond them. Thus, prevention of the runaway beam generation or the mitigation of the developed beam is of prime importance. In JET ITER-like wall, the runaway electron beams triggered by massive gas injection (MGI) coexists with a cold background plasma. Lines corresponding to the higher ionization states of argon are observed in VUV spectra outside of the runaway region suggesting a hot background plasma.Using the quantitative analysis of the VUV spectroscopy, the temperature profiles of the background plasmas are estimated using a synthetic line ratios method. The background plasmas at JET-ILW are found to be hotter than other tokamaks where mitigation of the runaway electron beam was unconditionally successful. The volume-averagedT(e)is found to increase linearly with the gas amount used to trigger the disruption and the electron density in the far scrape-off layer. It is independent of other background plasma properties. A 0D/1D power balance of the post-disruption physical systems is made using the characteristics of the background plasma. The collisional power loss of the runaway electron beam is the primary power source heating the background plasma.

The spontaneous pedestal formation above a power threshold at the edge of magnetically confined plasma is modelled for the first time in flux driven three-dimensional fluid simulations of electromagnetic turbulence with the code EMEDGE3D. The role of the collisional friction between trapped and passing particles is shown to be the key ingredient for shearing the radial electric field, hence stabilizing the turbulence, rather than the Maxwell and Reynolds stresses. The isotope effect, observed in many tokamaks worldwide, is recovered in EMEDGE3D simulations: the power threshold for pedestal formation is lower for Tritium than for Deuterium. The turbulence auto-correlation time is found to increase with the ion mass easing the radial electric shear stabilization, hence the pedestal formation.

In tokamak plasma, the edge region is characterized by larger gradients, higher resisitivity and a shorter distance to the critical beta compared to core plasma. By edge region, we mean the region where the pedestal forms when transiting from the L to H-mode: r/a = 0.85-1. This paper shows, using a three-dimensional electromagnetic fluid turbulent transport model, that the quasilinear hypothesis is valid in the edge region. The diamagnetic coupling is stabilizing while beta is destabilizing. The effect of beta on the characteristic times of turbulence is studied, as is its effect on the Reynolds stress and the Maxwell stress in these fluid simulations.

Magnetic triggering of edge localized modes (ELMs) in Ohmic H-mode plasmas was first reported in the TCV tokamak (Degeling et al 2003 Plasma Phys. Control. Fusion 45 1637). This method, showing reliable locking of the ELM frequency to an imposed axisymmetric vertical plasma oscillation, was also demonstrated in the ITER-relevant type-I ELM regime in ASDEX Upgrade (Lang et al 2004 Plasma Phys. Control. Fusion 46 L31) and JET (de la Luna et al 2015 Nucl. Fusion 56 026001). However, the mechanisms of the ELM triggering due to a vertical motion has not been studied extensively. The non-linear reduced MHD code JOREK-STARWALL has been extended for 3D free-boundary computations (Hölzl et al 2012 J. Phys.: Conf. Ser. 401 012010), which has allowed us to simulate for the first time realistic vertical oscillations together with ELM simulations in a single consistent scheme. Our simulations demonstrate that stable plasmas can be destabilized by the application of a vertical oscillation for ITER. During the vertical motion, a toroidal current is induced in the pedestal. The origin of this current is analysed in detail with the use of simulations and a simple analytical model, revealing that it arises from the compression of the plasma cross section due to its motion through an inhomogeneous magnetic field. Lower pedestal currents between ELMs require bigger vertical displacements to destabilize ELMs, which directly points towards the increased edge current as the ELM driving mechanism. Finally the ELM triggering shows a very weak dependence on the plasma velocity for ITER in agreement with experiments.

In order to contribute to the understanding of runaway electron generation mechanisms during tokamak disruptions, a test particle tracker is introduced in the JOREK 3D non-linear MHD code, able to compute both full and guiding center relativistic orbits. Tests of the module show good conservation of the invariants of motion and consistency between full orbit and guiding center solutions. A first application is presented where test electron confinement properties in a JET massive gas injection-triggered disruption simulation are investigated. It is found that electron populations initialised before the thermal quench (TQ) are typically not fully deconfined in spite of the global stochasticity of the magnetic field during the TQ. The fraction of "survivors" decreases from few tens down to a few tenths of percent as the electron energy varies from 1keV to 10MeV. The underlying mechanism for electron "survival" is the prompt reformation of closed magnetic surfaces at the plasma core and, to a smaller extent, the subsequent reappearance of a magnetic surface at the edge. It is also found that electrons are less deconfined at 10MeV than at 1MeV which appears consistent with a phase averaging effect due to orbit shifts at high energy.

Runaways are suprathermal electrons having sufficiently high energy to be continuously accelerated up to tens of MeV by a driving electric field (Connor and Hastie 1975 Nucl. Fusion 15 415). Highly energetic runaway electron (RE) beams capable of damaging the tokamak first wall can be observed after a plasma disruption (Reux et al 2015 Nucl. Fusion 55 129501). Therefore, it is of primary importance to fully understand their generation mechanisms in order to design mitigation systems able to guarantee safe tokamak operations. In a previous work, Sommariva et al (2018 Nucl. Fusion 58), a test particle tracker was introduced in the JOREK 3D non-linear MHD code and used for studying the electron confinement during a simulated JET-like disruption. It was found in Sommariva et al (2018 Nucl. Fusion 58) that relativistic electrons are not completely deconfined by the stochastic magnetic field taking place during the disruption thermal quench (TQ). This is due to the reformation of closed magnetic surfaces at the beginning of the current quench (CQ). This result was obtained neglecting the inductive electric field in order to avoid the unrealistic particle acceleration which otherwise would have happened due to the absence of collision effects. The present paper extends (Sommariva et al 2018 Nucl. Fusion 58) analysing test electron dynamics in the same simulated JET-like disruption using the complete electric field. For doing so, a simplified collision model is introduced in the particle tracker guiding center equations. We show that electrons at thermal energies can become RE during or promptly after the TQ due to a combination of three phenomena: a first REs acceleration during the TQ due to the presence of a complex MHD-induced electric field, particle reconfinement caused by the fast reformation of closed magnetic surfaces after the TQ and a secondary acceleration induced by the CQ electric field.

The fluid theory of a new type of electron temperature gradient instability is proposed. This mode is closely related to the short wavelength Alfvén mode in the regime k⊥2ρi2>1. Contrary to standard electron temperature gradient modes, which are mostly electrostatic, the considered mode is fundamentally electromagnetic and does not exist in the electrostatic limit. The mechanism of instability relies on gradients in both the electron temperature and magnetic field. It is suggested that this instability may be a destabilizing mechanism for collisionless microtearing modes, which are observed in a number of gyrokinetic simulations.

The degradation of plasma confinement in tokamaks caused by magnetic islands motivates to better understand their possible suppression using electron cyclotron current drive (ECCD) and to investigate the various strategies relevant for this purpose. In this work, we evaluate the efficiency of several control methods through nonlinear simulations of this process with the toroidal magneto-hydro-dynamic (MHD) code XTOR-2F (Lutjens and Luciani 2010 J. Comput. Phys. 229 8130-43), which has been extended to incorporate in Ohm's law a source term modeling the driven current resulting from the interaction of the EC waves with the plasma. A basic control system has been implemented in the code, allowing testing of advanced strategies that require feedback on island position or phase. We focus in particular on the robustness of the control strategies towards uncertainties that apply to the control and ECCD systems, such as the risk of misalignment of the current deposition or the possible inability to generate narrow current deposition.

The Ion Temperature Gradient (ITG) mode is known to play a crucial role in the ion energy transport in tokamaks. Additionally, ITG-driven turbulence is likely the main source for zonal flows and Geodesic Acoustic Mode (GAM) fluctuations, which are ubiquitously present in tokamaks. A number of kinetic [1, 2] and fluid [3–7] models describing ITG modes have been studied in the literature. Fluid models provide good physics insight into the instability mechanism and are useful for nonlinear numerical modeling. We present a two-fluid model of the toroidal ITG mode, which incorporates the ion inertial (polarization) and gyroviscous drifts into the energy equation at the same order as in the density evolution. This procedure involves the gyroviscous cancellation in both the energy and density equations, which allows the model to provide a consistent description of the ion Finite Larmour Radius (FLR) effects in both equations. The model is asymptically correct in the limit of small but finite ion FLR, k2 ⊥ρ 2 i < 1, and provides a qualitatively correct description in the limit of large ion FLR, k2 ⊥ρ 2 i > 1. The predictions of this model are compared with previous results [5–7] in the local limit. The model is then reformulated into a form suitable for the JOREK code in order to study the global mode structure in realistic tokamak geometry, accounting for all equilibrium profile effects.

1 CEA, IRFM, F-13108, Saint Paul-lez-Durance, France 2 Dep. Applied Physics, T.U. Eindhoven, P.O.B 513, 5600, Eindhoven, Netherlands 3 Max-Planck Institute for Plasma Physics, Boltzmannstr. 2, 85768, Garching Germany 4 Aix-Marseille Université, CNRS, PIIM UMR 7345, 13397, Marseille Cedex 20, France 5 EUROfusion Consortium, JET, Culham Science Center, Abingdon, OX14 3DB, UK * See the author list of ‘Overview of the JET results in support to ITER’, by X. Litaudon et al., to be published in Nuclear Fusion special issue: overview and summary reports from the 26 th

Challenges in plasma physics are wide. Investigation and advances are made in experiments but at the same time, to understand and to reach the experimental limits, accurate numerical simulations are required from systems of nonlinear equations. The numerical challenges of solving the associated fluid equations are discussed in this paper. Using the framework of the finite difference discretization, the most widely used methods for the problems linked to the diffusion or advection operators are presented.

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Tearing modes are MHD instabilities that reduce the performances of fusion devices. They can however be controlled and suppressed using Electron Cyclotron Current Drive (ECCD) as demonstrated in various tokamaks. In this work, simulations of islands stabilization by ECCD-driven current have been carried out using the toroidal nonlinear 3D full MHD code XTOR-2F, in which a current-source term modeling the ECCD has been implemented. The efficiency parameter is computed and its variations with respect to source width and location are computed. The influence of parameters such as current intensity, source width and position with respect to the island is evaluated and compared to the Modified Rutherford Equation. We retrieve a good agreement between the simulations and the analytical predictions concerning the variations of control efficiency with source width and position. We also show that the 3D nature of the current source term can lead to the onset of an island if the source term is precisely applied on a rational surface. We report the observation of a flip phenomenon in which the O-and X-Points of the island rapidly switch their position in order for the island to take advantage of the current drive to grow.

L-H transition features are reproduced using three-dimensional first-principles plasma edge turbulence simulations. A transport barrier is observed to form spontaneously above a threshold of the input power. The physical mechanism relies on the coupling between the equilibrium pressure gradient and the poloidal flow, through both the radial force balance and the neoclassical friction. Accounting for the actual radial profile and time evolution of the latter is key to the barrier formation. It is found that neoclassical friction acts as an energy source for the flow, which largely overcomes the sink due to the turbulent Reynolds stress during the whole barrier lifetime. Importantly, experimentally reported dynamical features are recovered during the formation and lifetime of the barrier. This includes dithering of the radial electric field, which is reminiscent of experimentally observed limit-cycle oscillations and quasi-periodic relaxation oscillations showing similarities with type-III ELMs. These rich dynamics emerge from interplay between turbulence, turbulence-driven flows and the equilibrium flow governed by force balance.

JOREK and IMAGINE A. Fil1, E. Nardon1, M. Hoelzl2, G.T.A Huijsmans3, F. Orain2, M. Becoulet1, P. Beyer4, G. Dif-Pradalier1, R. Guirlet1, H.R. Koslowski5, M. Lehnen3, J. Morales7, S. Pamela6, C. Reux1, F. Saint-Laurent1, C. Sommariva1 and JET contributors6,∗ 1CEA, IRFM, F-13108 Saint-Paul-Lez-Durance, France 2 Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching b. M., Germany 3ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance, France 4Aix-Marseille University, CNRS, PIIM UMR 7345, 13397 Marseille Cedex 20, France 5Forschungszentrum Juelich GmbH, Institute of Energy and Climate Research Plasma Physics (IEK-4), Trilateral Euregio Cluster, 52425 Juelich, Germany 6EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB, UK 7Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas, 1015 Lausanne, Switzerland ∗See the Appendix of F. Romanelli et al., Proceedings of the 25th IAEA Fusion Energy Conference 2014, Saint Petersburg, Russia