WEST full tungsten operation with an ITER grade divertor

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

Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

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

2D imaging X-ray spectrometer on WEST : results and technical challenges

International audienceA new imaging X-Ray spectrometer has been installed, aligned and used on the WEST tokamak. A typical experimental spectrum is presented for each of the 3 interchangeable crystals, measuring respectively the Ar XVII, Ar XVIII and Fe XXV spectra. The instrument function exhibits a distortion for the Ar XVII spectrum, presumably due to the crystal splitting in 2 stripes and non-parallelism of the crystal layers. The Ar XVII and Ar XVIII spectra also display blends of unidentified spectral lines, presumably W line

Fluctuation spectra and velocity profile from Doppler backscattering on Tore Supra

International audienceBackscattering of a microwave beam close to the cut-off allows for measurement of density fluctuations ñ( k ⊥ ) at a specified wave-number, selected by the scattering geometry k ⊥ = -2 k i , where k i is the beam wave-number at the reflection layer. On the Doppler reflectometry system installed on Tore Supra, both the scattering wave-number k ⊥ and the scattering localization (r/a) can be changed during the shot owing to the steppable probing frequency and the motorized antenna. Operating in O mode, the spatial and wave-number ranges depend essentially on density profile, typically probing 0.5 < r/a < 0.95 and 2 < k < 15 cm -1 . Wave number spectra are similar to those obtained with conventional scattering systems. The perpendicular fluctuation velocity in the laboratory frame is obtained from the Doppler shift of the frequency spectrum ω = k ⊥ v ⊥ . It is dominated by the plasma E r × B velocity. In the core, the latter is mainly due to the projection of the toroidal velocity, as this is shown by comparison with measurements by charge exchange recombination spectroscopy. In the set of analysed Tore Supra ohmic and ICRH plasmas, the observed rotation is consistent with a poloidal velocity in the electron diamagnetic direction and/or a toroidal velocity in the counter current direction. The detailed structure of the velocity profile, at the edge and in different plasma regimes, allows us then to get information on the radial electric field distribution. The dynamics of the fluctuation velocity can be studied from the time frequency analysis of the signal, for investigating intermittent behaviour and transient regimes

Inter-machine comparison of intrinsic toroidal rotation in tokamaks

Parametric scalings of the intrinsic (spontaneous, with no external momentum input) toroidal rotation observed on a large number of tokamaks have been combined with an eye towards revealing the underlying mechanism(s) and extrapolation to future devices. The intrinsic rotation velocity has been found to increase with plasma stored energy or pressure in JET, Alcator C-Mod, Tore Supra, DIII-D, JT-60U and TCV, and to decrease with increasing plasma current in some of these cases. Use of dimensionless parameters has led to a roughly unified scaling with M-A alpha beta(N), although a variety of Mach numbers works fairly well; scalings of the intrinsic rotation velocity with normalized gyro-radius or collisionality show no correlation. Whether this suggests the predominant role of MHD phenomena such as ballooning transport over turbulent processes in driving the rotation remains an open question. For an ITER discharge with beta(N) = 2.6, an intrinsic rotation Alfven Mach number of M-A similar or equal to 0.02 may be expected from the above deduced scaling, possibly high enough to stabilize resistive wall modes without external momentum input

Effects of Magnetic Shear on Toroidal Rotation in Tokamak Plasmas with Lower Hybrid Current Drive

Application of lower hybrid (LH) current drive in tokamak plasmas can induce both co- and countercurrent directed changes in toroidal rotation, depending on the core q profile. For discharges with q[subscript 0] < 1, rotation increments in the countercurrent direction are observed. If the LH-driven current is sufficient to suppress sawteeth and increase q[subscript 0] above unity, the core toroidal rotation change is in the cocurrent direction. This change in sign of the rotation increment is consistent with a change in sign of the residual stress (the divergence of which constitutes an intrinsic torque that drives the flow) through its dependence on magnetic shear.United States. Dept. of Energy (Contract DE-FC02-99ER54512)United States. Dept. of Energy (Fusion Research Postdoctoral Research Program