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One of the missions of the WEST tokamak is to test in a realistic tokamak environment, the ITER-like divertor plasma facing components made with tungsten. On exposed leading edges of monoblocks, poloidally-distributed cracks having an average spacing of 0.4 mm, running perpendicular to the cooling tube axis, were observed following the experimental campaigns in 2018. Damage may be induced by different processes which can lead to: brittle fracture below the Ductile to Brittle Transition temperature or ductile failure for which softening (recovery/recrystallization) process plays a major role. To improve our understanding about the leading edge damage process, numerical simulations were run here. The following results are specially studied: (i) the thermal gradients; (ii) the softening fraction gradient and (iii) the stress and strain distributions taking into account the mechanical properties of tungsten (elastic-viscoplastic) and the softening phenomenon. This paper describes the results obtained for a range of WEST steady state parallel heat flux (from 45 MW/ m2 to 70 MW/ m2 ) and disruption (600 MW/ m2 ) heat loading on the leading edge. Estimated results related to the plastic strain accumulation, give an interpretation of the premature cracking of the monoblock leading edge but do not explain the poloidal distribution. According to the numerical results, brittle fracture are expected under disruption as estimated normal stresses are beyond the material yield stress, along 88% of the leading edge.

The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m−2 for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 μm followed by a re-solidified tungsten bump of 200 μm in the JxB direction.

Plasma-facing components receive enormous heat fluxes under steady state and especially during transient conditions that can even lead to tungsten (W) melting. Under these conditions, the unimpeded ...

15th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC), Aix en Provence, FRANCE, MAY, 2015; International audience; Simulations of impurity trajectories in deuterium plasmas in the vicinity of the surface are performed by taking into account the magnetic sheath in conditions relevant for ITER and WEST. We show that the magnetic sheath has a strong effect on the average impact angle of impurities in divertor conditions and that it can lead to an increase of approximate to 60% at the gross erosion maximum for neon (Ne+4) compared to the case when only the cyclotron motion is considered. The evaluation of the net erosion has been undertaken by retaining local redeposition of tungsten (W). We investigate how it is affected by the sheath magnetic potential profile. The largest effect is however observed when an energy distribution is considered. In this case the number of particles that manage to exit the sheath is larger as it is dominated by the more energetic particles. The comparison with other work is also discussed. The application to a scenario of the WEST project is finally performed, which exhibits a moderate, however non negligible, erosion of the plasma facing components.

In the new CIEL configuration of Tore Supra, all the plasma facing components are actively cooled. The surface area of the wall that is covered with carbon measures about 15 m2 (2 to 4 m2 of which in close interaction with the plasma). Steady-state plasma conditions up to 4 min 25 s have been maintained in this configuration. In these experiments, the required gas injection to maintain the prescribed density remains constant during the whole discharge. The exhausted flux is also constant and equal to 40 ÷ 50% of the injected flux. Therefore, 50 ÷ 60% of the injected particles remain trapped in the vessel, the total retention being proportional to the plasma duration. Since the amount of gas recovered between shots or by He-glow discharges does not always balance the injected gas, it follows that a quantity of deuterium remains indefinitely trapped in the vessel, which appears as an infinite reservoir. This reservoir is believed to be dominated by co-deposited layers, as observed in several places of the vessel. The thickest deposits (up to 800 µm) are observed on the leading edge of the neutralizers of the pump limiter. They display a column-like shape (typical growth rate ~20 nm/s) and have a graphite-like structure. Their deuterium concentration is D/C ~ 1%. Conversely, in regions that are shadowed from the direct plasma flux, the deposits show a smoother shape and their deuterium content is typically ~10 ÷ 15%.