Heat loads on poloidal and toroidal edges of castellated plasma-facing components in COMPASS

Abstract

Dedicated experiments have been performed in the COMPASS tokamak to thoroughly study the power deposition processes occurring on poloidal and toroidal edges of castellated plasma-facing components in tokamaks during steady-state L-mode conditions. Surface temperatures measured by a high resolution infra-red camera are compared with reconstructed synthetic data from a 2D thermal model using heat flux profiles derived from both the optical approximation and 2D particle-in-cell (PIC) simulations. In the case of poloidal leading edges, when the contribution from local radiation is taken into account, the parallel heat flux deduced from unperturbed, upstream measurements is fully consistent with the observed temperature increase at the leading edges of various heights, respecting power balance assuming simple projection of the parallel flux density. Smoothing of the heat flux deposition profile due to finite ion Larmor radius predicted by the PIC simulations is found to be weak and the power deposition on misaligned poloidal edges is better described by the optical approximation. This is consistent with an electron-dominated regime associated with a non-ambipolar parallel current flow. In the case of toroidal gap edges, the different contributions of the total incoming flux along the gap have been observed experimentally for the first time. They confirm the results of recent numerical studies performed for ITER showing that in specific cases the heat deposition does not necessarily follow the optical approximation. Indeed, ions can spiral onto the magnetically shadowed toroidal edge. Particle-in-cell simulations emphasize again the role played by local non-ambipolarity in the deposition pattern.

The very high spatial resolution infrared thermography on ITER-like tungsten monoblocks in WEST Tokamak

“…1) [2]. The heat loads deposited per unit of area can then lead to a local melting of components and irreversible damages [3], [4]. Designs are proposed to reduce the risk induced by misalignments.…”

Conceptual design of reciprocating probes and material-testing manipulator for tokamak COMPASS Upgrade

“…Such high heat load is limiting even for estimated ∼ 10 2 -10 3 ms long discharges for most of materials. On the other hand, it provides a worldwide opportunity to further investigate priority tasks for ITER with tungsten melting [10] and pre-damaged material testing [11], or plasma deposition into gaps in between plasma facing components [1] (carried out in inner-wall limited discharges on COMPASS with two follow-up experiments performed in divertor of AUG and WEST). Experiment on liquid metals on COM-PASS tokamak in 2019 [2,7] was relevant thanks to the simple shape of a quarter of a cylinder of the tile used there, see figure 3(b).…”

RF wave coupling, plasma heating and characterization of induced plasma-material interactions in WEST L-mode discharges

“…targets [28], while some are reciprocating [29] either from the APL (Fig. 1) or from the top of the device, with design immune to sheath expansion [30], therefore favourable for measurements in RF environment.…”

“…In addition to these core diagnostics, the edge plasma is monitored by diagnostics such as probes and visible spectroscopy. Some of the probes are flush-mounted along with the divertor targets [28], while some are reciprocating [29] either from the APL (figure 1) or from the top of the device, with a design immune to sheath expansion [30], therefore favourable for measurements in RF environment.…”