Scrape-off layer properties of ITER-like limiter start-up plasmas in JET

Abstract

Recent experiments at JET combining reciprocating probe measurements upstream and infrared thermography at the plasma-facing components (PFC) on plasmas in limiter configurations show that the common approach to predicting the power load on the limiter underestimates the heat flux at the contact point by a factor 1.5–3. The current model and scaling laws used for predicting the power load onto the first wall during limiter current ramp-up/down in ITER are uncertain and a better understanding of the heat transport to the PFCs is required. The heat loads on PFCs are usually predicted by projecting the parallel heat flux associated with scrape-off layer (SOL) properties at the outer mid-plane (upstream) along the magnetic field lines to the limiter surface and deducing the surface heat flux through a cosine law, thus ignoring any local effect of the PFC on transport within the SOL. The underestimate of the heat flux is systematic in inner wall limiter configurations, independent of the plasma parameters, whereas in outer limiter configuration this is not observed, probably because of the much shorter SOL power decay length. Models that can explain this enhanced heat flux around the contact point are proposed and discussed although no definitive conclusion can be drawn.

Understanding and suppressing the near scrape-off layer heat flux feature in inboard-limited plasmas in TCV

“…In all previous dedicated experiments [3,4,6,7], the near SOL of inboard-limited plasmas had always been observed at the HFS. Still, no consensus could be reached on the presence of the near SOL on the LFS, being evident in C-Mod [7], not in DIII-D [6] and COMPASS [5], while no outboard measurements were available in TCV in Ref.…”

“…However, recent infrared (IR) thermography and Langmuir probes (LPs) measurements of inboardlimited L-mode plasmas in many tokamaks [3,4,5,6,7] have shown the ubiquitous presence of two regions in the SOL: a "near" SOL, extending typically a few mm from the LCFS, characterized by a steep gradient of the parallel heat flux, and a "far" SOL, typically a few cm wide, featuring flatter heat flux profiles. The parallel heat flux radial profile in the SOL is therefore better described by a sum of two exponentials q || (r u ) = q n exp(−r u /λ n ) + q f exp(−r u /λ f ) ,…”

“…(1). This can lead to severe damages to the limiter, as observed in JET [3]. The design of ITER First Wall (FW) panels has been changed to handle the heat flux associated with the near SOL [10].…”

Plasma turbulence in the scrape-off layer of the ISTTOK tokamak

“…[2][3][4][5] These theoretical investigations 6 focused mainly on the toroidally limited SOL, 7-9 a configuration that is relevant to the ITER start-up and ramp-down phases during which the inner or the outer vessel wall will be used as the limiting surface. 10,11 In this scenario, using low-frequency fluid models, the turbulent regimes were identified. It was found that drift waves (DW) and ballooning modes (BM) drive the plasma turbulent dynamics, with the resistive BM being the main drive in typical existing tokamak conditions, 12 a result in agreement with previous experimental results.…”

Effect of the limiter position on the scrape-off layer width, radial electric field and intrinsic flows

“…Since this part of the tokamak vessel is not designed to handle large particle and heat fluxes as are the divertor plates, the start-up scenario must be carefully tailored to minimize its power load. In the last few years, dedicated experiments have been conducted in a number of tokamaks in order to characterize the fluctuations and equilibrium profiles in the scrape-off layer (SOL) of limited plasmas [2][3][4][5][6][7][8]. Substantially large asymmetries have been observed between the different explored configurations, e.g.…”