Link do spotkania w aplikacji Microsoft Teams: https://teams.microsoft.com/meet/330417189011217?p=R3Lm2ghE8KDW6NCXvq

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Kod dostępu: CY6iu79t

Abstract

In tokamaks, the plasma density outside of the confined plasma, more specifically in the region spanning between just outside (few mm/cm) of the plasma boundary and the main chamber walls of the tokamak, has been long known to exhibit a lower radial decay than classic collisional/orbit transport would predict. This is because in this region (also dubbed the far scrape-off layer, or far-SOL) particle (and also heat) transport is dominated by highly convective intermittent turbulence, detectable as plasma 'blobs' (filaments) ejected from the confined plasma and which propagate radially outwards towards the walls. Empirically, radial density profile flattening in the SOL can vary strongly among different plasma operating scenarios, and lead to the formation of so-called 'density shoulders' with changing slopes in the far-SOL.

Radial SOL profile flattening is problematic not only from the point of view of safe power exhaust, but also it can lead to increased high-Z impurity production in the main chamber. In particular, with the recent decision to fully cover the inner walls of the next step device ITER (which is presently under construction in France) with tiles made of Tungsten (W), the W source from the main chamber poses a new risk. State of the art plasma-wall-interaction simulations for ITER, using different combinations of electron density (ne) and ion temperature (Ti) reached in front of the main chamber tiles as input, show that W sputtering from the main chamber could far outweigh other W sources. The combination of high ne and high Ti is the worst case scenario, yielding the highest W source. However, these inputs are hypothetical: At present there is no reliable prediction for what ne & Ti at that location in ITER will be. In order to constrain these simulations for the 'new' full-W ITER, it has become critical to advance the physics basis to extrapolate far-SOL profile measurements from current machines to ITER with sufficient accuracy.

In the present work, electron density profiles in the far-SOL of JET tokamak H-mode plasmas are carefully evaluated by using profile (FM-CW) reflectometry in conjunction with high resolution Thomson Scattering and far-infrared interferometry. The analysis is applied to a set of density ramp experiments covering a variety of conditions. Despite the dataset's heterogeneity, a turbulence-related quantity known as the 'edge turbulence control parameter' (alpha_t) evaluated at the plasma boundary radius is found to unify the measured radial decay length of electron density ne in the far-SOL for the entire dataset, typically to within ±25%. The alpha_t parameter was first introduced by T. Eich and co-workers [Nucl. Fusion 2020 056016] and was used to describe the turbulence state of the plasma boundary, in particular to describe the relative importance between drift-wave induced and interchange induced turbulent transport. In the present work, a direct proportionality between alpha_t and the density radial decay length is found over the entire alpha_t range accessed during these experiments. The above result identifies alpha_t as a good descriptor to quantify far-SOL density flattening on JET. It supports and further expands (to larger tokamak size and for a broad range of conditions) previous findings obtained on smaller (medium-sized) tokamak machines like TCV [A.Stagni Nucl. Fusion 62 (2022) 096031, A.Stagni Nucl. Fusion 64 (2024) 026016] or ASDEX Upgrade [A.Redl Nucl.Fusion 64 (2024) 086064] that also indicated that the SOL density flattening is set by the balance between drift wave transport and interchange transport. What renders this result particularly relevant is that the parameter alpha_t (unlike other previously used parameters linking to divertor detachment criteria, which rely on atomic physics) offers a comparatively straightforward route to extrapolate from current devices to ITER.

Another important aspect is what sets the radius of the outer inflection of SOL density shoulders and whether it is possible to control it by external means. Analysis of the JET experiments run with different plasma shapes and with a higher than usual clearance to the outer wall have revealed that the inflection radius is set by the high SOL density 'contact point' (nearest magnetic field line intersection) with the main chamber. The JET experiments show it is possible to move the outer inflection radius of SOL density shoulders away from the outer midplane region (main chamber low field side limiters) through careful equilibrium design, in this case by first intersecting field lines with the low field side divertor shoulder. This is good news because recent 2D simulations of tungsten erosion and edge-to-core transport of JET discharges [H.A. Kumpulainen, IAEA-FEC 2025, submitted to NF] have shown that moving the W source away from the near outer midplane to divertor vicinity reduces substantially the W penetration efficiency into the main plasma (by a factor of order 2000). Thus, the density shoulder results on JET suggest a possible path in ITER to greatly reduce the W ingress into the core plasma.