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International audience; The presence of magnetic dust can be an important issue for future fusion reactors where plasma breakdown is critical. Magnetic dust has been collected from contemporary fusion devices (FTU, Alcator C-Mod, COMPASS and DIII-D) that feature different plasma facing components. The results of morphological and elemental analysis are presented. Magnetic dust is based on steel or nickel alloys and its magnetism is generated by intense plasma-material interactions. In spite of the strong similarities in terms of morphology and composition, X-ray diffraction analysis revealed differences in the structural evolution that leads to non-trivial magnetic responses.

Boronization is a Plasma Facing Component (PFC) conditioning technique widely used in tokamak machines. The National Spherical Torus Experiment-Upgrade (NSTX-U) applied this conditioning, using a plasma glow with a deuterated Trimethyl-boron (d-TMB) and He mixture. The use of boronization during the campaign improved the plasma performance, allowing longer plasma discharges and H-mode access. The chemical state of an ATJ graphite sample, used as a proxy for the NSTX-U PFCs, was monitored in-situ using the Materials Analysis Particle Probe (MAPP) diagnostic and X-ray Photoelectron Spectroscopy (XPS). The XPS data showed a progressive rise (from

As part of an effort to validate the edge plasma model in the SOLPS-ITER code suite under ITER-relevant divertor plasma and neutral conditions, we report on progress in the modeling of the Alcator C-Mod divertor plasma with the new code. We perform simulations with a complete drifts model and kinetic neutrals, including effects of neutral viscosity, ion-molecule collisions and Lyα-opaque conditions, but assuming a pure deuterium plasma. Through a series of simulations with varying divertor geometries, we show the importance of including neutal leakage paths through the divertor substructure on the divertor plasma solution. Moreover, the impact of drifts on inner-outer target asymmetries is assessed. Including both effects, we achieve excellent agreement between simulations and upstream and outer target Langmuir Probe data. In absence of strong volumetric losses due to e.g. impurity radiation in our simulations, the strong inner target detachment observed experimentally remains elusive in our modeling at present. Keywords: SOLPS-ITER, Alcator C-Mod, Divertor modeling

A study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE (Rognlien et al., J. Nucl. Mater. 196, 347, 1992). The model parameters are based on the ADX tokamak concept design (LaBombard et al., Nucl. Fusion 55, 053020, 2015). Several long-legged divertor configurations are considered, in particular the X-point target configuration proposed for ADX, and compared with a standard divertor. For otherwise identical conditions, a scan of the input power from the core plasma is performed. It is found that as the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remaining stable. At low power the detachment front eventually moves to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, the operation window for a detached divertor in the standard divertor configuration is very small, or even non-existent; under the same conditions for long-legged divertors the detached operation window is quite large, in particular for the X-point target configuration, allowing a factor of 5–10 variation in the input power. These modeling results point to possibility of stable fully detached divertor operation for a tokamak with extended divertor legs., United States. Department of Energy (Contract DE-AC52-07NA27344)

Experiments indicate that the formation of density shoulders in the far scrape-off layer (SOL) is associated with an increase in cross-field transport and the propagation of coherent structures of high density known as blobs. It has been suggested that far SOL divertor collisionality (Λdiv) may be a controlling parameter in this physics, which could be challenging for heat flux mitigation schemes. At Alcator C-Mod, a new high spatial resolution divertor Langmuir probe array (rail probes) combined with a scanning Mirror Langmuir probe has been used to investigate this phenomenon in detail. Experiments were performed to mitigate divertor heat flux and thus increase Λdiv by two different methods and observe the divertor and scrape-off layer plasma response: (1) a 3-point density scan is found to increase the fluctuation intermittency at both the midplane and divertor, and produce a density shoulder at elevated Greenwald fraction; (2) a 3-point N2 seeding scan is found to increase the divertor strike point collisionality, by over two orders of magnitude yet no noticeable trend is observed in either the density profile or the fluctuation statistics. It is concluded that Λdiv near the strike point is not by itself a controlling parameter for SOL fluctuations and shoulder formation which is important for localized heat flux mitigation schemes targeting the strike point where heat flux is the greatest.

Growth of tungsten nano-tendrils (“fuzz”) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The depth of the W fuzz layer (600 ± 150 nm) is consistent with an empirical growth formula from the PISCES experiment. Re-creating the C-Mod exposures as closely as possible in Pilot-PSI experiment can produce nearly-identical nano-tendril morphology and layer thickness at surface temperatures that agree with uncertainties with the C-Mod W probe temperature data. Helium concentrations in W fuzz layers are measured at 1–4 at.%, which is lower than expected for the observed sub-surface voids to be filled with several GPa of helium pressure. This possibly indicates that the void formation is not pressure driven., United States. Dept. of Energy (Award DE-SC00-02060)

A model is developed which constrains heat width, λr based on global power balance, momentum conservation, pedestal stability and sheath heat transmission. The model relies on measurements of the ratio of separatrix to pedestal pressure; a ratio ∼5% is found to be expected for ITER. Applying this model indicates a constraint that the allowed λr ∼ 10–30 mm for ITER if the divertor is in the high-recycling regime as expected (T < 20 eV) while a λr ∼ 1–3 mm requires a separatrix pressure approximately equal to the top pedestal pressure in violation of physical reasoning and the concept of a pedestal. A weaker constraint is applied in the model that upstream separatrix temperature simultaneously satisfies power balance. The constrained model cannot satisfy power balance with λr < 3 mm, and in order to obtain λr ∼ 5 mm requires divertor plasma temperature >100 eV, a condition which would have very negative consequences for the divertor, but has never been observed experimentally., United States. Department of Defense (US DOE award DE-SC00-02060), United States. Department of Defense (contract DE-FC02-99ER54512), United States. Department of Defense (cooperative agreement DE-FC02- 04ER54698)