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Description
The Divertor Tokamak Test (DTT) facility is a large experiment under design and construction at the ENEA Research Centre in Frascati, Italy. Its main goal is to assess alternative solutions for the heat and power exhaust problem in future fusion plants [1]. To this end, different configurations will be tested, and safe operation must be ensured in all of them.
One of the major challenges in this framework is plasma–wall interaction (PWI) [2]. Materials may undergo erosion, limiting component lifetime, while eroded impurities can migrate into the core plasma, causing dilution and cooling. These effects must be carefully monitored and controlled in all DTT scenarios.
This work presents an overview of erosion/deposition and core plasma contamination studies for the DTT tungsten divertor using ERO2.0, a 3D Monte-Carlo impurity transport code [3]. Two magnetic configurations are analysed: a single-null positive triangularity (PT) scenario and a negative triangularity (NT) one. Both are investigated at full power with neon (Ne) seeding to achieve detachment, while an attached low-power PT case in pure deuterium (D) is also modelled. In PT scenarios, the impact of Edge Localised Modes (ELMs) is estimated using a novel approach based on literature data [4,5].
The two PT scenarios show comparable net erosion rates, driven by the higher power and the presence of heavier Ne ions in detached plasma. Impurity effects, modelled including oxygen (O) as a proxy, are more significant in the attached case with pure D plasma. For the intra-ELM phase, a D ion burst is modelled by varying its impact angle, energy, and affected surface area. Among these, the impact angle proves to be the most critical parameter, with erosion rates reaching a few tens of nm/s. From a migration perspective, W core contamination from the divertor remains below acceptable limits in all scenarios. The attached plasma exhibits enhanced prompt redeposition near the outer strike point due to higher temperatures, whereas the detached case shows superior screening, effectively suppressing W influx into the core. In the NT scenario, the absence of ELMs results in Ne and O erosion dominating, similarly to PT inter-ELM conditions. The extrapolation of the SOLPS-ITER plasma solution to the divertor surface is particularly relevant for erosion and migration estimates in this case, given the mesh distance to the surface.
Local simulations including the 3D geometry of the divertor tiles are also performed to assess optimal positions for PWI diagnostics in the divertor region.
[1] Romanelli F. et al., NF 64.11 (2024) 112015
[2] Roth J. et al., JNM 390–391 (2009) 1–9
[3] Romazanov J. et al., PS T170 (2017) 014018
[4] Kirschner A. et al., NME 18 (2019) 239244
[5] Kumpulainen H.A. et al., NME 33 (2022) 101264
