Abstract
Recent experiments on Tokamak à Configuration Variable have made significant progress toward partial detachment of the outer divertor in neutral beam heated H-mode plasmas in conventional and alternative divertor configurations. The heating power required to enter H-mode was measured in a range of divertor configurations. It is found that at the core densities most favourable for H-mode access, the L–H threshold power is largely independent of the poloidal flux expansion and major radius of the outer divertor, and in the snowflake minus configuration. A factor 2 reduction in the outer divertor power load was achieved in ELM-free (using a fuelling and nitrogen seeding) and ELMy (using nitrogen seeding) H-mode plasmas. No significant reduction in the outer divertor particle flux was achieved in the ELM-free scenarios, compared with ~30% reduction in the most strongly detached ELMy cases. The poloidal flux expansion at the outer divertor was not found to significantly alter the cooling of the divertor in the ELM-free scenarios.
Highlights
Gas puffing into the divertor from valves situated near the floor of Tokamak à Configuration Variable (TCV) consisted of deuterium fuelling directed toward the Scrape-Off Layer (SOL) of the outer leg and nitrogen seeding into the private flux region
Access to partial detachment in ELM-free and ELMy H-mode TCV discharges, where there is a noticeable reduction in the outer divertor power load, has been explored using fuelling and nitrogen injection from the divertor in a range of divertor configurations
It has been observed that the heating power required to cross the L–H transition threshold, where the density is close to the minimum in the PL–H/ne curve, is relatively insensitive to the outer target poloidal flux expansion or major radius in the single null configuration, nor is there a large difference in threshold power between these variants on the single null and the snowflake minus configurations studied
Summary
Several potential solutions are currently being explored to address this challenge They include introducing impurities into the confined plasma to exhaust a significant fraction of the heating power as radiation and alternatives to the conventional divertor configuration, such as the X-divertor [5, 6], snowflake [7], Super-X [8], X-point target [9] and the fluxtube-expansion divertor [10].
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