Progress on in-vessel poloidal field coils optimization design for alternative divertor configuration studies on the EAST tokamak

An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at reaching over 400 s long-pulse H-mode operations with a full metal wall and a divertor heat load of ˜10 MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water- cooled...

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Veröffentlicht in:	Fusion engineering and design 2019-09, Vol.146, p.2149-2152
Hauptverfasser:	Xiao, B.J., Luo, Z.P., Li, H., Li, G.Q., Wang, L., Wang, Z.L., Xu, G.S., Yao, D.M., Zhou, Z.B., Calabrò, G., Crisanti, F., Castaldo, A., Lombroni, R., Minucci, S., Ramogida, G.
Format:	Artikel
Sprache:	eng
Schlagworte:	Advanced magnetic divertor configurations Configuration management Configurations Cooling systems Design engineering Design optimization Distance between null points EAST Field coils In-vessel coils Liquid cooling Magnetic flux Plasma Plasma currents Plasma density Plasma equilibrium Poloidal flux Tokamak devices Vessels
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Beschreibung
Zusammenfassung:	An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at reaching over 400 s long-pulse H-mode operations with a full metal wall and a divertor heat load of ˜10 MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water- cooled W/Cu plasma face components (PFC) with minimized divertor volume, has been proposed to achieve Te,target
ISSN:	0920-3796 1873-7196
DOI:	10.1016/j.fusengdes.2019.03.126