Plans for Liquid Metal Divertor in Tokamak Compass

0491780 - ÚFP 2019 RIV RU eng J - Journal Article
Horáček, Jan - Entler, Slavomír - Vondráček, Petr - Adámek, Jiří - Šesták, David - Hron, Martin - Pánek, Radomír - Dejarnac, Renaud - Weinzettl, Vladimír - Kovařík, Karel - Van Oost, G.
Plans for Liquid Metal Divertor in Tokamak Compass.
Plasma Physics Reports. Roč. 44, č. 7 (2018), s. 652-656. ISSN 1063-780X. E-ISSN 1562-6938.
[International Symposium on Liquid metals Applications for Fusion (ISLA2017) /5./. Moscow, 25.09.2017-27.09.2017]
R&D Projects: GA MŠMT(CZ) LM2015045; GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-14228S; GA MŠMT(CZ) 8D15001
EU Projects: European Commission(XE) 633053 - EUROfusion
Institutional support: RVO:61389021
Keywords : liquid metal * plasma-facing component * tokamak * divertor
OECD category: Fluids and plasma physics (including surface physics)
Impact factor: 0.941, year: 2018
https://doi.org/10.1134/S1063780X18070024

The COMPASS tokamak (R = 0.56 m, a = 0.2 m, BT = 1.3 T, Ip ~ 300 kA, pulse duration 0.4 s) operates in ITER-like plasma shape in H-mode with Type-I ELMs. In 2019, we plan to install into the divertor a test target based on capillary porous system filled with liquid lithium/tin. This single target will be inclined toroidally in order to be exposed to ITER-relevant surface heat flux (20 MW/m2). Based on precisely measured actual heat fluxes, our simulations predict (for 45° inclination, without accounting for the lithium vapor shielding) the surface temperature rises up to 700°C within 120 ms of the standard ELMy H-mode heat flux with ELM filaments reaching hundreds MW/m2. Significant lithium vaporization is expected. The target surface will be observed by spectroscopy, fast visible and infrared cameras. The scientific program will be focused on operational issues (redeposition of the evaporated metal, ejection of droplets, if any) as well as on the effect on the plasma physics (improvement of plasma confinement, L–H power threshold, Zeff, etc.). After 2024, a closed liquid divertor may be installed into the planned COMPASS Upgrade tokamak (R = 0.84 m, a = 0.3 m, BT = 5 T, Ip = 2 MA, Pin = 8 MW, pulse duration ~2 s) with ITER-relevant heat fluxes loading the entire toroidal divertor.
Permanent Link: http://hdl.handle.net/11104/0285522