Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

0552289 - ÚFP 2022 RIV AT eng J - Journal Article
Gaspar, J. - Corre, Y. - Fedorczak, N. - Gunn, J. P. - Bourdelle, C. - Brezinsek, S. - Bucalossi, J. - Chanet, N. - Dejarnac, Renaud - Firdaouss, M. - Gardarein, J.-L. - Laffont, G. - Loarer, T. - Pocheau, C. - Tsitrone, E.
Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST.
Nuclear Fusion. Roč. 61, č. 9 (2021), č. článku 096027. ISSN 0029-5515. E-ISSN 1741-4326
EU Projects: European Commission(XE) 633053 - EUROfusion
Institutional support: RVO:61389021
Keywords : diagnostics comparison * divertor * heat load pattern * infrared * Langmuir probes * thermocouples
OECD category: Fluids and plasma physics (including surface physics)
Impact factor: 4.215, year: 2021
Method of publishing: Limited access
https://iopscience.iop.org/article/10.1088/1741-4326/ac1803

WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared thermography, thermal measurement with thermocouples and fibre Bragg grating embedded few mm below the surface and flush mounted Langmuir probes. A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions: from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1 cm) is obtained between thermal inversions and flush-mounted LP measurements. The peak heat flux from the whole set of diagnostics is in good agreement and mainly in the 20% range, while the heat flux decay length reported on the target shows significant discrepancy between diagnostics and location in the machine ( 40% range). Despite such discrepancy, heat flux decay length at target is found to scale mainly with the magnetic flux expansion through the variation of the X-point height, as expected. The improved plasma performances achieved during C4 enabled to reach significant heat load in the divertor, up to 6 MW m-2 with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW m-2 steady state) with about 7 MW of additional power in L-mode discharge. The heat load distribution is clearly asymmetric with a 3/4 and 1/4 distribution on the outer and inner strike point region respectively for the parallel heat flux.
Permanent Link: http://hdl.handle.net/11104/0327418