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

J. Gaspar (1, 2) , Yann Corre (3) , Clarisse Bourdelle (3) , S. Brezinsek (4) , J. Bucalossi (3) , N. Chanet (3) , R. Dejarnac (5) , N. Fedorczak (3) , M. Firdaouss (3) , J-L. Gardarein (2) , J. P. Gunn (3) , Guillaume Laffont (6) , T. Loarer (3) , C. Pocheau (3) , E. Tsitrone (3)

1 AMU - Aix Marseille Université
2 IUSTI - Institut universitaire des systèmes thermiques industriels
3 IRFM - Institut de Recherche sur la Fusion par confinement Magnétique
4 IEK-4 - Institut fur Energie und Klimaforschung - Plasmaphysik
5 Institute of Plasma Physics [Praha]
6 LCFO (CEA, LIST) - Laboratoire Capteurs Fibres Optiques (CEA, LIST)

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 (IR) thermography, thermal measurement with thermocouples (TC) and fiber Bragg grating (FBG) embedded few mm below the surface and flush mounted Langmuir probes (LP). 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 as function of the continuous progress achieved in WEST : from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (up to 8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) steady state experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1cm) 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/m² with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW/m² 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 close to the 2/3 – 1/3 commonly observed in forward-B divertor configuration.

Mots clés: instrumentation; nuclear instrumentation; tokamak; magnetic equilibrium; heat load; heat flux; plasma; diagnosis; Optical Fiber Sensor; Fiber Bragg Grating; sensor; thermal measurement; WEST; ITER; Langmuir probe; infrared thermography; divertor

Domaines: Instrumentations et Détecteurs [physics.ins-det]; Physique des plasmas [physics.plasm-ph]; Thermique [physics.class-ph]; Optique / photonique

J. Gaspar, Yann Corre, Clarisse Bourdelle, S. Brezinsek, J. Bucalossi, et al.. Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST. IAEA FEC 2021 - The 28th IAEA Fusion Energy Conference, May 2021, Nice (Virtual conference), France. ⟨cea-03252519⟩