Measure and properties of heat load profiles on the lower tungsten divertor of WEST in L-mode experiments

Measure and properties of heat load profiles on the lower tungsten divertor of WEST in L-mode experimentsN. Fedorczaka,*, J. Gasparb, J.P. Gunna, Y. Correa, A. Grosjeana, R. Mitteaua, C. Desgrangesa, R. Dejarnacc, M. Dimitrovac,e, Tsv. K. Popovd,e, J. Bucalossia, E. Tsitronea, T. Loarera, S. Brezinsekf and the WEST team* a CEA, IRFM, F-13108 Saint-Paul-Lez-Durance, Franceb Aix Marseille Univ, CNRS, IUSTI, Marseille, Francec Institute of Plasma Physics, The Czech Academy of Sciences, 182 00 Prague 8, Czech Republicd NIS-Faculty of Physics, St Kliment Ohridski University of Sofia, 1164 Sofia, Bulgariae Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, 1784 Sofia, Bulgariaf Forschungszentrum Jülich, Institut für Energie- und Klimaforschung Plasmaphysik, 52425 Jülich, Germany* See http://west.cea.fr/WESTteamDuring the first phase of WEST tokamak operation, the main lower divertor was composed of a mix of actively cooled tungsten monoblocs and uncooled tungsten-coated graphite plasma-facing components. It exhausted approximately 7.5GJ of conducted energy over 2400 pulses. A set of complementary diagnostics allowed characterizing properties of divertor heat & particle loads on the inertial components: arrays of embedded thermocouples and fiber Bragg gratings, infra-red thermography on both strike point areas within the same field-of-view, and arrays of flush-mounted Langmuir probes. Heat load properties are investigated across a wide range of L-mode conditions: Deuterium and Helium plasmas, from 1 to 9 MW of additional power (LHCD+ICRH), plasma currents from 300 kA to 800 kA at nominal magnetic field BT=3.6T, and various X-point height with respect to the flat divertor (impacting both strike point position & target magnetic flux expansion). We report on the continuous effort made to benchmark the diagnostics and improve the accuracy of heat load analyses. Flushed Langmuir probes were cross calibrated during slow strike point sweepings over the probe arrays, confirming the accurate alignment of the flushed collectors with respect to tile surface. In addition, comparisons with a protruding probe proved the accuracy of a simple model for effective collection area of flushed probes, validating the concept of flushed probes for measuring particle fluxes and electron temperature. Surface heat loads were conjointly estimated from embedded thermal sensors and infra-red thermography at . Arrays of fiber Bragg gratings were installed a few millimetres below the surface of uncooled tiles, providing local bulk temperature dynamics with a spatial resolution of 10mm. This resolution is sufficient to invert temperatures into surface heat load, assuming some geometrical properties of heat load pattern. In particular it was assumed that heat load patterns follow a common gaussian-exponential shape, with unknown widths and amplitude. This shape was confirmed by infra-red thermography. On the other hand, thermography was proven to suffer from ambience reflections and change of surface emissivity during the campaigns, which add difficulties in estimating absolute heat load values from thermography inversion. This critical issue in metallic devices with low surface emissivity is addressed by combining thermography and embedded thermal measurements [1]. Resulting reflection and emissivity maps are shown to vary strongly in space, possibly due to evolving surfaces under erosion and redeposition. Finally, global properties of the lower divertor heat flux pattern in WEST large aspect ratio tokamak will be discussed and confronted to current knowledge of L-mode diverted regimes. [1] J. Gaspar, this conference. *Corresponding author: tel.: +33442253712 e-mail: [email protected]

Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

As in many of today's tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, ${{q}_{||}}$ in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as ${{q}_{||}}={{q}_{0}}\text{exp} ~\left(-r/\lambda _{q}^{\text{omp}}\right)$ , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, $\lambda _{q}^{\text{omp}}$ . The initial choice of $\lambda _{q}^{\text{omp}}$ , which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with $R=\text{0}\text{.4--2}\text{.8}\,\text{m},\,{{B}_{0}}=\text{1}\text{.2--7}\text{.5}\,\text{T},\,{{I}_{\text{p}}}=\text{9--2500}\,\text{kA}.$ Measurements of $\lambda _{q}^{\text{omp}}$ in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar predicted values of $\lambda _{q}^{\text{omp}}$ mapped to the outside midplane. The engineering scaling with the highest statistical significance, $\lambda _{q}^{\text{omp}}=10{{\left({{P}_{\text{tot}}}/V\,\left(\text{W}\,{{\text{m}}^{-3}}\right)\right)}^{-0.38}}{{\left(a/R/\kappa \right)}^{1.3}}$ , dependent on input power density, aspect ratio and elongation, yields $\lambda _{q}^{\text{omp}}$   =  [7, 4, 5] cm for ${{I}_{\text{p}}}$   =  [2.5, 5.0, 7.5] MA, the three reference limiter plasma currents specified in the ITER heat and nuclear load specifications. Mapped to the inboard midplane, the worst case (7.5 MA) corresponds to $\lambda _{q}^{\text{imp}}\sim 57\pm 14$ mm, thus consolidating the 50 mm width used to optimize the FW panel toroidal shape