Validating the SOLPS-ITER drift model using C-Mod data*

To develop practical tokamak fusion reactors, accurate edge plasma transport modeling is critical. The need to properly model impurity transport is particularly acute: injected impurities are necessary to dissipate exhaust power to a level that ensures survival of divertor targets but core impurity contamination must be limited. The advanced capability of the SOLPS-ITER code to capture plasma drifts has made it a focal point of the tokamak community. On the Alcator C-Mod tokamak, exceptionally well-diagnosed experiments with impurity injection have been performed. Available diagnostics include upstream Thomson scattering, target Langmuir probes, divertor neutral pressure, multi-channel spectroscopy, and bolometry. SOLPS-ITER is applied to a 5.4 T, q$_{\mathrm{95}}=$4.9 EDA H-mode, a steady state ELM-free regime, with q$_{\mathrm{\vert \vert }}$ up to 0.4 GW m$^{\mathrm{-2}}$, which has steady phases with and without toroidally symmetric private flux region N$_{\mathrm{2}}$ injection. Initial results for the phase without N$_{\mathrm{2}}$ show clearly that drifts are needed to reproduce measured edge plasma profiles. Progress on modeling with drifts and with nitrogen injection will also be reported.