Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

The XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates. The simulation results compare well against the empirical scaling λ _q ∝ 1/Bγ_p obtained from present tokamak devices, where λ _q is the divertor heat-flux width mapped to the outboard midplane, γ = 1.19 as found by Eich et al (2013 Nucl. Fusion 53 093031), and B _P is the magnitude of the poloidal magnetic field at the outboard midplane separatrix surface. This empirical scaling predicts λ _q ≲ 1 mm when extrapolated to ITER, which would require operation with very high separatrix densities (n _sep/n _Greenwald > 0.6) (Kukushkin et al 2013 J. Nucl. Mater. 438 S203) in the Q = 10 scenario to achieve semi-detached plasma operation and high radiative fractions for acceptable divertor power fluxes. Using the same simulation code and technique, however, the projected λ _q for ITER's model plasma is 5.9 mm, which could be suggesting that operation in the ITER Q = 10 scenario with acceptable divertor power loads may be obtained over a wider range of plasma separatrix densities and radiative fractions. The physics reason behind this difference is, according to the XGC1 results, that while the ion magnetic drift contribution to the divertor heat-flux width is wider in the present tokamaks, the turbulent electron contribution is wider in ITER. Study will continue to verify further this important projection. A high current C-Mod discharge is found to be in a mixed regime: While the heat-flux width by the ion neoclassical magnetic drift is still wider than the turbulent electron heat-flux width, the heat-flux magnitude is dominated by the narrower electron heat-flux.