Scaling of the tokamak near the scrape-off layer H-mode power width and implications for ITER

Abstract

A multi-machine database for the H-mode scrape-off layer power fall-off length, λq in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under the auspices of the International Tokamak Physics Activity. Regression inside the database finds that the most important scaling parameter is the poloidal magnetic field (or equivalently the plasma current), with λq decreasing linearly with increasing Bpol. For the conventional aspect ratio tokamaks, the regression finds , yielding λq,ITER ≅ 1 mm for the baseline inductive H-mode burning plasma scenario at Ip = 15 MA. The experimental divertor target heat flux profile data, from which λq is derived, also yield a divertor power spreading factor (S) which, together with λq, allows an integral power decay length on the target to be estimated. There are no differences in the λq scaling obtained from all-metal or carbon dominated machines and the inclusion of spherical tokamaks has no significant influence on the regression parameters. Comparison of the measured λq with the values expected from a recently published heuristic drift based model shows satisfactory agreement for all tokamaks.

Partial detachment of high power discharges in ASDEX Upgrade

“…Without any power dissipation, about 45 MW/m 2 would be expected from the attached heat load width scaling [5]. The measurements indicate partial detachment, with a moderate midplane-target pressure drop of about a factor 3.…”

“…λ q is understood as the power width at the divertor entrance, S describes the Gaussian shaped power spreading. Typical AUG values (I p = 1 MA) are λ q = 2 mm, S= 1.5 mm and λ int = 4.5 mm, mapped to the outer midplane [4] [5]. Taking into account the flux expansion f x 5, values of λ int 2-2.5 cm are typical along the attached outer target.…”

Pedestal and edge electrostatic turbulence characteristics from an XGC1 gyrokinetic simulation

“…This near-and far-SOL definition is in relation with the divertor heatload width mapped to the outboard midplane that is < 4 mm or ∆ψ N = 0.01 for most of the DIII-D discharges [9]. Time traces of the electron density from this XGC1 simulation at two spatial points at the LFS midplane are shown in Figure 2, one in the middle of the pedestal (ψ N = 0.98) and the other at the boundary (in our definition) of the near-and far-SOL(ψ N = 1.01).…”

20 years of research on the Alcator C-Mod tokamak

“…C-Mod heat footprint data contributed to an international database, extending the range in B T , B P , plasma pressure, and heat flux to ITER-like values in multi-machine empirical scaling studies. 81 The unique diagnostic set on C-Mod also allowed an accurate determination of the sheath transmission factor that relates plasma properties upstream of the sheath to the heat flux conducted to the underlying material. Theoretical calculations predict a value for this factor %7, but experimental measurements of this critical quantity have ranged from 2 to 20 (with the values below 5, physically impossible).…”

“…74 Measurement of the divertor heat flux is only half the battle. Given the narrow deposition footprints that are currently predicted for ITER, 81 methods to reduce the power load to acceptable engineering limits must be found. One solution is to inject a small level of recycling impurities that would radiate near the plasma edge and spread the heat over a larger surface area.…”