Inter-ELM Power Decay Length for JET and ASDEX Upgrade: Measurement and Comparison with Heuristic Drift-Based Model

Abstract

Experimental measurements of the SOL power decay length (λq) estimated from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade, are presented. Data was measured by means of infrared thermography. An empirical scaling reveals parametric dependency λq/mm = 0.73 • B −0.78

Heat-flux footprints for I-mode and EDA H-mode plasmas on Alcator C-Mod

“…The target heat flux profiles are fit using a physics-based parameterization due to Eich [2] that convolves a Gaussian with a decaying exponential function:…”

“…This interest is due in large part to the desire to understand the physics that sets the heat-flux widths and to predict them in ITER [2,5,7,6], since they will play a crucial role in determining the difficulty of an acceptable solution for power handling in the ITER divertor. These multi-machine studies indicate that the width scales inversely with the poloidal field strength, B pol , which when extrapolated to ITER yields a ~1mm width [2,5,6]. Widths this small would substantially increase the difficulty of finding a divertor solution.…”

“…An especially useful functional form that has valuable physical interpretation has been recently proposed by Eich [2] and is now being used for multi-machine comparisons. This form fits the heat-flux footprints to a function that yields a measure of the perpendicular transport into the common-flux region, λ SOL (the λ q parameter in [2]), as well as a measure of the transport into the private flux zone (PFZ) along the outer leg of the divertor, w PFZ (the S parameter in [2]). Another measure of more direct relevance to the power handling requirement is the integral power width [9], λ Int , which relates the peak heat load to the total power incident on the target, i.e.…”

“…Recent work [1][2][3][4][5][6] has focused on measurements of the heat-flux "footprints" on divertor surfaces and heat-flux widths during H-mode confinement in various tokamaks (Alcator C-Mod, AUG, DIII-D, JET, and NSTX). This interest is due in large part to the desire to understand the physics that sets the heat-flux widths and to predict them in ITER [2,5,7,6], since they will play a crucial role in determining the difficulty of an acceptable solution for power handling in the ITER divertor. These multi-machine studies indicate that the width scales inversely with the poloidal field strength, B pol , which when extrapolated to ITER yields a ~1mm width [2,5,6].…”

Divertor asymmetry and scrape-off layer flow in various divertor configurations in Experimental Advanced Superconducting Tokamak

“…Simulations should be introduced into our research to analyze the experiment results, e.g., using SOLPS and heuristic drift-based model 26,27 to study the SOL flow and divertor performance.…”