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.

Abstract.Progress, since the ITER Physics Basis publication, in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are : energy transport in the SOL in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low and high Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas : refinement of the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a 2 major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed. Introduction.This chapter outlines the significant progress achieved since the ITER Physics Basis in understanding basic scrape-off layer (SOL) and divertor processes in a tokamak. The interaction of plasma with first-wall surfaces will have considerable impact on the performance of fusion plasmas, the lifetime of plasma facing components, and the retention of tritium in next step Burning Plasma E...

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

Type-I edge-localized modes (ELMs) have been mitigated at the JET tokamak using a static external n=1 perturbation field generated by four error field correction coils located far from the plasma. During the application of the n=1 field the ELM frequency increased by a factor of 4 and the amplitude of the D(alpha) signal decreased. The energy loss per ELM normalized to the total stored energy, DeltaW/W, dropped to values below 2%. Transport analyses shows no or only a moderate (up to 20%) degradation of energy confinement time during the ELM mitigation phase.