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...

First experiments with non-axisymmetric magnetic perturbations, toroidal mode number n = 2, produced by newly installed in-vessel saddle coils in the ASDEX Upgrade tokamak show significant reduction of plasma energy loss and peak divertor power load associated with type-I Edge Localized Modes (ELMs) in high-confinement mode plasmas. ELM mitigation is observed above an edge density threshold and is obtained both with magnetic perturbations that are resonant and not resonant with the edge safety factor profile. Compared with unperturbed type-I ELMy reference plasmas, plasmas with mitigated ELMs show similar confinement, similar plasma density and lower tungsten impurity concentration.

Abstract:Plasma-wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strikepoint tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor / Be first wall and all-W or all-C.One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q=10 ITER discharge [ 1 ] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4±3 higher, this margin has been adopted as uncertainty of the scaling.With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated:• It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ.• Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated.• For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms.• For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account.Finally the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.