The work of the ITPA SOL/divertor group is reviewed and implications for ITER discussed. Studies of near SOL gradients have revealed a connection to underlying turbulence models. Analysis of a multi-machine database shows that parallel conduction gradients near the separatrix scale as major radius. New SOL measurements have implicated low-field side transport as driving parallel flows to the inboard side. The high-n nature of ELMs has been elucidated and new measurements have determined that they carry ~10-20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. Analysis of ELM measurements imply that the ELM continuously loses energy as it travels across the SOL-larger gaps should reduce surface loads. The predicted divertor power loads for ITER disruptions has been reduced as a result of finding that the divertor footprint broadens during the thermal quench and that the plasma can lose up to 80% of its thermal energy before the thermal quench (not true for VDEs or ITBs). On the other hand predictions of power loading to surfaces outside the divertor have increased. Disruption mitigation through massive gas puffing has been successful at reducing divertor heat loads but estimates of the effect on the main chamber walls indicate 10s of kG of Be could be melted/mitigation. Estimates of ITER tritium retention have reduced the amount retained/discharge although the uncertainties are large and tritium cleanup may be necessary every few days to weeks. Long-pulse studies have shown that the fraction of injected gas that can be recovered after a discharge decreases with discharge length. The retention rate on the sides of tiles appears to ~ 1-3% of the ion flux to the front surface for C tiles and ~100x less for Mo tiles. T removal techniques are being developed based on surface heating and surface ablation although ITER mixed materials will make T removal more difficult. The use of mixed materials gives rise to a number of potential processes-e.g. reduction of surface melting temperatures (formation of alloys) and reduction of chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Behavior and characteristics of tungsten under impinging high heat-fluxes are investigated in view of the material choices for future devices such as ITER and DEMO. Experiments have been performed in the edge of the TEXTOR tokamak to study melt-layer motion, macroscopic melt layer erosion as well as the changes of the material properties. The parallel heat-flux ranges around q ∼ 45M W/m 2 allowing samples at an impact angle of 35 • to be exposed to 20 − 30M W/m 2 . Melt-layer motion perpendicular to the magnetic field is observed following a Lorentzforce originating from thermoelectric emission of the hot sample. Up to 3 g of tungsten are redistributed forming mountain like structures at the edge of the sample. The typical melt layer thickness is 1 − 1.5mm. Those hills are particularly susceptible to even higher heat-fluxes of up to the full q . Locally the temperature can reach up to 6000K, high levels of evaporation are causing significant erosion in form of continuous fine-spray (∼ 1 · 10 24 atoms m −2 s −1 ). Vaporshielding is occurring and hindering the further heating of the samples. In addition the formation of ligaments and splashes occurs several times during the melt phase ejecting droplets in the order of several 10µm up to 100µm probably caused by a Kelvin-Helmholtz instability evolving in the melt. In terms of material degradation several aspects are considered: formation of leading edges by redistributed melt, bubble formation and re-crystallization. Bubbles are occurring in sizes between µm and 200 µm while recrystallization increases the grain size up to 1.5 mm. The power handling capabilities are thus severely degraded. Melting of Tungsten in future devices is highly unfavorable and needs to be avoided especially in light of uncontrolled transients and possible unshaped PFCs

Recent experiments on JET have shown that type-I edge localized modes (ELMs) can be controlled by applying static low n = 1 external magnetic perturbation fields (EMPFs) produced by four external error field correction coils (EFCC) mounted far away from the plasma between the transformer limbs. When an n = 1 EMPF with an amplitude of a few Gauss at the plasma edge (ρ > 0.95) is applied during the stationary phase of a type-I ELMy H-mode plasma, the ELM frequency rises from ~30 Hz up to ~120 Hz and follows the applied perturbation field strength. The energy loss per ELM normalised to the total stored energy, ΔW ELM /W p , decreased from 7 % to below the resolution limit of the diamagnetic measurement (~ 2%). Transport analysis using the TRANSP code shows no or a modest reduction of the thermal energy confinement time because of the density pump-out, but when normalised to the IPB98(y,2) scaling the confinement shows almost no reduction. Stability analysis of mitigated ELMs shows that the operational point moves from intermediate n peeling-ballooning (wide mode) boundary to low-n peeling (narrow mode) boundary with n = 1 perturbation fields. The first results of ELM mitigation with the n = 2 EMPFs on JET demonstrate that the frequency of ELM can be increased by a factor of 3.5, only limited by the available EFCC coil current. During the application of the n = 1, 2 EMPFs, a reduction in the ELM size (ΔW ELM) and ELM peak heat fluxes on the divertor target by roughly the same factor as the increase of the ELM frequency has been observed. The reduction in heat flux is mainly due to the drop of particle flux rather than the change of the electron temperature. Similar plasma braking effect has been observed with n = 1 and n = 2 external fields when a same EFCC coil current was applied. Compensation of the density pump-out effect has been achieved by means of gas fuelling in low triangularity plasmas. An optimised fuelling rate to compensate the density pump-out effect has been identified. Active ELM control by externally applied fields offers an attractive method for next-generation tokamaks, e.g. ITER.

The operational domain for active control of type-I edge localized modes (ELMs) with an n = 1 external magnetic perturbation field induced by the ex-vessel error field correction coils on JET has been developed towards more ITER-relevant regimes with high plasma triangularity, up to 0.45, high normalized beta, up to 3.0, plasma current up to 2.0 MA and q 95 varied between 3.0 and 4.8. The results of ELM mitigation in high triangularity plasmas show that the frequency of type-I ELMs increased by a factor of 4 during the application of the n = 1 fields, while the energy loss per ELM, W/W , decreased from 6% to below the noise level of the diamagnetic measurement (<2%). No reduction of confinement quality (H 98Y ) during the ELM mitigation phase has been observed. The minimum n = 1 perturbation field amplitude above which the ELMs were mitigated increased with a lower q 95 but always remained below the n = 1 locked mode threshold. The first results of ELM mitigation with n = 2 magnetic perturbations on JET demonstrate that the frequency of ELMs increased from

The linear plasma device PSI-2 serves as a pilot experiment for the development of components, operational regimes and control systems for the linear plasma device JULE-PSI, which will be located in the nuclear environment allowing studies of radioactive and toxic samples. PSI-2 is also used for fusion reactor relevant plasma-material interaction studies. This contribution describes the PSI-2 layout and parameters and summarizes the recent scientific and technical progress in the project, including the installation of a target station for the sample manipulation and analyses.