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The far scrape-off layer transport is studied in ASDEX Upgrade H-mode discharges with high divertor neutral density N 0 , div , high power across the separatrix P sep and nitrogen seeding to control the divertor temperature. Such conditions are expected for ITER but usually not investigated in terms of turbulent SOL transport. At high N 0 , div and P sep the H-mode discharges enter a regime of high cross-field particle and power transport in the SOL which is accompanied by a significant change of the turbulence characteristic analogous to the transition from conductive to convective transport in L-mode. Parallel particle and power flux densities of several 10 23 m−2 s−1 and 10 MW m−2 have been detected about ∼40 to 45 mm outside the separatrix mapped to the outer mid-plane. The particle flux fall-off length reached up to 45 mm. This paper presents for the first time an empirical condition to enter the high transport regime in H-mode and the relation of this regime to changes in the filamentary transport.

The SOL power decay length ( λ q ) deduced from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade with carbon plasma facing components, are presented. Interpretation of the target heat load profiles is performed by using a 1D-fit function which disentangles the upstream λ q and an effective diffusion in the divertor (S), the latter essentially acting as a power spreading parameter in the divertor volume. It is shown that the so called integral decay length λ int is approximately given by λ int ≈ λ q + 1.64 × S . An empirical scaling reveals parametric dependency λ q / mm ≃ 0.9 · B T - 0.7 q cyl 1.2 P SOL 0 R geo 0 for type-I ELMy H-modes. Extrapolation to ITER gives λ q ≃ 1 mm. Recent measurements in JET-ILW and from ASDEX Upgrade full-W confirm the results. It is shown that a regression for the divertor power spreading parameter S is not yet possible due to the large effect of different divertor geometries of JET and ASDEX Upgrade Divertor-I and Divertor-IIb.

Multi-fluid plasma edge code packages, usually coupled to Monte-Carlo transport codes for neutrals are in wide use in the fusion community to simulate the SOL and divertor. A status of the current code validation effort is presented using experimental data of Ohmic and L-mode discharges from ASDEX Upgrade for He in forward toroidal field, BT, only and for D fueled in forward and reversed BT in lower single null configurations. While satisfactory agreement between simulated and experimental data can be obtained for both divertors as long as one of the two divertors is not strongly recombining, the qualitative behavior as a function of line averaged density cannot be recovered in the simulations for higher recycling regimes. This holds independently of field direction or fueling species. Semi-empirical models and thus far not included physics models are discussed.

In the present paper, inter-machine studies of Ion Cyclotron Wall Conditioning (ICWC) have been performed in limiter (TEXTOR) and divertor (ASDEX Upgrade, AUG) tokamaks in the presence of toroidal (0.2–2.35 T) and vertical (0–0.04 T) magnetic fields using the conventional ICRF antennas without modifications in hardware. The ICWC effect on both machines was studied by analyzing the removal rate of marker gases which have been loaded to the walls by glow discharge beforehand. Several factors were identified which could have a crucial impact on the conditioning efficiency: (i) RF power coupled to the plasmas; (ii) RF power absorption scheme; (iii) superimposing an additional vertical magnetic field on the toroidal field ( B V B T ). All the observed effects are analyzed in terms of RF plasma wave excitation/absorption and compared with the predictions from 1-D RF and 0-D transport codes. ICWC scenarios for ITER are proposed and analyzed.

A qualitative as well as quantitative evaluation of experimentally observed plasma parameters in the detached regime proves to be difficult for several tokamaks. A series of ohmic discharges have been performed in ASDEX Upgrade and DIII-D at similar as possible plasma parameters and at different line averaged densities, {bar n}{sub e}. The experimental data represent a set of well diagnosed discharges against which numerical simulations are compared. For the numerical modeling the fluid-code B2.5 coupled to the Monte Carlo neutrals transport code EIRENE is used. Only the combined enhancement of effects, such as geometry, drift terms, neutral conductance, increased radial transport and divertor target composition, explains a significant fraction of the experimentally observed asymmetries of the ion fluxes as a function of {bar n}{sub e} to the inner and outer target plates in ASDEX Upgrade. The relative importance of the mechanisms leading to detachment are different in DIII-D and ASDEX Upgrade.

Main chamber recycling has been quantified over a wide range (Ip = 0.6–1.1 MA, Bt = −3 to −1.9 T, ne = 0.3–1 × 1020 m−3, Ptot = 2.5–20 MW), both inter-ELM and during type-I ELMs. A vertical CO2 interferometer channel registers densities δneLV2 in excess of those expected from core/SOL parameters. δneLV2 is localized at the inner, lower divertor baffle in the main chamber. Its behavior follows the neutral fluxes measured by a nearby pressure gauge. Particle fluxes to the inner wall as well as densities exhibit parallel behavior. δneLV2 can be as high as 8 × 1019 m−2 and the ELM perturbation itself as large as 9 × 1019 m−2. ne at the inner wall reaches 2.5 × 1020 m−3 and fluxes ∼6 × 1023 s−1. For controlled conditions δneLV2 is linear with ne and increases with Ptot. A ne threshold increasing with Ip exists below which δneLV2 ∼ 0, corresponding to a Greenwald density

An experimental database of the power spreading into the target private flux region (S-factor) from ASDEX Upgrade and JET outer target data has been gathered and analysed. The power spreading, combined with the Scrape-Off Layer heat flux decay length λ q , determines the power exhaust channel width at the divertor target. In terms of main plasma parameters, we find that the poloidal magnetic field plays a major role S ∼ B pol - 1 in both L-mode and H-mode plasmas, similarly to λ q . Combining JET and AUG data in H-mode with open divertor we obtain a strong and beneficial major radius dependency and no plasma density influence is found. For closed AUG divII configurations in L-mode, however, larger upstream n e clearly broadens S . Interestingly, the scaling of S with divertor parameters, reveals that only target T e is needed to account for its variation. SOLPS simulations with fixed perpendicular transport reproduce most of the observed phenomenology, suggesting that the observed larger S in closed versus open divertor configuration is simply due to different target T e generated by larger parallel gradients.