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As in many of today’s tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, . The initial choice of , which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with Measurements of in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar predicted values of mapped to the outside midplane. The engineering scaling with the highest statistical significance, , dependent on input power density, aspect ratio and elongation, yields   =  [7, 4, 5] cm for   =  [2.5, 5.0, 7.5] MA, the three reference limiter plasma currents specified in the ITER heat and nuclear load specifications. Mapped to the inboard midplane, the worst case (7.5 MA) corresponds to mm, thus consolidating the 50 mm width used to optimize the FW panel toroidal shape. [ABSTRACT FROM AUTHOR]

This paper summarizes the status of the COMPASS tokamak, its comprehensive diagnostic equipment and plasma scenarios as a baseline for the future studies. The former COMPASS-D tokamak was in operation at UKAEA Culham, UK in 1992–2002. Later, the device was transferred to the Institute of Plasma Physics of the Academy of Sciences of the Czech Republic (IPP AS CR), where it was installed during 2006–2011. Since 2012 the device has been in a full operation with Type-I and Type-III ELMy H-modes as a base scenario. This enables together with the ITER-like plasma shape and flexible NBI heating system (two injectors enabling co- or balanced injection) to perform ITER relevant studies in different parameter range to the other tokamaks (ASDEX-Upgrade, DIII-D, JET) and to contribute to the ITER scallings. In addition to the description of the device, current status and the main diagnostic equipment, the paper focuses on the characterization of the Ohmic as well as NBI-assisted H-modes. Moreover, Edge Localized Modes (ELMs) are categorized based on their frequency dependence on power density flowing across separatrix. The filamentary structure of ELMs is studied and the parallel heat flux in individual filaments is measured by probes on the outer mid-plane and in the divertor. The measurements are supported by observation of ELM and inter-ELM filaments by an ultra-fast camera. [ABSTRACT FROM AUTHOR]