The recycling of D ions impinging onto a W divertor surface is a key input parameter into the power and momentum balance at the target boundary during SOL modeling. It is described by the ratio R of the flux of recombining D 2 molecules to the incident ion flux. In steady-state plasmas where the surface is in equilibrium with the incident flux R equals one due to particle conservation. However during transient events such as ELMs the evolution of R with time is not straight forward to predict. Therefore detailed diffusiontrapping calculations were performed taking into account the variations in power influx and particle energy during an ELM. They showed that in contrast to the naive expectation, that the ELM would deplete the surface and subsequently lead to "pumping" (R ≪ 1) of the incident flux by the empty surface, R ≈ 1 or even R > 1 occurs. This paper will first describe how the ELM was approximated in the 1D diffusion-trapping code and will then discuss the evolution of R during an ELM and in the inter ELM phase. Also an analytical picture of R will be developed which allows to qualitatively understand the evolution of R as calculated by the diffusion-trapping code.

Asymmetrical disruptions may occur during ITER operation and they may be accompanied by large sideways forces and rotation of the asymmetry. This is of particular concern because resonance of the rotating asymmetry with the natural frequencies of the vacuum vessel (and other in-vessel components) could lead to large dynamic amplification of the forces. A significant fraction of non-mitigated JET disruptions have toroidally asymmetric currents that flow partially inside the plasma and partially inside the surrounding vacuum vessel ("wall"). The toroidal asymmetries (otherwise known as appearance of 3-D structures) are clearly visible in the plasma current (I p) and the first plasma current moments. For the first time we present here the asymmetries in toroidal flux measured by the diamagnetic loops and also propose a physical interpretation. The presented data covers the period of JET operation with C-wall (JET-C from 2005 until late 2009) and with ITER-like wall (JET-ILW from 2011 until late 2014), during which pickup coil and saddle loop data at four toroidally orthogonal

Plasma Phys. Control. Fusion 58 (2016) 074005 (11pp) q imp mm, thus consolidating the 50 mm width used to optimize the FW panel toroidal shape.

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 NBIassisted 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.