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Abstract Damage and melting of ITER-like tungsten castellated blocks – made of tungsten top layer, copper inter-layer, copper alloy base (CuCrZr) – exposed to long pulse H-mode plasmas were investigated. Since the blocks were not actively cooled, the temperature of blocks was increased rapidly up to melting point of tungsten during H-mode plasmas resulting in severe damage such as cracks and melting of tungsten layer. It is observed that the copper inter-layer located at 5 mm below the surface of the tungsten was also melted indicating that the temperature of the interface had reached the melting point of copper, at least 1200 °C. Melted tungsten material was moved towards j × B direction (high filed side). The melting event was reproduced by 3D transient modeling using COMSOL with 2 MW/m2 heat load and explains the observation quantitatively well. This indicates that failure of cooling system during a long pulse operation with metal walls, especially melting of copper inter-layer will be critical. [ABSTRACT FROM AUTHOR]

Heat load on the misaligned leading edges of tungsten castellated blocks based on tungsten (W), oxygen-free high conductive copper (OFHC-Cu), and copper-chrome-zirconium (CuCrZr) alloy are studied by COMSOL analysis and infrared (IR) measurements in KSTAR. IR measurements show that 1–3 MW/m 2 of heat flux has been deposited on the blocks during the inter-ELM (edge localized mode) phase in H-mode plasmas. COMSOL analysis indicates that the temperature of 1 mm leading edge in KSTAR under 3 MW/m 2 would reach the recrystallization temperature within 2 s and will be melted within 30 s during a long pulse H-mode shot. Rounded and double chamfered blocks show much better thermal response meaning that shaping of divertor block enhances the heat load handling capability. It seems that a simple COMSOL analysis describes heat load patterns on the tungsten blocks of different shapes qualitatively well. Therefore, simple analysis would be useful to make a quick prediction on heat load patterns of blocks with arbitrary shapes. [ABSTRACT FROM AUTHOR]

In this paper, we report results of a dedicated experiment that gives the plasma penetration profiles inside a gap of a tokamak castellated plasma-facing component. A specially designed probe that recreates a gap between two tiles has been built for the purpose of this study. It allows to measure ion saturation profiles along the 2 sides and at the bottom of the gap for both poloidal and toroidal orientations. The novelty of such experiment is the real time measurement of the plasma flux inside the gap during a tokamak D-shaped discharge compared to previous experimental studies which were mainly post-mortem. This experiment was performed in the COMPASS tokamak and results are compared with particle-in-cell simulations. The plasma deposition is found to be asymmetric in both orientations with a stronger effect in poloidal gaps. The Larmor radius of the incoming ions plays a role in the plasma penetration only in poloidal gaps but seems to have little impact in toroidal gaps. Profiles are qualitatively well reproduced by simulations. Ion current is recorded at the bottom of a toroidal gap under certain conditions. [ABSTRACT FROM AUTHOR]

Abstract: A novel timber -joist with a castellated web is presented. While castellated webs are common in steel joists, this concept is new to timber joists. The openings provided by the castellation process provide for the easy passage of services during construction or subsequent remodelling of a building. The manufacturing process is described together with details of tests on twelve 241 mm and eleven 305 mm deep joists. The load–deflection response, failure loads and modes of failure are given. The mean failure moment and shear force for the 241 mm deep joists were found to be 8.2 kN m and 6.0 kN, respectively. The corresponding values for the 305 mm deep joists were 12.5 kN m and 6.9 kN. The predominant failure modes were tension failure in the web at the corner of the openings and shear failure in the web at mid-depth between the openings. The effect of the web openings on the joist stiffness is not significant but the shear capacity is reduced. The reduction in shear capacity compares favourably with the shear capacity of commercial joists of equivalent sizes with circular or rectangular openings. [Copyright &y& Elsevier]

Abstract: A castellated tungsten test limiter composed of detachable segments was exposed to plasma discharges in the TEXTOR tokamak operated with graphite main limiters. Dismantling of the limiter enabled the analysis of surfaces located inside the castellation. The emphasis was on the determination of: (i) deposition and fuel retention; (ii) material mixing and new compound formation on plasma-facing surfaces and in the grooves of castellation. The investigation performed by means of accelerator-based ion beam analysis methods, microscopy and X-ray diffraction has brought several essential results: (i) deuterium retention on plasma-facing surfaces and in the castellation of metal PFC is strongly related to the co-deposition with carbon; (ii) both carbon and deuterium are detected only in narrow belts, a few millimetre broad, down the gap with the decay length of around 1.2–1.8mm; (iii) the presence of copper droplets and tungsten oxide (WO2) has been identified in the gaps. Different pathways leading to the oxide formation are considered. [Copyright &y& Elsevier]

Abstract: The effect of discrete structures such as macrobrush or castellated surfaces on power handling and deuterium retention of plasma facing components is to be assessed since such geometrical configurations are needed for increasing the lifetime of the armour to heat-sink joint. Four small macrobrush W and W+1%La2O3 samples have been exposed in the Frascati Tokamak Upgrade (FTU) scrape-off layer up to the last closed flux surface by means of the Sample Introduction System. FTU is an all metal machine with no carbon source inside vacuum vessel; it exhibits ITER relevant energy and particle fluxes on the plasma facing components. Here, results on morphological surface changes (SEM), chemical composition (EDX) and deuterium retention (TDS) are reported. [Copyright &y& Elsevier]

Abstract: To insure the thermo-mechanical durability of ITER it is planned to manufacture the castellated armour of the divertor i.e. to split the armour into cells [W. Daener et al., Fusion Eng. Des. 61&62 (2002) 61]. This will cause an increase of the surface area and may lead to carbon deposition and tritium accumulation in the gaps in between cells. To investigate the processes of deposition and fuel accumulation in gaps, a castellated test-limiter was exposed to the SOL plasma of TEXTOR. The geometry of castellation used was the same as proposed for the vertical divertor target in ITER [W. Daener et al., Fusion Eng. Des. 61&62 (2002) 61]. After exposure the limiter was investigated with various surface diagnostic techniques. Deposited layers containing carbon, hydrogen, deuterium and boron were found both on top plasma-facing surfaces and in the gaps. The amount of deuterium in the gaps was at least 30% of that found on the top surfaces. [Copyright &y& Elsevier]