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Tritium retention in the castellated structure of beryllium limiters used in JET with the ITER-like wall (ILW) during the first (ILW1), third (ILW3) and all three (ILW1-3) campaigns were examined and evaluated. Tritium was deposited on the surfaces inside the castellation grooves together with deuterium, beryllium, oxygen, carbon and small amounts of metallic impurities such as nickel, copper and tungsten. The tritium content after the ILW1 campaign was greater than after the ILW3 campaign. This is attributed to the steadily decreasing amount of carbon impurities in JET from campaign to campaign. The majority of tritium was retained in shallow regions in the grooves, up to 2 mm from the entrance to the gap. It was comparable on all sides of the castellation, i.e. no difference has been detected between the toroidal and poloidal gaps. Secondly, the tritium retention in the gaps was similar on all specimens independent of their position in the tokamak, while the retention on the plasma-facing surfaces clearly depended on the tile position. The tritium deposition patterns in the castellation were also compared with the deuterium distribution determined in earlier studies., QC 20230403

Plasma-facing components have castellated structure for thermo-mechanical durability and integrity under high heat flux loads. However, fuel co-deposition in the grooves of the castellation may enhance fuel retention. In KSTAR, castellated tungsten tiles were tested to investigate the impact of tile shaping and misalignment on the retention. The tiles with poloidal and toroidal gaps of 0.5 mm were exposed at the divertor during a whole campaign encompassing 4364 s of plasma operation. Surfaces inside the gaps were analysed by means of 3He-based micro-NRA, ERDA and PIXE. Modelling of carbon deposition was performed with the impurity transport code 3D-GAPS assuming impurity penetration along the magnetic field lines with plasma-wetted areas defined by simple geometrical shadowing. The main deposited element is carbon with different concentration at the entrance of the groove, dependent on the tile shaping: 6 x 10(17) cm(-2) for a chamfered and misaligned gap and up to 283 x 10(17) cm(-2) for a flat and aligned gap. The deposition patterns are exponentially decreased to 4-10 x 10(16) cm(-2) inside the gap. Deuterium concentration in the gaps described above ranges, respectively, from 2 x 10(17) cm(-2) to 50 x 10(17) cm(-2) at the top of the groove and decreases to 1-4 x 10(16) cm(-2) following the carbon deposition trends. The highest carbon and deuterium densities are measured at the plasma-exposed side of the flat tile and aligned gap. Modelled deposition profiles reproduce qualitatively the experimentally observed trends., QC 20200415

Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in the divertor. This has allowed for a large scale test of castellated plasma-facing components (PFC). Procedures for sectioning the limiters into single blocks of castellation have been developed. This facilitated morphology studies of morphology of surfaces inside the grooves for limiters after experimental campaigns 2011–2012 and 2013–2014. The deposition in the 0.4–0.5mm wide grooves of the castellation is ‘shallow’. It reaches 1–2mm into the 12mm deep gap. Deuterium concentrations are small (mostly below 1 × 1018 cm−2 ). The estimated total amount of deuterium in all the castellated limiters does not exceed the inventory of the plasma-facing surfaces (PFS) of the limiters. There are only traces of Ni, Cr and Fe deposited in the castellation gaps. The same applies to the carbon content. Also low deposition of D, Be and C has been measured on the sides of the bulk tungsten lamellae pieces. Modelling clearly reflects: (a) a sharp decrease in the measured deposition profiles and (b) an increase in deposition with the gap width. Both experimental and modelling data give a strong indication and information to ITER that narrow gaps in the castellated PFC are essential. X-ray diffraction on PFS has clearly shown two distinct composition patterns: Be with an admixture of Be–W intermetallic compounds (e.g. Be22W) in the deposition zone, whilst only pure Be has been detected in the erosion zone. The lack of compound formation in the erosion zone indicates that no distinct changes in the thermo-mechanical properties of the Be PFC might be expected.

Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in the divertor. This has allowed for a large scale test of castellated plasma-facing components (PFC). Procedures for sectioning the limiters into single blocks of castellation have been developed. This facilitated morphology studies of morphology of surfaces inside the grooves for limiters after experimental campaigns 2011-2012 and 2013-2014. The deposition in the 0.4-0.5 mm wide grooves of the castellation is 'shallow'. It reaches 1-2 mm into the 12 mm deep gap. Deuterium concentrations are small (mostly below 1 x 10(18) cm(-2)). The estimated total amount of deuterium in all the castellated limiters does not exceed the inventory of the plasma-facing surfaces (PFS) of the limiters. There are only traces of Ni, Cr and Fe deposited in the castellation gaps. The same applies to the carbon content. Also low deposition of D, Be and C has been measured on the sides of the bulk tungsten lamellae pieces. Modelling clearly reflects: (a) a sharp decrease in the measured deposition profiles and (b) an increase in deposition with the gap width. Both experimental and modelling data give a strong indication and information to ITER that narrow gaps in the castellated PFC are essential. X-ray diffraction on PFS has clearly shown two distinct composition patterns: Be with an admixture of Be-W intermetallic compounds (e. g. Be22W) in the deposition zone, whilst only pure Be has been detected in the erosion zone. The lack of compound formation in the erosion zone indicates that no distinct changes in the thermo-mechanical properties of the Be PFC might be expected., For complete list of authors see http://dx.doi.org/10.1088/1741-4326/aa6864

A castellated tungsten test limiter composed of detachable segments was exposed to plasma discharges in the TEXTOR to kamak 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 oil 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, clown the gap with the decay length of around 1.2-1.8 mm; (iii) the presence of copper droplets and tungsten oxide (WO(2)) has been identified in the gaps. Different pathways leading to the oxide formation are considered., 8th International Symposium on Fusion Nuclear Technology Location: Heidelberg, GERMANY Date: SEP 30-OCT 05, 2007

To examine the performance of a castellated structure under plasma loading, a hemispherical solid tungsten brush limiter was exposed to the plasma in the TEXTOR-94 tokamak. Due to the thermally isolated column of W segments, IR camera showed a non-uniform temperature distribution. The maximum incident power density was calculated to be about 35-40 MW/m(2). Concerning impurity generation, the structure did not show any particular effects. During plasma exposure, only some minor cracks developed in one of the columns, however, the crack propagation was interrupted by a groove. It can be concluded that the W brush limiter had comparable performance and superior mechanical behaviour compared to a solid W limiter. To study erosion and long-range transport of W atoms, a graphite limiter was exposed simultaneously with the brush limiter. As a result, the deposited W atoms via long-range transportation were estimated to be 10(15) cm(-2) shot(-1) at 46.5 cm from the plasma centre of TEXTOR., QC 20100525