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Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET–ITER-like wall (ILW) during two campaigns in 2011–2012 (first campaign, ILW-1) and 2015–2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470°C to 520°C, which was higher than measured after ILW-1: 370°C to 430°C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns.

A number of 35 samples cored from JET divertor tiles exposed to JET plasmas in ILW1 (ITER Like Wall – first campaign), ILW2, ILW1 + ILW2 and ILW3 campaigns were analyzed using the Glow Discharge Optical Emission Spectrometry (GDOES) technique. The depth profiles of the elements contained in the surface layer of the tiles (W, Mo, Be, C, O and D) have been quantitatively determined up to a depth of 30–60 μm. The upper zone of Tile 3 (the lower vertical zone of the inner divertor) seems to be an erosion zone as it was demonstrated with a Mo marker layer in the ILW1 campaign. By increasing the input energy in the following campaigns this erosion did not increase, but it almost disappeared, probably due to significant amounts of Be that were transported from the main chamber and deposited on that area. The Be concentration increased in the upper zone of Tile3 in ILW2 and ILW3 by a factor of about 5 in comparison with the ILW1 campaign. It looks like the Be layer protects the tile surface from erosion. Depending on the amount of Be deposited on specific areas of the W-coated CFC (Carbon Fiber Composite) tiles, two mechanisms of D retention were identified. If the peak concentration of Be is less than ∼4.5 at.% D is mainly trapped into the defects of the W/Mo structure including W/Mo and Mo/C interfaces. When the peak concentration of Be exceeds ∼4.5 at.%, D is mainly retained in the Be deposits. The GDOES results were correlated with SEM analyses performed on the GDOES crater flanks and with the TDS (Thermal Desorption Spectroscopy), results obtained with the same samples after GDOES analyses. They are in good agreement with the corresponding results obtained by other authors using NRA (Nuclear Reaction Analysis), SIMS (Secondary Ion Mass Spectrometry) and TDS.

Carbon impurity transport and deposition were investigated in the Wendelstein 7-X stellarator by injecting isotopically labelled methane ((CH4)-C-13) into the edge plasma during the last plasma operations of its Operational Phase (OP) 1.2B experimental campaign. C-13 deposition was measured by secondary ion mass spectrometry (SIMS) on three upper divertor tiles located on the opposite side of the vessel to the(13)CH(4) inlet. The highest C-13 inventories were found as stripe-like patterns on both sides of the different strike lines. These high deposition areas were also analysed for their impurity contents and the depth profiles of the main elements in the layers. Layered deposition of different impurity elements such as Cr, Ni, Mo and B was found to reflect various events such as high metallic impurities during the OP1.2A and three boronizations carried out during OP1.2B.

Evaluation of hydrogen isotopes retention in the plasma-facing components (PFC)l is required for the design of a future fusion reactor. To obtain a set of relevant data on the isotope retention and their surface state (and chemistry), the analysis of PFC from the JET tokamak with the ITER-Like Wall (JET-ILW) is both useful and important. In previous studies, the correlation of surface chemical states and hydrogen isotope retention in the divertor samples from the first ILW campaign (ILW-1, 2011-2012, 19 h plasma operation, total energy 150 GJ) was determined. It was found that the major D desorption peaks were found at 643 K and 803 K. For samples at the strike point region, the D desorption temperature was clearly shifted towards higher T, indicating the stabilization of D trapping by higher heat load. In this work, the studied tiles were exposed during the 3rd campaign (ILW-3) which comprised in total 23 h of plasma with the total energy input of 245 GJ. The hydrogen isotope retention and chemical states of divertor tiles were studied by thermal desorption spectroscopy (TDS) and X-ray photoelectron spectroscopy (XPS). It was found that the major D desorption were located at 620 K and 800 K, which were consistent with the data for ILW-1. However, the total D retention was increased. The XPS analysis showed that major elements on the surface was Be with forming oxide state with co-existence of W. Carbon was found mainly on the outer divertor tiles. In the presentation, detailed XPS and TDS results will be shown and discuss the hydrogen isotope retention behavior in JET-ILW divertor tiles., 12 th International Conference on Tritium Science & Technology

A mistake regretfully appears in Section 3.2 of the paper in the Fig. 4a and b. The authors make a sincere apology for the mistake and would like to apologise for any inconvenience caused. During preparation of the plot for the publication a mistake had occurred in the y-axis unit. This has been attributed to a typo error in the y-axis value column used in the preparation of the plot. Therefore, the y-axis value is in wrong decade, it should appear in 1014instead of 1015. The values from the plot were not used in any calculations presented in the paper therefore the change does not affect any conclusions of the paper. Fig. 4(a) and (b) should be as follows: [Figure presented]

First wall components in controlled fusion devices undergo severe modification by various physical and chemical processes arising from plasma-wall interactions: material erosion, its transport in the plasma and re-deposition. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks. The importance of surface analysis in studies of fuel inventory and material migration is presented. Experimental procedures and analysis methods are briefly reviewed with emphasis on ion beam techniques which play a prominent role in studies of wall components exposed to hot plasmas. Practical aspects in the analytical approach are addressed and special instrumentation used in these studies is described. QC 20101117

ITER-Like Wall Project has been launched at JET in order to perform a fully integrated test of plasma-facing materials. During the next major shutdown a full metal wall will be installed: tungsten in the divertor and beryllium in the main chamber. Beryllium erosion is one of key issues to be addressed. Special marker tiles have been designed for this purpose. Test coupons of such markers have been manufactured and examined. The performance test under high power deposition was carried in the electron beam facility JUDITH. The results of material characterization before and after high heat flux loads are presented. The samples survived, without macroscopic damage, power loads of up to 4.5 MW/m2 for 10 s (surface temperature ∼650 °C) and 50 cyclic loads at 3.5 MW/m2 lasting 10 s each (surface temperature ∼600 °C).

The retention of tritium (T) by carbon based deposits on tokamak surfaces is of increasing concern to the fusion community as the scale of tritium retention by this mechanism could be a limiting factor for the operation of fusion reactors, such as ITER. Hence there is a need to investigate ways of mitigating T retention and also for detritiating surfaces by either desorption of T or removal of tritiated deposits. The results of the removal of codeposits from CFC tiles by pulsed laser ablation are reported here. The results show that it is possible to completely remove a 300μm thick hydrogen isotope rich carbon film at a rate of 12x10-3m2/hr by this method and that with optimisation of the laser parameters there is scope to improve the treatment rates to provide a useful tool for managing T inventory in tokamaks.

The JET Fusion Technology (FT) work programme was launched in 2000, in the frame of the European Fusion Development Agreement, to address issues related to JET and ITER. In particular, there are four topics related to tritium being investigated. Based on the experience gained on the existing tokamaks, first calculations indicate that in-vessel tritium retention could represent a burden for ITER operation. Therefore erosion/deposition studies are being performed in order to better understand the layer co-deposition and tritium retention processes in tokamaks. Moreover, testing of in-situ detritiation processes, in particular laser and flash lamp treatments, should assess detritiation techniques for in-vessel components in the ITER-relevant JET configuration. To reduce the constraints on waste disposal, dedicated procedures are being developed for detritiation of metals, graphite, carbon-fibre composites, process and housekeeping waste. During the operational and decommissioning phases of a fusion reactor, many processes will produce tritiated water. Key components for an ITER relevant water detritiation facility are being studied experimentally with the aim of producing a complete design that could be implemented and tested at JET. This paper describes these topics of the FT-programme, the strategy developed and the results obtained so far.

Estimates of erosion, deposition and H-isotope retention in JET from previous divertor campaigns have relied on analysis of in-vessel components removed at shutdowns. The components analysed have also provided an incomplete coverage of the vessel. In 2004, new diagnostics are being installed to give a more complete picture (such as smart tiles) and to provide some time resolution. The latter includes further quartz microbalances (QMB), following the successful operation of a prototype in 2002-2004 [H.-G. Esser, G. Neill, P. Coad, G.F. Matthews, D. Jolovic, D. Wilson, M. Freisinger, V. Philipps, Quartz microbalance: a time-resolved diagnostic to measure material deposition in JET, Fusion Eng. Des. 66-68 (2003) 855-860; H.-G. Esser, V. Philipps, M. Freisinger, G.F. Matthews, J.P. Coad, G.F. Neill, JET EFDA Contributors, Effect of plasma configuration on carbon migration measured in the inner divertor of JET using quartz microbalance, J. Nucl. Mater. 337-339 (2005) 84-87], which will also have temperature control. Other diagnostics include rotating collectors and deposition monitors [M. Mayer, V. Rohde, P. Coad, P. Wienhold, ASDEX Upgrade Team, JET EFDA Contributors, Carbon erosion and migration in fusion devices, Phys. Scr. T111 (2004) 55-59]. Units are also being installed to provide information on mirrors for ITER. QC 20101122

The Joint European Torus (JET) fusion machine is the only device capable of operation with tritium and Be, which makes it best suited to study ITER relevant issues. A large variety of activities are performed within the JET fusion technology task force. In this paper, some topics, such as erosion/deposition and material transport, flakes characterization and detritiation techniques, are highlighted. Recent results obtained using a pumping cryopanel and on plasma facing component (PFC) characterization, are given. Finally, issues that will be addressed in the forthcoming JET work-programme are presented, such as a beryllium main wall for JET and in situ laser detritiation.

Elemental mapping of cross sections of deposited layers on inboard tiles in the JET divertor after exposure to plasma operations with carbon wall are presented. The study was made using microbeam ion beam analysis methods in combination with optical microscopy and SEM. The surfaces had been exposed to plasma through different periods of operation (1998–2007, 2007–2009 and 1998–2009). The texture and composition of the layers are non-uniform. The physical structures include columnar, lamellar and disordered globular appearances. The distribution of trapped deuterium was frequently found to be lamellar, with well-defined sub layers with higher deuterium concentration. However, 3D regions with dimensions of about 100 μm with enhanced deuterium content were also found, both at the layer surfaces and in the layer cross sections. The distributions of beryllium and Inconel components were lamellar but did not otherwise show large non-uniformity on the same scale length as the deuterium.