[en] A major issue for future fusion power-generating plants is the retention of tritium within the machine, since this reduces the efficiency of the recycling process (leading to costs for extra tritium) and introduces additional safety considerations. The retention seen in existing machines is primarily associated with the re-deposition of carbon eroded from contact points between the plasma and the surrounding walls. In JET after operation with deuterium-tritium mixtures, the great majority of the tritium trapped in the vessel was in carbon-based deposits that flaked off and fell to the bottom of the vessel, where access is difficult. Future machines will also have an allowable limit for the amount of dust, and in JET the break-up of these flakes represents a possible source. The issues of tritium retention and associated dust in a Next-Step device are being addressed in several ways. Firstly, the use of carbon is being minimized. Carbon is the only material that is tolerant of high surface power loads resulting from off-normal events, so it is not yet clear if the material of the critical contact are may be substituted. Secondly, the deposition mechanisms are being studied, with the aim of minimising the quantity of deposits and, within them, the fraction of tritium retained. Thirdly, ways are being developed of removing any deposits that do form within the device, ideally in such a way that the tritium can be returned (as tritium gas) to the storage beds. This is coupled with work on the more general issues of de-tri t tritiation, such as methods of reducing the tritium content of carbon titles, and reducing medium-level waste to low-level waste for subsequent disposal. (Author)

[en] Rotating collectors (RC) were installed in JET during the period 2005-2007, each providing a time-resolved deposition pattern on the surface of a rotating silicon disc, which could be analysed once retrieved from the vessel. This paper reports results from the silicon disc removed from the RC located under the load-bearing septum replacement plate in JET in 2007. Nuclear reaction analysis results of the deposits on the disc have been correlated directly with the pattern of erosion and deposition observed by the quartz microbalance (QMB) located in an equivalent position. The thickest film in the time-resolved region (i.e. deposited in ∼60 pulses) was ∼250 nm, and the Be/C ratio was generally found to be 0.1 or lower, with two regions where the ratio rose to 0.2. The deposition observed with the QMB appears to be about a factor of four less.

[en] A complete global balance for material transport in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust in the divertor region. Following the end of the first JET ITER-like wall campaign a set of tiles has been removed from the main chamber and the divertor. This paper describes the initial tile surface profiling results for evaluating the erosion in the main chamber and deposition in the divertor. Tile profiling was performed on upper dump plate tiles and inner wall guard limiters made of beryllium and on inner divertor tiles made of tungsten coated carbon (C)-fibre composites. Additionally, the mass of dust collected from the JET divertor is also reported. Present results are compared with JET-C campaign results with the all-carbon C wall. (paper)

[en] The impact of edge localized modes (ELMs) carrying energies of up to 450 kJ on carbon erosion in the JET inner divertor is assessed by means of time resolved measurements using an in situ quartz microbalance diagnostic. The inner target erosion is strongly nonlinearly dependent on the ELM energy: a single 400 kJ ELM produces the same carbon erosion as ten 150 kJ events. The ELM-induced enhanced erosion is attributed to the presence of codeposited carbon-deuterium layers on the inner divertor target, which are thermally decomposed under the impact of ELMs

[en] The Fusion Technology Task Force (TFFT) has a wide-ranging series of programmes in the areas of waste management and safety, tritium recovery, tritium analysis and accounting, and testing components under development for ITER at JET. Examples have been presented here in the fields of waste management and safety. In waste management, the largest effort is currently on water de-tritiation, which is considered to be the most urgent and important topic affecting JET operations, and plant design is also required for ITER. It is also the most technically challenging of the waste detritiation issues. A complete design for a water de-tritiation plant for JET (a prototype for ITER), including optimised and tested catalysts, is expected within the next 2 years. TFFT safety programmes support the on-going work on safety in preparation for ITER, including tritium spreading and dust inhalation effects for worse-case accident scenarios. Effort is also going into documenting the operational experience of the JET machine with respect to reliability of mechanical components within the tritium boundary and radiation exposure, and inferring what lessons should be learnt for ITER

[en] 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 co-deposits from CFC tiles by pulsed laser ablation are reported here. The results show that it is possible to completely remove a 300 μm thick hydrogen isotope rich carbon film at a rate of 12*10^-3 m²/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. (authors)

[en] This contribution presents two-dimensional Monte-Carlo simulations of the local transport of hydrocarbons that are chemically eroded in the JET MkIIa divertor. The effect of a given background carbon flux flowing from the main plasma down to the divertor is also taken into account. The influence of local plasma temperatures and densities and the influence of different assumptions for the sticking of re-deposited particles hitting the tiles is analysed. Even under the assumption that the sticking of re-deposited hydrocarbon fragments is zero, a large amount (about 75%) of the eroded particles is re-deposited in the form of ionized carbon on the tiles. A reasonable agreement between the simulation and the experimental observation of the carbon deposition at the inner louvers in the MkIIa JET divertor is achieved, if high chemical erosion yields of about 20% together with a negligible effective sticking of hydrocarbons are assumed. Although there are experimental observations indicating such high erosion yields, such yields cannot be applied as a stationary value for the effective erosion since it would turn the inner divertor into a net erosion area in contrast to the experimental findings. The possibility of applying different erosion rates for (re)-deposited layers and the substrate material is discussed. To further improve the modelling a better knowledge about the formation and erosion of re-deposited layers is required. A first conclusion concerning the erosion and re-deposition behaviour in ITER will be drawn

[en] A build-up of co-deposits in remote areas of the divertor can contribute significantly to the overall fuel retention. The control of plasma-material interactions via the study and understanding of erosion-deposition of PFCs provides vital information for the efficient future operation of ITER. The major aim of this work is to reveal details of beryllium deposition and fuel (deuterium) retention on divertor plasma-facing componentsremoved from the JET ITER-Like Wall divertor after cumulative exposure during the first two (ILW−1 + 2) and all three (ILW−1 + 2 + 3) campaigns. Ion beam analysis techniques such as Rutherford backscattering spectrometry, nuclear reaction analysis and proton induced x-ray emission have been extensively used for post-mortem analyses of selected tiles from JET following each campaign and can provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles via implantation or co-deposition. The studied divertor tiles represent a unique set of samples, which have been exposed to plasmas since the beginning of the JET-ILW operation for three successive plasma campaigns. This is a comprehensive comparison of divertor components after these operation periods. The results presented summarise deposition and fuel retention on Tiles 4 (inner base) and 6 (outer base). Although the deposition pattern is similar to that determined after individual campaigns, D retention is not a cumulative process and is determined mainly by the last campaign, and the total Be deposit after the 3 campaigns (i.e. data 1 + 2 + 3 = tile exposed 2011–2016) is less than the sum of the deposits after each individual campaign (sum 1 + 2 + 3) for Tile 4 but greater for Tile 6. (topical issue article)

[en] In 2010, all the plasma-facing components were removed from JET so that the carbon-based surfaces could be replaced with beryllium (Be) or tungsten as part of the ITER-like wall (ILW) project. This gives unprecedented opportunities for post-mortem analyses of these plasma-facing surfaces; this paper reviews the data obtained so far and relates the information to studies of tiles removed during previous JET shutdowns. The general pattern of erosion/deposition at the JET divertor has been maintained, with deposition of impurities in the scrape-off layer (SOL) at the inner divertor and preferential removal of carbon and transport into the corner. However, the remaining films in the SOL contain very high Be/C ratios at the surface. The first measurements of erosion using a tile profiler have been completed, with up to 200 microns erosion being recorded at points on the inner wall guard limiters.

[en] The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing ¹³C-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project. (authors)