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Highlights • Operations using four carbon divertor configurations in JET are described. • Carbon with Be, Ni, etc eroded in the main chamber deposit at the inner divertor. • Carbon is preferentially eroded and migrates to the shadowed divertor corners. • Deuterium (and tritium from D-T fuelling) were co-deposited with the carbon. Abstract 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. [ABSTRACT FROM AUTHOR]

Abstract Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points: (a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1–1.4 g per campaign (19.1–23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JET-C). (b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2–10 μm in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters. (c) Tungsten dust occurs mainly as partly molten flakes originating from the W-coated tiles. [ABSTRACT FROM AUTHOR]

• First-ever long-term direct exposure of Ti in the tokamak environment. • Identification of different stages of Ti degradation up to complete disintegration. • Passive oxide layer on titanium can protect it from transformation into Ti hydride. • Titanium manufacturing process may influence its behaviour in a fusion reactor. Titanium activation foils for neutron studies were exposed inside the JET tokamak vessel during the entire 15 months long experimental campaign with deuterium fuelling. Afterwards, it was found that the structure of some foils was only slightly affected, while others were totally or partly disintegrated into dust: flakes and powder. The analyses indicated that defects produced in materials by cutting during the sample preparation eventually lead to the degradation of the titanium crystal structure upon the exposure in JET to neutral particles including hydrogen. In some cases, a passive oxide layer on the foil surface effectively prevented the access of hydrogen, and - by this - protected the metal against partial or complete transformation into the titanium hydride powder. [ABSTRACT FROM AUTHOR]

Abstract: An overview of several techniques considered for fuel and co-deposits removal is given. The methods were tested both on plasma-facing components from the TEXTOR tokamak and on laboratory-prepared layers: (a) chemical approach based on oxidative or nitrogen-assisted plasma; (b) photonic methods with laser-induced fuel desorption or ablation of co-deposits; (c) thermal desorption in vacuum or under oxidative conditions at a broad range of temperatures. The emphasis is on outstanding issues associated with every technique aiming at the reduction of fuel content: the efficiency of fuel and co-deposit removal, the surface state of PFC following the treatment and dust generation. [Copyright &y& Elsevier]

Abstract: 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 4.5MW/m2 for 10s (surface temperature ∼650°C) and 50 cyclic loads at 3.5MW/m2 lasting 10s each (surface temperature ∼600°C). [Copyright &y& Elsevier]

Abstract: The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. In order to perform a fully integrated material test JET has launched the ITER-like Wall Project with the aim of installing a full metal wall during the next major shutdown. As a result of R&D projects in 2005–2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies. [Copyright &y& Elsevier]

In the framework of the ITER Physics Department of the EUROfusion Consortium, JET mission is focused on preparing for the next step devices by developing high performance scenarios and testing reactor relevant techniques and technologies. In terms of scenario development, a complete scan in isotopic composition has already started and, according to the present schedule, should include full T operation and culminate in a 50/50 DT campaign by 2020. In DD, significant increases in input power and various adjustments in fueling have allowed reaching the IPBp8(y,2) scaling up to 3 MA. A complete set of diagnostic upgrades is being implemented to support operation and to guarantee adequate scientific exploitation of the experiments. With regard to the technology, the main enhancements and specific tests are focused on maximizing the returns from the unprecedented 14 MeV neutron field and the tritium fuel cycle. [ABSTRACT FROM AUTHOR]

Abstract Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified ("geological-like") mixed-material deposition layer which mainly included Be and Ni with the thickness of ∼2 μm. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper. [ABSTRACT FROM AUTHOR]

The ITER first mirrors are the components of optical diagnostic systems closest to the plasma. Deposition may build up on the surfaces of the mirror affecting their ability to fulfil their function. However, physics modelling of this layer growth is fraught with uncertainty. A new experiment is underway on JET, under contract to ITER, with primary objective to test if, under realistic plasma and wall material conditions and with ITER-like first mirror aperture geometry, deposits do grow on first mirrors. This paper describes the engineering design and analysis of this mirror test assembly. The assembly was installed in the 2014–15 shutdown and will be removed in the 2016–17 shutdown. [ABSTRACT FROM AUTHOR]

Ion cyclotron wall conditioning (ICWC) is being developed for ITER and W7-X as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the currentless conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-Juelich, Germany) proposes to explore several key aspects of ICWC. For this purpose we have designed an ICRF system made of a single strap antenna within a metallic box, connected to a feeding port and a pre-matching system. We discuss the design work of the antenna system with the help of the commercial electromagnetic software CST Microwave Studio ® . The simulation results for a given geometry provide input impedance matrices for the two-port system. These matrices are afterwards inserted into various circuit models to assess the accessibility of the required frequency range. The sensitivity of the matching system to uncertainties on plasma loading and capacitance values is notably addressed. With a choice of three variable capacitors we show that the system can cope with such uncertainties. We also demonstrate that the system can cope as well with the high reflected power levels during the short breakdown phase of the RF discharge, but at the cost of a significantly reduced coupled power. [ABSTRACT FROM AUTHOR]

Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was ∼1.5 μm. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation. [ABSTRACT FROM AUTHOR]

• Deuterium retention patterns and other impurities have been measured on 1/8 of the COMPASS divertor tiles with ion beam analysis in all spatial directions, including depth. • Using 2D spatial analysis, the deuterium retention pattern appears to be toroidally symmetric; symmetry breaks down when considering also depth information. • Deuterium retention is correlated to different light and heavy impurities, showing different correlations between elements at different depths. • The radial deuterium retention pattern is correlated to divertor probe signals in order to obtain a relationship between radial probe signal patterns and radial deuterium retention pattern. • Simple regression analysis shows that, under certain assumptions, deuterium retention can be predicted by probe signals, given that fit function parameters have been estimated by calibration measurements. • New contributions to the field. • 1/8 of the COMPASS divertor has been inspected for impurity and deuterium retention. • The ion beam analysis takes all three spatial dimensions into account to reveal subtle differences in deuterium co-deposition. • Extensive statistical studies have been carried out to reveal correlations between deuterium and impurity deposition, and between deuterium deposition and divertor probe signals. • This paper outlines how an empirical formula for determining fuel retention in future fusion devices can be estimated based on fuel retention calibration measurements and basic divertor probe signals like electron temperature, ion saturation current and parallel heat flux. Divertor tiles from the COMPASS tokamak have been examined with ion beam analysis for the determination of deuterium retention and co-deposits. A novel approach was used to draw 2D deuterium retention maps at different depths and correlated to the concentrations of other elements: boron, carbon, oxygen and metals. This approach, which employs all three spatial dimensions, revealed that the deuterium retention pattern is not toroidally symmetric at different depths while the overall deuterium pattern is toroidally symmetric. Analysis of visible-light camera records disclosed that parts of the divertor were not wetted by plasma. The radial profile of deuterium in both wetted and non-wetted ("shadowed") divertor regions was compared to strike point positions, angle of incidence, electron temperature T e , parallel heat flux Q par and ion saturation current I sat in order to find the influence of these parameters on deuterium retention. There are certain series of consecutive shots for which correlations between deuterium retention and separate plasma parameters (T e , I sat , Q par) are high. Combining the three parameters in a linear model or a power law model enabled reconstruction of the radial deuterium retention pattern. After fitting the model to one third of the data points, the deuterium retention for shadowed tiles was reproduced at the other two thirds of the data points within 20% deviation. This study shows the need for detailed tile analysis in three dimensions, the great benefits of broad statistical analysis in solving challenges in fusion, and motivates the investigation of COMPASS tiles after dismantling. [ABSTRACT FROM AUTHOR]

Highlights: [•] Be-contaminated JET mirrors have been laser-cleaned in vacuum. [•] The removal of the mixed-material layers was only successful for thin ones. [•] Laboratory-prepared Be layers were removed successfully from metal mirrors. [•] No preference for a particular wavelength range is apparent. [Copyright &y& Elsevier]

Abstract: Our studies were aimed to determine the damage threshold of molybdenum (Mo) and stainless steel (SS) mirrors to provide the maximum fluence which the mirror surfaces could withstand without affecting their reflectivity properties. A high repetition rate ytterbium fiber laser (20kHz, 1.06μm, 120ns) was applied. The experimental single-pulse and multiple-pulse damage thresholds were obtained. To calculate damage thresholds, a 1D analytical model which takes into account the temperature dependent absorptance and multiple-pulse damage based on plastic deformations accumulation was applied. The experimental damage thresholds and the theoretical ones are in a good agreement. Cleaning tests with the contaminated mirrors exposed in JET have been performed. [Copyright &y& Elsevier]

Abstract: The in-vessel tritium inventory control is one of the most ITER challenging issues which has to be resolved to fulfil safety requirements. This is due mainly to the presence of carbon as a constituent of plasma facing components (PFCs) which leads to a high fuel permanent retention. For several years now, physics studies and technological developments have been undertaken worldwide in order to develop reliable techniques which could be used in ITER severe environment (magnetic field, vacuum, high temperature) for in situ tritium recovery. The scope of this contribution is to review the present status of these achievements and define the remaining work to be done in order to propose a dedicated work program. Different treatment techniques (chemical treatments, photonic cleaning) will be reviewed. In the frame of ITER, they will be compared in terms of fuel removal rate as well as surface accessibility, type of production (gas or particulates), ability to clean mixed material. And lastly, consequences of bulk trapping observed in tokamak on the techniques currently under development will be addressed. [Copyright &y& Elsevier]

Abstract: The Joint European Torus (JET) fusion machine is the only device capable of operation with tritium and of handling Be and therefore is best suited to the study of tritium and fusion-related issues. A large variety of activities are performed within the JET fusion technology task force (FT-TF). In this paper, some topics such as erosion/deposition and material transport, characterisation of flakes and detritiation techniques are highlighted. Recent examples of results obtained on waste management studies are also given. Data on some ITER-relevant components that have been tested at JET, such as a pumping cryopanel and hardened optics fibers, are presented. In all fields, the work to be addressed in future JET work programmes is discussed. [Copyright &y& Elsevier]

Abstract: 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. [Copyright &y& Elsevier]

• Tritium distributions in Be limiter tiles from the JET with ITER-like wall were analyzed. • The largest tritium retention was observed at the center of outer poloidal limiter. • The inner wall guard limiter and upper dump plate showed smaller tritium retention than outer poloidal limiter. • Distributions of tritium found at castellation grooves were similar to those of metallic impurities and deuterium. Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis. [ABSTRACT FROM AUTHOR]