Divertor Tiles Research Articles

Design of a 10 MW/m2-heat-flux removal system for a W7-X divertor tile using triply periodic minimal surfaces

Design of a 10 MW/m2-heat-flux removal system for a W7-X divertor tile using triply periodic minimal surfaces

Co-deposited layers on gap surfaces of bulk tungsten divertor tiles in JET ITER-like wall: Directional effects and nanostructures

Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011–2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, “homogeneous” and “directional”, have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

Application of two-dimensional temperature response functions for reconstruction of divertor heat flux profile in commercial fusion reactors.

To keep the tritium breeding rate TBR > 1 and to meet the high heat load and neutron shielding requirements for the first wall and divertor in fusion demonstration (DEMO) reactors, the number of port plugs and other openings must be limited. To accomplish this, it is necessary to develop alternatives to the use of infrared (IR) thermography to determine the peak heat flux and the heat flux profile onto divertor targets. A divertor tile equipped with multiple temperature monitoring channels can be used to reproduce the temperature profile. To avoid the high temperatures and high neutron flux environment in a DEMO, the monitoring positions can be set well away from the irradiated surface. However, the spatial resolution of this method is lower than that provided by IR thermography. In the present work, we apply two-dimensional temperature response functions and the corresponding heat conduction model to temperature data obtained from a divertor tile surface in the large helical device to study the effects of the spatial resolution of the monitored temperature profile on the reconstructed heat flux profile. The findings provide information that will be useful in defining a method for embedding thermocouples into the divertor tiles of future DEMO reactors.

Experimental evidence of very short power decay lengths in H-mode discharges in the COMPASS tokamak

Analysis of heat fluxes in scrape-off layer (SOL) plasma is important for the prediction of divertor tile heat loads in future reactor-sized tokamak machines. Typically, the radial heat flux profile can be accurately characterized by the decay length parameter λ q . The predictions are then based on the dependence of λ q on plasma and machine parameters. In recent years, several empirical scaling models were derived from a collective database of decay lengths observed in several large tokamaks as well as two spherical tokamaks. Most recently, a report from the TCV tokamak showed a deviation from several of the proposed models (Maurizio et al (TCV team) 2021 Nucl. Fusion 61 024003). In this work, edge plasma profiles of ELMy H-mode discharges were collected from the COMPASS tokamak database for heat flux analysis. The COMPASS tokamak has a similar machine size and plasma parameters to the TCV tokamak, offering a valuable comparison. The λ q was measured in both downstream and upstream SOL utilizing a divertor probe array validated by an IR camera and Thomson scattering diagnostics. A comparison with the predictions of several scaling models indicated an occurrence of anomalously short (by a factor of 2–5) inter-ELM power decay lengths in both upstream and downstream, as well as a difference in λ q between these two regions. Possible causes related to the accuracy of magnetic reconstruction were eliminated and physics-based sources of apparent compression effects were hypothesized as a motivation for future research.

Overview of tritium retention in divertor tiles and dust particles from the JET tokamak with the ITER-like wall

Divertor tiles after Joint European Torus-ITER like wall (JET-ILW) campaigns and dust collected after JET-C and JET-ILW operation were examined by a set of complementary techniques (full combustion and radiography) to determine the total, specific and areal tritium activities, poloidal tritium distribution in the divertor and the presence of that isotope in individual dust particles. In the divertor tiles, the majority of tritium is detected in the surface region and, the areal activities in the ILW divertor are in the 0.5–12 kBq cm−2 range. The activity in the ILW dust is associated mainly with the presence of carbon particles being a legacy from the JET-C operation. The total tritium activities show significant differences between the JET operation with ILW and the earlier phase with the carbon wall (JET-C) indicating that tritium retention has been significantly decreased in the operation with ILW.

Performance of a liquid Sn divertor target during ASDEX upgrade L-mode and H-mode operation

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

Two-dimensional collisional particle model of the divertor sheath with electron emissive walls

A novel two-dimensional particle-in-cell (PIC) code, named Divertor Edge Simulator of Plasma-wall Interaction with Consistent COllisions (DESPICCO) and developed at CNR-ISTP, is capable of simulating the thin plasma layer of several millimeters, adjacent to the divertor tiles of a Tokamak fusion reactor. Here, kinetic effects and non-neutral plasma physics in the Debye sheath can be self-consistently captured by the PIC approach. The code is firstly benchmarked against literature one-dimensional codes and additional theoretical predictions for a magnetized sheath. Then, it is applied to a realistic divertor scenario featuring an attached plasma with monoblocks (MBs) radial misalignment and gaps, to compute the energy flux amplification factor at the exposed MB edge. A non-ambipolar local current density close to the leading edge and an average sheath heat transmission coefficient larger than the one predicted by classical sheath theory, are found. The effects of electron wall emission and plasma-gas collisions on the ion Mach number and on particle and energy fluxes to the walls are finally estimated to determine future guidelines for simulations. Ion collisions with recycled neutrals and both secondary and thermionic electron emission from the wall are found to have a relevant impact, with the overall effect of reducing by 25% the average ion impact energy, and by 15%–20% the total heavy particles energy flux to the walls, with relevant implications on the divertor wall erosion.

Retention of noble and rare isotope gases in plasma-facing components – Experience from the JET tokamak with the ITER-like wall

Plasma edge cooling, ion cyclotron wall conditioning and disruption mitigation techniques involve massive gas injection (by puffs or pellets) to the torus. A certain fraction remains in plasma-facing components (PFC) due to co-deposition and implantation. An uncontrolled release/desorption of such retained species affects the stability of plasma operation. The aim of this work was to determine the lateral and depth distribution of noble (3He, 4He, Ne, Ar), seeded (N2, Ne, Ar) and tracer gases (15N, 18O) in PFC retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) after three experimental campaigns (ILW-1, ILW-2, ILW-3). Results regarding the retention of those gases are shown as well as a comparison to the deuterium retention in the studied areas. Heavy ion elastic recoil detection analysis was used, being the only technique capable of detection and quantitative assessment of all elements, especially low-Z isotopes. Helium was found on the divertor Tile 5, locally up to 44·1015 3He cm−2 and 12·1015 4He cm−2, and on the limiters as well. Neon was found in two positions on the limiters, with up to 10·1015 Ne cm−2 and the 15N tracer on Be limiters exposed to ILW-3. A correlation of N retention with the N seeding rates for each campaign has also been found.

Multiphysics Modeling of Impurity Transport for FNSF Startup Scenario with ERO2.0

This study focuses on performing a multiphysics study using the ERO2.0 and UEDGE codes for two standard double null configurations for the Fusion Nuclear Science Facility: (a) 100% recycling and (b) 99% recycling. Results show that the main contributor to tungsten erosion along the divertor plates is impurities from the midplane waveguides. In addition, the standard high-recycling case (100% recycling) shows a significantly higher buildup of impurities along the divertor tiles during the startup phase, which can lead to a higher increase of energy loss in the plasma during steady-state operation. Last, for high recycling, anomalous diffusion can dominate over parallel field diffusion. The work performed in this study can be iteratively applied to a full operation scenario with additional physics such as those from neutrals, wall shaping, and additional external fields.

A multi-scale hybrid approach to the modelling and design of a novel micro-channel cooling structure for the W7X divertor

The second operating phase of the W7X stellarator, with an expanded set of plasma-facing components, includes the test of divertor tiles with a continuous heat load reaching 10 MW/m2. The divertor tiles are cooled by subcooled water. Here a novel cooling concept, based on a network of parallel arrays of micro-channels (MC) with sub-millimetre dimensions, is investigated on a 0.1 m × 0.1 m tile, realizable by Additive Manufacturing. Detailed CFD simulations of the mock-up are performed to check the cooling uniformity using a multi-scale approach, aiming at limiting the dimension of the computational grid without a major loss of accuracy. First, the detailed hydraulic and thermal characterization on a sub-domain with of a small group of MC is performed. Then, the block of MC is substituted with an equivalent porous strip (PS), calibrating the hydraulic and thermal characteristics of the porous medium. The model is verified on an array of MCs or PSs connected to the same manifolds, showing the capability to reproduce the pressure drop and temperature increase with maximum errors of 1.05% and ∽20% in nominal conditions, respectively. The numerical model of the entire tile equipped with PSs is then reliably adopted to evaluate the thermal-hydraulic performance of the cooling device.

Numerical simulation of erosion of misaligned W/Cu monoblocks caused by ITER-like type-Iedge localized modes

The thermal performance of the divertor W/Cu monoblock tiles under the ITER-like transient events has been one of the main concerns for ITER plasma facing components. Owing to the assembly tolerances during installation, the leading edge caused by misalignment between toroidal neighboring tiles will receive the extremely high cumulative heat flux and be damaged. In this work, we develop a two-dimensional heat conductivity model, including evaporation, radiation, melting process, and coupling cooling water condition, to investigate the thermal erosion of two shapes of tiles of W/Cu monoblock (unshaped and beveled tiles) with misalignment in a range from 0 to 0.3 mm, within the allowable maximum misalignment for ITER. To reflect the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma, the energy flux density distribution arriving at the castellated divertor tile surface is evaluated first by using a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo collisions code, and the obtained energy flux distribution is then used as input for the heat conduction model. The simulation results show that the maximum temperature of the unshaped tile with no misalignment is lower than that of the beveled tile under the steady-state inter-ELM heat flux, which increases more quickly than that of the beveled tile with misalignment increasing and will be larger than that of the beveled tile when misalignment is not less than 0.105 mm. Two shapes of the divertor tiles would melt and vaporize under typical heat flux density of a transient event of type-I edge localized modes (ELMs) for ITER, deposition energy of 1 MJ·m<sup>–2</sup> in a duration of 600 μs. The highest temperature, the maximum melting thickness, the maximum vaporization thickness of the unshaped tile with no misalignment are higher than those of the beveled tile except the melting volume ratio. The thermal erosion of the unshaped tile increases more remarkably than that of the beveled tile with misalignment increasing, and the melting volume ratio of the unshaped tile will exceed that of the beveled tile at a misalignment of 0.17 mm. In comparison with the unshaped tile, the beveled tile is more resistant to such a high heat flux of an ELM and misalignment.

Misalignment of magnetic field in DIII-D assessed by post-mortem analysis of divertor targets

We assess the toroidal magnetic field B t asymmetry in DIII-D due to a misalignment of the toroidal field coils with respect to the poloidal magnetic field coils and vacuum vessel. The peak-to-peak variation of the divertor strike point (SP) radius is measured to be 1 cm, with an n = 1 toroidal pattern. We use the centre of a narrow carbon deposition band on tungsten-coated divertor tiles just inside the outer strike point (OSP) as a proxy for the divertor SP location. The band occurred in a series of reverse B t discharges with the OSP positioned on the divertor inserts due to strong E × B drift transport of C from the inner to the outer SP through the private flux region. The variation in band radius (and hence the magnetic SP) is a (4.89 ± 0.31) mm shift toward (310 ± 4)° toroidal direction. These measurements agree well with previous measurements of the 3D magnetic field distribution (Luxon 2003 Nucl. Fusion 43 1813), simulations performed by the mafot field line integration code, and recent Langmuir probe measurements in the small-angle-slot (SAS) divertor (Watkins et al 2019 Nucl. Mater. Energy 18 46). Comparison of these measurements in the SAS divertor also indicates that there is the possibility of a tilt (in conjunction with the shift) of the B t coil field of (0.04 ± 0.07)° towards the toroidal angle of (215 ± 25)°. Previous measurements suggested a field misalignment of (4.6 ± 0.3) mm in the 270° toroidal direction, and a tilt of (0.06 ± 0.02)° toward the 114° toroidal direction, which is similar to the results reported here. These studies will be important for better understanding the radial variation of the toroidal strike line in DIII-D, for designing the new generation of SAS divertor, and for developing an understanding of the impact of error fields on tokamaks with tightly baffled slot divertors.

Dust Monitors in JET with ITER-like Wall for Diagnosis of Mobilized Particles and Co-Deposited Layers

Silicon plates were installed above the inner and outer divertor of the JET with the ITER-like wall (ILW) after the second and third ILW campaigns to monitor dust generation and deposition with the aim to determine the morphology and content of individual particles and co-deposits, including deuterium content. Particular interest was in metal-based particles: Be, W, steel, Cu. Ex-situ examination after two ILW campaigns was performed by a set of microscopy and ion beam methods including micro-beam nuclear reaction analysis and particle-induced X-ray emission. Different categories of Be-rich particles were found: co-deposits peeled-off from plasma-facing components (PFC), complex multi-element spherical objects, and solid metal splashes and regular spherical droplets. The fuel content on the two latter categories was at the level of 1 × 1016 at/cm-2 indicating that Be melting and splashing occurred in the very last phase of the second experimental campaign. The splashes adhere firmly to the substrate thus not posing risk of Be dust mobilisation. No tungsten droplets were detected. The only W-containing particles were fragments of tungsten coatings from the divertor tiles.

Studies of material deposition on the graphite divertor tile after the 2019 experimental campaign in EAST

• Detailed analysis of the deposits on the surface of the graphite tile from the divertor dome region has been made after the 2019 experimental campaign. • The porous globular deposits on exposed graphite tile were mainly stacked by columnar grains with size of 2–5 μm and small crystal particles with several hundreds of nanometers in size. • The thickness of the deposits varied from several micrometers to about 120 μm. • As the SiC coating on the graphite tile has a rough surface and exhibited valleys and hills, the deposits on the valleys were typically thicker than that on the hills. • Li was in the form of Li 2 CO 3 . And Mo and W were in MoO3 and WO3 respectively. • The total content of Li 2 CO 3 was found to be higher than 90 wt% with a corresponding Li density in deposits of about 7.22 μg/mm −2 in average. • The intensity of Li was strong and changed only slightly with the depth, which was different from other elements. Elements of W, Mo, Fe, Cu seemed to be more concentrated at the near surface with a depth of about 4 μm. Material erosion and deposition are one of the important topics related to plasma-wall interaction in tokamaks. Detailed analysis of the deposits on the surface of the graphite tile from the divertor dome region has been made after the 2019 experimental campaign. It is indicated that the deposits exhibited porous globular in morphology which were mainly stacked by columnar grains with size of 2–5 μm and small crystal particles with several hundreds of nanometers in size. The thickness of the deposits varied from several micrometers to about 120 μm. As the SiC coating on the graphite tile has a rough surface and exhibited valleys and hills, the deposits on the valleys were typically thicker than that on the hills. The deposits consisted of Li, C, O, W, Mo, Fe, Cu, Ni, et. mainly from the wall conditioning materials and plasma-facing materials. It was noted that Li was in the form of Li 2 CO 3 . And Mo and W were in MoO 3 and WO 3 respectively. The total content of Li 2 CO 3 was found to be higher than 90 wt% with a corresponding Li density in deposits of about 7.22 μg/mm −2 in average due to the routine Li wall conditioning. The depth profile of Li exhibited differently from the other elements. And W has a higher density than Mo, which could be due to the high W erosion from the upper divertor.

Reassessment of tritium content in CFC tiles exposed to the JET D-T campaign in 1997

• In 2019 two MkIIA divertor tiles (6IN3 and 4BN4) exposed during DTE1 were retrieved at CCFE for Thermal Desorption Spectroscopy (TDS) and pyrolysis analyses. A set of samples were prepared using a coring technique. The highest tritium (T) inventories were found in the shadowed corner of the inner divertor due to asymmetric deposition. TDS analyses indicated that T is desorbed at rather high temperatures with maximum release peaks at ∼590 and 820 o C. T amounts were also investigated as a function of depth from the tile surface up to a depth of ∼4.5mm and T was detected at these depths. Comparison was also made with old results obtained with full combustion technique and PIN-diode method a few years after the DTE1 experiment, allowing for the natural decay of T. Our results agree within a factor of ∼3 with these results. In 2019 two MkIIA divertor tiles (6IN3 and 4BN4) exposed during DTE1 were retrieved at CCFE for Thermal Desorption Spectroscopy (TDS) and pyrolysis analyses. A set of samples were prepared using a coring technique. The highest tritium (T) inventories were found in the shadowed corner of the inner divertor due to asymmetric deposition. TDS analyses indicated that T is desorbed at rather high temperatures with maximum release peaks at ∼590 and 820 o C. A few samples were reannealed at 850 o C using the same heating procedure and it turned out that a further ∼40-50 % of T was still released indicating that the annealing procedure used does not empty the sample completely. Pyrolysis results for thin disks cut from the surface of the tile were somewhat higher than the corresponding TDS results. T amounts were also investigated as a function of depth from the tile surface up to a depth of ∼4.5mm and T was detected at these depths. Comparison was also made with old results obtained with the pyrolysis technique and a PIN-diode method a few years after the DTE1 experiment, allowing for the natural decay and off-gassing of T. Our results agree within a factor of ∼3 with these results.

Characterization of the impurity features deposited on the boronization tungsten tiles exposed in KSTAR tokamak using laser-induced breakdown spectroscopy

• The boron and other impurity elements such as: carbon, silicon, calcium and copper deposition on the surface of tungsten divertor tiles removed from KSTAR has been measured by LIBS. • A subtraction method is proposed to eliminate the W line contribution for B deposition measurement. • The lateral-distribution of the boron and carbon on the three tungsten castellated tiles (Rounded, Chamfered and Double chamfered) were carried out at 10 −5 mbar vacuum condition. Boronization is an important wall conditioning procedure for realizing the steady-state high performance operation of tokamaks such as KSTAR and ITER. The measurement of boron and its distribution on the first wall and divertor has a significant impact on studying the plasma-wall interaction (PWI) such as impurity generation, transportation and deposition behaviors. In this study, the laser-induced breakdown spectroscopic (LIBS) diagnostic method was applied in the lab for investigating the boron impurity deposited on the surface of different shaped tungsten (W) castellated tiles which were removed from the divertor region in the Korea Superconducting Tokamak Advanced Research (KSTAR). Meanwhile, other various impurity elements comprising carbon, silicon, calcium and copper were also observed. A subtraction method is proposed to eliminate the W line contribution for B deposition measurement. The depth profile of the impurity elements in the deposited layer also has been studied by the variation of spectral intensity with consecutive laser pulses. The lateral-distribution of the boron and carbon on the three tungsten castellated tiles (Rounded, Chamfered and Double chamfered) were carried out at 10 −5 mbar vacuum condition. The distribution features of the impurities on the three different shaped tungsten tiles provide valuable information to understand PWI process in KSTAR. This study demonstrates that LIBS is a potential diagnostic approach for monitoring the impurity deposition in tokamak such as KSTAR or ITER.

3D deposition patterns of deuterium retention and impurities in the COMPASS divertor: a data-driven root cause analysis and prediction approach

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 Te, parallel heat flux Qpar and ion saturation current Isat 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 (Te, Isat, Qpar) 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.

Overview of the COMPASS results * *In the future we will refer to the author list of the paper as the COMPASS Team.

COMPASS addressed several physical processes that may explain the behaviour of important phenomena. This paper presents results related to the main fields of COMPASS research obtained in the recent two years, including studies of turbulence, L–H transition, plasma material interaction, runaway electron, and disruption physics: Tomographic reconstruction of the edge/SOL turbulence observed by a fast visible camera allowed to visualize turbulent structures without perturbing the plasma.Dependence of the power threshold on the X-point height was studied and related role of radial electric field in the edge/SOL plasma was identified.The effect of high-field-side error fields on the L–H transition was investigated in order to assess the influence of the central solenoid misalignment and the possibility to compensate these error fields by low-field-side coils.Results of fast measurements of electron temperature during ELMs show the ELM peak values at the divertor are around 80% of the initial temperature at the pedestal.Liquid metals were used for the first time as plasma facing material in ELMy H-mode in the tokamak divertor. Good power handling capability was observed for heat fluxes up to 12 MW m−2 and no direct droplet ejection was observed.Partial detachment regime was achieved by impurity seeding in the divertor. The evolution of the heat flux footprint at the outer target was studied.Runaway electrons were studied using new unique systems—impact calorimetry, carbon pellet injection technique, wide variety of magnetic perturbations. Radial feedback control was imposed on the beam.Forces during plasma disruptions were monitored by a number of new diagnostics for vacuum vessel (VV) motion in order to contribute to the scaling laws of sideways disruption forces for ITER.Current flows towards the divertor tiles, incl. possible short-circuiting through PFCs, were investigated during the VDE experiments. The results support ATEC model and improve understanding of disruption loads.

Deuterium and beryllium depth profiles into the W-coated JET divertor tiles after ITER-like wall 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.

An overview of tritium retention in dust particles from the JET-ILW divertor

Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles.