Focus on Plasma-Facing Materials and Components for Fusion Applications - Selected Papers from PFMC 18 Conference

One of the missions of the WEST tokamak is to test in a realistic tokamak environment, the ITER-like divertor plasma facing components made with tungsten. On exposed leading edges of monoblocks, poloidally-distributed cracks having an average spacing of 0.4 mm, running perpendicular to the cooling tube axis, were observed following the experimental campaigns in 2018. Damage may be induced by different processes which can lead to: brittle fracture below the Ductile to Brittle Transition temperature or ductile failure for which softening (recovery/recrystallization) process plays a major role. To improve our understanding about the leading edge damage process, numerical simulations were run here. The following results are specially studied: (i) the thermal gradients; (ii) the softening fraction gradient and (iii) the stress and strain distributions taking into account the mechanical properties of tungsten (elastic-viscoplastic) and the softening phenomenon. This paper describes the results obtained for a range of WEST steady state parallel heat flux (from 45 MW/ m² to 70 MW/ m² ) and disruption (600 MW/ m² ) heat loading on the leading edge. Estimated results related to the plastic strain accumulation, give an interpretation of the premature cracking of the monoblock leading edge but do not explain the poloidal distribution. According to the numerical results, brittle fracture are expected under disruption as estimated normal stresses are beyond the material yield stress, along 88% of the leading edge.

A sequence of fuel recovery methods was tested in JET, equipped with the ITER-like beryllium main chamber wall and tungsten divertor, to reduce the plasma deuterium concentration to less than 1% in preparation for operation with tritium. This was also a key activity with regard to refining the clean-up strategy to be implemented at the end of the 2nd DT campaign in JET (DTE2) and to assess the tools that are envisaged to mitigate the tritium inventory build-up in ITER. The sequence began with 4 days of main chamber baking at 320 °C, followed by a further 4 days in which Ion Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) were applied with hydrogen fuelling, still at 320 °C, followed by more ICWC while the vessel cooled gradually from 320 °C to 225 °C on the 4th day. While baking alone is very efficient at recovering fuel from the main chamber, the ICWC and GDC sessions at 320 °C still removed slightly higher amounts of fuel than found previously in isotopic changeover experiments at 200 °C in JET. Finally, GDC and ICWC are found to have similar removal efficiency per unit of discharge energy. The baking week with ICWC and GDC was followed by plasma discharges to remove deposited fuel from the divertor. Raising the inner divertor strike point up to the uppermost accessible point allowed local heating of the surfaces to at least 800 °C for the duration of this discharge configuration (typically 18 s), according to infra-red thermography measurements. In laboratory thermal desorption measurements, maintaining this temperature level for several minutes depletes thick co-deposit samples of fuel. The fuel removal by 14 diverted plasma discharges is analysed, of which 9, for 160 s in total, with raised inner strike point. The initial D content in these discharges started at the low value of 3%–5%, due to the preceding baking and conditioning sequence, and reduced further to 1%, depending on the applied configuration, thus meeting the experimental target.

One set of horizontal target elements of the Test Divertor Units (TDU), retrieved from the Wendelstein 7-X (W7-X) vessel after the end of second divertor Operation Phase (OP1.2B) in Hydrogen (H), were investigated by picosecond Laser-Induced Breakdown Spectroscopy (ps-LIBS). The Boron (B) distribution, H pattern and the material erosion/deposition pattern on these target elements were analyzed with high depth resolution and mapped in the poloidal direction of W7-X. From the spectroscopic analysis, B, H, Carbon (C) and Molybdenum (Mo) were clearly identified. A non-uniformly distributed B pattern on these divertor target elements was determined by the combination of B layer deposition during the three boronizations and W7-X plasma operation with multiple erosion and deposition steps of B. Like the TDU, the analyzed target elements are made of fine grain graphite, but have two marker layers which allow us to determine the material migration via the ps-LIBS technique. Two net erosion zones including one main erosion zone with a peak erosion depth of 6.5 μm and one weak erosion with a peak erosion of 1.3 μm were determined. Between two net erosion zones, a net deposition zone with width of 135 mm and a thickness up to 3.5 μm at the peak deposition location was determined by the ps-LIBS technique. The B distributions are correlated with the erosion/deposition pattern and the operational time in standard magnetic configuration of W7-X in the phases after the boronizations. The thickness of the containing B layer on these target elements also correlates with the erosion/deposition depth, in which the thickness of the containing B layer varies spatially in poloidal direction between 0.1 μm and 6 μm. Complementary, Focused Ion Beam combined with Scanning Electron Microscopy (FIB-SEM) was employed also to verify and investigate the deposition layer thicknesses at typical net erosion and net deposition zones as well as to identify the three boronizations in depth.

In this work the synergism between displacement damage creation and presence of hydrogen isotopes was studied. Tungsten samples were irradiated by 10.8 MeV W ions with or without the presence of D ions with two different energies of 300 eV/D and 1000 eV/D and different temperatures. In order to compare the results obtained with different exposure parameters the samples were afterwards additionally exposed to D ions at 450 K to populate the created defects. By increasing the W irradiation time, ion flux and energy, the increase of D concentration and D retention was observed as measured by nuclear reaction analysis and thermal desorption spectroscopy. By fitting the D depth profiles and D desorption spectra by the rate equation code MHIMS-R we could see that additional fill-levels were populated with higher flux and ion energy which ends up in higher final D concentration and retention as compared to experiments with lower D flux and energy.

Erosion and deposition is modelled with ERO2.0 for a hypothetical full-tungsten ITER for an ELM-free H-Mode baseline deuterium discharge. A parameter study considering seeding impurities (Ne, Ar, Kr, Xe) at constant percentages (0.05% to 1.0%) of the deuterium ion flux is done while neglecting their radiation cooling and core plasma compatibility. With pure deuterium plasma, tungsten main wall erosion is only due to charge exchange deuterium atoms and self-sputtering and there is only minor tungsten divertor sputtering. With a beryllium main wall, beryllium erosion is due to deuterium ions, charge exchange deuterium neutrals and self-sputtering. For this case, tungsten in the divertor is eroded by beryllium ions and self-sputtering. The simulations for full-tungsten device including seeded impurities leads to significant tungsten erosion in the divertor. In general, tungsten erosion, self-sputtering and deposition increase by factors larger than 50 at the main wall and 5000 in the divertor compared to pure deuterium plasma.

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.

Based on the conventional model of hydrogen retention in plasma-facing components, the question of hydrogen outgassing during and after plasma exposure is addressed in relation to mass spectrometry and laser-induced breakdown sprectroscopy (LIBS) measurements. Fundamental differences in retention and release data acquired by LIBS and by mass spectrometry are described analytically and by modelling. Reaction-diffusion simulations are presented that demonstrate possible thermal outgassing effects caused by LIBS. Advantages and limitations of LIBS as a tool for analysis of short term retention are discussed.

An assessment of the tritium (T) inventory in plasma facing components (PFC) during JET T and deuterium-tritium (DT) operations is presented based on the most comprehensive ex situ fuel retention data set on JET PFCs from the 2015-2016 ILW3 operating period is presented. The global fuel retention is 4.19 × 10²³ D atoms, 0.19% of injected fuel. The inner divertor remains the region of highest fuel retention (46.5%). The T inventory in PFCs at the end of JET operations is calculated as 7.48 × 10²² atoms and is informative for accountancy, clean-up efficacy and waste liability assessments. The T accumulation rate at the upper inner divertor during JET DT operations has been used to assess the requirements and frequency of operation of a new laser induced desorption diagnostic to be installed on JET for the final DT experiments in 2023.

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10–15 μm of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a 'V' shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned 'V' shape predict a substantial reduction in W sourcing and SOL accumulation in either B × ∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. This coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

Tungsten Langmuir probes retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) after the second ILW campaign were examined by nano-indentation, microscopy and x-ray diffraction in order to determine changes in mechanical properties and phase composition. Not-exposed probe served as a reference material. Two regions were studied: (i) recrystallized region below the tip and, (ii) the lower probe structure, called 'support structure'. A large difference between the hardness in the tip and the other region has been found: 5 GPa versus 15 GPa, respectively. The measured values of the Young's modulus in both zones of exposed probe are at the same level of 260 GPa. From the force-displacement curves, it can be concluded that the material in the tip has a smaller range of elastic deformations compared to that characteristic for the support structure. The values obtained for the material in its initial state are consistent with the available literature data for tungsten. With x-ray diffraction and microscopy only tungsten has been detected in the probe tip. It remained clean and free from impurities and undesirable compounds, which could have a negative impact on the probes electrical properties.

The morphology of beryllium coatings on the Inconel inner wall cladding tiles after JET-ILW campaigns was determined. The focus was on: (i) fuel retention and its share in the overall fuel inventory; (ii) the change of the layer structure and composition. The study is motivated in the view of planned D-T operation in JET. Four tiles were examined: the initial not exposed; one exposed to two campaigns (ILW1-2) and two facing the plasma during ILW1–3. As determined with ion beam and microscopy methods, the initial Be layer (9.0 μm thick) contained up to 4–5 at.% of impurities, mainly H, O, C, Ni. In the exposed tiles, the impurity content increases to 14–26 at.% (up to 20 at.% O, 1.7 at.% C, 1.0 at.% N, 1.3 at.% Ni and under 0.1 at.% W). The surface composition indicates gettering of O and a long-term retention of N. The Be thickness on the tile exposed to ILW1–2 was between 7.6 and 9.7 μm, thus indicating erosion in some areas, while the thickness after ILW1–3 increased to 10–12 μm. The D content was in the range 1.2–3.4×10¹⁷ cm⁻² after ILW1–2 and 3.2–10×10¹⁷ cm⁻² after ILW1–3 on most of the analyzed area, but in the limiter shadow values up to 58 ×10¹⁷ cm⁻² were measured. Taking into account the total area of the Be-coated inner wall cladding tiles, the lower limit of D inventory amounts to 5.3×10²² atoms corresponding to about 176 mg, i.e. somewhat greater than the amount determined on Be limiters. The formation and spalling-off of Be-O particles was revealed.

In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s⁻¹ averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 10¹² W-atoms/cm²s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 10¹⁷–10¹⁸ B + C atoms/cm² (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

Tritium (T) permeation through plasma-facing component (PFC) into the coolant is a major concern of fusion reactor operation. In this work, deuterium (D) permeation through CLAM steel, CLAM/CLAM and CLAM/Fe-Cr-Al samples prepared by hot isostatic pressing (HIP) are tested in a linear plasma device. Only the downstream surfaces of the samples are oxidized with controlled atmosphere to form permeation barrier. No significant effect on D diffusion and penetration can be observed for the joining interfaces, while the dense oxide layer at the downstream side plays an important role in suppressing D permeation. The downstream surface oxidization of CLAM/Fe-Cr-Al is found to effectively reduce D permeation flux by a factor up to 1000.

The elemental quantification of liquid metal divertor (LMD) surface is important for understanding the material erosion, migration, re-deposition, and fuel retention in Plasma-Facing Components (PFCs). Currently, LMD are attractive candidates for the short- and long-term operation of fusion devices like DEMO. Liquid metals can provide self-cooling, self-replenishing plasma-facing surfaces requiring very little upkeep. In a previous work, we studied Li and LiSn layers deposited on attachment screws in the COMPASS tokamak by means of Calibration-Free Laser-Induced Breakdown Spectroscopy (CF-LIBS). Several problems were encountered related to the detection of Sn in LiSn. Thus, in the present work, we are optimizing the experimental conditions for the detection of Sn I-II and Pb I-II in Pb-containing Sn-based alloys, performing the quantification of Pb in traces and in bulk quantities using CF-LIBS approach.

We analyzed data from ultrasonic testing and infrared thermography non-destructive examinations performed on a subset of 25 small scale water-cooled target mockups of four target designs developed in the 2nd R&D phase of WPDIV. Examinations were performed before high heat flux tests for the 25 mockups and after high heat flux tests for 18 of them. The detected manufacturing defects are various in size and location. Widths of defects range from 0.2 to 12 mm, which is the entire block width. Angular sizes of defects range from few degrees to 360°. Defects having an angular size less than 50° or a width less than 4 mm are likely to be missed by infrared thermography examinations. After high heat flux tests at 20 MW m⁻² up to 500 cycles, we noticed no significant evolution of pre-existing defects.

Laser-induced breakdown spectroscopy is applied successfully for plasma-wall interaction studies in several fusion devices and post-mortem analyses of plasma-facing materials. However, the quantitative as well as qualitative analysis of low hydrogen isotope contents in tungsten plasma-facing components is still challenging. A promising approach to increase the optical signal in laser-induced breakdown spectroscopy is to apply a second laser pulse to the laser-produced plasma. We present two setups for post mortem plasma-facing component analyses using different laser pulse properties and different excitation geometries. The enhancement factors and changes in spectral line shapes for double pulse compared to single pulse laser-induced breakdown spectroscopy are presented.

Material related limitations are one of the main challenges for the design of future fusion reactors. Tungsten (W) as the primary material choice is considered resilient against erosion, has the highest melting point of any metal and shows low activation after neutron irradiation. However, W is intrinsically brittle and faces operational embrittlement. To overcome these issues, W-based composites have been in development. W fiber-reinforced W composite materials (W_f/W) incorporate extrinsic toughening mechanisms allowing the redistribution of stress peaks and thus overcoming the intrinsic brittleness problem. In this contribution recent results on the incorporation of new textile preformes into W_f/W production will be given with a focus on the production via chemical vapor deposition of tungsten-based materials. The use of tungsten yarns, instead of single wires for the textile production is elaborated.

Once ITER commences full power operation, the ITER divertor will be exposed to high thermal and particle loads. Tungsten was chosen as plasma facing material for the ITER divertor. It is, therefore, of the utmost importance to understand the behavior of ITER-grade tungsten under conditions similar to those it will have to withstand inside the reactor. In this study, ITER-grade tungsten samples were exposed to stationary D/He(6%) plasma and ELM-like transient heat loads in the linear plasma device PSI-2. The effects of each kind of load was first studied separately, and the synergistic effects obtained when exposed to both loads simultaneously were then analyzed. Additionally, the hardness of a recrystallized tungsten sample after exposure to simultaneous loads was tested via nanoindentation. The results indicate that hydrogen and helium embrittlement worsens the cracking behavior of the material when exposed to the simultaneous loads compared to only heat loads. Additionally, bubbles of up to 1 μm are formed under the surface due to the synergistic effects at the highest heat load. The nanoindentation tests showed that plasma and heat loads increase the hardness of the material by 39%, but only plasma loads appeared to have no effect on it.

Tungsten-based materials are possible candidates as PFCs in future fusion devices. LIBS is one of the most suitable techniques for monitoring erosion and deposition processes including fuel retention, due to its versatility and ability to perform in situ measurements. By deploying ps-LIBS, instead of ns, the laser ablation occurs with fewer melting effects. This work compares ns- and ps- (CF)-LIBS characterization of WZr(D) samples, at the linear plasma generator at Magnum-PSI at the DIFFER. The laser energy has been optimized for both laser regimes, lowering the laser energy for the ns regime (from 19.9 mJ pulse⁻¹ to 7.4 mJ pulse⁻¹) to approximate to ps regime (0.3 mJ pulse⁻¹). All the experimental measurements have been performed at P_atm. The pure WZr samples have been analyzed in ambient air, while the WZrD sample measurements have been performed under Ar gas flow. The retained deuterium content varies from 4 at% to 0.3 at%.

Wendelstein 7-X (W7-X) is an optimized stellarator with a 3-dimensional five-fold modular geometry. The plasma-wall-interaction (PWI) investigations in the complex 3D geometry of W7-X were carried out by in situ spectroscopic observations, exhaust gas analysis and post-mortem measurements on a large number of plasma-facing components extracted after campaigns. The investigations showed that the divertor strike line areas on the divertor targets appeared to be the major source of carbon impurities. After multistep erosion and deposition events, carbon was found to be deposited largely at the first wall components, with thick deposits of >1 μm on some baffle tiles, moderate deposits on toroidal closure tiles and thin deposits at the heat shield tiles and the outer wall panels. Some amount of the eroded carbon was pumped out via the vacuum pumps as volatile hydrocarbons and carbon oxides (CO, CO₂) formed due to the chemical processes. Boron was introduced by three boronizations and one boron powder injection experiment. Thin boron-dominated layers were found on the inner heat shield and the outer wall panels, some boron was also found at the test divertor unit and in redeposited layers together with carbon. Local erosion/deposition and global migration processes were studied using field-line transport simulations, analytical estimations, 3D-WallDYN and ERO2.0 modeling in standard magnetic field configuration.

The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m⁻² for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 μm followed by a re-solidified tungsten bump of 200 μm in the JxB direction.

It is still challenging to achieve the quantitative analysis with a spatially resolved measurement for a non-uniform fuel retention in Plasma Facing Materials (PFMs). In this work, a long-pulse (τ_p ≈ 750 μs) Laser Induced Desorption-Quadrupole Mass Spectroscopy (LID-QMS) has been developed as a diagnostic approach for characterizing inhomogeneous deuterium (D) retention in tungsten (W) target exposed to D plasma for ∼1 h with the total ion flux of 4.0 × 10²¹ D/(m² s) at the substrate temperature of 400 K. The behavior of desorption of trapped D during LID was analyzed. 2D-distribution features of D retention in the exposed and shadowed areas of the W-target were investigated with a lateral resolution of about 650 μm and depth resolution of approximately 175 μm. The results show that the D retention is almost uniform along lateral dimension in the D plasma exposure area with an average fluence of (2.51 ± 0.40) × 10²⁰ D m⁻², while the value in the shadow area is only (0.25 ± 0.04) × 10²⁰ D m⁻² which is one order of magnitude smaller than that in the exposure region. In the depth dimension, about 86.5% of the retained D is desorbed by the first laser pulse, which indicates that the D-retention is mainly distributed in the region less than 6.6 μm based on the diffusion length. The micro-structures of W target exposed to D plasma after the long-pulse laser irradiation present swelling, blistering as well as bubble-bursting, which might be attributed to the thermodynamic interaction due to the rapid release of D particles from bulk to surface of W target during LID.

We report carbon impurity ion incident angles and deposition rates, along with silicon erosion rates, from measurements of micro-engineered trenches on a silicon surface exposed to L-mode deuterium plasmas at the DIII-D divertor. Post exposure ex-situ analysis determined elemental maps and concentrations, carbon deposition thicknesses, and erosion of silicon surfaces. Carbon deposition profiles on the trench floor showed carbon ion shadowing that was consistent with ERO calculations of average carbon ion angle distributions (IADs) for both polar and azimuthal angles. Measured silicon net erosion rates negatively correlated with the deposited carbon concentration at different locations. Differential erosion of surfaces on two different ion-downstream trench slope structures suggested that carbon deposition rate is affected by the carbon ion incident angle and significantly suppressed the surface erosion. The results suggest the C impurity ion incident angles, determined by the IADs and surface morphology, strongly affect erosion rates as well as the main ion (D, T, He) incident angles.

The ITER-Like-Wall project has been carried out at the Joint European Torus (JET) to test plasma facing materials relevant to ITER. Materials being tested include both bulk metals (Be and W) and coatings. Tritium accumulation mechanisms and release properties depend both on the wall components, their location in the vacuum vessel, conditions of exposure to plasma and to the material itself. In this study, bulk beryllium limiter tiles, plasma-facing beryllium coated Inconel components from the main chamber, bulk tungsten and tungsten coated carbon fibre composite divertor tiles were analysed. A range of methods have been developed and applied in order to obtain a comprehensive overview on tritium retention and behaviour in different materials of plasma facing components (PFCs). Tritium content and chemical state were studied by the means of chemical or electrochemical dissolution methods and thermal desorption spectroscopy. Tritium distribution in the vacuum vessel and factors affecting its accumulation have been assessed and discussed.

In order to map the complex plasma footprint observed on the lower divertor of the WEST experiment after its first phase of operation, we analysed deposits collected on ITER-like Plasma Facing Units (PFUs) exposed in the C3 (deuterium plasma) and C4 (deuterium and helium plasma) campaigns. Our results show that these deposits have multilayer structures mainly composed of tungsten, oxygen, boron and carbon, whose texture and composition vary along the radial direction. These traits allowed us to identify three types of deposits: deposits rich in boron (conditioning) in the low plasma flux area further away from the strike point, deposits rich in tungsten with traces of metallic compounds (Cu, Fe, Cr, Ni, Ag) in the high plasma flux area and deposits rich in boron and tungsten in the private flux area. In addition, we found more nanoparticles, voids and tungsten oxide layers in the deposits formed during C4 in comparison to that of C3.

The operating temperature window of solid tungsten (W) is dictated by its Ductile to Brittle Transition Temperature (DBTT) and re-crystallization temperature; roughly between 300-1300 °C. The brittleness of W at lower temperatures is exasperated when it undergoes recrystallization. We investigate here the thermal shock resistance of micro-porous W as a meta-material fabricated in a 3D open-cell network structure. We present experimental results for the effects of cyclic high-enthalpy arc-jet plasma on surface damage in three testing categories. Observed damage includes surface ablation of asperities, melting and solidification of W-fuzz on samples that have been exposed to a prior helium plasma, and micro-cracks at ligament triple junctions. Scanning electron microscope (SEM) observations show more micro-cracking on 54% and 23% foams compared to the 43% ones. In all tested samples, thermal expansion/contraction displacements were accommodated by ligament rotation and a network of micro-cracks. No large through-thickness crack were observed.

The lithium and boron powder have been successfully real-time injected into Experimental Advanced Superconducting Tokamak (EAST) high-confinement (H−) mode discharges, showing good wall conditioning effects, especially on the impurities control. Both of lithium and boron powder particles are gravitationally accelerated into the upper edge of a upper single null discharge. The lithium powder real-time injection effectively reduced the tungsten impurity content both in the plasma edge and core. Contrary to the lithium injection, the boron injection mainly reduced the impurities content in the plasma core, while the impurities increased in the edge. This mainly due to that there was an edge harmonic mode stimulated by boron injection, providing sufficient particles transport. Moreover, the impurities control through boron injection has much wider operation windows than lithium powder injection. Furthermore, there was accumulative wall conditioning effects after sequential boron powder injection discharges.

Dust particles and co-deposits were sampled for the first time from beryllium limiters and bulk tungsten divertor (both after ILW-3), and test mirrors from the main chamber after ILW-2 and ILW-3. The focus was on the morphology of molten particles and metal outgrowths. In parallel, the stability of beryllium layers under the impact of hot water was examined on limiters and Be coatings. The study performed by several microscopy methods including SEM, AFM, FIB, TEM and Be-sensitive EDX has revealed: (i) an asymmetric distribution of Be particles with the largest objects (125-550 μm) on side surfaces of the Be tile: (ii) the presence of highly porous particles, resembling blisters; (iii) very few thin flakes of co-deposits on the divertor, on the shadowed edge of lamella; (iv) the elemental composition and internal structure of the needle-shaped outgrowths on the mirrors: MoO; (v) no detectable impact of water on the beryllium morphology.

In continuation of earlier work on 3 MeV proton-damaged tungsten and reduced-activation steels we present new results on Eurofer97, Beryllium and W-5%Re sintered alloy irradiated <400 K. Methodical improvements result in largely reduced uncertainties. Beryllium is loaded using a 5 kV D₂⁺ ion-source to 6.3*10²¹ D m⁻² at 300 K. Eurofer97 and W-5Re are loaded in PSI-2 to 3*10²⁵ D m⁻². Irradiation and D-loading are conducted at ∼400 K. The D retention is measured by ³He μ-NRA. An exponential saturation fits the W-5Re D-retention data with R² = 0.99 . The retention increases by a factor 10.3 in W-5Re, similar as in W, but on a ~7 times lower level. Within ±25% uncertainty D Retention in Eurofer97 proves to be independent of displacement damage up to 6.3 DPA. Beryllium shows increased retention by a factor 3 up to the tested maximum of 0.08 DPA. The retention in beryllium starts saturating, but the limited DPA range allows fitting the data with exponentials and power-laws.

During H-mode plasma in ITER, type-I Edge Localized Modes can occur naturally leading to transient heat load (<10 GW m⁻²) on the tungsten divertor. Such high heat fluxes might therefore induce annealing effects, such as recovery and recrystallization, or even lead to surface melting and deterioration (e.g. cracks) of the plasma-facing components. To ensure sufficient lifetime of tungsten monoblocks is thus mandatory to understand how W will evolve when exposed to such transient heat loads. We present here results on the study of grain growth and damage threshold of tungsten induced by transient heat loads (from 1 to 15 ms, <3 GW m⁻²). Using a high-power laser facility, we simulate ELMs and we submitted samples to various heating durations up to melting temperature. Laser heating is combined to several imaging techniques, like SEM and optical microscopy, to study grain growth and damages on W. We observed that the growth kinetics strongly depends on the repetition rate with a rapid grain growth for 100 Hz heat loads. We eventually built a finite element methods simulations, benchmarked against our experimental results, to evaluate temperature gradients and recrystallized fraction of the material when exposed to unmitigated ELMs.

Experiments in JET with the Be/W ITER-like wall show that in pure hydrogen low-confinement mode (L-mode) plasmas the density limit is approximately 20% higher than their corresponding deuterium plasmas. The maximum achievable density in L-mode plasmas is limited by the magnetohydrodynamic stability of the m/n = 2/1 tearing mode. Studies with the edge fluid-neutral Monte-Carlo code package EDGE2D-EIRENE show that the density of hydrogen atoms inside the separatrix is two times lower than for deuterium in plasma conditions preceding the density limit. The difference between the isotopes is caused by the non-linear process at density limit onset which leads to more efficient dissociation and ionization of hydrogen molecules and atoms in hydrogen than in deuterium plasmas at the high-field X-point region at electron temperatures lower than 2 eV. The m/n = 2/1 island size is estimated to be 30% smaller islands for hydrogen than for deuterium cases for equal fuelling conditions and radial transport assumptions.

Purpose of WEST is to test key technologies for future device like ITER. In particular the actively cooled plasma facing component technology, the so-called tungsten monoblock, will equip the lower divertor, allowing studies and analyses of the plasma wall interaction. 456 plasma-facing components have been manufactured during the last years, using hot isostatic pressing as main bonding process. The manufacturing of the components was not completely straightforward; however, it was successful and main requirements have been reached. The key manufacturing steps, as well as the associated controls are described.

Carbon impurity transport and deposition were investigated in the Wendelstein 7-X stellarator by injecting isotopically labelled methane (¹³CH₄) into the edge plasma during the last plasma operations of its Operational Phase (OP) 1.2B experimental campaign. ¹³C deposition was measured by secondary ion mass spectrometry (SIMS) on three upper divertor tiles located on the opposite side of the vessel to the¹³CH₄ inlet. The highest ¹³C 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.

The activelly cooled plasma facing units (PFUs) constituting the WEST lower divertor must meet strict technical specifications before their installation into the WEST tokamak. The tests performed at CEA lead mainly: to provide information on the feasibility to attach mechanically PFUs on sectors, to ensure geometrical tolerances for the welding of PFUs to water manifolds, to check the PFU vacuum tightness and to confirm the PFUs heat exhaust capability. Using high heat flux (HHF) test facilities, such as HADES at CEA-Cadarache and GLADIS at IPP-Garching, ∼5% of the PFU production was tested. Infrared thermography (IR) tests were also performed (∼24% of the PFU production tested). We show that PFUs are with a quality in agreement to the requirements and that the assessement of the heat exhaust capability during the series production is needed. Based on statistical approaches, this work also provides information on the methods to assess the quality of tested components using statistic process control.

The castellated tungsten monoblock is considered one of the preferred reference designs of components faced to plasma in the divertor of a fusion reactor. The behavior of such components during the development of transient events (disruptions or/and of Edge Localized Modes (ELMs)) is still among the most critical issues for future thermonuclear devices. Experimental studies of castellated tungsten surface erosion have been performed within the powerful quasi-stationary plasma accelerator QSPA Kh-50 in conditions relevant to unmitigated ITER Type I ELMs. The surface energy load from the impacting hydrogen plasma streams was 0.9 MJ m⁻² during 0.25 ms to achieve pronounced melting of the target surface. The studies of solid and liquid particle generation were performed at sequentially oblique plasma exposure of the different chosen edges of the castellated structure. It was determined that most of the W droplets are ejected from the sharp leading edge of the target due to the development of instabilities in the plasma-liquid metal interface. At the same time, the formation of solid dust particles is mainly attributed to the cracking during the surface cooling after the plasma pulse. Particle velocities typically achieve tens m s⁻¹. Both the re-solidified droplets as well as dust particles have been collected on the special plate near the exposed surface. The maximal size of collected particles reached tens of micrometers. Such particles could either fly away with the next plasma pulses or mix with the material of the collecting plate in the course of repetitive plasma pulses. Thin (submicron) layers of re-deposited tungsten also formed on the surface of the collecting plate.

Hydrogen isotope trapping at lattice defects in neutron-irradiated tungsten (W), a leading candidate as plasma facing material, is an important problem determining tritium (T) inventory in a vacuum vessel of a future fusion reactor. In this study, W samples were irradiated with neutrons in the Belgium Reactor 2 at 563 K to 0.06 or 0.016 displacement per atom (dpa). After characterizing defects by positron lifetime measurements, deuterium (D) penetration under exposure to D plasma was examined at 563–773 K. Positron lifetime showed the presence of dislocations, monovacancies and relatively large vacancy clusters. These defects trapped D atoms with different values of binding energy. Dependence of D retention on plasma exposure temperature and damage level indicated that the concentrations of weak traps with smaller binding energy increased more significantly with damage level than those of strong traps.

In DEMO, the divertor must endure steady state loads of 10 MW m⁻² and transient thermal cycling up to 20 MW m⁻². A novel divertor target, termed the Spiral Plate Module (SPM) and designed to meet these loads, was initially optimized by a one-dimensional, steady-state model. The best design was a trade-off between the wall overheat (τ_s)—a figure of merit for cooling performanceand the pumping ratio (η_P)—a comparison of the pressure drop and the incident heat flux on a target surface. A three-dimensional, steady-state, conjugate heat transfer study showed significant correlation to the one-dimensional model. Compared to other divertor target concepts, the SPM achieved the lowest wall overheat of any target design. The hydraulic performance was also on par with comparable designs, demonstrating a very low pumping ratio. This novel, modular divertor target could be used in future fusion power plants to effectively cool the plasma facing components with low pressure drop.

Electron- (ECWC) and ion- (ICWC) cyclotron wall conditioning are essential means for controlled fusion to modify the surface state of plasma-facing components in order to reduce impurity generation and fuel accumulation in the wall. Development of ECWC and ICWC requires characterization of neutral particle fluxes generated in discharges, because neutrals enhance the homogeneity of the conditioning, which may contribute to remote or shadowed areas, especially in the presence of a permanent magnetic field (e.g. W7-X, ITER). A time-of-flight neutral particle analyzer (ToF-NPA) with 4.07 m flight distance is employed to measure time- and energy-resolved low energetic (<1 keV) neutral particle distributions. The ToF-NPA setup tested at the EXTRAP T2R reversed field pinch was installed at the TOMAS toroidal plasma facility to determine low energy neutral particle fluxes while investigating the impact of the gas pressure in the instrument and compatibility with low count rates during EC- and ICWC discharges. TOMAS has a major radius of 0.78 m and provides various plasma operation conditions: toroidal magnetic field up to 0.12 T, EC frequency 2.45 GHz with the power of 0.6–6 kW, IC frequency of 10–50 MHz with the power of up to 6 kW. Early results on the characterization of three phases (EC only, EC + IC, and IC only) of hydrogen discharges demonstrate: (i) the low energy (10–725 eV) neutrals distribution has been determined by the NPA system, (ii) the mixed EC + IC phase produces the highest population of neutral particles, while the EC only provides one order of magnitude lower rate, (iii) the neutrals produced in IC only have higher average energy (28 eV) than EC only (7 eV) and EC + IC (16 eV).

Plasma-facing components of future fusion reactors will have tungsten-based materials as armor. Annealed pure tungsten is brittle at room temperature restricting its use as plasma-facing material, whereas plastically deformed tungsten behaves in general more ductile at ambient temperatures. During operation as plasma-facing material at high temperatures, the deformation structure induced by plastic deformation becomes unstable. Restoration processes as recovery, recrystallization, and grain growth will alter the microstructure and impair the desired mechanical properties. In particular, recrystallization will reinstate the intrinsic brittleness of tungsten. Achieving a thorough understanding of the occurring restoration mechanisms (for long times at temperatures as close to the desired operation temperatures as possible) and quantifying the temperature-dependent recrystallization kinetics are essential for assessing the materials performance and an informed materials selection. The thermal stability of differently rolled pure tungsten plates is reviewed with the aim of predicting the materials lifetime; the impact of different activation energies on the selection of armor materials highlighted. The concept of a recrystallization temperature constituting a threshold temperature below which recrystallization does not occur is dismissed.

In this work the correlation of the neutron radiation damage and tungsten (W) microstructure is investigated. Further, the modification of the structure as a result of the neutron irradiation is assessed and its effect on the mechanical properties is determined. Forged bar (ITER grade), cold rolled sheet, and single crystalline tungsten materials were neutron irradiated at 600 °C to a damage of 0.12 displacements per atom. Neutron irradiation results in the formation of voids of almost the same size (larger than 1 nm) and dislocations detected by positron annihilation lifetime spectroscopy. All W grades have similar total dislocation densities, in the range of (1.6–2.4)×10¹⁴ m⁻², as determined by electrical resistivity measurements. After irradiation the hardness of all tungsten grades increases and the largest increase is that of the single crystal (47%), whereas the smallest that of the sheet (13%). Increase in the yield strength, correlated to the increase of the hardness, is also found. The largest increase is observed for the single crystal (25%) and the smallest for the sheet (6%). The different degrees of hardening and strengthening indicate that the microstructure of the different tungsten grades has a significant influence on their neutron radiation damage resistance.

16 MeV protons have been used to irradiate 300 μm thick macroscopic W samples in a pilot experiment to 0.006 dpa damage dose under low and high temperature scenarios of ∼373 K and ∼1223 K, respectively. The linear pre-Bragg region has been used for damage where the electronic loss (heat) in the sample amounts to 1.5 MW · m⁻². Post high-temperature irradiation, the W sample has been recrystallized as seen under the scanning electron microscope. Indentation measurements on the surface show a softening of 0.6 GPa post-recrystallization against an irradiation hardening of 0.8 GPa for the low-temperature irradiation scenario.

The tritium removal from the vacuum vessel is one of the key issues for the realization of a fusion reactor. As in situ tritium removal techniques from the plasma facing materials, helium glow discharge wall conditioning (GDWC) and RF heated plasma due to electron cyclotron wall conditioning (ECWC) and ion cyclotron wall conditioning (ICWC) are applied with Large Helical Device (LHD) which remains tritium produced by deuterium plasma experiment. As the result, it suggests that ICWC could remove the tritium more efficiently than the ECWC and the GDWC. In the ICWC operation, the high energy particles induced by the minority ion heating mode as the standard ICRF heating operation may play an important role in tritium removal from the plasma-facing materials due to the particle's bombardment.

Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s⁻¹ upto 12 g s⁻¹ depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping.

The MEMOS-U code, a significantly upgraded version of MEMOS-3D, has been developed to address macroscopic metallic melt motion in large-deformation long-displacement regimes, where melts spill onto progressively colder solid surfaces, that are ubiquitous in contemporary tokamaks and expected to be realized in ITER. The modelling of plasma effects, appearing via the free-surface boundary conditions, is discussed along with the sensitivity to external input. The crucial roles of convective and thermionic cooling are exemplified by simulations of ELM-induced tungsten leading edge melting. Key melt characteristics, revealed by previous MEMOS-U modelling of grounded sample exposures, are confirmed in new simulations of the recent floating sample exposures in ASDEX-Upgrade.

The net erosion and deposition patterns in the inner and outer divertor of WEST were determined after different experimental campaigns (C3 and C4) of the first operational phase using ion beam analyses and scanning electron microscopy techniques. The analyses were performed on four entire tiles from inertially cooled, W-coated divertor units with an additional Mo marker coating covered with a further W coating. Strong erosion occurred at the expected location of the inner and outer strike line area with a campaign-averaged net erosion rate of >0.1 nm s⁻¹. On the high field side of the inner strike line area, thick deposited layers were found (>10 μm; growth rate >1 nm s⁻¹), mainly composed of B, C, O, and W. Additionally, strong arcing was observed in this region. At the end of the C4 campaign, He discharges were performed to study the He-W interaction. Although the conditions for nanotendrils, i.e. fuzz formation were fulfilled around the outer strike line position, neither nanotendrils nor He bubbles (>10 nm) were observed at this area.

We report on the detection by means of Raman spectroscopy of amorphous beryllium deuteride, BeD₂, in magnetron sputtered deposits synthesized in two different laboratories and containing about 20 at% of deuterium. In contrast, this signature has not been found for the JET limiter samples studied coming from the inner, outer or upper limiters, even when coming from a deposition zone of the limiters. We give a way to disentangle that BeD₂ signature from other signatures falling in the same spectroscopic range and mainly related to other phenomena. We also analyze the Raman characteristics of the JET sample defects. These results could help in the interpretation of D thermal desorption spectra and in future analyses of JET thick Be deposit divertor tiles.

Tungsten erosion was quantified during inter/intra-ELM periods in He-dominated JET-ILW plasmas by optical emission spectroscopy. The intra-ELM tungsten sputtering in helium plasmas, which dominates the total W source, prevails by a factor of about 4 over inter-ELM sputtering in the investigated ELM frequency range from 90 Hz–120 Hz. He ions are mainly responsible for the W erosion during the ELMs in He plasmas. The strong in/out asymmetry of the ELM-induced W erosion is observed in He plasmas even at high ELM frequencies beyond 100 Hz. In Ohmic/L-mode plasmas and during the H-mode inter-ELM plasma phases both He²⁺ and Be²⁺ ionic species are major contributors to the W erosion. Their contribution depends on the electron temperature in the divertor: for T_e > 15 eV both species cause significant W sputtering, for T_e < 15 eV, B_e²⁺ ions are solely responsible for the W erosion. Tungsten erosion during in both inter and intra-ELM periods in He-dominated plasmas are significantly larger than in deuterium plasmas. It is 15–25 times larger during the inter-ELM phase and in L-mode discharges at T_e = 25–30 eV. On the other hand, the ELM-induced W source is by a factor of 3 larger than in D plasmas.

The development of advanced tungsten grades able to tolerate irradiation damage combined with thermo-mechanical loads is important for design of plasma-facing components for DEMO. The material microstructure (i.e. grain size, dislocation density, sub grains, texture) is defined by manufacturing and post heat treatment processes. In turn, the initial microstructure might have an important influence on the accumulation of neutron damage because irradiation defects interact with microstructural defects evolving into a new microstructural state. In this work, the microstructure and hardness of four tungsten grades is assessed before and after neutron irradiation performed at 600, 1000 and 1200 °C, up to a dose of ∼1.2 dpa. Experimental characterization involves hardness testing, energy dispersive spectroscopy, electron backscatter diffraction, and transmission electron microscopy. The investigated grades include Plansee and AT&M ITER specification tungsten, as well as fine grain tungsten produced by spark plasma sintering, and ultra-fine grain tungsten reinforced with 0.5 wt% ZrC particles.

We demonstrate a combinatorial approach integrating ion implantation followed by thermal annealing and simultaneous in situ ion beam analysis with thermal desorption spectroscopy in a single set-up. Atomic and molecular deuterium ions of 3 keV were implanted into bulk tungsten with doses exceeding 1 × 10²² ions m⁻². Depth profiling of both, protium and deuterium was performed by elastic recoil detection analysis, while simultaneously the outgassing rates of molecular deuterium by thermal desorption spectroscopy were monitored during temperature ramps from room temperature to ≈1400 K. The combination of the two techniques in situ is shown capable to identify the distinct retention behavior of deuterium at different temperatures and in different reservoirs, e.g. located close to the surface and diffused deep into the material. Ex-situ scanning electron microscopy confirmed blister formation, and recovery of the initial surface morphology after high temperature annealing, in analogy to comprehensive ex-situ studies.

The effect of grain boundary (GB) direction with respect to the exposed surface on blistering in tungsten (W) is studied. Four types of W materials were used in this work and classified according to GB directions: one with GBs parallel to the exposed surface (parallel GBs) and the other with GBs perpendicular to the exposed surface (perpendicular GBs). After exposure to high flux deuterium plasmas, blisters were observed and found to be related to GBs in all types of materials. It is found that large-size blisters were originated from parallel GBs rather than perpendicular GBs. Blisters were supposed to grow faster in parallel GBs than in perpendicular GBs. Moreover, the exfoliation of blister caps was observed in the materials with parallel GBs but not in that with perpendicular GBs. This study provides a new insight into the role of GB direction on blistering.

Tungsten–tantalum (W-Ta) alloys for the application to fusion reactor divertor were developed to overcome the issues of W related to the low temperature brittleness, embrittlement by recrystallization, and embrittlement by neutron irradiation. In the present study, the response to cyclic thermal shock tests as well as the fundamental mechanical properties of W-3%Ta and pure W were investigated, which simulated a transient heat load event in the actual divertor environment. Because it was concerned that the effect of deuterium (D) plasma exposure might be much more pronounced in the Ta-alloyed materials, the thermal shock tests were performed under the background steady state D-plasma exposure. Based on the present study, W-3%Ta might be a competitive alloy system from the viewpoint of thermo-mechanical response in the actual divertor environment as well as the fundamental mechanical properties and resistance to recrystallization.

We have investigated the applicability of Laser Induced Breakdown Spectroscopy (LIBS) for analyzing the changes in the composition and fuel retention of W and W-Ta coatings following exposure to D₂ or mixed D₂-N₂ plasma beams in the linear plasma device Magnum PSI. The exposed samples were characterized by in situ ns-LIBS and complementary analysis methods Secondary Ion Mass Spectroscopy, Energy Dispersive x-ray spectroscopy and Nuclear Reaction Analysis. In agreement with the used complementary analysis methods, LIBS revealed the formation of up to 400 nm thick co-deposited surface layer in the central region of the coatings which contained a higher concentration of the main plasma impurities, such as N, and metals, such as Ta and Mo, the latter originating mainly from the substrate and from the plasma source. The deuterium retention on the other hand was highest outside from the central region of the coatings.

Recent experiments have revealed that accelerated tungsten (W) nanostructure growth occurs during the exposure to helium (He) plasmas with an additional W deposition, and mm-thick large-scale fiberform nanostructures (LFNs) are formed when certain condition is satisfied. In order to reveal whether accelerated growth occurs in fusion device, it is important to understand the annealing process in addition to the growth process. In this study, we will perform thermal treatment of W LFNs. Detailed observation reveals that the LFNs shrink at a faster rate than that of conventional fuzz, though it takes longer time to fully be reintegrated to the surface because the thickness is orders of magnitude greater.