For future fusion devices, tungsten is the main candidate materials for the application as plasma facing materials (PFMs). However, considering the challenging operational condition with high thermal loading/thermal stress combining plasma exposure and neutron irradiation/embrittlement, one of the major concern for tungsten as PFMs is its intrinsic brittleness. To avoid cracking and components failure, toughening tungsten is widely investigated, among which tungsten fiber reinforced tungsten composites (Wf/W) are developed using an extrinsic toughening mechanism. Recently, a new type of aligned long fiber Wf/W (L-Wf/W) with dedicated weak interface have been prepared by powder metallurgy process, combing the advantages of superb damage resilience with a much easier production compared to conventional chemical vapor deposition process. In this work, the newly developed material is characterized, including, mechanical tests, high heat flux tests, exposure to plasma for erosion and fuel retention tests. The l-Wf/W composite could improve significantly the damage resilience compared to pure W without altering much of other properties.

Hydrogen embrittlement (HE) has persisted for more than a century as one of the most intractable problems in materials science. The prevailing view1 that diffusive H governs embrittlement has fostered the widespread assumption that H trapping at crystal defects mitigates HE. Here we overturn this conventional paradigm. Using plasma/ion irradiation of tungsten, we decouple – for the first time – H-induced crack nucleation from subsequent cavity propagation, and reveal nucleation as a two-stage mechanochemical fracture instability enabled by trapped H in the absence of diffusive H. In the first stage, H accumulation to a critical occupancy at dislocation cores acts as a chemical fuse, collapsing the local cohesive strength to a threshold at which infinitesimal external loads can trigger atomic decohesion. This bond rupture instantaneously enables the second stage: confined recombination of atomic hydrogen into molecular form. The abrupt release of chemical energy within an atomically restricted volume generates a transient inflation pressure that drives a dynamic, brittle jump to an internal macroscopic cavity. By separating mechanical decohesion triggering from energetic crack driving, our results provide a deterministic framework for the onset of H-induced crack nucleation under low-stress conditions. Furthermore, we place experimentally the classical H-enhanced decohesion model on an atomistic foundation and elevate it from phenomenology to prediction. Finally, by shifting the focus from experimentally elusive diffusive H to directly measurable trapped H, this work reframes HE as a deterministic, quantifiable instability, establishing a new paradigm for understanding and mitigating H-induced failure in high-strength metals.

Self-passivating metal alloys with reduced thermo-oxidation (SMART) have been considered as promising candidates for the first wall armor in fusion plants due to their ability to suppress the sublimation of radioactive substance during loss-of-coolant accidents (LOCAs). However, conventional manufacturing of SMART via fieldassisted sintering technology (FAST) typically introduces Cr-rich phases, which are regarded harmful for both oxidation resistance as well as tolerance to high heat peak loads. In this paper, the influence of FAST sintering temperature on the densification behavior, microstructures and hardness of W-11.4 wt% Cr alloys is studied to suppress the Cr-rich phases. The Cr-rich phases are found to precipitate when sintered at 1400 degrees C and could be suppressed within the temperature range of 1500-1600 degrees C. However, they precipitate again when sintering at 1700 degrees C due to the localized liquidation of Cr at the grain boundaries. In addition, pores are formed at the grain boundary junction when sintered above 1600 degrees C, resulting in the reduction of the sample density. Moreover, with the increase of sintering temperature, the grains grow from 0.30 mu m at 1400 degrees C to 2.56 mu m at 1700 degrees C, and the Vickers hardness decreases from 1086 at 1400 degrees C to 880 at 1700 degrees C. Based on that, the activation energy of the grain growth, Q, as well as the parameters of Hall-Petch relationship, sigma 0 and k, are fitted to be 380 kJ/mol, 2402 MPa and 473 MPa & sdot;mu m1/2, respectively. The results could serve for further optimization of microstructures and mechanical properties of SMART materials.

Atomic displacement – the fundamental process underlying diverse deformation and damage phenomena in metals, from irradiation defect production to stress-driven dislocation motion – is governed by interatomic cohesion strength. Here, lattice-dissolved hydrogen (LDH) occurring in metals under direct hydrogen exposure is identified to effectively weaken lattice cohesion, and thereby facilitating atomic displacement and dislocation movement upon plastic deformation in sub-threshold stress regime. This atomic-scale insight provides a physically transparent mechanism for hydrogen-enhanced localized plasticity implicated in hydrogen embrittlement. We quantitatively verify the hydrogen-induced lattice cohesion weakening effect on metal surfaces exposed to low-energy hydrogen plasma, where massive defects are generated despite the absence of sufficient ion momentum for direct displacement damage. By unprecedentedly quantifying the cohesion-weakening effect of LDH independently from defect-trapped H, we establish a new paradigm to understand hydrogen embrittlement.

Tungsten-based self-passivating metal alloys with reduced thermo-oxidation (SMART), which using Cr or Si as oxidation-resistant elements, have exhibited great potential for oxidation resistance application at high temperature compared with pure W. In this work, aluminum (Al) was added to W-Cr SMART systems to further improve the oxidation resistance, and the novel W-17.8Cr-6.4Al alloys were successfully developed using spark plasma sintering (SPS) technique. The influence of sintering pressure and temperature on the densification process was systematically investigated, and the oxidation behaviors were studied at 800 degrees C and 1000 degrees C. The sintering curves exhibit that there is the liquidation of Al above around 630 degrees C, and the rapid densification of WCr-Al alloys starts above 923-950 degrees C. The intermetallic compounds of (WCr)Al12 and (WCr)Al4 are generated when sintered at 650 degrees C and 700 degrees C, respectively. When sintered above 1000 degrees C, W-Cr-Al alloys mainly consist of two BCC phases of W-Cr-Al solution. Dense bulk W-Cr-Al alloys, with a density of around 11.6 g/cm3, could be manufactured by sintering above 1200 degrees C at 20 MPa for 15 min. At the end of oxidation, the oxides formed at 800 degrees C mainly consist of WO3, Cr2WO6 and Al2W3O12, while WO3 disappears at 1000 degrees C due to the volatilization of WO3 and the reaction with Cr2O3. In addition, no obvious variation trend of mass gain is observed when oxidized at 800 degrees C compared with pure W, while a double parabolic trend of mass gain is found during oxidization at 1000 degrees C. The parabolic oxidation parameter, kp, is around (7-8) x 10-5 mg2/(cm4 center dot min) and (4-5) x 10-4 mg2/(cm4 center dot min), respectively. Compared to the linear oxidation behavior of pure W, W-Cr-Al alloys exhibit its potential for oxidation resistance application at high temperature.

Plasma-facing materials (PFMs) for future fusion reactors require advanced mechanical and thermal properties to withstand the extreme challenges of high heat flux, plasma exposure, and neutron irradiation. Tungsten is one of the most suitable materials for use as a PFM in the divertor region. However, considering the high thermal loading/thermal stress combining plasma exposure and neutron irradiation/embrittlement, one of the major concerns for tungsten in PFMs is its intrinsic brittleness. To avoid cracking and components failure, tungsten toughening has been widely investigated, including the development of tungsten fiber-reinforced tungsten composites (Wf/W) using an extrinsic toughening mechanism, which could provide damage resilience against neutron embrittlement. Recently, a type of aligned long-fiber Wf/W (L-Wf/W) based on a powder metallurgical fabrication process was developed, demonstrating advanced fracture toughness while retaining other application-relevant properties. For L-Wf/W, the relatively easy production process suggests the feasibility and basis of industrialization. This work reports on the initial progress in industrializing L-Wf/W, with a focus on adapting the lab sintering process to a sintering process with industrial partner (Dr. Fritsch Sondermaschinen GmbH) and optimizing the process parameters. To improve the sinterability of tungsten and achieve higher density, various tungsten powders were explored, including commercial W powders, bimodal mixtures of different particle sizes, and granulated W powders. At the dedicated yttria interface, the thickness of yttria coating on the fibers was also optimized to ensure effective separation between the fibers and the matrix. Series of samples were produced with different dimensions up to 100 mm × 100 mm × 4 mm. After optimization, samples with 93% density and desired pseudo-ductility were prepared. Similarly to production in the lab, a major challenge in this work involved balancing the densification of the tungsten matrix with controlling fiber recrystallization and mitigating damage to the yttria interface.

Powder injection molding (PIM) has been used to produce nearly net-shaped samples of tungsten-based alloys. These alloys have been previously shown to have favorable characteristics when compared with standard ITER-grade tungsten. Six different alloys were produced with this method: W-1TiC, W-2Y2O3, W-3Re-1TiC, W-3Re-2Y2O3, W-1HfC and W-1La2O3-1TiC. These were tested alongside ITER-grade tungsten in the PSI-2 linear plasma device under ITER-relevant plasma and heat loads to assess their suitability for use in a fusion reactor. All materials showed good behavior when exposed to the lower pulse number tests (≤1000 ELM-like pulses), although standard tungsten performed slightly better, with no observable difference in surface roughness. High-power shots, namely one laser pulse of 1.6 GWm−2, revealed that samples containing yttria are more prone to melting and droplet ejection. After high pulse number tests (10,000 and 100,000 pulses), with and without plasma, the reference tungsten showed the most cracking and highest surface roughness of all materials, while the PIM samples seemed to have a higher resistance to cracking. This can be attributed to the higher ductility of these alloys, particularly those containing rhenium. This means that tungsten-based alloys, whether produced via PIM or other methods, could potentially be used in certain areas of a fusion reactor.

The severe environment and loads acting on plasma-facing components (PFCs) of future fusion power plants cause inevitable erosion of their armor. In situ regeneration of tungsten (W) armored PFCs by local deposition of material would open up the possibility of damage healing and compensation of eroded material. The wire-based laser metal deposition (LMD-w) process fulfils the necessary requirements for use in the reactor vessel. Process development for the deposition of W on W substrate has already been carried out and it has been proven that thermal induced damage in the PFM can be healed this way. In this study, W armored PFCs were coated using LMD-w and tested under fusion-relevant thermal loads in the electron beam facility JUDITH 2. One respectively two stacked layers, each similar to0.65 mm in height, were applied on the top surfaces of the W double tiles with surface areas of 28 x 12 mm(2) respectively, which are the characteristic dimensions for the plasma-facing surface of monoblocks. Some of the coated surfaces were also smoothed by laser remelting. In the electron beam facility JUDITH 2, the test components were exposed to steady state as well as combined steady state and transient thermal loads that are expected in the divertor area of the future DEMOnstration power plant. The coatings were tested with cyclic (200 and 1000 cycles) steady state thermal loading in the form of surface temperatures equivalent to heat fluxes on monoblock components of 10 MW m(-2) (similar to 1000 degrees C) and 15 MW m(-2) (similar to 1500 degrees C). To determine the performance of LMD-w layers under thermal loads that are expected during exposure to edged localized modes (ELMs), some layers were subjected to combined steady state and transient loading scenarios (0.13-0.55 GW m(-2), 10(3) to 10(5) pulses of 0.48 ms, 200-700 degrees C base-temperature). The temperature data obtained from the HHF experiments was processed and analyzed. Profile measurements on the coated surfaces before and after the high heat flux (HHF) exposure were used to investigate the influence of thermal stress on the deposited layers. Furthermore, cross-sectional micrographs of the test components were prepared and analyzed.

Tungsten (W) is a promising candidate material for the plasma-facing components in future fusion reactors. However, it has issues regarding the intrinsic brittleness as well as operational embrittlement. Tungsten fiber reinforced tungsten (Wf/W) composites overcome these issues by using extrinsic toughening mechanisms. Due to neutron irradiation and high-temperature recrystallization during fusion operation, the ductility of the fibers may degrade over a long period of service. Therefore, it is necessary to investigate the fracture behavior of Wf/W composites with brittle fibers. In the present work, the tungsten fibers were embrittled by carbonization. Subsequently, the continuous brittle Wf/W composite with yttrium oxide (Y2O3) interface was fabricated by the field-assisted sintering technology process. The microstructure of the prepared Wf/W composites was characterized. The fracture behavior and toughening mechanisms were discussed in detail based on the experimental results of 3-point bending tests and the corresponding finite element simulation. The composites show a pseudo-ductile fracture behavior. Cracks are hindered and deflected by the de-bonded fiber-matrix interface. The extrinsic toughening mechanisms of interface de-bonding, crack bridging, and fiber pull-out are active. This indicates the reinforcement concept still works even though all components in the Wf/W are brittle.

Successful upscaling of tungsten fiber-reinforced tungsten composites (Wf/W) on industrial level could represent an important milestone for future nuclear fusion reactors. The primary objective of these materials is to enhance the durability and operational lifespans of critical components. Developing mature manufacturing approaches remains a challenge, highlighting the need for innovative solutions. This study evaluates the feasibility of merging chemical vapor deposition (CVD) with spark plasma sintering (SPS) for producing such composites. This analysis indicates that combining CVD-W sealed tungsten fabrics with SPS requires additional manufacturing steps or the utilization of tungsten powders for effective sintering. The process is currently only suitable for simple textile structures utilizing single filaments, mitigating one of the main advantages of CVD. Configurations such as radially braided yarns are currently less compatible to the high stress levels during SPS. A key outcome of this work is the introduction of a thin secondary CVD-W interface into the composite design, substantially improving the stability of the yttria-interface and effectively shielding the W-fibers from potential matrix interactions. This innovation reduces issues such as carbon embrittlement and allows the potential integration of tungsten fibers into different matrix materials such as ceramics, broadening the potential application range of tungsten fiber-reinforcements.

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

Tungsten (W) is a promising candidate material for the plasma facing components in fusion reactors. However, it has issues regarding the intrinsic brittleness. Tungsten fiber reinforced tungsten composites (Wf/W) have been developed based on the concept of extrinsic toughening mechanisms and they show a pseudo-ductile behavior during the fracture process. In the present work, continuous fiber reinforced Wf/Y2O3/W composites were fabricated via a powder metallurgy (PM) process, and the microstructure and mechanical properties were characterized. The fracture behavior and toughening mechanisms were analyzed in detail combining the results of experiments and numerical simulation. The Wf/Y2O3/W composites is toughened by multiple mechanisms such as fiber bridging, crack bending and deflection, interface de-bonding and plastic deformation of fiber. The energy dissipation by interface de-bonding can be neglected. However, it is a necessary factor to ensure any extrinsic toughening mechanisms. The main contribution of the energy dissipation while composite failure is the plastic deformation of fibers.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Plasma-facing components (PFC) in nuclear fusion reactors are exposed to demanding conditions during operation. The combination of thermal loads, plasma exposure as well as neutron induced damage and activation limits the number of materials suitable for this application. Due to its properties, tungsten (W) is foreseen as plasma-facing material (PFM) for the future DEMOnstration power plant. It is considered suitable due to its exceptionally high melting point, excellent thermal conductivity, low tritium retention and low erosion resistance during plasma exposure. But even tungsten armored PFCs have a limited lifetime due to, among other factors, surface erosion and the resulting thickness reduction of the armor material.In-situ local deposition of tungsten by means of additive manufacturing (AM) could counteract surface erosion and thus increase the service life span of PFCs. After evaluation of the potential AM processes qualified for this task, the wire-based laser metal deposition (LMD-w) process was selected as the most suitable process. First trials were conducted to examine if it is possible to reliably deposit tungsten onto tungsten substrate using the LMD-w process. In these first studies, single welding beads were generated, and in later experiments, entire layers were created from several welding beads which are arranged next to each other. To ensure reproducibility of the results, the substrate temperature was kept constant. Further experiments aimed at the elimination or minimization of problems such as oxidation, occurrence of balling defects, porosity, cracking, surface waviness and insufficient connection to the substrate. To increase the welding bead quality, the input parameters like laser power, deposition velocity, wire feed rate, inert gas flow, as well as the wire position were optimized. Furthermore, stacking of several layers, as well as the remelting of an already created layer, were carried out and investigated. This study represents the first steps in testing the feasibility of an in-situ surface regeneration concept for PFCs.

Net tungsten erosion and deposition profiles are simulated for the outer and inner lower divertor of STEP, a spherical prototype fusion reactor targeting similar to 2040. In this contribution, previous modelling work [1] has been extended by studying the influence of various input parameters. Main aim is to analyse the influence of several modelling improvements on the erosion/deposition characteristics at the divertor targets. This comprises the consideration of an electron density decay according to the Boltzmann relation within the sheath region as well as a certain tungsten influx to the divertor originating from main wall erosion. Also, for the inner divertor improved background plasma parameters are applied taking into account a gradient along the flux surfaces compared to formerly constant electron temperature and density and ion temperature along the flux surfaces. ERO simulations have been performed for one selected plasma case with Ar seeding, both for the outer (with peak target T-e similar to 25 eV) and inner lower divertor (with peak target T-e similar to 3 eV). The simulated tungsten erosion and redeposition profiles do not significantly change for the cases studied if the Boltzmann-related decay of electron density within the sheath is considered. The assumption of tungsten within the background plasma can notably alter the overall erosion/redeposition behaviour at a background tungsten concentration of 1/10 with respect to the Ar divertor concentration. The inner divertor target plate shows net deposition everywhere, while at the outer one the net erosion zone becomes smaller and the maximum net erosion peak reduces by a factor of two. Lastly, the improved background plasma for the inner divertor has moderate effects, for instance the overall tungsten redeposition at the target plate increases from 88.7% to 94.4%.

Due to its unique properties tungsten is a promising candidate as plasma-facing-material (PFM) in future nuclear fusion reactors. Tungsten features an exceptionally high melting point, high thermal conductivity, low tritium inventory and comparatively low erosion rate under plasma loading [1]. But given the extreme loads on the PFM during operation of a fusion reactor, the lifetime of plasma-facing components (PFC)s is limited. Currently, it is planned to replace damaged PFCs when they reach the end of their service life. However, the lifetime of PFCs could be increased by in situ repair using additive manufacturing technology (AM) in the form of direct-energydeposition (DED). The wire-based laser metal deposition process (LMD-w) meets several necessary conditions for operation in the vessel and could be used for performing such in situ repairs. It was investigated if the LMD-w process is able to heal thermal induced surface cracks and roughening by remelting the substrate during deposition of tungsten. For this purpose, tungsten samples of 12 x 12 x 5 mm3, which later served as substrate plates for the LMD-w experiments, were treated with combined steady-state and transient thermal loads in the electron beam facility JUDITH 2. These samples were brazed to a copper cooling structure and exposed to 105 thermal shocks of 0.5 ms duration and an intensity of Labs = 0.55 GW m- 2 (FHF = 12 MW s0.5 m- 2) at a base temperature of Tbase = 700 degrees C. This way, edge localized mode (ELM) like thermal load damage was induced on the tungsten samples. On these samples, different LMD-w and laser remelting process strategies were performed. Subsequently, these samples were analyzed, and it was examined that the healing of the pre-damaged substrate material was successful. In parallel, the laser remelting process was modeled in a thermal transient finite element method (FEM) simulation in order to gain an insight into the temperatures prevailing in the material during the process.

The self-passivating yttrium-containing WCr alloy has been developed and researched as a potential plasma-facing armour material for fusion power plants. This study explores the use of yttria (Y2O3) powders instead of yttrium elemental powders in the mechanical alloying process to assess their applicability for this material. Fabricated through field-assisted sintering, WCr-Y2O3 ingots show Y2O3 and Cr-containing oxides (Cr-O and Y-Cr-O) dispersed at grain boundaries (GBs), while WCrY ingots contain Y-O particles at grain boundaries, both resulting from unavoidable oxidation during fabrication. WCr-Y2O3 demonstrates higher flexural strength than WCrY across all temperature ranges, ranging from 850 to 1050 MPa, but lower fracture toughness, between 3 and 4 MPa·√m. Enhanced oxidation resistance is observed in WCr-Y2O3, with lower mass gain as compared to WCrY during the 20-hour oxidation test. This study confirms the effectiveness of both yttria and yttrium in the reactive element effect (REE) for the passivation of WCr alloy, suggesting the potential of Y2O3-doped WCr for first wall applications in a fusion power plant.

Analysis of comparable discharges fuelled by either deuterium or protium reveals a clear relationship between the isotope mass and the H-mode density limit. Notably, the density limit is significantly lower in protium, showing a reduction of up to 35 % compared to identical deuterium plasma conditions. Within mixed H-mode density limit (HDL) plasmas, the maximum achievable density, or H-mode density limit, decreases with increasing protium concentration, denoted as cH. For instance, the highest corresponding maximum Greenwald fraction (fGW) of about 1.02 was observed in the pulse with the lowest cH value of 4.4 %. This fGW decreases to 0.96 at cH = 48 %. The average atomic mass, A, of the plasma species decreases in these pulses from the value of 1.96 (cH = 4.4%) down to 1.52 (cH = 48 %). Interestingly, the maximum achievable density appears to be largely unaffected by the applied power value, regardless of whether deuterium or protium is used, as well as under mixed H/D fuelling conditions. Additionally, the measured Greenwald fractions are agreed with a heuristic model based on the SOL pressure threshold of an MHD instability, as proposed by Goldston. This comparison, especially concerning the model's dependence on isotopic mass, shows full consistency between the measured and predicted Greenwald fractions.

The paper provides an overview of recent modelling of global material erosion and deposition in the fusion devices Wendelstein 7-X (W7-X), JET and ITER using the Monte-Carlo code ERO2.0. For validating the modelling tool in a three-dimensional environment, W7-X simulations are performed to describe carbon erosion from the graphite test divertor units, which were equipped in operational phase OP 1.2 and analysed post-mortem. Synthetic spectroscopy of carbon line emission is compared with experimental results from the divertor spectrometer measurement system, showing a good agreement in the e-folding lengths in the radial intensity profiles of carbon. In the case of metallic wall materials, earlier modelling of the Be/W environment in JET and ITER is revisited and extended with an updated set of sputtering and reflection data, as well as including the mixing model for describing the Be/W dynamics in the divertor. Motivated by recent H/D/T isotope experiments in JET, limited and diverted configuration pulses are modelled, showing the expected trend of both Be and W erosion increasing with isotope mass. For the JET diverted configuration pulses, it is shown that Be migrates predominantly to the upper part of the inner divertor where it initially leads to strong W erosion. With longer exposure time, the growth of a Be deposited layer leads to a reduction of W erosion in that region. A similar trend is observed in simulations of the ITER baseline Q = 10 scenario, however with a more symmetric Be migration pattern leading to deposition also on the outer divertor.