Enhanced deuterium retention in tantalum (Ta) cold spray coatings, compared to reference polycrystalline tantalum and tungsten materials, has been evaluated using the thermal desorption spectrometry technique. Tantalum coatings, deposited via cold spray technology on 316L stainless steel substrates, are proposed as plasma-facing material surfaces with hydrogen gettering functionality for advanced fusion concepts. The materials were exposed to 95 eV D ions at a flux of 1.6- 3.5x1021 D m-2 s-1. Retention was measured as a function of incident ion fluence and surface temperature. The results highlight an increased deuterium inventory in Ta cold spray coatings by a factor of 3.5 compared to polycrystalline tantalum and by two orders of magnitude compared to polycrystalline tungsten. A tendency for retention saturation in tantalum is observed at a fluence above 1x1024 D m-2. While deuterium retention gradually decreases with increasing surface temperature from 400 K to 925 K for polycrystalline tungsten, it remains constant for polycrystalline tantalum. In contrast, retention in Ta coatings significantly decreases when the surface temperature exceeds 750 K. The microstructure of the cold spray Ta coatings plays a crucial role in the dynamics of deuterium trapping and release. Tantalum also exhibits a superior resistance to blister formation compared to tungsten when subjected to a high dose of deuterium.

Tungsten (W) is a common plasma-facing material in nuclear fusion devices. It readily oxidizes in the presence of oxygen, forming tungsten oxides, particularly WO3, which may modify deuterium retention. This experimental study investigates mechanisms of the evolution of deuterium retention in a thermally stable similar to 100 nm WO3 layer grown by thermal oxidation of a W substrate at 1073 K under low pure oxygen pressure of 7 Pa. 500 eV D-2(+) deuterium implantation and Temperature Programmed Desorption (TPD) were used to explore deuterium trapping and release from the oxide as a function of incident ion fluence and storage time in ultra high vacuum. Upon deuterium irradiation, the formation of a W-rich layer on the surface of the oxide is evidenced with x-ray photoelectron spectroscopy. After successive implantation/TPD cycles corresponding to an accumulated deuterium fluence >10(21) D m(-2), the appearance of an amorphous oxide encapsulated between two W-rich layers is observed with transmission electron microscopy. Following this high fluence deuterium implantation, deuterium retention increases by a factor of 10. The findings provide insight into the behavior of tungsten oxides under deuterium implantation, emphasizing the importance of considering its thermal stability and its structural modifications at the surface of tungsten plasma facing components.

Understanding the physical–chemical processes ruling the interaction of particles (atoms, molecules, and ions) with surfaces is fundamental in several research fields, such as heterogeneous catalysis, astrochemistry, and nuclear fusion. In particular, the interaction of hydrogen isotopes with plasma facing materials represents a high-priority research task in the fusion community. Such studies are essential to ensure the successful operation of experimental fusion reactors, such as the tokamak ITER. In this work, we present a surface science apparatus developed to study ion-surface interaction in fusion relevant systems. It combines laser-based techniques with contaminant-free ion/molecular beams, mass spectrometry, and surface science tools such as low-energy electron diffraction and Auger electron spectroscopy. It allows to cover a wide range of sample temperatures, from 130 to 2300 K, by changing the heating rate of samples from 0.1 to 135 K/s and maintaining the linearity of the heating ramps, a powerful feature to gain insight on adsorption, absorption, and desorption mechanisms. Experimental calibration and performance are presented in detail. Moreover, to provide a factual overview of the experimental capabilities, we focus on two different applications: the protocol used to clean a W(110) single crystal sample and the development of laser temperature programmed desorption to study helium retention in tungsten.

Rate equation modelling is performed to simulate D2 and D2+D2+ exposure of the W(110) surface with varying coverage of oxygen atoms (O) from the clean surface up to 0.75 monolayer of O. Density Functional Theory (DFT) calculated energetics are used as inputs for the surface processes and desorption energies are optimized to best reproduce the Thermal Desorption Spectrometry (TDS) experiments obtained for D2 exposure. For the clean surface, the optimized desorption energies (1.10 eV-1.40 eV) are below the DFT ones (1.30 eV-1.50 eV). For the O covered surface, the main desorption peak is reproduced with desorption energies of 1.10 eV and 1.00 eV for 0.50 and 0.75 monolayer of O respectively. This is slightly higher than the DFT predicted desorption energies. In order to simulate satisfactorily the total retention obtained experimentally for D2+D2+ exposure, a sputtering process needs to be added to the model, describing the sputtering of adsorbed species (D atoms) by the incident D ions. The impact of the sputtering process on the shape of the TDS spectra, on the total retention and on the recycling of D from the wall is discussed. In order to better characterize the sputtering process, especially its products and yields, atomistic calculations such as molecular dynamics are suggested as a next step for this study.

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.

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).

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He + H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.

Future fusion reactors like ITER and DEMO will have all-tungsten (W) walls and long pulses. These features will make wall conditioning more difficult than in most of the existing devices. The W Environment Steady-state Tokamak (WEST) is one of the few long pulse (364 s) fusion devices with actively cooled W plasma-facing components in the world. WEST is a unique test bed to study impurity migration and plasma density control via reactor relevant wall conditioning techniques. The phase II of WEST operations began in 2022, after the installation of a new lower divertor, now entirely equipped with actively cooled, ITER grade, W monoblocks. After pump down, we baked WEST between 90 degrees C and 170 degrees C for similar to 2 weeks. After 82.5 h at 90 degrees C and 33 h at 170 degrees C, vacuum conditions were stable with a vessel pressure of 6x10(-5) Pa and mass spectra dominated by H-2 molecules. While at 170 degrees C, we performed similar to 40 h of D-2 glow discharge cleaning (GDC) and similar to 5 h of glow discharge boronization (GDB), using a 15 %-85 % B2D6-He mix and a total boron mass of similar to 12 g. This was the very first GDB at such high temperature for WEST. The whole wall conditioning sequence led to a similar to 10 times reduction of the H2O signal as well as to a similar to 3 times reduction of the O-2 signal, according to mass spectra. Once back to 70 degrees C, the vessel pressure was 5.5x10(-6) Pa and plasma restart was seamless with similar to 30 s cumulated over the very first 5 pulses and an Ohmic radiated power fraction F-rad = 0.6, showing successful conditioning of the new ITER grade divertor. The effect of the first, 'hot' GDB faded with a characteristic cumulative injected energy of 2.45 GJ and saturation towards F-rad similar to 0.8. After 1.4 h and 7.5 GJ of cumulative plasma time and injected energy, we carried out a second GDB, this time at 70 degrees C. This 'cold' GDB initially led to a much lower Ohmic F-rad = 0.3-0.4 but the effect lasted similar to 7 times less, with a characteristic cumulative injected energy of 0.37 GJ. At the end of the campaign, we cumulated similar to 3h and similar to 30 GJ through repetitive, minute long pulses without any boronization. Throughout this 4-weeks-long experiment, F-rad in the 4 MW heating phase evolved only marginally (from 0.5 to 0.55). This increase is mostly due to the build-up of re/co-deposited layers on both lower divertor targets.

We present a systematic study that quantifies deuterium (D) retention and ammonia (ND3) production from 316 L stainless steel (SS316L) following the implantation of D ions in conditions similar to the ones expected in the ITER tokamak, i.e. with kinetic energy below 300 eV. Using Temperature Programmed Desorption (TPD) after deuterium ion implantation at 250 eV/D, we show that deuterium retention increases linearly with the D fluence up to 1021 D+m-2, with a retention probability of 18%. For higher D fluence, deuterium retention increases sub-linearly. Analysis of the TPD spectra evolution with varying storage time in vacuum after D implantation, shows that D retention is influenced by D diffusion into the bulk of SS316L. Subsequent to D ion implantation, we evidence the efficient production of ND3 molecules during TPD, between 400 K and 750 K, from the nitrogen present naturally in SS316L. Up to 21% of the D release during TPD can be found in ND3 molecules, indeed. The fraction of ND3 in the total D release depends both on the D ion fluence and the nitrogen concentration profile in the bulk. At least 7% of the D release is found in the form of ND3 molecules, even at a fluence of 2 x 1021 D+m-2 and for a natural N concentration bulk profile. Both N diffusion and D diffusion into the bulk appear to dictate the kinetics of ND3 production. Our findings of efficient production of ND3 in D-implanted austenitic 316 L stainless steel underline the need for similar studies on reduced-activation ferritic/martensitic (RAFM) steels that contain similar content of nitrogen and will be used in fusion reactor prototypes.

In this work, we investigate the optical properties variation of thermally grown W oxides during the exposure to low energy deuterium (D) plasma. In-situ ellipsometry in the 400–1000 nm range was coupled to ex-situ diagnostics, such as X-ray photoelectron spectroscopy (XPS) and focused ion beam coupled to scanning electron microscopy (FIB-SEM), to probe the evolution of chemical and morphological properties of the W oxides. First, a 70 nm thick WO3 layer was exposed to D plasma at a surface temperature of 650 K. An important reduction of the oxide layer was observed by FIB-SEM, and in-situ ellipsometry showed the evolution of the optical constants n and k of the oxide towards the ones of pure W metal. Secondly, a 300 nm thick WO3 layer was exposed to D plasma at a surface temperature below 373 K. In order to follow the oxide evolution step by step, we alternated in-situ ellipsometry and XPS surface characterization. A quite similar evolution of the optical constants was observed, in particular an increase of the extinction coefficient k in the near infrared, which was linked to a progressive reduction of the oxidation level at the surface, as seen by XPS. Interestingly, the reduction of the oxide at 373 K was below the resolution of the FIB-SEM. This observation indicates that ellipsometry in the 400–1000 nm range is able to follow in-situ the reduction of WO3 oxides by D plasma with high surface sensitivity.

The pre-fusion power operation (PFPO) phase of ITER, as described in the ITER research plan with Staged Approach2, includes both hydrogen (H) and helium (He) plasma operations. In preparation for PFPO, both WEST and JET ran He plasma campaigns to study plasma-wall interactions in a tungsten environment. The campaigns included a back-and-forth transition between H or deuterium (D) and He plasma operation allowing the assessment of the achievable plasma content as well as the accessible wall reservoirs for respective species. The WEST changeovers included tokamak pulses with a fixed divertor configuration. The JET changeovers applied ion cyclotron wall conditioning (ICWC) and tokamak pulses including limiter phases and four different divertor configurations. Glow discharge conditioning (GDC) was applied to complete the changeovers. The results are characterized by subdivertor optical and mass spectrometric gas analyzers and spatially resolved optical emission spectroscopy. A He content of 96-97% after H operations is achieved by tens of ICWC pulses (JET) and several dedicated diverted plasmas (WEST and JET), while a fivefold is estimated to be required for the back transition. Effective pumping of the wall released species is a key parameter for a fast changeover. Upon applying higher heating power, the relative content of the fueled plasma species decreases. The JET gas balance analysis indicates that He operation may increase H retention. WEST divertor spectroscopy indicates a larger He inventory near the inner divertor strike line. He GDC has a clear effect on the He recycling light at the WEST divertor while D GDC did not reduce the long lasting He content observed in D pulses after the JET He campaign.

Removal of neutral hydrogen atoms in the plasma edge reduces the number of charge exchange events and thus, the net energy losses in the plasma, significantly improving performance of fusion devices. Effective control of the residual pressure of hydrogen isotopes (HIs) in the plasma edge may be achieved by utilizing a hydrogen absorbing first wall interface capable of withstanding the harsh fusion environment. In this study, we have investigated tantalum (Ta) coating deposited by cold spray technology on 316L stainless steel substrate as a potential plasma-facing material surface. High fluence low energy deuterium plasma irradiation experiments and subsequent thermal annealing cycles associated with thermal desorption spectrometry (TDS) demonstrated superior structural stability of the Ta coating. TDS experiments revealed the outgassing of deuterium (as measure of its retention) for cold spray Ta coatings to be three times higher than bulk Ta and two orders of magnitude greater than bulk polycrystalline W. X-ray photoelectron spectroscopy revealed evolution of oxidation states upon deuterium irradiation and a partial recovery of the metallic signature of Ta after the thermal treatment at 1100 K.

Helium (He) retention in tungsten (W) is a concern in fusion reactors since it could be detrimental to plasma facing components performance and influence the fusion fuel balance. He being not soluble in W, it tends to agglomerate on preexisting defects (vacancy, grain boundary), but it could in theory also self-trap (be immobilized on a non-preexisting vacancy) through the emission of a vacancy/self-interstitial W pair in the vicinity of a Hen interstitial cluster. In the present study, we prepared a pure single crystal W(1 1 0) sample with a clean surface in order to evidence the self-trapping of He in the W bulk at a sample temperature of 300 K and for a constant fluence of 2.0 × 1021 He+.m−2. At a He+ kinetic energy of 130 eV and a flux of 0.3 × 1017 He+.m−2.s−1, we only observed a small He desorption peak below 600 K. Rising the ion flux to 0.7 × 1017 He+.m−2.s−1, we observed the sudden appearance of two desorption peaks at 950 K and 1700 K. For the highest flux studied in this work, 5.0 × 1017 He+.m−2.s−1, an additional desorption peak at 1800 K and a desorption shoulder at 1900 K are observed. The temperature position of these He desorption peaks are consistent with the density functional theory literature and points to the occurrence of self-trapping once the 0.7 × 1017 He+.m−2.s−1 flux is attained at 300 K and to the possible realization of trap-mutation for the flux of 5.0 × 1017 He+.m−2.s−1. The present set of results should be used to constrain the development of He retention and He bubbles growth models based on ab initio quantities.

Stainless steels are widely used as structural materials due to their excellent corrosion resistance and fabrica-bility. In some applications, in nuclear fusion reactors for instance, these materials may be exposed to harsh conditions, including high heat flux and ion bombardment in a vacuum environment, which can significantly affect their surface properties and composition. In this work, we investigate the temperature-dependent evolution of the surface composition and optical properties (reflectivity, emissivity) of stainless steel 316L. X-ray photoelectron spectroscopy analyses, temperature programmed desorption and reflectivity measurements in the visible/near-infrared range were performed on polished samples from ambient temperature to 1000 K. It was found that around 750 K, the initial Fe/Cr surface abundance of 90/10 changes to 35/55 following the onset of Cr and CrO desorption at 575 K, with Cr scavenging oxygen initially bound to Fe. In addition, Mn segregated at the surface and oxidized, while small amounts of metallic Ni segregated. Despite these modifications in surface composition, no significant change in reflectivity up to 900 K was observed. These results provide insight into the behavior of stainless steel under extreme conditions which is relevant for understanding its performance and durability in high-temperature applications.

This paper explores the plasma parameters of helium and hydrogen Ion Cyclotron Wall Conditioning (ICWC) discharges performed in JET as part of a He/H fuelling changeover experiment. The conducted study shows that plasma with a higher density is formed in helium than in hydrogen. A distinct glow in the ion cyclotron resonance zone is observed throughout the discharge in He. In H-ICWC discharges, a lower radio-frequency coupling efficiency and coupled power was observed than in He-ICWC discharges. While the helium concentration decreased with the number of H-ICWC pulses and the same for hydrogen in He-ICWC, which is the intended result of the plasma wall interaction in the ICWC changeover procedure, the main features of hydrogen as well as the helium IC discharge do not change dramatically.

Tungsten (W) is the material selected for the divertor exhaust of the international nuclear fusion experiment ITER. In this harsh environment, the interactions of heat loads and ion fluxes with W can induce temporary or permanent evolution in the optical properties. Poor knowledge of such evolution during a plasma operation can lead to errors in temperature measurements performed by optical diagnostics. Therefore, it is of fundamental importance to characterize possible changes in W optical properties. In this work, we studied the role of morphology and temperature on the optical response of W. The reflectivities of five W samples with different roughness values (20–100 nm) were measured during laser annealing (25–800 °C) in the visible and near-infrared domains (500–1100 nm). We observed an increase in reflectivity after annealing and we demonstrated that it was due to a change in the chemical composition of the surface, in particular a reduction in the amount of native oxide. Moreover, we show that roughness does not sensibly vary in the investigated temperature range. By highlighting the role played by roughness and surface impurities (e.g., oxide), we provide insight in how W optical properties can evolve in tokamaks where high ion fluxes, heat loads, and impurities can induce the evolution of both the morphology and surface composition of W.

The knowledge of optical properties of beryllium is of crucial importance in fields such as nuclear fusion and aerospace applications. The optical properties of pure beryllium are known in the visible and infrared domains. Nevertheless, the role of different physico-chemical parameters such as composition and surface roughness, that is often neglected in first approximation, deserves dedicated comprehensive studies. In this work we have studied the optical properties of bulk beryllium and magnetron sputtering beryllium layers in the 500–2000 nm spectral range. Experimental measurements show that beryllium reflectivity strongly depends both on bulk fabrication procedure and on surface preparation. Different models allow us to perform a quantitative interpretation of reflectivity results and to study the influence of different parameters: (i) a multi-reflection interference model to understand the role of oxide layer, (ii) a Lorentz–Drude model for the bulk composition effect, (iii) scattering models for the surface roughness, and (iv) the Maxwell–Garnett model for the surface porosity. The calculated relative permittivity of the studied samples is used to evaluate the emissivity in the visible and infrared domain. Such evaluation, giving indications of possible evolution of optical properties of beryllium in a plasma environment, can provide a useful tool for thermography studies of tokamak walls.

Surface and near-surface morphology evolution of helium-irradiated tungsten due to thermal cycling up to the ITER-relevant temperature of 1350 K was studied using electron microscopy and positron annihilation spectroscopy techniques. Holes at the surface and bubbles in the near-surface of recrystallized polycrystalline tungsten samples were created by 75 eV helium plasma irradiation with the fluence of 3 × 10 23 He m −2 at the surface temperature of 1073 K. Subsequent annealing experiments were combined with a detailed electron microscopy analysis to investigate the shape and density changes of holes and helium bubbles with respect to grain orientation. We show that the initially circular holes and round bubbles became faceted upon heating with 1 K s −1 ramp up to 870 K. Annealing cycles up to 1350 K induced resulted in bubbles removal in the first 5 nm below the surface and surface smoothing. Electron energy loss spectroscopy measurements allowed estimation of helium gas density and pressure inside bubbles. Positron annihilation spectroscopy allowed to investigate the nature of defects and their evolution in the helium-irradiated tungsten with thermal cycling.

ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating ∼2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E inc > 20 eV, helium fluence >10 24 He/m 2 , surface temperature T surf > 700 °C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.