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In fusion devices like ITER, plasma-wall interactions are a significant concern due to the high heat fluxes, often tens of MW/m2, impacting the first wall. These intense heat fluxes can lead to the formation of hot spots on components facing the plasma, such as tungsten, used in divertor plates and antennas. This results in material erosion and plasma core contamination. Our study investigates the thermal behavior of tungsten surfaces under these conditions using fluid modeling and Particle-In-Cell (PIC) simulations. We examine the effects of thermionic electron emission on the sheath potential and heat transmission. The simulations reveal that thermionic emission can decrease the sheath voltage, increasing the surface temperature due to enhanced heat flux due to electrons. Additionally, we explore how the ratio between the spot size (S) and the surrounding surface length (Ly) influences the surface temperature. We find that a higher Ly/S ratio allows the surface to reach higher temperatures before the system enters the space-charge-limited regime, where thermionic current is maximized and considerably larger than the case where the entire surface is emissive (Ly=S).

The investigation of plasma facing materials (PFM) subjected to a large number (≥10,000) of thermal shocks is of interest to determine long term morphological changes which might influence component lifetime in and plasma performance of a fusion reactor. The electron beam facility JUDITH 2 was used to simulate these conditions experimentally. In this study eight different tungsten grades produced by powder injection molding (PIM) were investigated: Two pure tungsten grades, one with 2 wt% Y₂O₃, three with 1, 2 and 3 wt% TiC, and two with 0.5 and 1 wt% TaC. Samples of 10 × 10 × 4 mm³ were brazed to a copper cooling structure and subjected to 10⁵ thermal shocks of 0.5 ms duration and an intensity of L$_{abs}$=0.55 GW/m² (F$_{HF}$=12 MWs½/m2) at a base temperature of T$_{base}$ = 700 °C. The PIM grades showed damages in general comparable with a sintered and forged pure tungsten reference grade (>99.97 wt% W) that complies with the ITER specifications. One exception was the 2 wt% TiC doped material which failed early during the experiment by delamination of a large part of the surface. The Y₂O₃ doped material showed a comparatively good performance with respect to crack width (<15 μm) and roughening (R$_{a}$ = 0.75 μm), but showed melt droplets of ∼3–4 μm diameter, while the 1 wt% TiC doped material showed wide cracks (up to 50 μm) and strong roughening (R$_{a}$ = 2.5 μm). The paper discusses the post-mortem analysis of all grades, comparing them with respect to roughness (from laser profilometry), crack network characteristics and local melt droplet formation or other special morphological features (from SEM images) as well as crack depth (from metallographic cross sections).

Recrystallized, polycrystalline tungsten was self-damaged by 20MeV tungsten ions up to a calculated damage dose in the damage peak of 0.23dpa. The time to acquire this dose and hence the average damaging dose rate was varied from 6×10-3 to 4×10-6dpa/s, the latter coming close to the damage dose rate expected from fusion neutrons in future devices such as ITER and DEMO. One series was conducted at 295K and one at 800K to check for possible effects of defect evolution at elevated temperature. The created damage was decorated afterwards with a deuterium plasma at low ion energy of <15eV and low flux of 6×1019D/m2 until saturation to derive a measure for the defect density that can retain hydrogen isotopes. 3He nuclear reaction analysis (NRA) was applied to derive the deuterium depth profile and the maximum concentration in the damage peak. Neither for the 295K nor for the 800K series a variation in deuterium retention with damage dose rate was found. Keywords: Tungsten, Deuterium retention, Displacement damage, Plasma, NRA, Plasma–material interactions, Ion radiation effects

This research investigated fuel particle balance during long duration discharge in an all-metal plasma facing wall (PFW) through intensive QUEST execution. A simple wall model including the plasma-induced deposition layer that creates hydrogen (H) barriers, called the H barrier model, was established. A simple calculation, based on a combination of H state rate equations and the H barrier model, was applied to real plasma in the early phase of its longest discharge. The model accurately reconstructed the evolutions of electron density and wall-stored H over time, proper values are chosen for the parameters that are difficult to determine experimentally. Comparative calculations that used the H barrier and a fully reflective models, predicted significant impacts of wall models on the plasma density time response and value of electron density, indicating that a proper wall model should be developed for all-metal PFW devices. Keywords: Fuel particle balance, steady state operation, Hydrogen barrier model, QUEST

The paper presents complementary approaches based on experimental and numerical works to address the behavior of tokamak-relevant tungsten particles loaded with tritium. Sampling of particles inside the WEST tokamak have been realized thanks to an in situ particle collection system called Duster Box. This method allowed to identify various types of tungsten particles among them spherical shaped micro-particles between 5 µm and 30 µm in diameter. Based on these results a surrogate tungsten powder has been provided by means of spheroidization process and sieving method. Moreover, the powder tritium retention capacity was measured and specific activities of 90 MBq.g−1 and 280 MBq.g−1 were obtained for particles with 17 µm and 11.5 µm median diameters, respectively. Considering such tritium activities trapped in the particles, Monte-Carlo simulation were performed to estimate the electrostatic self-charging rates and the corresponding electrical charge carried by the radioactive tungsten dust. The results of these experiments provide robust data for the assessment of the dispersion of toxic/radioactive material in the environment that could follow a loss of containment.

Nuclear materials and energy 38, 101587 - (2024). doi:10.1016/j.nme.2024.101587 Published by Elsevier, Amsterdam [u.a.]