Arkadi Kreter

Neon (Ne) seeding is essential in ITER to mitigate divertor heat loads, but it significantly influences plasma-surface interactions and fuel retention in tungsten (W). Moreover, transient heat loads (THLs) from edge-localized modes (ELMs) further alter surface morphology and fuel retention behavior. This work investigates the combined effects of Ne seeding and ELM-like THLs on surface modification and deuterium (D) retention in rolled W. Experiments were carried out in the linear plasma device PSI-2 using pure D plasma and D + 10% Ne plasma at 500 K, with an ion energy of 40 eV, and a fluence of 1 × 10 ²⁶ D m ⁻² . Transient loads were applied using a pulsed laser (1 ms, 0.76 GW m ⁻² , 1000 cycles), either successively or simultaneously with plasma exposure. For plasma-only exposures, Ne seeding caused pronounced erosion and fine cracks on blister caps, resulting in approximately 32% lower total D retention compared with pure D plasma. THLs generated temperature and stress fields, thereby causing surface roughening and dynamic recrystallization, as well as increasing D retention in both the near-surface (< 4 µ m) and deeper regions. When combined with Ne seeding, the effects depended on the loading sequence. With sequential THLs followed by plasma exposure, Ne seeding reduced the near-surface and total D retention by about 44% and 27%, respectively, compared with the corresponding D-only condition. In contrast, under simultaneous THL + plasma exposure, Ne seeding increased near-surface and total D retention by approximately 62% and 16%, respectively, and produced an additional broad D release peak at around 1200 K. This peak could be attributed to the formation of high-binding-energy Ne-vacancy complexes (Ne _i -V complexes) induced by the synergistic interaction of Ne seeding and transient heat loading. These results demonstrate that both impurity seeding and THLs critically determine D retention behavior, emphasizing the importance of accounting for their synergistic effects when predicting fuel retention and optimizing operation scenarios in future fusion devices.

Controlling hydrogen isotope retention is a critical challenge for plasma-facing materials (PFMs) in fusion reactors. Chemical vapor deposited tungsten (CVD-W), featuring columnar grain structures, has attracted increasing attention as a candidate PFM. However, its surface blistering and deuterium (D) retention behavior under fusion-relevant, complex irradiation conditions remain insufficiently understood. In this work, CVD-W and ITER-like forged W both with grains elongated normal to the exposed surface were irradiated with pure D and mixed D + 5% He plasma, with and without prior W self-ion irradiation at damage levels of 0.2 and 2 dpa. Compared with ITER-like W, CVD-W exhibits substantially enhanced tolerance to D-induced surface blister formation and markedly reduced D retention under pure D plasma exposure. Regardless of displacement damage, He seeding, or their combination, CVD-W consistently retains less D than ITER-like W across all irradiation sequences studied. This persistently low D retention in CVD-W is primarily associated with its lower defect density (e.g. fewer grain boundaries) and a blister-resistant surface texture that limits blister-related D trapping. Furthermore, displacement damage and He seeding effectively suppress surface blistering, but exert opposing individual effects on D retention in both materials. Their combined effect shows a clear dependence on D fluence and material microstructure. This work highlights the advantages of CVD-W in suppressing hydrogen isotope retention under complex irradiation environments and provides valuable insights for the selection and microstructural design of advanced PFMs with improved resistance to hydrogen-induced blistering and reduced retention.

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.

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– 3.5 × 10 21 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 1 × 10 24 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.

This work presents the first experimental study on the near-threshold sputtering regime for monocrystalline low-index plane tungsten targets investigated using high-resolution emission spectroscopy. We analyzed the line shape emitted by sputtered atoms, which contains information on the angular and velocity distribution functions via Doppler broadening. Specifically, we report changes in the line profile of the resonant W I 498.4 nm transition during plasma exposure of polycrystalline and monocrystalline (100) and (111) tungsten targets at the linear plasma device PSI-2. Biasing the targets from −60 V to −100 V provided low-energy argon ions for near-threshold sputtering. The line shapes, measured along the angle of observation perpendicular to the normal of the sample, were significantly broader for the monocrystalline (100) and (111) compared to that of the polycrystalline target. In particular, the (111) target demonstrates a pronounced heart-shaped profile. The modeling captures this distribution via a ∝ cos(θ)exp(−bθ) function—θ is the polar angle—combined with a parameterized Thompson velocity distribution. Furthermore, comparing the experimental data to molecular dynamics simulations at 100 eV illustrates a reasonable agreement of the angular distribution function with the measurements.

Laser absorption spectroscopy provides high-resolution spectra of atomic transitions that reveal many often inaccessible features. The line shapes of krypton and xenon measured in magnetized plasmas are strongly affected by the contribution of the odd-numbered isotopes ⁸³Kr, ¹²⁹Xe and ¹³¹Xe due to their hyperfine structure, creating more challenging spectra in comparison to even-numbered ones. The lines originating from metastable levels of krypton and xenon with J = 2 (Kr I 760.4 nm) and J = 0 (Kr I 785.7 nm, Xe I 764.4 nm) were measured and analyzed in the linear plasma device PSI-2 in the field range of 22.5 mT–90 mT. Evaluating the Hamiltonian, including hyperfine and Zeeman interaction terms for these magnetic field strengths, unveils a deviation from the linear energy shift of the sublevels as a function of the magnetic field and from constant relative intensities that the weak field formulas provide. We prove that modeling the transitions in Xe using the weak field approximation, frequently used in magnetized plasma, becomes inadequate at ≈50 mT. In particular, the spectra of the ¹³¹Xe isotope show pronounced deviations from the weak field results. For krypton, however, the situation is less critical compared to xenon due to the low natural abundance of the odd-numbered isotope.

The impact of helium plasma exposure on the tungsten surface damage structure development and erosion has been investigated by comparing the impact of hydrogen plasma exposure. Crystal orientation dependence of the undulating surface structure formation and erosion rate is observed on the plasma-exposed tungsten surface independently from the plasma species. The top surface of the plasma exposed tungsten has a tendency to {100} plane independently from the initial surface orientation. Although hydrogen and/or helium cause no erosion in tungsten under incident ion energy exposure conditions below the sputtering threshold, inevitable minute impurities, like oxygen, play an essential role in erosion, and significant erosion can be observed even at 30 eV.

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.