The response of a variety of W material grades to nanostructure ‘fuzz’ formation is explored. W targets are exposed to He or D
2–0.2He plasmas in PISCES-B at 900–1320
K to below sputter threshold He
...+ ions of energy 25–60
eV for up to 2.2
×
10
4
s. SEM and XPS reveal nanoscopic reorganization of the W surface to a layer of ‘fuzz’ of porosity ∼90% as determined by a ‘fuzz’ removal/weight loss method. The variability of ‘fuzz’ growth is examined at 1120
K for 1
h durations: SR, SC and doped W grades – La
2O
3 (1% wt.), Re (5% and 10% wt.), and TiC (1.5% wt.) developed 2–3
μm thick ‘fuzz’ layers, while a VPS grade developed a layer 4
μm thick. An RC grade revealed additional ‘fuzz’ at deep (>100
μm) grain boundaries. However, heat treatment up to 1900
K produced reintegration of ‘fuzz’ with the bulk and He release at ∼1000
K and ∼1400–1800
K due to depopulation from vacancy complexes.
Sputtering yields of He-induced W ‘fuzzy’ surfaces bombarded by Ar have been measured in the linear divertor plasma simulator PISCES-B. It is found that the sputtering yield of a fuzzy surface, ...Yfuzzy, decreases with increasing fuzzy layer thickness, L, and saturates at ∼10% of that of a smooth surface, Ysmooth, at L>1μm. The reduction in the sputtering yield is suspected to be due mainly to the porous structure of fuzz, since the ratio, Yfuzzy/Ysmooth follows (1−pfuzz), where pfuzz is the fuzz porosity. Further, Yfuzzy/Ysmooth is observed to increase with incident ion energy, Ei. This may be explained by an energy dependent change in the angular distribution of sputtered W atoms, since at lower Ei, the angular distribution is observed to become more butterfly-shaped. That is, a larger fraction of sputtered W atoms can line-of-sight deposit/stick onto neighboring fuzz nanostructures for lower Ei butterfly distributions, resulting in lower ratio of Yfuzzy/Ysmooth.
This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. ...Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼1030m−2), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue.
Microscopic damage and D retention in tungsten have been investigated for low-energy (∼60–120eV), high flux (∼1022m−2s−1), high fluence (∼5×1025m−2) ion bombardment at moderate temperature ...(∼573–773K) in mixed species D+He plasmas in the linear divertor plasma simulators PISCES-A and B. A significant reduction in D retention and the formation of nanometer-sized He bubbles occur in W due to seeding of He into the D plasma. The volume fraction of He bubbles, estimated with TEM observations and ellipsometric measurements, exceeds the percolation threshold. The desorption mechanism that injected D atoms diffuse back to the surface through the percolating bubbles is suggested. The seeding of Be into D+He mixture plasma eliminates this He effect on the reduction in D retention.
Nanotendril "fuzz" will grow under He bombardment under tokamak-relevant conditions on tungsten plasma-facing materials in a magnetic fusion energy device. We have grown tungsten nanotendrils at low ...(50 eV) and high (12 keV) He bombardment energy, in the range 900-1000 °C, and characterized them using electron microscopy. Low energy tendrils are finer (~22 nm diameter) than high-energy tendrils (~176 nm diameter), and low-energy tendrils have a smoother surface than high-energy tendrils. Cavities were omnipresent and typically ~5-10 nm in size. Oxygen was present at tendril surfaces, but tendrils were all BCC tungsten metal. Electron diffraction measured tendril growth axes and grain boundary angle/axis pairs; no preferential growth axes or angle/axis pairs were observed, and low-energy fuzz grain boundaries tended to be high angle; high energy tendril grain boundaries were not observed. We speculate that the strong tendency to high-angle grain boundaries in the low-energy tendrils implies that as the tendrils twist or bend, strain must accumulate until nucleation of a grain boundary is favorable compared to further lattice rotation. The high-energy tendrils consisted of very large (>100 nm) grains compared to the tendril size, so the nature of the high energy irradiation must enable faster growth with less lattice rotation.
Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed ...in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a
Q
=
10 ITER discharge G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260 which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4
±
3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5
MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-
Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.
The formation of He bubbles in tungsten under exposure to a He plasma was systematically investigated using low energy (∼50eV) He+ ions with a wide fluence range (∼1023 to 1026m−2) in the linear ...divertor plasma simulator PISCES-A at several temperatures (523–973K). TEM observations after thinning exposed W samples with FIB revealed that the layer thickness (>30nm) of He bubbles largely exceeds the ion implantation range of a few nm. The size of He bubbles was found to increase with an increase in the sample temperature: it was around 10nm at 973K, while only small He bubbles of 1–2nm were observed at <773K. In addition, to obtain information on the initial formation behavior, in-situ TEM observations during He ion irradiation were also performed.
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