Polished W discs exposed to pure He plasma in the PISCES-B linear-divertor-plasma simulator at 1120 and 1320 K are found to develop deeply nanostructured surface layers consisting of a conglomerate of amorphous ‘nanorods’. The growth of the thickness of the nanostructured layer is explored for exposure times spanning 300–(2.2 × 104) s in He plasmas of density n e ∼ 4 × 1018 m−3 and temperature T e ∼ 6–8 eV where the average He-ion surface-impact energy is ∼60 eV, below the threshold for physical sputtering. A nanostructured layer in excess of 5 µm thick is observed for the longest exposure time explored. The kinetics of the layer growth are found to follow Fick's law, characterized by an effective diffusive mechanism with coefficients of diffusion: D 1120 K = 6.6 ± 0.4 × 10−12 cm2 s−1 and D 1320 K = 2.0± 0.5 × 10−11 cm2 s−1. The diffusion of He atoms in W is considered too rapid to explain the observed growth of surface modification and points to the interplay of other mechanisms, such as the availability of thermal vacancies and/or the slower diffusion of He through the forming nanostructured layer.

Abstract: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 strikepoint 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 [ 1 ] 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.

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Trapping of tritium at lattice damage from fusion neutron irradiation is expected to increase the tritium inventory in tungsten components in ITER. The magnitude of this increase depends on the concentration of traps that are produced, and on the depth to which the increased tritium retention extends into the material. Experiments to address these issues are described, in which displacement damage by ion irradiation was used as a surrogate for neutron damage. Irradiated samples were exposed to high flux deuterium plasma to simulate divertor conditions. The resulting deuterium content was measured by nuclear reaction analysis. Measurements were done at various damage levels up to those expected from the end-of-life neutron fluence in ITER. These experiments determine the number of traps produced by displacement damage and the rate at which they are filled during exposure to plasma. The role of defect annealing was explored through plasma exposures at various temperatures. In addition to trapping at damage, near-surface retention from internal precipitation was observed at lower temperatures. Addition of 5% helium to the deuterium plasma greatly reduced D retention by precipitation by localizing it closer to the surface. Results from these experiments indicate that the contribution to tritium inventory in ITER from trapping at neutron damage should be small.

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