Irradiation effects of H/He neutral beam on different forged tungsten materials

Abstract

Tungsten has been considered as the most promising plasma-facing materials (PFMs) for the future fusion reactor due to its excellent chemical and physical properties. In this work, different kinds of tungsten-based PFMs including pure W, W-K, and W-Y 2 O 3 alloys were prepared via powder metallurgy. Then, high-energy rate forging was used for the plastic deformation to refine the grain size of the samples and increase their densities. To evaluate the service performance of these tungstenbased PFMs under edge plasma conditions, the deformed samples were exposed to high-energy hydrogen (H) neutral beam doped with 6 at% helium (He). Sequentially, edge-localized mode-like transient loadings (1 ms, 100 cycles) were carried out on the specimens with neutral beam pre-loading to study the impact of H/He pre-loading on the generation and propagation of cracks. The microstructural changes of the exposed surfaces were observed by scanning electron microscopy with the help of focused ion beam. The experimental results indicate that H/He neutral beam irradiation could induce obvious surface damages of the tested samples. For example, pinhole and coral-like nanostructures were formed on the surfaces of the exposed samples, as the surface temperatures were relatively high. The cracking thresholds of the W-Y 2 O 3 samples with and without neutral beam pre-irradiations were both between 0.22 and 0.33 GW m −2 , while the cracking thresholds of pure W and W-K samples were less than 0.22 GW m −2. This result indicates that the H/He neutral beam pre-loading has no significant impact on the cracking threshold. In addition, the crack pattern on the surface of the exposed samples closely depends on the grain boundary pattern.

Vacancy‐Type Defects in H+6%He Neutral Beam Irradiated WK Alloy Probed by Slow Positron Beam

“…[ 39,40 ] The gradient of stress and temperature during irradiation drives the He nanobubbles to migrate to the surface and be released, leaving a pinhole on the surface. [ 8,41 ]…”

“…The He nanobubbles near the surface are broken under the gradient of temperature and H/He stress during the irradiation, releasing H and He particles and forming a large number of pinholes, as shown in Figure 3 of our previous work on these types of samples. [ 8 ] Almost all H atoms would have already been released at this high surface temperature of 800 °C, [ 35 ] which implies that much fewer H atoms can diffuse to the bulk layer and interact with intrinsic vacancy‐type defects. This explains why the S‐parameter in the bulk layer of WK‐800 is greater than that of WK‐400 but still smaller than that of WK‐UN.…”

“…Furthermore, the rupture of large helium bubbles near the surface was more severe, releasing more H/He atoms and causing a coral‐like structure on the surface, as shown in Figure 3 of our previous work on these types of samples. [ 8 ]…”

“…Therefore, the attention of researchers has shifted to ultrafine‐grained nanostructured W, alloying, and dispersion strengthening to strive for the optimization of W‐based PFMs with more suitable properties. [ 6–9 ] Plastic‐deformed W‐based materials, which have relatively ultrafine‐grained nanostructures, have demonstrated good tolerance, good resistance, and higher strength in transient thermal load experiments. [ 10 ] K‐Doped W‐based materials, which were first used to strengthen tungsten wires and have a long history in the lighting industry, are also regarded as important dispersal strengthening materials.…”

In situ measurements of low energy D plasma-driven permeation through He pre-damaged W

“…Tungsten (W) has been selected as the plasma facing material (PFM) for ITER and other fusion reactors in the future due to its outstanding properties such as low sputtering yield, high melting temperature and low hydrogen (H) isotope retention. PFM would suffer from high flux (∼10 20 -10 24 m −2 s −1 ) of deuterium (D), tritium (T) ions as well as helium (He) ion [1,2]. Considering the nuclear safety issue, permeation of T through wall materials of fusion reactors into the coolant system has been a major concern.…”