Recent studies of tungsten-based plasma-facing materials in the linear plasma device STEP

Yin, Hao; Wang, Jun; Guo, Wangguo; Cheng, Long; Yuan, Yue; Lü, Guang-Hong

doi:10.1007/s42864-019-00004-x

Cited by 30 publications

(15 citation statements)

References 45 publications

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“…The tested components were placed at the center of the neutral beam, as such loading conditions were different, depending on the distance between the sample position and the center of the beam. Before the official test, screening tests at different power densities (6,8,10,13, and 15 MW m −2 ) were carried out, and the surface temperatures of the central samples were obtained. Subsequently, a cyclic heat loading with a heat flux of 10 MW m −2 was performed for the W-Y 2 O 3 component, and 8 MW m −2 was conducted for the component consisting of pure W and W-K alloy.…”

Section: High-energy Neutral Beam Irradiationmentioning

confidence: 99%

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

et al. 2019

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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.

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Section: High-energy Neutral Beam Irradiationmentioning

confidence: 99%

“…In recent studies [5][6][7][8], different kinds of dispersionstrengthened tungsten materials processed by hot-plastic deformation have been developed. The obtained samples have fine grains and improved performances.…”

Section: Introductionmentioning

confidence: 99%

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

et al. 2019

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“…Under neutron irradiation, there will be collisions between energetic particles and lattice atoms, leading to a large number of displacement cascades, ultimately creating plenty of point defects (vacancies and self-interstitial atoms) in plasma-facing materials [16][17][18]. These defects may accumulate to form large-scale defects, which will induce the degradation of thermal and mechanical properties [8,10,[19][20][21][22]. Due to the high binding energies and low migration energy barriers, self-interstitial atoms (SIA) tend to cluster rapidly and eventually accumulate to form the interstitial dislocation loop (IDL) [23][24][25], which can be observed by transmission electron microscopy (TEM) [26,27].…”

Section: Introductionmentioning

confidence: 99%

Interaction of 1/2〈111〉 interstitial dislocation loop with hydrogen and helium in tungsten: molecular dynamics simulation

Xu,

Li,

Wang

et al. 2023

Mater. Res. Express

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The interaction of hydrogen and helium atoms with 1/2 <111> interstitial dislocation loop in tungsten is investigated by molecular dynamics simulation. The binding energies of hydrogen and helium atoms around dislocation loop are calculated by molecular statics method. The results show that the outer region of the loop is attractive to the two atoms and the inner region is repulsive. Notably, the maximum binding energies are located in the core region of the dislocation loop. We have also studied the influence factors of the interaction between the dislocation loop and two atoms: free volume, lattice distortion degree, the radius and shape of the dislocation loop. The results show that large free volume benefits the retention of hydrogen and helium atoms, especially for helium. The less lattice distortion caused by the impurity atom, the more favorable for the dislocation loop to trap it. In addition, the larger dislocation loop with higher defect concentration results in stronger capture ability for the hydrogen and helium atoms. The different dislocation loop shapes lead to different binding energy distribution patterns. And the hydrogen and helium atoms tend to occupy the groove region of the concave dislocation loop. Finally, we employ the nudged elastic band theory and dynamics method to investigate the diffusion pattern of the hydrogen atom in the dislocation loop and find that the hydrogen atom tends to migrate spirally around dislocation line. Based on the obtained results, a reasonable interpretation of the interaction behaviors between the dislocation loop with hydrogen and helium atoms are discussed, which can provide essential parameters for mesoscopic scale simulations.

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“…To this end, various materials (such as W [47][48][49][50], addition of Rh in W [51], use of bcc Fe [52,53], Ta [54], W-Ta [55], Ta/Fe [56], Pd [57], nanocrystalline Cu [58], SiOC/Crystalline Fe nanocomposite [59], W-K [60], reduced activation steel [61], ferritic [62], ferritic/martensitic steels [63], Be pebbles [64][65][66][67], Be and beryllides [68], graphite, carbon fiber composite [69]) and high Z atoms (Zr, No, Mo, Hf, Ta) [70] have been tested but none proved satisfactory [71][72][73][74]. All show rapid surface degradation exhibiting surface blisters [75][76][77][78] and formation of fuzz [51,[79][80][81][82] or under dense nanostructure [40] after bubble.…”

Section: Introductionmentioning

confidence: 99%

Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

Rafique¹

2023

Preprint

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Bubble (point defect) – a precursor of fuzz or under dense nanostructure formation is crystal lattice defect. Suitable selection of crystal lattice which inhibit Frenkel pair generation and intrinsically promotes selfinterstitial solid solution strengthening contributes effectively towards making plasma facing material. For this, interstitial sites, their size, amount / fraction, positions, tendency of occupation and diffusion parameters (e.g. activation energies (Q), activation volumes) are determined. Fcc iron carbon alloys (austenitic stainless steels AISI / SAE 321, fcc structure, Pearson code cF4, space group Fm3̅m) are proposed as suitable candidates. Along with their room temperature fcc structure having 12 interstitial positions (4 octahedral, 6 coordination sites and 8 tetrahedral, 4 coordination sites / unit cell) to allow insertion of self (iron) atoms, they have excellent corrosion resistance, thermal conductivity, and nonmagnetic properties. After their melting, casting, and machining to required dimensions and geometry, stabilizing heat treatment is applied to precipitate all carbon as TiC and prevent formation of Cr23C6 (sensitization). This resist heat and surface degradation and yield excellent architecture which not only inhibit Frankel pair generation but will also allow bulk assimilation or surface annihilation (loop punching) of this lattice point defect. A superior thermal, fluid, and structural design augment above

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Recent studies of tungsten-based plasma-facing materials in the linear plasma device STEP

Cited by 30 publications

References 45 publications

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

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

Interaction of 1/2〈111〉 interstitial dislocation loop with hydrogen and helium in tungsten: molecular dynamics simulation

Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

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