A5068088351- Fusion materials and technologies
- Nuclear Materials and Properties
- Advanced materials and composites
- Nuclear Physics and Applications
- Metal and Thin Film Mechanics
- High-Temperature Coating Behaviors
- Nuclear reactor physics and engineering
- Magnetic confinement fusion research
- Titanium Alloys Microstructure and Properties
- Ammonia Synthesis and Nitrogen Reduction
- Hybrid Renewable Energy Systems
- Nuclear materials and radiation effects
- Advanced Materials Characterization Techniques
- Hydrogen embrittlement and corrosion behaviors in metals
- Muon and positron interactions and applications
- Molten salt chemistry and electrochemical processes
- Hydrogen Storage and Materials
- Particle accelerators and beam dynamics
- Ion-surface interactions and analysis
- Nuclear physics research studies
- Advanced ceramic materials synthesis
Science and Technology on Surface Physics and Chemistry Laboratory
2022-2024
State Key Laboratory of Nuclear Physics and Technology
2021
Peking University
2021
Abstract The safety of future fusion reactors is critically dependent on the tritium (T) retention in plasma-facing materials. Hydrogen isotope (HI) exchange offers a method to redistribute HIs within solid materials, presenting feasible approach for removing T from bulk materials and trapped by strong trapping sites. Nonetheless, unraveling intricate mechanism behind HI remains an urgent yet formidable challenge. This study undertakes comprehensive investigation into tungsten across...
Abstract In fusion reactors, plasma-facing materials (PFMs) work under a complex irradiation environment. A lot of defects will be generated in tungsten (W), primary candidate for PFMs, leading to substantial amount hydrogen fuel retention. Among various induced defects, <100> interstitial dislocation loops can stably exist and trap isotopes, significantly threatening the safety reactors. Therefore, understanding mechanism behind trapping de-trapping behavior isotopes is an urgent...
Abstract A large-size potassium-doped tungsten (KW) plate with a thickness of 15 mm was fabricated via powder metallurgy technology and hot rolling. In order to appraise the irradiation resistance KW, surface deuterium (D) blistering D retention were studied on Fe 11+ pre-damaged (0, 0.05 0.5 dpa) KW pure (PW), which exposed ∼60 eV ∼5 × 10 21 m −2 s −1 plasmas at 500 K fluence ∼5.76 25 . The results indicate that alloy can better inhibit generation vacancy defects after ion damage compared...
This work investigates the irradiation hardening and deuterium (D) retention behaviour of tungsten (W) under synergistic irradiations heavy ions D plasma. 3.5 MeV iron (Fe13+) ion was performed on recrystallized W samples (RW) to produce displacement damage 1 dpa. Then, low-energy (38 eV) plasma exposure conducted at 500 K. It is found that Fe creates substantial vacancy-type defects dislocation loops/networks in RW. These irradiation-induced not only function as nucleation sites for...
The single-crystal lithium hydride (LiH) generally grows in a gradient temperature region with the Bridgman method. A stable and appropriate is crucial crystallization process. In this paper, variation of LiH growth calculated by finite element method (FEM). It shown that compact melted entirely after heating to 750 °C at 10 °C/min dual-temperature furnace holding for 2.4 h. margin was 46.5 5 cone point, respectively, decreased 33.7 °C, 28.6 25.6 16.5 when upper maintained lower cooled 680...
Abstract Tritium (T) retention in plasma-facing materials (PFMs) raises significant radiological safety concerns and adversely affects the self-sustained burning of T fusion reactors. Therefore, removal retained from PFMs has become an urgent task. Hydrogen isotope (HI) exchange proven to be effective method for removal. However, microscopic mechanisms, particularly role effects, remain insufficiently understood. For first time, this work employs path integral molecular dynamics account...
W-Ni-Fe alloy with excellent properties is widely used in aerospace and nuclear industry. In the actual production, hydrogen often as protective gas for preparation of alloy, which inevitably induces embrittlement. This work mainly investigates transport behavior deuterium W-2.46Ni-1.04Fe through gas-driven permeation testing range 723 K~923 K, contributes to understand embrittlement provides important guidance improving material properties. The resultant parameters...