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Materials Today · 2025 · IF 22

Continuous production of Cl-terminated MXenes by bubbling chemical vapor growth in molten salt

Jikai Zhang,Yi Tao,Yuxuan Ye,Haiyang Wang,Xinping Wu,Zicheng Luo,Yu Shi,Qi Zhao,Zongju Cheng,Yu Guo,Bin Li,Bo Mai,Qinyou An,Xiansheng Zhang,Zhiguo Du,Shubin Yang

Although 2D transition-metal carbides (MXenes) are usually synthesized by selective etching and chemical vapor deposition approaches, they are limited by the slow reaction kinetics, resulting in their continuous production challenging. Here, we demonstrate a bubbling chemical vapor growth (BCVG) method to continuously produce Cl-terminated MXenes by the reaction of CH 4 and TiCl 4 chemical bubbles with titanium powder in a molten salt. In the molten reaction system, titanium powder would effectively react with TiCl 4 to form highly active [TiCl 6 ] y – with low Ti Cl bond dissociation energies. The multivalent [TiCl 6 ] y – could efficiently incorporate the carbon species from CH 4 to form [Ti 6 CCl 18 ] clusters on the surface of bubbles in a low energy barrier, efficiently promoting the reaction kinetics. Subsequently, Cl-terminated MXenes were produced based on the following condensation reaction of [Ti 6 CCl 18 ] clusters. The resultant MXene (Ti 2 CCl x ) exhibits an expanded structure with strong interfacial polarization between its basal planes and the polymer matrix, delivering a good electromagnetic wave absorption performance with a minimum reflection loss (RL min ) value of –52.2 dB at a thin thickness of 2.84 mm.

CAS / MDL: 7789-24-4
研究领域: 材料科学 · 化学工程
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SEPARATION AND PURIFICATION TECHNOLOGY · 2024 · IF 8.1

Preliminary research on selective extraction of 99Mo from fuel salt for molten salt reactor

Yongsheng Niu,Lixin Sun,Haiying Fu,Qiang Dou,Yuan Qian,Qingnuan Li

Molybdenum-99 ( 99 Mo) is one of the most important radionuclides for nuclear medicine. This isotope of molybdenum is mainly produced as a result of fission and decay processes occurring in nuclear reactors. Molten salt reactors (MSRs), a unique reactor with the fuel dissolved in a molten salt , have attracted attention as a source of 99 Mo. In this work, the feasibility of fluoride volatility process as a potential technique for selective extraction of trace amount of 99 Mo from fuel salt in the form of volatile MoF 6 by nitrogen trifluoride (NF 3 ) is illustrated. The results showed that in the case of using LiF-BeF 2 -ZrF 4 -UF 4 -MoF 3 (64.4–29.7–5.0–0.5–0.4 mol.%) molten salt as the simulated fuel salt, MoF 3 reacted with NF 3 to form MoF 6 at or above 570℃, while the conversion of UF 4 to volatile UF 6 required higher temperatures (≥650℃), when both exposed to 20 % NF 3 /Ar mixture. Based on above “cold” experiments, fluoride volatility process in LiF-BeF 2 -ZrF 4 molten salt containing irradiated UF 4 was carried out, and the extraction rate of 99 Mo was determined by detecting the change of its γ activity in the molten salt. The results indicated that 99 Mo could be extracted from the molten salt containing a quantity of uranium with the extraction rate of 98 % under current experimental conditions, and the content of uranium in collected 99 Mo product was only 0.9 wt%.

研究领域: 核医学 · 化学工程 · 核能工程
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JOURNAL OF NUCLEAR MATERIALS · 2025 · IF 3.2

Solubility and dissolution kinetics of uranium trifluoride in 2LiF-BeF2 molten salt

Zhangqi Li,Shizhuan Xu,Peng Wang,Jinzhao Zou,Jian Chen,Changqing Cao,Jin Lin

Although uranium trifluoride (UF 3 ) holds promise for applications in molten salt reactors, its high-temperature physicochemical properties in molten salt systems have yet to be thoroughly explored. In this study, the solubility and dissolution kinetics of UF 3 in molten 2LiF-BeF 2 (66–34 mol %, FLiBe) eutectic salt were investigated using the isothermal saturation method within the temperature range of 823 K to 973 K. High-purity UF 3 compacts (∼99.70 %) were synthesized via an optimized solid-phase reaction protocol and subsequent compacting. The UF 3 -saturated FLiBe molten salts were prepared by immerse dissolution of nickel mesh-wrapped UF 3 compacts and bulk uranium in the molten salt, eliminating the filtration step. Experimental findings showed that the dissolution equilibrium of UF 3 in the FLiBe salt was 120 h. The solubility of UF 3 exhibited a linear increase (R 2 > 0.99) from 4.16 wt. % to 12.60 wt. % as the elevation of temperature. Crystallographic analysis confirmed that the typical UF 3 phase was the only uranium-bearing phase throughout the dissolution process, while deconvolution X-ray photoelectron spectroscopy (XPS) verified the exclusive presence of the U 3+ species under all temperature conditions. Notably, the dissolution kinetics conformed to a mass transport control model.

CAS / MDL: 7789-24-4
研究领域: 核化学 · 材料科学 · 物理化学 · 能源技术
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