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It is well known that the close-packed CF3-terminated solid surface is among the most hydrophobic surfaces in nature. Molecular dynamic simulations show that this hydrophobicity can be further enhanced by the atomic-scale roughness. Consequently, the hydrophobic gap width is enlarged to about 0.6 nm for roughened CF3-terminated solid surfaces. In contrast, the hydrophobic gap width does not increase too much for a rough CH3-terminated solid surface. We show that the CF3-terminated surface exists in a microscopic Cassie–Baxter state, whereas the CH3-terminated surface exists as a microscopic Wenzel state. This finding elucidates the underlying mechanism for the different widths of the observed hydrophobic gap. The cage structure of the water molecules (with integrated hydrogen bonds) around CH3 terminal assemblies on the solid surface provides an explanation for the mechanism by which the CH3-terminated surface is less hydrophobic than the CF3-terminated surface.

To develop a new separation technique for rare earth (RE) elements based on alloy diaphragms and molten salt electrolysis, a permeation experiment was conducted in LiCl–KCl eutectic melts containing RECl3 (RE = Dy, Nd, and La, 0.5 mol%) at 450 °C. In this technique, the alloy diaphragm functions as a bipolar electrode and RE ions permeate via three steps: (a) reduction of RE ions to form alloys on the anolyte side of the diaphragm, (b) diffusion of RE atoms in the diaphragm, and (c) oxidation of RE atoms to dissolve into the catholyte on the other side of the diaphragm. The experiment indicated that Dy selectively permeated through the alloy diaphragm consisting of RENi2. The permeated Dy/Nd molar ratio was ∼5, which was mainly determined by the selective alloy formation on the anolyte side of the diaphragm. However, the obtained Dy/Nd ratio was lower than that suggested by the preliminary experiment, in which a Ni substrate was simply alloyed in a LiCl–KCl–NdCl3–DyCl3 melt. Scanning electron microscopy observation and energy dispersive X-ray analysis of the diaphragm revealed that the lower selectivity compared to the preliminary experiment was attributable to the high diffusion rate of Nd atoms inside the alloy diaphragm.

The electrochemical reduction behavior of borosilicate glass, which is the main component of vitrified radioactive waste, was investigated in molten CaCl2 at 1123 K to establish a new nuclear waste disposal procedure. Cyclic voltammetry of borosilicate and silica glasses suggested that the reduction of B2O3 in borosilicate glass occurred at a more positive potential than that of SiO2. X-ray photoelectron spectroscopy confirmed that the B2O3 component was reduced to B or a B-Si compound at 0.9 V vs. Ca 2+ /Ca. The reduction products prepared by the potentiostatic electrolysis of borosilicate glass at 0.9 V had granular morphology and consisted of crystalline Si. The Al2O3 component was not reduced at 0.9 V and the Na2O component was suggested to be dissolved in molten CaCl2 during electrolysis.

The electrochemical reduction of borosilicate glass in molten CaCl 2 at 1123 K was investigated. The behaviors of the constituent elements, i.e., Si, B, Al, Na, and K, were estimated using potential-pO 2− diagrams constructed from the thermodynamic data for the species. The diagrams suggested that the first cathodic wave in the cyclic voltammogram results from the reduction of the B 2 O 3 component. The dissolution of the Na 2 O and K 2 O components, which was predicted from the diagrams, was confirmed by energy dispersive X-ray analysis of a borosilicate glass plate after immersion into molten CaCl 2 without electrolysis. The scanning electron microscopy/wavelength dispersive X-ray mappings and Raman spectrum for borosilicate glass reduced at 0.9 V vs. Ca 2+ /Ca indicated that SiO 2 and B 2 O 3 are reduced to Si and B-Si compound. The formation of Ca-Al-O compounds owing to the increase of O 2− ions is suggested. The pO 2− range during electrolysis at 0.9 V was indicated to be 2. In 2015, more than 11% of global electricity was produced by nuclear power plants.1 Indeed, nuclear power generation is expected to replace conventional fossil-fuel-based thermal power generation in order to accommodate the energy demands of the growing global population. Considering the controls on carbon dioxide emissions, nuclear energy is attractive as a low-carbon-emission power source. However, the disposal of radioactive waste is a serious problem for nuclear power generation, especially in countries frequently hit by volcanic tremors and earthquakes, such as Japan, where the selection of sites suitable for the geological disposal of radioactive wastes is very difficult. Therefore, the development of an alternative to geological disposal is required.In Japan, a new process for the disposal of radioactive wastes has been proposed. 2 In the first step of this process, long-lived fission products (LLFPs) such as 135 Cs, 79 Se, 93 Zr, and 107 Pd are separately recovered from high-level nuclear waste. They are then disposed as stable waste after their conversion by nuclear transmutation into shortlived or stable nuclides. If this process is established as a new disposal process for nuclear wastes, the amount of nuclear waste will be significantly reduced. This process has another potential advantage in that it may be used to recover valuable elements like platinum group metals, which can then be utilized for automobiles and fuel-cell catalysts. Moreover, if this process is realized, it is also applicable to the LLFPs already present in the large amount of existing vitrified waste. However, to accomplish this, each LLFPs must first be recovered from the vitrified waste.Recently, the authors proposed a new method to recover LLFPs from vitrified wastes by electrochemical reduction in molten CaCl 2 .The three-dimensional (3-D) network structure of the glass (Si-O) is electrochemically destroyed. The reduction product is then subjected to the separation of each element. This process is superior to the conventional wet process using HF ac...