The corrosion behavior of 20G and TP347H materials was investigated in molten LiCl-NaCl-KCl salt. The corrosion rates of these materials in molten chloride salt are high and are strongly affected by the alloying surface oxide formation. The 20G shows uniform surface corrosion with almost no protective oxide formation on the surface. In contrast, the austenitic steel TP347H exhibits better corrosion resistance in molten chloride salts due to its high Cr content. Owing to the highly corrosive nature of molten chloride salts, the Cl⁻ in molten salt could react with oxides and alloy, inducing intergranular corrosion of austenitic steel in molten chloride salt environments.

Transmission electron microscopy (TEM), second ion mass spectrum (SIMS) and atom probe tomography (APT) techniques are used to study the Li ion distribution in the oxide formed on the rolling surface (S_N) of Zircaloy-4 corroded in lithiated water with 0.01 M LiOH at 633 K/18.6 MPa. The results showed that the Li ions segregated in the grain boundaries and subgrain boundaries in the oxide film, but nearly no Li ions were found in the oxide around the interface between the oxide and matrix. Finally, we discussed the mechanism of the LiOH influence on the corrosion resistance of Zircaloy-4.

Supercritical water oxidation is a promising technology for decomposition of industrial wastewater and sludge. However, the system is operated under high temperature and pressure (usually higher than 500 °C and 25 MPa). Corrosion of component materials and salt deposition may lead to leakage or even the burst of the pressure vessel in the SCWO system, resulting in a high level of worry about the safe operation of the system. In this paper, the safety management and accident- control strategies are introduced according to a commercial-scale SCWO plant in China. The safety management strategy refers to the special design and operation strategies of some facilities. Different types of potential accidents are analyzed and the coping strategies for accidents of different levels of severity are described in detail. The strategies are valuable for the safe operation of commercial SCWO plants and other plants operated in high temperature and high pressure.

A 700 V pulsed laser was used for the surface treatment of Zircaloy-4. Phases including the treatment layer, morphology and the distributions of alloying elements of the treatment layer were detected via X-ray diffraction (XRD), scanning electron microscope (SEM) and transmission electron microscope (TEM). The results showed that the laser surface treatment (LST) layer is also α-Zr phase layer, the morphology of the treatment layer was “cauliflower-like” and the Fe-Cr precipitates in the LST layer were dissolved. The corrosion tests of the LST and the no-laser surface treatment (NLST) specimens were conducted in steam at 1100 °C using TGA (NETZSCH STA 449 F). The results showed that LST can enhance the corrosion resistance of the Zircaloy-4 in high-temperature steam. More microcracks distributed in the oxide film formed on the NLST specimen than on the LST specimen. And the volume fraction of the tetragonal zirconia (t-ZrO₂) phase in the oxide film on the surface of the LST specimen was higher than that of NLST specimen. The main reason for this phenomena could be attributed to the dissolving Fe-Cr precipitates and higher solid solution of Fe and Cr in the laser treatment layer.

Lead-cooled fast reactors exhibit strong inherent safety performance and good economic features, while material degradation due to corrosion and irradiation is still challenging. Oxide dispersion-strengthened steels are one of the promising candidates for fuel cladding materials. The effects of both irradiation and corrosion on ODS steel need to be further studied. In this work, MX-ODS steel was irradiated by Fe ions at 500 °C up to 46 dpa. Later, the as-received specimen and the irradiated specimen were used to conduct corrosion tests in oxygen-saturated Pb at 550 °C for 1 h. In the as-received specimen, discontinuous oxides penetrated by Pb and Pb in contact with steel matrix were observed, demonstrating unsatisfactory corrosion resistance of the material. However, in the irradiated specimen after corrosion experiment, a protective oxide layer formed and prevented Pb attack. The oxidation behavior differences between the two specimens can be attributed to the defects produced by irradiation and the structural discrepancy in oxides caused by the formation process. A possible mechanism of irradiation on the corrosion is discussed. In the as-received specimen, Fe atoms loss led to voids in the oxides, and lead penetrated the oxides through these voids. In the irradiated specimen, defects left by previous irradiation helped to form a more uniform oxide layer. The adhesive outer magnetite oxide and the Fe ions generated from where grain boundary oxidation developed retarded the presence of voids and made the oxide layer protective.