For lead‐cooled fast reactors and accelerator‐driven subcritical systems, the surface corrosion behavior of candidate structural materials in lead–bismuth eutectic (LBE) is a key issue, which determines whether the material is applicable. The candidate materials of two typical MAX phases, Ti3SiC2 and Ti3AlC2, were immersed in static LBE with saturated oxygen concentration at 500°C for up to 3000 h. The corrosion behaviors of Ti3SiC2 and Ti3AlC2 were analyzed by scanning electron microscope, energy‐dispersive X‐ray, X‐ray diffraction, and Raman spectra. The experimental results showed that elements interdiffusion between LBE and sample matrix occurred on both Ti3SiC2 and Ti3AlC2 surfaces, which led to the formation of the diffusion layer. The dominant component of the diffusion layer is PbTiO3, which makes the corroded surface fragile in a stress environment. Besides, there were differences in structures of corroded sample surfaces between Ti3SiC2 and Ti3AlC2. The corrosion layer of Ti3SiC2 consisted of two layers, while only one single layer formed on Ti3AlC2 surfaces. The stable oxide layer consisting of SiO2 and TiO2 can protect Ti3SiC2 samples from further LBE corrosion and maintain the integrity of the surfaces. For Ti3AlC2 samples, it is hard to form a continuous Al2O3 protective layer, thus no stable oxide layer was detected on the corroded surfaces. Compared with Ti3AlC2, Ti3SiC2 showed better corrosion resistance in LBE.

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In this paper, two typical candidate structural materials of 316L and T91 with different surface roughnesses were studied at temperatures from 200–500 ℃. The surface with different roughness was prepared by mechanical polishing on the sandpapers with particle sizes from 400 to 2000 mesh. The wetting test was carried out in a smart contact angle measuring device by using the sessile-drop method. Meanwhile, the microstructure of the liquid-solid surface was analyzed by scanning electron microscope (SEM). The results show that the surfaces of both materials are non-wetting to LBE in the tested temperature range. The contact angles of LBE drop on material surfaces decrease with increasing temperature in general. However, it appears to increase significantly at 400 ℃ for both two materials. Besides, the decrease of surface roughness can effectively inhibit the wettability of LBE on the material surface. In addition, compared with 316L, the wetting of the LBE to T91 surface is better, indicating the higher tendency of LME for T91 in practical application. These results can provide references for the prediction of the LME behavior of structural materials.

The reactor pressure vessel (RPV) is one of the most critical equipment in the pressurized water reactor, and its structural integrity is the key factor that determines the operational safety and service life of the reactor. In practical applications, the aging degree of RPV can be evaluated through the ductile-brittle transition temperature (DBTT) curve of the Charpy impact specimen pre-placed in RPV. However, due to the space limitation inside the reactor core, the available irradiation surveillance specimens are limited for mechanical testing. Especially, most reactors have faced the problem of life extension in recent years, and the impact data of the irradiation surveillance specimen is an important basis for the life extension of the reactor. One of the solutions is to reconstitute new Charpy specimens from the impacted ones to obtain more impact data. In this paper, the basic methods for the reconstitution of RPV material Ni-Cr-Mo-V steel are studied. By testing the hardness change of the Charpy impact broken specimen along the length direction, the maximum value of the plastic deformation zone of the impact fracture is obtained. Besides, based on the Gurson-Tvergaard-Needleman (GTN) model, the impact process of the material in the upper shelf temperature region is calculated by ABAQUS numerical simulation. Compared with the tested microhardness results of the material, the maximum length of the insert section of Charpy impact specimen reconstitution is confirmed. It shows that for Ni-Cr-Mo-V steel, the maximum length of the plastic deformation zone at the upper shelf temperature region is about 7 mm, and the insert length of the reconstituted specimens can be selected to 20 mm. The results can be used as an important reference for establishing the fabrication standard of the reconstituted Charpy specimen of the reactor pressure vessel.

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