Liquid lead-bismuth eutectic (LBE) is one of the candidate materials for advanced nuclear systems. The structural materials used in contact with LBE are selected according to the resistance to liquid metal corrosion, irradiation embrittlement, and compatibility with the coolant. However, simultaneous presence of mechanical strain and LBE environment can induce liquid metal embrittlement (LME) in these materials. In this study, a specimen from candidate ferritic-martensitic steel T91 was tested by Constant Extension Rate Tensile (CERT) test exposed to PbBi environment with oxygen concentration 6 × 10⁻⁶ wt % at 300 °C up to rupture. Post-test examination using scanning electron microscopy (SEM) showed a deep crack indicating features of LME in the plastic strained region of the tested specimen. Further investigations focused on characterization of the fracture path and microstructure determination using focused ion beam (FIB) and energy dispersive X-ray spectrometry/electron backscatter diffraction (EDX/EBSD). This observation revealed that the dominant LME failure mode of the observed crack is translath or transgranular and the crack stopped at the high-angle grain boundary. The role of oxides in the crack initiation is discussed.

Development of liquid lead cooled nuclear systems requires consideration of compatibility issues with the construction materials. In order to understand the corrosion or passivation behavior of the 316 L austenitic stainless steel, the steel specimens were exposed for 1000 h in liquid lead with 1 × 10−7 wt % oxygen level at 480 °C in static and flowing (velocity 1.6 m/s) conditions. Post-test microscopy investigation using scanning electron microscope with focused ion beam (FIB) was performed and it demonstrated significant differences in the formation of thin oxide layers in the two conditions. Maximum oxide thickness was 2 μm in the static lead (Pb) and less than 0.1 μm in the flowing Pb. In the static condition, oxide scale was not continuous and local corrosion attack was indicated; but in flowing condition the oxide layer was continuous without any corrosion attacks.

Fuel failure caused by fretting damage to cladding remains a relevant issue despite decades of research and development aimed at enhancing the physical parameters of fuel. This paper presents the results of experiments conducted at the Research Centre Řež on Zr-1%Nb alloy tube specimens covered with protective coatings made of chromium (Cr) and nitrogen (N) compounds. The experiments involved debris-fretting tests under dry conditions at room temperature as well as microscopic measurements of groove depths. A detailed analysis was performed using the Scanning Electron Microscopy, Energy-Dispersive X-Ray Spectroscopy, Electron Backscatter Diffraction, and Focused Ion Beam techniques. The objectives of the tests were (1) to compare the debris-fretting resistance between the reference Zr-1%Nb specimens and those of the same alloy coated with various compositions, and (2) to demonstrate the positive effects of coating applications on the endurance of fuel cladding. The conducted analysis revealed a significant advantage in using cladding with a thin, wear-resistant layer compared to standard cladding material, with the CrN-coated specimens exhibiting 36 times better fretting resistance.

Nuclear grade fuel claddings act as a barrier against release of fuel particles into the coolant water during nuclear power plant operation, handling, and wet or dry storage of the spent fuel rods. The integrity of claddings is always a critical issue during the reactor operation, storage, or loss of coolant accidents (LOCA). After Fukushima disaster, cladding materials are widely studied, as the neutron transparent zirconium (Zr)-based alloys or accident tolerant fuel claddings, which should reduce the high-temperature oxidation rate of Zr-based alloys and enhance the accident tolerance. One part of the long-term investigation of Zr-based claddings in the Research Center Řež contributes to the characterization of material behavior after creep testing and high-temperature oxidation, i.e., conditions that may occur during reactor operation or storage of the spent fuel. Then, the microstructure response can be investigated by means of standard as well as advanced microscopy methods. In this study, we introduce the methodology of the cladding microstructure complex characterization demonstrated on the reference (nonirradiated) Zr-1%Nb alloy after creep testing. In the Zr-1%Nb microstructure, the hydrogen is present under some conditions as the hydride phases ZrH or ZrH2 that cause the mechanical properties degradation. As well, the hydride reorientation can pose a risk. Therefore, currently used methods of hydrides evaluation are presented. Furthermore, we introduce the microstructure analysis of the zircalloy-4 cladding after high-temperature oxidation at the temperature 1300 °C simulating LOCA. For this study, the new automatic chemical analysis procedure was developed. The main issue is the determination of the oxygen concentration and alloying elements behavior after critical increase of oxygen in the Zr-alloy tube at LOCA, also concerning the hydrogen content in the microstructure. Based on the element concentration gradients, the phase transitions from the oxide ZrO2 to the prior β-Zr can be determined and contributes to the phase diagrams experimental verification. In our contribution, the significant role of the electron microscopy methods “from micro to nano-scale” in the nuclear research is emphasized in these two research topics.

Research and development of materials in the nuclear industry, including the assessment of irradiated components in nuclear power plants, rely heavily on research infrastructures that facilitate the preparation and analysis of radioactive samples. The metallographic preparation of samples must be carried out in shielded, hermetically sealed boxes and hot cells, as a large amount of dusty radioactive particles is released during the sample preparation process. Small samples, thin films for scanning and transmission electron microscopy (SEM, TEM), and powdered samples from biological shielding concretes are prepared in glove boxes, where a constant negative pressure is maintained. The enclosures of the boxes are constructed of thick-walled steel plates to shield against the ionizing radiation emitted by radioactive samples, protecting personnel. Subsequent microscopic analyses allow for the assessment of material degradation in operational nuclear power plant components caused by radiation-induced microstructure damage. These analyses focus both on existing materials, with the aim of extending the lifespan of nuclear power plants, and on newly tested materials irradiated as part of domestic and international programs, including those conducted at research reactors like LVR-15, operated by the Řež Research Center (CVŘ). Different analytical requirements call for the preparation of various types of samples. At CVŘ, X-ray diffraction analysis of powdered samples is primarily used to evaluate structural and phase changes in the cement and aggregate of concrete structures caused by radiation aging, which can impact the overall integrity of the structure. Monitoring these changes and predicting material behavior are essential for evaluating the safety, stability and durability of concrete used in biological shielding, containment structures, spent nuclear fuel storage pools and future deep radioactive waste repositories. Thin films for electron microscopy are prepared specifically to assess detailed changes, such as radiation-induced microstructure damage in reactor internals or fuel cladding, which result in dimensional changes and the degradation of mechanical properties due to neutron radiation.