Patricie Halodová

This article presents the development and commissioning of a supercritical carbon dioxide (sCO2) autoclave for high-temperature corrosion testing of structural materials under representative operating conditions. The facility is intended to support materials development and corrosion assessment for sCO2-based energy conversion systems, with particular relevance to advanced nuclear applications, including Generation IV reactors and small modular reactors (SMRs). The autoclave enables dynamic corrosion testing at elevated temperatures and pressures, with independent control and continuous monitoring of temperature, pressure, and moisture content in the sCO2 environment. The performance of the system is demonstrated through corrosion testing of high-temperature-resistant Ni-based alloys and the advanced austenitic steel Sanicro 25. The observed corrosion behavior, governed by coupled oxidation and carburization processes, is consistent with previously reported studies, validating the capability of the experimental setup. The developed autoclave allows materials to be evaluated under conditions closely approaching those expected in practical sCO2 power cycles, providing a reliable platform for materials screening and selection for advanced nuclear energy systems.

This study examines the effects of gamma and combined neutron?gamma irradiation on cement paste and mortar, focusing on the micromechanical response of the main hydrated phases and associated microstructural changes. Samples were irradiated under both neutron?gamma conditions. Nanoindentation was performed to assess the mechanical behavior of individual phases. Results show that irradiation induces stiffening of the primary hydrates and causes compaction and densification of C?S?H gels, including partial breakdown of fibrillar structures and reorganization of colloids into clustered, rosette-like morphologies. Irradiation also led to the formation of microcracks within the cement paste. While these microcracks did not affect the micromechanical response of individual phases, they are likely responsible for the reductions in compressive strength observed in cement composites under irradiation. The findings demonstrate that irradiation produces irreversible nanoscale and microscale modifications in cementitious materials, and that local phase stiffening combined with microcrack formation provides a mechanistic explanation for the degradation of bulk mechanical properties.

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.

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.

Concrete exposed to irradiation from nuclear reactor undergoes atomic-level degradation due to the influence of neutrons, secondary gamma rays, and associated gamma heating. The radiationinduced damage decreases the mechanical properties of the concrete and affects its performance as a construction material. The concrete samples analysed in this work were prepared from blended Portland cement CEM II 32.5R mixed with siliceous aggregates with a size of 0–2.5 mm. The material underwent irradiation in the LVR-15 research reactor in Research Centre Rež to the target neutron fluence of 1.6–1.8 · 1019 n cm−2 (E>0.1MeV) and received the gamma dose of ⁓ 500 MGy. The temperatures during irradiation were kept within a range of 50–60 °C. The mechanical properties of the individual phases in the cementitious matrix after neutron irradiation were assessed by the nanoindentation. To track the evolution of the nanomechanical properties of individual components under radiation, a comparison was made between the mechanical properties of neutron-irradiated material and pristine concrete.

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.

The presented article outlines and describes the processes involved in the creation and testing of the passivation layer that was created during hot functional tests (HFTs) at the third unit of the nuclear power plant (NPP) Mochovce. The article is divided into several sections. In Sec. 1 is the description of the hot tests on the third block, in particular the chemical regime, i.e., the formation of hydrogen or the necessary adjustment of the pH value. Sec. 2 of the article describes the evaluation of the passivation layer using instrumental techniques, i.e., Raman microscopy, scanning electron microscopy (SEM), and X-ray photoelectron spectroscopy. The results obtained by using these techniques provided a comprehensive view of the resulting layer and clearly defined its' structure. The measurement revealed that a passivation layer was formed, the inner layer of which is enriched with a chrome element. The upper part of the passivation layer was composed of iron oxides, mainly magnetite.