The Zr-2.5%Nb alloy used to manufacture the pressure tubes for the CANDU 600 power plant experiences continual corrosion when the reactor’s fuel channels are functioning normally. Analysis of the hydrogen accumulation circumstances, the creation of zirconium hydride platelets, evaluation of the platelets’ reorientation in the thermo–mechanical field at volumetric flaws, and determination of the hydrogen morphology are therefore significant goals of the worldwide research being conducted in this field. In this paper, pressure tube alloy samples with artificial V flaws were hydrided by the technique of electrolytic deposition of a hydrogen film layer, followed by a specific thermal treatment. Micrographs of the Zr-2.5%Nb alloy samples in the radial cross-section were taken to examine the hydride reorientation phenomenon that occurred when the samples were heated under mechanical stress. The complex stress field around a tip flaw promoted the formation of some circles from hydride filaments. To characterize the zirconium hydride morphology in the reorientation zone at the volumetric flaw tip, image analysis and processing techniques were applied. To better characterize the reorientation zone from the point of view of embrittlement and mechanical behavior, this paper proposes a novel metric for the morphology of hydrides located in the reorientation zone. This work’s findings are beneficial for the investigation of CANDU fuel channels’ structural integrity.

The Inconel 718 superalloy, a precipitation-hardenable material, is of particular interest for applications involving components operating under extreme conditions due to its excellent mechanical properties, high corrosion resistance at temperatures up to 700 °C, and good workability. At high temperatures, thermal transfer processes are crucial for temperature distribution across the component’s section, structural transformations, and variations in the alloy’s properties. The history of accidental overheating events is critical for the microstructure and properties of the alloy. Studies on thermal transfer in the Inconel 718 alloy available in the literature typically focus on the alloy in its as-delivered state. The experimental research presented in this paper examines the influence of repeated overheating history on the thermal diffusivity of the alloy.

The precipitation-hardenable nickel-based superalloy Rene 41 exhibits remarkable mechanical characteristics and high corrosion resistance at high temperatures, properties that allow it to be used in high-end applications. This research paper presents findings on the influence of thermal shocks on its microstructure, hardness, and thermal diffusivity at temperatures between 700 and 1000 °C. Solar energy was used for cyclic thermal shock tests. The samples were characterized using microhardness measurements, optical microscopic analysis, scanning electron microscopy coupled with EDS elemental chemical analysis, X-ray diffraction, and flash thermal diffusivity measurements. Structural transformations and the variation of properties were observed with an increase in the number of shocks applied at the same temperature and with temperature variation for the same number of thermal shocks.

The Zr-2.5%Nb alloy, from which the pressure tubes of the CANDU 600 power plant at the Cernavodă NPP are made, is permanently subjected to a corrosion process during the normal operation of the fuel channels in the reactor. At the same time, the formation of specific flaws on the inner surface of the pressure tube under the conditions of prolonged irradiation and absorption of deuterium from the cooling agent, heavy water, leads to the initiation of precursor centers for cracking under mechanical load, called in the specialized literature type cracking Delayed Hydride Cracking (DHC). Among the phenomena that can contribute to the initiation and triggering of the DHC mechanism is the phenomenon of fatigue under mechanical load under the conditions of the formation of defects such as "bearing pad fretting flaw" or "debris fretting flaw". These defects are the most important defects revealed by periodic inspections of CANDU fuel channels, as they are the precursors to DHC propagating cracks. Therefore, an important objective in the studies carried out internationally is to analyze and model the consequences of the fatigue phenomenon of the pressure tube, which may contain this type of flaw, under the conditions of hydrogen accumulation, the formation of zirconium hydride platelets and their reorientation in the thermo-mechanical field. The paper presents part of the experimental studies regarding the performance of thermo mechanical treatments for the reorientation of zirconium hydride platelets under high mechanical stresses in the Zr-2.5%Nb alloy, the micro-structural analysis and the definition of the HCC metric for characterizing the morphology of the hydrides located in the tube wall of CANDU pressure, carrying out analyzes by the finite element method of the mechanical stress field developed at the top of the defects of the "bearing pad fretting flaw" or the complex defect "bearing pad fretting flaw with debris fretting flaw". Finally, based Correspondence address: alexandra.jinga@nuclear.ro 266 Jinga A. at al. / Study of the influence of zirconium hydrides on fatigue crack… on the data obtained from the mechanical fatigue tests on samples with the mentioned type defects, the fatigue lifetime was determined in the format of the corresponding Basquin type equation, „the amplitude of the alternating stress versus the number of cycles to failure". The results of the work are useful in the structural integrity analysis of the CANDU fuel channels.

The paper presents a comparative experimental study of heat-transfer behavior in three alloys considered candidate materials for nuclear reactors: the austenitic stainless steel 316L, Zircaloy-4 (currently used in CANDU reactors), and an ODS alloy with a ferritic matrix. The investigation was conducted across five temperature intervals, each sample being subjected to a thermal shock through short-term overheating to the upper limit of its respective interval. The variation of thermal diffusivity in the three alloys was determined as a function of both measurement temperature and applied thermal shock, and trends in heat-transfer behavior were compared across the five temperature ranges. The experimental results show that up to 400 °C, Zircaloy-4 exhibits the highest thermal diffusivity, followed by the ODS alloy, with the lowest values measured for 316L steel. At approximately 450 °C, the ratio between 316L and the ODS alloy reverses. Beyond this point, increasing the temperature up to 900 °C is accompanied by a continuous rise in thermal diffusivity for both 316L stainless steel and Zircaloy-4. In contrast, for the ODS steel, increasing temperature leads to a continuous decrease in thermal diffusivity, reaching a minimum near the Curie point. The novelty of the study lies in the comparative assessment of the influence of temperature on the heat-transfer process in three alloys relevant to nuclear energy, covering the operating temperature ranges of CANDU and ALFRED reactors, as well as potential accidental overheating up to 900 °C. A particular feature of the work is the prior application of a short-duration overheating step produced using solar energy. The results are relevant not only for nuclear reactors but also for other high-temperature applications in corrosive environments.

In some Generation IV reactor configurations, embrittlement of the liquid metal can manifest in various forms, and this behaviour strongly depends on the specific solid–liquid couple. For the ALFRED demonstrator, which will be built at the RATEN ICN site in Romania, the study of embrittlement induced by molten lead in 316L stainless steel at temperatures of 350–450 °C is of interest. The purpose of the paper is to evaluate the effect of liquid metal embrittlement on the fracture mechanics properties of 316L tested in liquid lead. To do this, the “Normalisation Data Reduction Technique” (ASTM E1820) is used to obtain the J-R resistance curve for 316L steel in molten lead. In this way, the fracture mechanics parameters were obtained, indicating the fracture toughness of 316L steel in liquid lead with a saturated oxygen concentration at 350 °C. Optical and SEM examinations complement the analyses.