Search Results (42)

FeCrAl exhibits excellent resistance to high temperatures, corrosion, and irradiation, making it a prime candidate material for accident-tolerant fuel (ATF) cladding. This study investigates the high-temperature creep behavior of FeCrAl alloys with grain sizes of 12.0 μm and 9.9 μm under temperatures ranging from 450 °C to 650 °C and applied stresses between 75 and 200 MPa. The texture, grain morphology, grain orientation, and dislocation density of FeCrAl were characterized by electron backscatter diffraction (EBSD). The results indicate that temperature, applied stress, and grain size are the primary factors governing high-temperature creep behavior. The material texture showed no significant difference before and after creep. Large grains tend to engulf smaller ones during the creep process at lower temperatures and stresses, reducing the proportion of low-angle grain boundaries (LAGBs). In contrast, at higher temperatures or under higher stress, dislocations proliferate within grains, leading to a significant increase in the number of LAGBs. As the applied stress increases, the dominant creep mechanism tends to convert from grain boundary sliding to dislocation motion. Moreover, higher temperatures or smaller grain sizes lower the critical stress required to activate dislocation motion and significantly increase dislocation density, severely degrading the creep resistance. Full article

This study investigates the role of Al alloying in tailoring the oxidation resistance of AlTiCrZrNbTa refractory high-entropy alloy (RHEA) coatings on Zry-4 substrates under high-temperature steam environments. Coatings with varying Al contents (0–25 at.%) were deposited via magnetron sputtering and subjected to oxidation tests at 1000–1100 °C. The results demonstrate that Al content critically governs oxidation kinetics and coating integrity. The optimal performance was achieved at 10 at.% Al, above which a dense, continuous composite oxide layer (Al₂O₃, TiO₂, Cr₂O₃) formed, effectively suppressing oxygen penetration and maintaining strong interfacial adhesion. Indentation tests confirmed enhanced mechanical integrity in Al-10 coatings, with minimal cracking post-oxidation. Excessive Al alloying (≥17 at.%) led to accelerated coating oxidation. This work establishes a critical Al threshold for balancing oxidation and interfacial bonding, providing a design strategy for developing accident-tolerant fuel cladding coatings. Full article

The development of FeCrAl alloys has commenced for use as nuclear fuel cladding material, intended to serve as an enhanced accident-tolerant alternative to Zr-based alloys. In this study, the Fe-13Cr-4Al alloy, specifically designed for advanced accident-tolerant fuel (ATF) cladding, was carefully prepared through vacuum induction melting and hot-working processes. Mechanical properties and serrated flow behavior of this alloy were investigated through tensile tests at temperatures ranging from 200 to 800 °C. Intriguingly, serrations emerged within a specific temperature range, accompanied by unique mechanical behavior characteristics indicative of dynamic strain aging (DSA). Additionally, the alloy’s fracture modes showed a transition from a mix of ductile and cleavage fracture features to fully ductile fracture as the temperature increased. This study offers insights into the mechanical properties and serration behaviors of FeCrAl alloys, highlighting their potential for use in nuclear fuel cladding. Full article

In the nuclear industry, coated cladding is a topical problem and it is chosen as the near-term and most promising ATF (Accident-Tolerant Fuel) cladding concept. The main objective of this concept is to enhance the accident tolerance of nuclear power plants and accordingly, the performance of cladding is expected to be improved. This work assesses the corrosion performance of a Zircalloy-4 alloy coated with a thin chromium coating by MS (magnetron sputtering), tested under a CANDU (CANada Deuterium Uranium) reactor primary circuit simulated condition (LiOH solution, 10 MPa, 310 °C, pH = 10.5). The anticorrosive performance is evaluated by a gravimetric analysis, a metallographic analysis, X-ray diffraction, electronic microscopy, and electrochemical methods. A four times less gain mass was noticed compared to uncoated Zircaloy-4, indicating a smaller corrosion rate. The SEM micrographs illustrate that the coatings are still adherent, and they are keeping the initial morphological characteristics during the autoclaving process. A SEM cross-section analysis shows values of the thickness of the coatings between 0.8 and 1.46 µm. By XRD, the presence of Cr₂O₃ oxide is identified. Electrochemical testing confirms good stability and good corrosion performance of Cr coating over time under autoclave conditions. Full article

After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron–chromium–aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2. Full article

Silicon carbide (SiC) ceramic matrix composite (CMC) cladding is currently being pursued as one of the leading candidates for accident-tolerant fuel (ATF) cladding for light water reactor applications. The morphology of fabrication defects, including the size and shape of voids, is one of the key challenges that impacts cladding performance and guarantees reactor safety. Therefore, quantification of defects’ size, location, distribution, and leak paths is critical to determining SiC CMC in-core performance. This research aims to provide quantitative insight into the defect’s distribution under multi-scale characterization at different length scales before and after different Transient Reactor Test Facility (TREAT) irradiation tests. A non-destructive multi-scale evaluation of irradiated SiC will help to assess critical microstructural defects from production and/or experimental testing to better understand and predict overall cladding performance. X-ray computed tomography (XCT), a non-destructive, data-rich characterization technique, is combined with lower length scale electronic microscopic characterization, which provides microscale morphology and structural characterization. This paper discusses a fully automatic workflow to detect and analyze SiC-SiC defects using image processing techniques on 3D X-ray images. Following the XCT data analysis, advanced characterizations from focused ion beam (FIB) and transmission electron microscopy (TEM) were conducted to verify the findings from the XCT data, especially quantitative results from local nano-scale TEM 3D tomography data, which were utilized to complement the 3D XCT results. In this work, three SiC samples (two irradiated and one unirradiated) provided by General Atomics are investigated. The irradiated samples were irradiated in a way that was expected to induce cracking, and indeed, the automated workflow developed in this work was able to successfully identify and characterize the defects formation in the irradiated samples while detecting no observed cracking in the unirradiated sample. These results demonstrate the value of automated XCT tools to better understand the damage and damage propagation in SiC-SiC structures for nuclear applications. Full article

The nuclear industry is focusing some efforts on increasing the operational safety of current nuclear reactors and improving the safety of future types of reactors. In this context, the paper is focused on testing and evaluating the corrosion behavior of a thin chromium coating, deposited by Electron Beam Physical Vapor Deposition on Zy-4. After autoclaving under primary circuit conditions, the Cr-coated Zy-4 samples were characterized by gravimetric analysis, optical microscopy, SEM with EDX, and XRD. The investigation of the corrosion behavior was carried out by applying three electrochemical methods: potentiodynamic measurements, EIS, and OCP variation. A plateau appears on the weight gain evolution, and the oxidation kinetics generate a cubic oxidation law, both of which indicate a stabilization of the corrosion. By optical microscopy, it was observed a relatively uniform distribution of hydrides along the samples, in the horizontal direction. By SEM investigations it was observed that after the autoclaving period, the coatings with thickness from 2 to 3 µm are still adherent and maintain integrity. The XRD diffractograms showed a high degree of crystallinity with the intensity of chromium peaks higher than the intensity of zirconium peaks. Electrochemical results indicate better corrosion behavior after 3024 h of autoclaving. Full article

After the Fukushima nuclear accident, the development of new accident-tolerant fuel cladding materials has become a research hotspot around the world. Due to its outstanding corrosion resistance, radiation resistance, and creep properties at elevated temperatures, the oxide dispersion strengthened (ODS) FeCrAl alloy, as one of the most promising candidate materials for accident-tolerant fuel cladding, has been extensively studied during the past decade. Recent research on chemical composition design as well as its effects on the microstructure and mechanical properties has been reviewed in this paper. In particular, the reasonable/optimized content of Cr is explained from the aspects of oxidation resistance, radiation resistance, and thermal stability. The essential role of the Al element in oxidation resistance, high-temperature stability, and workability was reviewed in detail. The roles of oxide-forming elements, i.e., Y (Y₂O₃), Ti, and Zr, and the solid solution strengthening element, i.e., W, were discussed. Additionally, their reasonable contents were summarized. Typical types of oxide, i.e., Y–Ti–O, Y–Al–O, and Y–Zr–O, and their formation mechanisms were also discussed in this paper. All aspects mentioned above provide an important reference for understanding the effects of composition design parameters on the properties of nuclear-level ODS FeCrAl alloy. Full article

Since 2011, there has been an international effort to evaluate the behavior of newer fuel rod materials for the retrofitting of existing light water reactors (LWR). These materials include concepts for the cladding of the fuel and for the fuel itself. The materials can be broadly categorized into evolutionary or improved existing materials and revolutionary or innovative materials. The purpose of the newer materials or accident-tolerant fuels (ATF) is to make the LWRs more resistant to loss-of-coolant accidents and thus increase their operation safety. The benefits and detriments of the three main concepts for the cladding are discussed. These include (i) coatings for existing zirconium alloys; (ii) monolithic iron–chromium–aluminum alloys; and (iii) composites based on silicon carbide. The use of ATF materials may help extend the life of currently operating LWRs, while also being a link to material development for future commercial reactors. Full article

Preparing protecting coatings on the surface of Zr claddings has been regarded as one of the accident tolerant fuel (ATF) strategies. In this study, a series of nitride-modified layers were in situ grown by hollow cathode plasma nitriding on the surface of Zr metal. The influence of nitriding currents and time on the phases, composition, microstructure and corrosion resistance of the modified layers was investigated by X-ray diffraction (XRD), X-ray Photoemission Spectroscopy (XPS), transmission electron microscope (TEM), scanning electron microscopy (SEM) with energy dispersive spectrometer (EDS) and potentiodynamic polarization curves. The ZrO₂ layer with loose microstructure and cracks prefers to form under low nitriding current of 0.4 A, which also causes poor corrosion resistance. The high temperature caused by high nitriding currents (0.6 A and 0.8 A) promote the formation of compact nanocrystalline layers, made up of nitride and oxynitride. Below the nanocrystalline layer, it is Zr₂N caused by N penetration. Besides this, a double-layer structure of the nanocrystalline layer, i.e., an equiaxed crystal zone with a grain size of ~10–50 nm on the surface and a long strip grain region beneath it was observed. The compact nitride/oxynitride layer with excellent interface bonding can improve the corrosion resistance effectively. Full article

In order to meet the growing energy demand, more environmentally friendly and efficient GEN-IV reactors have emerged. Additionally, nuclear structural materials need larger more safety margins for accident scenarios as a result of the Fukushima accident. In order to extend the failure time and lessen the effect of accidents, this design method for accident-tolerant fuel materials calls for cladding materials to maintain good corrosion resistance and mechanical properties during a beyond design basis accident (BDBA). Accidents involving nuclear reactors would undoubtedly result in higher temperatures, which would make it much harder for materials to withstand corrosion. Oxide dispersion strengthened (ODS) FeCrAl alloys have shown promise as candidate materials because of their extraordinarily slow reaction rates under high-temperature steam. However, the addition of the Al element renders the alloy’s high-temperature mechanical properties insufficient. In particular, studies on the alloy’s creep properties are extremely rare, despite the fact that the creep properties are crucial in the real service environment. Therefore, this paper focuses on the creep properties of ODS FeCrAl alloy, summarizes and analyzes the research results of this material, and provides a reference for future research and applications. Full article

To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect experiments, and more complex phenomena can be investigated using integral-effect tests, the behavior of an entire system as complex as a nuclear power plant core must be investigated using computer code modeling. One of the most commonly used computer codes for the assessment of severe accidents is MELCOR 2.2. In version 18019, the authors enabled the modeling of the behavior of the nuclear fuel with FeCrAl cladding (namely, alloy B136Y3) for the first time, using the GOX model. The ability of this model to reasonably accurately predict the behavior of FeCrAl cladding in accident conditions with quenching was verified in this work by modeling the QUENCH-19 experiment carried out in the Karlsruhe Institute of Technology on the QUENCH experimental device and by subsequent comparison of the MELCOR calculation results with the experiment. This article proves that the GOX model can be used to evaluate the behavior of FeCrAl cladding and that the results can be considered conservative. Full article

FeCrAl is regarded as one of the most promising cladding materials for accident-tolerant fuel at nuclear fission reactors due to its comprehensive properties of inherent corrosion resistance, excellent irradiation resistance, high-temperature oxidation resistance, and stress corrosion cracking resistance. In this work, the irradiation response of FeCrAl irradiated by 2.4 MeV He²⁺ ions with a fluence of 1.1 × 10¹⁶ cm⁻² at room temperature was studied using X-ray diffraction, transmission electron microscopy, and nanoindentation. The characterization results of structural and mechanical properties showed that only black-dot defects exist in irradiated FeCrAl samples, and that the hardness of the irradiated samples was 11.5% higher than that of the unirradiated samples. Similar to other types of radiation defects, black-dot defects acted as fixed defect obstacles and hindered the movement of slip dislocations moving under the applied load, resulting in a significant increase in the hardness of FeCrAl. Importantly, this work points out that irradiation-induced black-dot defects can significantly affect the mechanical properties of materials, and that their contribution to radiation hardening cannot be ignored. Full article

FeCrAl alloys have been suggested as one of the most promising fuel cladding materials for the development of accident tolerance fuel. Creep is one of the important mechanical properties of the FeCrAl alloys used as fuel claddings under high temperature conditions. This work aims to elucidate the deformation feature and underlying mechanism during the creep process of nanocrystalline FeCrAl alloys using atomistic simulations. The creep curves at different conditions are simulated for FeCrAl alloys with grain sizes (GS) of 5.6–40 nm, and the dependence of creep on temperature, stress and GS are analyzed. The transitions of the mechanisms are analyzed by stress and GS exponents firstly, and further checked not only from microstructural evidence, but also from a vital comparison of activation energies for creep and diffusion. Under low stress conditions, grain boundary (GB) diffusion contributes more to the overall creep deformation than lattice diffusion does for the alloy with small GSs. However, for the alloy with larger GSs, lattice diffusion controls creep. Additionally, a high temperature helps the transition of diffusional creep from the GB to the dominant lattice. Under medium- and high-stress conditions, GB slip and dislocation motion begin to control the creep mechanism. The amount of GB slip increases with the temperature, or decreases with GS. GS and temperature also have an impact on the dislocation behavior. The higher the temperature or the smaller the GS is, the smaller the stress at which the dislocation motion begins to affect creep. Full article

Accident-tolerant fuel concepts have been developed recently in diverse research programs. Recent research has shown clear advantages of Cr-coated Zr cladding over bare cladding tubes regarding oxidation behavior under the design basis loss-of-coolant accident condition. However, limited data are available about the hydriding behavior of the Cr coating. For that purpose, Cr-coated Zricaloy-4 tubes were tested to investigate the effects of hydriding, oxidation, and postquench ductility behavior on coated Zr cladding. A high-power impulse magnetron sputtering (HiPIMS) process was used to produce a high-density coating on the Zircaloy-4 tube surface. Coated and uncoated Zircaloy-4 tube specimens underwent one-sided hydriding in a tube furnace filled with pure hydrogen gas at 425 °C. The tubing specimen ends were sealed with Swagelok plugs before the hydriding runs. For uncoated specimens, H analysis of the hydrided specimens indicated that the H content increased as the test time and initial pressure increased. However, almost no change was observed for the coated specimens that were hydrided under the same test conditions. After one-sided hydriding, the hydrided coated and uncoated specimens were exposed to steam at high temperatures for two-sided oxidation studies to simulate accident conditions. The coated specimens showed a slower oxidation: oxygen pickup was 50% lower than the uncoated specimens tested under the same conditions. Ring compression testing was performed to evaluate the embrittlement behavior of the Cr-coated specimens after hydriding and oxidation. The results indicated that the HiPIMS coating provides excellent protection from hydriding and oxidation at high temperatures. Full article