Search Results (62)

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

Nuclear fuel pellets are subject to stress for long periods during the in-pile operation, and this study on high-temperature creep performance is of great significance for predicting the in-pile behaviors and safety evaluation of fuel elements. In the present study, a mixture of ZrC (50 wt%), SiC (46 wt%), and Si (4 wt%) powder was ball-milled for 24 h and then evaporated to obtain ZrC–SiC composite material. ZrC–SiC composite was adopted as the matrix, with ZrO₂ surrogate kernel TRSIO particles and dispersion coated particle fuel pellets prepared with different TRISO packing fractions using the Spark Plasma Sintering (SPS) process. This study on compressive creep performances was conducted under a temperature range of 1373–2073 K and a stress range of 5–250 MPa, elucidating the creep behavior and mechanism of dispersed coated particles fuel pellets, and obtaining the variation laws of key parameters such as creep stress exponents and activation energy with TRISO packing fraction. The results showed that creep stress exponents of the surrogate fuel pellets are between 0.89 and 2.12. The activation energies for high temperature–low stress creep (1873–2073 K, 5–50 MPa) are 457.81–623.77 kJ/mol, and 135.14–161.59 kJ/mol for low temperature high stress creep (1373–1773 K, 50–250 MPa). Based on the experimental results, a high-temperature creep model was established, providing a valuable reference for the research and application of a ceramic matrix dispersed with coated particle fuels. 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

This study investigates the enhancement of thermal conductivity in silicon carbide (SiC) matrix pellets for accident-tolerant fuels via atomic layer deposition (ALD) of alumina (Al₂O₃) coatings. Pressure-holding ALD protocols ensured precursor saturation, enabling precise coating control (0.09 nm/cycle). The ALD-coated Al₂O₃ layers on SiC particles were found to be more uniform while minimizing surface oxidation compared to traditional mechanical mixing. Combined with yttria (Y₂O₃) additives and spark plasma sintering (SPS), ALD-coated samples achieved satisfactory densification and thermal performance. Results demonstrated that 5~7 wt.% ALD-Al₂O₃ + Y₂O₃ achieved corrected thermal conductivity enhancements of 14~18% at 100 °C., even with reduced sintering aid content, while maintaining sintered densities above 92% T.D. (theoretical density). This work highlights ALD’s potential in fabricating high-performance, accident-tolerant SiC-based fuels for safer and more efficient nuclear reactors, with implications for future optimization of sintering processes and additive formulations. 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

U₃Si₂ is a long term, accident-tolerant nuclear fuel candidate for light-water reactors because of its superior thermal conductivity and increased uranium density when compared to traditional uranium dioxide (UO₂). While reducing internal thermal stresses and increasing efficiency, U₃Si₂ exhibits energetic oxidation during certain off-normal and accident scenarios, which include coolant or steam exposure. To mitigate this, Nb is investigated as an alloy constituent to enhance corrosion resistance and increase mechanical strength. The work presented investigates the response of Nb-alloyed U₃Si₂ to steam atmospheres. A thermogravimetric analysis is conducted in flowing steam to T > 1000 °C to assess oxidation resistance. The phase characterization of as-melted, thermally annealed and post-oxidation compositions with up to 12 vol% Nb by powder X-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy is reported. Full article

Large-grained UO₂ is considered a potential accident-tolerant fuel (ATF) due to its superior fission gas retention capabilities. Irradiation experiments for cerium dioxide (CeO₂), used as a surrogate fuel, is a common approach for evaluating the performance of UO₂. In this work, spark plasma sintered CeO₂ pellets with varying grain sizes (145 nm, 353 nm, and 101 μm) and a relative density greater than 93.83% were irradiated with 4 MeV Xe ions at a fluence of 2 × 10¹⁵ ions/cm² at room temperature, followed by annealing at 600 °C for 3 h. Microstructure, including dislocation loops and bubble morphology of the irradiated samples, has been characterized. The average size of dislocation loops increases with increasing grain size. Large-sized dislocation loops are absent near the grain boundary because the boundary absorbs surrounding defects and prevents the dislocation loops from coalescing and expanding. The distribution of bubbles within the grain is uniform, whereas the large-sized and irregularly shaped xenon bubbles observed in the small grain exhibit pipe diffusion along the grain boundaries. The bubble diameter in the large-grained pellet is the smallest. As the grain size increases, the volumetric swelling of the irradiated pellets decreases while the areal density of Xe bubbles increases. Elemental segregation, which tends to occur at dislocation loops and grain boundaries, has been analyzed. Large-grained CeO₂ pellet with lower-density grain boundaries exhibits better resistance to volumetric swelling and elemental segregation, suggesting that large-grained UO₂ pellets could serve as a potential ATF. 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

The simulation of fuel composition requires coupled calculations of neutron transport and burnup. It is generally assumed that the neutron flux density and cross-sections remain constant within a burnup step. However, when there are strong absorber poisons present, the reaction rates of the absorbers change too rapidly over time, necessitating extremely fine step sizes to ensure computational accuracy, which in turn leads to low computational efficiency. As a type of accident tolerant fuel (ATF), fully ceramic micro-encapsulated (FCM) fuel is a promising new type of nuclear fuel. Accelerated algorithms for burnup calculations of FCM fuel containing gadolinium isotopes have been developed based on the ALPHA code, including the projected predictor–corrector (PPC), the log-linear rate (LLR), and the high-order predictor–corrector (HOPC) methods (including CE/LI, CE/QI, LE/LI, and LE/QI). The performances of different algorithms under the two forms of Gd₂O₃ existence were analyzed. The numerical results show that the LE/QI method performs the best overall. For Gd₂O₃ existing in both forms, the LE/QI algorithm can maintain accuracy with a burnup step size of up to 1.0 GWd/tU, keeping the infinite multiplication factor k_inf within 100 pcm, and it exhibits high accuracy in simulating the atomic number densities of Gd-155 and Gd-157 throughout the burnup process. Full article

A 700 V pulsed laser was used for the surface treatment of Zircaloy-4. Phases including the treatment layer, morphology and the distributions of alloying elements of the treatment layer were detected via X-ray diffraction (XRD), scanning electron microscope (SEM) and transmission electron microscope (TEM). The results showed that the laser surface treatment (LST) layer is also α-Zr phase layer, the morphology of the treatment layer was “cauliflower-like” and the Fe-Cr precipitates in the LST layer were dissolved. The corrosion tests of the LST and the no-laser surface treatment (NLST) specimens were conducted in steam at 1100 °C using TGA (NETZSCH STA 449 F). The results showed that LST can enhance the corrosion resistance of the Zircaloy-4 in high-temperature steam. More microcracks distributed in the oxide film formed on the NLST specimen than on the LST specimen. And the volume fraction of the tetragonal zirconia (t-ZrO₂) phase in the oxide film on the surface of the LST specimen was higher than that of NLST specimen. The main reason for this phenomena could be attributed to the dissolving Fe-Cr precipitates and higher solid solution of Fe and Cr in the laser treatment layer. 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

Cr coatings with a thickness of about 19 μm were synthesized on Zr-4 cladding using plasma-enhanced arc ion plating. A Zr-Cr micro-diffusion layer was formed via Cr ion cleaning before deposition to enhance the interface bonding strength. Cr coatings exhibit an obvious columnar crystal structure with an average grain size of 1.26 μm using SEM (scanning electron microscopy) and EBSD (electron backscatter diffraction) with a small amount of nanoscale pores on the surface. A long-term aqueous test at 420 ± 3 °C, 10.3 ± 0.7 MPa and isothermal oxidation tests at 900~1300 °C in air were conducted to evaluate the Cr-coated Zr-4 cladding. All the results showed that the Cr coatings had a significant protective effect to the Zr-4 alloy. However, the corrosion deterioration mechanism is different. A gradual thinning of the Cr coating was observed in a long-term aqueous test, but a cyclic corrosion mechanism of void initiation–propagation–cracking at the oxide film interface is the main corrosion characteristic of the Cr coating in isothermal oxidation. Different corrosion models are constructed to explain the corrosion mechanism. 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