Kun Mo

Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe–Cr model alloys (10–16 at.%) irradiated at 300 and 450 °C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0–1.3 nm were observed. The precipitate density varied significantly from 1.1 × 1023 to 2.7 × 1024 1/m3, depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe–10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.

The microstructure modifications of a high-energy Xe implanted U3Si2, a promising accident tolerant fuel candidate, were characterized and are reported upon. The U3Si2 pellet was irradiated at Argonne Tandem Linac Accelerator System (ATLAS) by an 84 MeV Xe ion beam at 300 °C. The irradiated specimen was then investigated using a series of transmission electron microscopy (TEM) techniques. A dense distribution of bubbles were observed near the range of the 84 MeV Xe ions. Xe gas was also found to accumulate at multiple types of sinks, such as dislocations and grain boundaries. Bubbles aggregated at those sinks are slightly larger than intragranular bubbles in lattice. At 300 °C, the gaseous swelling strain is limited as all the bubbles are below 10 nm, implying the promising fission gas behavior of U3Si2 under normal operating conditions in light water reactors (LWRs).

The study of grain growth kinetics in nano-grained UO2 samples is reported. Dense nano-grained UO2 samples with well-controlled stoichiometry and grain size were fabricated using the spark plasma sintering technique. To determine the grain growth kinetics at elevated temperatures, a synchrotron wide-angle X-ray scattering (WAXS) study was performed in situ to measure the real-time grain size evolution based on the modified Williamson-Hall analysis. The unique grain growth kinetics of nanocrystalline UO2 at 730 °C and 820 °C were observed and explained by the difference in mobility of various grain boundaries.

The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

The nickel-base Alloy 617 has been considered as the lead candidate structural material for the intermediate heat exchanger (IHX) of the Very-High-Temperature Reactor (VHTR). In order to assess the long-term performance of Alloy 617, thermal aging experiments up to 10,000 h in duration were performed at 1000 °C. Subsequently, in-situ synchrotron wide-angle X-ray scattering (WAXS) tensile tests were carried out at ambient temperature. M23C6 carbides were identified as the primary precipitates, while a smaller amount of M6C was also observed. The aging effects were quantified in several aspects: (1) macroscopic tensile properties, (2) volume fraction of the M23C6 phase, (3) the lattice strain evolution of both the matrix and the M23C6 precipitates, and (4) the dislocation density evolution during plastic deformation. The property?microstructure relationship is described with a focus on the evolution of the M23C6 phase. For aging up to 3000 h, the yield strength (YS) and ultimate tensile strength (UTS) showed little variation, with average values being 454 MPa and 787 MPa, respectively. At 10,000 h, the YS and UTS reduced to 380 MPa and 720 MPa, respectively. The reduction in YS and UTS is mainly due to the coarsening of the M23C6 precipitates. After long term aging, the volume fraction of the M23C6 phase reached a plateau and its maximum internal stress was reduced, implying that under large internal stresses the carbides were more susceptible to fracture or decohesion from the matrix. Finally, the calculated dislocation densities were in good agreement with transmission electron microscopy (TEM) measurements. The square roots of the dislocation densities and the true stresses displayed typical linear behavior and no significant change was observed in the alloys in different aging conditions.

In situ X-ray diffraction and small angle scattering measurements during tensile tests were performed on 9–12% Cr ferritic/martensitic steels. The lattice strains in both particle and matrix phases, along two principal directions, were directly measured. The load transfer between particle and matrix was calculated based on matrix/particle elastic mismatch, matrix plasticity and interface decohesion. In addition, the void or damage evolution during the test was measured using small angle X-ray scattering. By combining stress and void evolution during deformation, the critical interfacial strength for void nucleation was determined, and compared with pre-existing void nucleation criteria. These comparisons show that models overestimate the measured critical strength, and require a larger particle size than measured to match the X-ray observations.

The microstructure investigations on a high-energy Xe-implanted U3Si2 pellet were performed. The promising accident tolerant fuel (ATF) candidate, U3Si2, was irradiated by 84?MeV Xe ions at 600?°C at Argonne Tandem Linac Accelerator System (ATLAS). The characterizations of the Xe implanted sample were conducted using advanced transmission electron microscopy (TEM) techniques. An oxidation layer was observed on the sample surface after irradiation under the 10-5?Pa vacuum. The study on the oxidation layer not only unveils the readily oxidation behavior of U3Si2 under high-temperature irradiation conditions, but also develops an understanding of its oxidation mechanism. Intragranular Xe bubbles with bimodal size distribution were observed within the Xe deposition region of the sample induced by 84?MeV Xe ion implantation. At the irradiation temperature of 600?°C, the gaseous swelling strain contributed by intragranular bubbles was found to be insignificant, indicating an acceptable fission gas behavior of U3Si2 as a light water reactor (LWR) fuel operating at such a temperature.

Radiation-induced defect structures are known to elevate material yield strength and reduce material ductility so that small strains induce plastic instability. This process is commonly known as flow localization. Recent research indicates that the flow localization in face-centered cubic (FCC) materials is controlled by critical stress, the true stress at the onset of necking. Critical stress is found to be independent of irradiation dose, but have strong temperature dependence. Here simplified 9–12% ferritic/martinsetic steels are examined using X-ray diffraction and small angle scattering under in situ tensile deformation, in order to elucidate the controlling mechanisms and temperature dependence of critical stress. It is found that the critical stress for the onset of necking is linearly correlated with critical interfacial strength, which in turn determines the void nucleation. The effects of temperature and particle size on critical stress are correspondingly determined by how temperature and particle size influence the critical interfacial strength.

In this study, two kinds of 14Cr ODS alloys (14Cr-Al and 14Cr-Ti) were investigated to reveal the different effects between Al and Ti on the microstructures and mechanical properties of 14Cr ferritic ODS alloys. The microstructure information such as grains, minor phases of these two alloys has been investigated by high-energy X-ray diffraction and transmission electron microscopy (TEM). The in situ synchrotron X-ray diffraction tensile test was applied to investigate the mechanical properties of these two alloys. The lattice strains of different phases through the entire tensile deformation process in these two alloys were analyzed to calculate their elastic stresses. From the comparison of elastic stress, the strengthening capability of Y2Ti2O7 is better than TiN in 14Cr-Ti, and the strengthening capability of YAH is much better than YAM and AlN in 14Cr-Al ODS. The dislocation densities of 14Cr-Ti and 14Cr-Al ODS alloys during tensile deformation were also examined by modified Williamson-Hall analyses of peak broadening, respectively. The different increasing speed of dislocation density with plastic deformation reveals the better strengthening effect of Y-Ti-O particles in 14Cr-Ti ODS than that of Y-Al-O particles in 14Cr-Al ODS alloy.