Wei-Ying Chen

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.

Microstructural changes resulted from neutron irradiation and post-irradiation annealing in a high-temperature ultra-fine precipitate strengthened (HT-UPS) stainless steel were characterized using transmission electron microscopy (TEM) and atom probe tomography (APT). Three HT-UPS samples were neutron-irradiated to 3 dpa at 500?°C, and after irradiation, two of them were annealed for 1?h?at 600?°C and 700?°C, respectively. Frank dislocation loops were the dominant defect structure in both the as-irradiated and 600?°C post-irradiation-annealed (PIAed) samples, and the loop sizes and densities were similar in these two samples. Unfaulted dislocation loops were observed in the 700?°C PIAed sample, and the loop density was greatly reduced in comparison with that in the as-irradiated sample. Nano-sized MX precipitates were observed under TEM in the 700?°C PIAed sample, but not in the 600?°C PIAed or the as-irradiated samples. The titanium-rich clusters were identified in all three samples using APT. The post-irradiation annealing (PIA) caused the growth of the Ti-rich clusters with a stronger effect at 700?°C than at 600?°C. The irradiation caused elemental segregations at the grain boundary and the grain interior, and the grain boundary segregation behavior is consistent with observations in other irradiated austenitic steels. APT results showed that PIA reduced the magnitude of irradiation induced segregations.

This study characterizes the microstructural evolution of single-phase complex concentrated solid- solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mecha- nisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr 18 Fe 27 Mn 27 Ni 28 , Cr 15 Fe 35 Mn 15 Ni 35 , and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the Cr- FeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr ++ ions to vari- ous levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irra- diation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr 18 Fe 27 Mn 27 Ni 28 and 10 nm in Cr 15 Fe 35 Mn 15 Ni 35 . This difference in loop growth kinet- ics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.#171118

In this study, a novel alloy of modified Inconel 718 produced by laser powder bed fusion is studied before and after in-situ Kr irradiation up to 3 dpa at 200 and 450 °C. Before irradiation, the microstructure consists of dislocation cells having a misorientation angle less than 5° and with an average size of ~500 nm. There are also second phase particles of MC type carbides, Laves phase and oxides such as Y-O, Y-(Ti)-Al-O. While the microstructure consists of stacking fault tetrahedra, faulted and perfect loops after irradiation at 200 °C, dislocation loops are the primary defects at 450 °C. With increasing dose, the size of the defects remains similar at 200 °C while it increases at 450 °C. This has been attributed to the existence of vacancy type defects at 200 °C and the different defect transport mechanisms at different temperatures. Moreover, matrix and second phase particle compositions seem to be similar after irradiation. The sink strengths of the structures have been calculated and superior radiation resistance of this alloy has been attributed to the existence of fine cell boundaries stabilized by the second phase particles produced by additive manufacturing.

In this work, the ion irradiation responses of a Fe-based nanostructured ferritic alloy or ‘NFA’ (Fe–9Cr–2W–0.2V–0.4Ti–0.3Y2O3) and a Cr3C2@SiC-NFA composite were assessed. In-situ ion irradiation with TEM observation was carried out by using 1 MeV Kr++ ions at doses of 0, 1, 3, 5, 10 dpa and temperatures of 300 °C and 450 °C. Both the NFA and Cr3C2@SiC-NFA samples showed significant dislocation density after 10 dpa at 300 °C. However, the Cr3C2@SiC-NFA composite showed a significantly lower dislocation loop density and a smaller average loop size during the irradiation at 450 °C as opposed to the NFA. At 300 °C, 1/2<111> type dislocation loops were observed in both the NFA and Cr3C2@SiC-NFA samples. Interestingly, at 450 °C, <100> type loops were dominant in the NFA sample while 1/2<111> type loops were still dominant in the Cr3C2@SiC-NFA sample. The results were discussed based on the large surface sink effects and enhanced interstitial-vacancy recombination at higher temperatures. The additional Si element in the Cr3C2@SiC-NFA sample might have played a significant role in determining the dominant loop types.

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.

High entropy alloys (HEAs) have been considered for applications in nuclear reactors due to their promising mechanical properties, corrosion and radiation resistance. It has been suggested that sluggish diffusion kinetics and lattice distortion of HEAs can enhance the annihilation of irradiation-induced defects, giving rise to a higher degree of tolerance to irradiation damage. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages over conventional austenitic stainless steels (SS), we performed in-situ ion irradiation experiments with 1 MeV krypton at 300 °C on two HEAs and a 316H SS under an identical irradiation condition. The irradiation introduced a high density of dislocation loops in all materials, and the microstructural evolution as a function of dose was similar for HEAs and 316H SS. Nanoindentation tests showed that the degree of irradiation hardening was also comparable between them. The similar microstructural evolution and irradiation hardening behavior between the HEAs and 316H indicate that, at low temperatures (≤300 °C), the irradiation damage of fcc alloys is not sensitive to compositional variation and configurational entropy.

To evaluate the potential of high entropy alloys (HEAs) for nuclear applications, two HEAs, Al0.3CoCrFeNi and CoCrFeMnNi, and a conventional Type 316H stainless steel (SS) were irradiated with 1 MeV krypton ions at 500 °C up to 1 dpa, and examined in-situ with a transmission electron microscope (TEM). After irradiation, a high density of ordered L12 nanoparticles was observed in Al0.3CoCrFeNi. In contrast, no phase transformation was observed in CoCrFeMnNi and 316H SS. In the thin foil regions of TEM samples, stacking-fault tetrahedra were observed in the HEAs. In the thick foil regions, interstitial dislocation loops were observed for all alloys, and the areal loop density increased linearly with foil thickness. The Al0.3CoCrFeNi had the largest loop size (the lowest density), followed by the CoCrFeMnNi and then the 316H SS. The higher loop nucleation rate in the 316H SS was attributed to carbon content. The degree of irradiation hardening was slightly lower for the HEAs than for the 316H SS, which is a promising sign for the nuclear application of HEAs at high temperatures.

Fe and Fe-Cr (Cr = 10-16 at.%) specimens were neutron-irradiated at 300 °C to 0.01, 0.1 and 1 dpa. The TEM observations indicated that the Cr significantly reduced the mobility of dislocation loops and suppressed vacancy clustering, leading to distinct damage microstructures between Fe and Fe-Cr. Irradiation-induced dislocation loops in Fe were heterogeneously observed in the vicinity of grown-in dislocations, whereas the loop distribution observed in Fe-Cr is much more uniform. Voids were observed in the irradiated Fe samples, but not in irradiated Fe-Cr samples. Increasing Cr content in Fe-Cr results in a higher density, and a smaller size of irradiation-induced dislocation loops. Orowan mechanism was used to correlate the observed microstructure and hardening, which showed that the hardening in Fe-Cr can be attributed to the formation of dislocation loops and &#945;&#8242; precipitates.

The objective of this study is to understand irradiation-induced and assisted defect evolution in binary body center cubic (bcc) Fe-based alloys. The broader class of bcc ferritic alloys are leading candidates for advanced nuclear fission and fusion applications, in part due to their exceptional void swelling resistance. However, their irradiated microstructure evolution is sensitive to solute species present, since these solutes can act as traps for irradiation-induced defects due to the surrounding tensile or compressive stress fields. Here, three alloys (Fe- 9.5%Cr, Fe-4.5%P, and Fe-2.3%N) are selected for study because they systematically exhibit varying solute sizes and solute positions (i.e., substitutional or interstitial). Ex situ and in situ ion irradiations reveal that Fe-P has a considerably finer and denser population of irradiation-induced defects than Fe-Cr and Fe-N at the same irra- diation conditions, which is attributed to strong defect trapping at undersized substitutional P, consequently hindering the development of extended defects. Meanwhile, oversized substitutional solutes (e.g., Cr) and interstitial solutes (e.g., N) may also suppress dislocation loop development due to weak solute-defect trapping.