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Abstract 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. [ABSTRACT FROM AUTHOR]

Irradiation induced microstructural changes in the NF709 austenitic stainless steel were investigated under 1 MeV Kr ion irradiations to a dose of 3 dpa at room temperature (RT), 300, 400, 500 and 600 °C, and a dose of 20 dpa at RT. The irradiation-induced defects and the stability of precipitates were characterized with transmission electron microscopy (TEM). Frank dislocation loops were observed in all the irradiated samples, and the loop sizes were much larger at 600 °C than those at lower temperatures. “Raft” defect structures, formed through self-alignment of small dislocation loops, were also observed in all irradiated samples. M 23 C 6 precipitates were amorphorized under irradiations at RT and 300 °C, but remained to be crystalline at 400 °C and above. MX precipitates were stable under irradiations at RT up to 20 dpa, and at temperatures below 600 °C to 3 dpa. At 600 °C, some MX precipitates were observed to dissolve during in-situ irradiation, suggesting possible precipitate instability at this irradiation temperature. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Alloy 709 stainless steel (A709 SS) has gained an increased interest in nuclear applications due to its advanced properties at high temperatures over the widely used Type 316 SS. However, how this alloy behaves under irradiation with concurrent environmental factors, such as mechanical deformation and/or thermal annealing, has not been studied. In this work, we used in-situ ion irradiation inside a transmission electron microscope (TEM) to investigate the effects of irradiation on the microstructures of A709 SS samples with different prior processing histories, including as-received, annealed, and creep-deformed. It was found that, compared to room temperature irradiation, 600°C irradiation led to larger dislocation loop sizes and lower loop densities. With 600°C irradiation, the creep-deformed sample, which had the highest precipitate density and the highest dislocation density, had the smallest loop size and the lowest loop density compared to others. This observation has been rationalized with quantitative evaluation of the sink strength factors in each sample. This study shows that in-service deformation can significantly affect the irradiation performance in structural materials at reactor operating conditions. [ABSTRACT FROM AUTHOR]

To evaluate the potential of high entropy alloys (HEAs) for nuclear applications, two HEAs, Al 0.3 CoCrFeNi 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 L1 2 nanoparticles was observed in Al 0.3 CoCrFeNi. 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 Al 0.3 CoCrFeNi 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. Bright-field TEM micrographs of Al0.3CoCrFeNi, CoCrFeMnNi and 316H SS irradiated with 1 MeV Kr ions at 500 °C to 1 dpa. All images were taken at similar foil thickness near 110 zone axis with g = 200. The two high entropy alloys exhibited significantly larger size and smaller density of dislocation loops as compared with the 316H stainless steels. Image 1 [ABSTRACT FROM AUTHOR]

Additive manufacturing as a new processing technique can produce unique microstructure that is difficult to achieve using conventional techniques. In this study, we have investigated the creep behavior of 316 L stainless steel produced by a laser powder bed fusion process at temperatures of 550, 600 and 650 °C and stresses between 175 and 300 MPa. We found that additively-manufactured 316 L stainless steel had a higher stress dependence of the minimum creep rate than conventionally-made Type 316 SS, which could be attributed to the dislocation cell structure resulting from the printing process. The dislocation cell structure was unstable under creep, evolving into a uniform dislocation structure under the test conditions. While internal porosity in AM 316 L SS may serve as nucleation sites of creep voids and may be responsible for a relatively lower creep life, additively-manufactured 316 SS did not show inferior creep ductility when compared with conventionally-made 316 SS. The creep life of AM 316 L SS could be improved by stabilizing dislocation cell structure and/or reducing internal porosity through an optimized additive manufacturing process. [ABSTRACT FROM AUTHOR]

Single crystal CeO2, as a surrogate material to UO2, was irradiated with 500keV xenon ions at 800°C while being observed using in situ transmission electron microscopy (TEM). Experimental results show the formation and growth of defect clusters including dislocation loops and cavities as a function of increasing atomic displacement dose. At high dose, the dislocation loop structure evolves into an extended dislocation line structure, which appears to remain stable to the high dose levels examined in this study. A high concentration of cavities was also present in the microstructure. Despite high atomic displacement doses, the specimen remained crystalline to a cumulated dose of 5×1015 ions/cm2, which is consistent with the known stability of the fluorite structure under high dose irradiation. Kinetic Monte Carlo calculations show that oxygen mobility is substantially higher in hypo-stoichiometric UO2/CeO2 than hyper-stoichiometric systems. This result is consistent with the ability of irradiation damage to recover even at intermediate irradiation temperatures. [ABSTRACT FROM AUTHOR]

Refractory alloys consisting of reduced activity elements are attractive for nuclear applications to allow safe post-service recycling and meet fusion reactor requirements. Here, ion irradiation response was studied for a newly developed refractory high entropy alloy, HfTaTiVZr, with low activation constituent elements. Ion irradiation led to around 20% hardening for the high entropy alloy compared to 50% hardening for stainless steel under identical conditions. Superior irradiation resistance of the concentrated alloy was explained by the reduced mobility of point defects and self-healing ability. [ABSTRACT FROM AUTHOR]