Yong Yang

AGR-1 and AGR-2 tristructural-isotropic (TRISO) fuel particles were fabricated using slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed ²³⁵U enrichments. Extensive microstructural and analytical characterizations were conducted to correlate those differences with the fuel kernels’ responses to neutron irradiations in terms of irradiated fuel microstructure, fission products’ chemical and physical states, and fission gas bubble evolutions. The studies used state-of-the-art transmission electron microscopy (TEM) equipped with energy-dispersive x-ray spectroscopy (EDS) via four silicon solid-state detectors with super sensitivity and rapid speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen in order to create representative irradiation conditions with fuel burnup in the range of 10.8 to 18.6% fissions per initial metal atom (FIMA) and time-average volume-average temperatures varying from 1070 to 1287°C. The ²³⁵U enrichment was 19.74 wt.% and 14.03 wt.% for the AGR-1 and AGR-2 fuel kernels, respectively. The TEM results showed significant microstructural reconstructions in the irradiated fuel kernels from both the AGR-1 and AGR-2 fuels. There are four major phases: fuel matrix of UO₂ and UC, U₂RuC₂, and UMoC₂—in the irradiated AGR-2 fuel kernel. Zr and Nd form a solid solution in the UC phase. The UMoC₂ phase often features a detectable concentration of Tc. Pd was mainly found to be located in the buffer layer or associated with fission gas bubbles within the UMoC₂ phase. EDS maps qualitatively show that rare-earth fission products (Nd et al.) preferentially reside in the UO₂ phase. In contrast, in the irradiated AGR-1 fuel kernel, no U₂RuC₂ or UMoC₂ precipitates were positively identified. Instead, there was a high number of rod-shaped precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The differences in irradiated fuel kernel microstructural and micro-chemical evolution when comparing AGR-1 and AGR-2 TRISO fuel particles may result from a combination of irradiation temperature, fuel geometry, and chemical composition. However, irradiation temperature probably plays a more deterministic role. Limited electron energy-loss spectroscopy (EELS) characterizations of the AGR-2 fuel kernel show almost no carbon in the UO₂ phase, but a small fraction of oxygen was detected in the UC/UMoC₂ phase.

Cast austenitic stainless steels (CASSs) are used in the cooling system of light water reactors (LWRs) for components with complex shapes, such as pump casings, valve bodies, coolant piping, etc. The CF grades of CASS alloys are the cast equivalents of 300-series stainless steels (SSs) and show excellent mechanical properties and corrosion resistance. In contrast to the fully austenitic microstructure of wrought SSs, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite and are vulnerable to thermal aging embrittlement. The service performance of CASS alloys is of concern after long-term exposure to high-temperature coolant. In this work, we studied the effects of thermal aging and ferrite content on the fracture resistance of CASS alloys. Crack growth rate and fracture toughness J–R curve tests were performed on aged and unaged CASS alloys in simulated light water reactor environments. The impact of thermal aging on the cracking susceptibility was investigated and the effect of ferrite content was evaluated. Significant embrittlement was observed in the CASS alloys after aging at 400°C. To understand the embrittlement mechanism, microstructural characterizations were performed with transmission electron microscope. The thermal aging produced G-phase precipitates and phase separation in the ferrite, but did not affect the microstructure of austenite. Consequently, the ferrite was hardened considerably after thermal aging while the hardness of austenite phase remained unchanged. The difference in hardness created a high incompatible strain at the interface between ferrite and austenite, leading to fracture at phase boundaries.

Cast austenitic stainless steels (CASS) possess excellent corrosion resistance and mechanical properties and are used alongside with wrought stainless steels (SS) in light water reactors for primary pressure boundaries and reactor core internal components. In contrast to the fully austenitic microstructure of wrought SS, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite. The delta ferrite is critical for the service performance since it improves the strength, weldability, corrosion resistance, and soundness of CASS alloys. On the other hand, the delta ferrite is also vulnerable to embrittlement when exposed to reactor service temperatures and fast neutron irradiations. In this study, the combined effect of thermal aging and neutron irradiation on the degradation of CASS alloys was investigated. Neutron-irradiated CASS specimens with and without prior thermal aging were tested in simulated light water reactor environments for crack growth rate and fracture toughness. Miniature compact-tension specimens of CF-3 and CF-8 alloys were tested to evaluate the extent of embrittlement resulting from thermal aging and neutron irradiation. The materials used are static casts containing more than 23% delta ferrite. Some specimens were thermally aged at 400 °C for 10,000 hours prior to the neutron irradiation to simulate thermal aging embrittlement. Both the unaged and aged specimens were irradiated at ∼320°C to a low displacement damage dose of 0.08 dpa. Crack growth rate and fracture toughness J-integral resistance curve tests were carried out on the irradiated and unirradiated control samples in simulated light water reactor environments with low corrosion potentials. While no elevated crack propagation rates were detected in the test environments, significant reductions in fracture toughness were observed after either thermal aging or neutron irradiation. The loss of fracture toughness due to neutron irradiation seemed more evident in the samples without prior thermal aging. Transmission electron microscope (TEM) examination was carried out on the thermally aged and neutron irradiated specimens. The result showed that both neutron irradiation and thermal aging can induce significant changes in the delta ferrite. A high density of G-phase precipitates was observed with TEM in the thermally aged specimens, consistent with previous results. Similar precipitate microstructures were also observed in the neutron-irradiated specimens with or without prior thermal aging. A more extensive precipitate microstructure can be seen in the samples subjected to both thermal aging and neutron irradiation. The similar precipitate microstructures resulting from thermal aging and neutron irradiation are consistent with the fracture toughness results, suggesting a common microstructural origin of the observed embrittlement after thermal aging and neutron irradiation.

Nuclear energy provides more than 10% of electrical power internationally and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M₂₃C₆ precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments.