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Competing microstructural evolution mechanisms can exist simultaneously when duplex stainless steels are operating for several decades in a high temperature service environment. Such competition between different microstructural evolution pathways can be difficult to ascertain using simple model alloy systems necessitating detailed microstructural analysis of phase transformation mechanisms in complex alloys. Thus, duplex stainless steels with complex but well understood chemistries were used to investigate the relative importance of different heterogeneous nucleation sites – specifically, spinodal decomposition and Cu clustering – on Ni-Si-Mn precipitation during thermal aging. Precipitation of Ni-Si-Mn particles in ferrite-bearing steels during thermal aging and irradiation can greatly change mechanical properties. Using duplex stainless steels with custom-modified compositions along with advanced microstructural characterization and first-passage kinetic Monte Carlo simulations, it is revealed that while the interface between Cr and Fe formed during spinodal decomposition can be a pathway for solute diffusion, it is not a preferred site for Ni-Si-Mn precipitation. Instead, the presence of a higher concentration of Cu leads to the formation of small Cu-rich clusters with high energy interfaces that act as nucleation sites for Ni-Si-Mn particles. These results will inform predictive models for the use of precipitation-hardened alloys for extended operation at high temperatures.

First-Passage Kinetic Monte Carlo (FPKMC) simulations of species migration in duplex stainless steels were performed in order to establish relationships between alloy content, segregation behavior, and the formation of Ni–Si–Mn rich particles in cast duplex stainless steels during thermal aging. The Ni–Si–Mn-rich second phase forms after extended aging at reactor operating temperatures and degrades alloy properties. Simulations of Ni–Si–Mn cluster formation were validated through comparison with experimental results obtained through Atom Probe Tomography (APT) on similar alloy compositions, identifying several trends. First, Cu promotes the formation of Ni–Si–Mn clusters, but only when Ni or Mn prefer segregation to the surface of Cu particles; without the segregation of these species, the critical composition for the clusters to form was not achieved. Second, the width of precipitate-denuded zones near γ/δ interfaces increases with decreasing Cu content. This finding was in strong agreement with APT and Scanning Transmission Electron Microscopy results, further validating our model. Finally, our model predicts that Ni–Si–Mn cluster formation will be the most extensive when the Si:Mn ratio is approximately 1:1 and the least extensive when one of these key elements is less concentrated (Mn ≤ 0.5 at.% or Si ≤ 0.5 at.%). The implications of these results on how to improve alloy properties are discussed.

We have made observations of noble metal phase fission-product agglomerates and gaseous xenon within the fuel-cladding interaction (FCI) zone of a high-burnup UO2 fuel. The FCI is the boundary between the UO2 pellet outer surface and the inner wall of the oxidized Zr-liner/cladding of the fuel rod. These fission-product agglomerates are well known to occur within the spent fuel matrix, and although radionuclides have been reported by others, we reveal aspects of their speciation and morphology. That they occur as discrete particles in the oxidized Zr liner, suggests the occurrence of hitherto unknown processes in the FCI zone during reactor operation, and this may have implications for the long-term storage and disposal of these types of materials. As expected, the particle agglomerates, which ranged in size from the nanometer scale to the micrometer scale, contained mainly Mo, Ru, Tc, Rh, and Pd; however, we also found significant quantities of Te associated with Pd. Indeed, we found nanometer scale separation of the distinct Pd/Te phase from the other fission products within the particles. Often associated with the particles was concentrations of uranium, sometimes appearing as a “cloud” with a tail emanating from the fuel into the oxidized cladding liner. Many of the noble metal phase particles appeared as fractured clusters separated by Xe-gas-filled voids. Possible mechanisms of formation or transport in the cladding liner are presented.

Cast austenitic stainless steels (CASS) are excellent alloys because they combine high corrosion resistance with high strength and toughness. However, they are susceptible to embrittlement upon long-term thermal aging at elevated temperatures. Thus, the microstructural evolution pathways during thermal aging need to be understood to predict and potentially prevent embrittlement. Atom probe tomography was used to identify and quantify the microstructural evolution pathways of the δ-ferrite in different CASS alloys aged for up to 10,000 h at temperatures between 290 °C and 400 °C. The four steels – CF8, CF8M, CF3, and CF3M − which vary by Mo and C concentration, each experienced spinodal decomposition of the δ-ferrite, and precipitation of G-phase clusters and Cu clusters attached to the G-phase. There were large differences in the extent of these features due to their Mo and C concentration. Using radial distribution function analysis, the interactions of constituent elements was found to determine the evolution of these features, with Mo and C specifically influencing the movement of Cr, Ni, Si, Mn, and Cu atoms due to their relative miscibility with these elements. The results will help inform predictive models for the use of duplex stainless steels for extended operation at high temperatures.

Neutron irradiation tests were carried out on 3D-printed SiC derived from binderjet additive manufacturing and chemical vapor infiltration. Irradiation was carried to 2.3 dpa over a temperature range of 400–850 °C. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. No degradation in strength was observed in the material before or after irradiation, at various temperatures, or in different orientations. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

Mechanical testing and microstructural characterization were performed on short-term thermally aged cast austenitic stainless steels (CASS) to understand the severity and mechanisms of thermal-aging degradation experienced during extended operation of light water reactor (LWR) coolant systems. Four CASS materials–CF3, CF3M, CF8, and CF8M–were thermally aged for 1500 h at 290 °C, 330 °C, 360 °C, and 400 °C. All four alloys experienced insignificant change in strength and ductility properties but a significant reduction in absorbed impact energy. The primary microstructural and compositional changes during thermal aging were spinodal decomposition of the δ-ferrite into α/α′, precipitation of G-phase in the δ-ferrite, segregation of solute to the austenite/ferrite interphase boundary, and growth of M23C6 carbides on the austenite/ferrite interphase boundary. These changes were shown to be highly dependent on chemical composition, particularly the concentration of C and Mo, and aging temperature. The low C, high Mo CF3M alloys experienced the most spinodal decomposition and G-phase precipitation coinciding the largest reduction in impact properties.

Radiation induced segregation and precipitation phenomena in an in-service boiling water reactor 304 stainless steel component were investigated using directly correlated 3D-atom probe tomography and scanning transmission electron microscopy. Significant quantitative differences in measured segregation at grain boundaries were found between the atom probe and energy dispersive spectroscopy measurements of the exact same locations. In particular, a much stronger Si segregation (~10 atomic% via atom probe versus ~4 atomic% via electron microscopy) and different Cr profile shapes were detected that are critical to models of radiation induced segregation and stress corrosion cracking behavior. These quantitative differences highlight the need for comparative microscopy and critical evaluation of limitations in each analytical method. Elemental segregation to dislocations and conjoined-clusters were also highlighted by atom probe; confirming and expanding upon what has been observed in test reactor neutron and accelerator-based ion irradiations.

Deployment of additively manufactured materials into nuclear energy systems requires investigation of the full range of the unique environmental effects on these materials. Hydrogen isotopes are common gaseous species in nuclear reactors and exhibit ample high mobility in most materials. While hydrogen isotopes mobility in wrought stainless steel is well understood, this property is not thoroughly studied for its additively manufactured variants. In this study, we investigated the deuterium permeation and retention in 316L stainless steel manufactured by laser powder bed fusion. The results showed that the deuterium permeability in the as-built additively manufactured 316L stainless steel (AM SS316L) is greater than that of the reference wrought 316L stainless steel by a factor of 2.8. However, the stress-relieved and solution-annealed AM SS316L samples exhibit lower deuterium permeability in comparison with the as-built condition. Following the solution annealing, the deuterium permeability of AM SS316L is comparable with that of the reference wrought SS316L. Deuterium retention in the as-built AM SS316L is 45% higher than that of the wrought 316L stainless steel and slightly higher than that of the two thermally annealed AM SS316L materials. Transmission electron microscopy and positron annihilation lifetime spectroscopy were used to obtain microstructural information used to determine deuterium permeation and retention behavior in the studied materials.

Fabrication of nuclear reactor components using additive manufacturing (AM) methods is now a practical option since the AM technologies have advanced to allow for building of complex parts with high quality materials. To assess the mechanical performance of printed components in reactor-relevant conditions and to build a property database for the AM 316L stainless steel (SS), mechanical testing and characterization were performed before and after neutron irradiation. Miniature tensile specimens were irradiated at the High Flux Isotope Reactor (HFIR) to 0.2 and 2 displacements per atom (dpa) at 300 and 600°C. The AM 316L SS was tested in the as-built, stress-relieved, and solution-annealed conditions, and the wrought (WT) 316L SS in solution-annealed condition as a reference alloy. The baseline test result showed that the AM 316L SS, regardless of the post-build heat treatment, had higher strength than the WT 316L SS, but similar ductility. Post-irradiation tensile testing was conducted at RT, 300°C, and 500°C for selected irradiation conditions. Neutron irradiation induced significant changes in the mechanical behavior of the AM stainless steels, including both hardening and softening. Although the as-built 316L steel after 300°C irradiation showed necking just after yielding, the overall property changes of the as-printed alloy became less significant after 600°C irradiation. Irradiation-induced ductilization was also observed after the higher temperature irradiation. In general, the strength change was smaller in the relatively stronger as-built and stress-relieved AM SSs than in the solution-annealed AM and WT SSs. These relatively lower strength 316L SSs overall retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after the higher temperature (600°C) irradiation. It is a positive assessment for the AM 316L materials that no embrittlement was observed within the test and irradiation conditions of the experiment.

Nuclear reactor environments provide a unique scientific and engineering challenge wherein materials must tolerate prolonged exposure to concurrent irradiation, elevated temperatures, and corrosive media. However, uncontrolled variability in material composition and structure often prohibits truly single-variable experiments that can reveal basic aspects of environmental damage. Magnetron sputtering is used here to provide a more controlled model system for these fundamental studies, yielding reproducible single-crystal Fe and FeCr thin films containing 8 and 18 at.% Cr. Electron microscopy is used to determine the systematic correlations between growth conditions and the resulting film microstructure and surface morphology. It is found that the substrate temperature and applied radio frequency (RF) bias can be tuned to obtain consistent homogeneous and single crystal films with a minimal amount of Ar impurities from the RF bias process. Epitaxial, single-crystal Fe films are obtained on MgO substrates at 500 °C with 10 Watt (W) RF bias deposition. However, when Cr is alloyed with Fe, higher substrate temperatures (600 °C) and applied RF biases (15 W) are required to achieve a similar epitaxial single-crystal FeCr film. Accelerated molecular dynamics simulations reveal that Cr impedes surface transport, explaining the need for higher temperature and bias during the growth of the Cr-bearing films.

Cast austenitic stainless steels (CASSs) have been extensively used for the large components of nuclear reactor primary coolant systems. Since the cast steels inevitably contain degradable metastable phases and replacement of the large coolant system components is impractical, the thermal embrittlement of CASS components has been a serious concern in the extended-term operation of nuclear power plants. This study aimed to systematically measure and analyze the effect of long-term thermal aging on the Charpy impact toughness to provide a comprehensive understanding of thermal degradation behavior and a practical aging model to predict the degree of thermal degradation in the cast stainless steels. The materials tested in the research include eight CASS alloys (two CF3s, one CF3M, three CF8s, and two CF8Ms) and two reference wrought materials (304L and 316L), in which the nominal δ-ferrite content ranges from ~2% to 33%. These stainless steels have been thermally aged at two light water reactor (LWR) temperatures (290 and 330 °C) and at two accelerated-aging temperatures (360 and 400 °C) for up to 30,000 h; these include both under-aged and over-aged conditions relative to the extended service lifetime (80 years). Charpy impact testing was performed for aged and non-aged specimens, and the impact (absorbed) energy parameters were correlated with a new aging parameter (A). Both the reduction of impact fracture toughness and the shift of ductile-brittle transition temperature were strongly dependent on the δ-ferrite content and degree of thermal aging. A linear relationship was found between the increasing rate of the index transition temperature T41J and aging parameter A; base on which an empirical model was proposed for prediction of the transition temperature as a simple function of the aging parameter (A) and δ-ferrite content (F). Finally, the critical aging parameter for embrittlement (AC) was evaluated and compared with the existing δ-ferrite content criteria.

Thermal degradation of cast stainless steels was studied to provide an extensive knowledgebase for the assessment of structural integrity during extended operations of reactor coolant systems. The CF3 and CF8 series cast stainless steels with relatively low (5–12%) δ-ferrite contents were thermally aged at 290–400 °C for up to 10,000 h and tested to measure changes in tensile and impact properties. The aging treatments caused significant reduction of tensile ductility, but only slight softening or negligible strength change. The thermal aging also caused significant reduction of upper shelf energy and large shift of ductile-brittle transition temperature (ΔDBTT). The most influential factor in thermal degradation was ferrite content because of the major degradation mechanism occurring in the phase, while the nitrogen and carbon contents caused only weak effects. An integrated model is being developed to correlate the mechanical property changes with microstructural and compositional parameters.

Metallic uranium is an important material for many applications, especially nuclear energy for low enriched metallic fuel forms and new reactor concepts. Due to the its high electropositivity, production pathways to the metallic form are limited. One of the historically most common synthesis routes involves the heating of a uranium (IV) halide, in this study UF4, with a highly electropositive metal like calcium. This synthesis is referred to as “bomb reduction” due to the temperatures and pressures released by the resulting exothermic reaction. This synthesis route is important for production, but it has also been shown to separate a decay product, thorium, from the parent uranium. This fractionation could be important for purification, but also presents the opportunity for radiochronometric dating of the reduction date. Unfortunately, little characterization has been performed on the products of this reaction. The present study performed bomb reduction on ∼10 g of thorium-doped (2000 ppm) UF4 followed by scanning electron microscopy and energy dispersive spectroscopy. This characterization revealed morphological features of the metal product and slag, the latter of which displayed a wide range of features that indicates a complex reaction in which many variables are involved. Initial characterizations also identified thorium-rich particles, which were extracted and analyzed via transmission electron microscopy and atom probe tomography. These characterizations identified a new thorium-bearing phase, Al9-xFe7+xTh2Si