Correlation Between the Microstructure and Mechanical Properties of Irradiated Fe-9Cr ODS

The growing global demand for energy will increasingly call upon fusion reactors and Generation IV nuclear fission reactors to supply safe and reliable energy worldwide. Ferritic/martensitic (F/M) alloys are leading candidates for structural components in these reactors because of their high strength, dimensional stability, and low activation. In novel reactor concepts, these materials will be subject to extreme operating conditions, accumulating doses of irradiation up to a few hundred displacements per atom (dpa) at temperatures as high as 600°C. Oxide dispersion strengthened (ODS) F/M alloys containing a dispersion of Y-Ti-0 nanoclusters have been developed to operate at even higher temperatures

Irradiation-induced NanoCluster Evolution

Oxide dispersion strengthened steel (ODS) and commercial ferritic-martensitic (F-M) alloys are widely accepted candidate structural materials for designing advanced nuclear reactors. Nanoclusters embedded in the steel matrix are key microstructural features of both alloy types. Irradiation from nuclear fusion and fission affects the morphology of these nanoparticles, altering the performance of the alloys and potentially decreasing their usable lifetime. Thus, it is important to understand the effect of irradiation on these nanoparticles in order to predict long-term nuclear reactor performance. It was found that the evolution of nanoclusters in each material is different depending on the experimental irradiation parameters. The Nelson-Hudson-Mazey (NHM) model has been refined based on previous experimental work, and has been shown to be an effective model to simulate irradiation-induced nanocluster evolution in ODS and F-M steels. In this work, an NHM simulation tool was developed for nanoHUB, with a simplified user interface that enables rapid prediction of the effect of irradiation on the size of nanoclusters in a variety of Fe-based steels

Enhanced radiation damage tolerance of amorphous interphase and grain boundary complexions in Cu-Ta

Amorphous interfacial complexions are particularly resistant to radiation damage and have been primarily studied in alloys with good glass-forming ability, yet recent reports suggest that these features can form even in immiscible alloys such as Cu-Ta under irradiation. In this study, the mechanisms of damage production and annihilation due to primary knock-on atom collisions are investigated for amorphous interphase and grain boundaries in a Cu-Ta alloy using atomistic simulations. Amorphous complexions, in particular amorphous interphase complexions that separate Cu and Ta grains, result in less residual defect damage than their ordered counterparts. Stemming from the nanophase chemical separation in this alloy, the amorphous complexions exhibit a highly heterogeneous distribution of atomic excess volume, as compared to a good glass former like Cu-Zr. Complexion thickness, a tunable structural descriptor, plays a vital role in damage resistance. Thicker interfacial films are more damage-tolerant because they alter the defect production rate due to differences in intrinsic displacement threshold energies during the collision cascade. Overall, the findings of this work highlight the importance of interfacial engineering in enhancing the properties of materials operating in radiation-prone environments and the promise of amorphous complexions as particularly radiation damage-tolerant microstructural features

Benzoate dioxygenase from<em> Ralstonia eutropha</em> B9 – unusual regiochemistry of dihydroxylation permits rapid access to novel chirons

Oxidation of benzoic acid by a microorganism expressing benzoate dioxygenase leads to the formation of an unusualipso,orthoarenecis-diol in sufficient quantities to be useful for synthesis.</p

Continuous-flow transfer hydrogenation of benzonitrile using formate as a safe and sustainable source of hydrogen †

The continuous catalytic transfer hydrogenation of benzonitrile to benzylamine is demonstrated using a palladium on carbon catalyst with triethylammonium formate as reducing agent. Solvent choice was critical in overcoming rapid catalyst deactivation. A 15-fold increase in catalyst productivity was observed in flow compared to batch, which was achieved using an ethanol–water solvent in combination with intermittent catalyst regeneration by washing with water

The Comparison of Microstructure and Nanocluster Evolution in Proton and Neutron Irradiated Fe–9%Cr ODS Steel to 3 DPA at 500 °C

A model Fe-9%Cr oxide dispersion strengthened (ODS) steel was irradiated with protons or neutrons to a dose of 3 displacements per atom (dpa) at a temperature of 500 °C, enabling a direct comparison of ion to neutron irradiation effects at otherwise fixed irradiation conditions. The irradiated microstructures were characterized using transmission electron microscopy and atom probe tomography including cluster analysis. Both proton and neutron irradiations produced a comparable void and dislocation loop microstructure. However, the irradiation response of the Tie-Ye-O oxide nanoclusters varied. Oxides remained stable under proton irradiation, but exhibited dissolution and an increase in Y:Ti composition ratio under neutron irradiation. Both proton and neutron irradiation also induced varying extents of Si, Ni, and Mn clustering at existing oxide nanoclusters. Protons are able to reproduce the void and loop microstructure of neutron irradiation carried out to the same dose and temperature. However, since nanocluster evolution is controlled by both diffusion and ballistic impacts, protons are rendered unable to reproduce the nanocluster evolution of neutron irradiation at the same dose and temperature

PM-HIP Mechanical Data Archive

This is the mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). Data in this archive include uniaxial tensile testing data, fractography of selected tensile bars, and nanoindentation. The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300ºC and 400ºC through the DOE Office of Nuclear Energy, Nuclear Science User Facilities (NSUF) project 15-8242.THIS DATASET IS ARCHIVED AT DANS/EASY, BUT NOT ACCESSIBLE HERE. TO VIEW A LIST OF FILES AND ACCESS THE FILES IN THIS DATASET CLICK ON THE DOI-LINK ABOV

Nanoindentation Investigation of Chloride-Induced Stress Corrosion Crack Propagation in an Austenitic Stainless Steel Weld

Transgranular chloride-induced stress corrosion cracking (TGCISCC) is a mounting concern for the safety and longevity of arc welds on austenitic stainless steel (AuSS) nuclear waste storage canisters. Recent studies have shown the key role of crystallography in the susceptibility and propagation of TGCISCC in SS weldments. Given that crystallography underlies mechanical heterogeneities, the mechanical-crystallographic relationship during TGCISCC growth must be understood. In this study, welded SS 304L coupons are loaded in four-point bend fixtures and then boiled in magnesium chloride to initiate TGCISCC. Nanoindentation mapping is paired with scanning electron microscopy (SEM) electron backscatter diffraction (EBSD) to understand the correlation between grain orientation, grain boundaries, and hardening from TGCISCC propagation. The nanoindentation hardness of individual grains is found to not be a controlling factor for TGCISCC propagation. However, intragranular hardness is generally highest immediately around the crack due to localized strain hardening at the crack tip. This work shows that nanoindentation techniques can be useful in understanding CISCC behaviors when paired with electron microscopy

Microstructure Impacts on Mechanical Properties in a High Temperature Austenitic Stainless Steel

Austenitic and super-austenitic stainless steels are a critical component of the spectrum of high temperature materials. With respect to power generation, alloys such as Super 304H and NF709 span a gap of capability between ferritic and martensitic high chromium steels and nickel-based alloys in boiler tube applications for both conventionally fired boilers and heat-recovery steam generators (HRSG). This research explores a wrought version of a cast austenitic stainless steel, CF8C-Plus or HG10MNN, which offers promise in creep strength at relatively low cost. Various manufacturing techniques have been employed to explore the impact of wrought processing on nano-scale microstructure and ultimately performance, especially in high temperature creep. Transmission electron microscopy has been used to quantify and characterize the creep-strengthening particles examining the relationship between traditional melting and extrusion as compared to powder metallurgy