Understanding and removing FIB artifacts in metallic TEM samples using flash electropolishing
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36 MATERIALS SCIENCE↗
Engineering topics
Prabhakaran, Ramprashad
Publications and source records attributed to Prabhakaran, Ramprashad.
Transmission electron microscopy of ion irradiated ODS MA956 samples
Oxide dispersion strengthened (ODS) alloys are promising candidate materials for the next generation advanced nuclear reactors due to their superior irradiation resistance and mechanical properties. To better understand the effect of irradiation on MA956, it is essential to study higher dose (50-100 dpa) samples, so that the general trend of microstructural evolution and the resulting radiation-hardening can be deduced. Currently, ion irradiations are considered the only way to achieve doses beyond ~50 dpa in a practical time frame relevant to alloy and welding development programs. This dataset contains TEM characterization results of ODS MA956 samples that were ion irradiated under different conditions: Sample #5 (1.25 dpa at 190℃); Sample #9 (50 dpa at 190℃); Sample #15 (1.25dpa under 320℃); and Sample 17 (25 dpa at 320℃). TEM characterization focused on the irradiation induced defects (dislocation lines and loops) using the on-zone axis bright field STEM technique. These data were collected using a FEI Tecnai G2 F30 S/TEM at Microscopy and Characterization Suite (MaCS) at Center for Advanced Energy Studies (CAES), Idaho Falls, ID. This project (ion irradiation and TEM studies) was supported by the U.S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-05ID14517 as part of Nuclear Science User Facilities award #18-14784 (PI: Ramprashad Prabhakaran, PNNL).
Prabhakaran, Ramprashad↗
Preliminary Characterization and Evaluation on ShAPE Manufactured 316H and ODS Steels
This study provides the first- of- a- kind results of direct tube formation through shear assisted processing and extrusion (ShAPE) for oxide dispersion strengthened (ODS) steel material; previously only bar was successfully made. The Advanced Materials and Manufacturing Technology (AMMT) program develops cross-cutting technologies in support of a broad range of nuclear reactor technologies and maintains U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of AMMT is to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy. Solid-state advanced manufacturing techniques can overcome some of the challenges in liquid-based additive manufacturing processes and should therefore be considered in material design and manufacturing as well. The work presented in this report forms part of a study on solid-state additive manufacturing techniques of 316 stainless steels and ODS steel components and supports the vision and goals of the AMMT program relevant to accelerate the development and deployment of advanced manufacturing processes. Achieving this can provide a safety improvement through larger safety margins, economic benefit for higher efficiency during operation, and a cost reduction through more effective manufacturing processes and less waste.
36 MATERIALS SCIENCE↗
Manufacturing ODS Steels from GARS Powders by Friction Consolidation and Extrusion
Previous research suggested that friction-based processing is a promising method for fabricating oxide dispersion-strengthened (ODS) steel. In this study, we combined friction consolidation and friction extrusion to successfully manufacture ODS steel rods using precursor powder made with gas atomization reaction synthesis. We examined the microstructure evolution from the initial powder to the final extruded rod, which revealed the dispersion process of Y. Additionally, by comparing the microstructures of three rods extruded at different temperatures, we showed that low-temperature friction extrusion effectively enhanced microstructure uniformity and prevented grain coarsening, leading to improved mechanical properties. Furthermore, our findings provide practical guidelines for adjusting processing parameters in the production of ODS steel using friction-based processing.
36 MATERIALS SCIENCE↗
The effect of neutron irradiation on conventional and nanocrystalline nickel: NSUF Work Package UA-22PN080501 Milestone M3UA-22PN0805012 (Final Report)
Even though nanocrystalline materials (20-100 nm) present an unprecedented potential, scientific knowledge related to the effect of neutron irradiation on the mechanical properties and microstructure is still scarce. Most of the past studies were conducted using ion irradiation which may not have the same effect as neutron irradiation because of the smaller irradiation volume and the higher dose rate. To reach a firm conclusion on the potential of nanocrystalline materials for nuclear reactor applications, extensive study of model metals with different stacking fault energy (SFE) is required to elucidate their behavior in radiation environments. Nanocrystalline copper and nickel are typically chosen because they are commonly used as model FCC metals in studies of radiation effects. Nickel is an FCC metal with a high stacking-fault energy (~125 mJ/m 2 ) compared to copper (~45 mJ/m 2 ). Hence, microcrystalline and nanocrystalline nickel samples were irradiated in the INL’s Advanced Test Reactor (ATR), as a part of FY08 North Carolina State University NSUF Irradiation Experiment# 96 to evaluate the irradiation behavior of these materials. The objective of this FY20 NSUF project# 19122 is to perform PIE (at PNNL NSUF facility) on previously ATR-neutron irradiated (1.2 and 2.6 dpa; 80-89°C) nanocrystalline and microcrystalline nickel samples to investigate the changes in mechanical properties and microstructures and evaluate whether nanocrystalline nickel is relatively more radiation resistant compared to conventional microcrystalline nickel. To perform PIE at PNNL, sixteen neutron irradiated specimens (microcrystalline and nanocrystalline) were transferred from the NSUF Nuclear Fuels and Materials Library at INL. Experimental techniques such as SEM/EBSD, XRD, TEM, Vickers microhardness and tensile testing were employed to characterize the effect of neutron irradiation on the microstructure and mechanical properties of nanocrystalline nickel and compared them with corresponding characteristics of microcrystalline nickel.
36 MATERIALS SCIENCE↗
Irradiation-induced shift in the thermodynamic stability of phases and the self-healing effect in transformative high entropy alloys
Here, this work investigated Fe 40 Mn 20 Cr 15 Co 20 Si 5 high entropy alloy (CS-HEA), which exhibits transformation induced plasticity (TRIP) from γ-fcc → ε-hcp, as a probable candidate for nuclear applications. CS-HEA is an extensively-explored, low stacking fault energy alloy with superior strength, ductility, fatigue resistance, and corrosion resistance. This study delved into the effect of irradiation on the shift in thermodynamic stability of the phases and thus radiation tolerance. The evolution of phases, lattice parameters, and transformation volume, V γ→ε , were evaluated from X-ray diffraction experiments along with the mechanical response from nanoindentation. The alloy exhibited a recently-proposed novel self-healing mechanism possible due to the TRIP effect to minimize irradiation damage by restraining the γ-fcc → ε-hcp transformation via thermal aid; this self-healing mechanism was confirmed by transmission electron microscopy. The results were corroborated by a negative change in V γ→ε and a low |V γ→ε |, which is an important criterion for recovery of parent γ-fcc phase. Thus, this alloy was deemed a good radiation-tolerant candidate for nuclear application.
36 MATERIALS SCIENCE↗
Manufacturing Oxide Dispersion Strengthened (ODS) steel plate via cold spray and friction stir processing
Oxide dispersion strengthened (ODS) steels, traditionally fabricated by ball milling and conventional powder metallurgy techniques to achieve bulk form, followed by intricate rolling and thermal treatment steps to achieve plate or sheet form. Here, we present a novel processing route that combines cold spray (CS) with friction stir processing (FSP) to manufacture ODS steel plate directly from gas atomization reaction synthesis (GARS)-prepared powder, thus no rolling steps involved. Microstructural and mechanical characterizations were performed to assess the quality and properties of the resulting ODS steel plate. Our findings demonstrate that the slightly porous CS deposited layer was fully consolidated after FSP, yielding a fully dense ODS steel plate that exhibited a favorable tradeoff between strength and ductility upon extraction from the substrate. Furthermore, through microstructural analysis, we revealed the presence of an appreciable density (∼10 22 /m 3 ) of nano-sized oxide particles, with the majority being smaller than 5 nm via the combined CS + FSP fabrication route. This work serves as a first proof-of-concept demonstration of the manufacturing approach described herein, offering a possible alternative route for producing ODS steel plates.
36 MATERIALS SCIENCE↗
Feasibility Evaluation of a Solid Phase Scalable HEA Cladding Manufacturing Route
First of a kind development result on two low-energy solid-phase processes applied on an irradiation-resistant alloy, NiCoFeCrCu 0.12 , are achieved and demonstrate moderate feasibility of successful tube fabrication using shear assisted processing and extrusion (ShAPE™) and friction stir layer deposition as a bulk manufacturing process. The scope of the work is performed in four phases: 1) direct tube manufacturing of the irradiation-resistant high-entropy alloy (HEA) composite with increased strength, 2) co-shear lining manufacturing process for the increased strength and corrosion-resistant, irradiation-tolerant HEAs, 3) ShAPE of the radially gradient corrosion resistance alloy, and 4) alloy development and fabrication enabled through friction stir additive manufacturing processes among others. This report describes the development activities from April to December 2023 to manufacture a direct customizable thin-walled tubular product from irradiation-tolerant composite high-entropy alloys (C-HEAs) while the overall project is continuing in 2024.
36 MATERIALS SCIENCE↗
Influence of processing on secondary phase formation and microstructural evolution at U-10Mo alloy and Zr interlayer interfaces
Thermo-mechanical processing of uranium-10 wt. % molybdenum (U-10Mo) fuel plates leads to microstructure changes at the U-10Mo/Zr interfaces. Secondary phases formed at this interface are particularly important to interfacial bond strength, process optimization, and maintaining structural integrity of the U-10Mo fuel plates during irradiation. In this work, we determined the phases and phase transformation products occurring at the interface of the U-10Mo fuel and Zr interlayer when the fuel plate is subjected to short and long hot isostatic pressure times. Interfacial morphology, structure and composition of phases formed, and relative hardness across the U-10Mo/Zr interfaces were studied using a multi-length scale, multi-modal characterization approach involving electron microscopy, atom probe tomography, and atomic force microscopy. Here, results highlight that the extent of phase transformations, secondary phase formation, and hardness variability across interfaces can be controlled by modifying processing parameters. Phase diagram construction and thermodynamic calculations were performed using the Thermocalc software to identify expected phases formed at interfaces during the maximum hold temperature of 560 °C experienced during HIP.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermomechanical Properties of Neutron Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials
A thermal neutron absorber material composed of Al 3 Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al 3 Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4T m . Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al 3 Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al 3 Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al 3 Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of –6.3% for the annealed irradiated 36.5 vol% specimen.
36 MATERIALS SCIENCE↗
Microstructural Characterization of Ion Irradiated ODS MA956 Alloy
Oxide Dispersion Strengthened (ODS) alloys are considered as promising nuclear materials for the next generation advanced nuclear reactors due to their improved elevated temperature mechanical properties and radiation damage resistance. MA956 is a commercial ODS alloy with superior oxidation resistance and creep strength at elevated temperatures. The improved properties for ODS alloys can be attributed to the presence of dispersed nanosized oxide particles. These oxide particles can act as barriers to dislocation motion and provide additional sinks and sites for point defect recombination. Hence, their stability can greatly affect the properties of the irradiated ODS alloys. Besides, irradiation will induce defects into the ODS alloys, and these defects can develop further as irradiation dose increases. Furthermore, the defects propagation can significantly undermine the mechanical properties of the bulk materials, resulting in irradiation hardening and lower ductility. Therefore, it is necessary to have a good understanding of the defect’s evolution with irradiation. Hence, efforts are ongoing to understand the effect of ion irradiation on ODS MA956 with emphasis on oxide particles and evolution of defects as a function of dose (2.5, 50 and 100 dpa).
07 ISOTOPE AND RADIATION SOURCES↗