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Meher, Subhashish

Publications and source records attributed to Meher, Subhashish.

36 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Hierarchical microstructures: a potential route to enhanced stability in structural materials for advanced nuclear reactors

The drive to increase efficiency in nuclear energy systems is leading to the need for materials that operate at higher temperatures and stress levels for extended periods, while maintaining stable microstructures to ensure their performance is not compromised. A novel route to producing materials that can perform well in such environments is the creation of hierarchical microstructures. A hierarchical microstructure is a microstructure in which features are present at multiple length scales simultaneously. In this work, a hierarchical microstructure is fabricated in a nickel-base superalloy, featuring nanometer-size gamma precipitates inside larger gamma-prime particles, which are in turn embedded in the gamma matrix phase. The hierarchical features of the microstructure lead to enhanced stability of the gamma-prime precipitates during annealing; the particle size does not follow the expected t^(1/3) growth law predicted by the classic LSW theory. Phase-field simulations are used to understand the unexpected stability of the gamma-prime precipitates.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Development of a Laser Ultrasonics-based Approach for Rapid Screening of High Entropy Alloys

This project utilized a laser ultrasonic technique to systematically study temperature-induced evolution of material properties in a set of interrelated binary alloys and a high entropy alloy (HEA) fabricated using arc-melting and spark plasma sintering processes. This technique involved the use of a nanosecond duration, high-intensity pulsed laser to thermo-elastically generate ultrasonic waves that propagate in the bulk of the metal alloy. Sub-nanometer-scale displacements associated with the propagating bulk ultrasonic waves were detected along the epicenter on the opposite surface of the sample using a 1 GHz bandwidth photorefractive interferometer. Phase transformations and microstructural changes were inferred from the temperature-dependent trends in the bulk acoustic velocities and features in the ultrasonic epicentral waveforms measured in the binary alloys. These inferences were then correlated with electron/optical microscopy observations and predictions using calculations of phase diagrams (CALPHAD). The results showed that the laser-generated ultrasonic pulses were strongly influenced by changes in material microstructure and could accurately track thermally driven phase transformations and detect the presence of microscale heterogeneities (grain boundaries, dendritic structures, etc.) in the set of binary alloy samples. The measurement approach was then applied to a quinary HEA sample for estimating phase transition temperature and determining microstructural heterogeneity. The rapid, non-contact and non-destructive ultrasonic testing approach demonstrated here is amenable to high throughput combinatorial investigations that can be applied to graded composition HEAs produced using advanced manufacturing methods. When paired with atomistic simulations and CALPHAD modeling, this approach can overcome the bottlenecks faced by current material characterization methods in efficiently screening the vast discovery space of HEAs that spans over a hundred million unique quinary alloy compositions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A laser ultrasonics-based approach for rapid screening of high entropy alloys

This project demonstrates a laser ultrasonics-based characterization methodology for rapid metallic materials design and discovery via in situ determination of phases, microstructures and elastic properties with respect to temperature. Ultrasonic waves are strongly affected by material microstructure, and therefore serve as a facile means to probe elastic properties, phase content and their size distributions and volume fractions. In this study, a laser ultrasonic technique is used to systematically study the evolution of properties in a set of interrelated simple binary alloys and high entropy alloys (HEA). Phase transformations and microstructural changes inferred from the ultrasonic signals will be correlated with electron microscopy data and predictions using calculations of phase diagrams (CALPHAD). The non-contact and non-destructive ultrasonic testing approach developed in this study could overcome several limitations associated with current material characterization methods for materials discovery. It is expected that laser ultrasonics-based methodology developed in this work will be utilized to evaluate novel graded composition HEAs currently being developed at the Idaho National Laboratory (INL) using advanced manufacturing methods based on direct energy deposition and spark plasma sintering processes, and contribute to accelerate the discovery of HEAs.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A laser ultrasonics-based approach for rapid screening of high entropy alloys

The primary objective of this seed project is to develop a laser ultrasonics-based characterization methodology for rapid metallic materials design and discovery via in situ determination of phases, microstructures and elastic properties with respect to temperature. Ultrasonic waves are strongly affected by material microstructure, and therefore, serve as a facile means to probe elastic properties, phase content and their size distributions and volume fractions. In this study, a laser ultrasonic technique will be used to systematically study the evolution of properties in a set of interrelated simple binary alloys and high entropy alloys (HEA). Phase transformations and microstructural changes inferred from the ultrasonic signals will be correlated with electron microscopy data and predictions using calculations of phase diagrams (CALPHAD). The non-contact and non-destructive ultrasonic testing approach developed in this study could overcome several limitations associated with current material characterization methods for materials discovery. It is expected that the products of the proposed work will be utilized to evaluate novel graded composition HEAs currently being developed at the Idaho National Laboratory (INL) using advanced manufacturing methods based on direct energy deposition and spark plasma sintering processes, and contribute to accelerate the discovery of HEAs.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Microstructural and Micro-Chemical Evolutions in the Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles

AGR-1 and AGR-2 TRISO fuel particles were fabricated with slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. To correlate those differences with the fuel kernel responses to neutron irradiations in terms of irradiated fuel microstructure, fission products chemical and physical states, and fission gas bubble evolutions, extensive microstructural and analytical characterizations were conducted. The studies used a state of art transmission electron microscopy (TEM) equipped with Energy-dispersive X-ray spectroscopy (EDS) of four silicon solid-state detectors which have super sensitivity and fast 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 to represent a representative irradiation conditions with a fuel burnup within the range from 10.8 to 18.6% FIMA, and the time-average volume-average temperatures vary from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% for the AGR1 fuel kernels and 14.03 wt.% for the AGR-2 fuel kernels. The TEM results show that there were significant microstructural reconstructions in the irradiated fuel kernels for both the AGR-1 and AGR-2 fuels. There are four major phases including fuel matrix of UO2 and UC, U2RuC2, and UMoC2 in the irradiated AGR2 fuel kernel. Zr and Nb form solid solution in the UC phase. UMoC2 phase often shows a detectable concentration of Tc. Pd was found to mainly locate in the buffer layer or to be associated with fission gas bubble within the UMoC2 phase. The EDS maps qualitatively show that the rare-earth fission products (Nb, et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there is a high number of rod-shape precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The difference of microstructural and micro-chemical evolutions in irradiated fuel kernels between the AGR-1 and AGR-2 TRISO fuel particle may result from a combined factor of irradiation temperature, fuel geometry and chemical composition. However, the irradiation temperature probably play a more deterministic role. Limited electron energy loss spectroscopy (EELS) characterizations on the AGR2 fuel kernel show that there is nearly no carbon in the UO2 phase while a small fraction of oxygen was detected in the UC/UMoC2 phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Micro- and Nano-techniques for the Study of Fission Product Precipitation in SiC Layer

The complexity of fission product distribution and composition within the SiC layer of tristructural isotropic (TRISO) coated particles, irradiation effects on the SiC structure as well as the variable nature of metallic fission product release, has led to the exploration of multiple micro- and nano-characterization techniques. The fine scale of the fission product precipitates necessitates unique learnings and application of electron microscopic techniques with irradiated fuel during the Advanced Gas Reactor (AGR)-1 experiment analysis. A summarized discussion on the specific advanced techniques with associated method development for TRISO coated particles is provided followed by the down selected techniques currently considered to provide the highest impact. Techniques considered are electron probe micro-analyzer (EPMA), scanning transmission electron microscopy (STEM), and Precession electron diffraction (PED). SiC grain boundary characteristics are evaluated due to the mobility of specific fission products through grain boundaries. PED was performed for understanding the role of grain boundary character on fission product transport. One set of comparative analyses between AGR-1 and AGR-2 particles leads to a finding that the AGR-1, high-Ag-retention particle, had statistically more coincidence site lattice (CSL)-related grain boundaries (but fewer low-angle grain boundaries), compared to the AGR-2 particle. This implies that CSL-related grain boundaries may have a direct influence on Ag retention. However, precipitates were found on a relatively small fraction of CSL-related grain boundaries in both particles. In another set of comparison between AGR-1 and AGR-2 particles that were subjected to safety testing, there is significant differences in average SiC grain boundary distributions, but the fission product distributions on these grain boundaries are very similar.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Electron Microscopic Examination of an Irradiated TRISO Coated Particle of AGR-2 Experiment: AGR2-222-RS019

The transmission electron microscopic (TEM) examination of an irradiated tristructural isotropic (TRISO) coated particle from advanced gas reactor (AGR) 2 fuel Compact 2-2-2 is presented in this report. Compact 2-2-2 refers to the compact in Capsule 2 at Level 2 of Stack 2. The examination on AGR2 222 RS019 is focused on the interface between the silicon carbide (SiC) to inner pyrolytic carbon (IPyC) layer, central part of SiC, and the interface between SiC and outer pyrolytic carbon layer. The SiC grain boundary distribution and associated fission product precipitates were characterized by scanning transmission electron microscopy (STEM) energy dispersive spectroscopy (EDS), and precession electron diffraction (PED) in the transmission electron microscope. Two separate locations on SiC layer were selected for on the basis of the intact of the buffer layer in the irradiated TRISO particle. The key observation in this particle is that the compositional nature of grain boundary precipitates is more complex near IPyC/SiC layer where Pd, Pd-Ag, Pd-U containing precipitates can be observed. However, the compositions of precipitate towards outer most region of SiC mainly contain Pd. Ag has been observed in the IPyC/SiC layer in one of the two locations. Comparison of the results to previously irradiated and analyzed, Variant 3, safety tested AGR-1 TRISO particles (AGR1-433-001 and AGR1 222-RS004) showed the grain boundary distributions have some noticeable differences. The differences in coincident site lattice grain boundary and misorientation angles are possibly inherent to the fabrication process. Neutron irradiation induced void distribution has been found to be nonuniform. A high concentration of smaller voids is found at stacking faults compared to those in the SiC matrix. Larger void sizes are observed in the SiC location that is near to the broken buffer layer. This void size variation with integrity of buffer layer can potentially affect the fission product retention. However, the average void size continuously decreases along the radius of SiC in both locations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Advanced Electron Microscope and Micro Analysis of TRISO coated Particles: FY2020 Overview

Objectives Understanding Effects of Irradiation on TRISO layers Fission product chemistry and behavior in UCO kernel Identify and Understand Fission Product Transport Mechanisms in TRISO Coated Particles Outcomes and Impact Impact on Performance Improve Predictive Behavior Modeling Kernel Behavior: Release from kernel; release from whole particle Known Fission Product Transport Mechanisms

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

An Overview of Neutron Irradiated SiC Layer Behavior using Micro-and Nano-Characterization Techniques

The complexity of fission product distribution and composition within the SiC layer of tristructural isotropic (TRISO) coated particles, irradiation effects on the SiC structure as well as the variable nature of metallic fission product release, has led to the exploration of multiple micro- and nano-characterization techniques. The fine scale of the fission product precipitates necessitates unique learnings and application of electron microscopic techniques with irradiated fuel during the Advanced Gas Reactor (AGR)-1 experiment analysis. A summarized discussion on the specific advanced techniques with associated method development for TRISO coated particles is provided followed by the down selected techniques currently considered to provide the highest impact. Techniques considered are electron probe micro-analyzer (EPMA), scanning transmission electron microscopy (STEM), and Precession electron diffraction (PED). SiC grain boundary characteristics are evaluated due to the mobility of specific fission products through grain boundaries. PED was performed for understanding the role of grain boundary character on fission product transport. One set of comparative analyses between AGR-1 and AGR-2 particles leads to a finding that the AGR-1, high-Ag-retention particle, had statistically more coincidence site lattice (CSL)-related grain boundaries (but fewer low-angle grain boundaries), compared to the AGR-2 particle. This implies that CSL-related grain boundaries may have a direct influence on Ag retention. However, precipitates were found on a relatively small fraction of CSL-related grain boundaries in both particles. In another set of comparison between AGR-1 and AGR-2 particles that were subjected to safety testing, there is significant differences in average SiC grain boundary distributions, but the fission product distributions on these grain boundaries are very similar.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Grain Boundary Characteristics of SiC in Irradiated, AGR-2 TRISO Particles

Microstructural characterization of the SiC layer was carried out on neutron irradiated, tristructural isotropic (TRISO) particles from the Advanced Gas Reactor (AGR)-2 experiment. The SiC grain boundary distribution in each particle was characterized with precession electron diffraction in the transmission electron microscope. Generally, the distributions in the AGR-2 TRISO particles (fabricated at the pilot scale by an industrial vendor) were similar to the Variant 3, AGR-1 TRISO particles made on a laboratory-scale at Oak Ridge National Laboratory. Slightly more random, high-angle grain boundaries in conjunction with slightly less low-angle grain boundaries were found in the AGR-2 TRISO particles compared to AGR-1 TRISO particles. Plotting the Ag-110m retention exhibited by both Variant 3 AGR-1 and AGR-2 irradiated TRISO particles against various grain boundary types revealed Ag retention correlate with the random, high-angle and twin grain boundary fraction in the center area of the SiC layer. Observations associated with localized areas of precipitation also will be reported in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Understanding of a Novel Irradiation-induced Nanostructuring Process in SiC layer of TRISO fuel particles via Transmission Electron Microscopy

Metallic fission products are found to transport through the SiC layer of the tristructural isotropic (TRISO) nuclear fuel particle and various studies have been undertaken to understand the transport behavior [1, 2]. While the grain boundaries in SiC acts as a primary pathway to dissipate fission products in the form of multi-phased complex compounds, intragranular fission product precipitation has always been observed. Research on neutron irradiation damage tolerance of silicon carbide (SiC) and its composites, used in TRISO fuel, light water reactors (LWR) fuel cladding etc., has traditionally ignored the irradiation-induced “polymorphism” of SiC. In the post-irradiation advanced microscopy effort at Idaho National Laboratory, it has been recently discovered that the intragranular precipitation of fission products in SiC layer of TRISO coated fuels takes place via a dual step nucleation mechanism that involves cubic (?) ?hexagonal (?) SiC polymorphic transition, and subsequent metamorphosis of ?-SiC into fission product precipitates, as shown in Figure 1 [3]. This precipitation behavior cannot be explained by any reported nucleation mechanism. The extent of this precipitation has been observed to be influenced by the nature of the neutron irradiation induced damage structures. These damage structures (e.g., defect morphology and density) are also indicative of the irradiation temperature and neutron fluence levels experimented by the individual TRISO particles. The distinction of neutron irradiation damage structures as a function of the fuel kernel type, burn up level and fabrication histories of TRISO coated particles from the Advanced Gas reactor (AGR-1 and AGR-2) experiments has been examined and compared. This discovery not only adds new fundamental understanding of physical behavior of polymorphic ceramics under irradiation, but also provides the following beneficial traits for high temperature applications. 1) Since the extent of precipitation of the desired phase is directly dependent on the amount of precursor phase, the volume fraction of desired phase can be easily controlled by irradiation parameters. This eliminates the conventional limitation of volume fraction of a desired phase in the system as outlined by the phase diagram of the specific system. 2) Most of the precipitate-strengthened materials for high temperature applications suffer from degradation of size and shape of precipitate upon prolonged thermal exposures. But, precipitates are confined to the shape and size of the morphological template of surrogate phase (e.g. a Pd silicide precipitate is confined by the ?-SiC as shown in Figure 1(c)) in the present nucleation pathway. Hence, there is no possibility of desired phase to undergo any morphological degradation at high temperature that eventually will extend the lifetime of these materials at high temperature.

42 ENGINEERING↗
DOE OSTI2018

A Novel Dual-Step Nucleation Pathway in Crystalline Solids under Neutron Irradiation

Innovations in nanostructuring of inorganic crystalline solids are often limited by prerequisite critical nucleation energy and solute supersaturation for formation of a phase. This research provides direct evidence supporting the viability of an unconventional irradiation-induced nanostructuring process, via transmission electron microscopy, that circumvents these preconditions. Using polymorphic silicon carbide (SiC) as a prototype, a surprising two-step nucleation route is demonstrated through which nanoscale distribution of the second phase is achieved by reaction of solutes with neutron irradiation-induced precursors. In the first step, nanoscale α–SiC precipitates in a β–SiC matrix unexpectedly nucleate heterogeneously at structural defects. This occurs at significantly lower temperatures compared with the usual β→α transition temperature. Subsequently, α–SiC precipitate acts as a surrogate template for its structural and compositional transition into a fission product precipitate, palladium silicide. These discoveries provide a modern view of irradiation engineering in polymorphic ceramics for advanced applications.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Effect of High Si Content on U 3 Si 2 Fuel Microstructure

The development of U 3 Si 2 as an accident tolerant nuclear fuel has gained research interest due to its promising high uranium density and improved thermal properties. In the present study, three samples of U 3 Si 2 fuel with varying silicon content have been fabricated by a conventional powder metallurgical route. Microstructural characterization via scanning and transmission electron microscopy reveals the presence of other stoichiometry of uranium silicide such as USi and UO 2 in both samples. The detailed phase analysis by x-ray diffraction shows the presence of secondary phases, such as USi, U 3 Si, and UO 2 . The samples with higher concentrations of silicon content of 7.5 wt.% displays additional elemental Si. Furthermore, these samples also possess an increased amount of the USi phase as compared to that in the conventional sample with 7.3 wt.% silicon. The optimization of U 3 Si 2 fuel performance through the understanding of the role of Si content on its microstructure has been discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Three-dimensional nanoscale characterisation of materials by atom probe tomography

The development of three-dimensional (3D), characterization techniques with high spatial and mass resolution is crucial for understanding and developing advanced materials for many engineering applications as well as for understanding natural materials. In recent decades, atom probe tomography (APT) which combines a point projection microscope and time-of-flight mass spectrometer has evolved to be an excellent characterization technique capable of providing 3D nanoscale characterization of materials with sub-nanometer scale spatial resolution, with equal sensitivity for all elements. This review discusses the current state as of beginning of the year 2016 of APT instrumentation, new developments in sample preparation methods, experimental procedures for different material classes, reconstruction of APT results, the current status of correlative microscopy, and application of APT for microstructural characterization in established scientific areas like structural materials as well as new applications in semiconducting nanowires, semiconductor devices, battery materials, catalyst materials, geological materials and biological materials. Finally, a brief perspective is given regarding the future of APT.

Atom Probe Tomography, characterization, microstru↗
DOE OSTI2015

Determination of solute site occupancies within γ' precipitates in nickel-base superalloys via orientation-specific atom probe tomography

In this study, the analytical limitations in atom probe tomography such as resolving a desired set of atomic planes, for solving complex materials science problems, have been overcome by employing a well-developed unique and reproducible crystallographic technique, involving synergetic coupling of orientation microscopy with atom probe tomography. The crystallographic information in atom probe reconstructions has been utilized to determine the solute site occupancies in Ni-Al-Cr based superalloys accurately. The structural information in atom probe reveals that both Al and Cr occupy the same sub-lattice within the L1 2 -ordered g precipitates to form Ni 3 (Al,Cr) precipitates in a Ni-14Al-7Cr(at.%) alloy. Interestingly, the addition of Co, which is a solid solution strengthener, to a Ni-14Al-7Cr alloy results in the partial reversal of Al site occupancy within g precipitates to form (Ni,Al) 3 (Al,Cr,Co) precipitates. This unique evidence of reversal of Al site occupancy, resulting from the introduction of other solutes within the ordered structures, gives insights into the relative energetics of different sub-lattice sites when occupied by different solutes.

36 MATERIALS SCIENCE↗