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Wharry, J. P.

Publications and source records attributed to Wharry, J. P..

6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Method for Evaluating Irradiation Effects on Flow Stress in Fe-9%Cr ODS Using TEM In Situ Cantilevers

Transmission electron microscopic (TEM) in situ mechanical testing has become a widely utilized tool for simultaneously measuring mechanical properties and understanding fundamental deformation mechanisms in irradiated and nuclear materials. Although tensile and compression specimen geometries are amongst the most common, opportunities remain for investigating alternative geometries that could provide unique insights into the plasticity of irradiated materials. This research demonstrates a new TEM in situ cantilever beam configuration. Cantilevers are produced from as-received and proton irradiated (1 dpa, 500°C) Fe-9%Cr oxide dispersion strengthened (ODS) steel. Flow stress is measured using a TEM in situ depth-sensing mechanical testing holder. A 200 MPa increase in flow stress is measured due to irradiation. Size effects arise when the intrinsic (i.e. microstructural) size approaches the extrinsic (i.e. external dimensions) size and can be described using a power law relationship as a function of the material microstructure and cantilever dimensions.

36 MATERIALS SCIENCE↗
DOE OSTI2016

In Situ TEM Microcompression Pillar Size Effects in Fe-9Cr ODS

The objective of this paper is to ascertain the existence of a specimen size threshold, below which the material no longer exhibits bulk mechanical behavior when tested using in situ transmission electron microscopic (TEM) micro-compression pillars. In this work, sub-micron sized pillars of a model Fe-9Cr oxide dispersion strengthened (ODS) alloy were compressed in TEM in situ at room temperature, and attempts have been made to quantify the resulting data. Results are compared to mechanical property values in the archival literature, and a discussion on the validity of quantitative in situ TEM micro-compression data for ODS steel is included. In situ TEM mechanical testing is a relatively new technique enabling concurrent TEM imaging and mechanical testing of sub-micron-sized electron-transparent specimens. Due to limited sample volume of many nuclear materials, in situ TEM mechanical testing holds great promise for advancing our understanding of deformation phenomena. But it is well known that the dramatically reduced specimen size required for in situ testing can affect quantitative and qualitative deformation behaviors. However, Kiener, et al. used in situ TEM micro-compression pillars on irradiated and unirradiated single-crystal copper. They observed size independence in the irradiated material, while the unnirradiated material displayed a size threshold below which the mechanical properties deviated from those of bulk Cu. They assert that the more densely-spaced obstacles in irradiated Cu allow it to display bulk properties even in specimens having ∼ 100 nm dimensions. Similar to Kiener's findings, we hypothesize that unirradiated ODS will demonstrate a size threshold, but the in situ TEM micro-compression technique can be extended from single crystal materials to engineering alloys. An Fe-9Cr model ODS alloy was chosen for this study because of the broad interest in ODS for advanced nuclear reactors. ODS alloys exhibit high irradiation tolerance and high temperature creep resistance. New insight into deformation processes gained through in situ TEM mechanical testing could be valuable for the validation of ODS alloys for next generation nuclear reactors. The fact that the compressive yield stress and elastic modulus do not follow distinct trends as a function of specimen size is likely due to the high density of obstacles to dislocation motion. The ODS microstructure contains such obstacles as oxide nanoparticles, a fine grain structure, a high density of dislocation lines, and carbide precipitates. Since the deformation behavior is governed by the minimum separation distance between obstacles [1], it is plausible that the high obstacle density enables ODS alloy micropillars to exhibit bulk-like mechanical properties even prior to irradiation. Future work will focus on: (1) developing finite element models to help explain the difference in elastic moduli between bulk and micropillar tests, (2) understanding the spread in yield stress measurements at fixed micropillar volumes, and (3) extending this study to irradiated ODS. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Comparison of Ion and Neutron Irradiations to 3 dpa at 500 deg. C in Ferritic-Martensitic Alloys

The growing global demand for energy will increasingly call upon advanced nuclear fission reactors to supply safe and reliable electricity. The structural and fuel cladding components of these reactors will be subject to extreme conditions of irradiation damage up to several hundred displacements per atom (dpa) at temperatures as high as 700 deg. C. Ferritic-martensitic (F-M) steels are leading candidates for these challenging conditions due to their strength and dimensional stability under irradiation. In order to accelerate the process for evaluating F-M alloys, charged particles are increasingly being used to emulate neutron irradiations. Charged particle irradiations allow the possibility of conducting irradiation experiments within a shorter time period (i.e. at a rate up to 4 orders of magnitude faster) and with minimal radioactivation of the material, enabling lower cost and faster turnaround of post irradiation examination and analysis. However, the irradiation dose rate, damage cascade morphologies, and irradiation damage depth profiles all differ widely between protons, heavier ions, and neutrons. Currently, there is limited understanding of the significance of these physical differences and how they manifest in the irradiated microstructure and mechanical properties of F-M steels. The objective of this study is to evaluate charged particles as a surrogate for neutron irradiations in F-M alloys by assessing common irradiation conditions using Fe{sup ++} ions, protons, and neutrons. Keeping the temperature and dose consistent enables isolation of the effects of each irradiating particle and their respective dose rates and cascade morphologies. Specimens of HCM12A and HT9 were irradiated separately with protons, self-ions, and neutrons to a common dose of 3 dpa at 500 deg. C. The microstructure of each specimen was characterized with TEM and LEAP and compared to evaluate proton and self-ion irradiation as a surrogate for neutron irradiation at these conditions. - Proton, self-ion, and neutron irradiations produced consistent sizes and number densities of grains, carbides, dislocation lines, and dislocation loops. Voids were absent in all conditions. The charged particle irradiations have successfully reproduced the neutron irradiation results for these microstructural features without a temperature shift. - Nano-cluster evolution of Si-Mn-Ni-P and Cu-rich clusters was markedly different for each irradiating particle type. Some cluster morphology trends could be explained by differences in cascade volume for differing irradiating species. But dose rate likely also influences cluster evolution. - Cr-rich nanoclusters nucleated only in the neutron-irradiated HCM12A and HT9. This is likely due to thermal aging at 500 deg. C over a longer experimental time, as compared to the proton and self-ion irradiations. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of Irradiation on Strengthening of a Model Fe-9Cr Oxide Dispersion Strengthened Alloy

The growing energy demand will increasingly call upon advanced nuclear reactors to supply safe and reliable energy worldwide. However, structural and cladding materials in these reactors will be subject to extreme conditions of up to a few hundred displacements per atom (dpa) at temperatures as high as 600 deg. C. Oxide dispersion strengthened (ODS) alloys are leading candidates for these components because their dispersion of Y-Ti-O nanoclusters provides high strength at elevated temperatures and dimensional stability under irradiation. But recent work has suggested that the oxides are unstable under irradiation, which has serious implications on mechanical performance. Thus, the objective of this work is to understand the change in strengthening mechanisms of an Fe-9Cr ODS alloy under irradiation. We use two approaches to assess strengthening of a proton-irradiated and neutron-irradiated Fe-9Cr ODS. First, nanoindentation directly measures changes in yield strength. Second, microstructure measurements are passed through models of two strengthening mechanisms. Results of these two approaches are compared to infer the strengthening mechanisms in the irradiated ODS alloy. A model Fe-9Cr ODS steel was irradiated to 3 dpa at 500 deg. C using either 2.0 MeV protons or a fast neutron spectrum. Both irradiation types induced the formation of dislocation loops and voids. Both proton and neutron irradiation resulted in dissolution of oxide nanoclusters, though neutron irradiation led to much more extensive oxide nano-cluster dissolution. The irradiated microstructure was used to predict irradiation hardening by considering both the precipitation strengthening and solid solution strengthening mechanisms. Due to more extensive oxide nano-cluster dissolution under neutron irradiation, the solid solution strengthening mechanism must be considered in order for the predicted change in yield strength to be comparable to that measured by nanoindentation. (authors)

36 MATERIALS SCIENCE↗