Caleb Massey

Profile Information

Name: Dr. Caleb Massey
Institution: Oak Ridge National Laboratory
Position: R&D Staff Member
h-Index
ORCID: 0000-0003-1093-3958
Expertise: Atom Probe Tomography, FeCrAl, Ion Beam Irradiation, Mechanical Properties, Microscopy, Neutron Irradiation, Oxide Dispersion-Strengthened Alloy (ODS), Uranium Alloys, Zircaloy-4

Publications:

		"An Atom Probe Tomography Study of the Through Wall Attenuation Effect on Cu-rich Precipitate Formation in a Reactor Pressure Vessel Steel" Philip Edmondson, Caleb Massey, Mikhail Sokolov, Thomas Rosseel, Journal of Nuclear Materials Vol. 543 2021 Link
		"An Atom Probe Tomography Study of the Through Wall Attenuation Effect on Cu-rich Precipitate Formation in a Reactor Pressure Vessel Steel" Philip Edmondson, Caleb Massey, Thomas Rosseel, Journal of Nuclear Materials Vol. 545 2021 Link High-Cu weld material harvested from an ex service reactor pressure vessel (RPV) steel from Unit 1 of the decommissioned Zion Nuclear Generating Station has been characterized using atom probe tomography. Samples taken from 4 different positions through the thicknesses of the pressure vessel wall from the water-side to the air-side were characterized, along with an unirradiated baseline material. In the baseline material, no precipitates were found and the Cu was observed to be fully in solid solution; however, scanning transmission electron microscopy combined with energy dispersive spectroscopy (STEM-EDS) revealed the presence of -Cu that form during processing of the material and results in the concomitant decrease of matrix Cu. Following irradiation, a high number density of nano-scale Cu-rich precipitates (CRPs) were observed, uniformly distributed throughout the matrix. The Cu content within the CRPs was found to be 30-35 at.% regardless of location in the wall. No statistically significant variation in the compositions, mean radius, number density, or volume fraction as a function of location within the wall was observed. The measured matrix Cu level excluding CRPs contribution was found to be 90 appm higher than the solubility limit suggesting that further nucleation and growth of the CRPs under continued operations would have occurred. These results clearly demonstrate that the neutron energy attenuation has no significant effect on the precipitation kinetics of CRPs regardless of location in the wall in high-Cu RPV steels under irradiation.
		"Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding" David Hoelzer, Caleb Massey, Sebastien Dryepondt, Kinga Unocic, Bruce Pint, Journal of Nuclear Materials Vol. 501 2018 59-71 Link Low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys were developed as accident tolerant fuel cladding because of their excellent oxidation resistance at very high temperature, high strength and improved radiation tolerance. Fe-12Cr-5Al wt.% gas atomized powder was ball milled with Y2O3+FeO, Y2O3+ZrO2 or Y2O3+TiO2, and the resulting powders were extruded at 950?°C. The resulting fine grain structure, particularly for the Ti and Zr containing alloys, led to very high strength but limited ductility. Comparison with variants of commercial PM2000 (Fe-20Cr-5Al) highlighted the significant impact of the powder consolidation step on the alloy grain size and, therefore, on the alloy mechanical properties at T?<?500?°C. These low-Cr compositions exhibited good oxidation resistance at 1400?°C in air and steam for 4?h but could not form a protective alumina scale at 1450?°C, similar to observations for fine grained PM2000 alloys. The effect of alloy grain size, Zr and Ti additions, and impurities on the alloy mechanical and oxidation behaviors are discussed.
		"Influence of mechanical alloying and extrusion conditions on the microstructure and tensile properties of Low-Cr ODS FeCrAl alloys" Caleb Massey, Sebastien Dryepondt, Philip Edmondson, Kurt Terrani, Steven Zinkle, Journal of Nuclear Materials Vol. 512 2018 227-238 Link
		"Microchemical evolution of irradiated additive-manufactured HT9" Pengyuan Xiu, Caleb Massey, T.M. Kelsey Green, Stephen Taller, dieter Isheim, Niyanth Sridharan, JNM Vol. 559 2022 Link
		"Post irradiation examination of nanoprecipitate stability and α′ precipitation in an oxide dispersion strengthened Fe-12Cr-5Al alloy" Caleb Massey, Philip Edmondson, Kevin Field, David Hoelzer, Kurt Terrani, Steven Zinkle, Scripta Materialia Vol. 162 2018 94-98 Link
		"Post-irradiation examination of commercial tantalum alloys following neutron irradiation" Caleb Massey, Journal of Nuclear Materials Vol. 591 2024 154906 Link Post irradiation examination of two commercial tantalum alloys (based on Ta-8%W) revealed significant irradiation hardening following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile samples of commercial Ta alloys T-111 and ASTAR-811C were irradiated to a total neutron fluence (E > 0.1 MeV) in the range (2-4)x1021 n/cm2 (0.39-0.75 dpa) at an irradiation temperature near 900 °C. Tensile testing conducted at room temperature and at 800 °C revealed significant irradiation hardening under all test conditions. Samples that were tensile tested at room temperature exhibited brittle failure, whereas ductility was maintained for elevated temperature testing. The irradiation hardening and embrittlement noted for both alloys was attributed to irradiation-induced dislocation loop formation as confirmed through scanning transmission electron microscopy analysis. Corresponding energy dispersive x-ray spectroscopy revealed radiation-induced segregation of Hf in these loop populations. This Hf enrichment is shown to result in eventual co-precipitation of elongated (Hf,O)-rich precipitates along the dislocation loops. These results show that irradiation hardening is enhanced via irradiation-enhanced precipitation at elevated irradiation temperatures. Consequently, the minimum recommended operating temperature window for Ta alloys should be increased to account for observed hardening/embrittlement at irradiation temperatures as high as 900 °C (1,173 K).

NSUF Articles:

U.S. DOE Nuclear Science User Facilities Awards 30 Rapid Turnaround Experiment Research Proposals - Awards total nearly $1.2 million The U.S. Department of Energy (DOE) Nuclear Science User Facilities (NSUF) has selected 30 new Rapid Turnaround Experiment (RTE) projects, totaling up to approximately $1.2 million. These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Wednesday, April 26, 2017 - Calls and Awards
RTE 2nd Call Awards Announced - Projects total approximately $1.6 million These project awards went to principal investigators from 26 U.S. universities, eight national laboratories, two British universities, and one Canadian laboratory. Tuesday, May 14, 2019 - Calls and Awards
NSUF Research Helps Support Student Researchers - Leading and assisting on NSUF projects, students can get hands on experience The opportunities for student-led research are nearly endless Friday, January 24, 2020 - Newsletter
NSUF awards 28 Rapid Turnaround Experiment proposals - Approximately $1.74M has been awarded. The new call closes June 28. Thursday, June 1, 2023 - Calls and Awards

Accomplishments

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Authored an NSUF-supported publication

NSUF Submitter

Submitted an RTE Proposal to NSUF

Accomplished Researcher

Awarded 3+ RTE Proposals

Accomplished Collaborator

Collaborated on 3+ RTE Proposals

NSUF Reviewer

Reviewed an RTE Proposal

NSUF Supported Research

Nano-precipitate Response to Neutron Irradiation in Model ODS FeCrAl Alloy 125YF - FY 2017 RTE 2nd Call, #961

Nano-precipitate Stability and a'-Precipitation in ODS and Wrought FeCrAl Alloys - FY 2019 RTE 2nd Call, #1747

Quantifying the effect of simultaneous vs. sequential irradiation on creep performance of additively manufactured austenitic stainless steel - FY 2023 RTE 2nd Call, #4654

The Role of Helium on Microstructure Evolution in A709 - FY 2024 Super RTE Call, #5012

NSUF Research Collaborations

Co-Location of Solute Clusters and Dislocations in Additively Manufactured 316L Stainless Steels - FY 2024 RTE 1st Call, #4841

Evolution of Heterogeneous 316LSS Microstructures Under Neutron Irradiation - FY 2024 RTE 1st Call, #4862

Imaging of Irradiation Effects in Tantalum Alloys for Fast-Spectrum Self-Powered Neutron Detectors - FY 2021 RTE 1st Call, #4331

Investigating nitrogen effects on the mechanical properties and microstructure evolution in neutron irradiated HT-9 steel - FY 2023 RTE 2nd Call, #4651

Microstructural Origin of Irradiation Hardening and Embrittlement in Irradiated Second Generation FeCrAl Alloys - FY 2023 RTE 1st Call, #4554

Neutron Irradiation Effects on the Tensile Properties of Wire Arc Additive Manufactured Grade 91 Steel - FY 2022 RTE 1st Call, #4437

The Role of Dislocation Cell Walls on Cavity Nucleation in Additively Manufactured 316H Steel - FY 2024 RTE 1st Call, #4838

Towards Understanding Fuel Clad Chemical Interactions in FeCrAl Alloys - FY 2020 RTE 2nd Call, #4108