Peter Chou
- Name
- Dr. Peter Chou
- Institution
- Electric Power Research Institute
- Position
- Principal Technical Leader
- Affiliation
- ANS
- h-Index
- ORCID
- 0000-0001-6203-7974
- Biography
Dr. Peter Chou is a Principal Technical Leader in the Nuclear Sector of the Electric Power Research Institute (EPRI), an independent, nonprofit organization for public interest energy and environmental research that conducts research, development, and demonstration projects for the benefit of the public in the United States and internationally. He manages research on stress corrosion cracking and irradiation-assisted stress corrosion cracking in the primary environments of light-water reactors, focusing on the link between material microstructure and material reliability, and on the application of advanced characterization and micromechanical testing for industrial research.
- Expertise
- Atom Probe Tomography, Irradiated Stainless Steel, Irradiation Assisted Stress Corrosion Cracking, Micromechanical Testing, Stress Corrosion Cracking (SCC)
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"Correlative STEM-APT characterization of radiation-induced segregation and precipitation of in-service BWR 304 stainless steel"
Timothy Lach, Kayla Yano, Danny Edwards, Thak Sang Byun, Peter Chou,
Journal of Nuclear Materials
Vol. 549
2021
Link
Radiation induced segregation and precipitation phenomena in an in-service boiling water reactor 304 stainless steel component were investigated using directly correlated 3D-atom probe tomography and scanning transmission electron microscopy. Significant quantitative differences in measured segregation at grain boundaries were found between the atom probe and energy dispersive spectroscopy measurements of the exact same locations. In particular, a much stronger Si segregation (~10 atomic% via atom probe versus ~4 atomic% via electron microscopy) and different Cr profile shapes were detected that are critical to models of radiation induced segregation and stress corrosion cracking behavior. These quantitative differences highlight the need for comparative microscopy and critical evaluation of limitations in each analytical method. Elemental segregation to dislocations and conjoined-clusters were also highlighted by atom probe; confirming and expanding upon what has been observed in test reactor neutron and accelerator-based ion irradiations. |
| "Irradiation Assisted Stress Corrosion Cracking Susceptibility of Alloy X-750 Exposed to BWR Environments" Peter Chou, John Jackson, Sebastien Teysseyre, TMS-2017 conference [unknown] |
B and Li grain boundary segregation in irradiated LWR steels - FY 2016 RTE 2nd Call, #645
In-situ mechanical testing of neutron-irradiated 304SS exhibiting unusual deformation and fracture behavior with respect to temperature - FY 2021 RTE 1st Call, #4325
Mechanistic Insight Through TEM Characterization of Intergranular Irradiation-Assisted Stress Corrosion Crack Tips in As-Irradiated vs. Post-Irradiation Annealed Specimens - FY 2021 RTE 1st Call, #4346
Radiation-accelerated, strain-induced martensitic transformation and its potential impact on performance of 304 stainless steel irradiated to high doses in PWRs following plant life extension - FY 2020 RTE 1st Call, #2954
About Us
The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Through peer-reviewed proposal processes, the NSUF provides researchers access to neutron, ion, and gamma irradiations, post-irradiation examination and beamline capabilities at Idaho National Laboratory and a diverse mix of university, national laboratory and industry partner institutions.
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