Yachun Wang
- Name
- Dr. Yachun Wang
- Institution
- Idaho National Laboratory
- Position
- Nuclear Engineer
- Affiliation
- Nuclear Science &Technology (NS&T)
- h-Index
- 10
- ORCID
- 0000-0002-4952-3633
- Biography
- Yachun (Ya) Wang is a Nuclear Engineer in the Irradiated Fuels and Materials Department (C610) at Idaho National Laboratory (INL). She earned her Ph.D. in Nuclear Science and Engineering from Rensselaer Polytechnic Institute (RPI) in 2020 and subsequently joined INL as a Research Postdoctoral Associate. Yachun’s research spans pre- and post-irradiation examination of nuclear fuels and materials, with expertise in corrosion mechanisms of ceramic materials, molten-salt-exposed NiCr alloys, U-10Zr metallic fuels, and cladding materi als. She employs a broad suite of characterization techniques, including X-ray diffraction, SEM, FIB/PFIB, TEM, and XPS. Additionally, she has accumulated hands-on experience in in-situ microscale mechanical testing of structural materials.
- Expertise
- ceramic, Electron Microscopy, Fuel Cladding Chemical Interaction (FCCI), Irradiated Cladding, Tensile Properties
| "Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding" Xiang Liu, Luca Capriotti, Tiankai Yao, Jason Harp, Michael Benson, Yachun Wang, Fei Teng, Lingfeng He, https://www.sciencedirect.com/science/article/pii/S002231152031196X#ack0001 Vol. 544 2021 Link | ||
| "Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system" Yachun Wang, Xiang Liu, Daniel Murray, Fei Teng, Wen Jiang, Mukesh Bachhav, Laura Hawkins, Emmanuel Perez, Cheng Sun, Xianming Bai, Jie Lian, Colin Judge, John Jackson, Robert Carter, Lingfeng He, Materials Science & Engineering Vol. 849 2022 Link |
| 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 |
| Department of Energy Nuclear Science User Facilities Awards 29 Rapid Turnaround Experiment Proposals - Awarded projects total nearly $1.14M in access awards Tuesday, June 8, 2021 - News Release, Calls and Awards |
In-situ TEM Heating Investigation of M23C6 Stability in Neutron Irradiated HT9 - FY 2024 Super RTE Call, #24-5078
In-situ high temperature micro-tensile testing of reactor irradiated HT9 cladding - FY 2023 RTE 3rd Call, #23-4673
A first Investigation in Lanthanide-induced Grain Boundary Embrittlement in HT9 Cladding via In-situ Micro-tensile Testing - FY 2021 RTE 1st Call, #21-4345
TEM Characterization of Neutron Irradiated Nd2Zr2O7 and Its Thermal Recovery Behavior - FY 2019 RTE 2nd Call, #19-1814
Evaluation of the Mechanical Properties of Phases in an Annular U-10Zr Irradiated Fuel - FY 2026 RTE 1st Call, #26-5661
Machine Learning-Based Analysis of Post-Irradiation Examination Data for Fuel-Cladding Chemical Interaction Characterization in Metallic Fuels - FY 2025 Super RTE Call, #25-5555
In-Situ Synchrotron X-ray Diffraction Study During Tensile Deformation of Neutron Irradiated MA957 ODS Alloy - FY 2025 RTE 2nd Call, #25-5409
Microstructural characterization of U-10wt.%Zr specimens from MFF-2 experiment for FAST experiment benchmarking - FY 2025 RTE 2nd Call, #25-5391
Probing the Effects of Temperature, Radiation, and Fuel Cladding Chemical Interaction on Irradiated HT-9 Cladding via In-Situ Micro-Tensile Testing - FY 2025 RTE 1st Call, #25-5285
High-Temperature In-Situ Micro-Tensile Testing of FFTF-Irradiated HT9 Cladding from U-10Zr Fuel Pins - FY 2025 RTE 1st Call, #25-5295
Nano indentation hardness of U10Zr - FY 2024 RTE 3rd Call, #24-5181
In-Situ TEM Nanomechanical Testing of Neutron Irradiated U-10Zr - FY 2024 RTE 3rd Call, #24-5177
In-situ TEM study of microstructural evolution and deformation in FFTF irradiated HT-9 cladding - FY 2024 RTE 3rd Call, #24-5142
Nanoindentation Creep Testing and Characterization of High Temperature Irradiated HT-9 Cladding - FY 2024 RTE 2nd Call, #24-4973
Determining Mechanical Properties of the Phases Formed of Irradiated U-19Pu-10Zr - FY 2023 RTE 3rd Call, #23-4751
In-situ Micro-tensile Testing for Measuring Grain Boundary Strength of NiCr Alloys under Simultaneous Irradiation and Corrosion Environments - FY 2020 RTE 1st Call, #20-3028