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Yongfeng Zhang

Yongfeng Zhang

Assistant Professor
Associate Chair for Undergraduate Studies
gscholar website
lab website

Dr Yongfeng Zhang research interest lies in materials aging and degradation in extreme conditions such as radiation, high temperature, stress and corrosive media using microstructure-based modeling. In such harsh environments, materials are subject to significant property degradations induced by a variety of mechanisms, which limit their performance. By tracking microstructure evolution and constructing structure-property correlations, the objective of Dr. Zhang’s research is to uncover the degradation mechanisms in materials under extreme conditions. The research outcome will help predict the degradation rates and guide the development of novel materials for applications in extreme conditions.

Dr. Zhang obtained his PhD degree in 2009 from Rensselaer Polytechnic Institute. After that he joined Idaho National Laboratory, INL. Prior to join University of Wisconsin, He is a senior Staff Scientist at INL and leads the Computational Microstructure Science Group in the Fuel Modeling and Simulation Department.

Department

Nuclear Engineering & Engineering Physics

Contact

917, Engineering Research Building
1500 Engineering Dr
Madison, WI
p: (608) 890-3779

  • PhD 2009, Rensselaer Polytechnic Institute
  • MS 2004, University of Science and Technology of China
  • BS 2001, University of Science and Technology of China

  • Nuclear fuels and materials
  • Defect thermodynamics and kinetics
  • Irradiation effect
  • Mechanical degradation
  • Cossion and stress corrosion cracking

Affiliated Departments

  • 2020 US Nuclear Regulatory Commission Nuclear Education Program, Faculty Development award
  • 2016 Idaho National Laboratory, Exceptional Contribution Program Award
  • 2016 Idaho National Laboratory, Expanded Recognition Award, Nuclear Science & Technology Directorate
  • 2016 Idaho National Laboratory, Laboratory Director’s Award for Leadership
  • 2014 Idaho National Laboratory, Exceptional Contribution Program Award
  • University of Science and Technology of China, Guanghua Scholarship
  • University of Science and Technology of China, Outstanding student scholarship (3 times)

  • Jin, M., Miao, J., Chen, B., Khafizov, M., Zhang, Y., & Hurley, D. H. (2024). Extended defects-enhanced oxygen diffusion in ThO2. Computational Materials Science, 235, 112842.
  • Manzoor, A., & Zhang, Y. (2024). Interplay between thermal vacancy and short-range order in complex concentrated alloys. Journal of Alloys and Compounds, 982, 173788.
  • Schneider, A., Andersson, D., & Zhang, Y. (2024). Mechanistic study of moisture corrosion of FeCr alloys in molten salts by ab-initio molecular dynamics simulations. Communications Materials, 5(1), 91.
  • Zhang, Y. (2023). A review of void and gas bubble superlattices self-organization under irradiation. Frontiers in Nuclear Engineering, 2, 1110549.
  • Beeler, B., Zhang, Y., Jahid Hasan, A., Park, G., Hu, S., & Mei, Z. (2023). Analyzing the effect of pressure on the properties of point defects in γU–Mo through atomistic simulations. MRS Advances, 8(1), 1--5.
  • David, Rapha"elle,, Rezwan, A. A., & Zhang, Y. (2023). Impact of defect annihilation mechanism on the grain size dependence and the dimensionality effect of radiation induced segregation. Materialia, 30, 101851.
  • Yu, Z., Kautz, E., Zhang, H., Schneider, A., Kim, T., Zhang, Y., Lambeets, S., Devaraj, A., & Couet, A. (2023). Irradiation damage reduces alloy corrosion rate via oxide space charge compensation effects. Acta Materialia, 253, 118956.
  • Griesbach, C., Gerczak, T., Zhang, Y., & Thevamaran, R. (2023). Microstructural heterogeneity of the buffer layer of TRISO nuclear fuel particles. Journal of Nuclear Materials, 574, 154219.
  • Di Lemma, F. G., Salvato, D., Capriotti, L., Williams, W. J., Teng, F., Zhang, Y., & Yao, T. (2023). Microstructure and phase evolution in the U-10Zr fuel investigated by in situ TEM heating experiments.. Journal of Nuclear Materials, 583, 154475.
  • Hoffman, A. K., Zhang, Y., Arivu, M., He, L., Sridharan, K., Wu, Y., Islamgaliev, R. K., Valiev, R. Z., & Wen, H. (2023). Novel effects of grain size and ion implantation on grain boundary segregation in ion irradiated austenitic steel. Acta Materialia, 246, 118714.

  • E M A 471 - Intermediate Problem Solving for Engineers (Spring 2025)
  • E P 471 - Intermediate Problem Solving for Engineers (Spring 2025)
  • M S & E 790 - Master's Research or Thesis (Spring 2025)
  • M S & E 990 - Research and Thesis (Spring 2025)
  • N E 990 - Research and Thesis (Spring 2025)
  • M S & E 423 - Nuclear Engineering Materials (Fall 2024)
  • M S & E 790 - Master's Research or Thesis (Fall 2024)
  • M S & E 990 - Research and Thesis (Fall 2024)
  • N E 423 - Nuclear Engineering Materials (Fall 2024)
  • N E 890 - Pre-Dissertator's Research (Fall 2024)
  • N E 990 - Research and Thesis (Fall 2024)
  • M S & E 790 - Master's Research or Thesis (Summer 2024)
  • N E 890 - Pre-Dissertator's Research (Summer 2024)
  • N E 990 - Research and Thesis (Summer 2024)
  • M S & E 790 - Master's Research or Thesis (Spring 2024)
  • N E 541 - Radiation Damage in Metals (Spring 2024)
  • N E 699 - Advanced Independent Study (Spring 2024)
  • N E 890 - Pre-Dissertator's Research (Spring 2024)
  • M S & E 423 - Nuclear Engineering Materials (Fall 2023)
  • M S & E 790 - Master's Research or Thesis (Fall 2023)
  • N E 423 - Nuclear Engineering Materials (Fall 2023)
  • N E 699 - Advanced Independent Study (Fall 2023)
  • N E 890 - Pre-Dissertator's Research (Fall 2023)
  • M S & E 790 - Master's Research or Thesis (Summer 2023)
  • N E 890 - Pre-Dissertator's Research (Summer 2023)