headshot of Jun Liao

Jun Liao

Adjunct Assistant Professor
Mechanical Engineering & Materials Science


Dr. Liao is an expert in thermal hydraulics of nuclear reactors and in the development of modeling and simulation computer codes. His responsibility includes development of safety analysis computer codes and safety analysis methodologies for both light water and non light water reactors. He has developed and licensed WCOBRA/TRAC-TF2 systems code and the FULL SPECTRUM LOCA (FSLOCA) methodology for PWR LOCA analysis.He played important roles in the development of the Westinghouse AP1000 plant and of the Westinghouse SMR, he is currently developing eVinci micro reactor and the Westinghouse lead fast reactor (LFR). Dr. Liao is a member of the American Nuclear Society (ANS) and American Society of Mechanical Engineers (ASME) and serves in two committees of ANS and ASME. He is currently an adjunct professor in University of Pittsburgh and an associate editor of Journal of Nuclear Engineering and Radiation Science (ASME).

Research interests: Nuclear reactor thermal hydraulics; Advanced reactor design; Nuclear safety and accident analysis; Nuclear systems computer code; microchannel heat exchanger; multiphase flow and heat transfer in energy systems; apply machine learning to improve safety of reactor.


Ph.D, Aerospace Engineering, University of Florida, 2000 - 2005

Master of Science, Xi'an Jiaotong University, 1997

Bachelor of Engineering, Huazhong University of Science and Technology, 1994

Liao, J., Ferroni, P., Wright, R.F., Bachrach, U., Scobel, J.H., Sofu, T., Tentner, A.M., Lee, S.J., Epstein, M., Frignani, M., & Tarantino, M. (2021). Development of phenomena identification and ranking table for Westinghouse lead fast reactor's safety. PROGRESS IN NUCLEAR ENERGY, 131, 103577.Elsevier BV. doi: 10.1016/j.pnucene.2020.103577.

Hua, T.Q., Lee, S.J., Liao, J., Moisseytsev, A., Ferroni, P., Karahan, A., Paik, C.Y., Tentner, A.M., & Sofu, T. (2020). Development of Mechanistic Source Term Analysis Tool SAS4A-FATE for Lead- and Sodium-Cooled Fast Reactors. Nuclear Technology, 206(2), 206-217.Informa UK Limited. doi: 10.1080/00295450.2019.1598715.

Liao, J., & Utley, D. (2020). Study on Reactor Vessel Air Cooling for Westinghouse Lead Fast Reactor. Nuclear Technology, 206(2), 191-205.Informa UK Limited. doi: 10.1080/00295450.2019.1599614.

Liao, J., Ohkawa, K., Brown, W.L., & Wright, R.F. (2019). A Study on Onset of Liquid Entrainment in Low Pressure Depressurization System of Advanced Passive Pressurized Water Reactors. Journal of Nuclear Engineering and Radiation Science, 5(4).ASME International. doi: 10.1115/1.4042796.

Liao, J. (2016). System scaling analysis for modeling small break LOCA using the FULL SPECTRUM LOCA evaluation model. ANNALS OF NUCLEAR ENERGY, 87, 443-453.Elsevier BV. doi: 10.1016/j.anucene.2015.09.014.

Liao, J., Kucukboyaci, V.N., & Wright, R.F. (2016). Development of a LOCA safety analysis evaluation model for the Westinghouse Small Modular Reactor. Annals of Nuclear Energy, 98, 61-73.Elsevier BV. doi: 10.1016/j.anucene.2016.07.023.

Liao, J., Frepolil, C., & Ohkawa, K. (2015). Cold leg condensation model for analyzing loss-of-coolant accident in PWR. NUCLEAR ENGINEERING AND DESIGN, 285, 171-187.Elsevier BV. doi: 10.1016/j.nucengdes.2015.01.011.

Liao, J.Liao, J. (2015). A Horizontal Stratified Gas-Liquid Two-Phase Flow Model for the Two-Fluid Model in the WCOBRA/TRAC-TF2 PWR Safety Analysis Code. Nuclear Engineering and Design, 295, 239-250.

Liao, J., Mei, R., & Klausner, J.F. (2008). A study on the numerical stability of the two-fluid model near ill-posedness. International Journal of Multiphase Flow, 34(11), 1067-1087.Elsevier BV. doi: 10.1016/j.ijmultiphaseflow.2008.02.010.

Liao, J., Mei, R., & Klausner, J.F. (2006). A film boiling model for cryogenic chilldown at low mass flux inside a horizontal pipeline. Heat and Mass Transfer, 42(10), 891-900.Springer Science and Business Media LLC. doi: 10.1007/s00231-006-0143-5.

Liao, J., Mei, R., & Klausner, J.F. (2004). The influence of the bulk liquid thermal boundary layer on saturated nucleate boiling. International Journal of Heat and Fluid Flow, 25(2), 196-208.Elsevier BV. doi: 10.1016/j.ijheatfluidflow.2003.11.012.

Liao, J. (2019). Importance of PIRT to the Safety of Heat Pipe Based Micro Nuclear Reactor. In ANS Winter Meeting.ANS.Washington DC.

Liao, J. (2019). Data-Driven Safety Margin Management using Reduced Order Modeling. In ANS Winter Meeting.ANS.Washington DC.

Liao, J. (2019). The Westinghouse Lead Fast Reactor Program. In Top Fuel/GLOBAL.Seattle, WA.

Liao, J. (2019). The Importance of Phenomena Identification and Ranking Table in Lead Fast Reactor Development. In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.Portland, USA.

Liao, J. (2019). Chromium-Coated Cladding Effects in the Context of 10 CFR 50.46c. In Top Fuel/GLOBAL.Seattle, USA.