Longxiang Zhu

Biography:

Dr. Longxiang Zhu is now an assistant research scientist at Department of Nuclear Engineering, Chongqing University. He works in nuclear thermal-hydraulic, including fluid mechanics and heat transfer, two-phase flow phase distribution and interfacial transfer behavior, modeling and development of nuclear thermal-hydraulic codes, and Machine Learning in two-phase flow. His work appears onInternational Journal of Heat and Mass Transfer,Progress in Nuclear Energy, among other peer-reviewed journals.

Education Background:

Xi’an Jiaotong University, Ph.D., College of Nuclear, 2016/09-2021/06

University of Illinois at Urbana-Champaign, Visiting scholar, NPRE, 2018/09-2020/11

Chongqing University, B.S., Department of Nuclear Engineering, 2012/09-2016/06

Research Interests:

[1] Fluid mechanics and heat transfer

[2] Two-phase flow phase distribution and interfacial transfer behavior

[3] Modeling and development of nuclear thermal-hydraulic codes

[4] Machine Learning in two-phase flow

Selected Publications：

[1]Longxiang Zhu*, Taiyang Zhang, Joseph L. and Bottini, Caleb S. Brooks.“Two-dimensional quantitative study of boiling flow evolution in vertical inner-heated annulus channel.”International Journal of Heat and Mass Transfer, 183 (2022): 122190.（SCI,EI）

[2] Zhiee Jhia Ooi*,Longxiang Zhu, Joseph L. Bottini, and Caleb S. Brooks. “Identification of flow regimes in boiling flows in a vertical annulus channel with machine learning techniques.”  International Journal of Heat and Mass Transfer, 185 (2022): 122439.（SCI,EI）

[3]Longxiang Zhu, Zhiee Jhia Ooi, Joseph L. Bottini, Caleb S. Brooks*, and Jianqiang Shan. “Bubble diameter distribution and intergroup mass transfer in the multigroup two-fluid model.” International Journal of Heat and Mass Transfer, 163 (2020): 120456.（SCI,EI）

[4]Longxiang Zhu, Zhiee Jhia Ooi, Caleb S. Brooks*, and Jianqiang Shan. “Modeling sensitivity in prediction of interfacial area concentration in boiling flow.” Progress in Nuclear Energy, Vol. 133, 103638. DOI: 10.1016/j.pnucene.2021.103638.（SCI,EI）

[5] Joseph L. Bottini,Longxiang Zhu, Zhiee Jhia Ooi, Taiyang Zhang, and Caleb S. Brooks*. "Experimental study of boiling flow in a vertical heated annulus with local two-phase measurements and visualization." International Journal of Heat and Mass Transfer 155 (2020): 119712.（SCI,EI）

[6]Longxiang Zhu*, Zhiee Jhia Ooi, Taiyang Zhang, Caleb S. Brooks, and Liangming Pan. "Flow regime identification for boiling flow and study of feed-in information in unsupervised machine learning." In 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH -19. 2022.

[7]Longxiang Zhu*, Taiyang Zhang, Zhiee Jhia Ooi, Caleb S. Brooks, and Liangming Pan. "Validation of interfacial mass transfer closures in drift flux model for condensation flow." In 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH -19. 2022.

[8]Longxiang Zhu, Joseph L. Bottini, and Caleb S. Brooks*. “Sensitivity of intergroup mass transfer in two-group IATE for subcooled boiling flow.” In 2020 ANS Winter Meeting and Nuclear Technology Expo, Chicago, 2020.

[9]Longxiang Zhu, Zhiee Jhia Ooi, and Caleb S. Brooks*. "Current Intergroup mass transfer limitations in the multi-group two-fluid model." In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH -2019. 2019.（EI）

[10]Longxiang Zhu, Zhiee Jhia Ooi, and Caleb S. Brooks*. "Current capability of interfacial area transport approaches in subcooled boiling." In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-2019. 2019.（EI）

[11]Longxiang Zhuand Jianqiang Shan*. "Interfacial Drag Force Improvement in Two-Fluid Model." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers Digital Collection, 2018.（EI）

[12] Bottini, Joseph L.,Longxiang Zhu, Zhiee Jhia Ooi, Taiyang Zhang, and Caleb S. Brooks*. "A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel." In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH- 2019. 2019.（EI）