教育及工作经历

（一）2020.1-至今 西安交通大学 能动学院核科学与技术系 副教授

（二）2018.1-2019.12 西安交通大学 能动学院核科学与技术系 讲师

（三）2012.6-2017.12 西安交通大学 核科学与技术系，硕博连读

（四）2008-1-2012.6，西安交通大学，核工程与核技术系，本科

研究领域（方向）

（一）核反应堆热工水力基础理论与试验

（二）先进核动力系统热工水力与安全

科研项目

（一）热流局部集中下弥散型板燃料元件沸腾临界及熔化行为机理研究，国家自然科学基金，负责人

（二）事故容错燃料碳化硅包壳表面临界热流密度特性机理研究，国家自然科学基金，负责人

（三）基于分离效应试验数据的评估基准题开发联合研究，生态环境部核与辐射安全中心，负责人

（四）海洋条件对CHF的影响机理研究，中广核研究院有限公司，负责人

（五）运动条件下反应堆设备流固热耦合分析程序开发与验证试验，上海核工院研究设计院股份有限公司，负责人

学术及科研成果、专利、论文

[1]K. Zhang,Y.D. Hou,W.X. Tian,Y.Q. Fan,G.H. Su,S.Z. Qiu. Experimental investigations on single-phase convection and steam-water two-phase flow boiling in a vertical rod bundle[J]. Experimental Thermal and Fluid Science,2017,80.

[2]W.X. Tian,K. Zhang,Y.D. Hou,Y.P. Zhang,S.Z. Qiu,G.H. Su. Hydrodynamics of two-phase flow in a rod bundle under cross-flow condition[J]. Annals of Nuclear Energy,2016,91.

Kailun Guo, Chenglong Wang, Dalin Zhang, et al. INVESTIGATIONS OF NEAR-WALL BUBBLE BEHAVIOR IN WIRE HEATERS POOL BOILING[J]. Thermal Science, 2021, 25(5):3957 - 3967.

Wenxi Tian,  Qiang Lian,  Suizheng Qiu, et al. Research of two‐phase density wave instability in reactor core channels with rolling motion[J]. International Journal of Energy Research, 2020, 44(9):7323-7341.

Yu Liang,  Dalin Zhang,  Yutong Chen, et al. An experiment study of pressure drop and flow distribution in subchannels of a 37-pin wire-wrapped rod bundle[J]. Applied Thermal Engineering, 2020, 174

Chenglong Wang,  Hao Qin,  Wenxi Tian, et al. Transient analysis of tritium transport characteristics of thorium molten salt reactor with solid fuel[J]. Annals of Nuclear Energy, 2020, 141:107337-107337.

Di Liu,  Wenxi Tian,  Mengmeng Xi, et al. Study on safety boundary of flow instability and CHF for parallel channels in motion[J]. Nuclear Engineering and Design, 2018, 335:219-230.

Jian Ge,  Wenxi Tian,  Suizheng Qiu, et al. CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger[J]. Nuclear Engineering and Design, 2018, 337:27-37.

Juanli Zuo,  Wenxi Tian,  Ronghua Chen, et al. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump[J]. Nuclear Engineering and Design, 2013, 263:1-9.

Juanli Zuo,  Wenxi Tian,  Ronghua Chen, et al. Two-dimensional numerical simulation of single bubble rising behavior in liquid metal using moving particle semi-implicit method[J]. Progress in Nuclear Energy, 2013, 64:31-40.

Wenxi  Tian, Yuki  Ishiwatari, Satoshi  Ikejiri, et al. Numerical Simulation on Direct Contact Condensation of Single Bubble in Subcooled Water using MPS method[J]. AIP Conference Proceedings, 2010, 1207(1):933.

Kailun Guo, Chenglong Wang, Dalin Zhang, et al. INVESTIGATIONS OF NEAR-WALL BUBBLE BEHAVIOR IN WIRE HEATERS POOL BOILING[J]. Thermal Science, 2021, 25(5):3957 - 3967.

Wenxi Tian,  Qiang Lian,  Suizheng Qiu, et al. Research of two‐phase density wave instability in reactor core channels with rolling motion[J]. International Journal of Energy Research, 2020, 44(9):7323-7341.

Yu Liang,  Dalin Zhang,  Yutong Chen, et al. An experiment study of pressure drop and flow distribution in subchannels of a 37-pin wire-wrapped rod bundle[J]. Applied Thermal Engineering, 2020, 174

Chenglong Wang,  Hao Qin,  Wenxi Tian, et al. Transient analysis of tritium transport characteristics of thorium molten salt reactor with solid fuel[J]. Annals of Nuclear Energy, 2020, 141:107337-107337.

Di Liu,  Wenxi Tian,  Mengmeng Xi, et al. Study on safety boundary of flow instability and CHF for parallel channels in motion[J]. Nuclear Engineering and Design, 2018, 335:219-230.

Jian Ge,  Wenxi Tian,  Suizheng Qiu, et al. CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger[J]. Nuclear Engineering and Design, 2018, 337:27-37.

Juanli Zuo,  Wenxi Tian,  Ronghua Chen, et al. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump[J]. Nuclear Engineering and Design, 2013, 263:1-9.

Juanli Zuo,  Wenxi Tian,  Ronghua Chen, et al. Two-dimensional numerical simulation of single bubble rising behavior in liquid metal using moving particle semi-implicit method[J]. Progress in Nuclear Energy, 2013, 64:31-40.

Wenxi  Tian, Yuki  Ishiwatari, Satoshi  Ikejiri, et al. Numerical Simulation on Direct Contact Condensation of Single Bubble in Subcooled Water using MPS method[J]. AIP Conference Proceedings, 2010, 1207(1):933.

[3] K. Zhang,Y.Q. Fan,W.X. Tian,K.L. Guo,S.Z. Qiu,G.H. Su. Pressure drop characteristics of two-phase flow in a vertical rod bundle with support plates[J]. Nuclear Engineering and Design,2016,300.

[4]K. Zhang, Y.D. Hou, W.X. Tian, et al. Experimental investigation on steam-water two-phase flow boiling heat transfer in a staggered horizontal rod bundle under cross-flow condition[J]. Experimental Thermal and Fluid Science, 2018, 96:192-204.

[5]K. Zhang, Y.D. Hou, W.X. Tian, et al. Flow pattern effect on two-phase pressure drops in vertical upward flow across a horizontal tube bundle[J]. Annals of Nuclear Energy, 2018, 120:253-264.

[6]Y.P. Zhang, K. Zhang, J. Wang, et al. Research on pressure drop characteristics in inverted half U-tube bundle under two-phase cross-flow condition[J]. Annals of Nuclear Energy, 2018, 120:265-271.

[7]K. Zhang, Y.D. Hou, W.X. Tian, et al. Experimental investigations on single-phase convection and two-phase flow boiling heat transfer in an inclined rod bundle[J]. Applied Thermal Engineering, 2019, 148:340-351.

[8]K. Zhang, Y.D. Hou, W.X. Tian, et al. Theoretical prediction of single bubble motion in vertically upward two-phase flow across inclined tube bundles[J]. Annals of Nuclear Energy, 2019, 128:422-432.

[9]K. Zhang, H.B. You, Y.K. Zhou, et al. Experimental investigations on leak flow rate characteristics of water through axial artificial microcracks of steam generator tubes under back pressure conditions[J]. Nuclear Engineering and Design, 2019, 353:110285-110285.

[10]K. Zhang, Y.J. Huang, H.B. You, et al. Experimental investigations on single-phase convection and critical heat flux in vertical tubes under oscillatory flow condition[J]. Annals of Nuclear Energy, 2020, 143:107433-107433.