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姓 名 杨军 性 别
职 称 教授(博导) 毕业学校
个人主页 http://nuclear.energy.hust.edu.cn/
联系方式
邮 箱 yang_jun@hust.edu.cn
通讯地址 湖北省武汉市华中科技大学动力楼 430074
个人资料简介
杨军,华中科技大学教授、博士生导师,核工程与核技术系主任,能源与动力工程学院院长助理。 2015年入选国家第十二批千人计划青年人才,湖北省第六批百人计划。 先后从事先进反应堆热工水力与安全分析,热工系统程序验证开发,先进沸水堆设计,小型堆临界热通量分析,安全壳过滤排放系统设计等方面的研究。在权威期刊和国际会议发表论文近30篇。另出版专著一部,图书章节多部,英文科技报告40余部。
Researcher ID:  J-3541-2014

Google Scholar: http://scholar.google.ch/citations?user=kvqPeyoAAAAJ

2009-   美国核学会会员
2013-   中国核学会会员,湖北省核学会理事

 

担任以下核心期刊评审:

Nuclear Engineering and Design
Progress in Nuclear Energy
Annals of Nuclear Energy
Nuclear Science and Engineering

教育及工作经历

    2000              清华大学工程物理系学士
    2003              中国原子能科学研究院硕士
    2010              美国普渡大学(Purdue University)核工程系博士
    2011-2013     美国威斯康星大学(University of Wisconsin)博士后
    2013-2015     瑞士保罗谢尔研究院(Paul Scherrer Institut)研究人员
    2015-             华中科技大学能源与动力工程学院教授

研究方向

    主要研究方向:

    反应堆热工水力和安全分析
    汽液两相流流动及传热
    先进反应堆非能动安全系统设计
    堆芯临界热通量实验分析
    安全壳过滤降压系统设计

科研项目

    1,  先进反应堆非能动安全系统失冷事故在大型集成化热工水力试验台架上的实验模拟
    2,  第三代非能动反应堆到集成化实验台架的比例缩放设计建造
    3,  非能动安全壳冷却系统(PCCS)验证实验
    4,  热工系统程序对反应堆和试验台架的事故模拟
    5,  马克-I型反应堆193号通用安全问题实验研究
    6,  TIRGA型研究堆以及小型堆在自然循环条件下的临界热通量实验分析
    7,  反应堆安全壳过滤降压系统设计分析以及程序模拟

代表性论文与专利

    专著及章节

    (1)  J. Yang, “Evaluation of Passively Safe Boiling Water Reactor Design with Integral Tests and Codes”, ISBN:1249069718; 978-1249069713, published by ProQuest, July 2012 [Link]
    (2)  M. Anderson, J. Yang et al., “Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)”, Chapter 7: Critical Flow during depressurization from supercritical conditions, IAEA-TECDOC-1746, pp. 201-266, ISBN:978–9201076144, ISSN:1011–4289, published by International Atomic Energy Agency(IAEA), August 2014 [Link]


     期刊文章 (注*为通讯作者)
    (1)  J. Yang*, M. S. Greenwood, M. Angelis, M. Avery, M. Anderson, M. Corradini, J. Matos, F. Dunn, E. Feldman, “Study of Critical Heat Flux in Natural Convection Cooled TRIGA Reactors with Single Annulus and Rod Bundle Geometries”, Nuclear Science and Engineering, 180 (2015), pp. 1-13 [Link]
    (2)  J. Yang*, J. Lim, S. W. Choi, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii, “Behaviors of Passive Safety System under a Feed Water Line Break LOCA on a Generation III+ Boiling Water Reactor”, Progress in Nuclear Energy 83 (2015)pp. 35-42 [Link]
    (3)  J. Lim, J. Yang, S. W. Choi, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii*, “Assessment of Passive Safety System Performance under Gravity Driven Cooling System Drain Line Break Accident”,Progress in Nuclear Energy, 74 (2014), pp. 136-14 [Link]
    (4)  J. Lim, S. W. Choi, J. Yang, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii*, “Assessment of Passive Safety System Performance under Main Steam Line Break Accident”, Annals of Nuclear Energy, 64(2014) pp. 287-294 [Link] [Link]
    (5)  S. Rassame*, M. Griffiths, J. Yang, D. Y. Lee, P. Ju, S.W. Choi, T. Hibiki, and M. Ishii, “Experimental investigation of void distribution in Suppression Pool during the initial blowdown period of a Loss of Coolant Accident using air–water two-phase mixture,” Annals of Nuclear Energy, 73 (2014) pp. 53-67  [Link]
    (6)  S. Rassame*, M. Griffiths, J. Yang, D. Y. Lee, P. Ju, S.W. Choi, T. Hibiki, and M. Ishii, “Experimental Investigation of Void Distribution in Suppression Pool Over the Duration of a Loss of Coolant Accident Using Steam-Water Two-Phase Mixture,” Annals of Nuclear Energy, 75, (2014) pp. 570-580 [Link]
    (7)  J. Yang*, S. W. Choi, J. Lim, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii, “Counterpart experimental study of ISP-42 PANDA tests on PUMA facility”, Nuclear Engineering and Design, 258 (2013) pp. 249-257 [Link]
    (8)  J. Yang*, S. W. Choi, J. Lim, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii, “Assessment of performance of BWR passive safety systems in a small break LOCA with integral testing and code simulation”,Nuclear Engineering and Design, 247 (2012) pp. 128-135 [Link]
    (9)  M. Avery, J. Yang*, M. Anderson, M. Corradini, J. Matos, F. Dunn, E. Feldman, “Critical Heat Flux in TRIGA-Fueled Reactors Cooled By Natural Convection”, Nuclear Science and Engineering 172(2012) pp. 249-258 [Link]


    会议论文摘选
    (1)  J. Yang, D. Suckow, F. Michel, T. Lind, “Assessment of RELAP5/TRACE against VEFITA Thermal-Hydraulic Level Swell Tests”, International Conference for Advanced Nuclear Power Plants(ICAPP-14), Paper No. 14154,Charlotte, USA, April 6-9, 2014 [Link]
    (2)  D. Suckow, J. Yang, “Experiments for Containment Venting Filter Assessment in the VEFITA Facility”, Cooperative Severe Accident Research Program (CSARP-2013), Bethesda, USA, September 17-19, 2013 [Link]
    (3)  J. Yang, M. Angelis, M. S. Greenwood, M. Avery, M. Anderson, M. Corradini, J. Matos, F. Dunn, E. Feldman, “Study of Critical Heat Flux in Natural Convection Cooled TRIGA Reactors with Single Annulus and Rod Bundle Geometries”, The 15th International Topical Meeting on Nuclear Reactor Thermal-hydraulics (NURETH-15), Paper 065, Pisa, Italy, May 12-15, 2013 [Link]
    (4)  J. Yang, “Effect Of Vacuum Breaker on Passive Containment Cooling System during A Small Break LOCA”, Advances in Thermal Hydraulics (ATH-12), Paper 6627, San Diego, CA, USA, November 11-15,2012
    (5)  J. Yang, M. Avery, M. Anderson, M. Corradini, “Assessment of TRACE Code against Critical Heat Flux Experiment”, The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), N9P0003, Kaohsiung, Taiwan, September 9-13, 2012
    (6)  J. Yang, “Assessment of Reactor Passively Safe System Performance with Integral Tests on a Large Scale Test Facility”, International Workshop on Nuclear Safety and Severe Accidents (NUSSA 2012), Paper 006, Beijing, China, September 7-8, 2012
    (7)  J. Yang, S. W. Choi, J. Lim, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii, “Counterpart Experimental Study of ISP–42 PANDA Tests on PUMA Facility: Phase A”, The 20th International Conference on Nuclear Engineering (ICONE-20), Paper 54021, Anaheim, USA, July 30-August 3, 2012
    (8)  J. Yang, M. Avery, M. Angelis, M. Anderson, M. Corradini, “Critical Heat Flux in Natural Convection Cooled TRIGA Reactors with Hexagonal Bundle”, International Conference for Advanced Nuclear Power Plants (ICAPP-12), Paper 12033, Chicago, USA, June 24-28, 2012 [Link] [Link]
    (9)  J. Yang, S. W. Choi, J. Lim, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii, “Scaling, Experiment, and Code Validation on an Integral Testing Facility.”, The 14th International Topical Meeting on Nuclear Reactor Thermal-hydraulics (NURETH-14), Paper 409, Toronto, Canada, September 25-30, 2011[Link]
    (10)  J. Yang, S. W. Choi, J. Lim, D. Y. Lee, S. Rassame, T. Hibiki, M. Ishii, “Performance analysis of passively safe BWR with experimental and numerical simulation”, The 19th International Conference on Nuclear Engineering (ICONE-19), Paper 43961, Makuhari, Japan, May 2011 [Link] (10)  J. Yang, “Parameter Study of RELAP5 LOCA Simulation on Initial and Boundary Conditions”,Transactions of the American Nuclear Society, 108 (2013) pp. 980-983
    (11)  J. Yang, M. Angelis, M. Greenwood, “Critical Heat Flux in Natural Convection TRIGA Reactors with 2×2 Bundle”, Transactions of the American Nuclear Society, 107 (2012) pp. 1357-1360 [Link]
    (12)  J. Yang, M. Angelis, M. Corradini, “Critical Heat Flux in Vertical Channels at Zero Flow Condition”, Transactions of the American Nuclear Society, 106 (2012) pp. 1046-1048 [Link]
    (13)  M. Angelis, J. Yang, M. Greenwood, “Critical Heat Flux under Natural and Forced Convection”,Transactions of the American Nuclear Society, 106 (2012) pp. 1068-1070

所获荣誉和奖励

    2014 教育部春晖杯创业大赛优胜奖