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    Development of Safety Analysis Code TACOS and Application to Fuel Qualification Test Loop 

    Source: Journal of Nuclear Engineering and Radiation Science:;2017:;volume( 003 ):;issue: 001:;page 11002
    Author(s): Jiang, Chaofei; Tian, Wenxi; Qiu, Suizheng; Su, Guanghui
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: In this study, transient analysis code of SCWRs (TACOS), with the ability of simulating transients or accidents under both supercritical water (SCW) conditions and subcritical water conditions, has been developed with ...
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    Development of a Thermal Hydraulic Analysis Code and Transient Analysis for a Fluoride Salt Cooled High Temperature Test Reactor 

    Source: Journal of Nuclear Engineering and Radiation Science:;2015:;volume( 001 ):;issue: 001:;page 11007
    Author(s): Xiao, Yao; Hu, Lin; Qiu, Suizheng; Zhang, Dalin; Guanghui, Su; Tian, Wenxi
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: The fluoridesaltcooled hightemperature reactor (FHR) is an advanced reactor concept that uses hightemperature tristructural isotropic (TRISO) fuel with a lowpressure liquid salt coolant. Design of the fluoridesaltcooled ...
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    Simulation on Pellet–Cladding Mechanical Interaction of Accident Tolerant Fuel With Coated Cladding 

    Source: Journal of Nuclear Engineering and Radiation Science:;2019:;volume( 005 ):;issue: 001:;page 11015
    Author(s): Deng, Yangbin; Wu, Yingwei; Zhang, Dalin; Tian, Wenxi; Su, G. H.; Qiu, Suizheng
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: In this study, based on the code Fuel ROd Behavior Analysis (FROBA), a thermal–mechanical analysis code initially developed for traditional UO2-Zr fuel elements by our research group, a modified version was developed to ...
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    Neutronic and Thermo-Hydraulic Analyses of Water-Cooled Blanket Based on Pressurized/Supercritical Water Conditions for CFETR 

    Source: Journal of Nuclear Engineering and Radiation Science:;2019:;volume( 005 ):;issue: 004:;page 41302
    Author(s): Cheng, Jie; Wu, Yingwei; Su, G. H.; Qiu, Suizheng; Tian, Wenxi
    Publisher: American Society of Mechanical Engineers (ASME)
    Abstract: China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and future fusion power plant. As one of the ...
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    The Development of Candling Module Code in Module In vessel Degraded Analysis Code MIDAC and the Relevant Calculation for CPR1000 During Large Break LOCA 

    Source: Journal of Nuclear Engineering and Radiation Science:;2016:;volume( 002 ):;issue: 002:;page 21002
    Author(s): Wang, Jun; Fan, Yuqiao; Zhang, Yapei; Ni, Xinghe; Tian, Wenxi; Corradini, Michael L.; Su, Guanghui; Qiu, Suizheng
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: The occurrence of Fukushima has increased the focus on the development of severe accident codes and their applications. As a part of Chinese “National Major Projects,â€‌ a module invessel degraded analysis code (MIDAC) ...
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