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检索条件"机构=Science and Technology on Reactor System Design"
626 条 记 录,以下是561-570 订阅
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Natural convective heat transfer from a heated slender vertical tube in a cylindrical tank  16
Natural convective heat transfer from a heated slender verti...
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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015
作者: Xian, Lin Jiang, Guangming Yu, Hongxing Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China Nuclear Power Institute of China Chengdu610041 China
Natural convective heat transfer in enclosures is widely researched in extensive range of engineering application such as passive residual heat removal system (PRHRS) in nuclear plant, solar collectors and cooling of ... 详细信息
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Large eddy simulation of turbulent flow in the rod bundle with different spacer grids  23
Large eddy simulation of turbulent flow in the rod bundle wi...
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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015
作者: Wei, Zonglan Yu, Zhang Liu, Songtao Science and Technology on Reactor System Design Technology Laboratory NPIC Chengdu Sichuan China Nuclear Power Design and Research Sub-Institute NPIC Chengdu Sichuan China
Large eddy simulation(LES) of the swirling flow in the rod bundle subchannels with spacer grids are presented. According to the rod bundle flow benchmark experiment, the 5×5 rod bundle with a split-type spacer gr... 详细信息
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CFD simulation of single-phase flow in a 5×5 rod bundle with spacer grid  23
CFD simulation of single-phase flow in a 5×5 rod bundle wit...
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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015
作者: Li, Quan Jiao, Yongjun Chen, Jie Yu, Junchong Department of Engineering Physics Tinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Spacer grid is an important component in PWR fuel assemblies for its significant influence on thermal-hydraulic characteristics of the reactor core. In this study, single-phase CFD technology is used to study the flow... 详细信息
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Application of STPA to the digital reactor protection system in NPP for system safety analysis  23
Application of STPA to the digital reactor protection system...
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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015
作者: Liu, Zhaohui Wu, Zhiqiang Yang, Xiaohua School of Computer Science and Technology University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China
In NPP, the digital control system which integrated software and hardware are increasingly used to improve dependability and introduce new functionality. Traditional safety analysis can get a good result when handling... 详细信息
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The Software safety analysis based on sfta for reactor power regulating system in nuclear power plant  23
The Software safety analysis based on sfta for reactor power...
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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015
作者: Liu, Zhaohui Liao, Longtao Wu, Zhiqiang Yang, Xiaohua School of Computer Science and Technology University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China
The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control syst... 详细信息
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Research of the bundle CHF prediction based on the minimum DNBR point and the BO point methods  16
Research of the bundle CHF prediction based on the minimum D...
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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015
作者: Liu, Wei Shan, Jianqiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China School of Nuclear Science and Technology Xi'an Jiao Tong University Xi'an710049 China
There are two ways in the development of CHF correlation in rod bundles: one is based on round tube CHF correlations, considering the effects of spacers and CHF promoters, effects of cold walls, etc. The other one is ... 详细信息
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Thermal hydraulic analysis on the success criteria of steam generator tube rupture accident  23
Thermal hydraulic analysis on the success criteria of steam ...
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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015
作者: Ran, Fu Dan, Wu Ding, Shuhua Qian, Peng Libo, Qian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China Nuclear Power Institute of China Chengdu Sichuan China
The SGTR studies correspond to the double-ended guillotine rupture of a steam generator tube, which allows unimped blowdown from both ends of the severed tube. Occurrence of the SGTR a ccident leads to an inc rease in... 详细信息
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Research on Post Nonlinear Geometric Algorithm for BSS of Mixed Vibration Simulation Signals from the Impeller Pumps
Research on Post Nonlinear Geometric Algorithm for BSS of Mi...
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The second International Symposium on Socially and Technically Symbiotic systems (STSS2015);The 7th International Symposium on Symbiotic Nuclear Power systems (ISSNP2015)
作者: Sun Shi-yan XIA Hong CHEN Zhi-hui YANG Bo LIU Mei-ru Fundamental Science on Nuclear Safety and Simulation Technology Laboratory College of Nuclear Science and TechnologyHarbin Engineering University Science and Technology on Reactor System Design Technology Laboratory Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China
In nuclear power plants,for rotating equipment,a normal method for monitoring the modes and diagnosing its fault is monitoring and analyzing its vibration or noise ***,the detected signals are usually the non-linear m... 详细信息
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Neutron transport study based on assembly modular ray tracing MOC method
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Hedongli Gongcheng/Nuclear Power Engineering 2015年 第3期36卷 157-161页
作者: Tian, Chao Zheng, Youqi Li, Yunzhao Li, Shuo Chai, Xiaoming School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an710049 China State Nuclear Power Software Development Center Beijing102200 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China
It is difficulty for the MOC method based on Cell Modular Ray Tracing to deal with the irregular geometry such as the water gap between the PWR lattices. Hence, the neutron transport code NECP-Medlar based on Assembly... 详细信息
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Analysis of RCS depressurization effects on containment temperature for Qinshan phase II nuclear power plant  22
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2014 22nd International Conference on Nuclear Engineering, ICONE 2014
作者: Lili, Liu Ming, Zhang Jian, Deng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Chengdu China
A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containme... 详细信息
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