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检索条件"机构=National Key Laboratory of Science and Technology on Reactor System Design Technology"
1956 条 记 录,以下是421-430 订阅
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Calculation strategy for the determination of plasticity correction factors  13th
Calculation strategy for the determination of plasticity cor...
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13th Asia-Pacific Symposium on Engineering Plasticity and its Applications, AEPA 2016
作者: Du, Juan Shao, Xue Jiao Kan, Qian Hua Zhang, Ying Fu, Xiao Long Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Southwest Jiaotong University Chengdu610031 China
This paper presents an investigation about the plasticity correction factor, Ke, proposed in the RCC-M code, for the correction of elastic stress range exceeds twice the yield stress which results from both mechanical... 详细信息
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Fuel assembly design for supercritical water-cooled reactor
Fuel assembly design for supercritical water-cooled reactor
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作者: Linna, Feng Fawen, Zhu Science and Technology on Reactor System Design Technology Laboratory Third Section of Huafu Road Huayang Town Shuangliu County Chengdu Sichuan Province610213 China
The supercritical water-cooled reactor (SWCR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to t... 详细信息
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Research on integrated managing system design of NPP
Research on integrated managing system design of NPP
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2014 International Conference on Manufacturing technology and Electronics Applications, ICMTEA 2014
作者: Li, Feng-Yu Yan, Yu-Kun Zhang, Long-Fei Chen, Zhi Liao, Long-Tao Xiao, Kai Naval University of Engineering 125 mailbox Wuhan China Science and technology on reactor system design technology laboratory Nuclear Power Institute of China Chengdu China
In order to improve safety, economy and reliability of NPP(nuclear power plant) and to make the best of human, material and information resources in the NPP, a specific function model developing method for integrated ... 详细信息
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Numerical study on single Droplet's rise in steam separator by using VOF method
Numerical study on single Droplet's rise in steam separator ...
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2014 ANS Winter Meeting and Nuclear technology Expo
作者: Zhang, Di Luo, Qi Huang, Wei Wang, Kan Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
来源: 评论
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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Safety design and evaluation of loss of forced flow accident of CRS1000
Safety design and evaluation of loss of forced flow accident...
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2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018
作者: Dan, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Changshun Road No.328 ShuangLiu County Chengdu610213 China
CSR1000 (China supercritical water reactor, 1000MWe) was developed by NPIC, as BWR, the pressure vessel's reactor and directly circulate loop was adopted, however, the coolant will encounter double flow pass in th... 详细信息
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Development and Verification of a New Depletion, Activation and Radiation Source Term Calculation Code  23rd
Development and Verification of a New Depletion, Activation ...
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23rd Pacific Basin Nuclear Conference, PBNC 2022
作者: Wen, Xingjian Zhai, Zian Tang, Songqian Tian, Chao Liu, Zhouyu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China Xi’an Jiaotong University Xi’anShaanxi China
Existing depletion and source term calculation codes lack flexible interfaces, which is difficult to meet the actual engineering design needs. Based on the Chebyshev rational approximation method, a new depletion, act... 详细信息
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2-D/1-D method for neutron transport based on large-scale parallel computation
2-D/1-D method for neutron transport based on large-scale pa...
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2014 ANS Winter Meeting and Nuclear technology Expo
作者: Wu, Wenbin Li, Qing Wang, Kan Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
来源: 评论
Enhanced thermal isolation in porous thermal barrier coatings by the formation of pore guided thermal-shock cracks
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science China(Technological sciences) 2023年 第4期66卷 1007-1017页
作者: ZHEN Yu WU KaiJin LIU MengQi ZHENG SongLin HE LingHui YU Yin NI Yong CAS Key Laboratory of Mechanical Behavior and Design of Materials Department of Modern MechanicsCAS Center for Excellence in Complex System MechanicsUniversity of Science and Technology of ChinaHefei 230026China National Key Laboratory of Shock Wave and Detonation Physics China Academy of Engineering PhysicsInstitute of Fluid PhysicsMianyang 621900China
Pore structure design is an effective strategy to tailor the thermal isolation capability of thermal barrier coatings(TBCs).Pursuing optimal porosity is crucial to balance the requirements of thermal isolation and mec... 详细信息
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Preliminary Development of a Simulation Capability for Zircaloy Clad Ballooning in LOCA
Preliminary Development of a Simulation Capability for Zirca...
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International Conference on Water reactor Fuel Performance Meeting, WRFPM 2023
作者: Li, A. Wei Wu, B. Xiaoli State Key Laboratory for Strength and Vibration of Mechanical Structures Xi’an Jiaotong University Xi’an710049 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China
Zircaloy clad is widely known to be vulnerable to ballooning and burst in a loss-of-coolant accident (LOCA) typical of high-temperature steam environment, due to internal and external pressure difference and degraded ... 详细信息
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