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检索条件"机构=Science Technology on Reactor System Design Technology Laboratory"
1277 条 记 录,以下是101-110 订阅
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design and Implementation of DCS design List Automatic Generation and Check system Based on LabVIEW  7th
Design and Implementation of DCS Design List Automatic Gener...
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7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023
作者: Deng, Xiao-Jun Zhang, Xu Feng, Shi-Man Peng, Hao Yao, Ying-Fan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
The DCS configuration design process involves a large number of variables and multiple lists, and designers need to spend more time on the preparation, checking and maintenance of the lists. Manual compilation of file... 详细信息
来源: 评论
Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast reactor  23rd
Atomic Scale Simulation on Liquid Metal Embrittlement Induce...
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23rd Pacific Basin Nuclear Conference, PBNC 2022
作者: Bida, Zhu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Lead-Cooled fast reactor, which uses liquid lead or lead-bismuth alloy as coolants, is one of the six main reactors of the fourth generation reactor system. Most recently, the Lead-Cooled fast reactor received a lot o... 详细信息
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NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHARACTERISTICS IN LBE-COOLED HELICAL CRUCIFORM FUEL AND WIRE-WRAPPED FUEL  31
NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHA...
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2024 31st International Conference on Nuclear Engineering, ICONE 2024
作者: Qi, Zhang Haoyu, Wang Cao, Junxian Yuanming, Li Chenxi, Li Zhenhai, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
The lead bismuth eutectic (LBE) cooled fast reactor in one of the General 4 advanced reactors, which plays significant role in the nuclear fuel cycle. In the recent years, the wire-wrapped fuel with four wires surroun... 详细信息
来源: 评论
Simulation of esprit of hualong secondary passive residual heat removal system by using relap5  16
Simulation of esprit of hualong secondary passive residual h...
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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015
作者: Feng, Li Science and Technology on Reactor System Design Technology Laboratory ChangShun Avenue 328 ShuangLiuCounty ChengDu China
A 3rd generation reactor called HuaLong is developed by China National Nuclear Corporation (CNNC). S econdary passive system (PRS) is designed to remove the residual heat during 72 hours after the reactor trips. To ve... 详细信息
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Effect of clamping failure on flow induced vibration and fretting wear of fuel rods
Effect of clamping failure on flow induced vibration and fre...
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ASME 2018 Pressure Vessels and Piping Conference, PVP 2018
作者: Huan-Huan, Qi Zhi-Peng, Feng Fu-Rui, Xiong Nai-Bin, Jiang Qian, Huang Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Fuel rods are subjected to both axial and lateral flow in the reactor core. In this study, we present a study on the flow induced vibration (FIV) and fretting wear of fuel rod with failed clamping at grids. First, acc... 详细信息
来源: 评论
EXPLORATION AND APPLICATION OF ENSEMBLE LEARNING METHOD ON PREDICTING TWO-PHASE FLOW KEY PARAMETERS OF RECTANGULAR CHANNEL  29
EXPLORATION AND APPLICATION OF ENSEMBLE LEARNING METHOD ON P...
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2022 29th International Conference on Nuclear Engineering, ICONE 2022
作者: Huang, Qingyu Yu, Yang Zeng, Hui Yang, Hui Zhang, Siyuan Yang, Kunlin Lin, Yuanfeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Among the current mainstream nuclear reactor thermal-hydraulic calculation system analysis softwares (such as RELAP5, TRACE, etc.), different empirical or semi-empirical relationships are employed to calculate two-pha... 详细信息
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Comparative study on thermal stress analysis and fatigue curve in stress and fatigue calculation of nuclear equipment  28
Comparative study on thermal stress analysis and fatigue cur...
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2021 28th International Conference on Nuclear Engineering, ICONE 2021
作者: Shao, Xuejiao Xie, Hai Yixiong, Y. Zhang, Liping Fu, Xiaolong Mi, Xue Li, Hui Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
In the stress and fatigue evaluation of primary nuclear equipment, thermal stress calculation and fatigue curve have a great impact on the calculation results. There is a problem of stability in the step integral when... 详细信息
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Research on the effect of grid clamping failure on flow elastic stability and vortex shedding of fuel rod  26
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2018 26th International Conference on Nuclear Engineering, ICONE 2018
作者: Huan-Huan, Qi Zhi-Peng, Feng Nai-Bin, Jiang Qian, Huang Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of gr... 详细信息
来源: 评论
STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCLEAR reactor UNDER OCEAN CONDITION  29
STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCL...
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2022 29th International Conference on Nuclear Engineering, ICONE 2022
作者: Cheng, Kun Chu, Xiao Qiu, Zhifang Xian, Lin Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Different from land-based reactors, floating reactors are affected by the ocean condition. Numerical simulation is conducted to study the effects of heaving and rolling motion on typical floating nuclear reactor (FNR)... 详细信息
来源: 评论
Flow-induced vibration analysis of the check valve  25
Flow-induced vibration analysis of the check valve
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25th International Congress on Sound and Vibration 2018: Hiroshima Calling, ICSV 2018
作者: Shuai, Liu Peng, Feng Zhi Zhou, Jiang Xiao Xi, Lv Shou, Zhang Feng Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
As an important over-current adjustment unit in the nuclear power plant piping system, check valve often causes some vibration and noise problems. The study of the existing valves mainly focuses on the simple or fixed... 详细信息
来源: 评论