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检索条件"机构=Science and Technology on Reactor System Design Technology"
2723 条 记 录,以下是21-30 订阅
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Research on Hybrid Modeling Method Based on Mechanism Model and BP Neural Network  4
Research on Hybrid Modeling Method Based on Mechanism Model ...
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4th Asia Conference on Information Engineering, ACIE 2024
作者: Chen, Jie Zheng, Yanqiu Chu, Qing Jiang, Guanfu Li, Jing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China
reactor system modeling is the premise of reactor control system design. The common commercial thermal and hydraulic analysis programs like RELAP5, RETRAN have complex structure and slow operational speed, which are n... 详细信息
来源: 评论
ANALYSIS OF SMR reactor COOLANT system IN APROS  31
ANALYSIS OF SMR REACTOR COOLANT SYSTEM IN APROS
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2024 31st International Conference on Nuclear Engineering, ICONE 2024
作者: Zhu, Ye Xingbo, Wang Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China
The reactor coolant system (RCS) is the primary system of the nuclear power plant primary cycle. Some small module reactors (SMR) applied once-through steam generator and integrated reactor vessel. The characteristic ... 详细信息
来源: 评论
RESEARCH ON THE OPTIMIZATION OF THE reactor COOLANT system STRUCTURE  30
RESEARCH ON THE OPTIMIZATION OF THE REACTOR COOLANT SYSTEM S...
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30th International Congress on Sound and Vibration, ICSV 2024
作者: Yanli, Yuan YeXianhui Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China
The dynamic analysis of reactor coolant system is the key step of mechanics evaluation, and with the continuous improvement of seismic design benchmarks and system design requirements, it is urgent to carry out the re... 详细信息
来源: 评论
Transient fuel performance analysis of UO_(2)–BeO fuel with composite SiC coated with Cr cladding based on multiphysics method
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Nuclear science and Techniques 2023年 第12期34卷 93-108页
作者: Chun‑Yu Yin Shi‑Xin Gao Sheng‑Yu Liu Rong Liu Guang‑Hui Su Li‑Bo Qian School of Nuclear Science and Technology Xi’an Jiaotong UniversityXi’an 710049China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China School of Electric Power South China University of TechnologyGuangzhou 510640China
The transient multiphysics models were updated in CAMPUS to evaluate the accident-tolerant fuel performance under accident *** is a fuel performance code developed based on *** simulated results of the UO_(2)–Zircalo... 详细信息
来源: 评论
Annealing-induced gradient nanostructured FeCrAlTiMo high-entropy alloy coatings with significantly enhanced wear resistance
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Tungsten 2024年 第3期6卷 544-548页
作者: Jian Yang Ming-Yang Zhou Ji-Jun Yang Key Laboratory of Radiation Physics and Technology of Ministry of Education Institute of Nuclear Science and TechnologySichuan UniversityChengdu 610064China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China
Recently,the amorphous high-entropy alloy(HEA)coatings exhibit an impressive resistance to the lead–bismuth eutectic(LBE)corrosion and ion irradiation,showing great application potential for lead-cooled fast reactors... 详细信息
来源: 评论
Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water
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Chinese Journal of Mechanical Engineering 2023年 第4期36卷 326-338页
作者: Jun Wang Haojie Li Zhengyang Li Yujie Lei Quanyao Ren Yongjun Jiao Zhenbing Cai Key Lab of Advanced Technologies of Materials Tribology Research InstituteSouthwest Jiaotong UniversityChengdu 610031China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China
In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly ***,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested ... 详细信息
来源: 评论
A Tutorial on Quantized Feedback Control
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IEEE/CAA Journal of Automatica Sinica 2024年 第1期11卷 5-17页
作者: Minyue Fu IEEE the School of System Design and Intelligent Manufacturing Shenzhen Key Laboratory of Control Theory and Intelligent Systems and Centre for Control Science and TechnologySouthern University of Science and TechnologyShenzhen 518055China
In this tutorial paper, we explore the field of quantized feedback control, which has gained significant attention due to the growing prevalence of networked control systems. These systems require the transmission of ... 详细信息
来源: 评论
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
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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
来源: 评论
A DEEP LEARNING-BASED APPROACH FOR PREDICTING THE STOCHASTIC PROCESS OF reactor ACCIDENTS  30
A DEEP LEARNING-BASED APPROACH FOR PREDICTING THE STOCHASTIC...
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30th International Conference on Nuclear Engineering, ICONE 2023
作者: Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
Although the prediction of system behavior plays a crucial role in accident management, there is very limited research on this subject in the nuclear industry. Accident process prediction methods are usually implement... 详细信息
来源: 评论