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检索条件"机构=Science and Technology on Reactor System Design Technology"
2733 条 记 录,以下是101-110 订阅
排序:
Preliminary research on the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel
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International Journal of Advanced Nuclear reactor design and technology 2019年 1卷 51-56页
作者: Changbing, Tang Yongjun, Jiao Yuanming, Li Yi, Zhou Hua, Pang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China No. 328 Section 1 Changshun Avenue ChengduSichuan610041 China
FCM fuel (fully ceramic micro-encapsulated fuel) which is a promising ATF (accident tolerant fuel) candidate fuel. Establishing the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel accura... 详细信息
来源: 评论
Application of feedforward predictive control in DC furnace coordination system
Application of feedforward predictive control in DC furnace ...
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2020 International Symposium on Automation, Mechanical and design Engineering, SAMDE 2020
作者: Deng, Zhiguang Zhengxi, H.E. Biwei, Z.H.U. Jialiang, Z.H.U. Xin, L.V. Qian, W.U. National Key Labortory of Science and Technology on Reactor System Design Technology Chengdu610213 China
In order to solve the control difficulties such as large dynamic delay, serious coupling between machine and furnace and strong nonlinear, the feedforward and feedback predictive control is proposed to control the coo... 详细信息
来源: 评论
Multi-objective Optimization of Non-uniform Beam for Minimum Weight and Sound Radiation
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Transactions of Tianjin University 2017年 第4期23卷 380-393页
作者: Furui Xiong Mengxin He Yousef Naranjani Qian Ding Jianqiao Sun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China School of Mechanical Engineering Tianjin University School of Engineering University of California Merced
A multi-objective optimization of non-uniform beams is presented for minimum radiated sound power and weight. The transfer matrix method is used to compute the structural and acoustic responses of a non-uniform beam a... 详细信息
来源: 评论
Research of Advanced Control Algorithm in Primary Loop Control system  5th
Research of Advanced Control Algorithm in Primary Loop Contr...
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5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020
作者: Deng, Zhi-Guang Zhu, Bi-Wei Wu, Qian He, Peng Xiang, Mei-Qiong Xu, Tao Qing, Yue National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610213 China
The primary loop system of nuclear power plant is a large-scale control system with serious coupling between loops and complex objects. In order to solve the problem of integrated control of primary loop core power, s... 详细信息
来源: 评论
Assessment of Coupled Effect of Steam Pipes on Seismic Analysis of reactor Coolant system  7
Assessment of Coupled Effect of Steam Pipes on Seismic Analy...
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7th International Conference on Environmental science and Civil Engineering, ESCE 2021
作者: Li, Lijuan Liu, Zhenyu Li, Hui Ye, Xianhui Ai, Honglei National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu Sichuan China
reactor coolant system is connected with auxiliary pipes, surge pipes and steam pipes. In order to facilitate the dynamic analysis of reactor coolant system, it is necessary to decouple the pipes connected with the re... 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
Optimization of spatial structure designs of control rod using Monte Carlo code RMC
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Frontiers in Energy 2021年 第4期15卷 974-983页
作者: Hao LUO Mancang LI Shanfang HUANG Minyun LIU Kan WANG Department of Engineering Physics Tsinghua UniversityBeijing 100084China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology China National Nuclear Corporation Nuclear Power Institute of ChinaChengdu 610213China
Control rod is the most important approach to control reactivity in reactors,which is currently a cluster of pins filled with boron carbide(B4C).In this case,neutrons are captured in the outer region,and thus the inne... 详细信息
来源: 评论
Sensitivity analysis for dynamical response of reactor coolant system based on optimus
Sensitivity analysis for dynamical response of reactor coola...
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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference
作者: Yanli, Yuan Xianhui, Ye Lijuan, Li Feng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at t... 详细信息
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Development and verification of SNTA code system for SCWR core steady state analysis  23
Development and verification of SNTA code system for SCWR co...
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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015
作者: Wang, Lianjie Zhao, Wenbo Yang, Ping Ma, Yongqiang Lu, Di Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
A coupled neutronics/thermal-hydraulics three dimensional code system SNTA is developed for SCWR core steady state analysis by modular coupling the improved neutronics nodal methodological code and SCWR thermal-hydrau... 详细信息
来源: 评论
A NUMERICAL SIMULATION METHOD FOR THE VIBRATION REDUCTION OF MR ISOLATOR BASED ON IMPEDANCE TEST  30
A NUMERICAL SIMULATION METHOD FOR THE VIBRATION REDUCTION OF...
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30th International Congress on Sound and Vibration, ICSV 2024
作者: Wenzheng, Zhang Peng, Xiangfeng Zhihao, Yuan Guojiang, Liao Xiaozhou, Jiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China
It’s important to choose suitable method to simulate the vibration reduction of the isolators. For a long time, scholars and engineers have been working hard to reduce the error between numerical simulation and exper... 详细信息
来源: 评论