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Research of SSAE-GPC in Coordinated Control system of Nuclear Power Plant 5th

Research of SSAE-GPC in Coordinated Control System of Nuclea...

5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Deng, Zhi-Guang Wu, Qian Zhu, Bi-Wei Lv, Xin Zhu, Jia-Liang Xu, Si-Jie Wang, Xue-Mei National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610213 China

ISBN: (纸本)9789811634550

For the problem of poor control effect of conventional PID controllers in complex systems, this paper fully combines the advantages of deep learning in feature extraction, regression prediction, and predictive control in dealing with multivariable, strong coupling and other issues. Firstly, the predictive model controller is constructed through the stacked sparse auto-encoder (SSAE) deep network, and then the stepped generalized predictive control (SGPC) is used as the rolling optimization controller. Thus, the SSAE-SGPC neural network predictive controller is built. Finally, a simulation experiment is carried out in the dual-input dual-output multi-variable coordinated control system of nuclear power plant. Through a series of simulations such as set value disturbance, internal disturbance and external disturbance, compared with conventional PID control, SSAE-GPC controller has better control effect. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Controllers

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Structures and energetics of semicoherent interfaces of precipitates in hcp/bcc systems:A molecular dynamics study

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Journal of Materials science & technology 2021年第8期67卷 50-60页

作者： Jin-Yu Zhang Fu-Zhi Dai Zhi-Peng Sun Wen-Zheng Zhang Key Laboratory of Advanced Materials(MOE) School of Materials Science and EngineeringTsinghua UniversityBeijing 100084China Science and Technology of Advanced Functional Composite Laboratory Aerospace Research Institute of Materials&Processing TechnologyBeijing 100076China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610041China

α/β(hcp/bcc)interfaces are of great importance in the microstructure development and the mechanical properties of titanium and zirconium *** work contributes to the study of interface energetics and interfacial structures of the precipitate in the hcp/bcc system based on a simulation study using molecular statics(MS)and molecular dynamics(MD).The input orientation relationship(OR)was calculated based on the O-line *** on the energy of the interfaces containing the invariant line(IL),two preferred facets were determined by the Wulff construction,which explained the observed orientations of the habit plane(HP)and the side facet(SF).The deviation of the observed precipitate morphology from the equilibrium shape was discussed in terms of interface *** structures of the interfaces surrounding a three-dimensional(3 D)precipitate,including the preferred facets and the end face,were obtained at the atomic *** simulated dislocation structures and atomic structures in these interfaces are in good agreement with those of the experimental observations for Ti-Cr alloys.A method was suggested for modifying the O-cell structure with the guidance of the relaxed structure,yielding consistency between the calculated dislocation structure based on the generalized O-element approach and the simulation results.

关键词： Interface Misfit dislocation Precipitate

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Research on the domestic materials of forging Z12CN13 stainless steel for Hold-Down Spring of reactor

Research on the domestic materials of forging Z12CN13 stainl...

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Energy Materials 2014

作者： Li, Ning Ying, Luo Xin, Xia Hua, Du National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610041 China

ISBN: (纸本)9781119027942

Effects of main alloying elements, quenching and tempering temperatures and quenching cooling velocity on mechanical properties of Z12CN13 stainless steel were obtained by controlling the material composition and optimizing the heat treatments. Evaluation of domestic forging Z12CN13 stainless steel for Hold-Down Spring of reactor had been completed, and all parameters met the requirement of design criteria national standard. The main alloying elements should be selected from upper limits to reach the technical standards of the strength and toughness properties. The strength, plasticity and toughness properties of domestic forging Z12CN13 stainless steel will decrease with increasing the quenching temperature. On the contrary, the strength decreases, plasticity and the toughness properties will increase with the increasing of tempering temperature. The tensile strength and impact toughness of materials are influenced obviously with the quenching cooling velocity. Improving the cooling velocity could meet the technical standards of the strength and toughness properties. The comprehensive mechanical properties and combination of strength and toughness is the best when the domestic forging Z12CN13 stainless steel is quenched at 985°C and then cooling fast and aged at 640°C.

关键词： Stainless steel

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Research and design on the Control Rod Drive Mechanism Power Supply and Control Strategy for the Nuclear Power Plants Based on Silicon-Controlled Rectifier 10

Research and Design on the Control Rod Drive Mechanism Power...

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10th IEEE International Power Electronics and Motion Control Conference, IPEMC 2024 ECCE Asia

作者： Xu, A. Mingzhou Gao Zheng, B. Zhao, C. Zerun Ping Yang, D. Peng, E. Yusheng Science and Technology on Reactor System Design Technology Laboratory China Southwest Jiaotong University China

ISBN: (纸本)9798350351330

Due to the characteristics of large output current ripple and long current rise and fall times in the current rod control power supply basedsilicon-controlled rectifier, a decoupling nonlinear adaptive control is *** proposed control strategy can improve the rod control power supply of the speed of respond and the steady-state current accuracy. Practical application has shown that this method solves the output current ripple and the rise and fall time of the current, meeting the working requirements of rod control power supply in nuclear power plants. © 2024 IEEE.

关键词： Nuclear fuels

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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

ISBN: (纸本)9789811634550

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, stabilizer pressure and water level, and steam generator water level, this paper applies improved predictive control algorithms to each sub-control system on the basis of mechanism modeling. For the strong non-linear problem of water level control, a multi-model predictive controller is constructed through linearization at each typical operating point. The simulation results show that the application of advanced predictive control algorithm to the primary loop control system has achieved good control performance, and it also lays a solid foundation for subsequent practical engineering applications. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Water levels

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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 reactor coolant system. Although some international standards stipulating requirements for the decoupling of subsystems from the main system, in order to further explore the influence of decoupling steam pipes, the seismic response of reactor coolant system with or without steam pipes is simulated by using ANSYS program. Firstly, the modal analysis of reactor coolant system is carried out to analysis the influence of steam pipes on the resonant frequency of reactor system. Then, the standardized seismic dynamic analysis of reactor system with or without steam pipes is carried out. The nozzle load (inlet and outlet nozzle load of main device, such as reactor pressure vessel, steam generator, pump) and support load (of main device such as reactor pressure vessel, steam generator, pump) are output and compared. According to nozzle load and support load, it can be seen that the steam pipes have little influence on the seismic analysis of reactor coolant system, i.e. steam pipes can be decoupled in the seismic analysis of reactor coolant system. © Published under licence by IOP Publishing Ltd.

关键词： Steam generators

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CRRL: Contrastive Region Relevance Learning Framework for Cross-city Traffic Prediction

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Information Fusion 2025年 122卷

作者： Dai, Genan Yi, Wenfeng Cao, Jinzhou Gong, Zhaoya Fu, Xianghua Zhang, Bowen College of Big Data and Internet Shenzhen Technology University Shenzhen China China College of Applied Science Shenzhen University Shenzhen China School of Urban Planning & Design Peking University Shenzhen Graduate School Shenzhen China Key Laboratory of Earth Surface System and Human-Earth Relations of Ministry of Natural Resources of China Peking University Shenzhen Graduate School Shenzhen China

Accurate traffic prediction plays a pivotal role in smart urban systems by enabling effective traffic management and improving the quality of life for residents. While intra-city traffic prediction has been extensively studied, cross-city traffic prediction remains a challenging task due to data scarcity in target cities, domain gaps between cities, and the risk of negative transfer. To address these challenges, we propose a novel Contrastive Region Relevance Learning (CRRL) framework. CRRL leverages contrastive learning to align region-level spatiotemporal patterns and transfer high-quality knowledge between cities. Specifically, CRRL integrates three key modules: (1) a Dual-branch Spatiotemporal Encoder (DSE) to capture region-pair and high-order region group relationships, (2) a Pseudo-Label Generation (PLG) module for aligning cross-city embedding similarities, and (3) a Reliable Region Selection (RRS) module for contrastive learning within consistent regions. Extensive experiments on real-world datasets demonstrate that CRRL achieves state-of-the-art performance in cross-city traffic prediction under data-scarce scenarios, showcasing its practicality and effectiveness in addressing urban traffic challenges. © 2025

关键词： Contrastive Learning

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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 accurately, would be very helpful for the optimal design of FCM fuel. In this research, TRSIO particles were assumed to be uniformly distributed in the silicon carbide (SiC) matrix, and the corresponding representative volume element (RVE) models were established. Based on ABAQUS commercial software, the thermal effects and irradiation effects of FCM fuel were all introduced into this numerical simulation by coding the UMAT subroutine, and then the irradiation - thermal - mechanical coupling behavior simulation method of FCM fuel was preliminary established. Compared with the simulation result of BISON software in open literature, the simulation results of this numerical simulation method are basically consistent with those of BISON software, the effectiveness of this numerical simulation method was preliminary verified. © 2019 Xi'an Jiaotong University

关键词： Irradiation

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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 accurately and efficiently. The multi-objective particle swarm optimization technique is applied to search the Pareto optimal solutions that represent various compromises between weight and sound radiation. Several constraints are imposed, which substantially reduce the volume fraction of feasible solutions in the design space. Two non-uniform beams with different boundary conditions are studied to demonstrate the multi-objective optimal designs of the structure. © 2017, Tianjin University and Springer-Verlag Berlin Heidelberg.

关键词： Acoustic generators Acoustic properties Acoustic wave propagation Acoustic wave scattering Optimal systems Pareto principle Particle swarm optimization (PSO) Transfer matrix method

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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 to simulate the fretting damage evolution process of cladding in the PWR *** can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal *** fretting wear behavior of Zr-4 alloy under different temperature environments was *** addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light *** show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.

关键词： Fretting wear Cladding High temperature and high pressure Zirconium alloy

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