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design of Emulation system for Safety DCS 5th

Design of Emulation System for Safety DCS

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

作者： Peng, Hao Zhang, Xu Deng, Zhi-Guang Chen, Qi Zhang, Yu Jiang, Wei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811634550

DCS (Digital Control system) is a very important part in safety protection in NPP (Nuclear Power Plant). Virtual DCS realizes the algorithm configuration of physical DCS running on general computers and gets rid of the limitation of hardware. In this paper, an emulation system (virtual DCS) is designed using virtual DPU technology. The virtual DCS can be used to verify the algorithm logic with the closed-loop system, which is composed by virtual DCS and digital reactor processing model together. The virtual DCS and controlled object are closed to test the control effects and in this way to verify the design of the NPP DCS. For the purpose of scientific research, training and design verification, virtual DCS should have malfunction simulation and other functions. These functions reflect the superiority of emulation technology. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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design and Simulation Verification of Intelligent Perception and Adaptive Control system for Control Rod Drive Mechanism 30

Design and Simulation Verification of Intelligent Perception...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Yang, Hui Lin, Yuanfeng Zeng, Hui Huang, Qingyu Zhang, Siyuan Yang, Kunlin Wang, Yuanmei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Nuclear power plant status monitoring and equipment fault diagnosis are important guarantees for the safe operation of nuclear power plants. With the development of artificial intelligence technology, fault diagnosis of nuclear reactor equipment has entered the stage of intelligent diagnosis and prediction. The control rod drive mechanism (CRDM) is the actuator of the reactor control and nuclear safety protection system. In the normal state, CRDM realizes the stepping movement of the control rod through the alternating action of the three sets of coil currents. When an abnormality in the coil current is detected, it means that the CRDM movement fails, and in more serious cases, it may cause the control rod to accidentally drop. Therefore, in order to prevent the occurrence of control rod stuck, sliding rod and driving failure in the reactor and to ensure the safe and economic operation of the reactor, a necessary measure is to effectively monitor the operation status of the CRDM. In this paper, the overall architecture of an intelligent perception and adaptive control system for CRDM is studied. By analyzing the multi-condition and multi-mode coupling relationship in the time dimension and spatial correlation between the coil current and vibration signals, a condition discrimination and fault diagnosis model for CRDM based on hybrid network of CNN+BiGRU is established. Furthermore, a digital adaptive control model of control rod drive system is built based on the real-time simulation system dSPACE. Through functional verification, the system can perform real-time perception and identification of normal operation conditions, fault diagnosis and adaptive control based on the coil current and vibration signals. The recognition accuracy rate for various operation conditions reaches more than 99.35%, and the control accuracy of the sequential current is improved around ±4% of the rated value under various external environment changes, providing a supportive reference for

关键词： Failure analysis

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Nuclear Safety-Class DCS Hardware Logic Implementation Scheme Based on Modularization design 6th

Nuclear Safety-Class DCS Hardware Logic Implementation Schem...

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6th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2021

作者： Hu, Yan-Liang Zhang, Xu Chen, Shi-Yong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811911804

At present, nuclear safety-class DCS can deal with the common cause failure of digital control system software system by building hardware logic control circuit. Different from the digital control system, the hardware logic control circuit is mainly designed and realized by switches, relays and delay relays, which improves the diversity of the system. However, when the safety-class DCS put into operation, the hardware logic control system fault troubleshooting, regular test and other work is relatively heavy. In order to improve the testability and maintainability of safety-class DCS hardware logic control system, this paper proposes a scheme based on the modular hardware logic design method and the matching hardware logic test device. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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Application of Optimal design in Nuclear Safety-Class DCS design 6th

Application of Optimal Design in Nuclear Safety-Class DCS De...

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6th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2021

作者： Zhang, Xu Yao, Zhang Chen, Shi-Yong Peng, Hao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811911804

In the design process of nuclear safety-class digital control system (DCS), there may be mistakes in the details of various drawings for the large number of drawings and heavy design task. In safety-class DCS, with the trait of redundancy channel and equipment, cabinets within the same system have high similarity. At the same time, different units also have high similarity. According to the characteristics of these requirements, the deep development of engineering design software has realized a series of functions such as automatic drawing design, content automatic check and drawing automatic conversion. Combining the above design ideas with software automation technology, the design process is optimized, the engineering design efficiency is greatly improved, and the engineering design cycle is shortened. The optimization design method in this paper has been applied to the safety-class DCS design of Unit 1&2 of Zhangzhou nuclear power plant, and has achieved good results. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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In-situ TEM observation of the evolution of helium bubbles in Mo during He^(+)irradiation and post-irradiation annealing

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Chinese Physics B 2021年第8期30卷 102-112页

作者： Yi-Peng Li Guang Ran Xin-Yi Liu Xi Qiu Qing Han Wen-Jie Li Yi-Jia Guo College of Energy Xiamen UniversityXiamen 361102China Fujian Research Center for Nuclear Engineering Xiamen 361102China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

The evolution of helium bubbles in purity Mo was investigated by in-situ transmission electron microscopy(TEM)during 30 keV He^(+)irradiation(at 673 K and 1173 K)and post-irradiation annealing(after 30 keV He^(+)irradiation with the fluence of 5.74×10^(16)He^(+)/cm^(2)at 673 K).Both He^(+)irradiation and subsequently annealing induced the initiation,aggregation,and growth of helium *** had a significant effect on the initiation and evolution of helium *** higher the irradiation temperature was,the larger the bubble size at the same irradiation fluence would *** 1173 K irradiation,helium bubbles nucleated and grew preferentially at grain boundaries and showed super large size,which would induce the formation of *** the same time,the geometry of helium bubbles changed from sphericity to *** polyhedral bubbles preferred to grow in the shape bounded by{100}*** statistical analysis of the characteristic parameters of helium bubbles,the functions between the average size,number density of helium bubbles,swelling rate and irradiation damage were ***,an empirical formula for calculating the size of helium bubbles during the annealing was also provided.

关键词： helium bubbles in-situ TEM observation ion irradiation annealing molybdenum

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RESEARCH ON VIBRATION CHARACTERISTICS OF FLOATING RAFT ISOLATION system BASED ON STRUCTURAL PARAMETERS 30

RESEARCH ON VIBRATION CHARACTERISTICS OF FLOATING RAFT ISOLA...

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30th International Congress on Sound and Vibration, ICSV 2024

作者： Wang, Yu Lai, Jian-Yong Cai, Long-Qi Li, Zhao-Wen Liu, Jia Wang, Ran Liu, Shuai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789090390581

As an important noise source for nuclear power systems, pump equipment needs continuous operation and always in a high vibration level. With the development of vibration isolation technology, floating raft vibration isolation is increasingly favored by people as a new and effective vibration control method. At present, the research on the floating raft vibration isolation system mostly focuses on the influence of the basic stiffness, the damping, and the mass-ratio. And there were few studies on the structural style of the floating raft. In order to explore the effects of the structural parameters on the vibration characteristics of floating raft system. The floating raft vibration isolation system of the pump equipment was researched in this paper. The finite element analysis model of the floating raft vibration isolation system was established by ANSYS. And the Control-Variable method was used to explore the effects of parameters such as the elasticity and density of the raft structure. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Nuclear energy

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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

关键词： alloy resistance wear

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Investigations on friction resistance of forced circulation in a circular tube under inclination and rolling conditions

Investigations on friction resistance of forced circulation ...

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18th International Conference on Nuclear Engineering, ICONE18

作者： Du, Sijia Hong, Zhang Department of Engineering Physics National Key Laboratory of Science and Technology on Reactor System Design Technology of Nuclear Power Institute of China China National Key Laboratory of Science and Technology on Reactor System Design Technology of Nuclear Power Institute of China China

ISBN: (纸本)9780791849323

The friction resistance under marine conditions is very important to a marine reactor system in the ocean. The traditional empirical or semi-empirical relationships were used in thermo-hydraulic system programmes at the present time, while the applicability of these relationships under ocean conditions was not validated adequately. Experiments have been conducted to study the friction resistance of forced circulation when the test tube was at inclined attitude and rolling model. The experimental setup consists of stainless steel cylinder tube as test section with 10mm inside diameter and 1200mm length, is subjected to a constant wall heat flux boundary condition. The investigation covers Reynolds number range from 1.0e+4 to 1.2e+5, and the test section angles of inclination including 10°, 20°and 30°, amplitudes of rolling angle including 10°, 15°and 20°with the rolling periods of 10s, 15s and 20s. Pressure drop was measured to calculate the coefficient of friction resistance. CFD method was used to simulate the flow characteristics under inclination and rolling conditions, the results were compared with experimental data. The investigations showed that the influence of inclination and rolling to friction coefficient is not obvious. © 2010 by ASME.

关键词： Reynolds number

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MATLAB/Simulink-Based Simulation Study of Small Pressurized Water reactor Nuclear Steam Supply system 4

MATLAB/Simulink-Based Simulation Study of Small Pressurized ...

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4th Asia Conference on Information Engineering, ACIE 2024

作者： Chen, Jie Xiao, Kai Mengwei, Mengwei Zhao, Zhao Huang, Ke Yang, Pengcheng Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9798350371925

Accurate and reliable modeling and simulation is the key prerequisite for intelligent control technology research of nuclear power plant. For the nuclear steam supply system (NSSS) of small pressurized water reactor under bi-stable steady state operation scheme, this paper firstly establishes the models of the core, regulator, steam generator and other components. Then, the nuclear steam supply system platform (NSSSP) of small pressurized water reactor is built based on Matlab/Simulink. Simulations are carried out to analyze the normal operation transient and load rejection transient in the water supply flow control mode. The accuracy and reliability of NSSSP are verified to beconsistent with the commercial program RETRAN. © 2024 IEEE.

关键词： Simulink

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A combined reactivity control pattern on a floating nuclear power plant 27

A combined reactivity control pattern on a floating nuclear ...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Zhou, Bingyan Song, Danrong Chen, Zhang Chai, Xiaoming Xiang, Hongzhi Hu, Yuying Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

In this study, a new combined reactivity control pattern is presented to lower the power peaking factor on a floating nuclear power plant. In the pattern the whole cycle length is divided into several boron-adjusting periods. During one boron-adjusting period, the soluble boron concentration remains unchanged and the core reactivity change is controlled by control rods. Boron concentration would only be changed at the end of a boron-adjusting period. In the whole cycle the boron concentration would be changed for two to four times, which is acceptable for a floating nuclear power plant. To get an optimized solution, a model of the combined reactivity control pattern was established, which contained boron concentration and control rod pattern (CRP) as coupled variables. A particle swarm optimization with novel mutation operator (PSONM) was applied to optimize the solution. The model was practiced based on a small modular reactor ACP100 core. The solution was optimized separately for equal division of two, three, four and five boron-adjusting periods. The results show that the combined reactivity control pattern can get a lower power peak factor than the original rod-controlled-only pattern without shortening the cycle length. Therefore, the safety of the floating nuclear power plant can be improved. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

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