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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Peiying, Li Jian, Deng Lei, Zhong Libo, Qian Rong, Cai Yugao, Ma Hongxing, Yu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

With liquid metal like lead-bismuth alloy(LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt=A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental results, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Liquid metals

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Experimental Research on Data Management Method of Instrument and Control Emulation system 1

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7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Peng, Hao Zhang, Xu Tang, Lei Wu, Yan-qun Zhang, Yu Zheng, Meng-bing Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

The instrument and control (I&C) emulation system that developed based on virtual technology is an effective mapping of physical I&C system. It can realistically reflect the properties of physical system, and can be applied to verify profit intelligent control strategies. In the stage of solution design, it is definitely necessary to analyze data storage and interaction procedures of each process or thread that represents various functions in system. Moreover, some simulation functions, for instance, saving and loading initial conditions, involve the management for various data types. Therefore, overall management for internal and external data is indispensable. Refer to above requirements, this paper proposes a set of data management methods that include interface design and process optimization for data interaction in I&C system. The method has been proven to be beneficial for data management in physical system, and it should have significant engineering application value in I&C system. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Information management

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EMC PERFORMANCE OF DIFFERENT CABLE ROUTING STYLES FOR I&C systemS OF NPPS 29

EMC PERFORMANCE OF DIFFERENT CABLE ROUTING STYLES FOR I&C SY...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Yang, Youwei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886373

In the nuclear power plants, cables are signal transport links between separated signal processes that must be allowed to communicate with one another in a specified manner. Nowadays, for the instrumentation and control system of nuclear power reactor, there have always two ways to place the cables, i.e. above the cabinet or underneath the cabinet. One of the two ways is often preferred on account of the local installation environment or the human factor, while the electromagnetic compatibility (EMC) performance of these two ways is seldom reported. The pitfall may compromise the electromagnetic compatibility performance of the cabinet, or the system or installation it is used in, even when all the components used in the cabinet have excellent EMC compliance individually. In this paper, based on the numerical simulation technology, numerical models were built for the instrumentation and control system with the two different cable routing styles. Radiated emission, radiated susceptibility and crosstalk of cables connected to the cabinets are mainly researched. Results show that the different cable routing styles mainly affect the radiated susceptibility performance and the way that underneath the cabinet is preferred. Copyright © 2022 by ASME.

关键词： Electromagnetic compatibility

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design of Emulation system for Safety DCS 5th

Design of Emulation System for Safety DCS

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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： 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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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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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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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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Effects of rolling on characteristics of system under forced circulation and natural circulation

Effects of rolling on characteristics of system under forced...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Cai, Rong Yue, Nina Fang, Hongyu Chen, Baowen Liu, Lili Ma, Zehua Science and Technology on Reactor System Design Technology Laboratory Chengdu China Cnnc Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology Chengdu China Xi'an Jiaotong University Xi'an Shaanxi China

ISBN: (纸本)9780791883785

The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermalhydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Nuclear fuels

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