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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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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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Computational homogenization of thermomechanical properties for Fully Ceramic Microencapsulated fuel with 3D printed SiC matrix

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Annals of Nuclear Energy 2025年 222卷

作者： Chen, Yiwen Lu, Qi Liu, Changwen Li, Wei Zhang, Dalin Su, G.H. State Key Laboratory of Multiphase Flow in Power Engineering Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an710049 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu61213 China State Key Laboratory for Strength and Vibration of Mechanical Structures Xi'an Jiaotong University Xi'an710049 China

Fully Ceramic Microencapsulated (FCM) fuel, fabricated using 3D printing technology, incorporates traditional Tristructural-Isotropic (TRISO) fuel particles embedded within a 3D-printed SiC matrix. This fuel type shows great potential for compact, high-temperature nuclear reactors. However, comprehensive performance analysis remains challenging due to the multilayered structure, complex material properties, and the irradiation behavior of numerous TRISO particles within the matrix. To address these challenges, this study aims to apply finite element (FE)-based computational homogenization to evaluate the fuel's thermomechanical properties. First, a detailed finite element analysis (FEA) of a single TRISO particle's performance was conducted, focusing primarily on the Buffer-IPyC gap size change during irradiation, a critical factor influencing the fuel's thermal performance. Second, the homogenization of a single TRISO particle was performed using FEA, accounting for the changes in the Buffer-IPyC gap size. Finally, homogenized TRISO particles were randomly distributed and perfectly bonded within a 3D-printed SiC matrix to form a Representative Volume Element (RVE). The FE homogenization of the RVE was then conducted to derive the effective thermomechanical properties of the FCM fuel. The packing fraction of TRISO particles in the matrix ranged from 10 % to 50 %, with temperature and burnup conditions spanning 800–1600 K and 0–16 % FIMA, respectively. Results show that irradiation significantly affects the effective properties of the fuel, though this impact diminishes as burnup increases. Additionally, the effective thermal conductivity of the FCM fuel decreases with increasing TRISO packing fraction, assuming the thermal conductivity of the SiC matrix exceeds that of TRISO particles. This study provides a valuable reference for the design and optimization of FCM fuel for future nuclear applications. © 2025 Elsevier Ltd

关键词： Thermal conductivity of solids

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Error analysis and control of the acoustic overlapping mesh technique

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Applied Acoustics 2025年 238卷

作者： Wang, Youjiang Yang, Zhuangtao Liu, Zhenghao Ding, Yongle Zhang, Tao Zhang, Daiyu Yu, Long Key Laboratory of Marine Intelligent Equipment and System Ministry of Education Shanghai200240 China School of Ocean & Civil Engineering Shanghai Jiao Tong University Shanghai200240 China Kunming Branch of the 705 Research Institute China State Shipbuilding Corporation Limited Kunming China School of Marine Science and Technology Northwestern Polytechnical University Shaanxi Xi'an China Marine Design and Research Institute of China Shanghai China Xi'an Precision Machinery Research Institute Xi'an China School of Naval Architecture & Ocean Engineering Jiangsu University of Science and Technology Zhenjiang212003 China

Accurate and affordable methods for predicting the underwater radiated noise of ships will facilitate the development of effective noise control techniques. The acoustic overlapping mesh (AOM) technique, which enables the direct integration of volumetric noise terms in the Lighthill and FWH (Ffowcs Williams and Hawking) acoustic analogies, is a promising way to obtain such a method. In this paper, we improved the AOM technique by extending the overlapping mesh idea to the surface noise terms, further reducing the acoustic computational burden. Meanwhile, three causes for numerical errors of the AOM technique are identified and analyzed theoretically and numerically, leading to preliminary guidelines for choosing the acoustic mesh dimensions. The results suggest that keeping the cell-size-to-wavelength ratio (h/λ) lower than 0.05 avoids a sharp increase of the error from Cause-I, and keeping h/λ lower than 0.005 helps maintain this error less than 1 dB. The errors from the other causes (Cause-II and III) can be maintained less than 1 dB by keeping the distance-to-cell-size ratio (r/h) larger than 7.97. The simulation case also shows that the AOM technique reduces the acoustic computational time by more than 99% and 90% for the total and surface noise, respectively, while maintaining a noise prediction difference of around 1 dB. Using a non-uniform acoustic mesh instead of the uniform one helps achieve a much better with a similar acoustic computational time. © 2025 Elsevier Ltd

关键词： Numerical methods

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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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