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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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Transient fuel performance analysis of UO_(2)–BeO fuel with composite SiC coated with Cr cladding based on multiphysics method

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Nuclear science and Techniques 2023年第12期34卷 93-108页

作者： Chun‑Yu Yin Shi‑Xin Gao Sheng‑Yu Liu Rong Liu Guang‑Hui Su Li‑Bo Qian School of Nuclear Science and Technology Xi’an Jiaotong UniversityXi’an 710049China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China School of Electric Power South China University of TechnologyGuangzhou 510640China

The transient multiphysics models were updated in CAMPUS to evaluate the accident-tolerant fuel performance under accident *** is a fuel performance code developed based on *** simulated results of the UO_(2)–Zircaloy fuel performance under accident conditions were compared with those of the FRAPTRAN code and the experimental data to verify the correctness of the updated ***,multiphysics models of the UO_(2)–BeO fuel and composite SiC coated with Cr(SiC_(f)/SiC-Cr)cladding were implemented in ***,the fuel performance of the three types of fuel cladding systems under Loss of Coolant Accident(LOCA)and Reactivity Insertion Accident(RIA)conditions was evaluated and compared,including the temperature distribution,stress distribution,pressure evolution,and cladding failure *** results showed that the fuel temperature of the UO_(2) fuel under accident conditions without pre-irradiation was lower after being combined with SiC_(f)/SiC-Cr ***,the centerline and outer surface temperatures of the UO_(2)–BeO fuel combined with SiC_(f)/SiC-Cr cladding reduced further under accident *** cladding temperature increased after the combination with the SiC_(f)/SiC-Cr cladding under accident conditions with *** addition,the use of SiC_(f)/SiC-Cr cladding significantly reduced the cladding hoop strain and plenum pressure.

关键词： Accident condition Fuel performance UO2–BeO fuel SiCf/SiC-Cr cladding

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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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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

ISBN: (纸本)9789090390581

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 experimental results of the vibration level drop of the isolators. Due to the standing wave effect and the limitations of analytical models, it is difficult to simulate the vibration level drop of isolators well solely based on dynamic stiffness or damping. In this paper, an analysis model for the vibration reduction effect of metal-rubber (MR) vibration isolators based on impedance test data is established. The origin impedance and cross point impedance data is tested on the testing platform. The four-terminal parameter method is applied to calculate the stiffness and damping value of the MR vibration isolators. The calculation results show that compared with commonly used numerical simulation methods in the past, the model based on frequency dependent stiffness and damping can further reduce the error between the numerical simulation and the experimental results. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Damping

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NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHARACTERISTICS IN LBE-COOLED HELICAL CRUCIFORM FUEL AND WIRE-WRAPPED FUEL 31

NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHA...

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2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Qi, Zhang Haoyu, Wang Cao, Junxian Yuanming, Li Chenxi, Li Zhenhai, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791888261

The lead bismuth eutectic (LBE) cooled fast reactor in one of the General 4 advanced reactors, which plays significant role in the nuclear fuel cycle. In the recent years, the wire-wrapped fuel with four wires surrounds the cylindrical fuel is developed and promoted, which may be accept by the LBE-cooled fast reactor. The helical cruciform fuel of the Lightbridge corporation has similar geometrical features with the wire-wrapped fuel. The cross-section of helical cruciform fuel consists of the convex blades and the concave elbows, and the cross-section is rotated around the axis. The blade and elbow of helical cruciform fuel function as the wires while they are part of the nuclear fuel. In the present work, the comparison on the thermal-hydraulic and flow-induced vibration (FIV) characteristics of the helical cruciform fuel and wire-wrapped fuel is carried out by numerical simulation. The fluid-structure interaction is solved by the coupling between the fluid and solid domains, and the transient turbulence fluctuation is solved by large eddy simulation (LES). The results show that the convective heat transfer is enhanced by the spiral structures even though the hot spot phenomenon is caused simultaneously. The hot spot temperature of the helical cruciform fuel is larger than the wire-wrapped fuel duo to the non-uniform distributed heat flux. As the coolant in the flow channel is blocked by the blades and wires, the directional transverse flow is generated and thus the transverse mixing is enhanced. The one-way mixing with opposite direction occurs in the two sides of the single rod. There are similarities in the FIV characteristics for the spiral shaped fuels. The maximum vibration displacement and dominant vibration frequency of the helical cruciform fuel and wire-wrapped fuel are almost the same. © 2024 by ASME.

关键词： Fluid structure interaction

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STUDY ON DYNAMIC RESPONSE OF reactor CORE BARREL UNDER PUMP-INDUCED PULSATING PRESSURE 30

STUDY ON DYNAMIC RESPONSE OF REACTOR CORE BARREL UNDER PUMP-...

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30th International Congress on Sound and Vibration, ICSV 2024

作者： Ye, Xianhui Cai, Fengchun Huang, Xuan Feng, Zhipeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789090390581

The flow-induced vibration (FIV) problem is widely present in nuclear power plants and has caused numerous engineering accidents. Therefore, it has been highly valued by the nuclear engineering community for a long time. The pump-induced pulsating pressure is an important excitation source of FIV. The characteristics of pump-induced pulsating pressure are periodic, it may cause resonance and have the adverse impact on the fatigue life of reactor coolant system. In order to study the vibration of reactor core barrel (CB) of newly designed HPR1000 (Hua-long Pressurized reactor) under pump-induced pulsating pressure, the finite element model of CB is established and the dynamic response characteristics are analyzed. Combined with turbulent excitation, the influence of CB under flow induced load was comprehensively evaluated .The stress analysis shows that the maximum stress of FIV at each position of CB is less than the fatigue stress limit, and the structure is safe. However, for the newly designed reactor or the replacement of the reactor coolant pump, the pump-induced vibration of the CB and reactor internal should be paid enough attention to. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Stress analysis

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Research on The FPGA Reconfiguration technology Based on RS Code for Nuclear Applications 15

Research on The FPGA Reconfiguration Technology Based on RS ...

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15th IEEE Global Reliability and Prognostics and Health Management Conference, PHM-Beijing 2024

作者： Chen, Qi Wang, Fanyu He, Jinyu Zhao, Yang Wang, Heng Huang, Qichang Lei, Minjie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (数字)9798350354010

ISBN: (纸本)9798350354010

Field Programmable Gate Array (FPGA) is now widely used in nuclear safety Instrumentation and Control (I&C) systems and operates core features like communication, signal capturing and processing, peripheral controlling etc. Those features are crucial to the entire system, which means the state of the FPGA is vital to the system as well. In nuclear I&C industry, single event effect (SEE) is a common hazard faced by the nuclear facilities which are exposed to the high radiation environment, because it may cause register upsets on static random-access memory (SRAM) devices and interfere the well-functioning of them. One popular method to solve this issue is fault diagnose, while this method only finds out the errors, not corrects them. To improve the solution, forward error correction technology, which is widely used in communication industry, is proposed to combine with the FPGA reconfiguration to check and rectify the upsets. Reed-Solomon (RS) Code is chosen to reinforce the core configuration data and to correct the faults by reconfiguration when diagnose problems occurs. Experiment was taken on the demo board to verify the viability of this strategy. The results indicate that this strategy can improve the FPGA against single event upset (SEU), thereby increase the system reliability in this aspect. © 2024 IEEE.

关键词： Forward error correction

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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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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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RESEARCH OF MODEL FOR LOCA NONLINEAR DYNAMIC ANALYSIS OF reactor COOLANT SYSTEM 30

RESEARCH OF MODEL FOR LOCA NONLINEAR DYNAMIC ANALYSIS OF REA...

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30th International Congress on Sound and Vibration, ICSV 2024

作者： Yanli, Yuan Xianhui, Ye Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789090390581

The Loss of Coolant Accident (LOCA) is one of the most serious accidents in nuclear power plants. According to the requirements of the United States Nuclear Regulatory Commission (USNRC) SPR, the dynamic response analysis of LOCA transients occurring in the reactor coolant system (RCS) must be performed to verify the design of equipment, supports, anti-whip devices and civil structures[1]. At present, the LOCA dynamic response of reactor coolant systems is mainly analysed using a single loop model, and there are few relevant literatures on LOCA analysis with the complete loop model. The influence on the destructive loop due to the hydraulic load of the unbroken loop cannot be considered when using a single loop model for LOCA dynamic response analysis, which may lead to unreasonable analysis results. In this paper, the two models for nonlinear LOCA dynamic analysis of reactor coolant system are established: one is single loop, and the other is three-loop. The nonlinear time-history dynamic analysis of LOCA is analysed by ANSYS software, and the analysis results from the two models are compared. The results show that the loads from the three-loop LOCA dynamic analysis model are consistent with those of the single loop LOCA analysis. By comparing the results calculated by single loop LOCA analysis model with different boundary conditions at inlet nozzles, it is found that the expansion displacements due to the internal pressure at the inlet nozzles of the reactor pressure vessel cannot be ignored during the LOCA dynamic response analysis. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Nuclear power plants

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