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检索条件"机构=Science and Technology on Reactor System Design Technology"

共 2774 条记录，以下是11-20 订阅

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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...

引用

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...

引用

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-...

引用

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 ...

引用

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...

引用

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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Key technology of a New Potential Economical Desalination Method for Small Nuclear Power 23rd

Key Technology of a New Potential Economical Desalination Me...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Li, Siguang Gan, Yiran Li, Yi Liang, Tiebo Science and Technology on Reactor System Design Technology Laboratory Sichuan Chengdu China

ISBN: (纸本)9789811988981

With the rapid development of distributed energy, small nuclear power has been widely concerned by researchers because of its inherent advantages. Keeping circulating water in a low salinity state is an important factor in economically maintaining the system operation. Therefore, it is necessary to investigate a specialized desalination method for small nuclear power. In this paper, expert knowledge and comparative analysis were introduced to study the advantages and disadvantages of some traditional desalination methods according to the recent requirements for small nuclear power. In addition, a potential economical namely hydrate-based desalination method was discussed, the key technologies that are urgent to be solved for this method were summarized and one of these technologies, phase equilibria of gas hydrate prediction, was studied using a thermodynamic model. Results show that each traditional desalination method cannot well meet the requirements for economy, energy and environment. Referring to the national requirements for carbon emission, the hydrate-based method is much suitable for small nuclear power due to its significant advantages in the fields of economy, energy and environment. In addition, the proposed thermodynamic model shows good performance compared with the published hydrate equilibrium data. Furthermore, the kinetics of hydrate formation and decomposition is considered another key to promoting the development of hydrate-based method. Finally, we hope the solution to the above problem can provide the idea of designing small nuclear power in the future. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Desalination

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INVESTIGATION ON THE APPLICATION OF MRF DAMPER BASED ON-OFF SEMI-ACTIVE VIBRATION CONTROL ON THE reactor COOLANT PUMP 30

INVESTIGATION ON THE APPLICATION OF MRF DAMPER BASED ON-OFF ...

引用

30th International Congress on Sound and Vibration, ICSV 2024

作者： Liao, Guojiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789090390581

The seismic design of the reactor coolant pump is important for the nuclear power plant. In this study, the magnetorheological fluid damper (MRF damper) is introduced to the supporting structure of reactor coolant pump, which is referred to HPR1000. MRF damper is a kind of intelligent damper, whose damping force can be controlled by the external magnetic current within several milliseconds according to the external excitation environment. In this study, considering both the controllable damping property of MRF damper and the complexity of the control strategy, the relative velocity/displacement based ON-OFF control algorithm is proposed to control the damping force of MRF damper. Numerical simulation demonstrates that the absolute acceleration and the relative displacement of the reactor coolant pump under seismic excitation are suppressed greatly with the help of MRF damper. The MRF damper is efficient in improving the seismic safety of the nuclear grade reactor coolant pump. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Damping

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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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Research on Hybrid Modeling Method Based on Mechanism Model and BP Neural Network 4

Research on Hybrid Modeling Method Based on Mechanism Model ...

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4th Asia Conference on Information Engineering, ACIE 2024

作者： Chen, Jie Zheng, Yanqiu Chu, Qing Jiang, Guanfu Li, Jing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9798350371925

reactor system modeling is the premise of reactor control system design. The common commercial thermal and hydraulic analysis programs like RELAP5, RETRAN have complex structure and slow operational speed, which are not conductive to the transplantation and improvement of reactor system. In order to realize the rapid development of reactor control system, this paper carries out the mechanism modeling of the reactor system in Matlab/Simulink, and then introduce BP neural network to establish the hybrid model, which improves the modeling accuracy to support the design of reactor system. For small pressurized water reactor, we firstly take analytical modeling method to build the core mechanism model and its state space model through three conservation principles and the operating mechanism of core. Based on analyses, we can conclude that the parameter error and structural error of the mechanism model is caused by the assumption and simplification during modeling, and choose the parameter with the most influence (the feedback coefficient of the coolant temperature) as the corrected parameter in the BP neural-based hybrid modeling process. After that, according to the dual constant operation scheme of small pressurized water reactors, and the gradient descent method of BP network, the hybrid model of core is built and the parameter error and structure error of the mechanism modeling are corrected online. The simulation results show that the proposed hybrid modeling method can accurately simulate the reactor operate state. Thus, the proposed hybrid model is accurate and effective. © 2024 IEEE.

关键词： Pressurized water reactors

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