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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...

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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 ...

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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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ANALYSIS OF SMR reactor COOLANT system IN APROS 31

ANALYSIS OF SMR REACTOR COOLANT SYSTEM IN APROS

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2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Zhu, Ye Xingbo, Wang Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9780791888247

The reactor coolant system (RCS) is the primary system of the nuclear power plant primary cycle. Some small module reactors (SMR) applied once-through steam generator and integrated reactor vessel. The characteristic of these reactors differs widely from the tradition nuclear power plant. APROS Nuclear simulation and analysis software, with the Comprehensive plant model covering reactor island, turbine island, electrical and automation systems, has been successfully used in a series of major nuclear power plant projects, including power upgrade, modernization, safety improvement, and new plant projects. A reactor coolant system model of SMR that contains both process and control elements is developed and simulated using APROS. Steady states analysis is conducted under different power load and both force/natural cycle. Analysis of Transient condition including increasing/decreasing power load and force/natural circulation switching is also conducted to acquire dynamic thermal hydrate parameter. Reference and guidance can be provided for system optimization and safety analysis. Copyright © 2024 by ASME.

关键词： Nuclear power plants

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RESEARCH ON THE OPTIMIZATION OF THE reactor COOLANT system STRUCTURE 30

RESEARCH ON THE OPTIMIZATION OF THE REACTOR COOLANT SYSTEM S...

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30th International Congress on Sound and Vibration, ICSV 2024

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

ISBN: (纸本)9789090390581

The dynamic analysis of reactor coolant system is the key step of mechanics evaluation, and with the continuous improvement of seismic design benchmarks and system design requirements, it is urgent to carry out the research on seismic capacity optimization of reactor coolant system. The reactor coolant system has the character that the design parameters are numerous and the structure is nonlinear, and when the structure is optimized, it is difficult to determine the key parameters which affect the dynamic response of the reactor coolant system prominently, which is also the difficulty in the traditional optimization according to the designer’s experience. In this paper, the sensitivity analysis of reactor coolant system dynamic response is carried out based on the OPTIMUS integrated platform first, and the key design parameters affecting the dynamic response of reactor coolant system are identified;then the optimization model for dynamic response of reactor coolant system is established according to the sensitivity analysis results, taking the support stiffness and support gap as the design variables,taking the nozzle loads of the main equipment as the objective function;finally, the sequential quadratic programming method is adopted to solve the optimization problem. The optimization analysis results show that the nozzle loads of the main equipment can be effectively decreased when appropriate design parameters are chosen. Through the research, a set of optimization analysis methods and processes applicable to the reactor coolant system and generalizable to other systems and equipment in the reactor were established,which can provide data support for the improvement of reactor system. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Sensitivity analysis

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A numerical method for predicting bursting strength of composite rocket motor case considering filament winding process-induced stress

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Chinese Journal of Aeronautics 2025年第2期38卷 505-518页

作者： Jiqiu LIANG Fei ZHAO Dan WU Anxin DING Hubei Key Laboratory of Advanced Aerospace Propulsion Technology Wuhan 430040China System Design Institute of Hubei Aerospace Technology Academy Wuhan 430040China School of Materials Science and Engineering Wuhan University of TechnologyWuhan 430070China State Key Laboratory of Advanced Technology for Materials Synthesis and Processing Wuhan University of TechnologyWuhan 430070China

A numerical method to predict the bursting strength of filament wound composite rocket motor case is proposed *** method can evaluate the longitudinal stress evolution of each composite layer as impregnated filaments with fiber tension are wound layer by layer,and consider the effects of accumulated stress and deformation during filament winding on the bursting strength of composite ***∅520 mm composite cases as a case study,the filament-winding-process-induced stress and deformation as well as progressive damage behavior are numerically predicted,followed by a comparison with experimental *** numerical results show that the predicted bursting pressures for composite cases manufactured on the mandrels with and without a flexible component are 14.20 MPa and 21.40 MPa,*** values exhibit slight deviation from the measured pressures of 13.50 MPa and 21.57 ***,the predicted damage locations,situated respectively in the dome and cylinder,agree well with the experimental *** observations indicate that use of flexible component reduces the load-bearing capacity of the ***,it validates the reliability and accuracy of the proposed numerical method in predicting the bursting strength of composite cases.

关键词： Composite case Fiber tension Bursting strengt hProgressive damage analysis Filament winding

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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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Heat dissipation of carbon shell in ZrC-SiC/TaC coating to improve protective ability against ultrahigh temperature ablation

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Journal of Advanced Ceramics 2024年第7期13卷 1080-1091页

作者： Yuyu Zhang Xuemeng Zhang Hongkang Ou Bozhe Wang Jia Sun Qiangang Fu Science and Technology on Thermostructural Composite Materials Laboratory Shaanxi Key Laboratory of Fiber Reinforced Light-Weight CompositesNorthwestern Polytechnical UniversityXi’an 710072China System Design Institute of Hubei Aerospace Technology Academy Wuhan 430040China

To efficiently decrease ablation heat accumulation and improve the ability of ZrC-SiC/TaC coatings to protect carbon/carbon(C/C)composites,a thermally conductive nanonetwork with a ceramic@carbon core-shell structure was designed and ***-derived SiC/TaC with a graphene carbon shell was synthesized and introduced into a ZrC coating by supersonic atmospheric plasma spraying(SAPS).Graphene shell paths increased the heat transfer capability by lowering the surface temperature to approximately 200℃during oxyacetylene *** heat dissipation of the graphene shell in the ZrC-SiC/TaC@C coating reduced the volatilization of low-melting-point phases and delayed the sintering of ZrO_(2) ***,the graphene shell in ZrC-SiC/TaC@C coating decreased the mass and linear ablation rates by 91.4%and 93.7%compared to ZrC-SiC/TaC coating,*** work provided a constructive idea for improving the ablation resistance of the coatings by incorporating carbon nanomaterials as a function of heat dissipation.

关键词： carbon/carbon(C/C)composites carbon nanomaterials core-shell structure polymer-derived ceramic(PDC) ablation resistance

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