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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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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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design and Implementation of DCS design List Automatic Generation and Check system Based on labVIEW 7th

Design and Implementation of DCS Design List Automatic Gener...

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

作者： Deng, Xiao-Jun Zhang, Xu Feng, Shi-Man Peng, Hao Yao, Ying-Fan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789819934546

The DCS configuration design process involves a large number of variables and multiple lists, and designers need to spend more time on the preparation, checking and maintenance of the lists. Manual compilation of files is a lot of work, and prone to human error. In this process, the use of automatic software to assist in the design can effectively improve efficiency. Based on the practical requirements of safety-class DCS design project of a nuclear power plant unit 1 and 2, this paper designs and implements a set of automatic generating and checking system of accounting list, which implements the automatic generation and typesetting of list files, and obtains good application effect. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast reactor 23rd

Atomic Scale Simulation on Liquid Metal Embrittlement Induce...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

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

ISBN: (纸本)9789811988981

Lead-Cooled fast reactor, which uses liquid lead or lead-bismuth alloy as coolants, is one of the six main reactors of the fourth generation reactor system. Most recently, the Lead-Cooled fast reactor received a lot of attention due to its safety, economy, and sustainability advantages. However, the liquid metal embrittlement (LME) could be a potential threat to the reliability of structure materials used in the Lead-Cooled fast reactor. Several experimental results show that the segregation of alloy element in grain boundary has a close association with liquid metal embrittlement. In this paper, by applying atomic simulation techniques, including molecular dynamics and Monte Carlo method, the segregation of lead element on poly-crystal ferritic and the resulting mechanical properties variations are simulated with the newest interatomic potential. The simulation results show that the lead element segregates on grain boundary. Moreover, such segregation of lead on grain boundary induced hardening and embrittlement of poly-crystal systems in the following uniaxial tensile testes. The atomic-scale mechanisms for such mechanical properties variation induced by lead element segregation are discussed carefully. In conclusion, these results verify that the segregation of lead element could be one of the critical mechanisms for LME, which well agree with previous experimental evidences. The present study may allow us to understand the mechanism of LME better. It can serve as a good reference for selecting more suitable materials for the Lead-Cooled fast reactor. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Grain boundaries

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A DEEP LEARNING-BASED APPROACH FOR PREDICTING THE STOCHASTIC PROCESS OF reactor ACCIDENTS 30

A DEEP LEARNING-BASED APPROACH FOR PREDICTING THE STOCHASTIC...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Although the prediction of system behavior plays a crucial role in accident management, there is very limited research on this subject in the nuclear industry. Accident process prediction methods are usually implemented as data-based methods due to the fast inference speed. However, the existing methods used for reactor accidents often ignore the predictive capability of the forecasting models over long-term patterns, or do not analyze the confidence of the prediction. In this paper, we have proposed a method for accident process prediction based on a Temporal Fusion Transformer (TFT) model. On the one hand, the method leverages multiple types of covariates as auxiliary, such as static labels of accidents and other monitoring data, to improve the prediction accuracy;on the other hand, it carries out uncertainty assessment of the predicted sequences. The method proposed in this paper is applied to MBLOCA post-accident prediction of HPR1000. Extensive results show that the method significantly outperforms the NiHiTS, Nbeats, Transformer, LSTM, GRU and RNN etc. models in terms of prediction accuracy on the test dataset. Furthermore, the uncertainty analysis of the prediction results allows the method to be applied to production scenarios of nuclear reactors, which is of high safety demands. © 2023 by JSME.

关键词： Forecasting

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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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