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

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Preliminary research on the thermal-mechanical coupling behavior simulation method of m3 fuel 28

Preliminary research on the thermal-mechanical coupling beha...

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2021 28th International Conference on Nuclear Engineering, ICONE 2021

作者： Changbing, Tang Yongjun, Jiao Yuanming, Li Kun, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Chengdu China

ISBN: (纸本)9780791885253

M3 fuel which microencapsulated TRISO particles in zirconium metal matrix is an integrated fuel designed to replace the traditional pellet-cladding fuel rod. The fuel design can effectively solve the problems of LWR fuel, such as water side corrosion, abrasion, pellet-cladding interaction, stress corrosion cracking and internal pressure bracing. Because of the multiple coating structure of TRISO particles and complex contacting behavior and deformation between TRISO particle and metal matrix, M3 fuel exhibits complex in-pile behavior. In order to establish the simulation method for the M3 fuel, the three-dimensional model was established by using ABAQUS software. The complex contacting behavior and deformation of coated layers and matrix for M3 fuel was considered. The results indicated that bonded strength between TRISO particle and Zr matrix has great effect on the stress distribution of SiC layer and matrix. The vulnerable area of Zr matrix located in the nearest position of TRISO particle to TRISO particle. The maximum tensile hoop stress of SiC layer was up to 150 MPa under strong bonding state, while SiC layer suffered compressive stress during all operation time for the weak bonding state. The failure probability of SiC layer was about 7 10-3 under strong bonding state which much higher than the weak ones. Thus, the integrity of SiC layer can be maintained by controlling the bonding strength between TRISO particle and Zr matrix during manufacturing process. © 2021 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Silicon carbide

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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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A new turbulent Prandtl number model for forced fully-developed pipe heat transfer of liquid metals 27

A new turbulent Prandtl number model for forced fully-develo...

引用

27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Peiying, Li Hongxing, Yu Sijia, Du Yaou, Shen Mingyu, Yan Hongsheng, Yuan Yu, Liu Xiaoli, Wu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

With the proposal of liquid metals as coolant of the reactors, the use of accurate heat transfer correlation and reliable turbulent prandtl number（Prt）model of these low prandtl number fluids in turbulent flow is essential. The heat transfer of Liquid metals differ from traditional coolants for their good molecular heat conductivity and this property leads to a separation of the spatial extension of thermal and viscous boundary layers. A constant turbulent prandtl number is proved to be inappropriate for complex conditions for it depends on the Reynolds and molecular Prandtl number. This paper proposes a new turbulent prandtl number model Prt=C+B/Pet, which can be used for accurately predicting the heat transfer for liquid metal flows. The constants C and B of the expression are obtained from the analytic solution of renormalization group procedure. B is further corrected with the physical properties for different liquid metals. The results are in a good agreement with experimental data over the range of experimentally accessible Prandtl numbers. The Nusselt numbers calculated using the proposed model with the Lyon integral are found to be well consistent with the experimental data for forced fully-developed pipe flows with constant heat flow boundary conditions. Computational fluid dynamics(CFD) analysis is also carried out for pipe geometries. With the proposed model for turbulent prandtl number, CFD codes using RANS approach give similar results for local Nusselt numbers measured in the experiments. Copyright © 2018-2019 by JSME

关键词： Liquid metals

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Research on the vibration reduction and sound radiation index properties of the cabin with internal equipment 26

Research on the vibration reduction and sound radiation inde...

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26th International Congress on Sound and Vibration, ICSV 2019

作者： Wenzheng, Zhang Bihao, Wang Xiaozhou, Jiang Bin, Lan Haiyang, Song Fengchun, Cai Xianhui, Ye Jiabin, Feng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781999181000

Suitable vibration and noise indices which provide technical direction for vibration and noise reduction, play a very important role in structure analysis and validated test. The contribution of foundation impedance is studied by analyzing the vibration reduction index of a single degree of freedom system. The finite element method (FEM) and boundary element method (BEM) are applied to simulate the vibration and noise indices of a cabin along with internal equipment. The vibration reduction and sound radiation indices of the internal isolator have been calculated to analyze the variation trend with frequency. The results indicate that the difference between vibration level difference and insert loss is very small in most frequency band. Interestingly, the power flow difference and radiated sound power are opposite in phase in middle and high frequency band, which have been proved to be appropriate elements to evaluate the vibration and noise indices of the cabin. The analysis method and procedure can be used to objectively evaluate the vibration reduction property of similar vibration isolators © Proceedings of the 26th International Congress on Sound and Vibration, ICSV 2019. All rights reserved.

关键词： Finite element method

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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...

引用

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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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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Experimental Research on Data Management Method of Instrument and Control Emulation system 1

引用

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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Investigation of turbulent prt model and the segmentation rule for lbe turbulent heat transfer

Investigation of turbulent prt model and the segmentation ru...

引用

2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Peiying, Li Jian, Deng Lei, Zhong Libo, Qian Rong, Cai Yugao, Ma Hongxing, Yu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

With liquid metal like lead-bismuth alloy(LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt=A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental results, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Liquid metals

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EMC PERFORMANCE OF DIFFERENT CABLE ROUTING STYLES FOR I&C systemS OF NPPS 29

EMC PERFORMANCE OF DIFFERENT CABLE ROUTING STYLES FOR I&C SY...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

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

ISBN: (纸本)9780791886373

In the nuclear power plants, cables are signal transport links between separated signal processes that must be allowed to communicate with one another in a specified manner. Nowadays, for the instrumentation and control system of nuclear power reactor, there have always two ways to place the cables, i.e. above the cabinet or underneath the cabinet. One of the two ways is often preferred on account of the local installation environment or the human factor, while the electromagnetic compatibility (EMC) performance of these two ways is seldom reported. The pitfall may compromise the electromagnetic compatibility performance of the cabinet, or the system or installation it is used in, even when all the components used in the cabinet have excellent EMC compliance individually. In this paper, based on the numerical simulation technology, numerical models were built for the instrumentation and control system with the two different cable routing styles. Radiated emission, radiated susceptibility and crosstalk of cables connected to the cabinets are mainly researched. Results show that the different cable routing styles mainly affect the radiated susceptibility performance and the way that underneath the cabinet is preferred. Copyright © 2022 by ASME.

关键词： Electromagnetic compatibility

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design of Emulation system for Safety DCS 5th

Design of Emulation System for Safety DCS

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5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Peng, Hao Zhang, Xu Deng, Zhi-Guang Chen, Qi Zhang, Yu Jiang, Wei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811634550

DCS (Digital Control system) is a very important part in safety protection in NPP (Nuclear Power Plant). Virtual DCS realizes the algorithm configuration of physical DCS running on general computers and gets rid of the limitation of hardware. In this paper, an emulation system (virtual DCS) is designed using virtual DPU technology. The virtual DCS can be used to verify the algorithm logic with the closed-loop system, which is composed by virtual DCS and digital reactor processing model together. The virtual DCS and controlled object are closed to test the control effects and in this way to verify the design of the NPP DCS. For the purpose of scientific research, training and design verification, virtual DCS should have malfunction simulation and other functions. These functions reflect the superiority of emulation technology. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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