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Determination of Hardware Logic Drawing Simulation Modeling Criteria and Its Application in Nuclear Safety-Class DCS 6th

Determination of Hardware Logic Drawing Simulation Modeling ...

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6th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2021

作者： Zhang, Xu Chen, Shi-Yong Yao, Zhang Deng, Zhi-Guang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811911804

In digital control system (DCS) design, independent hardware logic circuit design is usually used to deal with common causes failures of software. Due to the distinct difference between the design criteria of electrical design diagram and logic schematic diagram, the hardware logic verification should be carried out in the form of translation simulation. In this paper, the modeling criteria of DCS hardware logic simulation are summarized. The modeling criteria take into account both the standard engineering hardware logic drawing specifications and the characteristics of Simulink platform, and the DCS hardware logic electrical design information is translated to the Simulink platform for automatic redrawing according to specific criteria. A complete hardware logic simulation system was established based on the modeling criteria, and the application was verified in HPR 1000 Zhangzhou nuclear power plant project. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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A revised method to simulate the dynamic characteristic of vibration isolator of main pump 27

A revised method to simulate the dynamic characteristic of v...

引用

27th International Congress on Sound and Vibration, ICSV 2021

作者： Zhang, Wenzheng Xiong, Furui Liu, Shuai Yuan, Zhihao Jiang, Xiaozhou Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9788378807995

It's important to choose suitable kinetic parameters for vibration isolators to simulate the dynamic characteristic. The kinetic parameters have significant influence on the vibration reduction of vibration isolators. In this paper, the kinetic parameters of two kinds of metal-rubber (MR) vibration isolators of main pump are measured to accurately acquire the dynamic characteristic. 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. Unfortunately, the method can't simulation the influence of the peak value behind the standing wave frequency, which is caused by the internal resonance of the vibration isolators. A revised rod vibration model is adopted to simulate the internal resonance, because the rod vibration model can express the high order peak. The numerical results are in good agreement with experimental results. The revised method can be used to simulate the dynamic characteristic of vibration isolators. © "Advances in Acoustics, Noise and Vibration - 2021" Proceedings of the 27th International Congress on Sound and Vibration, ICSV 2021. All rights reserved.

关键词： Kinetic parameters

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DATA DRIVEN METHODS FOR BREAK SIZE AND LOCATION ESTIMATION IN LOCA BASED ON DEEP LEARNING 29

DATA DRIVEN METHODS FOR BREAK SIZE AND LOCATION ESTIMATION I...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Yuan, Peng Deng, Jian Qiu, Zhifang Zhu, Dahuan Huang, Tao Li, Zhongchun Du, Peng Xu, Qinglan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886427

Loss of coolant accident (LOCA) is one of the most important accidents in thermal hydraulic design of nuclear power plant. The traditional analysis method is difficult to realize the rapid prediction of the size and location of the break in the LOCA, while machine learning provides an idea for the rapid diagnosis of the initial cause of LOCA. In this paper HPR1000 which is independently designed by China was taken as the object, the process of LOCA under various break location and sizes is studied by using the advanced reactor system analysis code ARSAC (Advanced reactor system Analysis Code). The key thermal hydraulic parameters including temperature, pressure, flow rate and water level in the transient process of LOCA are analyzed, and a series of data sets are generated. Based on convolution neural network algorithm of deep learning, the transient thermal hydraulic parameters of HPR1000 under different working conditions are learned and the accident diagnosis model is obtained. By comparing with the test data, the trained model can quickly and accurately predict the size and location of LOCA, which can be subsequently extended to the diagnosis of other kinds of accidents in nuclear power plants and has certain application prospects. Copyright © 2022 by ASME.

关键词： Location

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COMPARISON OF NUMERICAL ANALYSIS METHODS OF COOLANT ENVIRONMENTAL FATIGUE BASED ON STRAIN RATE CONTROL 29

COMPARISON OF NUMERICAL ANALYSIS METHODS OF COOLANT ENVIRONM...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Shao, Xuejiao Xie, Hai Zhang, Yixiong Xiong, Furui Wang, Xinjun Shi, Kaikai Yu, Mingda Huang, Xuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886397

In this paper, the transformed strain rate of environmental fatigue correction factor (Fen), which is used to consider the influence of PWR primary coolant environment on material fatigue, is studied. EPRI's guideline and theAFCEN's probationary phase rule (RPP) No.3 of Volume VI in RCC-M 2017 edition provide two different calculation methods of transformed strain rate separately. NUREG CR-6909 gives two versions of prediction models of Fen for Low alloy steel, nickel base alloy and austenitic stainless steel which are the three main materials of the primary circuit system. In this paper, the model and transients of Sample 2 in EPRI's guideline are selected for calculation and comparison, so as to verify whether the analysis method is reasonable. Firstly, the fatigue analysis in air is carried out, and then based on the combination, the detailed calculation of strain rate is carried out to calculate environmental fatigue results. The analysis results are compared with the results of guideline, and the valuesare close. Then, the main pipeline safety injection nozzle components with small fatigue margin in PWR are selected to carry out environmental fatigue analysis. In the analysis, different methods of calculating modified strain rates are used, and different prediction models in CR-6909 are *** coefficient of Fenintegratedin RPP3 isconsidered, and the effects of various calculation methods on the results are compared. The results show that for austenitic stainless steel, the results of RCC-M RPP3 method are mostly less than those of EPRI guidelines. Compared with the prediction model of NUREG CR-6909 version 0, the prediction model version 1 greatly decreases the Fen results. Copyright © 2022 by ASME.

关键词： Strain rate

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Acoustic indeced vibration research of dryer in steam generator 26

Acoustic indeced vibration research of dryer in steam genera...

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26th International Congress on Sound and Vibration, ICSV 2019

作者： Xianhui, Ye Fengshou, Zhang Fengchun, Cai Wenzheng, Zhang Furui, Xiong Haiyang, Song Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781999181000

The acoustic induced vibration of dryer (also called secondary separator) in steam generator in nuclear power plants is a concern in recent years. Periodic vibrations of long-term operation can lead to fatigue damage of structural components, posing a safety hazard. In this paper, the finite element analysis model of the dryer structure in steam generator of "Hualong-1" nuclear power plant was established, and the acoustic induced vibration characteristics of the structure were studied. Firstly, the acoustic excitation of the dryer structure was discussed, and the excitation inputs were obtained, including the maximum static pressure and the maximum pressure power spectrum density. Then the finite element model was established by ANSYS and the weak parts of the dryer structure were determined by the vibration characteristics combined with the acoustic excitation characteristics analysis. Finally, the structural strength and fatigue characteristics of the dryer structure under two different inputs of the acoustic load were investigated. The analysis of vibration characteristics shows that the main weak parts of the dryer structure are the vertical partition plate and the perforated plate. The static pressure analysis shows that the stress intensity of the two weak plates is less than the limit. In the study of pressure power spectral density, the maximum dynamic stress intensity and alternating stress of the vertical partition plate and the perforated plate are also less than the limit. The research results indicate that the acoustic induced vibration response of the dryer structure is within the acceptable range © Proceedings of the 26th International Congress on Sound and Vibration, ICSV 2019. All rights reserved.

关键词： Steam generators

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A Novel Model for Dynamic Process Study of reactor Control Rod Drive Mechanism 26

A Novel Model for Dynamic Process Study of Reactor Control R...

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26th International Conference on Electrical Machines and systems, ICEMS 2023

作者： Yang, Yun Xu, Qiwei Zhang, Xuefeng Yu, Tianda Chen, Xinan Zhao, Yizhou Chongqing University National Key Laboratory of Power Transmission Equipment Technology Chongqing China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9798350317589

The magnetic jack type control rod drive mechanism (CRDM) is a critical electromagnetic actuator utilized in reactors, with its movable components subject to the combined effects of electromagnetic force, liquid resistance, spring force, and other factors, resulting in highly nonlinear and uncertain motion characteristics. This article establishes a novel model for the dynamic process analysis of the CRDM. The dynamic electromagnetic force and inductance of the CRDM are analyzed based on the equivalent magnetic network (EMN) model, taking into account mesh reconstruction and magnetic saturation. Through kinematic analysis and parameter coupling analysis, electromagnetic and kinematic parameters during the CRDM lifting process are obtained. Finally, finite element analysis (FEA) and experimentation are conducted to validate the accuracy of the proposed model. © 2023 IEEE.

关键词： Kinematics

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A method for predicting the configuration of corium pool in RPV lower plenum

A method for predicting the configuration of corium pool in ...

引用

2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018

作者： Liu, Lili Yu, Hongxing Chen, Liang Jian, Deng Chunrui, Deng Dan, Zhang Xiaoli, Wu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780894487552

The configuration of a corium pool in the lower plenum of reactor vessel is very important for validity of maintaining reactor vessel integrity under IVRERVC (In-Vessel corium Retention through External reactor Vessel Cooling) condition. A method is designed to predict the configuration of corium pool for an Advanced China PWR, call ACP1000 after entire melt dropping into the lower plenum. It takes into account both thermo-chemical reaction of oxides and metals in corium and the influence of different paths of corium relocated from core region into the lower plenum. The reasons why a three-layer pool has not been observed in the MASCA and COSMOS tests are given by this method. The method is applied to investigate the corium configuration of ACP1000 after a hypothetical station blackout (SBO) accident. It is shown that the stratification of the corium pool is related to the relocation paths of the corium. In the case of downward relocation, a two-layer melt pool with a metal layer on the top is formed. For the sideward relocation, the configuration shows a stratified pool consisting of a dense metal layer on the bottom, an oxide layer in the middle and a light metal layer on the top. © 2018 American Nuclear Society. All rights reserved.

关键词： Pressurized water reactors

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EVALUATION OF THE PARAMETER IMPORTANCE CHANGES OVER TIME FOR NUCLEAR reactor TRANSIENTS 29

EVALUATION OF THE PARAMETER IMPORTANCE CHANGES OVER TIME FOR...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Xiong, Qingwen Song, Gongle Du, Peng Huang, Tao Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886410

The uncertainty analysis and sensitivity analysis are two important parts in the safety analysis of nuclear reactor accidents, and the main purpose of the sensitivity analysis is to quantify the importance of the input on the target output. In typical sensitivity analysis for the nuclear reactor transients, specific techniques such as the regression analysis or correlation coefficient analysis are utilized to calculate the sensitivity measures of the inputs on the target output at scalar points, for instance the peak cladding temperature (PCT). However, the phenomena and parameters vary significantly over time during the transients, thus the influence of each input also changes. In order to evaluate the importance changes of the inputs on the target output over time, a time-dependent global sensitivity analysis technique was developed base on the B-spline function system of the functional data analysis, the landmark registration strategy, and an efficient moment-independent sensitivity analysis method. The developed technique was adopted in the sensitivity analysis of a large break loss of coolant accident (LBLOCA) of the nuclear reactor, and the time-dependent sensitivity measures of the inputs on the maximum cladding temperature were acquired and analyzed. Results showed that the proposed technique is capable of evaluating the importance changes of the key parameters in nuclear reactor accident transients. Copyright © 2022 by ASME.

关键词： Sensitivity analysis

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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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design and Implementation of Translation-Based Virtual DCS Based on Simulink 5th

Design and Implementation of Translation-Based Virtual DCS B...

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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： Zhang, Xu Deng, Zhi-Guang Ma, Quan Liu, Ming-Ming Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811634550

The translation-based virtual DCS based on Simulink can realize the function of virtualizing the physical DCS. This paper presents a method to virtualize physical DCS by using Simulink to realize the functions of running, freezing, acceleration and deceleration, setting and acquiring variable values, malfunction simulation, snapping and loading initial conditions, etc. Virtual DCS realizes the configuration translation by translating the configuration files of the physical DCS platform to Simulink platform, generating codes by using the configuration, and compiling the codes into algorithm files. The virtual DCS is combined with process model software to form a closed-loop, which simulates the level control system of the steam generator of a megawatt pressurized water reactor nuclear power plant for verification. Effective results have been achieved, which proves that the translation virtual DCS can meet the simulation effect of the simulator. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Simulink

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