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Research of Advanced Control Algorithm in Primary Loop Control system 5th

Research of Advanced Control Algorithm in Primary Loop Contr...

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

作者： Deng, Zhi-Guang Zhu, Bi-Wei Wu, Qian He, Peng Xiang, Mei-Qiong Xu, Tao Qing, Yue National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610213 China

ISBN: (纸本)9789811634550

The primary loop system of nuclear power plant is a large-scale control system with serious coupling between loops and complex objects. In order to solve the problem of integrated control of primary loop core power, stabilizer pressure and water level, and steam generator water level, this paper applies improved predictive control algorithms to each sub-control system on the basis of mechanism modeling. For the strong non-linear problem of water level control, a multi-model predictive controller is constructed through linearization at each typical operating point. The simulation results show that the application of advanced predictive control algorithm to the primary loop control system has achieved good control performance, and it also lays a solid foundation for subsequent practical engineering applications. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Water levels

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Assessment of Coupled Effect of Steam Pipes on Seismic Analysis of reactor Coolant system 7

Assessment of Coupled Effect of Steam Pipes on Seismic Analy...

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7th International Conference on Environmental science and Civil Engineering, ESCE 2021

作者： Li, Lijuan Liu, Zhenyu Li, Hui Ye, Xianhui Ai, Honglei National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu Sichuan China

reactor coolant system is connected with auxiliary pipes, surge pipes and steam pipes. In order to facilitate the dynamic analysis of reactor coolant system, it is necessary to decouple the pipes connected with the reactor coolant system. Although some international standards stipulating requirements for the decoupling of subsystems from the main system, in order to further explore the influence of decoupling steam pipes, the seismic response of reactor coolant system with or without steam pipes is simulated by using ANSYS program. Firstly, the modal analysis of reactor coolant system is carried out to analysis the influence of steam pipes on the resonant frequency of reactor system. Then, the standardized seismic dynamic analysis of reactor system with or without steam pipes is carried out. The nozzle load (inlet and outlet nozzle load of main device, such as reactor pressure vessel, steam generator, pump) and support load (of main device such as reactor pressure vessel, steam generator, pump) are output and compared. According to nozzle load and support load, it can be seen that the steam pipes have little influence on the seismic analysis of reactor coolant system, i.e. steam pipes can be decoupled in the seismic analysis of reactor coolant system. © Published under licence by IOP Publishing Ltd.

关键词： Steam generators

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Preliminary research on the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel

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International Journal of Advanced Nuclear reactor design and technology 2019年 1卷 51-56页

作者： Changbing, Tang Yongjun, Jiao Yuanming, Li Yi, Zhou Hua, Pang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China No. 328 Section 1 Changshun Avenue ChengduSichuan610041 China

FCM fuel (fully ceramic micro-encapsulated fuel) which is a promising ATF (accident tolerant fuel) candidate fuel. Establishing the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel accurately, would be very helpful for the optimal design of FCM fuel. In this research, TRSIO particles were assumed to be uniformly distributed in the silicon carbide (SiC) matrix, and the corresponding representative volume element (RVE) models were established. Based on ABAQUS commercial software, the thermal effects and irradiation effects of FCM fuel were all introduced into this numerical simulation by coding the UMAT subroutine, and then the irradiation - thermal - mechanical coupling behavior simulation method of FCM fuel was preliminary established. Compared with the simulation result of BISON software in open literature, the simulation results of this numerical simulation method are basically consistent with those of BISON software, the effectiveness of this numerical simulation method was preliminary verified. © 2019 Xi'an Jiaotong University

关键词： Irradiation

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Multi-objective Optimization of Non-uniform Beam for Minimum Weight and Sound Radiation

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Transactions of Tianjin University 2017年第4期23卷 380-393页

作者： Furui Xiong Mengxin He Yousef Naranjani Qian Ding Jianqiao Sun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China School of Mechanical Engineering Tianjin University School of Engineering University of California Merced

A multi-objective optimization of non-uniform beams is presented for minimum radiated sound power and weight. The transfer matrix method is used to compute the structural and acoustic responses of a non-uniform beam accurately and efficiently. The multi-objective particle swarm optimization technique is applied to search the Pareto optimal solutions that represent various compromises between weight and sound radiation. Several constraints are imposed, which substantially reduce the volume fraction of feasible solutions in the design space. Two non-uniform beams with different boundary conditions are studied to demonstrate the multi-objective optimal designs of the structure. © 2017, Tianjin University and Springer-Verlag Berlin Heidelberg.

关键词： Acoustic generators Acoustic properties Acoustic wave propagation Acoustic wave scattering Optimal systems Pareto principle Particle swarm optimization (PSO) Transfer matrix method

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Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water

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Chinese Journal of Mechanical Engineering 2023年第4期36卷 326-338页

作者： Jun Wang Haojie Li Zhengyang Li Yujie Lei Quanyao Ren Yongjun Jiao Zhenbing Cai Key Lab of Advanced Technologies of Materials Tribology Research InstituteSouthwest Jiaotong UniversityChengdu 610031China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly ***,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR *** can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal *** fretting wear behavior of Zr-4 alloy under different temperature environments was *** addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light *** show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.

关键词： Fretting wear Cladding High temperature and high pressure Zirconium alloy

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Study on flow modes of open natural circulation with low height difference under low power conditions

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International Journal of Advanced Nuclear reactor design and technology 2020年 2卷 10-14页

作者： Yi, Li Jianchuang, Sun Hang, Peng Cheng, Xiang Quan, Biao Cao, Xiaxin Jian, Zhou Ming, Ding Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin150001 China

Open natural circulation systems are widely applied in the field of energy and chemical industries. Under some special conditions, open natural circulation systems with low height difference need to be applied, but research on this type of systems is not sufficient. In this paper, the flow modes of sucha natural circulation system were experimentally studied at low power level by adopting steam heating. In addition, the flow modes and physical phenomena of the natural circulation system were analyzed in detail under different heating power conditions. It is found that there are four flow modes at different power levels in the natural circulation system due to the influence of non-condensable gas and subcooled boiling. The inlet and outlet temperatures of heating section are analyzed in different flow modes. The result shows that the outlet temperature of heating section can be used as a judgment criterion for different flow modes. © 2019

关键词： Chemical industry

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Research Progress of Metal-Based Shielding Materials for Neutron and Gamma Rays

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Acta Metallurgica Sinica(English Letters) 2021年第12期34卷 1609-1617页

作者： Bin Gan Shichang Liu Zhen He Fucai Chen Haoxuan Niu Jingchang Cheng Bo Tan Bo Yu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu610213China State Key Laboratory of Light Alloy Casting Technology for High-End Equipment Shenyang Research Institute of Foundry Co.Ltd.Shenyang110022China

The radiation generated by nuclear reaction is harmful to human body and equipment,thus the radiation shielding materials that employ the shielding ability from neutron and gamma rays are the best candidates according to application situations and radiation *** this paper,the researches of metal-based neutron and gamma rays or multiple purpose shielding materials are systematically summarized,and the respective and principal problems of these materials with respect to shielding effectiveness and other performances,such as corrosion,mechanical properties,manufacture,etc.,are ***,the prospect of shielding materials is outlined,which suggests that the development of highly efficient and multiply functional radiation shielding materials with good environmental compatibility is one of the future development trends.

关键词： Radiation shielding materials Metal-based Boron-containing shielding materials High density alloys

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Sensitivity analysis for dynamical response of reactor coolant system based on optimus

Sensitivity analysis for dynamical response of reactor coola...

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

作者： Yanli, Yuan Xianhui, Ye Lijuan, Li Feng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system. According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis. © 2020 American Society of Mechanical Engineers (ASME). All rig

关键词： Sensitivity analysis

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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...

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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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STUDY ON HYDROGEN RISK VENTING MITIGATION MEASURE OF SMALL STEEL CONTAINMENT DURING SEVERE ACCIDENT 29

STUDY ON HYDROGEN RISK VENTING MITIGATION MEASURE OF SMALL S...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zou, Zhiqiang Zhang, Hang Zhang, Ming Peng, Huanhuan Xu, Youyou Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886465

The hydrogen risk in small steel containment is one of the most important safety problems of small modular reactor (SMR), effective measures need to be taken to mitigate the risk of high concentration hydrogen in a small free volume during severe accidents. In the early stage of a severe accident caused by a loss of coolant accident, the pressure in the containment may rise rapidly, probably triggering the immediate venting of the containment, which can protect the integrity of the containment in the stage of pressure rise in the short term. In the later stage of a severe accident, the generation of hydrogen from reactor core may cause hydrogen combustion or even explosion, which may cause the possibility of overpressure in the containment. In this paper, Computational Fluid Dynamics (CFD) GASFLOW-MPI code is used to analyze the effect of containment venting measures on hydrogen risk in small steel containment of SMR during severe accident initiated by direct vessel injection line break loss-of-coolant. The analysis results show that the venting action can effectively depressurize the containment, the larger the area of the venting pipe is, the more obvious the depressurization effect is. However, hydrogen cannot be effectively discharged in the middle and later stage of the accident, so the risk of hydrogen cannot be completely eliminated, and the possibility of flame acceleration in containment cannot be ruled out for the reason that the negative pressure generated by water vapor condensation will suck in the air outside the containment and form a combustible gas cloud again. Copyright © 2022 by ASME.

关键词： Risk assessment

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