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

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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Application of feedforward predictive control in DC furnace coordination system

Application of feedforward predictive control in DC furnace ...

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2020 International Symposium on Automation, Mechanical and design Engineering, SAMDE 2020

作者： Deng, Zhiguang Zhengxi, H.E. Biwei, Z.H.U. Jialiang, Z.H.U. Xin, L.V. Qian, W.U. National Key Labortory of Science and Technology on Reactor System Design Technology Chengdu610213 China

In order to solve the control difficulties such as large dynamic delay, serious coupling between machine and furnace and strong nonlinear, the feedforward and feedback predictive control is proposed to control the coordinated system on the basis of multi-model predictive control, and the static feedforward part designs the feed line feedforward of coal quantity against coal quality disturbance on the basis of the accurate balance of energy. Thus, the reference total fuel quantity under the target load instruction is determined, and the dynamic part fully considers the dynamic compensation of the unit energy storage, gives a certain overshoot fuel quantity to the boiler regulation, speeds up its response, and ensures the rapidity of the unit. The real-time simulation results of virtual DPU (Distributed Processing Unit) show that the designed feedforward-feedback predictive control can quickly respond to load change, improve the resistance to fuel disturbance, reduce the fluctuation of main steam pressure, and better ensure the safe, economic and stable operation of the coordination system. © Published under licence by IOP Publishing Ltd.

关键词： Feedback

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

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

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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： 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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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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Optimization of spatial structure designs of control rod using Monte Carlo code RMC

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Frontiers in Energy 2021年第4期15卷 974-983页

作者： Hao LUO Mancang LI Shanfang HUANG Minyun LIU Kan WANG Department of Engineering Physics Tsinghua UniversityBeijing 100084China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology China National Nuclear Corporation Nuclear Power Institute of ChinaChengdu 610213China

Control rod is the most important approach to control reactivity in reactors,which is currently a cluster of pins filled with boron carbide(B4C).In this case,neutrons are captured in the outer region,and thus the inner absorber is ***,the lifetime of the control rod is challenged due to the high reactivity worth loss resulted from the excessive degradation of B4C in the high flux *** this work,some control rod designs are proposed with optimized spatial structures including the spatially mixed rod,radially moderated rod,and composite control rod with small-sized *** control rod worth and effective absorption cross section of these designs are computed using the Monte Carlo code RMC.A long-time depletion calculation is conducted to evaluate their burnup *** the spatially mixed rod,rare-earth absorbers are combined with B4C in spatial *** with the homogenous B4C rod,mixed designs ensure more sufficient reactivity worth in the lifetime of the *** minimum reactivity loss at the end of the cycle is only 1.8%from the dysprosium titanate rod,while the loss for pure B4C rod is nearly 12%.For the radially moderated design,a doubled neutronic efficiency is achieved when the volume ratio of moderator equals approximately 0.3,while excessive moderating may lead to the failure of control *** control rod with small-sized pins processes an enhanced safety performance and saves the investment in *** rod worth can be further enhanced by introducing small moderator pins,and the reactivity loss caused by the reduction of absorbers is sustainable.

关键词： control rod optimized spatial structure neutronic performance burnup stability

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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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Development and verification of SNTA code system for SCWR core steady state analysis 23

Development and verification of SNTA code system for SCWR co...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Wang, Lianjie Zhao, Wenbo Yang, Ping Ma, Yongqiang Lu, Di Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

A coupled neutronics/thermal-hydraulics three dimensional code system SNTA is developed for SCWR core steady state analysis by modular coupling the improved neutronics nodal methodological code and SCWR thermal-hydraulic sub-channel code. The problem of CSR1000 core is studied to verify SNTA. The results calculated by SNTA are agreed well with those by CASIR and SRAC. SNTA is more efficient than CASIR and SRAC whose neutronics modules are based on the Finite Difference Method. The numeric results show that SNTA can be well applied to SCWR core steady state analysis and core concept design. Copyright © 2015 by JSME.

关键词： Finite difference method

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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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