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

作者： Deng, Xiao Deng, Li-Ping Huang, Wei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784888982566

Water-lubricated thrust bearing is one of the key parts in canned motor pump, for example, the RCP in AP1000, and it functions to balance axial loads. A calculation model which can handle all flow state lubrication performance for water-lubricated thrust bearing has been presented. The model first includes laminar and turbulent Reynolds' equation. Then to get continuous viscosity coefficients cross critical Reynolds number, a transition zone which ranges based on engineering experience is put up, through which Hermite interpolation is employed. The model is numerically solved in finite difference method with uniform grids. To accelerate the calculation process, multigrid method and line relaxation is adopted within the iteration procedure. A medium sized water-lubricated tilting pad thrust bearing is exampled to verify the calculation model. Results suggest that as rotating speed enlarges, lubrication state distribution of the thrust bearing gradually tends to turbulent lubrication from the intersection corner of pad outer diameter and pad inlet to the opposite, minimum water film thickness increases approximately linearly, maximum water film pressure has little change, meanwhile the friction power grows nearly in exponential law which could result in bad effect by yielding much more heat. Copyright © 2015 by JSME.

关键词： Iterative methods

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A study of heat transfer calculation method of reactor vessel metallic insulation 22

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2014 22nd International Conference on Nuclear Engineering, ICONE 2014

作者： Dapeng, Yan Luo, Ying Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791845899

Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter "view factor", and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster. Copyright © 2014 by ASME.

关键词： Heat radiation

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Research on the response of core barrel under random turbulence excitation 25

Research on the response of core barrel under random turbule...

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25th International Congress on Sound and Vibration 2018: Hiroshima Calling, ICSV 2018

作者： Huang, Xuan Wu, Wan-Jun Qi, Huan-Huan Feng, Zhi-Peng Liu, Shuai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9781510868458

The flow-induced vibration of reactor structures is directly related to nuclear safety and has long been closely watched by the nuclear engineering community. A large number of theoretical, experimental and numerical analysis research work are carried out at home and abroad. In one hand, the test means cost a lot of money and time;on the other hand due to the restriction of test conditions, we only can get the measurement results about part of the structure under specific conditions. Therefore, from the design point of view, the theoretical or numerical analysis methods should be developed to predict structural flow induced vibration response in order to compensate for the lack of test methods. For the purpose of completing the flow-induced vibration analysis of the components in the reactor, it is necessary to analyse the vibration characteristics of the components in the reactor. According to the fluid excitation situation, this paper based on the test study two kinds of core barrel calculation method of forced vibration response under turbulent excitation. The advantages and disadvantages of these two methods are described from the aspects of calculation model, calculation input, analysis method and calculation result. © 25th International Congress on Sound and Vibration 2018, ICSV 2018: Hiroshima Calling. All rights reserved.

关键词： Vibration analysis

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A study on the sealing behaviors of the laterally one side restrained C-ring under internal pressure by numerical simulation 25

A study on the sealing behaviors of the laterally one side r...

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2017 25th International Conference on Nuclear Engineering, ICONE 2017

作者： Yuanyuan, Dong Ying, Luo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888982566

C-ring, which is one of the main sealing elements of reactor Pressure Vessels, is composed of a close-wound helical spring surrounded by two metal linings. C-ring has better resilience and sealing characteristics. Just because of the complex structure of C-ring, it was studied by experiments almost. There were also some numerical simulations on C-ring. With these methods, the compression-decompression process of C-ring under ideal condition can be simulated. But them cannot be used to study the sealing behavior of laterally one side restrained C-ring under internal pressure, which is fit for practical condition. This paper will establish a 3D numerical simulation method to study the C-ring which is mentioned as above. In this paper, it can be obtained that with internal pressure forcing and sealing groove squeezing, sealing performance of C-ring will reduce;when p>20MPa, the ratio of reducing becomes bigger;when p>140MPa, it will generate leakage;inner lining and outer lining will generate biggish plastic deformation, while the stress distribution in spring changes much. © Copyright 2017 ASME.

关键词： Numerical methods

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Neutronic feasibility analysis of fully ceramic microencapsulated fuel for commercial PWRs

Neutronic feasibility analysis of fully ceramic microencapsu...

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Top Fuel 2016: LWR Fuels with Enhanced Safety and Performance

作者： Mancang, Li Dan, Wang Yingrui, Yu Dong, Qin Wenjie, Li Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9780894487309

The fully ceramic microencapsulated (FCM) fuel is based on tristructural isotropic (TRISO) coated particles embedded in silicon carbide (SiC) matrix, which is among the accident tolerant fuel concept options. With the strength of high fission product retention, the FCM fuel provides potential superior safety performance. Compared to the conventional standard UO2 fuel, the deficiency of FCM fuel is its considerable lower fissile inventory, which results in a reduced reactor power or/and life. Supposing the FCM fuel is used as direct replacement of the fuel in commercial PWRs, the spectrum is softened due to the usage of SiC as matrix, and the moderator temperature coefficient is possible to be positive at the beginning of life, leading to high risk during operation. A neutronic feasibility study of FCM fuel utilized in commercial PWRs is performed in the paper. The fuel enrichment and undermoderated lattice are evaluated by analyzing the typical TRISO particle with UO2 kernels which is widely used in high temperature gas reactors and that with enlarged UN fuel kernels. It was found that under the mandatory constraint of low enriched uranium, the FCM fuel with UO2 kernels, although relatively mature, is only capable of small reactor designs, and the assembly lattice needs to be modified. The FCM fuel with UN kernels, when directly loading in the commercial PWRs, can yield comparable reactor power and life;however, a higher fuel enrichment and the fuel design and fabrication need to be optimized.

关键词： Silicon carbide

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The implicit runge kutta method for solving point burnup equations

The implicit runge kutta method for solving point burnup equ...

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2018 Transactions of the American Nuclear Society, ANS 2018

作者： Cai, Yun Peng, Xingjie Li, Qing Qin, Dong Jiang, Zhumin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

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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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CFD simulation of direct contact condensation phenomenon of vapor in subcooled water tank 23

CFD simulation of direct contact condensation phenomenon of ...

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

作者： Ji, Yu Quan, Yan Zhang, Yining Zhang, Haochun Wang, Yahui Harbin Institute of Technology Harbin Heilongjiang China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9784888982566

The direct contact condensation (DCC) is a significant phenomenon in nuclear reactor and its balance facilities, together with other chemical engineering systems. DCC occurs when the vapor is injected into the subcooled water, contacts and condenses on the interface directly. The DCC phenomenon accompanied with the heat transfer and mass transfer will result in the temperature and pressure fluctuations in the tank. This paper discusses the DCC in the subcooled water tank from the view of computation fluid dynamics. The numerical simulation was performed with the CFD software, Fluent, in which the direct contact condensation phenomenon was modeled with the Euler-Euler framework for two phases flow, and the evaporation & condensation model was adopted to simulate the mass transfer process. During the simulation process, the flow field and temperature field are derived. In addition, the shape and size of plume jet are also investigated. Copyright © 2015 by JSME.

关键词： Heat transfer

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Investigation on the Influence of Entrained Air to Internal Flow of Centrifugal Pumps under Cavitation Condition

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International Journal of Fluid Machinery and systems 2022年第1期15卷 56-63页

作者： Fanjie, Deng Jianping, Yuan Song, Yang Biaobiao, Wang Qiaorui, Si Research Center of Fluid Machinery Engineering and Technology Jiangsu University Jiangsu Zhenjiang212013 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu610000 China

The introduction of air is often used in engineering to delay the occurrence of cavitation. A steady simulation on a single-stage and single-suction centrifugal pump was used to study the influence of the air on the internal flow under cavitation condition at 1% inlet void fraction. Steady simulation based on SST k-ω turbulence model and Rayleigh-Plesset cavitation model included in Ansys CFX 17.0 were processed to obtain vapor volume fraction, gas volume fraction and the turbulent energy of the impeller under the condition of natural cavitation and cavitation with entrained air. A homogeneous two-phase flow model was adopted to describe the air-water mixed flow. The experimental cavitation performance curve provides boundary condition support for the numerical simulation. In conclusion, it shows that the head of the pump keeps steady after 1% air entrained like natural cavitation. Then it goes worse when the cavitation number decreases but with lower NPSHr which means cavitation performance is improved by 1% air entrained. It indicated that the entrained air changes the distribution of the vapor in the passage of the impeller. However, the descent of the pressure over a certain value will also do harm to the internal flow of the model pump whether it is ventilati ng or not. The internal flow of the pump will be seriously damaged and even be blocked completely due to more and more vapor generated causing a dramatic decline of the head. © 2022, Turbomachinery Society of Japan. All rights reserved.

关键词： Cavitation

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PREDICTION OF CRITICAL HEAT FLUX BASED ON RANDOM FOREST ALGORITHM 29

PREDICTION OF CRITICAL HEAT FLUX BASED ON RANDOM FOREST ALGO...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Sun, Zhuo Han, Mingyang Yuan, Peng Gong, Lei Chen, Di Yuan, Guanghui Zhang, Shuang Nuclear Power Institute of China Chengdu China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9780791886410

The accurate prediction of critical heat flux(CHF) is of great significance to the safe operation of the reactor. At present, most of the traditional methods for predicting CHF are based on relational fitting in mathematical form, which are not accurate enough and may can not taking full advantage of experimental data. Therefore, a prediction model based on random forest algorithm (RF) is proposed, the model is based on CHF 2006 look-up table, and the accuracy and smoothness of the model have been verified when predicting CHF. In this paper, the variation trend of CHF predicted value and actual value with pressure, mass flow rate and thermodynamic quality is given. It has been verified that the error of this method is acceptable, the predicted value is in good agreement with the experimental value, and conforms to the corresponding physical laws, which can provide a reference for the development of high-precision CHF prediction models or CHF-related experiments. Copyright © 2022 by ASME.

关键词： Heat flux

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