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

共 626 条记录，以下是171-180 订阅

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Research on thermal diffusion coefficient calculation for subchannels with spacer grid using CFD methodology 14

Research on thermal diffusion coefficient calculation for su...

引用

14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Chen, Xi Du, Sijia Liu, Yu Qian, Libo Liu, Lili Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

Thermal Diffusion Coefficient (TDC) is one of the most important parameters in subchannel analysis codes. Abundant research results demonstrate that TDC is mainly determined by the structures of fuel assembly, especially by the mixing vanes of spacer grids. Traditionally, TDC is acquired by thermal-hydraulic tests. However, hydraulic tests cost a great deal of money and need long time, and it's impossible to test every grid design. Another new way to obtain coefficients economically and fast is really needed for decades. Recently, with great development of Computational Fluid Dynamics, CFD has been validated as an analysis method for nuclear engineering, especially for single phase calculation. This paper established CFD methodology in TDC calculation for a 5x5 rod bundle with spacer grids. Firstly, CFD simulations were carried out to get sunchannel exit temperature including different conditions, such as different temperature, pressure, and flow rate. Then subchannel code CORTH was used to make subchannel analysis with varied TDC. Finally, we compared the TDC using CFD method with that come from tests, and validation results show that the methodology was right. In conclusion, the feasibility of TDC calculation methodology is preliminarily validated, which provided references for developments and design of new fuel assemblies. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Computational fluid dynamics

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Influence of Impedance Characteristics of Perforated Plate on Pump-Induced Pulsation in reactor 23rd

Influence of Impedance Characteristics of Perforated Plate o...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Yu, Sun Fengchun, Cai Sheng, Qian Yuzhao, Huangfu Ke, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811988981

In primary circuit, pump-induced pulsation (PIP) of coolant may cause fatigue failure of reactor components. In this article, a three-dimensional numerical model of PIP was established by COMSOL, theory of acoustic impedance of perforated plate was introduced and the perforated plate construction in reactor, such as flow baffle plate (FBP) was modeled by impedance boundary condition. In this paper, the calculation results of PIP pressure load were compared considering and without considering the impedance boundary condition. This study finds that differences of pressure amplitude on both sides of impedance boundary are significant when excitation frequency is coupled with the acoustic mode of computational model or when excitation frequency is relatively high. In addition, there are great differences of acoustic resonance frequency between two calculation models. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Boundary conditions

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Numerical research on rotating speed influence and flow state distribution of water-lubricated thrust bearing 23

Numerical research on rotating speed influence and flow stat...

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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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Computation of fluid thrust force from the pressure release tank of nuclear power plant pressurizer 17

Computation of fluid thrust force from the pressure release ...

引用

17th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2017

作者： Xiong, Fu-Rui Jiang, Nai-Bin Wu, Ge Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

Pressure release tank is a key component of nuclear power plant safeguard. Evaluating the fluid thrust force from high pressure and high temperature flow jet under postulated accidents is an essential task for nuclear system design and safety estimation. In this work, flow jet impingement from pressure release tank to nearby building roof is computed via a two-stage strategy. First, the RELAP5 code is applied to estimate thermal-hydraulic conditions at pressure release tank outlet. The thermal-hydraulic computational model includes key components of pressure release tank and its connection to the pressurizer. Based upon the thermal-hydraulic estimation within the pressure release tank, detailed computational fluid dynamics (CFD) simulation is performed with full considerations of how flow thrust is applied on nearby structures. The two-stage computational strategy provides a practical paradigm for fluid thrust force estimation. © 2016 Association for Computing Machinery Inc. All Rights Reserved.

关键词： Computational fluid dynamics

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

引用

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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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...

引用

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...

引用

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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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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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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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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