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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Lu, Xifeng Bai, Xiaoming Zheng, Liangang Zhang, Rui Yuan, Feng Yuan, Yanli Huang, Qian Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9784888983051

In the nuclear power plant, the trunnion welding attachment is widely used to connect the pipe and support. According to ASME Code Cases the stress caused by the welded attachment is considered as the additional stress of the pipe. However, the calculation of the additional stress is very complex and time-consuming. A method based on the concept of stress indices is proposed in present work. To investigate the influence of size and constraint and the stress indices, the finite element analysis for pipe with different sizes trunnion welding attachment under different constraint forms are done. The stress indices are calculated for each case, respectively. Furthermore, the fitted expressions for stress indices are proposed for general case. To verify this method, an analysis according to ASME Code Case for piping with trunnion is done. The comparison of the result between present method and the ASME Code Case method show slight different. Therefore, the stress indices method for the stress analysis of pipe with trunnion welding attachment can be an alternative method for ASME Code Cases. The advantage of present method in that it simplified the analysis process by introduces the stress indices. This method is effective and practical during the design and analysis of piping system in the nuclear power plant. Copyright © 2019 by JSME

关键词： Welding

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Model development of Uo2-zr Plate-Type fuel behavior at early phase of severe accident and molten fuel meat relocation 16

Model development of Uo2-zr Plate-Type fuel behavior at earl...

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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015

作者： Zhuohua, Zhang Shinian, Peng Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

ISBN: (纸本)9781510811843

According to former study on oxygen diffusion, Nb-Zr solid reaction and Uo2-zr solid reaction, models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of Uo2-zr plate. The new models can supply theoretical elements for safety analysis of core assembled with dispersion plate-type fuel under severe accident.

关键词： Uranium dioxide

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Abstract

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应用力学学报 2015年第1期32卷 I0001-I0016页

作者： Feng Zhipeng Zang Fenggang Zhang Yixiong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China 610213 Chengdu China

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Uncertainty quantification research of reactor system under seismic load 27

Uncertainty quantification research of reactor system under ...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Huang, Qian Xiong, Furui Bai, Xiaoming Yu, Xiaofei Zheng, Liangang Lu, Xifeng Song, Haiyang Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9784888983051

In the dynamic analysis of reactor system, the key mechanical input parameters are uncertain due to the design change, manufacturing, installation and calculation deviation, which affects the dynamic response, load distribution and final mechanical evaluation results. In order to quantify the influence of the uncertainty on seismic responses and load distribution, the uncertainty quantification of key structural input parameters of reactor system under seismic load is studied. Firstly, according to the fundamental characters of key input parameters (contact stiffness and gap among various assemblies), maximum entropy principle is adopted to construct the probability density function(PDF). Then, Markov Chain Monte Carlo technique is applied to sample according to the PDF, and the input and output data pool is constructed via finite element transient computations. According to the results, probabilistic distributions of dynamic responses caused by uncertainty from inputs parameters are evaluated and the reliability of nominal model is investigated. The studies show that the probabilistic distribution of reactor system dynamic response differs greatly with different output mechanical parameters, degrees of freedom and frequency bands. The uncertainty quantification method and process for reactor system are proposed in this paper. This method can be extended to other equipments easily. Therefore, it is an effective tools for the reliability of dynamic response. Copyright © 2018-2019 by JSME

关键词： Markov processes

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Analysis of contribution degree of vibration transmission of bolted structure 29

Analysis of contribution degree of vibration transmission of...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Liu, Shuai Li, Wanyou Huang, Xuan Xiong, Furui Jiang, Xiaozhou Deng, Dliwei Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Chengdu China Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9780791886359

In practical engineering, the vibration transmission between bolted structures supported by cantilever beams involves many factors, including the friction between plates and the bolt connection relationship. The vibration transmission effect is the embodiment of the coupling of various factors, so it is impossible to accurately and quantitatively evaluate the influence degree of various factors. In this paper, the influence factors of vibration between bolted plates in the form of cantilever beam are studied by numerical calculation method. The effects of friction coefficient, bolt preload, temperature and internal and external pressure on vibration characteristics are studied. Finally, the contribution of different influencing factors to vibration transmission is given. The preliminary research shows that the bolt connects the two plates into a whole structure, and its stiffness changes little with the friction coefficient, bolt preload, temperature and internal and external pressure. The natural frequency of each order increases with the increase of friction coefficient, bolt preload and internal and external pressure, decreases with the increase of temperature, and the overall trend changes little. At the same time, the bolt preload is verified by test, the difference of natural frequency under the same preload is compared and analyzed, the numerical calculation is compared with the test value, and the vibration contribution of preload is verified. © 2022 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Cantilever beams

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THE STUDY of COUPLED FUEL PERFORMANCE ANALYSIS and NEUTRONICS with COMSOL and RMC 29

THE STUDY of COUPLED FUEL PERFORMANCE ANALYSIS and NEUTRONIC...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Liu, Zhenhai Qi, Feipeng Zhou, Yi Li, Quan Chen, Ping Chen, Hao Ma, Chao Zhao, Bo Li, Chenxi Science and Technology on Reactor System Design Technology Laboratory Npic Chengdu China

ISBN: (纸本)9780791886366

Predicting fuel behavior in the reactor is a typical multiphysics coupling problem. The traditional fuel performance analysis adopts a decoupling way. To provide a higher fidelity tool for fuel performance simulation, the multi-physics coupling software COMSOL and 3D Monte Carlo neutron transport code RMC are coupled. The fuel rod thermo-mechanical coupling analysis module is constructed in COMSOL. RMC is innovatively wrapped as a component of COMSOL to communicate with the fuel performance module. Grid mapping algorithm is set up using COMSOL's functionality. A carefully designed segregation solver sequence is used to control the coupling process. The fine power distribution calculated by RMC is transferred to COMSOL fuel thermal calculation, and the fuel temperature and the coolant temperature and density calculated by COMSOL are transferred to RMC. In a Picard iterative way, the fuel rod lattice level neutron thermomechanical tight coupling is realized. The coupling is demonstrated in a fuel rod BOL steady state case. The fuel performance analysis is enhanced by the higher fidelity neutron flux solution. © 2022 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Iterative methods

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Key technology of a New Potential Economical Desalination Method for Small Nuclear Power 23rd

Key Technology of a New Potential Economical Desalination Me...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Li, Siguang Gan, Yiran Li, Yi Liang, Tiebo Science and Technology on Reactor System Design Technology Laboratory Sichuan Chengdu China

ISBN: (纸本)9789811988981

With the rapid development of distributed energy, small nuclear power has been widely concerned by researchers because of its inherent advantages. Keeping circulating water in a low salinity state is an important factor in economically maintaining the system operation. Therefore, it is necessary to investigate a specialized desalination method for small nuclear power. In this paper, expert knowledge and comparative analysis were introduced to study the advantages and disadvantages of some traditional desalination methods according to the recent requirements for small nuclear power. In addition, a potential economical namely hydrate-based desalination method was discussed, the key technologies that are urgent to be solved for this method were summarized and one of these technologies, phase equilibria of gas hydrate prediction, was studied using a thermodynamic model. Results show that each traditional desalination method cannot well meet the requirements for economy, energy and environment. Referring to the national requirements for carbon emission, the hydrate-based method is much suitable for small nuclear power due to its significant advantages in the fields of economy, energy and environment. In addition, the proposed thermodynamic model shows good performance compared with the published hydrate equilibrium data. Furthermore, the kinetics of hydrate formation and decomposition is considered another key to promoting the development of hydrate-based method. Finally, we hope the solution to the above problem can provide the idea of designing small nuclear power in the future. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Desalination

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Thermal-mechanical coupling behavior analysis of solid-state constrained component in advanced nuclear energy system

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Energy 2025年 330卷

作者： Xuan Yang Quan Li Jing Zhang Yingwei Wu Yanan He Chenxi Li Mingjun Wang Kailun Guo Chenglong Wang G.H. Su Wenxi Tian Suizheng Qiu State Key Laboratory of Multiphase Flow in Power Engineering School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an 710049 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610213 China

With the advantage of phase change heat transfer, heat pipes have the potential to replace flowing coolants for removing fission heat from nuclear reactors. However, the solid-state constrained heat transfer configuration poses mutual constraints between mechanical properties and heat transfer, and their long-term thermal-mechanical coupling behavior requires further investigation. Currently, research on long-term behavior is limited to either individual/local system components or a single physical field. In this paper, an analysis method for coupled thermal-mechanical behavior is proposed and verified, comprehensively considering thermal-mechanical properties, interactions between components, and the fission gas release. This method is employed to analyze the operating characteristics of a solid-state constrained component. The results indicate that prolonged operation leads to complete contact between structural components, generating high contact pressure that enhances heat transfer but increases creep. The release of gaseous fission products, accumulated over operating time, results in a synchronous increase in both gap and external contact pressures, reaching 5.2 MPa and 4.8 MPa, respectively. This process reduces the gas gap conductance, leading to elevated system peak temperatures and a reduction in temperature safety margins by 43 K. After heat pipe failure, continued operation significantly increases the local creep strain, up to 3.9 times that under normal conditions. The gap size and fuel gap pressure should be optimized to enhance gap heat transfer and reduce component creep. Excessive fission gas release should be avoided in fuel configuration. Reducing the system power following a single heat pipe failure can mitigate component creep and extend the operational lifespan.

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Ratcheting Behavior of SA508-3 Steel at Elevated Temperature:Experimental Observation and Simulation

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Acta Metallurgica Sinica(English Letters) 2017年第9期30卷 822-828页

作者： Jun Tian Yu Yang Li-Ping Zhang Xue-Jiao Shao Juan Du Qian-Hua Kan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Applied Mechanics and Structure Safety Key Laboratory of Sichuan Province School of Mechanics and EngineeringSouthwest Jiaotong University

A series of monotonic uniaxial tensile tests, strain-controlled and stress-controlled cyclic tests of SA508-3 steel were conducted from 25 to 350℃. Results showed that the steel exhibited temperature-dependent cyclic softening characteristic and obvious ratcheting behavior, and dynamic strain aging was observed in the range of 250-350℃. Based on experimental observations, a temperature-dependent cyclic plastic constitutive model was proposed by introducing the nonlinear cyclic softening and kinematic hardening rules, and the dynamic strain aging was also considered into the constitutive model. Comparisons between experiments and simulations were carded out to validate the proposed model at elevated temperature.

关键词： Ratcheting Constitutive model Dynamic strain aging Cyclic loading Elevated temperature

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Analysis of RCS depressurization effects on containment temperature for Qinshan phase II nuclear power plant 22

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2014 22nd International Conference on Nuclear Engineering, ICONE 2014

作者： Lili, Liu Ming, Zhang Jian, Deng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Chengdu China

ISBN: (纸本)9780791845936

A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containment during a station blackout (SBO) induced accident was analyzed. The sensitivity calculation indicated that the hydrogen generation rate obviously increased due to RCS depressurization in a critical stage. The results show that RCS depressurization can play an important role in hydrogen generation rate and total accumulation, and the temperature of the containment atmosphere is highly influenced by hydrogen combustion. High temperature induced by hydrogen combustion may degrade the equipment and instruments capabilities. Based on this analysis, a feasible strategy of RCS depressurization for mitigating the accident consequence is provided for developing the capacity of the SBO treatment of Qinshan Phase II Nuclear Power Plant (QSP-II NPP). Copyright © 2014 by ASME.

关键词： Nuclear power plants

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