检索结果-内蒙古大学图书馆

SSRN 2022年

作者： Chen, Mingjing Liu, Hanzhou Chen, Deqi Zhong, Lei Liu, Haidong Yu, Hongxing Yu, Ruan Feng, Peng Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing 400044 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Key Laboratory of Aerospace Thermophysics Ministry of Industry and Information Technology School of Energy Science and Engineering Harbin Institute of Technology Harbin150001 China

Discrete droplets have received extensive attention in reflooding processes under Loss Of Coolant Accident (LOCA) conditions due to their large vapor-liquid interface area, droplet collision, droplet evaporation, and other behaviors that have a significant impact on the heat transfer of Dispersed Flow Film Boiling (DFFB). Unfortunately, the accurate prediction of the characteristics of droplet velocity and distribution under DFFB cannot be achieved at present, due to the lack of sufficient visual data support. For this reason, a visualization experiment of high-temperature surface reflooding in a single-rod channel was carried out. In this paper, the effect of the thermal parameter, such as input heat flux on the transient droplet behavior, which include the droplet size and velocity, has been analaysed in detail. Furthermore, to provide physical support for building droplet distribution models containing more physical mechanisms in the future, several droplet formation and breakup mechanisms, which include the Helmholtz instability, capillary breakage and rewetting area sputter, has been discussed in this paper. © 2022, The Authors. All rights reserved.

关键词： Drop breakup

来源：评论

学校读者我要写书评

暂无评论

Cover Picture: Materials and Corrosion. 11/2024

引用

Materials and Corrosion 2024年第11期75卷 1399-1399页

作者： Wei Zhang Mingyang Zhou Huifang Yue Danmin Peng Xi Qiu Jijun Yang Key Laboratory of Radiation Physics and Technology of Ministry of Education Institute of Nuclear Science and Technology Sichuan University Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

来源：评论

学校读者我要写书评

暂无评论

Study in pool of research cooling reactor strategy at for accident pressure situation vessel

引用

Kerntechnik 2020年第2期85卷 88-97页

作者： Zhang, D. Ran, X. Zhang, Y. Zhang, H. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

Pressure vessel in pool layout is usually used in high power and middle pressure research reactor, Flow inversion and residual heat removal is important problem. The heat removal strategy at accident situation for a pressure vessel in pool layout research reactor was studied in the paper, many cases was researched for the ability of heat removal and safety of core. © Carl Hanser Verlag, München.

关键词： Research reactors

来源：评论

学校读者我要写书评

暂无评论

Evaluation of typical interfacial area transport models for bubbly flow

引用

International Journal of Advanced Nuclear reactor design and technology 2019年 1卷 10-18页

作者： Liu, Hang Lai, Jianyong Li, Yi Zhang, Yulong Yu, Xiaoquan Liu, Minghao Pan, Liangming Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Ministry of Education Chongqing400044 China

Interfacial area concentration is one of the most important parameters in two-phase flow. This parameter also reflects the interfacial area of mass, momentum and energy transfer in two-fluid model. The prediction of interfacial area concentration is essential to close the interfacial transfer terms in the two-fluid model. For the modeling of the one-dimensional interfacial area transport equation for adiabatic bubbly flow, existing typical sink and source terms have been reviewed and evaluated by the experimental data of vertical pipes and rod bundles. The interactive mechanisms of bubbles and turbulent eddies have also been analyzed, including bubbles breakup and bubbles coalescence. The evaluation results show the interfacial area transport model of Hibiki and Ishii (Liu and Hibiki) can provide the most accurate prediction of interfacial area concentration for pipes with the relative error of 8.22% for bubbly flow. Their model has also been evaluated by the experiment data of rod bundle with the relative error of 21.89% for bubbly flow. © 2019 Xi'an Jiaotong University

关键词： Two phase flow

来源：评论

学校读者我要写书评

暂无评论

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

引用

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.

关键词：

来源：评论

学校读者我要写书评

暂无评论

Correction to:Study on the resource utilization of high fluorine-containing organic waste through fluidized incineration

引用

Waste Disposal and Sustainable Energy 2023年第2期5卷 245-245页

作者： Wenhan Li Zengyi Ma Jianhua Yan Qianming Huang Xingjian Wen Zian Zhai Bochen Huang Shuang Wang Yongqiang Chen Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu610213China State Key Laboratory of Clean Energy Utilization Institute for Thermal Power EngineeringZhejiang UniversityHangzhou310027China Zhejiang Xianchuang Energy&Technology Co.Ltd Jinhua321000China

Correction to:Waste Disposal&Sustainable Energy(2022)4:117-129 https://***/10.1007/s42768-022-00101-7 The section Confict of interest'has been amended:'Jian-hua Yan is the Editor in-Chief of Waste Disposal&Sustain-able Energy.'The revised'Confict of interest'is as follows:Jianhua Yan is the Editor-in-Chief of Waste Disposal&Sustainable *** behalf of all authors,the corresponding author states that there is no conflict of interest.

关键词： Sustainable fluorine behalf

来源：评论

学校读者我要写书评

暂无评论

Numerical Simulation of Flow-Thermal Coupling Field Inside Upper Plenum and Hot Legs of Pressurized Water reactor (Pwr)

SSRN

引用

SSRN 2022年

作者： Sun, Ziyun Zhou, Xinzhi He, Zhengxi Xu, Tao Zhu, Jialiang Dong, Chenlong College of Electronics And Information Engineering Sichuan University No. 24 South Section 1 1st Ring Road Sichuan Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China No. 328 Section 1 Changshun Avenue Sichuan Chengdu China

The temperature and flow rate of coolant in the hot legs of pressurized water reactor (PWR) systems directly reflect the nuclear power and the heat transfer state of the reactor core, and are key parameters for reactor power control and safety protection. Most of existing studies have been conducted to analyze the steady-state thermal hydraulic features in the upper plenum and hot legs from a spatial scale, but less from a temporal scale to describe and analyze the transient features. In order to meet the requirements of key parameters measurement and safe operation, this paper established a reasonably simplified 3D geometrical model of the upper plenum and hot legs of the HPR1000 reactor based on a multi-physics field simulation platform and investigated the steady-state and transient features of the flow-thermal coupling field of coolant under numerical simulation, focusing on the coupling analysis of the flow-thermal coupling field on the spatial and temporal scales. The analysis of the steady-state features shows that the coolant flowing into the upper plenum from the upper core plate is characterized by a high middle temperature and a low peripheral temperature. As a result, there is an uneven coolant temperature distribution at the inlet of the hot legs that the temperature differences between the cold lower part and hot upper part are relatively large. However, with the axial flow of coolant, the temperature distribution in hot legs tends to be more uniform and the temperature differences decrease. The simulation results show that the flow of the low temperature coolant in hot legs dominates the change of coolant temperature distribution. The analysis of the transient features shows that a small fluctuation of flow rate at the core outlet causes changes in the turbulent kinetic energy of coolant in hot legs, but the effect on coolant temperature distribution is relatively weak. These such variations on the temporal scale can change the coupling relations of c

关键词： Multiphysics

来源：评论

学校读者我要写书评

暂无评论

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

引用

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

来源：评论

学校读者我要写书评

暂无评论

Acoustic indeced vibration research of dryer in steam generator 26

Acoustic indeced vibration research of dryer in steam genera...

引用

26th International Congress on Sound and Vibration, ICSV 2019

作者： Xianhui, Ye Fengshou, Zhang Fengchun, Cai Wenzheng, Zhang Furui, Xiong Haiyang, Song Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781999181000

The acoustic induced vibration of dryer (also called secondary separator) in steam generator in nuclear power plants is a concern in recent years. Periodic vibrations of long-term operation can lead to fatigue damage of structural components, posing a safety hazard. In this paper, the finite element analysis model of the dryer structure in steam generator of "Hualong-1" nuclear power plant was established, and the acoustic induced vibration characteristics of the structure were studied. Firstly, the acoustic excitation of the dryer structure was discussed, and the excitation inputs were obtained, including the maximum static pressure and the maximum pressure power spectrum density. Then the finite element model was established by ANSYS and the weak parts of the dryer structure were determined by the vibration characteristics combined with the acoustic excitation characteristics analysis. Finally, the structural strength and fatigue characteristics of the dryer structure under two different inputs of the acoustic load were investigated. The analysis of vibration characteristics shows that the main weak parts of the dryer structure are the vertical partition plate and the perforated plate. The static pressure analysis shows that the stress intensity of the two weak plates is less than the limit. In the study of pressure power spectral density, the maximum dynamic stress intensity and alternating stress of the vertical partition plate and the perforated plate are also less than the limit. The research results indicate that the acoustic induced vibration response of the dryer structure is within the acceptable range © Proceedings of the 26th International Congress on Sound and Vibration, ICSV 2019. All rights reserved.

关键词： Steam generators

来源：评论

学校读者我要写书评

暂无评论

Investigation of turbulent Prt model and the segmentation rule for LBE turbulent heat transfer 18

Investigation of turbulent Prt model and the segmentation ru...

引用

18th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2019

作者： Li, Peiying Yu, Hongxing Li, Meifu Du, Sijia Yan, Mingyu Liu, Yu Zhu, Dahuan Zhang, He Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

With liquid metal like lead-bismuth alloy(LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt=A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental r esults, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism. © 2019 American Nuclear Society. All rights reserved.

关键词： Liquid metals

来源：评论

学校读者我要写书评

暂无评论

建议与咨询留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

时间限定

文献类型

馆藏选择

核心期刊

语言

文献类型

帮助

文字说明：

检索规则说明：

检索范例：

分类表

所选分类

限定检索结果

文献类型

馆藏范围

日期分布

学科分类号

主题

机构

作者

语言

请选择保存的检索档案：

请选择收藏分类：

通借通还

建议与咨询 留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

时间限定

文献类型

馆藏选择

核心期刊

语言

文献类型

帮助

文字说明：

检索规则说明：

检索范例：

分类表

所选分类

限定检索结果

文献类型

馆藏范围

日期分布

学科分类号

主题

机构

作者

语言

请选择保存的检索档案： 新增检索档案 确定 取消

请选择收藏分类： 新增自定义分类 确定 取消

通借通还

建议与咨询留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

请选择保存的检索档案：

请选择收藏分类：