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Investigation into the flow and heat transfer characteristics in liquid metal cooled narrow rectangular channel with rough surfaces

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Case Studies in Thermal Engineering 2025年 72卷

作者： Xue Zhang Wenpei Feng Xiaolong Yang Hongxing Yu Jian Deng Sijia Du Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610213 China College of Nuclear Technology and Automation Engineering Chengdu University of Technology Chengdu 610059 China

This study investigates the impact of roughness elements on fluid flow and heat transfer characteristics in narrow rectangular channels. The results show that when the fluid flows over roughness elements, it induces vortices near the wall, disturbing the flow and heat transfer boundary layers. This enhances heat transfer while simultaneously increasing flow resistance. Among different roughness types, cylindrical roughness elements demonstrate a more substantial improvement in both heat transfer efficiency and friction factor compared to cubic elements. The Nusselt number is found to be non-monotonic related to the height of the roughness elements, while the friction factor increases with height. Additionally, larger roughness element diameters lead to higher Nusselt numbers and friction factors. Further analysis reveals that under conditions where the spacing and height remain unchanged, an increase in the roughness elements quantity results in a diminished enhancement of heat transfer. The base edge length of the roughness elements has a more significant impact on heat transfer capacity than the additional heated area, while the additional area primarily affects the flow resistance, causing the friction factor to increase with the number of roughness elements.

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Prediction of boiling crisis in a full-size rod bundle with spacer grids 14

Prediction of boiling crisis in a full-size rod bundle with ...

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

作者： Li, Quan Li, Yuanming Avramova, M. Huang, Yongzhong Liu, Zhenhai Ma, Chao Qi, Feipeng Zhao, Bo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Department of Nuclear Engineering North Carolina State University NC27695 United States

In this paper, two-phase CFD approach is used to simulate the boiling crisis in PSBT rod bundle with spacer grids. The 5×5 rod bundle contains 17 spacer grids, which are finely modeled through UG code. Eulerian two-fluid model coupled with the extended wall boiling model is used to simulate the subcooled boiling flow. The used sub-models for nucleation site density and bubble departure diameter both take the effects of high pressure into consideration. The calculated surface averaged void fraction in the central four sub-channels shows good agreement with the experimental data at high void fraction conditions, while higher than the measured data at low void fraction conditions. Based on the two-phase simulations, the void fraction distribution in the plane section on which thermocouples are placed is used as the criteria for CHF prediction. The numerical simulation of boiling two-phase flow and prediction method of CHF provide new approach for the optimization of spacer grid and fuel assembly. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Forecasting

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Preliminary development on thermal-hydraulic analysis code for the spent fuel rod under the condition of spray cooling

Preliminary development on thermal-hydraulic analysis code f...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Chao, Guo Jian, Deng Rong, Cai Yugao, Ma Lili, Liu Yuhao, Zhang Siyu, Lv Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Nuclear Science and Engineering School North China Electric Power University Changping District Beijing102206 China

ISBN: (纸本)9780791883785

According to spray cooling characteristics for failure of the cooling system in the spent fuel pool, we establish a model for analyzing spray cooling of a single spent fuel rod and develop a code named SFSC which is used to calculate the thermalhydraulic characteristics of film on the spent fuel rod under the condition of spray cooling. The results obtained from SFSC consist of film thickness, film temperature, film evaporation, fuel pellet temperature of spent fuel rod and cladding temperature of spent fuel rod at different height of spent fuel rod are shown in this paper. At the same time, we establish same model by RELAP and compare results calculated by RELAP with that calculated by SFSC. The good agreement of two results demonstrates that SFSC is effective. SFSC fills in the blanks of thermal-hydraulic calculation for the spent fuel rod under the condition of spray cooling. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Cooling

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The stress indices study for the pipe with trunnion welding attachment 27

The stress indices study for the pipe with trunnion welding ...

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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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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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Development of self-healing coating on fuel cladding for water-cooling nuclear reactors 14

Development of self-healing coating on fuel cladding for wat...

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

作者： Duan, Z.G. Xin, G.S. Zhou, Y. Li, Y.M. Abe, H. Yang, H.L. Kano, S. McGrady, J. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Nuclear Professional School School of Engineering University of Tokyo Tokai Ibaraki319-1188 Japan

A novel scheme for a bilayer coating with self-healing ability is proposed in this study. The candidate materials for the coatings and the potential self-healing reaction are assessed in high-temperature aqueous environments and high-temperature air. The pure Cr2O3 layer and the composite of Cr2O3 and MoO3 are the candidate materials for the outer layer and inner layer, respectively, due to their compatibility under normal condition. Fe2O3-MoO3 reactions exhibit a potential ability to heal the cracks because of a high reaction rate under normal condition. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Light water reactors

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Predicting Reverse Flow Characteristics in Equal-Height-Difference Passive Heat Removal system for Ocean Nuclear Power

SSRN

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SSRN 2022年

作者： Wang, Zhiwei He, Yanping Duan, Zhongdi Huang, Chao Li, Mingzhi Liu, Shiwen Hao, Chengming State Key Laboratory of Ocean Engineering Shanghai Jiao Tong University Shanghai200240 China Institute of Marine Equipment Shanghai Jiao Tong University Shanghai200240 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

The equal-height-difference passive heat removal system (EHDPHRS) via natural circulation driving and utilizes ocean as infinite heat sink for heat dissipation, showing broad application prospect in ocean nuclear power field;but the heat removal pipe normally encounters the reverse flow of the cold seawater into the pipe, leading to operation instability and even condensation-induced water hammer. In this study, the integration of the volume of fluid model and the condensation model based on surface renewal theory is used to simulate the reverse flow characteristics in heat removal pipe under single-and two-phase natural circulation conditions. The interfacial heat and mass transfer are calculated by user-defined functions. The numerical results reveal that the buoyancy force ( [[EQUATION]] ) and condensation driving force ( [[EQUATION]] ) both accelerate the reverse flow, while the inertial force ( [[EQUATION]] ) restrains the reverse flow;the reverse flow length presents monotonous decrease with increasing flow velocity under single-phase conditions and parabolic-like trend for two-phase flow. A prediction correlation for reverse length is developed using dimensionless Reynolds number, Grashof number and vapor volume fraction, covering single- and two-phase conditions. The predicted reverse length by the correlation agrees with 94% of the results within deviations of [[EQUATION]] 15%. © 2022, The Authors. All rights reserved.

关键词： Forecasting

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Stratification and heat transfer of molten corium pool for in-vessel retention

Stratification and heat transfer of molten corium pool for i...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Xiang, Qingan Deng, Jian Zhu, Dahuan Wu, Xiaoli Bi, Jinsheng Chen, Baowen Cai, Rong Qian, Libo YugaoMa Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Nuclear and radiation safety center Ministry of Environment Protection of P.R. China Beijing China

ISBN: (纸本)9780791883778

In-vessel retention (IVR) consists in cooling the molten corium contained in the lower head of reactor vessel by natural convection and reactor cavity flooding. The general approach which is used to study IVR problems is a "bounding" approach which consists in assuming a specified corium pool stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium pool. Traditionally molten corium pool in the lower head was expected to stratify into two-layer with the dense oxide pool at the bottom and the light metal layer on the top. Based on the MASCA experiments, the increased density of the metal layer attributed to a transfer of uranium metal leads to inverse stratification with a heavy metal layer relocating below the oxide pool. This behavior can be explained by physicochemical interaction between the oxidic and metallic phases of the corium pool. Therefore, a methodology which couples physicochemical effects and thermal hydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for calculate stratification probability of two-layer and three-layer configuration, analyze the safety margin of IVR in two-layer and three-layer configuration, and evaluate the lower head heat thermal failure probability. Copyright © 2020 ASME

关键词： Heat flux

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Experimental Study on Droplet Behavior and Mechanisms During Reflooding in a Single-Rod Channel

SSRN

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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

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Enhanced Toughness and Ductility of T91 Steel after Exposure to Liquid Lbe Saturated by Oxygen at 350-500 °C

SSRN

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SSRN 2024年

作者： Zhou, Zhikun Tian, Chenwen Dong, Wenbo Du, Juan Liu, Yunpeng Wahab, Magd Abdel Chen, Ziguang Department of Engineering Mechanics School of Aerospace Engineering Huazhong University of Science and Technology Wuhan430074 China Laboratory Soete Faculty of Engineering and Architecture Ghent University Technologiepark Zwijnaarde 46 ZwijnaardeB-9052 Belgium Institute of Intelligent Manufacturing Jianghan University Wuhan430056 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China College of Engineering Yuan Ze University Taiwan Hubei Key Laboratory of Engineering Structural Analysis and Safety Assessment 1037 Luoyu Road Wuhan430074 China

The mechanical tensile properties of T91 steel after being oxidized in liquid LBE saturated by oxygen at 350-500 °C for up to 1,000 h are studied. Duplex oxide scales are found on T91 steels after the LBE exposure, and the thickness evolution is revealed. Oxidized T91 steels with surface oxide scale yield enhanced toughness and ductility, which is mainly attributed to the inner oxide layer. We show that the enhancing effect relies on the presence of the oxide scale, not its thickness. The possible enhancing effect of the oxide scale on the toughness of T91 steel by impeding the dislocation from emitting out of the steel surfaces is discussed and qualitatively verified by micro-scale elastoplastic simulations based on a newly proposed discrete dislocation dynamics and peridynamics superposition scheme. © 2024, The Authors. All rights reserved.

关键词： Ductility

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