Query：
学者姓名：苏光辉
Refining：
Year
Type
Indexed by
Source
Complex
Co-Author
Language
Clean All
Abstract ：
The thermionic conversion performance of the single-cell thermionic fuel element (TFE) is influenced by fuel mass transfer and other parameters. With the consideration of current flow and heat transfer, the thermionic conversion model and the modified heat transfer model were implemented in FROBA-THERMION, a steady-state performance analysis code for the single-cell TFE. Afterward, both implemented models were verified by comparing with experiments or calculation results of KATET (a complex code developed by the Research Inst of SIA 'Lutch' of Russia to simulate single-cell TFE life behavior). Further, a steady-state performance simulation of the single-cell TFE was carried out. The results indicate that the thermionic conversion efficiency will be decreased due to the emitter temperature flattening caused by fuel mass transfer. In addition, the influence of several key parameters on thermionic conversion efficiency and measures to improve the thermionic conversion efficiency were analyzed. © 2022 Elsevier Ltd
Keyword ：
Fuel mass transfer; Thermionic conversion efficiency; Thermionic converter; Thermionic fuel element
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Lu, K. , Yao, H. , He, Y. et al. Thermionic conversion performance analysis of the single-cell thermionic fuel element based on FROBA-THERMION code [J]. | Annals of Nuclear Energy , 2023 , 181 . |
| MLA | Lu, K. et al. "Thermionic conversion performance analysis of the single-cell thermionic fuel element based on FROBA-THERMION code" . | Annals of Nuclear Energy 181 (2023) . |
| APA | Lu, K. , Yao, H. , He, Y. , Wu, Y. , Zhang, J. , Liao, H. et al. Thermionic conversion performance analysis of the single-cell thermionic fuel element based on FROBA-THERMION code . | Annals of Nuclear Energy , 2023 , 181 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
Sodium boiling will occur in some hypothetical accident cases (such as additional reactivity introduction and heat discharge capacity deterioration, and Local fault propagation). Thus sodium boiling experiments are necessary for the safety analysis and the development of serious accident programs in sodium-cooled fast re-actors. In the present study, the boiling two-phase heat transfer experiments of liquid sodium were carried out on the boiling liquid sodium test loop. The experimental results showed that the entire boiling process can be divided into four stages. The heat transfer coefficient of the boiling two phases in the rod bundle channel in-creases gradually with the increase of heat flux and grows very slowly with the increase of system pressure. Based on the 54 groups of rod channels obtained, the new correlation of transfer heat coefficient of boiling two-phase liquid sodium was developed in a 7-rod bundle. The new correlation can well predict the experimental data of boiling two-phase liquid sodium obtained by other scholars, and the prediction error is within +/- 50\% . Finally, mean relative deviation (MRD), absolute mean relative deviation (MARD), and root mean square deviation (RMSRD) were introduced to evaluate the accuracy of the correlation quantitatively.
Keyword ：
Boiling two-phase Correlation developed Heat transfer Liquid sodium Rod bundle
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Hou, Yandong , Wang, Liu , Zhang, Kui et al. Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle - Part II: Heat transfer characteristics [J]. | ANNALS OF NUCLEAR ENERGY , 2022 , 183 . |
| MLA | Hou, Yandong et al. "Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle - Part II: Heat transfer characteristics" . | ANNALS OF NUCLEAR ENERGY 183 (2022) . |
| APA | Hou, Yandong , Wang, Liu , Zhang, Kui , Wang, Mingjun , Wu, Yingwei , Tian, Wenxi et al. Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle - Part II: Heat transfer characteristics . | ANNALS OF NUCLEAR ENERGY , 2022 , 183 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
The two-phase pressure drop characteristics of liquid sodium are very significant for the safety and serious incident program development analysis of sodium-cooled fast reactor (SFR). In this paper, the boiling two-phase pressure drop experiments of liquid sodium were conducted in a 7-rod bundle with the mass flux range of 80 -390 kg/(m2.s), heat flux up to 120 kW/m2 and the absolute pressure range of 7.5 -100 kPa. The experimental results show that the frictional multiplier of boiling two-phase liquid sodium in the rod bundle channel decreases with the increase ofXLM. Some existing two-phase pressure drop correlations in the literatures were assessed and compared with the experimental data obtained in the present study. Results showed that these correlations could not predict the current experiments well because of the different geometries and measurements of high tem-perature environment. The new correlation for the two-phase friction multiplier of liquid sodium was developed in the 7 rod bundle. The predicted value of the new correlation was within 30 % of the experimental data in the rod bundle, which sufficiently demonstrates that the new correlation developed in this paper can well predict the boiling two-phase frictional pressure drop of liquid sodium in the rod bundle.
Keyword ：
Boiling two-phase Liquid sodium Pressure drop Rod bundle Two-phase friction multiplier
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Hou, Yandong , Wang, Liu , Zhang, Kui et al. Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle-Part I: Pressure drop characteristics [J]. | ANNALS OF NUCLEAR ENERGY , 2022 , 183 . |
| MLA | Hou, Yandong et al. "Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle-Part I: Pressure drop characteristics" . | ANNALS OF NUCLEAR ENERGY 183 (2022) . |
| APA | Hou, Yandong , Wang, Liu , Zhang, Kui , Wang, Mingjun , Wu, Yingwei , Tian, Wenxi et al. Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle-Part I: Pressure drop characteristics . | ANNALS OF NUCLEAR ENERGY , 2022 , 183 . |
| Export to | NoteExpress RIS BibTex |
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Wang, Chenglong , Wei, Shiying , Tian, Wenxi et al. 撤稿声明: Core design and analysis of a lead-bismuth cooled small modular reactor (Retraction of Vol 133, Pg 511, 2019) [J]. | ANNALS OF NUCLEAR ENERGY , 2022 , 179 . |
| MLA | Wang, Chenglong et al. "撤稿声明: Core design and analysis of a lead-bismuth cooled small modular reactor (Retraction of Vol 133, Pg 511, 2019)" . | ANNALS OF NUCLEAR ENERGY 179 (2022) . |
| APA | Wang, Chenglong , Wei, Shiying , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . 撤稿声明: Core design and analysis of a lead-bismuth cooled small modular reactor (Retraction of Vol 133, Pg 511, 2019) . | ANNALS OF NUCLEAR ENERGY , 2022 , 179 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
Heat pipes utilize the latent heat of vaporization and lie on the capillary forces to maintain the circulation of working fluid. High-temperature heat pipes applied in the nuclear silent thermal-electrical reactor are investigated and optimized. Nine wick structures and several design variables are investigated in the single/multi-objective optimization. The improved network thermodynamic model of a heat pipe is adopted, and the thermal resistance, mass, and transport capacity of wicks are selected as the objective functions. Multi-objective optimization results are obtained by Pareto-optimal points. Comparative results reveal that mesh screen and annular artery present a better performance (0.00180 K/W) in both single and multiple objective optimizations. Potassium is preferred for the wick of the circular artery and rectangular artery, while sodium is preferred for the wick of isosceles-triangular groove and sintered fibers. And with the operating temperature (800 similar to 950 K), thermal resistance is decreased with the simultaneous rise of mass (165.2%) and transport capacity (154.6%) in the mesh screen. This work makes it possible to improve the operational performance of a heat pipe and provides a reference for the selection of wick structure and heat pipe design.
Keyword ：
liquid-metal heat pipe multi-objective optimization single-objective optimization
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Tian, Zhixing , Liu, Yu , Wang, Chenglong et al. Single/multi-objective optimization and comparative analysis of liquid-metal heat pipe [J]. | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2022 , 46 (12) : 17521-17539 . |
| MLA | Tian, Zhixing et al. "Single/multi-objective optimization and comparative analysis of liquid-metal heat pipe" . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH 46 . 12 (2022) : 17521-17539 . |
| APA | Tian, Zhixing , Liu, Yu , Wang, Chenglong , Guo, Kailun , Zhang, Dalin , Tian, Wenxi et al. Single/multi-objective optimization and comparative analysis of liquid-metal heat pipe . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2022 , 46 (12) , 17521-17539 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
The power system of heat pipe reactor is divided into dynamic conversion and static conversion. Segmented thermoelectric generator (STEG) is a typical static conversion, but there is not enough literature to study its application in heat pipe reactor. Therefore, we have to design the STEG used in the heat pipe reactor. Firstly, we successfully simulated the STEG in COMSOL 5.5 and optimized the geometry and performance. The numerical simulation under steady-state conditions is carried out to determine the optimal STEG geometry. Then, the optimal STEG is connected with a heat pipe to form a single-channel model for simulation to explore the performance. The maximum thermoelectric performance can reach 15.75% for a single STEG, and the maximum stress is about 270 MPa. In the single-channel model, thermoelectric conversion efficiency reduces from 15.75% to 15.63%. This work provides a preliminary basis for numerical simulation in the combination between the STEG and heat pipe reactor.
Keyword ：
Heat pipe reactor Segmented thermoelectric generator Thermoelectric efficiency
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Zhang, Yin , Guo, Kailun , Wang, Chenglong et al. Numerical analysis of segmented thermoelectric generators applied in the heat pipe cooled nuclear reactor [J]. | APPLIED THERMAL ENGINEERING , 2022 , 204 . |
| MLA | Zhang, Yin et al. "Numerical analysis of segmented thermoelectric generators applied in the heat pipe cooled nuclear reactor" . | APPLIED THERMAL ENGINEERING 204 (2022) . |
| APA | Zhang, Yin , Guo, Kailun , Wang, Chenglong , Tang, Simiao , Zhang, Dalin , Tian, Wenxi et al. Numerical analysis of segmented thermoelectric generators applied in the heat pipe cooled nuclear reactor . | APPLIED THERMAL ENGINEERING , 2022 , 204 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
Heat pipe cooled reactor (HPR) is an ideal technology and a potential future niche where reliability and simplicity are key requirements. The heat transfer capacity of heat pipes determines the application range of HPR. Heat transfer limits (HTLs) of heat pipes are investigated, and the models of HTLs are evaluated in this study. Various HTLs of heat pipes with different filling ratios are tested. For the capillary limit, there is a sudden dryout in the evaporator section, in which the model of Chi is applicable for the horizontal conditions with a deviation of 19.0% but not for the inclined conditions. There is a possibility of sonic limit for a temperature rise in the adiabatic and condenser sections. The sonic limit predicted by Levy model is always higher than the experimental results. The entrainment limit is accompanied by a sudden rise in temperature of the evaporator section and temperature fluctuation in the condenser section, which can be estimated by using the wave-induced model with a relative error of 45.3% and is affected by the wick and inclination angle. The Knudsen number is used to determine the vapor state and the recommended critical Knudsen number of 0.01 is less than experimental data, which neglects the effect of inclination angle. The experimental results provide references for the heat pipe design and the accuracy of the present models are estimated.
Keyword ：
Experimental investigation Heat pipe cooled reactor Heat transfer limits Sodium heat pipe
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Tian, Zhixing , Zhang, Jiarui , Wang, Chenglong et al. Experimental evaluation on heat transfer limits of sodium heat pipe with screen mesh for nuclear reactor system [J]. | APPLIED THERMAL ENGINEERING , 2022 , 209 . |
| MLA | Tian, Zhixing et al. "Experimental evaluation on heat transfer limits of sodium heat pipe with screen mesh for nuclear reactor system" . | APPLIED THERMAL ENGINEERING 209 (2022) . |
| APA | Tian, Zhixing , Zhang, Jiarui , Wang, Chenglong , Guo, Kailun , Liu, Yu , Zhang, Dalin et al. Experimental evaluation on heat transfer limits of sodium heat pipe with screen mesh for nuclear reactor system . | APPLIED THERMAL ENGINEERING , 2022 , 209 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
Heat pipe cooled reactor with static conversion is gaining more and more attention because of its unique advantages such as mobility, high reliability, simple structure, long lifetime, and high-power density. A 100 kW heat pipe reactor with the three-segment type thermoelectric generators namely NUSTER-100 is designed by Xian Jiaotong University. Based on the finite element software COMSOL Multiphysics, a multi-physics coupling analysis platform is established for the static heat pipe reactor NUSTER-100. The heat transfer models of the heat pipe and thermoelectric models of the three-segment type TEGs are supplemented in the simulation model. The thermal-electrical coupling characteristics of the static heat pipe reactor are investigated under different operation conditions. For the steady state condition, the temperature difference of the heat pipe between evaporator and condenser stabilizes at around 110 K, the main temperature gradient about 850 K appears between the two side of TEGs. The total output power of the NUSTER system is 123 kW and the system thermoelectric conversion efficiency is 12.3%. For the three rows unloading condition, the maximum temperature increases to 1483 K which is 133 K higher than that on the steady state condition. The output electrical power of the reactor system is 114.5 kW and the system thermoelectric conversion still could reach 11.5%. For the whole reactor unloading condition, the reactor core temperature increases by 393 K compared with the steady state and the maximum temperature is 1743 K which is lower than the core melting temperature. The operating temperatures of the heat pipes are still within the normal operating temperature range of the sodium heat pipe. The heat transfer capacity of the heat pipe is still sufficient to ensure that the core temperature is kept within the safe limits. In summary, the unloading condition will cause the core temperature to rise significantly. Therefore, the system safety characteristics under accident transient conditions need to be focused coupling with the unloading condition in the further research. (c) 2021 Elsevier Ltd. All rights reserved.
Keyword ：
Heat pipe Heat pipe reactor Multi-physics coupling platform TEG Thermal-electrical coupling characteristics
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Tang, Simiao , Liu, Xiao , Wang, Chenglong et al. Thermal-electrical coupling characteristic analysis of the heat pipe cooled reactor with static thermoelectric conversion [J]. | ANNALS OF NUCLEAR ENERGY , 2022 , 168 . |
| MLA | Tang, Simiao et al. "Thermal-electrical coupling characteristic analysis of the heat pipe cooled reactor with static thermoelectric conversion" . | ANNALS OF NUCLEAR ENERGY 168 (2022) . |
| APA | Tang, Simiao , Liu, Xiao , Wang, Chenglong , Zhang, Dalin , Su, G. H. , Tian, Wenxi et al. Thermal-electrical coupling characteristic analysis of the heat pipe cooled reactor with static thermoelectric conversion . | ANNALS OF NUCLEAR ENERGY , 2022 , 168 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
Nuclear thermal propulsion utilizes the nuclear reactor rather than the combustion chamber to yield thermal energy. Propellant hydrogen could dissociate in the high-temperature reactor, which has an important effect on thermal hydraulic performance of the reactor. In this study, a one-dimensional steady-state analysis code has been developed for studying the behavior of hydrogen flowing through the high-temperature coolant channel. Thermal dissociation and real gas thermophysical property models of hydrogen were proposed and considered in the calculation models. It was found that the model validation deviations of thermophysical properties were within +/- 5% in the range of 200 - 3000 K and 0.01 - 10.0 MPa. Developed models were reliable and accurate with validation. Thermal-hydraulic behaviors of hydrogen in NRX-A6 reactor channel were analyzed. When dissociation occurred, the variation of properties was larger than those without dissociation, which enhanced heat transfer. The degree of dissociation was small and xH,out was just 0.456% under the design condition. The power density was the most significant influence factor, especially under the high power density. xH,out was 3.2 times than that of design condition as the power density grew 20%. This study can provide an approach to study hydrogen dissociation phenomena in nuclear thermal propulsion reactors.
Keyword ：
Code validation Dissociation Hydrogen Nuclear thermal propulsion Thermal hydraulics Thermophysical properties
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | Fang, Yuliang , Wang, Chenglong , Tian, Wenxi et al. Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor [J]. | NUCLEAR ENGINEERING AND DESIGN , 2022 , 392 . |
| MLA | Fang, Yuliang et al. "Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor" . | NUCLEAR ENGINEERING AND DESIGN 392 (2022) . |
| APA | Fang, Yuliang , Wang, Chenglong , Tian, Wenxi , Zhang, Dalin , Su, Guanghui , Qiu, Suizheng . Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor . | NUCLEAR ENGINEERING AND DESIGN , 2022 , 392 . |
| Export to | NoteExpress RIS BibTex |
Abstract ：
针对目前雾化研究尚未有能够模拟从入口连续相到雾化场全过程的数值模拟方法的现状,提出了一种全流场的液相雾化破碎数值模拟方法.该方法基于最大不稳定波增长率破碎理论,在喷雾头出口位置构建液膜初级破碎模型-PBM,链接VOF和DPM模型,从而形成全流场的数值模拟方法.本文使用VOF+PBM+DPM模型,基于FLUENT求解器,对一种旋流式大型喷雾头进行了数值模拟.全流场模型计算得到的喷雾场流量密度分布与实验结果进行比较,两者符合较好,表明了该模型计算雾化全过程的有效性.
Cite：
Copy from the list or Export to your reference management。
| GB/T 7714 | 兰治科 , 李勇 , 苏光辉 et al. 连续液相碎化过程数值模拟研究 [J]. | 当代化工研究 , 2021 , (10) : 4-6 . |
| MLA | 兰治科 et al. "连续液相碎化过程数值模拟研究" . | 当代化工研究 10 (2021) : 4-6 . |
| APA | 兰治科 , 李勇 , 苏光辉 , 昝元锋 . 连续液相碎化过程数值模拟研究 . | 当代化工研究 , 2021 , (10) , 4-6 . |
| Export to | NoteExpress RIS BibTex |
Export
| Results： |
Selected to |
| Format： |
