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学者姓名：田文喜
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Abstract ：
For the compact PWR with cartridge fuel component, the application of flow partition design can significantly flatten the core outlet temperature distribution, improve the outlet subcooling, or reduce the coolant flow, to make more effective use of the coolant and greatly improve the core design performance. However, the three-dimensional power distribution shape of the core will change greatly during its lifetime, and it is difficult to reasonably distribute the flow. In this study, the closed channel core flow partition and water density feedback program is designed by studying the basis of flow partition and designing an optimization method that considers various factors. The optimization search of core flow distribution is carried out to maximize the subcooling degree at the component outlet and flatten the core outlet temperature distribution. In addition, the minimum coolant mass flow and the maximum outlet temperature are estimated under the given core thermal power, three-dimensional distribution and minimum outlet subcooling.
Keyword ：
Code development Flow rate distribution Optimization method Sub-channel
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| GB/T 7714 | Cai, J. Y. , Zhang, K. , Chen, L. et al. Program development for mass flowrate distribution optimization in the nuclear power plant [J]. | ANNALS OF NUCLEAR ENERGY , 2023 , 184 . |
| MLA | Cai, J. Y. et al. "Program development for mass flowrate distribution optimization in the nuclear power plant" . | ANNALS OF NUCLEAR ENERGY 184 (2023) . |
| APA | Cai, J. Y. , Zhang, K. , Chen, L. , Xia, B. Y. , Chen, R. H. , Tian, W. X. et al. Program development for mass flowrate distribution optimization in the nuclear power plant . | ANNALS OF NUCLEAR ENERGY , 2023 , 184 . |
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Abstract ：
In a light water reactor, nuclear fuel element melting is critical. This paper will introduce the experimen-tal studies on the melting and relocation behaviors of the single fuel rod and 3 x 3 fuel bundle imitators based on aluminum and zinc as alternative materials. The eutectic reaction can occur between Al and Zn. Moreover, the lower melting points of Al and Zn allow us to observe the experimental phenomena directly. In this paper, three single-rod experiments and a rod bundle experiment are discussed in de-tail. The results of the experiments are analyzed at both macro and micro levels. In the experiments of slowly increasing temperature, the melting phenomena of fuel rod imitators mainly include rod defor-mation, eutectic reaction, melting, erosion, and molten materials relocation. Among them, the existence of a eutectic reaction between materials will make the materials melt under their melting point, which significantly influences the melting time and behavior of the rod. Through this study, the nuclear fuel rods melting and relocation behaviors under the influence of the eutectic reaction can be observed, and the experimental results can provide data support for the development of related models. (c) 2021 Elsevier B.V. All rights reserved.
Keyword ：
Alternative materials Eutectic reaction Melting and relocation Microstructure analysis Visual research
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| GB/T 7714 | Li, Yonglin , Tian, Wenxi , Chen, Ronghua et al. Research on the nuclear fuel rods melting behaviors by alternative material experiments [J]. | JOURNAL OF NUCLEAR MATERIALS , 2022 , 559 . |
| MLA | Li, Yonglin et al. "Research on the nuclear fuel rods melting behaviors by alternative material experiments" . | JOURNAL OF NUCLEAR MATERIALS 559 (2022) . |
| APA | Li, Yonglin , Tian, Wenxi , Chen, Ronghua , Feng, Tangtao , Qiu, Suizheng , Su, G. H. . Research on the nuclear fuel rods melting behaviors by alternative material experiments . | JOURNAL OF NUCLEAR MATERIALS , 2022 , 559 . |
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Abstract ：
Lead-cooled fast reactor (LFR) adopts the liquid metal lead as the coolant and is beneficial to the fuel sus-tainability and high safety. LFR is suffering from the serious thermal stratification due to the pool type design and high operation temperature, leading to the threatening thermal stress on the structure. In this paper, the three-dimensional CFD model of ELSY lead pool is established and the thermal stratification phenomena under the forced and natural circulations are investigated. The results show that under steady-state conditions, ELSY has temperature stratification along the wall in the upper lead pool where the steam generator is located, and a high temperature concentration zone and a speed stagnant zone in the lower plenum. Under natural circulation conditions, when the water decay heat removal system (W-DHR), isolation condenser (IC) and reactor vessel air cooling system (RVACS) are put into operations, there is also obvious temperature stratification in the steam generator area. This paper puts forward some corresponding optimization suggestions on the problems in the simulation. The above results can provide an appropriate and effective reference for the design of ELSY in the future.(c) 2022 Published by Elsevier Ltd.
Keyword ：
CFD ELSY Forced circulation Natural circulation Thermal stratification
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| GB/T 7714 | Dong, Zhengyang , Qiu, Hanrui , Wang, Mingjun et al. Numerical simulation on the thermal stratification in the lead pool of lead-cooled fast reactor (LFR) [J]. | ANNALS OF NUCLEAR ENERGY , 2022 , 174 . |
| MLA | Dong, Zhengyang et al. "Numerical simulation on the thermal stratification in the lead pool of lead-cooled fast reactor (LFR)" . | ANNALS OF NUCLEAR ENERGY 174 (2022) . |
| APA | Dong, Zhengyang , Qiu, Hanrui , Wang, Mingjun , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . Numerical simulation on the thermal stratification in the lead pool of lead-cooled fast reactor (LFR) . | ANNALS OF NUCLEAR ENERGY , 2022 , 174 . |
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Abstract ：
In this study, the transient behavior of two-layer corium pool was studied based on COPRA facility. The transient change of two-layer corium pool caused by power change and the effect of stratification and stratified plate thickness on heat transfer characteristics were investigated. Then maximum heat flux ratios (MHFRs) of twolayer corium pools was proposed and investigated using CFD code FLUENT. The heat flux ratio is the ratio of heat flux to qCHF of vessel wall. The maximum of heat flux ratios at all angles is the MHFR of the corium pool. The MHFRs of two-layer corium pool under radiation top surface condition and isothermal top surface condition were obtained. By comparing the values and positions of the MHFRs, the degree of approaching boiling crisis of local reactor pressure vessel wall can be quantitatively analyzed. The results show that MHFR appears at metallic layer wall under radiation top surface condition and at upper part of the oxide layer wall under isothermal top surface condition. The results of this research can provide references for IVR (In-Vessel Retention) safety analysis and ERVC (External-Reactor Vessel Cooling) methods such as Nano-coating measure.
Keyword ：
COPRA experiment Maximum heat flux ratio Solidification and melting Transient behavior Two-layer corium pool
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| GB/T 7714 | Ge, Kui , Zhou, Yukun , Zhang, Yapei et al. Transient behavior and maximum heat flux ratios of two-layer corium pool [J]. | INTERNATIONAL JOURNAL OF THERMAL SCIENCES , 2022 , 179 . |
| MLA | Ge, Kui et al. "Transient behavior and maximum heat flux ratios of two-layer corium pool" . | INTERNATIONAL JOURNAL OF THERMAL SCIENCES 179 (2022) . |
| APA | Ge, Kui , Zhou, Yukun , Zhang, Yapei , Wu, Shihao , Tian, Wenxi , Su, G. H. et al. Transient behavior and maximum heat flux ratios of two-layer corium pool . | INTERNATIONAL JOURNAL OF THERMAL SCIENCES , 2022 , 179 . |
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Abstract ：
Model reduction is a method that maps full-order conservation equations into lower-order subspaces or establish a data-driven surrogate model to reduce the complexity of the entire physical system, which has been widely applied in various fields in recent years. Compared with computational fluid dynamics (CFD) simulations, reduced-order model (ROM) can quickly and instantly obtain simulation results at low cost, which provides an economical alternative approach for the research and design process which need large number of repetitive simulations. In this paper, a deep-learning ROM was developed based on the proper orthogonal decomposition (POD) and machine learning (ML) method. The rapid estimation of two significant thermal hydraulic parameters in steam generator (SG), including the void fraction and tem-perature, was carried out by ROM. By POD mode analysis, the order for void fraction and temperature field was reduced by 88.3% and 96.7%, respectively. An artificial neural network was trained to reflect the implicit nonlinear mapping relationship between the CFD inputs and feature coefficients. The ROM was validated by comparing the predicted results with refined CFD results. The maximum absolute errors of void fraction and temperature are 0.1 and 0.03 K with speedup on the order of 10 4 , indicating that the developed ROM can quickly and accurately estimate the thermal hydraulic characteristics of SG un-der different operating conditions. This work may provide a novel approach for the parameter sensitivity analysis and optimization design of SG and give valuable reference for the digital twin and the real-time online monitoring of the SG. (c) 2022 Elsevier Ltd. All rights reserved.
Keyword ：
CFD Deep -learning POD Reduced -order models Steam generators
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| GB/T 7714 | He, Shaopeng , Wang, Mingjun , Zhang, Jing et al. A deep-learning re duce d-order model for thermal hydraulic characteristics rapid estimation of steam generators [J]. | INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER , 2022 , 198 . |
| MLA | He, Shaopeng et al. "A deep-learning re duce d-order model for thermal hydraulic characteristics rapid estimation of steam generators" . | INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER 198 (2022) . |
| APA | He, Shaopeng , Wang, Mingjun , Zhang, Jing , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . A deep-learning re duce d-order model for thermal hydraulic characteristics rapid estimation of steam generators . | INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER , 2022 , 198 . |
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Abstract ：
This study investigated the vapor condensation characteristics in an inclined pipe under natural convection conditions experimentally and numerically. The inclined blind-end concentric pipe had an inner diameter of 134 mm and a length of 680 mm, which is cooled by cooling water. The experimental pressure range was 0.2-0.6 MPa, and the dimensionless mass number range was 0.44 to 202. The fog formation phenomenon was observed in the experiment. An empirical correlation with a broad range of dimensionless mass numbers was developed based on the experimental results to predict the heat transfer coefficient (HTC). When the air mass fraction was large and the HTC was low, the effect of fog formation on the HTC had to be considered. In the numerical simulation, this study aimed to develop an HTC prediction model with broader applicability. Based on the diffusion boundary layer theory, this study improved Peterson's model by adding fog effect and wave effect. The model showed high adaptability to this experiment and other experiments in the natural convection condensation database. The improved model performed significantly better, with 96% of the data falling within an error zone of 30%. The wide range of dimensionless mass number test data might complement the natural convection condensation experimental database. (c) 2021 Elsevier Ltd. All rights reserved.
Keyword ：
Condensation Fog formation Heat and mass transfer Modeling Suction effect
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| GB/T 7714 | Tan, Bing , Cai, Jiejin , Zhao, Jiyun et al. Experimental and theoretical study of vapor/air mixture condensation inside an inclined blind-end pipe in natural convection with considering fog formation [J]. | INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER , 2022 , 184 . |
| MLA | Tan, Bing et al. "Experimental and theoretical study of vapor/air mixture condensation inside an inclined blind-end pipe in natural convection with considering fog formation" . | INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER 184 (2022) . |
| APA | Tan, Bing , Cai, Jiejin , Zhao, Jiyun , Hibiki, Takashi , Tian, W. X. , Wu, Y. W. . Experimental and theoretical study of vapor/air mixture condensation inside an inclined blind-end pipe in natural convection with considering fog formation . | INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER , 2022 , 184 . |
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Abstract ：
As one of the most widely used types of heat exchangers, the shell-and-tube heat exchanger (STHX) offers the advantages of being low-cost, easy to clean, and highly reliable under high pressure and temperature conditions. Generally, a large number of tube bundles are used in the STHX to increase the heat transfer capacity. The three-dimensional (3D) two-phase flow simulation of the STHX is made exceedingly difficult if the tube bundle region is full-size modeled. Therefore, simplified models and approaches are required to meet the urgent demands of STHX engineering numerical simulation, especially for 3D analyses. Herein, we have developed an OpenFOAM solver "PorousDriftFoam " suitable for the two-phase flow and the boiling heat transfer numerical simulation of the STHX. The porous media model was applied to simplify the tube bundle region and the drift-flux model (DFM) was adopted in the two-phase flow computational fluid dynamics (CFD) simulation for the STHX. The heat transfer tubes are regarded as the solid region of porous media, while the shell side is considered the fluid region. We calculate the coupling heat transfer between the tube and the shell sides and consider the resistance introduced by the heat transfer tubes and support plates in the STHX. Two international benchmarks, the FRIGG test and the MB-2 experiments, were selected to fully validate the solver. For the FRIGG experiment, the predicted void fraction stands in good agreement with the experimental data, with an average absolute error of 0.07 and an average relative error of 13.3%. For the MB-2 experiment, the predicted values of pressure drop and temperature fit well with the experimental data. The developed solver could be applied in the 3D two-phase flow and heat transfer numerical simulation of a typical STHX in the industry.
Keyword ：
OpenFOAM Porous media model Shell-and-tube heat exchanger Two-phase flow
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| GB/T 7714 | He, Shaopeng , Wang, Mingjun , Tian, Wenxi et al. Development of an OpenFOAM solver for numerical simulations of shell-and-tube heat exchangers based on porous media model [J]. | APPLIED THERMAL ENGINEERING , 2022 , 210 . |
| MLA | He, Shaopeng et al. "Development of an OpenFOAM solver for numerical simulations of shell-and-tube heat exchangers based on porous media model" . | APPLIED THERMAL ENGINEERING 210 (2022) . |
| APA | He, Shaopeng , Wang, Mingjun , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . Development of an OpenFOAM solver for numerical simulations of shell-and-tube heat exchangers based on porous media model . | APPLIED THERMAL ENGINEERING , 2022 , 210 . |
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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
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| 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 . |
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Abstract ：
In this study, a set of mathematic models suitable for flow boiling simulation and CHF prediction in rod bundle channels of Pressurized Water Reactor (PWR), including Eulerian two-phase model combined with wall boiling model, were established using CFD method. The models were validated against the international flow boiling and Critical Heat Flux (CHF) experiment benchmark. The results show that the error between the calculated CHF and the experimental data is less than 10%. Then a 3-D model of 5 x 5 rod bundles with cold wall was built and the detailed three-dimensional thermal-hydraulic characteristics were investigated, especially emphasising on the cold wall effect. The existance of cold wall would affect the coolant thermal mixing intensity, leading to a certain degree influence of the CHF. Results show that under the same inlet conditions, the calculated CHF value after introducing the cold wall is higher by 11.4%, while the overall power decreases due to the existence of cold wall. Besides, the pressure drop and temperature in the channel would decrease, while the distribution of void fraction basically remain unchanged, which is still mainly affected by the spacer grids and mixing vanes. Compared with the modified W-3 formula, the error of the CHF value calculated by CFD method is less than 10%, and the CFD method can show the three-dimensional distribution characteristics and evolution process of physical parameters.
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| GB/T 7714 | Zhang, Ji , Wang, Mingjun , Chen, Chong et al. CFD investigation of the cold wall effect on CHF in a 5 x 5 rod bundle for PWRs [J]. | NUCLEAR ENGINEERING AND DESIGN , 2022 , 387 . |
| MLA | Zhang, Ji et al. "CFD investigation of the cold wall effect on CHF in a 5 x 5 rod bundle for PWRs" . | NUCLEAR ENGINEERING AND DESIGN 387 (2022) . |
| APA | Zhang, Ji , Wang, Mingjun , Chen, Chong , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . CFD investigation of the cold wall effect on CHF in a 5 x 5 rod bundle for PWRs . | NUCLEAR ENGINEERING AND DESIGN , 2022 , 387 . |
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Abstract ：
Pool boiling heat transfer is widely applied in nuclear engineering fields. The influence of heating surface orientation on the pool boiling heat transfer has received extensive attention. In this study, the heating surface with different roughness was adopted to conduct pool boiling experiments at different inclination angles. Based on the boiling curves and bubble images, the effects of inclination angle on the pool boiling heat transfer and critical heat flux were analyzed. When the inclination angle was bigger than 90 degrees, the bubble size increased with the increase of inclination angle. Both the bubble departure frequency and critical heat flux decreased as the inclination angle increased. The existing theoretical models about pool boiling heat transfer and critical heat flux were compared. From the perspective of bubble agitation model and Hot/Dry spot model, the experimental phenomena could be explained reasonably. The enlargement of bubble not only could enhance the agitation of nearby liquid but also would cause the bubble to stay longer on the heating surface. Consequently, the effect of inclination angle on the pool boiling heat transfer was not conspicuous. With the increase of inclination angle, the rewetting of heating surface became much more difficult. It has negative effect on the critical heat flux. This work provides experimental data basis for heat transfer and CHF performance of pool boiling. (c) 2021 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).
Keyword ：
Critical heat flux Inclination angle Pool boiling Theoretical and experimental analysis
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| GB/T 7714 | Wang, Chenglong , Li, Panxiao , Zhang, Dalin et al. Experimental study on the influence of heating surface inclination angle on heat transfer and CHF performance for pool boiling [J]. | NUCLEAR ENGINEERING AND TECHNOLOGY , 2022 , 54 (1) : 61-71 . |
| MLA | Wang, Chenglong et al. "Experimental study on the influence of heating surface inclination angle on heat transfer and CHF performance for pool boiling" . | NUCLEAR ENGINEERING AND TECHNOLOGY 54 . 1 (2022) : 61-71 . |
| APA | Wang, Chenglong , Li, Panxiao , Zhang, Dalin , Tian, Wenxi , Qiu, Suizheng , Su, G. H. et al. Experimental study on the influence of heating surface inclination angle on heat transfer and CHF performance for pool boiling . | NUCLEAR ENGINEERING AND TECHNOLOGY , 2022 , 54 (1) , 61-71 . |
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