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学者姓名:田文喜

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Experimental investigations on single-phase convection and two-phase flow boiling heat transfer in an inclined rod bundle EI
期刊论文 | 2019 , 340-351 | Applied Thermal Engineering
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Abstract :

Single-phase convection and steam-water two-phase flow boiling heat transfer experiments in an electrically heated 4 × 25 staggered inclined rod bundle were carried out for the following range: 15 kg m−2 s−1 ≤ G ≤ 50 kg m−2 s−1, 20.0 kW m−2 ≤ q ≤ 55.0 kW m−2, 0.05 ≤ xout ≤ 0.79 and 117 kPa ≤ Pin ≤ 260 kPa. Single-phase convective results show that the independence principle can be applied to the case that all tubes were heated in inclined rod bundles. With respect to two-phase results, the local flow boiling heat transfer coefficient increases from the bottom row to the eleventh row and then can be considered as a constant value from the eleventh row to the top row. An increasing heat flux results in a decrease of the flow boiling heat transfer coefficient. However, no significant effects of mass velocity and quality were observed. The inclination angle has a small effect on the flow boiling heat transfer coefficient at low heat fluxes. However, when heat flux is high, the flow boiling heat transfer coefficient in the inclined rod bundle is the minimum while that in the vertical rod bundle is the maximum compared with that in the horizontal rod bundle. In a general, a Chen-type correlation was developed to predict 98.5 percent of local flow boiling heat transfer coefficient data in the inclined rod bundle with a maximum deviation of ±20%. © 2018 Elsevier Ltd

Keyword :

Experimental investigations Flow boiling Flow boiling heat transfer Inclination angles Rod bundles Single-phase convections Steam water two phase flow Two-phase flow boiling

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GB/T 7714 Zhang, K. , Hou, Y.D. , Tian, W.X. et al. Experimental investigations on single-phase convection and two-phase flow boiling heat transfer in an inclined rod bundle [J]. | Applied Thermal Engineering , 2019 : 340-351 .
MLA Zhang, K. et al. "Experimental investigations on single-phase convection and two-phase flow boiling heat transfer in an inclined rod bundle" . | Applied Thermal Engineering (2019) : 340-351 .
APA Zhang, K. , Hou, Y.D. , Tian, W.X. , Zhang, Y.P. , Su, G.H. , Qiu, S.Z. . Experimental investigations on single-phase convection and two-phase flow boiling heat transfer in an inclined rod bundle . | Applied Thermal Engineering , 2019 , 340-351 .
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Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment EI
期刊论文 | 2019 , 111 , 174-182 | Progress in Nuclear Energy
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Abstract :

A new simple core degradation experiment will be conducted to investigate the distribution of mass and energy during a core degradation process, and the International Standard Problem (ISP) No.31 is chosen as a pre-simulation for the new experiment. The pre-simulation has been conducted using the widely accepted severe accident analysis software MELCOR. Firstly, the numerical analysis model and oxidation model are described, and all the input parameters are in accord with experiment conditions. Then, numerical results are validated by experimental measurements and SCDAP/RELAP5 results. Simulations results agree well with measured data. It is indicated that MELCOR has the capability of predicting the behaviors of fuel elements in reflood correctly. Finally, the visually spatial temperature distribution is obtained by TECPLOT, and the behaviors of molten fuel elements are directly reflected. The evolution of peak temperature in fuel rods during the experiment period can be visible. The peak temperature firstly appeared in outer heated rods of the fourth ring and later showed in a heated fuel rod of the second ring. The behaviors of molten fuel elements are visible in the figures of spatial temperature distribution, and it does help researchers to understand the migration behaviors of molten material accompanied by effective mitigation measures for a severe accident. © 2018 Elsevier Ltd

Keyword :

CORA Experiment condition International standards MELCOR Mitigation measures Numerical analysis models Peak temperatures Spatial temperature distribution

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GB/T 7714 Feng, Tangtao , Tian, Wenxi , Song, Ping et al. Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment [J]. | Progress in Nuclear Energy , 2019 , 111 : 174-182 .
MLA Feng, Tangtao et al. "Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment" . | Progress in Nuclear Energy 111 (2019) : 174-182 .
APA Feng, Tangtao , Tian, Wenxi , Song, Ping , Wang, Jun , Wang, Mingjun , Li, Longze et al. Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment . | Progress in Nuclear Energy , 2019 , 111 , 174-182 .
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Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 EI
期刊论文 | 2019 , 112 , 63-74 | Progress in Nuclear Energy
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Abstract :

Pressurized water reactors (PWRs) use Steam Generators (SG) to transfer the reactor core heat to the secondary loop. The SG contains thousands of heat transfer tubes and tubes plugging operation is always adopted to solve the heat transfer tubes rupture problems caused by the complex operation conditions, including thermal stress, radiation environment and material corrosion. However, tubes plugging operation has great impact on the SG hydraulic performance which leading to some safety issues in nuclear power plants (NPP). In this paper, the AP1000 SG primary side three-dimensional thermal hydraulic simulation model is built using porous media method, in which the important parameters are achieved employing the separate tube simulations. An innovative grid mark method is proposed to realize the flexible tubes plugging conditions. The tubes plugging fractions varies from 5% to 20%, and the effects of tubes plugging locations are also considered. The detailed flow fields in whole AP1000 SG primary side are achieved. The large eddies and some small vortexes are generated in the inlet side of bottom channel head. The momentum loss mechanism is revealed in the whole SG primary loop. Results show that the pressure drop in primary loop increases with the plugging fraction and varies with different tubes plugging positions. The pressure drop goes up to 492.77 KPa in case that the tubes plugging fraction reaches 20%. The tubes plugging position affects the results significantly in case that the plugging fraction is larger than 15%. This work is meaningful for the depth understanding of AP1000 SG tube plugging operations and provides a guideline for the SG maintenance strategy during the whole plant lifetime. © 2018 Elsevier Ltd

Keyword :

AP1000 Hydraulic characteristic Hydraulic performance Maintenance strategies Porous media methods Porous medium Radiation environments Thermal-hydraulic simulations

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GB/T 7714 Zhao, Xiaohan , Wang, Mingjun , Chen, Chong et al. Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 [J]. | Progress in Nuclear Energy , 2019 , 112 : 63-74 .
MLA Zhao, Xiaohan et al. "Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000" . | Progress in Nuclear Energy 112 (2019) : 63-74 .
APA Zhao, Xiaohan , Wang, Mingjun , Chen, Chong , Wang, Xi , Ju, Haoran , Tian, Wenxi et al. Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 . | Progress in Nuclear Energy , 2019 , 112 , 63-74 .
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Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle EI
期刊论文 | 2019 , 578-587 | Applied Thermal Engineering
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Abstract :

The single-phase thermal hydraulic characteristics of liquid metal sodium are very essential for the design and safety analysis of sodium-cooled fast reactor (SFR). In this paper, the pressure drop and heat transfer features of single-phase liquid sodium were experimentally investigated in a 7 rod bundle with the velocity range of 0–4 m/s, heat flux up to 120 kW/m2 and the absolute pressure range of 0–0.2 MPa. The corresponding Reynolds number ranges from 4000 to 40,000, and the Pe number varies from 0 to 340. It was found that the critical Re number for transition-turbulent flow of single-phase liquid sodium is 13,500 in the hexagonal 7-rod bundle. Then the effects of relative axial position, wall heat flux and Re number on the heat transfer were discussed, respectively. Some existing correlations in the literatures were assessed and compared with the experimental data. Results indicated that these correlations could not predict the current experiments well because of the different geometries and working fluids. The new correlations for the friction factor and Nu number calculations were proposed based on the current experimental data. For 98.5% of heat transfer data produced by the other researchers, the prediction error of the new correlation is less than 30%. For most of the experimental data, it is less than 20%, which sufficiently proves that the correlation developed in this paper could give a good prediction of the experimental data obtained by other researchers. © 2018 Elsevier Ltd

Keyword :

Different geometry Heat transfer data Liquid sodium Liquid sodium flows Rod bundles Single-phase liquids Sodium cooled fast reactors (SFR) Thermal hydraulics

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GB/T 7714 Hou, Yandong , Wang, Liu , Wang, Mingjun et al. Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle [J]. | Applied Thermal Engineering , 2019 : 578-587 .
MLA Hou, Yandong et al. "Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle" . | Applied Thermal Engineering (2019) : 578-587 .
APA Hou, Yandong , Wang, Liu , Wang, Mingjun , Zhang, Kui , Zhang, Xisi , Hu, Wenjun et al. Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle . | Applied Thermal Engineering , 2019 , 578-587 .
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Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate EI
期刊论文 | 2019 , 881-888 | Applied Thermal Engineering
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Abstract :

In order to solve the problem of non-convergence of CHF directly calculated by FLUENT under atmospheric pressure and low flow rate, a CFD methodology was proposed based on four equation drift flux model and an improved RPI wall boiling model to predict the CHF and thermal-hydraulics characteristics in the flow channel formed by the outer wall of the RPV and the inner wall of the insulation. The governing equations and flow boiling models were added into FLUENT solver, and then worked with Mixture multiphase models by user defined functions (UDFs). The developed CFD models for CHF prediction were validated by using experimental data, and the prediction results had a quite good agreement with the experimental data with deviations less than 20%. It indicated that the CFD methodology proposed in this study had a good convergence at atmospheric pressure and low flow rate. Meanwhile the CFD methodology could be qualified to predict the characteristics of CHF, and it provided a potential way to predict the CHF in the flow channel formed by the outer wall of the RPV and the inner wall of the insulation under IVR conditions of nuclear power plants. © 2018 Elsevier Ltd

Keyword :

CFD methodologies Drift flux modeling Flow boiling models Governing equations Multiphase model Numerical predictions Severe accident User Defined Functions

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GB/T 7714 Zhang, Yapei , Zhang, Rui , Tian, Wenxi et al. Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate [J]. | Applied Thermal Engineering , 2019 : 881-888 .
MLA Zhang, Yapei et al. "Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate" . | Applied Thermal Engineering (2019) : 881-888 .
APA Zhang, Yapei , Zhang, Rui , Tian, Wenxi , Su, G.H. , Qiu, Suizheng . Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate . | Applied Thermal Engineering , 2019 , 881-888 .
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Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5 EI
期刊论文 | 2019 , 112 , 209-224 | Progress in Nuclear Energy
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Abstract :

Nuclear power and thermionic conversion can serve as a compact, durable energy source for the space exploration and exploitation. In this paper, the modified Reactor Excursion and Leak Analysis Program5 (RELAP5) with the implement of NaK-78 eutectic alloy (78%K and 22%Na) properties and heat transfer correlations is adopted to analyze the thermal-hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including the thermionic fuel elements (TFEs), reactor core, radiator, coolant loop and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector, reflector, moderator and the reactivity insertion effects of control drums and safety drums are considered. The steady state condition and three severe transient accidents including reactivity insertion accident (RIA), loss of flow accident (LOFA) and loss of coolant accident (LOCA), are simulated and analyzed. The steady state calculated results agree well with the design values. During the three accidents, the moderator plays a dominant role in the positive temperature reactivity feedback. The coolant has at least 50 K temperature margin to the boiling point. The fuel and TFE components are all below their melting temperature. The progress of these accidents provide relatively sufficient time for operator's response. The calculation results prove that the reactor is a safe and reliable system. © 2018

Keyword :

Heat transfer correlation Reactivity insertion RELAP5 modification Space nuclear reactors Steady-state condition Thermal-hydraulic analysis Thermionic fuel elements Transient thermal-hydraulic

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GB/T 7714 Tang, Simiao , Sun, Hao , Wang, Chenglong et al. Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5 [J]. | Progress in Nuclear Energy , 2019 , 112 : 209-224 .
MLA Tang, Simiao et al. "Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5" . | Progress in Nuclear Energy 112 (2019) : 209-224 .
APA Tang, Simiao , Sun, Hao , Wang, Chenglong , Tian, Wenxi , Qiu, Suizheng , Su, G.H. et al. Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5 . | Progress in Nuclear Energy , 2019 , 112 , 209-224 .
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Effect of stratified interface instability on thermal focusing effect in two-layer corium pool EI
期刊论文 | 2019 , 359-370 | International Journal of Heat and Mass Transfer
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In this study, the effect of the stratified interface instability on the thermal focusing effect in two-layer corium pool were investigated by numerical simulations performed with CFD code Fluent. The Rayleigh numbers (Ra′) obtained in this study range from 109 to 1015. By setting different decay heat power and turbulence intensity, crust of different melting degree at stratified interface can be obtained. Through the comparison of the corium pools with the crust of different melting degree, the differences of temperature distribution and boundary heat flux distribution are obtained. The coupling mechanism of two layers of corium pools and a new criterion for the occurrence of stratified interface instability are also presented. The results show that when the crust is slightly damaged, the thermal focusing effect is intensified by the reduced thermal resistance due to the crust failure at the interface and the unevenness of the thickness of the crust on the side wall of the metal layer, and if the crust is highly damaged, the thermal focusing effect is weaken by the melting of the crust at the wall of the lower head. The results of this study can provide reference for reactor IVR (In-Vessel Retention) safety analysis and optimization design. © 2018 Elsevier Ltd

Keyword :

Boundary heat flux distribution Coupling mechanism Interface instability Melting of the crusts Optimization design Thermal focusing Turbulence intensity Two-layer

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GB/T 7714 Ge, K. , Zhang, Y.P. , Tian, W.X. et al. Effect of stratified interface instability on thermal focusing effect in two-layer corium pool [J]. | International Journal of Heat and Mass Transfer , 2019 : 359-370 .
MLA Ge, K. et al. "Effect of stratified interface instability on thermal focusing effect in two-layer corium pool" . | International Journal of Heat and Mass Transfer (2019) : 359-370 .
APA Ge, K. , Zhang, Y.P. , Tian, W.X. , Su, G.H. , Qiu, S.Z. . Effect of stratified interface instability on thermal focusing effect in two-layer corium pool . | International Journal of Heat and Mass Transfer , 2019 , 359-370 .
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Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor Scopus
会议论文 | 2018 , 9 | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Copyright © 2018 ASME In sodium-cooled fast reactor (SFR), thermal gradient is the paramount factor of assembly transient bowing, that may cause great reactivity change, accelerate wrapper vibration wear, hindering the motion of control/shutdown rods, or worse yet, threatening the integrity of assemblies. However, because of the complexity of multi-assembly contact and interaction problem, it is difficult to assess the impact of core deformation on reactor performance safety. The Core Assembly Deformation Test Facility (CADTF) is designed to perform a series of thermal bowing tests by Xi`an Jiao Tong University (XJTU) to investigate the core deformation behaviors under thermal gradient. In this paper, a finite element model was established to simulate the mechanical response of single assembly under different flat-to-flat thermal gradient. The single assembly restrained bowing test performed in CADTF is chosen to validate the model. In the model, the measured temperature distribution as well as temperature-dependent elastoplastic and thermal expansion properties were taken into consideration. To ensure the model reliability, iterative computation is conducted by adjusting the friction coefficient of the load pads to match the calculated and measured contact force. According to the results, it can be seen that the three-dimensional displacement of assembly shows relatively good agreement with the experimental data. Therefore, it can be concluded that the model is capable of performing core deformation analysis for SFR.

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GB/T 7714 Tian, Wenxi , Qiu, Suizheng . Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor [C] . 2018 .
MLA Tian, Wenxi et al. "Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor" . (2018) .
APA Tian, Wenxi , Qiu, Suizheng . Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor . (2018) .
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Pressure drop experiments of liquid sodium flowing in a 7-rod bundle Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME Pressure drop experiments was conducted for liquid sodium in an electrically heated 7-rod bundle. The electrically heated 7-rod bundle was placed in a hexagonal tube. In the experiment,the heat flux ranges from 0~300 kw·m-2,mass velocity from 40~450 kg·m-2·s-1, system pressure from 10~200 KPa and the average temperature of liquid sodium from 350~650℃.The effects of the heat flux, system pressure and the average temperature of liquid sodium on the pressure drop was in-depth analyzed. A new correlation for pressure drop was developed based on the experimental data of liquid sodium in a 7-rod bundle.

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GB/T 7714 Tian, Wenxi . Pressure drop experiments of liquid sodium flowing in a 7-rod bundle [C] . 2018 .
MLA Tian, Wenxi . "Pressure drop experiments of liquid sodium flowing in a 7-rod bundle" . (2018) .
APA Tian, Wenxi . Pressure drop experiments of liquid sodium flowing in a 7-rod bundle . (2018) .
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Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation Scopus
会议论文 | 2018 , 5
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Copyright © 2018 ASME. In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.

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GB/T 7714 Tian, Wenxi . Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation [C] . 2018 .
MLA Tian, Wenxi . "Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation" . (2018) .
APA Tian, Wenxi . Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation . (2018) .
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