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学者姓名:田文喜

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Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5 SCIE
期刊论文 | 2019 , 30 (1) | NUCLEAR SCIENCE AND TECHNIQUES
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Abstract :

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5 (RELAP5), with the implementation of sodium-potassium eutectic alloy (NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal-hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including 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, moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-II system model, an electrical ground test of the fully integrated TOPAZ-II system, the V-71 unit, is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal-hydraulic characteristics of TOPAZ-II are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

Keyword :

Thermal-hydraulic analysis RELAP5 modification Space nuclear reactor TOPAZ-II

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GB/T 7714 Wang, Cheng-Long , Liu, Tian-Cai , Tang, Si-Miao et al. Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5 [J]. | NUCLEAR SCIENCE AND TECHNIQUES , 2019 , 30 (1) .
MLA Wang, Cheng-Long et al. "Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5" . | NUCLEAR SCIENCE AND TECHNIQUES 30 . 1 (2019) .
APA Wang, Cheng-Long , Liu, Tian-Cai , Tang, Si-Miao , Tian, Wen-Xi , Qiu, Sui-Zheng , Su, Guang-Hui . Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5 . | NUCLEAR SCIENCE AND TECHNIQUES , 2019 , 30 (1) .
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Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5 EI SCIE
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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Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment EI SCIE
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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Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate EI SCIE
期刊论文 | 2019 , 149 , 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 , 149 : 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 149 (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 , 149 , 881-888 .
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Theoretical investigation of two-phase flow instability between parallel channels of natural circulation in rolling motion SCIE
期刊论文 | 2019 , 343 , 257-268 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

In present work, the two-phase flow instability between rectangular parallel channels of natural circulation under static and rolling conditions was coupled studied theoretically. Models including two-phase flow instability, natural circulation system components, and the additional force were established in combination based on the homogenous model. A computational program was written in FORTRAN language which was solved by Gear mull-value method by using control volume integrating method. The program was validated with experiments, and the results matched well with the experiment data. The marginal stability boundary (MSB) maps under different parameters were obtained by using nondimensional numbers N-sub and N-pch. The influence of different kinds of pressure drop, inlet subcooling temperature of heating channels, system pressure, valve resistance, venturi flowmeters resistance, structure height, rolling condition, and the interaction effect of natural circulation and two-phase flow instability between rectangular parallel channels were analyzed. The results show that with the increase in system pressure and venturi flowmeters resistance, the system stability of natural circulation is enhanced. The influence of inlet subcooling temperature is nonlinear. The increase in valve resistance leads to the instability of system. The increase in structure height does not change system stability significantly. The rise in rolling angle and period both reduce the system stability.

Keyword :

Two-phase flow instability Rolling motion Natural circulation Parallel channels Marginal stability boundary

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GB/T 7714 Wang, Xiaoyan , Tian, Wenxi , Huang, Siyang et al. Theoretical investigation of two-phase flow instability between parallel channels of natural circulation in rolling motion [J]. | NUCLEAR ENGINEERING AND DESIGN , 2019 , 343 : 257-268 .
MLA Wang, Xiaoyan et al. "Theoretical investigation of two-phase flow instability between parallel channels of natural circulation in rolling motion" . | NUCLEAR ENGINEERING AND DESIGN 343 (2019) : 257-268 .
APA Wang, Xiaoyan , Tian, Wenxi , Huang, Siyang , Chen, Ronghua , Zhang, Dalin , Qiu, Suizheng et al. Theoretical investigation of two-phase flow instability between parallel channels of natural circulation in rolling motion . | NUCLEAR ENGINEERING AND DESIGN , 2019 , 343 , 257-268 .
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Experimental study of the steam condensate dripping behavior on the containment dome SCIE
期刊论文 | 2019 , 346 , 131-139 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

Under accident conditions of CAP1400, recycling of condensate film from the inner containment wall is a significant aspect for the water level maintenance of the in-containment refueling water storage tank, whilst dripping from the containment dome is the dominant factor that causing the condensate loss. Therefore, experimental investigations on the dripping phenomena were carried out in this study. A pressure vessel was set to simulate the condensing atmosphere in the containment, in which a 1.5 x 0.6 m(2) rotatable test section was suspended with similar surface condition as CAP1400. Experimental results show that the condensate flow patterns could be divided into four types. It was found that dripping was triggered by the condensate mass flow flux exceeding the critical value on an unobstructed condensing surface. Meanwhile, the dripping fraction increases with the difference between condensate mass flow flux and the critical value. Besides, the effects of inclination, bulk pressure, air concentration etc. on dripping were experimental analyzed. In general, this study hopes to provide data support and theoretical guidance for the further studies of the condensate loss under accident conditions.

Keyword :

Containment dome Dripping Steam condensation Condensate film

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GB/T 7714 Chen, Ronghua , Zhang, Penghui , Ma, Pan et al. Experimental study of the steam condensate dripping behavior on the containment dome [J]. | NUCLEAR ENGINEERING AND DESIGN , 2019 , 346 : 131-139 .
MLA Chen, Ronghua et al. "Experimental study of the steam condensate dripping behavior on the containment dome" . | NUCLEAR ENGINEERING AND DESIGN 346 (2019) : 131-139 .
APA Chen, Ronghua , Zhang, Penghui , Ma, Pan , Tan, Bing , Wang, Zhangli , Zhang, Di et al. Experimental study of the steam condensate dripping behavior on the containment dome . | NUCLEAR ENGINEERING AND DESIGN , 2019 , 346 , 131-139 .
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Effect of stratified interface instability on thermal focusing effect in two-layer corium pool EI SCIE
期刊论文 | 2019 , 133 , 359-370 | International Journal of Heat and Mass Transfer
WoS CC Cited Count: 1
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Abstract :

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 , 133 : 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 133 (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 , 133 , 359-370 .
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Experimental investigations on single-phase convection and two-phase flow boiling heat transfer in an inclined rod bundle EI SCIE
期刊论文 | 2019 , 148 , 340-351 | Applied Thermal Engineering
WoS CC Cited Count: 1
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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 , 148 : 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 148 (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 , 148 , 340-351 .
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Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 EI SCIE
期刊论文 | 2019 , 112 , 63-74 | Progress in Nuclear Energy
WoS CC Cited Count: 1
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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 SCIE
期刊论文 | 2019 , 149 , 578-587 | Applied Thermal Engineering
WoS CC Cited Count: 2
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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 , 149 : 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 149 (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 , 149 , 578-587 .
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