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学者姓名:苏光辉

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< Page ,Total 29 >
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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Effects of nitrogen and carbon monoxide on the detonation of hydrogen-air gaseous mixtures EI
期刊论文 | 2019 , 343 , 1-10 | Nuclear Engineering and Design
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Abstract :

Nitrogen inhibition is considered as a mitigation measure against chemically sensitive mixtures in industry and carbon monoxide is possibly generated during molten corium concrete interaction in a severe accident of nuclear power plants. To study the effects of the N2 and CO on the detonation of H2-air mixtures, a detonation facility of 78 mm inner diameter and 10 m length is set up. Key measured parameters includes detonation cell size, flame velocity, lean and rich detonation limits. All the measured detonation parameters are theoretically predicted by CJ theory or one-dimensional ZND model. Detailed chemical kinetics mechanism for H2-air and H2-CO-air mixtures is coupled with reactive Euler equations. Experiments and theoretically analysis has been performed mostly at 0.101 MPa and 293 K. Results shows that N2 increases the detonation cell size and narrows down the detonable range significantly. Especially when N2 concentration is more than 43%, all mixtures are unable to detonate. Moreover, when the added N2 concentration is larger than 20%, detonation velocity does not increase with hydrogen concentration for rich mixtures. The above effects of N2 is explained by the replacement of O2 with N2 and by the weak chemical reaction of N2. CO significantly decreases the cell size of lean H2-air mixtures and increased the cell size for rich H2-air mixtures. For lean H2-air mixtures, the added CO linearly decreases the lean detonation limit, thus increasing the possibility of detonation to occur in the severe accidents. Therefore, the contribution of CO must be carefully considered in the safety assessment. Results also shows that pure CO is difficult to detonate in the air, while small quantity of hydrogen can significantly enhance the rate of CO oxidation reactions. © 2018 Elsevier B.V.

Keyword :

Detailed chemical kinetic Detonation cell sizes Detonation parameter Hydrogen concentration Measured parameters Mitigation measures Nitrogen inhibition Severe accident

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GB/T 7714 Chen, Yongzheng , Liu, Bo , Zhang, Y.P. et al. Effects of nitrogen and carbon monoxide on the detonation of hydrogen-air gaseous mixtures [J]. | Nuclear Engineering and Design , 2019 , 343 : 1-10 .
MLA Chen, Yongzheng et al. "Effects of nitrogen and carbon monoxide on the detonation of hydrogen-air gaseous mixtures" . | Nuclear Engineering and Design 343 (2019) : 1-10 .
APA Chen, Yongzheng , Liu, Bo , Zhang, Y.P. , Zhang, D.L. , Revankar, Shripad T. , Tian, W.X. et al. Effects of nitrogen and carbon monoxide on the detonation of hydrogen-air gaseous mixtures . | Nuclear Engineering and Design , 2019 , 343 , 1-10 .
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Conceptual design and coupled neutronic/thermal-hydraulic/mechanical research of the supercritical water cooled ceramic blanket for CFETR EI
期刊论文 | 2019 , 138 , 272-281 | Fusion Engineering and Design
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Abstract :

A conceptual structure design of the supercritical water cooled ceramic blanket has been proposed for phase-II of Chinese Fusion Engineering Test Reactor (CFETR). In this work, the coupled neutronic/thermal-hydraulic/mechanical analyses were performed specifically for the optimized blanket concept as well as the design process. Firstly, the 3D-1D-3D coupling approach was applied in the neutronic and thermal-hydraulic calculations after the materials were carefully selected. The different results of the 3D model and 1D model have been compared and analyzed. Secondly, the detailed design of the typical outboard equatorial blanket was carried out and introduced based on the coupling analysis. Then, the neutronic, thermal-hydraulic and mechanical characteristics of the optimized blanket were studied and verified. The 3D neutronic calculations of the blanket indicated the tritium breeding ratio (TBR) is 1.21, which can meet the requirement for tritium self-sufficiency. The thermal-hydraulic calculations proved that all the involved materials can be effectively cooled to their allowable temperatures with the coolant temperature reaching 500 °C. According to thermo-mechanics calculation results, the blanket structure (first wall, caps, coolant pips, etc.) can certainly sustain the structure stress as well as the thermal stress in steady state. The analyses show good performances of the blanket and prove the feasibility of the conceptual design preliminarily. © 2018 Elsevier B.V.

Keyword :

Blanket designs CFETR Coupled analysis Hermo-mechanical Neutronics Thermal hydraulics

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GB/T 7714 Cheng, Jie , Wu, Yingwei , Cui, Shijie et al. Conceptual design and coupled neutronic/thermal-hydraulic/mechanical research of the supercritical water cooled ceramic blanket for CFETR [J]. | Fusion Engineering and Design , 2019 , 138 : 272-281 .
MLA Cheng, Jie et al. "Conceptual design and coupled neutronic/thermal-hydraulic/mechanical research of the supercritical water cooled ceramic blanket for CFETR" . | Fusion Engineering and Design 138 (2019) : 272-281 .
APA Cheng, Jie , Wu, Yingwei , Cui, Shijie , Su, G.H. , Chen, Yi-tung . Conceptual design and coupled neutronic/thermal-hydraulic/mechanical research of the supercritical water cooled ceramic blanket for CFETR . | Fusion Engineering and Design , 2019 , 138 , 272-281 .
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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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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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Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME In order to enhance the inherent safety of sodium-cooled fast reactors, innovative hydraulically suspended absorber rod (HSR) passive shut-down system have been proposed for China demonstration fast reactor. In this study, based on the functional and performance requirements, a full-scale experimental setup has been designed and fabricated for the analysis of the HSR as applied to the prototype reactor. The main characteristic of the test facility is the actuation of the mobile safety rod is triggered by coolant flow rate decrease in the primary loop below half the nominal value and then the rod inserts into the stationary sleeve by gravity. The objective is to investigate the dynamic performance of HSR and establish the laws of its movement at lowering the flow rate modeling the coastdown of primary circulating pump. A series of tests have been performed, including start-up, steady-state operation, loss of flow accident, sensitivity analysis and reliability test. This study also focused on the effect of various factors on scram time, the effect of pump coasting time, rod weight, gap between rod and guide tube, bypass holes, cone angle of rod, flow rate and fluid temperature are analyzed. The experimental results demonstrate the functionality and reliability of the HSR, which would lay foundation for further optimization design.

Keyword :

Dynamic performance Experiment Passive shutdown system SFR

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GB/T 7714 Song, Jian , Wu, Yingwei , Tian, Wenxi et al. Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor [C] . 2018 .
MLA Song, Jian et al. "Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor" . (2018) .
APA Song, Jian , Wu, Yingwei , Tian, Wenxi , Qiu, Suizheng , Su, Guanghui . Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor . (2018) .
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Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k − ε model (SKE) and the realizable k − ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.

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GB/T 7714 Wang, Shibao , Zhang, Dalin , Wang, Chenglong et al. Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR [C] . 2018 .
MLA Wang, Shibao et al. "Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR" . (2018) .
APA Wang, Shibao , Zhang, Dalin , Wang, Chenglong , Song, Ping , Chen, Jing , Qiu, Suizheng et al. Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR . (2018) .
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Transient thermal behaviors of SBO accident for a 200MW OFNP under heaving motion conditions Scopus
会议论文 | 2018 , 4
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Copyright © 2018 ASME For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Keyword :

Heaving conditions Offshore floating nuclear plants RELAP5 SBO accident analysis

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GB/T 7714 Yan, Qiqi , Luo, Simin , Zhang, Yapei et al. Transient thermal behaviors of SBO accident for a 200MW OFNP under heaving motion conditions [C] . 2018 .
MLA Yan, Qiqi et al. "Transient thermal behaviors of SBO accident for a 200MW OFNP under heaving motion conditions" . (2018) .
APA Yan, Qiqi , Luo, Simin , Zhang, Yapei , Liu, Limin , Su, Guanghui , Qiu, Suizheng . Transient thermal behaviors of SBO accident for a 200MW OFNP under heaving motion conditions . (2018) .
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