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学者姓名:田文喜

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CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger EI SCIE Scopus
期刊论文 | 2018 , 337 , 27-37 | NUCLEAR ENGINEERING AND DESIGN
WoS CC Cited Count: 1 SCOPUS Cited Count: 1
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Abstract :

Passive residual heat removal heat exchanger (PRHR HX) is a key equipment in advanced passive safety pressurized water reactors (PWRs), such as AP1000. Immerged in the In-containment refueling water storage tank (IRWST), the C-shape PRHR HX removes the residual heat using natural convection and boiling heat transfer during postulated accidents. Therefore, the safety operation of the PRHR HX is very important for the nuclear power plant (NPP). In this paper, three dimensional CFD simulation of the secondary side fluid flow and heat transfer of the PRHR HX is performed. The drift flux model is used to simulate the two phase flow phenomenon in the IRWST. The tube region is modeled by the porous media approach. The flow resistance in the tube region is calculated by empirical pressure drop correlation and two phase flow multiplier. The heat transfer rate from the primary side to the secondary side fluid is also evaluated by widely used heat transfer empirical correlations. Additional source terms corresponding to the flow resistance and removed heat in the tube region are added to the momentum and energy equation, respectively. The governing equations are solved by the commercial CFD package FLUENT. Three dimensional distributions of the fluid velocity, temperature and void fraction are obtained. The heat transfer characteristics of the tube bundle is analyzed.

Keyword :

CFD Porous media approach Drift flux model PRHR HX

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GB/T 7714 Ge, Jian , Tian, Wenxi , Qiu, Suizheng et al. CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger [J]. | NUCLEAR ENGINEERING AND DESIGN , 2018 , 337 : 27-37 .
MLA Ge, Jian et al. "CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger" . | NUCLEAR ENGINEERING AND DESIGN 337 (2018) : 27-37 .
APA Ge, Jian , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger . | NUCLEAR ENGINEERING AND DESIGN , 2018 , 337 , 27-37 .
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CFD investigation on thermal hydraulics of the passive residual heat removal heat exchanger (PRHR HX) EI SCIE Scopus
期刊论文 | 2018 , 327 , 139-149 | NUCLEAR ENGINEERING AND DESIGN
WoS CC Cited Count: 1
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Abstract :

Passive residual heat removal (PRHR) system is a very important component of the passive safety systems in advanced passive safety pressurizer water reactors (PWRs) such as AP1000. The passive residual heat removal heat exchanger (PRHR HX) is a C-shape tube bundle heat exchanger immerged in the In-containment refueling water storage tank (IRWST) which removes the core decay heat during the accident transients. The performance of the PRHR HX is significant for the safety of the nuclear power plant (NPP). In this paper, the thermal hydraulics characteristics of the PRHR HX in the IRWST is analyzed using computational fluid dynamics (CFD). The tube region is modeled by the porous media approach along with the distributed resistance method. Heat transfer from the primary side fluid inside the tube to the secondary side fluid in the IRWST is considered. The simulation is carried out by the commercial CFD package FLUENT. The calculation of the flow resistance and heat transfer in the tube region is implemented using the User Defined Functions (UDF) in FLUENT based on the local flow conditions. Three dimensional distributions of the fluid velocity and temperature in the IRWST are obtained and thermal stratification is observed. The PRHR HX heat transfer capacity and the primary side fluid temperature distribution inside tubes are analyzed.

Keyword :

Passive heat removal CFD Porous media approach PRHR HX

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GB/T 7714 Ge, Jian , Tian, Wenxi , Qiu, Suizheng et al. CFD investigation on thermal hydraulics of the passive residual heat removal heat exchanger (PRHR HX) [J]. | NUCLEAR ENGINEERING AND DESIGN , 2018 , 327 : 139-149 .
MLA Ge, Jian et al. "CFD investigation on thermal hydraulics of the passive residual heat removal heat exchanger (PRHR HX)" . | NUCLEAR ENGINEERING AND DESIGN 327 (2018) : 139-149 .
APA Ge, Jian , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . CFD investigation on thermal hydraulics of the passive residual heat removal heat exchanger (PRHR HX) . | NUCLEAR ENGINEERING AND DESIGN , 2018 , 327 , 139-149 .
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Improvement of RELAP5 code for density wave instability analysis in parallel narrow rectangular channels EI SCIE Scopus
期刊论文 | 2018 , 122 , 241-255 | ANNALS OF NUCLEAR ENERGY
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Abstract :

As one of the most common types of oscillatory instability, density wave oscillation (DWO) is of practical importance in the general analysis of two-phase boiling systems. Due to the confinement effect of narrow space, the two-phase phenomena in thin rectangular channels are probably different from those in conventional pipes with large diameters. As the widely accepted system analysis code, RELAP5 has the capability to simulate the DWO phenomena in narrow channels, but the accuracy is limited under the condition of different pressures, flow rates, or subcoolings. In this paper, modifications for related constitutive models are proposed based on single-variable sensitivity analysis. Two-phase friction model and nucleate boiling heat transfer model are selected as the most effective models to improve the predictability of RELAP5. The influence of both models mentioned above is analyzed and It reveals that they are related to two-phase pressure drop and the vapor void fraction distribution along the channel, respectively. The modified calculation is assessed against the experimental results under the wide range of pressure 1-10 MPa, and these comparisons are performed in the map of stability boundary with two dimensionless parameters, phase change number Npch and subcooling number Nsub. Then the heating power and the period of flow oscillation under threshold condition are assessed in detail. The results show that a better agreement can be obtained between the modified calculations and the measured experimental data. (C) 2018 Elsevier Ltd. All rights reserved.

Keyword :

Density wave instability Narrow rectangular channel Stability boundary RELAP5 modification

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GB/T 7714 Lian, Qiang , Tian, Wenxi , Qiu, Suizheng et al. Improvement of RELAP5 code for density wave instability analysis in parallel narrow rectangular channels [J]. | ANNALS OF NUCLEAR ENERGY , 2018 , 122 : 241-255 .
MLA Lian, Qiang et al. "Improvement of RELAP5 code for density wave instability analysis in parallel narrow rectangular channels" . | ANNALS OF NUCLEAR ENERGY 122 (2018) : 241-255 .
APA Lian, Qiang , Tian, Wenxi , Qiu, Suizheng , Su, Guanghui . Improvement of RELAP5 code for density wave instability analysis in parallel narrow rectangular channels . | ANNALS OF NUCLEAR ENERGY , 2018 , 122 , 241-255 .
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Flow Instability in Parallel Double Channels under Motion Condition Based on Modified RELAP5 EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (5) , 875-880 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
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Abstract :

As the marine nuclear power device is operated widely, the thermal hydraulic phenomenon caused by motion conditions is considered significantly. Since many parallel channels exist in the reactor core, the study of flow instability in parallel double channels under single and coupled motion conditions was carried out by modifying the source term in momentum equation adopted in RELAP5 code. The marginal stability boundary (MSB) curves under static and motion conditions were compared, and it is shown that the effect of different motion conditions on flow instability is very small under forced circulation. With the same subcooled number (Nsub), the difference between phase change numbers (Npch) of static condition and motion condition is under 2%. © 2018, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Flow instabilities Forced circulations Marginal stability Marine nuclear power Momentum equation Motion conditions Static conditions Thermal-hydraulic phenomena

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GB/T 7714 Lian, Qiang , Liu, Di , Tian, Wenxi et al. Flow Instability in Parallel Double Channels under Motion Condition Based on Modified RELAP5 [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (5) : 875-880 .
MLA Lian, Qiang et al. "Flow Instability in Parallel Double Channels under Motion Condition Based on Modified RELAP5" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 5 (2018) : 875-880 .
APA Lian, Qiang , Liu, Di , Tian, Wenxi , Qiu, Suizheng , Su, Guanghui . Flow Instability in Parallel Double Channels under Motion Condition Based on Modified RELAP5 . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (5) , 875-880 .
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Study on safety boundary of flow instability and CHF for parallel channels in motion EI SCIE Scopus
期刊论文 | 2018 , 335 , 219-230 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

The safety boundary of flow instability and critical heat flux(CHF) for parallel channels was theoretically investigated in static and motion. For flow instability, a parallel-channel instability mechanistic model was adopted. For CHF, the liquid sublayer dryout mechanism model was used. Also, a unified form of additional forces caused by motion was derived for both flow instability and CHF model. An in-house code was developed combining the flow instability model and the CHF model in motion. The safety boundary of twin rectangular parallel channels with length of 40 mm, width of 2 mm and height of 1 m was calculated in static, inclination, heaving, pitching and rolling motions. The results show that the safety boundary consists of CHF lines and instability boundary. The heating power of the parallel channels is limited by flow instability or CHF depending on the mass flux. All the motions have little influence on the instability boundary. The effects of longitudinal inclination, heaving and pitching motions on the safety boundary are no more than 1%. However, the CHF is obviously reduced in transverse inclination due to flow maldistribution and in rolling motion due to pulsating flow.

Keyword :

Motion Flow instability Parallel channel Critical heat flux

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GB/T 7714 Liu, Di , Tian, Wenxi , Xi, Mengmeng et al. Study on safety boundary of flow instability and CHF for parallel channels in motion [J]. | NUCLEAR ENGINEERING AND DESIGN , 2018 , 335 : 219-230 .
MLA Liu, Di et al. "Study on safety boundary of flow instability and CHF for parallel channels in motion" . | NUCLEAR ENGINEERING AND DESIGN 335 (2018) : 219-230 .
APA Liu, Di , Tian, Wenxi , Xi, Mengmeng , Chen, Ronghua , Qiu, Suizheng , Su, G. H. . Study on safety boundary of flow instability and CHF for parallel channels in motion . | NUCLEAR ENGINEERING AND DESIGN , 2018 , 335 , 219-230 .
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Experimental investigation of entrainment effect on the countercurrent flow in the Hot Leg and Pressurizer Surge Line assembly of third-generation passive nuclear reactors EI SCIE Scopus
期刊论文 | 2018 , 335 , 326-338 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

The effect of entrainment from the Hot Leg (HL) to the Pressurizer Surge Line (PZR-SL) on the characteristics of Countercurrent Flow (CCF), is experimentally studied using air-water in a transparent test section, 1/4 scaled-down model of the HL and PZR-SL assembly of AP1000. The typical CCF behaviors are recorded with video cameras for understanding of mechanisms of Countercurrent Flow Limitation (CCFL) affected by entrainment. Two entrainment flow regimes, stratified-flow entrainment and slug-flow entrainment, are found to occur under certain HL pipe water levels and air velocity conditions. The entrainment effect on the characteristics of CCF in the WA section is analyzed in detail based on the two-phase air and water Kutateladze numbers, the transient hydraulic parameters and the flow behaviors. The stratified-flow entrainment does not always influence CCFL extent, unless the entrainment is strong enough and reverse flow of water occurs in the PZR-SL pipe. However, the slug-flow entrainment markedly influences CCFL extent. Two empirical correlations are developed respectively for high HL pipe water level situation with slugging entrainment and for medium and low HL pipe water level situation without slugging entrainment. The present data are compared and verified with the Small-Break Loss-of-Coolant Accidents (SB-LOCAs) tests data on AP600 Scaled Integral Test Facility (APEX). The present experimental research provide insight on the CCF flow phenomena in the HL and PZR-SL assembly for the third-generation passive nuclear reactors during the Fourth Stage Automatic Depressurization Stage (ADS-4) of SBLOCAs and support theoretical model development and the validation of computational codes.

Keyword :

Hot Leg and Pressurizer Surge Line Assembly Visual air-water experiments Third-generation passive nuclear reactors Entrainment effect Countercurrent Flow Limitation

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GB/T 7714 Yu, Jiangtao , Zhang, Dalin , Shi, Leitai et al. Experimental investigation of entrainment effect on the countercurrent flow in the Hot Leg and Pressurizer Surge Line assembly of third-generation passive nuclear reactors [J]. | NUCLEAR ENGINEERING AND DESIGN , 2018 , 335 : 326-338 .
MLA Yu, Jiangtao et al. "Experimental investigation of entrainment effect on the countercurrent flow in the Hot Leg and Pressurizer Surge Line assembly of third-generation passive nuclear reactors" . | NUCLEAR ENGINEERING AND DESIGN 335 (2018) : 326-338 .
APA Yu, Jiangtao , Zhang, Dalin , Shi, Leitai , Tian, Wenxi , Su, G. H. , Qiu, S. Z. . Experimental investigation of entrainment effect on the countercurrent flow in the Hot Leg and Pressurizer Surge Line assembly of third-generation passive nuclear reactors . | NUCLEAR ENGINEERING AND DESIGN , 2018 , 335 , 326-338 .
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Experimental research on the characteristics of steam-water counter-current flow in the Pressurizer Surge Line assembly EI SCIE Scopus
期刊论文 | 2018 , 96 , 180-191 | EXPERIMENTAL THERMAL AND FLUID SCIENCE
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Abstract :

To investigate Counter-Current Flow (CCF) characteristics in the Pressurizer Surge Line (PZR SL) assembly of the large advanced passive nuclear reactors, steam-water CCF experiments are carried out on a stainless steel test section, which is a 1:4 scaled-down model of the AP1000 PZR SL assembly. The inner diameter of the test PZR SL pipe is 90 mm, which also belongs to the large-diameter pipes, as same as the prototype PZR SL pipe, in the field of Counter-Current Flow Limitation (CCFL) researches. The present steam-water CCFL experiments are conducted under normal pressure and saturated temperature, with the PZR simulator collapse water level ranging from 350 to 900 mm. CCFL becomes severer at higher steam flow rate, and the most limiting CCFL effect locates between the PZR simulator bottom head and the vertical part of PZR SL pipe. The onset of CCFL and zero liquid penetration (ZP) are two critical conditions in the CCF development process, dividing the process into three stages: Before-CCFL, Partial-CCFL, and CCFL-ZP. According to the local CCFL conditions, the development of the CCFL process is also divided into four-regions since onset of CCFL. The present steam-water CCF data are well normalized in terms of dimensionless Kutateladze (Ku) numbers, and a Ku-type empirical partial CCFL correlation is developed. The comparisons of the present CCF data and partial CCFL correlation with the CCF data and correlations of former experiment researches validate that the present empirical partial CCFL correlation is conservative to predict steam-water partial CCFL in the prototype PZR SL assembly of the large advanced passive nuclear reactors.

Keyword :

Partial CCFL correlation Advanced passive nuclear reactors Steam-water counter-current flow Pressurizer surge line assembly

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GB/T 7714 Yu, Jiangtao , Zhang, Dalin , Shi, Leitai et al. Experimental research on the characteristics of steam-water counter-current flow in the Pressurizer Surge Line assembly [J]. | EXPERIMENTAL THERMAL AND FLUID SCIENCE , 2018 , 96 : 180-191 .
MLA Yu, Jiangtao et al. "Experimental research on the characteristics of steam-water counter-current flow in the Pressurizer Surge Line assembly" . | EXPERIMENTAL THERMAL AND FLUID SCIENCE 96 (2018) : 180-191 .
APA Yu, Jiangtao , Zhang, Dalin , Shi, Leitai , Wang, Zhiwei , Tian, Wenxi , Su, G. H. et al. Experimental research on the characteristics of steam-water counter-current flow in the Pressurizer Surge Line assembly . | EXPERIMENTAL THERMAL AND FLUID SCIENCE , 2018 , 96 , 180-191 .
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Heat Transfer Analysis of In-vessel Melt Retention in Ellipsoidal and Spherical Lower Heads EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (7) , 1294-1299 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
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Abstract :

In the severe accident, the core melt with high temperature is redistributed in the pressure vessel and exerts a thermal load on the pressure vessel wall, which may lead to the failure of pressure vessel. Based on the Fortran90 language, in-vessel melt retention (IVR) analysis code IVRASA-ELLIP in severe accident of ellipsoidal lower head was developed, which was used to analyze the heat transfer and IVR of the pressure vessel with the ellipsoidal lower head under the severe accident. IVRASA-ELLIP was used to calculate the heat transfer of the VVER-1000, and the results of the wall heat flux, the oxide crust thickness and the wall thickness of the pressure vessel were obtained. The results of heat transfer of AP1000 were obtained through IVRASA code, which were compared with the results of VVER-1000. The results show that with the same initial parameters, the wall heat flux of the ellipsoidal lower head is smaller than that of the spherical lower head, the amount of ablation of the pressure vessel wall with the ellipsoidal lower head is smaller than that with the spherical lower head, and the oxide crust thickness in the pressure vessel with the ellipsoidal lower head is thicker than that with the spherical lower head. © 2018, Science Press. All right reserved.

Keyword :

Ellipsoidal lower head Heat transfer analysis High temperature Initial parameter Molten pool Severe accident Wall heat flux Wall thickness

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GB/T 7714 Liu, Fang , Zhang, Yapei , Tian, Wenxi et al. Heat Transfer Analysis of In-vessel Melt Retention in Ellipsoidal and Spherical Lower Heads [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) : 1294-1299 .
MLA Liu, Fang et al. "Heat Transfer Analysis of In-vessel Melt Retention in Ellipsoidal and Spherical Lower Heads" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 7 (2018) : 1294-1299 .
APA Liu, Fang , Zhang, Yapei , Tian, Wenxi , Su, Guanghui , Qiu, Suizheng . Heat Transfer Analysis of In-vessel Melt Retention in Ellipsoidal and Spherical Lower Heads . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) , 1294-1299 .
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Flow pattern effect on two-phase pressure drops in vertical upward flow across a horizontal tube bundle EI SCIE Scopus
期刊论文 | 2018 , 120 , 253-264 | ANNALS OF NUCLEAR ENERGY
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Abstract :

Shell-side two-phase pressure drops under cross-flow condition play an important role in the operation of heat exchangers. However, no well-verified general correlation for predicting two-phase friction multipliers in horizontal tube bundles is available in the open literatures. In the present paper, based on a combination of Kanizawa and Ribastki flow regime criteria with the modified Mao and Hibiki flow regime criteria, the flow patterns of two-phase vertically upward flow across horizontal tube bundles in the open literatures were identified. The results show that flow patterns have a significant effect on two-phase friction multipliers. For the bubbly flow, finely dispersed bubbly flow, and annular flow, the mass velocity effect on two-phase friction multipliers can be neglected. However, the two-phase friction multipliers of cap bubbly flow and churn flow decrease with an increasing mass velocity. Then, a general dimensionless correlation in terms of the Martinelli parameter and the two-phase Froude number were developed, which is able to accurately predicting gas-liquid two-phase flow resistances of churn flow in staggered horizontal tube bundle under cross-flow condition. New correlations for predicting two-phase friction multipliers of bubbly flow, finely dispersed bubbly flow and annular flow both in staggered and in-line horizontal tube bundles were also developed and can give excellent representations for the existing data in the open literatures. (C) 2018 Elsevier Ltd. All rights reserved.

Keyword :

Two-phase friction multipliers Horizontal tube bundles Cross-flow Flow patterns New correlations

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GB/T 7714 Zhang, K. , Hou, Y. D. , Tian, W. X. et al. Flow pattern effect on two-phase pressure drops in vertical upward flow across a horizontal tube bundle [J]. | ANNALS OF NUCLEAR ENERGY , 2018 , 120 : 253-264 .
MLA Zhang, K. et al. "Flow pattern effect on two-phase pressure drops in vertical upward flow across a horizontal tube bundle" . | ANNALS OF NUCLEAR ENERGY 120 (2018) : 253-264 .
APA Zhang, K. , Hou, Y. D. , Tian, W. X. , Zhang, Y. P. , Su, G. H. , Qiu, S. Z. . Flow pattern effect on two-phase pressure drops in vertical upward flow across a horizontal tube bundle . | ANNALS OF NUCLEAR ENERGY , 2018 , 120 , 253-264 .
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Experimental investigation on steam-water two-phase flow boiling heat transfer in a staggered horizontal rod bundle under cross-flow condition EI SCIE Scopus
期刊论文 | 2018 , 96 , 192-204 | EXPERIMENTAL THERMAL AND FLUID SCIENCE
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Abstract :

Single-phase convection and steam-water two-phase flow boiling heat transfer experiments in an electrically heated 4 x 16 staggered horizontal rod bundle under cross-flow condition have been carried out for the fol- lowing range: 45 kg m(-2)s(-1) <= G <= 383 kg m(-2)s(-1), 20.0 kW m(-2) <= q <= 55.0 kW m(-2), 0.02 <= x(out) <= 0.57 and 112 kPa <= P-in <= 190 kPa. The major effects on local flow boiling heat transfer coefficients in horizontal rod bundles were analyzed. Firstly, the row number has a significant effect on local flow boiling heat transfer coefficients since the vapor bubbles generated from the lower rods impinge on the surface of the upper rods and enhance the turbulence there. Secondly, an increasing heat flux contributes to an increase of the average bundle flow boiling heat transfer coefficient at all mass velocities. Thirdly, the local quality can also slightly enhance the flow boiling heat transfer. Fourthly, the flow boiling heat transfer can be enhanced with an increasing mass velocity under lower heat fluxes. In a general, a Chen-type correlation was developed to predict local flow boiling heat transfer coefficients in horizontal rod bundles with a maximum deviation of +/- 20%.

Keyword :

Horizontal rod bundle Flow boiling Cross-flow Heat transfer experiments

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GB/T 7714 Zhang, K. , Hou, Y. D. , Tian, W. X. et al. Experimental investigation on steam-water two-phase flow boiling heat transfer in a staggered horizontal rod bundle under cross-flow condition [J]. | EXPERIMENTAL THERMAL AND FLUID SCIENCE , 2018 , 96 : 192-204 .
MLA Zhang, K. et al. "Experimental investigation on steam-water two-phase flow boiling heat transfer in a staggered horizontal rod bundle under cross-flow condition" . | EXPERIMENTAL THERMAL AND FLUID SCIENCE 96 (2018) : 192-204 .
APA Zhang, K. , Hou, Y. D. , Tian, W. X. , Zhang, Y. P. , Su, G. H. , Qiu, S. Z. . Experimental investigation on steam-water two-phase flow boiling heat transfer in a staggered horizontal rod bundle under cross-flow condition . | EXPERIMENTAL THERMAL AND FLUID SCIENCE , 2018 , 96 , 192-204 .
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