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学者姓名:苏光辉

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< Page ,Total 58 >
Effects of nitrogen and carbon monoxide on the detonation of hydrogen-air gaseous mixtures EI SCIE
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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Utilization of 3D fuel modeling capability of BISON to derive new insights in performance of advanced PWR fuel concepts SCIE
期刊论文 | 2019 , 516 , 271-288 | JOURNAL OF NUCLEAR MATERIALS
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Abstract :

The Internally and eXternally cooled Annular Fuel (IXAF) and Lightbridge's metallic Helical Cruciform Fuel (HCF) are two innovative fuels, which could allow significant power uprate (similar to 20%) in current Pressurized Water Reactors (PWRs) while maintaining or improving safety margins. Both fuels would substantially benefit from 3D modeling to properly assess their performance in a PWR core and to further their technology readiness level. For IXAF, the possible misalignment of the annular pellets creates a mismatch in inner and outer gap size, which will result in an azimuthally asymmetric temperature distribution and different heat split between the inner and outer channel. For HCF, the fuel deformation requires 3D consideration due to its inherit helical geometry. In this work, the first attempt of building thermo-mechanical capability for analysis of these fuels was made in BISON fuel performance code. The 3D fuel behavior under normal operation along with loss of coolant accident (LOCA) and reactivity initiated accident (RIA) were simulated. The preliminary key insights derived from the 3D fuel performance modeling were the potential excessive ballooning of IXAF during LOCA and excessive swelling of HCF during RIA. However, overall both fuels did show promising and improved performance at 20% uprated conditions compared to traditional solid fuel geometry and material. (C) 2019 Elsevier B.V. All rights reserved.

Keyword :

Annular fuel LB-LOCA BISON Lightbridge fuel RIA

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GB/T 7714 Deng, Yangbin , Shirvan, Koroush , Wu, Yingwei et al. Utilization of 3D fuel modeling capability of BISON to derive new insights in performance of advanced PWR fuel concepts [J]. | JOURNAL OF NUCLEAR MATERIALS , 2019 , 516 : 271-288 .
MLA Deng, Yangbin et al. "Utilization of 3D fuel modeling capability of BISON to derive new insights in performance of advanced PWR fuel concepts" . | JOURNAL OF NUCLEAR MATERIALS 516 (2019) : 271-288 .
APA Deng, Yangbin , Shirvan, Koroush , Wu, Yingwei , Su, Guanghui . Utilization of 3D fuel modeling capability of BISON to derive new insights in performance of advanced PWR fuel concepts . | JOURNAL OF NUCLEAR MATERIALS , 2019 , 516 , 271-288 .
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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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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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Conceptual design and coupled neutronic/thermal-hydraulic/mechanical research of the supercritical water cooled ceramic blanket for CFETR EI SCIE
期刊论文 | 2019 , 138 , 272-281 | Fusion Engineering and Design
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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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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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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 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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Numerical study on thermal deformation behaviors of the single subassembly in sodium-cooled fast reactors based on Euler-Bernoulli beam theory SCIE
期刊论文 | 2019 , 345 , 28-39 | NUCLEAR ENGINEERING AND DESIGN
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Due to the special geometry and compact arrangement of subassemblies in the sodium-cooled fast reactor, thermal deformation of the assemblies are easily triggered by uneven temperature distribution in the reactor core, which is negative for reactor safety. Therefore, it is necessary for researchers and engineers to conduct quantitative analysis towards thermal deformation behaviors of subassemblies, and furthermore, give reliable evaluation results on aftermath of the assembly thermal deformation. In the present study, thermal deformation analysis code FADAC, which was developed based on Euler-Bernoulli beam theory, was applied to make numerical investigation for thermal deformation behaviors of a single subassembly in sodium-cooled fast reactor. In order to preliminarily assess the capacity for thermal deformation, different temperature gradient conditions were considered and analyzed in detail for free bowing of the subassembly, furthermore, with the aid of numerical analysis for thermal axial forces and thermal bending moments, axial displacements and deflections were calculated finally. Besides, in order to predict the comprehensive results of the assembly deformation under both thermal load and mechanical load, restrained bowing were also analyzed for different temperature conditions. All of the simulation results were in good accordance with the experimental data. The present numerical research is of great significance to assembly deformation research in the sodium-cooled fast reactor, and will surely lay a solid foundation for deformation analysis towards multi-subassemblies.

Keyword :

Sodium-cooled fast reactor Thermal deformation Subassembly Deflection

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GB/T 7714 Ma, Zhenhui , Ma, Zehua , Wu, Yingwei et al. Numerical study on thermal deformation behaviors of the single subassembly in sodium-cooled fast reactors based on Euler-Bernoulli beam theory [J]. | NUCLEAR ENGINEERING AND DESIGN , 2019 , 345 : 28-39 .
MLA Ma, Zhenhui et al. "Numerical study on thermal deformation behaviors of the single subassembly in sodium-cooled fast reactors based on Euler-Bernoulli beam theory" . | NUCLEAR ENGINEERING AND DESIGN 345 (2019) : 28-39 .
APA Ma, Zhenhui , Ma, Zehua , Wu, Yingwei , Wang, Mingjun , Su, G. H. . Numerical study on thermal deformation behaviors of the single subassembly in sodium-cooled fast reactors based on Euler-Bernoulli beam theory . | NUCLEAR ENGINEERING AND DESIGN , 2019 , 345 , 28-39 .
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