• Complex
  • Title
  • Author
  • Keyword
  • Abstract
  • Scholars
Search
High Impact Results & Cited Count Trend for Year Keyword Cloud and Partner Relationship

Query:

学者姓名:秋穗正

Refining:

Source

Submit Unfold

Co-Author

Submit Unfold

Language

Submit

Clean All

Export Sort by:
Default
  • Default
  • Title
  • Year
  • WOS Cited Count
  • Impact factor
  • Ascending
  • Descending
< Page ,Total 76 >
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
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5 SCIE
期刊论文 | 2019 , 30 (1) | NUCLEAR SCIENCE AND TECHNIQUES
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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) .
Export to NoteExpress RIS BibTex
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
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
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
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
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
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
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
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools EI SCIE Scopus
期刊论文 | 2018 , 111 , 293-302 | ANNALS OF NUCLEAR ENERGY
WoS CC Cited Count: 3 SCOPUS Cited Count: 5
Abstract&Keyword Cite

Abstract :

Due to the high-Rayleigh-number natural convection and strong turbulence in the volumetrically heated corium melt pools, it is difficult to capture the heat transfer characteristics in detail. The Large Eddy Simulation (LES) methods are more and more applicable to nuclear industry and have become powerful tools to analyze the complicated turbulent and multi-phase flows. In this paper, the Wall-Modeled LES (WMLES) method was employed for the simulation of three typical corium pool heat transfer experiments: BALI, LIVE and COPRA experiments. The melt pool temperature and heat flux from numerical simulations were in good comparison with experimental data. Then numerical simulation for the transient formation of two-layer corium pool was performed with prototypical corium to obtain the heat transfer data in reactor situation. There was a distinctive gap between the temperatures in two layers. The heat flux showed a decrease at the two-layer interface and a huge jump at the metal layer side. The thin metal layer on the top at the early stage of the two-layer formation will threaten the vessel integrity. (C) 2017 Elsevier Ltd. All rights reserved.

Keyword :

Turbulent heat transfer Large eddy simulation Corium pool IVR

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Zhang, Luteng , Luo, Simin , Zhang, Yapei et al. Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools [J]. | ANNALS OF NUCLEAR ENERGY , 2018 , 111 : 293-302 .
MLA Zhang, Luteng et al. "Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools" . | ANNALS OF NUCLEAR ENERGY 111 (2018) : 293-302 .
APA Zhang, Luteng , Luo, Simin , Zhang, Yapei , Tian, Wenxi , Su, G. H. , Qiu, Suizheng . Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools . | ANNALS OF NUCLEAR ENERGY , 2018 , 111 , 293-302 .
Export to NoteExpress RIS BibTex
Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method EI SCIE Scopus
期刊论文 | 2018 , 203 (2) , 194-204 | NUCLEAR TECHNOLOGY
Abstract&Keyword Cite

Abstract :

In detailed previous work by the authors, an innovative decay heat removal (DHR) system has been proposed and designed for the Integral Inherently Safe Light Water Reactor ((IS)-S-2-LWR). The current paper studies the inadvertent actuation of one DHR system train during (IS)-S-2-LWR normal operation due to a false signal or operator action. The RELAP5 code is used to perform a one-dimensional study, and important thermal-hydraulic characteristics, including primary loop coolant flow rate, pressure, temperature, DHR primary-side flow rate, and coolant temperature, are achieved during this transient. Then, a detailed computational fluid dynamics simulation utilizing STARCCM+ is carried out to investigate the coolant mixing characteristics in the downcomer and lower plenum and obtain the local thermal-hydraulic conditions at the reactor core inlet. It is found that as a consequence of inadvertent DHR actuation, the maximum overcooling at the reactor core inlet is about 3 K, which would not result in significant reactivity insertion. Furthermore, a more severe transient of inadvertent DHR operation with intermediate loop break is studied, and the results show that this would not lead to more significant overcooling to the (IS)-S-2-LWR core compared with inadvertent DHR operation without intermediate loop break. This work is an indispensable supplement for DHR system comprehensive assessment in the (IS)-S-2-LWR project.

Keyword :

mixing phenomenon inadvertent operation Integral inherently safe light water reactor decay heat removal system computational fluid dynamics analysis

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Wang, Mingjun , Manera, Annalisa , Petrov, Victor et al. Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method [J]. | NUCLEAR TECHNOLOGY , 2018 , 203 (2) : 194-204 .
MLA Wang, Mingjun et al. "Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method" . | NUCLEAR TECHNOLOGY 203 . 2 (2018) : 194-204 .
APA Wang, Mingjun , Manera, Annalisa , Petrov, Victor , Qiu, Suizheng , Tian, Wenxi , Su, G. H. . Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method . | NUCLEAR TECHNOLOGY , 2018 , 203 (2) , 194-204 .
Export to NoteExpress RIS BibTex
Review of conceptual design and fundamental research of molten salt reactors in China EI SCIE Scopus
期刊论文 | 2018 , 42 (5) , 1834-1848 | INTERNATIONAL JOURNAL OF ENERGY RESEARCH
WoS CC Cited Count: 8 SCOPUS Cited Count: 16
Abstract&Keyword Cite

Abstract :

Molten salt reactor (MSR) as 1 candidate of the generation IV advanced nuclear power systems attracted more attention in China due to its top ranked in fuel cycle and thorium utilization. Two types of MSR concepts were studied and developed in parallel, namely the MSR with liquid fuel and that with solid fuel. Abundant fundamental research including the neutronics modeling, thermal-hydraulics modeling, safety analysis, material investigation, molten salts technologies etc. were carried out. Some analysis software such as COUPLE and FANCY were developed. Several experimental facilities like high-temperature fluoride salt experiment loop have been constructed. Some passive residual heat removal systems were designed, and 1 test facility is under construction. The key MSR techniques including the extraction and separation of molten salt and construction of N-base alloy have been mastered. Based on these fundamental research, Chinese Academy of Sciences has completed the design of thorium-based MSRs with solid fuel and liquid fuel and is promoting their construction in the near future. In China, future efforts should be paid to the material, online fuel processing, Th-U fuel cycle, component design, and construction and thermal-hydraulic experiments for MSR, which are rather challenging nowadays.

Keyword :

fundamental research thorium utilization solid fuel molten salt reactor liquid fuel

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Zhang, Dalin , Liu, Limin , Liu, Minghao et al. Review of conceptual design and fundamental research of molten salt reactors in China [J]. | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2018 , 42 (5) : 1834-1848 .
MLA Zhang, Dalin et al. "Review of conceptual design and fundamental research of molten salt reactors in China" . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH 42 . 5 (2018) : 1834-1848 .
APA Zhang, Dalin , Liu, Limin , Liu, Minghao , Xu, Rongshuan , Gong, Cheng , Zhang, Jun et al. Review of conceptual design and fundamental research of molten salt reactors in China . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2018 , 42 (5) , 1834-1848 .
Export to NoteExpress RIS BibTex
Numerical analysis of the granular flow and heat transfer in the ADS granular spallation target EI SCIE Scopus
期刊论文 | 2018 , 330 , 59-71 | NUCLEAR ENGINEERING AND DESIGN
WoS CC Cited Count: 2 SCOPUS Cited Count: 2
Abstract&Keyword Cite

Abstract :

The Accelerator-Driven System (ADS) is one of the most effective tools to deal with the spent fuels by nuclear transmutation. In the ADS, the neutrons could be produced by bombarding a proper spallation target with a concentrated beam of high-energy accelerator protons. Subsequently, the produced neutrons drive the nuclear transmutation in the sub-critical reactor. Therefore, the spallation target is an essential part of the ADS. Recently researchers proposed a gravity-driven dense granular target, which has both solid and fluid target characteristics. This study developed the FOCUS code to simulate the granular flow and heat transfer behaviors involved in the gravity-driven dense granular target. A solid sedimentation experiment was performed and simulated by FOCUS code. The simulation results were consistent with the findings in the experiment, which proved FOCUS to be an appropriate code to simulate the granular flow. Then the sensitivity analyses were performed to investigate the influence of the inlet flow rate and the target area shape on the state of the granular flow in the ADS granular spallation target. The simulation results showed that the velocity of each granular spallation target was related with the height of piled-up particles in the targets area. The geometry size of the target area obviously affected the flow state of targets. Meanwhile the inlet flow rate had a slight influence on the stable flow, unless the inlet flow was so large that targets overflow the target area. Furthermore, the study also calculated the heat transfer between each target and found that heat conduction between two contacted granular spallation targets was the main process in the targets area. The present work would be instructive for the ADS system design.

Keyword :

Heat conduction Granular spallation target Numerical simulations FOCUS code Accelerator-Driven System

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Chen, Ronghua , Guo, Kailun , Zhang, Yanshi et al. Numerical analysis of the granular flow and heat transfer in the ADS granular spallation target [J]. | NUCLEAR ENGINEERING AND DESIGN , 2018 , 330 : 59-71 .
MLA Chen, Ronghua et al. "Numerical analysis of the granular flow and heat transfer in the ADS granular spallation target" . | NUCLEAR ENGINEERING AND DESIGN 330 (2018) : 59-71 .
APA Chen, Ronghua , Guo, Kailun , Zhang, Yanshi , Tian, Wenxi , Qiu, Suizheng , Su, G. H. . Numerical analysis of the granular flow and heat transfer in the ADS granular spallation target . | NUCLEAR ENGINEERING AND DESIGN , 2018 , 330 , 59-71 .
Export to NoteExpress RIS BibTex
10| 20| 50 per page
< Page ,Total 76 >

Export

Results:

Selected

to

Format:
FAQ| About| Online/Total:3101/54984290
Address:XI'AN JIAOTONG UNIVERSITY LIBRARY(No.28, Xianning West Road, Xi'an, Shaanxi Post Code:710049) Contact Us:029-82667865
Copyright:XI'AN JIAOTONG UNIVERSITY LIBRARY Technical Support:Beijing Aegean Software Co., Ltd.