• 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 >
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
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
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
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 , 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 .
Export to NoteExpress RIS BibTex
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
Abstract&Keyword Cite

Abstract :

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

Keyword :

Containment dome Dripping Steam condensation Condensate film

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
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
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 , 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 .
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
WoS CC Cited Count: 1
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
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
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 , 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 .
Export to NoteExpress RIS BibTex
Analysis of ADS Granular Target Melting Accident Based on FOCUS Code EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (8) , 1431-1437 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
Abstract&Keyword Cite

Abstract :

In the study, the FOCUS code developed based on moving particle semi-implicit (MPS) method was used to simulate the movement and heat transfer behavior of the granular target and the flow process of molten materials, aiming at the accelerator driven subcritical system (ADS) granular target melting accident. The smaller size of the granular target will cause the larger rate of the heat transfer and melting. The conduction of granular target is the main heat transfer process. © 2018, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Accelerator driven subcritical systems Flow process Heat transfer behavior Heat transfer process Moving particle semiimplicit method

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Li, Chenxi , Wen, Yan , Guo, Kailun et al. Analysis of ADS Granular Target Melting Accident Based on FOCUS Code [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (8) : 1431-1437 .
MLA Li, Chenxi et al. "Analysis of ADS Granular Target Melting Accident Based on FOCUS Code" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 8 (2018) : 1431-1437 .
APA Li, Chenxi , Wen, Yan , Guo, Kailun , Chen, Ronghua , Wu, Yingwei , Tian, Wenxi et al. Analysis of ADS Granular Target Melting Accident Based on FOCUS Code . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (8) , 1431-1437 .
Export to NoteExpress RIS BibTex
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
Abstract&Keyword Cite

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

Cite:

Copy from the list or Export to your reference management。

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 .
Export to NoteExpress RIS BibTex
Multi-layer Clad and Non-rigid Pellet Mechanical Modeling and Its Application EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (7) , 1308-1315 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
Abstract&Keyword Cite

Abstract :

Based on the finite difference method, the mechanical model of multi-layer cylindrical body was established for accident tolerant fuel (ATF). The new model was implemented in FRAPCON 4.0 code to replace the original Fracas model. By the simulation for the same research objective, the characteristics of the new and old models were analyzed and the rationality and advancement of modification were validated. The comparison indicates that most of parameters agree well, demonstrating that the non-rigid fuel assumption is reasonable to some extent for the traditional fuel. However, the old model is over conservative in contact pressure prediction, and the upgraded code provides more reasonable result. Finally, the potential application of the multi-layer mechanical model and the upgraded FRAPCON 4.0 in the fuel behavior analysis was predicted. © 2018, Science Press. All right reserved.

Keyword :

Accident tolerant fuels Behavior analysis Contact pressures Cylindrical bodies FRAPCON 4.0 upgrade Mechanical model Research objectives Traditional fuels

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Deng, Yangbin , He, Yanan , Wu, Yingwei et al. Multi-layer Clad and Non-rigid Pellet Mechanical Modeling and Its Application [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) : 1308-1315 .
MLA Deng, Yangbin et al. "Multi-layer Clad and Non-rigid Pellet Mechanical Modeling and Its Application" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 7 (2018) : 1308-1315 .
APA Deng, Yangbin , He, Yanan , Wu, Yingwei , Tian, Wenxi , Qiu, Suizheng , Su, Guanghui . Multi-layer Clad and Non-rigid Pellet Mechanical Modeling and Its Application . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) , 1308-1315 .
Export to NoteExpress RIS BibTex
10| 20| 50 per page
< Page ,Total 76 >

Export

Results:

Selected

to

Format:
FAQ| About| Online/Total:2651/65150543
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.