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学者姓名:秋穗正

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Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 EI
期刊论文 | 2019 , 112 , 63-74 | Progress in Nuclear Energy
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Abstract :

Pressurized water reactors (PWRs) use Steam Generators (SG) to transfer the reactor core heat to the secondary loop. The SG contains thousands of heat transfer tubes and tubes plugging operation is always adopted to solve the heat transfer tubes rupture problems caused by the complex operation conditions, including thermal stress, radiation environment and material corrosion. However, tubes plugging operation has great impact on the SG hydraulic performance which leading to some safety issues in nuclear power plants (NPP). In this paper, the AP1000 SG primary side three-dimensional thermal hydraulic simulation model is built using porous media method, in which the important parameters are achieved employing the separate tube simulations. An innovative grid mark method is proposed to realize the flexible tubes plugging conditions. The tubes plugging fractions varies from 5% to 20%, and the effects of tubes plugging locations are also considered. The detailed flow fields in whole AP1000 SG primary side are achieved. The large eddies and some small vortexes are generated in the inlet side of bottom channel head. The momentum loss mechanism is revealed in the whole SG primary loop. Results show that the pressure drop in primary loop increases with the plugging fraction and varies with different tubes plugging positions. The pressure drop goes up to 492.77 KPa in case that the tubes plugging fraction reaches 20%. The tubes plugging position affects the results significantly in case that the plugging fraction is larger than 15%. This work is meaningful for the depth understanding of AP1000 SG tube plugging operations and provides a guideline for the SG maintenance strategy during the whole plant lifetime. © 2018 Elsevier Ltd

Keyword :

AP1000 Hydraulic characteristic Hydraulic performance Maintenance strategies Porous media methods Porous medium Radiation environments Thermal-hydraulic simulations

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GB/T 7714 Zhao, Xiaohan , Wang, Mingjun , Chen, Chong et al. Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 [J]. | Progress in Nuclear Energy , 2019 , 112 : 63-74 .
MLA Zhao, Xiaohan et al. "Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000" . | Progress in Nuclear Energy 112 (2019) : 63-74 .
APA Zhao, Xiaohan , Wang, Mingjun , Chen, Chong , Wang, Xi , Ju, Haoran , Tian, Wenxi et al. Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000 . | Progress in Nuclear Energy , 2019 , 112 , 63-74 .
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Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle EI
期刊论文 | 2019 , 578-587 | Applied Thermal Engineering
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Abstract :

The single-phase thermal hydraulic characteristics of liquid metal sodium are very essential for the design and safety analysis of sodium-cooled fast reactor (SFR). In this paper, the pressure drop and heat transfer features of single-phase liquid sodium were experimentally investigated in a 7 rod bundle with the velocity range of 0–4 m/s, heat flux up to 120 kW/m2 and the absolute pressure range of 0–0.2 MPa. The corresponding Reynolds number ranges from 4000 to 40,000, and the Pe number varies from 0 to 340. It was found that the critical Re number for transition-turbulent flow of single-phase liquid sodium is 13,500 in the hexagonal 7-rod bundle. Then the effects of relative axial position, wall heat flux and Re number on the heat transfer were discussed, respectively. Some existing correlations in the literatures were assessed and compared with the experimental data. Results indicated that these correlations could not predict the current experiments well because of the different geometries and working fluids. The new correlations for the friction factor and Nu number calculations were proposed based on the current experimental data. For 98.5% of heat transfer data produced by the other researchers, the prediction error of the new correlation is less than 30%. For most of the experimental data, it is less than 20%, which sufficiently proves that the correlation developed in this paper could give a good prediction of the experimental data obtained by other researchers. © 2018 Elsevier Ltd

Keyword :

Different geometry Heat transfer data Liquid sodium Liquid sodium flows Rod bundles Single-phase liquids Sodium cooled fast reactors (SFR) Thermal hydraulics

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GB/T 7714 Hou, Yandong , Wang, Liu , Wang, Mingjun et al. Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle [J]. | Applied Thermal Engineering , 2019 : 578-587 .
MLA Hou, Yandong et al. "Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle" . | Applied Thermal Engineering (2019) : 578-587 .
APA Hou, Yandong , Wang, Liu , Wang, Mingjun , Zhang, Kui , Zhang, Xisi , Hu, Wenjun et al. Experimental study of liquid sodium flow and heat transfer characteristics along a hexagonal 7-rod bundle . | Applied Thermal Engineering , 2019 , 578-587 .
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Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate EI
期刊论文 | 2019 , 881-888 | Applied Thermal Engineering
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Abstract :

In order to solve the problem of non-convergence of CHF directly calculated by FLUENT under atmospheric pressure and low flow rate, a CFD methodology was proposed based on four equation drift flux model and an improved RPI wall boiling model to predict the CHF and thermal-hydraulics characteristics in the flow channel formed by the outer wall of the RPV and the inner wall of the insulation. The governing equations and flow boiling models were added into FLUENT solver, and then worked with Mixture multiphase models by user defined functions (UDFs). The developed CFD models for CHF prediction were validated by using experimental data, and the prediction results had a quite good agreement with the experimental data with deviations less than 20%. It indicated that the CFD methodology proposed in this study had a good convergence at atmospheric pressure and low flow rate. Meanwhile the CFD methodology could be qualified to predict the characteristics of CHF, and it provided a potential way to predict the CHF in the flow channel formed by the outer wall of the RPV and the inner wall of the insulation under IVR conditions of nuclear power plants. © 2018 Elsevier Ltd

Keyword :

CFD methodologies Drift flux modeling Flow boiling models Governing equations Multiphase model Numerical predictions Severe accident User Defined Functions

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GB/T 7714 Zhang, Yapei , Zhang, Rui , Tian, Wenxi et al. Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate [J]. | Applied Thermal Engineering , 2019 : 881-888 .
MLA Zhang, Yapei et al. "Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate" . | Applied Thermal Engineering (2019) : 881-888 .
APA Zhang, Yapei , Zhang, Rui , Tian, Wenxi , Su, G.H. , Qiu, Suizheng . Numerical prediction of CHF based on CFD methodology under atmospheric pressure and low flow rate . | Applied Thermal Engineering , 2019 , 881-888 .
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Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5 EI
期刊论文 | 2019 , 112 , 209-224 | Progress in Nuclear Energy
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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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Effect of stratified interface instability on thermal focusing effect in two-layer corium pool EI
期刊论文 | 2019 , 359-370 | International Journal of Heat and Mass Transfer
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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 : 359-370 .
MLA Ge, K. et al. "Effect of stratified interface instability on thermal focusing effect in two-layer corium pool" . | International Journal of Heat and Mass Transfer (2019) : 359-370 .
APA Ge, K. , Zhang, Y.P. , Tian, W.X. , Su, G.H. , Qiu, S.Z. . Effect of stratified interface instability on thermal focusing effect in two-layer corium pool . | International Journal of Heat and Mass Transfer , 2019 , 359-370 .
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Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME In order to enhance the inherent safety of sodium-cooled fast reactors, innovative hydraulically suspended absorber rod (HSR) passive shut-down system have been proposed for China demonstration fast reactor. In this study, based on the functional and performance requirements, a full-scale experimental setup has been designed and fabricated for the analysis of the HSR as applied to the prototype reactor. The main characteristic of the test facility is the actuation of the mobile safety rod is triggered by coolant flow rate decrease in the primary loop below half the nominal value and then the rod inserts into the stationary sleeve by gravity. The objective is to investigate the dynamic performance of HSR and establish the laws of its movement at lowering the flow rate modeling the coastdown of primary circulating pump. A series of tests have been performed, including start-up, steady-state operation, loss of flow accident, sensitivity analysis and reliability test. This study also focused on the effect of various factors on scram time, the effect of pump coasting time, rod weight, gap between rod and guide tube, bypass holes, cone angle of rod, flow rate and fluid temperature are analyzed. The experimental results demonstrate the functionality and reliability of the HSR, which would lay foundation for further optimization design.

Keyword :

Dynamic performance Experiment Passive shutdown system SFR

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GB/T 7714 Song, Jian , Wu, Yingwei , Tian, Wenxi et al. Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor [C] . 2018 .
MLA Song, Jian et al. "Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor" . (2018) .
APA Song, Jian , Wu, Yingwei , Tian, Wenxi , Qiu, Suizheng , Su, Guanghui . Characterization and experimental investigation for the dynamic performance of the hydraulically suspended passive shutdown system in China sodium-cooled fast reactor . (2018) .
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Experimental studies on the thermal-hydraulics of dowtherm a through the pebble bed with internal heat generation Scopus
会议论文 | 2018 , 5
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Copyright © 2018 ASME. The Fluoride-salt-cooled High temperature Reactors (FHRs) are an advanced concept using a novel combination of high-temperature coated-particle fuel, low-pressure fluoride-salt coolant and air-Brayton power conversion system. Prismatic fuel or pebble fuel are adopted for the conceptual core designs of FHRs like TMSR-SF, MK1 PB-FHR and SM-AHTR. The high-Prandtl-number FLiBe is mainly adopted as the primary coolant, which specifies in high melting and boiling point and high volumetric capacity. The experimental results obtained from the air, water or inert gas prove reliable for the Prandtl number vary from 0.7 to 7. Little experimental research has been conducted to prove applicability of the above results to the high-Prandtl fluid, fluoride salts in the packed pebble bed. In this paper, a pebble bed experimental facility has been designed and constructed for the FHRs to explore the thermal-hydraulic characteristics of fluoride salts in the reactor pebble bed core. Dowtherm A is adopted as a simulant fluid for the fluoride salts. The cylindrical test section is packed with steel pebbles. The electromagnetic induction heating system is used to provide internal heat source for the pebble beds. The forced flow and convective heat transfer of high-Prandtl-number fluid in the pebble bed with internal heat generation are investigated in the experiment. The fluid inlet temperature and mass flow rate are studied on the thermal-hydraulic characteristics.

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GB/T 7714 Liu, Limin , Zhang, Dalin , Li, Linfeng et al. Experimental studies on the thermal-hydraulics of dowtherm a through the pebble bed with internal heat generation [C] . 2018 .
MLA Liu, Limin et al. "Experimental studies on the thermal-hydraulics of dowtherm a through the pebble bed with internal heat generation" . (2018) .
APA Liu, Limin , Zhang, Dalin , Li, Linfeng , Yang, Yichen , Wang, Chenglong , Qiu, Suizheng . Experimental studies on the thermal-hydraulics of dowtherm a through the pebble bed with internal heat generation . (2018) .
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COPRA experiments on melt pool behavior with eutectic NANO3-KNO3 simulant Scopus
会议论文 | 2018 , 6A | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Copyright © 2018 ASME. The COPRA experiments were performed to study the natural convection heat transfer behavior in a large-scale homogeneous melt pool inside the reactor pressure vessel lower plenum. The test section consists of a two-dimensional 1/4 circular slice with an inner radius of 2.2 m. A non-eutectic binary mixture 20%NaNO3-80%KNO3 was selected as melt simulant in the previous tests and the Rayleigh number of the melt pool reached up to 1016. In this paper, the working fluid was a eutectic binary mixture 50%NaNO3-50%KNO3. The melt pool temperature, heat flux distribution and crust thickness were obtained in the experiments with different heating powers. Results from the eutectic molten salt tests can be applied for posttest calculations and comparative analyses.

Keyword :

COPRA Heat transfer Melt pool Severe accident

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GB/T 7714 Qiu, Suizheng , Su, G. H. . COPRA experiments on melt pool behavior with eutectic NANO3-KNO3 simulant [C] . 2018 .
MLA Qiu, Suizheng et al. "COPRA experiments on melt pool behavior with eutectic NANO3-KNO3 simulant" . (2018) .
APA Qiu, Suizheng , Su, G. H. . COPRA experiments on melt pool behavior with eutectic NANO3-KNO3 simulant . (2018) .
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Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k − ε model (SKE) and the realizable k − ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.

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GB/T 7714 Wang, Shibao , Zhang, Dalin , Wang, Chenglong et al. Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR [C] . 2018 .
MLA Wang, Shibao et al. "Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR" . (2018) .
APA Wang, Shibao , Zhang, Dalin , Wang, Chenglong , Song, Ping , Chen, Jing , Qiu, Suizheng et al. Validation of a code and effect of turbulence model on predicting thermal stratification phenomena in the upper plenum of SFR . (2018) .
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Thermal-hydraulic analysis of TOPAZ-II with modified RELAP5 Scopus
会议论文 | 2018 , 6A | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Copyright © 2018 ASME. With the advantages of high reliability, high power density and long life, small nuclear power reactor has become one of the most excellent space power options in the space missions. TOPAZ-II is the most mature space nuclear power reactor based on thermionic conversion. In this paper, the thermo-physical and transport properties of NaK-78 and heat transfer correlations for liquid metals are implemented into the RELAP5 code. The modified RELAP5 has already been accessed to analyze the thermal-hydraulic characteristics of the space reactor cooled by NaK-78. A RELAP5 model including the core, TFEs, radiator, coolant loop and volume accumulator is developed. Temperature reactivity feedback, TFE emitter, TFE collector, moderator and the reactivity insertion effects of control drums and safety drums are modeled in the point reactor kinetics equations with six-group delayed neutrons. To V&V the integrated TOPAZ-II system model, the steady state is simulated and analyzed. The steady state calculated results are in good agreement with the designed values. On the basis of V&V, a hypothetical reactivity insertion accident is simulated and analyzed. During the accident, the automatic control system is assumed to be malfunctioned, 0.01$ positive reactivity is introduced for 500s and then control drums start to rotate inward. The maximum temperatures of fuel and emitter are below the melting temperature, respectively. The maximum temperature of coolant is 940K with 160K margin from boiling. With the rotating of control drums, the reactor reaches critical again.

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GB/T 7714 Qiu, Suizheng , Tian, Wenxi . Thermal-hydraulic analysis of TOPAZ-II with modified RELAP5 [C] . 2018 .
MLA Qiu, Suizheng et al. "Thermal-hydraulic analysis of TOPAZ-II with modified RELAP5" . (2018) .
APA Qiu, Suizheng , Tian, Wenxi . Thermal-hydraulic analysis of TOPAZ-II with modified RELAP5 . (2018) .
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