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< Page ,Total 89 >
Experimental study on molten corium-concrete interaction with simulant metal and oxide EI
期刊论文 | 2022 , 167 | Annals of Nuclear Energy
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Abstract :

Under hypothetical Reactor Pressure Vessel (RPV) failure accidents in Light Water Reactor (LWR), Molten Corium-Concrete Interaction (MCCI) will cause erosion of cavity concrete, possibly resulting in containment failure due to basemat penetration and overpressure. The CINA (Corium-Concrete Interaction Apparatus) experiment was conducted to investigate the MCCI of metallic and oxidic corium with siliceous concrete. Simulant melt material metallic iron and oxidic alumina generated by exothermic thermite chemical reaction was used to investigate the two-dimensional erosion of a cylindrical crucible in CINA experiment. The crucible was fabricated from siliceous concrete with an inner diameter of 300 mm and a height of 500 mm containing reinforcement (rebars). Decay heat in the melt was simulated by subsequent sustained addition of thermite. During the experiment, the melt temperature was monitored by two thermocouples and a two-color optical endurance pyrometer. Besides, the phenomenon that appeared on the melt surface was recorded by the video camera placed above the crucible and the formation process of the crust anchoring was observed. After the experiment, the crucible was split in half by a wire saw to accurately measure the ablation depth. Axial-radial ablation depths were 25 mm and 13 mm, respectively. Current findings contribute to the further understanding of MCCI mechanisms and the optimization of MCCI mitigation strategies. © 2021

Keyword :

Alumina Aluminum oxide Erosion Light water reactors Nuclear power plants Concretes Nuclear fuels Hydraulics Pressure vessels Nuclear reactor accidents Thermocouples Ablation Video cameras

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GB/T 7714 Xu, Zhichun , Zhang, Yapei , Wu, Zijie et al. Experimental study on molten corium-concrete interaction with simulant metal and oxide [J]. | Annals of Nuclear Energy , 2022 , 167 .
MLA Xu, Zhichun et al. "Experimental study on molten corium-concrete interaction with simulant metal and oxide" . | Annals of Nuclear Energy 167 (2022) .
APA Xu, Zhichun , Zhang, Yapei , Wu, Zijie , Zhan, Dekui , Su, G.H , Tian, Wenxi et al. Experimental study on molten corium-concrete interaction with simulant metal and oxide . | Annals of Nuclear Energy , 2022 , 167 .
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From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction EI
期刊论文 | 2022 , 165 | Annals of Nuclear Energy
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Abstract :

When a severe accident occurs in a nuclear reactor, fuel–coolant interaction (FCI) may occur and cause steam explosion. Energetic FCI or steam explosion will threaten the integrity of the containment and even cause radioactive leakage. Therefore, the research and evaluation of the morphology and development of the melt in the FCI process is very important. In recent decades, many scholars have conducted theoretical and experimental research on this phenomenon, and formed a large number of mathematical and physical models for the degradation, evolution and relocation of the melt. This study retrospects and summarizes the experimental and theoretical research on the FCI phenomenon, and describes the behavior of the melt in the entire evolution process from the melt jet to debris bed. In addition, on the basis of literature review, the models of each stage are classified according to the formation mechanisms and applicable conditions. Finally, in the present study, the future development direction is prospected on the ground of the previous research results. © 2021 Elsevier Ltd

Keyword :

Debris Coolants Nuclear reactors Fuels

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GB/T 7714 Sun, Ruiyu , Wu, Liangpeng , Ding, Wen et al. From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction [J]. | Annals of Nuclear Energy , 2022 , 165 .
MLA Sun, Ruiyu et al. "From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction" . | Annals of Nuclear Energy 165 (2022) .
APA Sun, Ruiyu , Wu, Liangpeng , Ding, Wen , Chen, Ronghua , Tian, Wenxi , Qiu, Suizheng et al. From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction . | Annals of Nuclear Energy , 2022 , 165 .
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Numerical investigation on thermohydraulic performance of high temperature hydrogen in twisted rod channels EI SCIE
期刊论文 | 2021 , 161 | ANNALS OF NUCLEAR ENERGY
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Nuclear thermal propulsion (NTP) is one of the most promising approaches for deep space exploration and manned interstellar voyage. As the propellant and coolant of NTP reactors, hydrogen provides higher specific impulse. In this paper, the thermohydraulic performance of high temperature hydrogen flowing through the twisted rod bundles were investigated numerically by contrasting with the cruciform rod bundles and the circular rod bundles. The heat transfer models were validated by literature data and the relative errors are within 20%. Twisted rods could enhance lateral mixing and heat transfer since the twisted lobes and valleys could enhance radial heat transfer and produce a strong rotational flow near the walls disturbing boundary layer, which reduces hot spot factors and assures the thermal safety. The largest relative difference of the J(F) factor between the twisted and other non-twisted channels are 28.0% and 6.1% respectively in this model. (C) 2021 Published by Elsevier Ltd.

Keyword :

Flow characteristics Heat transfer performance Hydrogen Twisted rod channels Nuclear thermal propulsion

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GB/T 7714 Fang, Yuliang , Qin, Hao , Wang, Chenglong et al. Numerical investigation on thermohydraulic performance of high temperature hydrogen in twisted rod channels [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 161 .
MLA Fang, Yuliang et al. "Numerical investigation on thermohydraulic performance of high temperature hydrogen in twisted rod channels" . | ANNALS OF NUCLEAR ENERGY 161 (2021) .
APA Fang, Yuliang , Qin, Hao , Wang, Chenglong , Zhou, Lei , Zhang, Jing , Zhang, Dalin et al. Numerical investigation on thermohydraulic performance of high temperature hydrogen in twisted rod channels . | ANNALS OF NUCLEAR ENERGY , 2021 , 161 .
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Validation of TRACE capability to simulate unprotected transients in Sodium Fast Reactor using FFTF LOFWST Test #13 EI SCIE
期刊论文 | 2021 , 164 | ANNALS OF NUCLEAR ENERGY
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Abstract :

The US NRC system code TRACE has been modified at PSI for application to liquid-metal-cooled reactor. An unprotected loss-of-flow-without-scram test performed at the Fast Flux Test Facility (FFTF) provides an opportunity to enhance the validation base of TRACE to transient analysis for sodium-cooled fast reac-tor (SFR). The FFTF primary system model was created with TRACE and initial core flow distribution and pressure drop in each segment of primary loop were reproduced using available data. In addition, a full-core model was built with the Serpent-2 Monte Carlo code to compute reactivity feedback parameters and delayed neutron information for point kinetics model in TRACE. Transient movement of sodium free level in Gas Expansion Modules (GEM) which was designed as a passive safety device of FFTF was sim-ulated with TRACE using a level tracking model. A good agreement between measured and calculated total reactivity indicated a reasonable validity of modeling of feedback effects and of predicted sodium level in GEM. Multi-dimensional thermal-hydraulics effects in the FFTF vessel especially thermal strati-fication phenomenon which was directly related to natural circulation flow rate in primary loop were simulated with three three-dimensional VESSEL components in TRACE. Transient evolution of sodium temperatures at the Post-Irradiation Open Test Assembly (PIOTA) outlet was predicted in a good agree-ment with the measurements. The need of a more accurate thermal-hydraulic simulation of the inter-assembly gaps corresponding to the fuel region was discovered to obviously improve the estimation of inter-assembly heat transfer. This study represented an important step towards the validation of the TRACE code to SFR and some suggestions for further development work are proposed. (c) 2021 Elsevier Ltd. All rights reserved.

Keyword :

Inter-assembly heat transfer Reactivity feedback Level tracking Natural circulation TRACE

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GB/T 7714 Wang, Shibao , Mikityuk, Konstantin , Dorde, Petrovic et al. Validation of TRACE capability to simulate unprotected transients in Sodium Fast Reactor using FFTF LOFWST Test #13 [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 164 .
MLA Wang, Shibao et al. "Validation of TRACE capability to simulate unprotected transients in Sodium Fast Reactor using FFTF LOFWST Test #13" . | ANNALS OF NUCLEAR ENERGY 164 (2021) .
APA Wang, Shibao , Mikityuk, Konstantin , Dorde, Petrovic , Zhang, Dalin , Su, Guanghui , Qiu, Suizheng et al. Validation of TRACE capability to simulate unprotected transients in Sodium Fast Reactor using FFTF LOFWST Test #13 . | ANNALS OF NUCLEAR ENERGY , 2021 , 164 .
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Capillary Characteristic Investigation of Porous Nano-Structured Stainless Steel Wire Mesh EI
期刊论文 | 2021 , 42 , 58-62 | Nuclear Power Engineering
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Abstract :

In order to fabricate the heat pipe wick structures with high capillary performance, and to support the design and manufacture the heat pipe with high power level, the oxidation and reduction method is adopted to fabricate the nano-structured stainless steel Dutch wire meshes, and their capillary characteristics are investigated in this study. The morphology of wire mesh is observed by SEM, and HSV and IR camera are adopted to record the droplet spreading and capillary rising processes respectively. The capillary performance coefficient is achieved based on capillary rising theory. Investigations show that the superhydrophilic nano-structures can be achieved when fabricated temperature is around 96. Nano-structures improve the capillary performance significantly, and the capillary performance factor can reach 4 mm. Copyright ©2021 Nuclear Power Engineering. All rights reserved.

Keyword :

Superhydrophilicity Heat pipes Stainless steel Wire Mesh generation Nanostructures

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GB/T 7714 Guo, Kailun , Wang, Chenglong , Zhang, Dalin et al. Capillary Characteristic Investigation of Porous Nano-Structured Stainless Steel Wire Mesh [J]. | Nuclear Power Engineering , 2021 , 42 : 58-62 .
MLA Guo, Kailun et al. "Capillary Characteristic Investigation of Porous Nano-Structured Stainless Steel Wire Mesh" . | Nuclear Power Engineering 42 (2021) : 58-62 .
APA Guo, Kailun , Wang, Chenglong , Zhang, Dalin , Qiu, Suizheng , Su, Guanghui , Tian, Wenxi . Capillary Characteristic Investigation of Porous Nano-Structured Stainless Steel Wire Mesh . | Nuclear Power Engineering , 2021 , 42 , 58-62 .
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CFD study on onset of liquid entrainment through ADS-4 branch line in AP1000 EI SCIE
期刊论文 | 2021 , 380 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

The consequences of liquid entrainment through the fourth stage Automatic Depressurization System (ADS-4) of Advanced Passive nuclear power reactor (AP1000) are decreased reactor core coolant inventory and increased resistance to quick depressurization of the reactor primary coolant system. Therefore, the accurate knowledge of the Onset of Liquid Entrainment (OLE) is critical to nuclear reactor safety (NRS) assessment. In this study, threedimensional computational fluid dynamics (CFD) simulations of OLE were carried out using a two-fluid Eulerian model of ANSYS Fluent CFD code along with the renormalization group (RNG) kappa - epsilon turbulence model for each phase. In OLE phenomena, the interfacial drag has a significant effect but available choices of interfacial drag coefficient in Fluent were not suitable due to the presence of droplet clusters in entrainment. Therefore, a modified interfacial drag coefficient accounting for droplet clusters was used. The transient simulations were performed with this modified CFD model and validated against ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) data. The comparison between CFD calculations and experimental data shows a good agreement. Furthermore, the effects of the liquid mass flow rates and single inlet on OLE were also investigated with the validated CFD model. To improve the understanding of OLE phenomena, the liquid volume fraction and gas velocity field distributions were also studied.

Keyword :

AP1000 ADS-4 CFD Nuclear reactor safety Onset of liquid entrainment

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GB/T 7714 Khan, Irfan , Wang, Mingjun , Zhang, Yapei et al. CFD study on onset of liquid entrainment through ADS-4 branch line in AP1000 [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 380 .
MLA Khan, Irfan et al. "CFD study on onset of liquid entrainment through ADS-4 branch line in AP1000" . | NUCLEAR ENGINEERING AND DESIGN 380 (2021) .
APA Khan, Irfan , Wang, Mingjun , Zhang, Yapei , Tian, Wenxi , Su, Guanghui , Qiu, Suizheng . CFD study on onset of liquid entrainment through ADS-4 branch line in AP1000 . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 380 .
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Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles EI
期刊论文 | 2021 , 55 (11) , 1930-1938 | Atomic Energy Science and Technology
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At present, the prediction of the critical heat flux (CHF) in the rod bundles is mostly based on the experimental correlations, which is limited by the specific application range, and cannot be effectively extrapolated or the prediction accuracy is reduced. In order to meet the prediction requirements of CHF in different light water reactors, it is necessary to develop a wide range of CHF prediction methods for different geometries and thermal boundaries. Based on the sub-channel analysis method, the two types of critical phenomena were considered in the paper, which were departure from nucleate boiling (DNB) and dryout, and the CHF in the bundles through coupling sub-channel analysis code and the CHF mechanistic model was calculated. By comparing with the experimental data of CHF, it is proved that the coupling code has better prediction accuracy for the CHF in the rod bundles. © 2021, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Light water reactors Nucleate boiling Heat flux Codes (symbols) Forecasting

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GB/T 7714 Gui, Minyang , Tian, Wenxi , Wu, Di et al. Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles [J]. | Atomic Energy Science and Technology , 2021 , 55 (11) : 1930-1938 .
MLA Gui, Minyang et al. "Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles" . | Atomic Energy Science and Technology 55 . 11 (2021) : 1930-1938 .
APA Gui, Minyang , Tian, Wenxi , Wu, Di , Chen, Ronghua , Zhang, Kui , Su, Guanghui et al. Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles . | Atomic Energy Science and Technology , 2021 , 55 (11) , 1930-1938 .
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Flow and Heat Transfer Characteristic of He-Xe Gas Mixture with Helical Wire Structure EI
期刊论文 | 2021 , 55 (6) , 991-999 | Atomic Energy Science and Technology
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In the design of the open grid gas-cooled space reactor core, the helical wires are used on the fuel rods, which will have great influence on the flow and heat transfer characteristics of the working mass. In this paper, the CFD method was used to analyze the flow and heat transfer characteristics of He-Xe gas mixture in an annulus wrapped with a helical wire and the import and export parameters spatial distributions such as temperature, pressure, flow rate and fluid density were obtained. The results show that the introduction of helical wire greatly increases the Fanning friction factor, and some helical wire structures will reduce the Nusselt number by 20%-30%. Moreover, the thermal-hydraulic performance ratios of the five helical wire structures studied in this paper are all less than 1. The research results are of great significance to the thermal design of gas-cooled space reactor core and to improve the safety of the system. © 2021, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Reactor cores Gases Wire Gas mixtures Computational fluid dynamics Heat transfer

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GB/T 7714 Chen, Siyuan , Qin, Hao , Wang, Chenglong et al. Flow and Heat Transfer Characteristic of He-Xe Gas Mixture with Helical Wire Structure [J]. | Atomic Energy Science and Technology , 2021 , 55 (6) : 991-999 .
MLA Chen, Siyuan et al. "Flow and Heat Transfer Characteristic of He-Xe Gas Mixture with Helical Wire Structure" . | Atomic Energy Science and Technology 55 . 6 (2021) : 991-999 .
APA Chen, Siyuan , Qin, Hao , Wang, Chenglong , Zhang, Yapei , Zhang, Dalin , Qiu, Suizheng et al. Flow and Heat Transfer Characteristic of He-Xe Gas Mixture with Helical Wire Structure . | Atomic Energy Science and Technology , 2021 , 55 (6) , 991-999 .
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Analysis of Startup Characteristics for Alkali Metal High Temperature Heat Pipe EI
期刊论文 | 2021 , 55 (6) , 1015-1023 | Atomic Energy Science and Technology
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The startup and operation transients of alkali metal high temperature heat pipes were studied in this paper. Based on the traditional heat pipe frozen startup model, a three-stage model depending on vapor flow transition was established, and a heat pipe startup transient analysis code, HPSTAC, was developed. The relative deviation between the code simulation result and experimental value is less than 15.7%. The HPSTAC code was used to perform the startup transient of single sodium-potassium heat pipe and sensitivity analysis. The results show that 450, 660 and 1 550 s after the starting, the heat pipe enters the second, third, and quasi-steady states respectively, and the vapor region tends to a consistent temperature of 837 K. The startup surrounding temperature mainly affects the heating rate of the condensation section, and there is a threshold on the heat flux influence of stage process during heat pipe startup. © 2021, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Sensitivity analysis Heat pipes Heat flux Metal analysis Transient analysis

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GB/T 7714 Zhang, Zeqin , Wang, Chenglong , Sun, Hao et al. Analysis of Startup Characteristics for Alkali Metal High Temperature Heat Pipe [J]. | Atomic Energy Science and Technology , 2021 , 55 (6) : 1015-1023 .
MLA Zhang, Zeqin et al. "Analysis of Startup Characteristics for Alkali Metal High Temperature Heat Pipe" . | Atomic Energy Science and Technology 55 . 6 (2021) : 1015-1023 .
APA Zhang, Zeqin , Wang, Chenglong , Sun, Hao , Zhang, Dalin , Qiu, Suizheng , Su, Guanghui et al. Analysis of Startup Characteristics for Alkali Metal High Temperature Heat Pipe . | Atomic Energy Science and Technology , 2021 , 55 (6) , 1015-1023 .
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Analysis of Thermal-hydraulic Characteristic of Solid Heat Pipe Reactor Simulator Device EI
期刊论文 | 2021 , 55 (6) , 984-990 | Atomic Energy Science and Technology
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Heat pipe reactor has the advantages of small size, simple structure and high inherent safety, which has a wide application prospect and research significance. In this paper, CFD method was used to analyze the thermal-hydraulic characteristics of a heat pipe reactor simulator device under steady state conditions, and the results were compared with the experimental results. The results show that the relative temperature error of each measuring point of the heat pipe is less than 5.5%, and the relative temperature error of each measuring point at the hot end of the thermoelectric generator is less than 3.1%. This study provides theoretical guidance and method support for numerical simulation of heat pipe reactor. © 2021, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Temperature measurement Numerical methods Buoyancy Thermoelectric energy conversion Heat pipes Thermoelectric equipment

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GB/T 7714 Zhang, Yin , Wang, Chenglong , Tang, Simiao et al. Analysis of Thermal-hydraulic Characteristic of Solid Heat Pipe Reactor Simulator Device [J]. | Atomic Energy Science and Technology , 2021 , 55 (6) : 984-990 .
MLA Zhang, Yin et al. "Analysis of Thermal-hydraulic Characteristic of Solid Heat Pipe Reactor Simulator Device" . | Atomic Energy Science and Technology 55 . 6 (2021) : 984-990 .
APA Zhang, Yin , Wang, Chenglong , Tang, Simiao , Li, Jian , Zhang, Dalin , Qiu, Suizheng et al. Analysis of Thermal-hydraulic Characteristic of Solid Heat Pipe Reactor Simulator Device . | Atomic Energy Science and Technology , 2021 , 55 (6) , 984-990 .
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