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学者姓名:秋穗正

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< Page ,Total 89 >
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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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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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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Energy allocation optimization of the gas-cooled space nuclear reactor EI SCIE
期刊论文 | 2021 , 196 | APPLIED THERMAL ENGINEERING
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Abstract :

The development of the space vehicles puts forward higher requirements on the energy supply, and the megawatt gas-cooled space nuclear reactors are promising to satisfy the demand. The energy allocation optimization on the space nuclear reactors takes critical part in the system design. The optimization between the energy conversion efficiency and radiator mass of the gas-cooled space nuclear reactor is conducted and analyzed. The models of direct Brayton cycle, heat pipe radiator, and liquid droplet radiator are established. The working performance of the thermodynamic cycle coupled with heat pipe radiator (HPR) and liquid droplet radiator (LDR) are obtained respectively, and compared. The area and weight of the HPR increases linearly with the radiant power. The calculation results show that, for space nuclear reactors adopting HPR, decreasing the coolant temperature at the reactor core outlet (T-1) from 1500 K to 1200 K will increase the specific surface area from 1200 m(2)/MW to about 3000 m(2)/MW with the same energy conversion efficiency. The radiant power of the LDR can be regulated by operation mode with the radiator weight remains unchanged. The mass of LDR is only about 10% of HPR for case with electricity power P-e = 0.5 MW, T-1 = 1500 K, showing significant advantage in the mass optimization. This paper may contribute to the energy management and allocation optimization of the gas-cooled space nuclear reactor.

Keyword :

Liquid droplet radiator Energy allocation optimization direct Brayton cycle Gas-cooled space nuclear reactor Heat pipe radiator

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GB/T 7714 Qin, Hao , Wang, Chenglong , Tian, Wenxi et al. Energy allocation optimization of the gas-cooled space nuclear reactor [J]. | APPLIED THERMAL ENGINEERING , 2021 , 196 .
MLA Qin, Hao et al. "Energy allocation optimization of the gas-cooled space nuclear reactor" . | APPLIED THERMAL ENGINEERING 196 (2021) .
APA Qin, Hao , Wang, Chenglong , Tian, Wenxi , Qiu, Suizheng , Su, Guanghui . Energy allocation optimization of the gas-cooled space nuclear reactor . | APPLIED THERMAL ENGINEERING , 2021 , 196 .
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Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER EI SCIE
期刊论文 | 2021 , 383 | NUCLEAR ENGINEERING AND DESIGN
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The heat pipe reactor is featured with the passive heat discharge, anti-single point failure and silent operation. Therefore, it could be applied for the power system of underwater vehicles. In this paper, the concept of Unattended Uranium-fueled Portable Reactor (UPR) for the NUclear Silence ThermoElectric Reactor (NUSTER) power system is introduced and the reactor core with a thermal power of 1 MW is neutronics designed. In the core, highly enriched UO2 fuel, sodium heat pipe, molybdenum alloy matrix, beryllium oxide and beryllium reflector are adopted and the reactivity is controlled by four sliding reflectors and four safety rods. In design process, the NECP-MCX code is used to carry out the neutronics calculation. At the same time, the heating unit, core loading, control system and power distribution of the core are described. Furthermore, criticality safety, burnup characteristics, power distribution of the core are specifically analyzed. The results show that the proposed core design meets the requirements of criticality safety and operating life.

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GB/T 7714 Du, Xianan , Tao, Yushan , Zheng, Youqi et al. Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 383 .
MLA Du, Xianan et al. "Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER" . | NUCLEAR ENGINEERING AND DESIGN 383 (2021) .
APA Du, Xianan , Tao, Yushan , Zheng, Youqi , Wang, Chenglong , Wang, Yongping , Qiu, Suizheng et al. Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 383 .
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Experimental and numerical investigation on characteristics of MCCI with exothermic thermite EI SCIE
期刊论文 | 2021 , 384 | NUCLEAR ENGINEERING AND DESIGN
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Investigation was conducted experimentally and numerically for ex-vessel melt behavior and concrete interaction during Molten Corium-Concrete Interaction (MCCI) in current research. The CINA (Corium-Concrete Interaction Apparatus) experiment was carried out with a total of 94 kg melt generated by exothermic thermite chemical reaction to react in a two-dimensional (2-D) cylindrical siliceous crucible with an inner diameter of 300 mm and a height of 500 mm. Meanwhile, the variation of melt temperature was monitored during the experiment. Besides, the accurate ablation depth and thickness of stratified layers were measured after the experiment. Numerical simulation was used to study the mechanism of high-temperature melt heat transfer and concrete ablation during the MCCI process. The slag film model to calculate the melt/concrete interface heat transfer coefficient and the stratified melt model were both used to simulate the above experiment. The results showed that the numerical results were in good agreement with experiment measurements by the slag film model under separated layers condition.

Keyword :

Melt stratification MCCI Severe accident Concrete ablation

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GB/T 7714 Xu, Zhichun , Zhang, Yapei , Zhang, Kui et al. Experimental and numerical investigation on characteristics of MCCI with exothermic thermite [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 384 .
MLA Xu, Zhichun et al. "Experimental and numerical investigation on characteristics of MCCI with exothermic thermite" . | NUCLEAR ENGINEERING AND DESIGN 384 (2021) .
APA Xu, Zhichun , Zhang, Yapei , Zhang, Kui , Wu, Zijie , Zhan, Dekui , Su, G. H. et al. Experimental and numerical investigation on characteristics of MCCI with exothermic thermite . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 384 .
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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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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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CFD modeling of liquid entrainment through vertical T-junction of fourth stage automatic depressurization system (ADS-4) EI SCIE
期刊论文 | 2021 , 159 | ANNALS OF NUCLEAR ENERGY
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In this study, a three-dimensional (3D) transient computational fluid dynamics (CFD) model is presented to investigated liquid entrainment in gas-liquid flow through vertical T-junction of fourth stage automatic depressurization system (ADS-4) in AP1000. A specialized CFD model for the liquid entrainment phenomenon study, which is based on coupled Eulerian-Eulerian VOF (volume of fluid) formulation with suitable interfacial drag and standard kappa - epsilon turbulence model for each phase was proposed. The entrainment rate was calculated and results were validated with ADETEL experimental data, which was built at XJTU-NuTheL. The good agreement between CFD calculations and experimental data demonstrated that the proposed CFD model could reasonably predict entrainment within the studied range. The effect of vertical to horizontal branch diameter ratio and gas mass flow rates on liquid entrainment were also studied. In addition, the liquid volume fraction distributions and velocity field were investigated to develop an understanding of the entrainment process. The relatively high demand for computational resources due to very small timestep size and small grid size to accommodate high flow velocities was found to be a challenge. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

CFD Eulerian-Eulerian VOF ADS-4 T-junction Liquid entrainment

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GB/T 7714 Khan, Irfan , Wang, Mingjun , Basit, Muhammad Abdul et al. CFD modeling of liquid entrainment through vertical T-junction of fourth stage automatic depressurization system (ADS-4) [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 159 .
MLA Khan, Irfan et al. "CFD modeling of liquid entrainment through vertical T-junction of fourth stage automatic depressurization system (ADS-4)" . | ANNALS OF NUCLEAR ENERGY 159 (2021) .
APA Khan, Irfan , Wang, Mingjun , Basit, Muhammad Abdul , Tian, Wenxi , Su, Guanghui , Qiu, Suizheng . CFD modeling of liquid entrainment through vertical T-junction of fourth stage automatic depressurization system (ADS-4) . | ANNALS OF NUCLEAR ENERGY , 2021 , 159 .
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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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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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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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