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学者姓名:田文喜

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< Page ,Total 58 >
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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Vapor condensation with low air mass fraction inside water seal branch tube under free convection EI SCIE
期刊论文 | 2021 , 379 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

The pressure vessel has used a newly designed water seal structure, and the sealed water is from condensation at low hydrogen concentration. To get the condensation capacity of the water seal branch tube under the pressure vessel normal operation, the experimental loop named WASETEL was constructed, which is aimed at studying the heat transfer during the process of air-steam mixture condensation inside the inclined branch tube at the condition of low air mass fraction. The inner tube is diameter 0.134 m with length 0.68 m which is connected with diameter 1.4 m, high 2.5 m pressure vessel. A canon camera was set to record the images and videos of the condensate height through glass tube in communicating vessel liquidometer. The experiment was conducted under the pressure ranged from 0.2 to 0.6 MPa, air mass concentration ranged from 0.49 to 6.42% and wall subcooling ranged from 11 to 40 K. The test showed HTC (heat transfer coefficient) would sharply deteriorate with the small air fraction, compared with pure steam condensation, about 0.49% mass fraction of the air will reduce heat transfer coefficient by 70-80%. When air concentration is relatively small, those widely used relations can not predict the condensation well, therefore, a new heat transfer empirical correlation with air mass fraction, pressure and subcooling is proposed, covered all the experimental data within the error band +/- 13%.

Keyword :

New empirical correlation Natural condensation Water seal

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GB/T 7714 Tan, Bing , Wang, Xi , Wu, Y. W. et al. Vapor condensation with low air mass fraction inside water seal branch tube under free convection [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 379 .
MLA Tan, Bing et al. "Vapor condensation with low air mass fraction inside water seal branch tube under free convection" . | NUCLEAR ENGINEERING AND DESIGN 379 (2021) .
APA Tan, Bing , Wang, Xi , Wu, Y. W. , Tian, W. X. , Qiu, S. Z. , Su, G. H. . Vapor condensation with low air mass fraction inside water seal branch tube under free convection . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 379 .
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Performance analysis of PRHRS in primary and secondary circuit for offshore floating nuclear plant EI SCIE
期刊论文 | 2021 , 164 | ANNALS OF NUCLEAR ENERGY
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Abstract :

In order to conduct performance analysis of passive residual heat removal system (PRHRS), a system analysis code applicable to offshore floating nuclear plant (OFNP) was developed based on existing code by adding motion condition model and the OFNP system with PRHRS in primary and secondary circuit was modeled in this study. The station blackout (SBO) accident was calculated under stationary and motion conditions with PRHRS in primary and secondary circuit operating respectively to compare the performance of the two PRHRS. The results show that the heat removal capability of PRHRS in secondary circuit is better than that in primary circuit with the same arrangement. The influence of heaving condi-tion on the primary PRHRS is greater than the secondary PRHRS, while the secondary PRHRS is more sus-ceptible to trimming and pitching conditions and heeling and rolling conditions have little influence on both of two PRHRS. (c) 2021 Published by Elsevier Ltd.

Keyword :

(PRHRS) Passive residual heat removal system&nbsp Offshore floating nuclear plant (OFNP) Station blackout (SBO) accident Motion conditions

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GB/T 7714 Ma, Yichao , Zhang, Jing , Wang, Mingjun et al. Performance analysis of PRHRS in primary and secondary circuit for offshore floating nuclear plant [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 164 .
MLA Ma, Yichao et al. "Performance analysis of PRHRS in primary and secondary circuit for offshore floating nuclear plant" . | ANNALS OF NUCLEAR ENERGY 164 (2021) .
APA Ma, Yichao , Zhang, Jing , Wang, Mingjun , Tian, Wenxi , Chen, Ronghua , Wu, Yingwei et al. Performance analysis of PRHRS in primary and secondary circuit for offshore floating nuclear plant . | ANNALS OF NUCLEAR ENERGY , 2021 , 164 .
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CFD simulation of thermal hydraulic phenomena in enclosed cavity of nuclear power plants EI SCIE
期刊论文 | 2021 , 151 | Annals of Nuclear Energy
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The 'dead-leg' in nuclear power plants (NPP) are pipes, which are connected with the primary loop and isolated by two check valves setting at both ends. It is a typical enclosed cavity. As the temperature and pressure in primary side is higher, the coolant in tube is heated by the valve and becoming two phase state, resulting in the corrosion of valves. In present work, the CFD method was employed to study the single phase and two phase thermal hydraulic phenomena in enclosed cavity. The buoyancy was calculated by solving compressible governing equations, and the mixture multiphase model and improved compressible phase change model were used to calculate the boiling process. The models are validated against with the experiment data and the simulation results show great agreement. The coolant natural convection occurs and the significant thermal stratification exists in the vertical direction. The void fraction increases at first and tends to a stable value finally. The vapor accumulates in the upper part of the pipe and the corrosion in the region is more serious. The trend of pressure increasing is consistent with void fraction and self-pressurization phenomenon is caused by the compression of vapor. Increasing the initial pressure could reduce boiling process greatly and it is an effective means to alleviate 'dead-leg' phenomenon. This work is meaningful for the deep understanding of thermal hydraulic phenomena in enclosed cavity and provides valuable guidance for engineering. © 2020 Elsevier Ltd

Keyword :

Void fraction Nuclear fuels Thermal Engineering Coolants Nuclear energy Pipeline corrosion Nuclear power plants Two phase flow

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GB/T 7714 Chen, Chong , Wang, Xingjun , Wang, Mingjun et al. CFD simulation of thermal hydraulic phenomena in enclosed cavity of nuclear power plants [J]. | Annals of Nuclear Energy , 2021 , 151 .
MLA Chen, Chong et al. "CFD simulation of thermal hydraulic phenomena in enclosed cavity of nuclear power plants" . | Annals of Nuclear Energy 151 (2021) .
APA Chen, Chong , Wang, Xingjun , Wang, Mingjun , Qi, Yubo , Zhang, Jing , Tian, Wenxi et al. CFD simulation of thermal hydraulic phenomena in enclosed cavity of nuclear power plants . | Annals of Nuclear Energy , 2021 , 151 .
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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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Neurtonics/Thermal-hydraulic analyses of the CFETR HCCB blanket for multiple operation modes under the poloidal nonuniform neutron wall loading condition EI SCIE
期刊论文 | 2021 , 168 | FUSION ENGINEERING AND DESIGN
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Helium Cooled Ceramic Breeder (HCCB) blanket is one significant component of China Fusion Engineering Test Reactor (CFETR). The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation for CFETR, and both of them are influenced by the neutron wall loading (NWL) distribution on the blanket. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the CFETR HCCB blanket have been preliminarily investigated. However, as the nuclear heating rate in each blanket component is adopted as the internal heat source for the thermal-hydraulic analyses, it's necessary to further investigate the influences of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. Besides, CFETR aims to cover multiple fusion powers (from 200 MW to 1.5 GW) at the present stage. Therefore, it's also essential to compare the different influences of the poloidal nonuniform NWL on blanket performances for multiple operation modes.

Keyword :

Neurtonics Thermal-hydraulic CFETR Poloidal nonuniform NWL Multiple operation modes HCCB blanket

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GB/T 7714 Cui, Shijie , Lang, Yueheng , Lian, Qiang et al. Neurtonics/Thermal-hydraulic analyses of the CFETR HCCB blanket for multiple operation modes under the poloidal nonuniform neutron wall loading condition [J]. | FUSION ENGINEERING AND DESIGN , 2021 , 168 .
MLA Cui, Shijie et al. "Neurtonics/Thermal-hydraulic analyses of the CFETR HCCB blanket for multiple operation modes under the poloidal nonuniform neutron wall loading condition" . | FUSION ENGINEERING AND DESIGN 168 (2021) .
APA Cui, Shijie , Lang, Yueheng , Lian, Qiang , Wan, Haoyu , Zhang, Dalin , Tian, Wenxi et al. Neurtonics/Thermal-hydraulic analyses of the CFETR HCCB blanket for multiple operation modes under the poloidal nonuniform neutron wall loading condition . | FUSION ENGINEERING AND DESIGN , 2021 , 168 .
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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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