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学者姓名:田文喜

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Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor Scopus
会议论文 | 2018 , 9 | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Abstract :

Copyright © 2018 ASME In sodium-cooled fast reactor (SFR), thermal gradient is the paramount factor of assembly transient bowing, that may cause great reactivity change, accelerate wrapper vibration wear, hindering the motion of control/shutdown rods, or worse yet, threatening the integrity of assemblies. However, because of the complexity of multi-assembly contact and interaction problem, it is difficult to assess the impact of core deformation on reactor performance safety. The Core Assembly Deformation Test Facility (CADTF) is designed to perform a series of thermal bowing tests by Xi`an Jiao Tong University (XJTU) to investigate the core deformation behaviors under thermal gradient. In this paper, a finite element model was established to simulate the mechanical response of single assembly under different flat-to-flat thermal gradient. The single assembly restrained bowing test performed in CADTF is chosen to validate the model. In the model, the measured temperature distribution as well as temperature-dependent elastoplastic and thermal expansion properties were taken into consideration. To ensure the model reliability, iterative computation is conducted by adjusting the friction coefficient of the load pads to match the calculated and measured contact force. According to the results, it can be seen that the three-dimensional displacement of assembly shows relatively good agreement with the experimental data. Therefore, it can be concluded that the model is capable of performing core deformation analysis for SFR.

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GB/T 7714 Tian, Wenxi , Qiu, Suizheng . Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor [C] . 2018 .
MLA Tian, Wenxi 等. "Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor" . (2018) .
APA Tian, Wenxi , Qiu, Suizheng . Numerical and experimental investigation on core assembly thermal-gradient-induced deformation of sodium-cooled fast reactor . (2018) .
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Pressure drop experiments of liquid sodium flowing in a 7-rod bundle Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME Pressure drop experiments was conducted for liquid sodium in an electrically heated 7-rod bundle. The electrically heated 7-rod bundle was placed in a hexagonal tube. In the experiment,the heat flux ranges from 0~300 kw·m-2,mass velocity from 40~450 kg·m-2·s-1, system pressure from 10~200 KPa and the average temperature of liquid sodium from 350~650℃.The effects of the heat flux, system pressure and the average temperature of liquid sodium on the pressure drop was in-depth analyzed. A new correlation for pressure drop was developed based on the experimental data of liquid sodium in a 7-rod bundle.

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GB/T 7714 Tian, Wenxi . Pressure drop experiments of liquid sodium flowing in a 7-rod bundle [C] . 2018 .
MLA Tian, Wenxi . "Pressure drop experiments of liquid sodium flowing in a 7-rod bundle" . (2018) .
APA Tian, Wenxi . Pressure drop experiments of liquid sodium flowing in a 7-rod bundle . (2018) .
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Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation Scopus
会议论文 | 2018 , 5
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Copyright © 2018 ASME. In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.

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GB/T 7714 Tian, Wenxi . Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation [C] . 2018 .
MLA Tian, Wenxi . "Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation" . (2018) .
APA Tian, Wenxi . Effect of rolling motion on flow instability of parallel rectangular channels of natural circulation . (2018) .
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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.

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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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Conceptual design and neutronics/thermal-hydraulic coupling optimization analyses of two typical helium cooled solid breeder blanket modules for cfetr phase II Scopus
会议论文 | 2018 , 5
WoS CC Cited Count: 1
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Copyright © 2018 ASME. Chinese Fusion Engineering Test Reactor (CFETR) is a new test Tokamak device which is now being designed in China to make the transition from the International Thermonuclear Experimental Reactor (ITER) to the future Fusion Power Plant (FPP). Breeding blanket is the key component of fusion reactor which is mainly responsible for the tritium self-sufficiency and fusion energy conversion. In the past few years, three kinds of blanket conceptual design schemes have been proposed and tested in parallel for CFETR Phase I, in which the helium cooled solid breeder (HCSB) blanket concept is acknowledged as the most promising one. However, nowadays, the design phase of CFETR has gradually changed from I to II aiming for the future DEMO operation condition, the main parameters of which are quite different from the previous one. Therefore, it’s necessary to perform conceptual design and various analyses for the HCSB blanket under the new working condition. In this work, firstly, a new conceptual design scheme of HCSB blanket for Phase II is put forward. Then, the radial build arrangements, of the two typical blanket modules are optimized by using the NTCOC. This work can provide valuable references for further conceptual design and neutronics/thermal-hydraulic coupling analyses of the HCSB blanket for CFETR Phase II.

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GB/T 7714 Tian, Wenxi . Conceptual design and neutronics/thermal-hydraulic coupling optimization analyses of two typical helium cooled solid breeder blanket modules for cfetr phase II [C] . 2018 .
MLA Tian, Wenxi . "Conceptual design and neutronics/thermal-hydraulic coupling optimization analyses of two typical helium cooled solid breeder blanket modules for cfetr phase II" . (2018) .
APA Tian, Wenxi . Conceptual design and neutronics/thermal-hydraulic coupling optimization analyses of two typical helium cooled solid breeder blanket modules for cfetr phase II . (2018) .
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Hydraulic characteristics research on SG under tube plugging operations using Fluent Scopus
会议论文 | 2018 , 9 | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Copyright © 2018 ASME Steam Generator (SG) is a critical equipment in the nuclear power plant, it is the huge heat exchanger in reactor system which can achieve removing fission energy from the reactor system effectively to ensure safety of the whole nuclear system. It is located between the primary and the secondary loop in reactor system act as the intermediate hub of energy and the security barrier in nuclear power plant. Generally, there are numerous of U-shaped heat transfer tubes in SG, it is one of the weakest structures throughout the primary loop system. So the integrity of the SG especially its heat transfer tubes is important to the safety of reactor operation. The degradation problem of heat transfer tubes together with ruptures accidents often occur under suffer environments in reactors, which include thermal stress, mechanical stress and so on, it is noteworthy that this kind of accidents is inevitable due to the limited properties of existing materials. The performance of the SG is seriously affected by the number of failure tubes. Plugging operations through various mechanical means is the most common method to solve the tubes ruptures problems which can reduce the economic losses to the utmost extent. However, plugging operations will make huge impact on the thermal hydraulic performances of both sides of SG. It's meaningful to research the characteristics of the plugging affects under different operations. In this paper the hydraulic characteristics of primary side in AP1000 SG under a certain fraction of heat transfer tube plugging conditions is researched. Three dimensional hydraulic characteristics of primary side coolant in SG under different plugging conditions are obtained by using the thermal hydraulic software FLUENT. The typical plugging fraction in this simulation model is 10 percent, and the effect of plugging locations also be considered through changing the plugging positions using the zone marking method. The results shows that the pressure drop under the structure integrated SG is 358.01MPa which is accordance with the results from Westinghouse 343KPa. The pressure drop values varies when changing positions of the plugging tubes under the same plugging fraction condition. The flow fields in bottom head also change meanwhile and the maximum pressure drop can reach up to 388.05KPa when the plugging fraction is 10%. The growth rate become significant when tube plugging fraction larger than 5%, and differences between maximum and minimum values of total pressure drop under different plugging positions become larger gradually. Finally the local resistance coefficients and flow field distributions of primary side in SG under various plugging conditions are obtained which is meaningful for the reactor safety and it can be a good reference for the maintenance of SG.

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GB/T 7714 Tian, Wenxi , Su, G. H. , Qiu, Suizheng . Hydraulic characteristics research on SG under tube plugging operations using Fluent [C] . 2018 .
MLA Tian, Wenxi et al. "Hydraulic characteristics research on SG under tube plugging operations using Fluent" . (2018) .
APA Tian, Wenxi , Su, G. H. , Qiu, Suizheng . Hydraulic characteristics research on SG under tube plugging operations using Fluent . (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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Development of boron dilution model in COBRA-EN Scopus
会议论文 | 2018 , 4 | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Copyright © 2018 ASME The dilution characteristic of concentrated boron in reactor core is very important for reactivity control during the Small Break Loss of Coolant Accident (SB-LOCA). Nonuniform distribution of the boron concentration may increase the risk of reactivity accidents. However, tracking the boron concentration inside the primary coolant system meets lots of challenges. Hence the development of a thermal-hydraulic code with accurate boron tracking model is necessary. This paper presents a multidimensional boron transport model based on subchannel software COBRA-EN. The original model gives a poor prediction in tracking the boron concentration of high Re, since it only considers the mass exchange caused by pressure difference and ignores the mass exchange put forward by the turbulent diffusion. In this paper, a new tracking model was proposed and implemented in the COBRA-EN code, in which the turbulent diffusion was considered. Then a 3×3 fuel bundle model was adopted to verify the proposed model. The CFD study was also conducted to investigate the feasibility of the new tracking model. The results indicate: (1) the modified boron tracking model are more close to the actual than the original model; (2) the modified model was verified by FLUENT; (3) the model shows defects at low Re due to the ignorance of molecule diffusion.

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GB/T 7714 Qiu, Suizheng , Tian, Wenxi , Su, G. H. . Development of boron dilution model in COBRA-EN [C] . 2018 .
MLA Qiu, Suizheng et al. "Development of boron dilution model in COBRA-EN" . (2018) .
APA Qiu, Suizheng , Tian, Wenxi , Su, G. H. . Development of boron dilution model in COBRA-EN . (2018) .
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Experimental study on spray pattern of pressure-swirl nozzle in reactor containment Scopus
会议论文 | 2018 , 9 | 2018 26th International Conference on Nuclear Engineering, ICONE 2018
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Copyright © 2018 ASME Spraying system plays an important role in the safety of PWR. To ensure homogeneous spraying of the containment, the layout of nozzles on the spray header was taken into consideration in design. In this paper, an experimental study was conducted to obtain spray characteristics data, including spray cone angle and 2-D spray flux distribution for the purpose of achieving optimal design of the spraying system. According to the specialty of the spray field involved, a testing loop with four pressure-swirl nozzles was established for the study. Spray cone angles were obtained by photograph method. The volume flux distribution was measured by collecting the spray droplet along the cross-section diameters. Based on the experimental data, typical spray flux distributions were obtained. The flux distribution results were used to build 3-D coverage models. Then these models were used to calculate the overall spray coverage in the containment. The present work introduces the experimental study of spray behavior of a typical pressure-swirl nozzle in containment and the method to evaluate spray coverage through building 3-D spray flux distribution models. The work is expected to be helpful for the optimization design of spraying systems.

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GB/T 7714 Tian, Wenxi , Suizheng, Q. I.U. . Experimental study on spray pattern of pressure-swirl nozzle in reactor containment [C] . 2018 .
MLA Tian, Wenxi et al. "Experimental study on spray pattern of pressure-swirl nozzle in reactor containment" . (2018) .
APA Tian, Wenxi , Suizheng, Q. I.U. . Experimental study on spray pattern of pressure-swirl nozzle in reactor containment . (2018) .
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Optimization of moving particle semi-implicit (MPS) method and studying of flow instability Scopus
会议论文 | 2018 , 8
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Copyright © 2018 ASME. Multiphase flow widely exists in the nature and engineering. The two-phase flow is the highlight of the studies about the flow in the vessel and steam explosion in nuclear severe accidents. The Moving Particle Semi-implicit (MPS) method is a fully-Lagrangian particle method without grid mesh which focuses on tracking the single particle and concerns with its movement. It has advantages in tracking complex multiphase flows compared with gird methods, and thus shows great potential in predicting multiphase flows. The objective of this thesis is to develop a general multiphase particle method based on the original MPS method and thus this work is of great significance for improving the numerical method for simulating the instability in reactor severe accident and two-phase flows in vessel. This research is intended to provide a study of the instability based on the MPS method. Latest achievements of mesh-free particle methods in instability are researched and a new multiphase MPS method, which is based on the original one, for simulating instability has been developed and validated. Based on referring to other researchers’ papers, the Pressure Poisson Equation(PPE), the viscosity term, the free surface particle determination part and the surface tension model are optimized or added. The numerical simulation on stratification behavior of two immiscible flows is carried out and results are analyzed after data processing. It is proved that the improved MPS method is more accurate than the original method in analysis of multiphase flows. In this paper, the main purposes are simulating and discussing Rayleigh-Taylor (R-T) instability and Kelvin-Helmholtz (K-H) instability. R-T and K-H instability play an important role in the mixing process of many layered flows. R-T instability occurs when a lower density fluid is supported by another density higher fluid or higher density fluid is accelerated by lower density fluid, and the resulting small perturbation increases and eventually forms turbulence. K-H instability is a small disturbance for two different densities, such as waves, at the interface of the two-phase fluid after giving a fixed acceleration in the fluid. Turbulence generated by R-T instability and K-H instability has an important effect in applications such as astrophysics, geophysics, and nuclear science.

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GB/T 7714 Guo, Kailun , Chen, Ronghua , Qiu, Suizheng et al. Optimization of moving particle semi-implicit (MPS) method and studying of flow instability [C] . 2018 .
MLA Guo, Kailun et al. "Optimization of moving particle semi-implicit (MPS) method and studying of flow instability" . (2018) .
APA Guo, Kailun , Chen, Ronghua , Qiu, Suizheng , Tian, Wenxi , Su, Guanghui , Wu, Junmei . Optimization of moving particle semi-implicit (MPS) method and studying of flow instability . (2018) .
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