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Featured researches published by Su Guanghui.


Nuclear Engineering and Design | 2002

Theoretical and experimental study on density wave oscillation of two-phase natural circulation of low equilibrium quality

Su Guanghui; Jia Dou-nan; Kenji Fukuda; Guo Yujun

A theoretical and experimental study of density wave oscillation (DWO) in natural circulation is presented in this paper. Experiments were performed on a natural circulation test facility. The influences of mass flow rate, pressure, inlet subcooling, heat flux and exit quality on DWO were analyzed. The marginal stability boundary (MSB) of DWO was obtained. A criterion of two-phase natural circulation, which predicts the stability thresholds, was developed by lumped parameter method. It is a function of non-dimensional parameters, such as phase change number Npch, subcooling number Nsub, Froude number, Fr, geometry number Nl and friction number τ. The geometry number and friction number are first defined in this paper. A correlation of DWO period was also obtained, it is also a function of the above non-dimensional parameters. The results of the present criterion and period correlation were compared with those of the experimental data and references. It is shown that they agree very well.


Nuclear Engineering and Design | 2003

Analysis of the critical heat flux in round vertical tubes under low pressure and flow oscillation conditions. Applications of artificial neural network

Su Guanghui; Koji Morita; Kenji Fukuda; Mark Pidduck; Jia Dou-nan; Jaakko Miettinen

Artificial neural networks (ANNs) for predicting critical heat flux (CHF) under low pressure and oscillation conditions have been trained successfully for either natural circulation or forced circulation (FC) in the present study. The input parameters of the ANN are pressure, mean mass flow rate, relative amplitude, inlet subcooling, oscillation period and the ratio of the heated length to the diameter of the tube, L/D. The output is a nondimensionalized factor F, which expresses the relative CHF under oscillation conditions. Based on the trained ANN, the influences of principal parameters on F for FC were analyzed. The parametric trends of the CHF under oscillation obtained by the trained ANN are as follows: the effects of pressure below 500 kPa are complex due to the influence of other parameters. F will increase with increasing mean mass flow rate under any conditions, and will decrease generally with an increase in relative amplitude. F will decrease initially and then increase with increasing inlet subcooling. The influence curves of mean mass flow rate on F will be almost the same when the period is shorter than 5.0 s or longer than 15 s. The influence of L/D will be negligible if L/D>200. It is found that the minimum number of neurons in the hidden layer is a product of the number of neurons in the input layer and in the output layer.


Journal of Nuclear Science and Technology | 2002

Application of an Artificial Neural Network in Reactor Thermohydraulic Problem: Prediction of Critical Heat Flux

Su Guanghui; Kenji Fukuda; Dounan Jia; Koji Morita

A new method for predicting Critical Heat Flux (CHF) with the Artificial Neural Network (ANN) method is presented in this paper. The ANNs were trained based on three conditions: type I (inlet or upstream conditions), II (local or CHF point conditions) and III (outlet or downstream conditions). The best condition for predicting CHF is type II, providing an accuracy of ±10%. The effects of main parameters such as pressure, mass flow rate, equilibrium quality and inlet subcooling on CHF were analyzed using the ANN. Critical heat flux under oscillation flow conditions was also predicted.


Journal of Nuclear Science and Technology | 2002

Applications of Artificial Neural Network for the Prediction of Flow Boiling Curves

Su Guanghui; Kenji Fukuda; Koji Morita; Mark Pidduck; Dounan Jia; Tatsuya Matsumoto; Ryo Akasaka

An artificial neural network (ANN) was applied successfully to predict flow boiling curves. The databases used in the analysis are from the 1960s, including 1,305 data points which cover these parameter ranges: pressure P=100–1,000 kPa, mass flow rate G=40–500 kg/m2-s, inlet subcooling ΔTsub =0–35°C, wall superheat ΔTw = 10–300°C and heat flux Q=20–8,000kW/m2. The proposed methodology allows us to achieve accurate results, thus it is suitable for the processing of the boiling curve data. The effects of the main parameters on flow boiling curves were analyzed using the ANN. The heat flux increases with increasing inlet subcooling for all heat transfer modes. Mass flow rate has no significant effects on nucleate boiling curves. The transition boiling and film boiling heat fluxes will increase with an increase in the mass flow rate. Pressure plays a predominant role and improves heat transfer in all boiling regions except the film boiling region. There are slight differences between the steady and the transient boiling curves in all boiling regions except the nucleate region. The transient boiling curve lies below the corresponding steady boiling curve.


Nuclear Science and Techniques | 2008

Steady thermal hydraulic analysis for a molten salt reactor

Dalin Zhang; Suizheng Qiu; Changliang Liu; Su Guanghui

The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.


Journal of Nuclear Science and Technology | 2003

A Theoretical Model of Annular Upward Flow in a Vertical Annulus Gap

Su Guanghui; Junli Gou; Kenji Fukuda; Dounan Jia

This paper presents a separated flow model of annular upward flow in a vertical narrow annular channel with bilateral heating. The theoretical model is based on fundamental conservation principles: the mass, momentum and energy conservation equations of liquid films and the momentum conservation equation of vapor core. Through numerically solving the equations, liquid film thickness, radial velocity and temperature distribution in liquid films, heat transfer coefficient of inner and outer tubes and axial pressure gradient are obtained. The predicted results of heat transfer coefficients and axial pressure gradient are compared with the experimental data and good agreements between them are found.


Journal of Nuclear Science and Technology | 2005

Experimental Study on Coolability of Particulate Core-metal Debris Bed with Oxidization, (II) Fragmentation and Enhanced Heat Transfer in Zircaloy Debris Bed

Ken-ichiro Sugiyama; Hiroomi Aoki; Su Guanghui; Yoshihiro Kojima

The oxidization and coolability characteristics of the particulate Zircaloy debris bed, which is deposited under the hard debris and through which first vapor penetrates and then water penetrates, are studied in the present paper. In the vapor penetration experiments, it is found that Zircaloy debris particles are effectively broken into small pieces after making thick oxidized layer with deep clacks by rapid oxidization under the condition that vapor with 20 cm/s penetrates for 30 to 70 min at an initial debris bed temperature of 1,030°C. It is also confirmed in the water penetration experiments that the oxidized particle debris bed has potentiality of high coolability when water penetrates through the fully oxidized particle bed because of a high capillary force originating from those particles with deep cracks on their surfaces. Based on the present study, a new scenario for the appearance and disappearance of the hot spot in the TMI-2 accident is posssible. The particulate core-metal debris bed is fir...


Nuclear Science and Techniques | 2007

Experimental research on flow instability in vertical narrow annuli

Geping Wu; Suizheng Qiu; Su Guanghui; Dounan Jia

Abstract A narrow annular test section of 1.5mm gap and 1800mm length was designed and manufactured, with good tightness and insulation. Experiments were carried out to investigate characteristics of flow instability of forced-convection in vertical narrow annuli. Using distilled water as work fluid, the experiments were conducted at pressures of 1.0 ∼ 3.0MPa, mass flow rates of 3.0 ∼ 25kg/h, heating power of 3.0 ∼ 6.5kW and inlet fluid temperature of 20°C, 40°C or 60°C. It was found that flow instability occured with fixed inlet condition and heating power when mass flow rate was below a special value. Effects of inlet subcooling, system pressure and mass flow rate on the system behavior were studied and the instability region was given.


Nuclear Science and Techniques | 2006

Experimental research on dryout point of flow boiling in narrow annular channels

Ge-Ping Wu; Suizheng Qiu; Su Guanghui; D.N. Jia; Dong-Hua Lu

Abstract An experimental research on the dryout point of flow boiling in narrow annular channels under low mass flux with 1.55 mm and 1.05 mm annular gap, respectively, is conducted. Distilled water is used as working fluid and the range of pressure is limited within 2.0∼4.0 MPa and that of mass flux is 26.0∼69.0 kg·m−2·s−1. The relation of critical heat flux (CHF) and critical qualities with mass flux and pressure are revealed. It is found that the critical qualities decrease with the increasing mass flux and increase with the increasing inlet qualities in externally heated annuli. Under the same conditions, critical qualities in the outer tube are always larger than those in the inner tube. The appearance of dryout point in bilaterally heated narrow annuli can be judged according to the ratio of qo/qi.


Nuclear Science and Techniques | 2006

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

Junli Gou; Suizheng Qiu; Su Guanghui; Dounan Jia

Abstract This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

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Qiu Suizheng

Xi'an Jiaotong University

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Tian Wenxi

Xi'an Jiaotong University

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Zhang Dalin

Xi'an Jiaotong University

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Suizheng Qiu

Xi'an Jiaotong University

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Dounan Jia

Xi'an Jiaotong University

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Jia Dou-nan

Xi'an Jiaotong University

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D.N. Jia

Xi'an Jiaotong University

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Dalin Zhang

Xi'an Jiaotong University

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Song Jian

Xi'an Jiaotong University

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