核动力工程

Severe Accident Mitigation Measure and Severe Accident Sequence of CPR1000 Nuclear Power Plant

LUO Bang-qi, LIN Ji-ming

2010, 31(S1): 1-3,7.

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The passive autocatalytic recombiners,pressurizer depressurization function extension （prevention high pressure melt core） and containment pressure relief and filter releasing system are used as the prevention and mitigation measures of severe accidents to prevent and mitigate the consequences of severe accidents in CPR100 nuclear power plant,to ensure the integrity of the containment of CPR1000 nuclear power plant under severe accidents and thus to protect the surrounding residents from the dangers of nuclear radiation. Based on the spectrum analysis of severe incidents,the LBLOCA + loss of safety injection pumps,station blackout,loss of all feed water +anticipated transient without scram and MSLB + loss of containment spray are selected as the important severe accident sequence.

Model for Analysis of Containment Failure by Melt-through during Severe Accident of Reactor

CHEN Zhi-yun, XU Shao-hua, CHEN Wen-zhen

2010, 31(S1): 4-7.

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The containment failure by melt-through during severe accident is analyzed and calculated based on the contact melting around the heat source. A new and more rational analysis model of contact melting is established under certain hypothesis,and the melting velocity and the time needed for melt-through are calculated. The results are compared with some related literatures. The mass of heat source and the heat transfer ratio k as the main factor on melt-through are analyzed,and the conclusions are drawn that the melting velocity is in direct ratio with the mass of heat source and approximately in direct ratio with the heat transfer ratio,and however the relation of the time needed for melt-through with the mass of heat source and k is in the opposite way. It is indicated that in order to delay the melt-through additional engineering device to reduce the mass of single heat source and enhance the upside heat transfer is of great importance.

Study on Total Loss of Main Startup and Auxiliary Feedwater Under SOP

RAN Xu, FANG Hong-yu

2010, 31(S1): 8-10.

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In this paper,the accident of total loss of main,startup and auxiliary feedwater is studied by CATHARE simulation. The accident analysis and the comparison between SOP and EOP indicate that the core is not uncovered and is safe in the accident under SOP. The feature and application of SOP in accident is also demonstrated.

Preliminary Research on Nuclear Reactor Thermal-Hydraulic Multi-scale Coupled Simulation

LIU Yu, ZHANG Hong, JIA Bao-shan

2010, 31(S1): 11-15.

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Recently,the nuclear safety analysis more and more emphasize on the fine and realistic analy-sis. Thus thermal-hydraulic multi-scale coupled simulation approach,including the system,component and local scales,has been proposed in some countries. And the three scales can be coupled together effectively through one coupling method. Based on the similar idea,this paper makes a multi-scale coupling framework of RELAP5,COBRA4 and CFX. Verification of current coupled code has been carried out by testing calcula-tion of two simplified problems.

Design Modification and Analysis of Helium Purification System for Gas-Cooled Reactors

CAO Jian-hua, WANG Jie, FU Xian-gang, Jean-Charles Robin

2010, 31(S1): 16-20.

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To avoid the corrosion of high temperature materials,the contaminants inside the helium loops of gas-cooled reactors need to be controlled by helium purification systems. Based on the mature technologies of oxidation and adsorption with CuO bed,Molecular Sieve bed and Active Charcoal bed,a modified design was given through the analysis and comparison of other typical helium purification systems. Results show that,the modified design could significantly reduce the system energy consumption. The effect of the efficiency of recuperators on system energy consumptions was also analyzed,and the lowest system energy consumption does not always occur at the highest efficiency of the No.2 recuperator in the modified design.

Thermal-Hydraulic Analysis of Sub-channel on ADS Principle Verification Device

LIU Zhan, YANG Yan-hua

2010, 31(S1): 21-23,28.

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The present paper performs the steady-state thermal-hydraulic analysis of sub-channel on ADS principle verification device using the COBRA-TF code. The calculation results from COBRA-TF code are proved fully consistently by RELAP5,which shows that the results about the steady-state thermal-hydraulic analysis are correct,and it provides the basic research on the design of ADS principle verification device and safety.

Applicability of "Chen" Type Methods for Predicting Boiling Heat Transfer in Mini-Channels

YU Kai-qiu, SUN Li-cheng, YAN Chang-qi

2010, 31(S1): 24-28.

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Chen correlation and other correlations based on Chen correlation are the most commonly used methods for predicting the flow boiling heat transfer. However,some recent researches show that great errors are caused when these correlations are used for mini-channels. Based the experimental results from Yan and Lin et al.,a new database including 2505 data for 11 liquids were built,covering diameter from 0.21 to 6.05mm. The applicability of these methods for prediction of the flow boiling heat transfer in mini-channels was evaluated. The results show that the Chen method and the Chen-type correlations are not quite suitable for mini-channels; the correlation for calculating the nucleate boiling heat transfer coefficient in Chen correla-tion is no longer applicable when the vapor quality is high.

Effect Study of Flow Boundary Layer Characteristics in Narrow Rectangular Channel by Longitudinal Vortex

HUANG Jun, HUANG Yan-ping, WANG Qiu-wang, MA Jian

2010, 31(S1): 29-33.

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Longitudinal Vortex （LV） is produced by Longitudinal Vortex Generators （LVG）. The heat transfer enhancement mechanism of LV is based on the reducing and damaging of flow boundary layer. With relative long influence distance and simple structure,the LVGs can be used in narrow channels with flat surface. In this paper,Phase Doppler Particle Analyzer has be used to measure the flow field in narrow rectangular channel with LV. The experimental data of smooth and LVGs channels showed that the boundary layer is damaged by LV in single-phase water flow,and the calculation results of CFD showed that the velocity field with LV can be simulated by SST model in narrow rectangular channel.

Flow Field Characteristics of Bubble Growth of Subcooled Flow Boiling in Vertical Narrow Rectangular Channel

YUAN De-wen, PAN Liang-ming, CHEN De-qi, LI Long-jian

2010, 31(S1): 34-38,48.

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A growth model of bubbles in sub-cooling flow boiling was established by VOF two-phase flow model,and the bubble growth process was numerically simulated by UDF interface procedures. The factors such as the velocity of mainstream speed,sub-cooling degree of the mainstream,the degree of the heating wall and vapor-liquid contact angle were considered in the simulation. The growth curve of bubbles in subcooling flow boiling was also obtained. The velocity field of bubble growth process was investigated atdifferent working conditions. The relationship between the bubble inner velocity field and the velocity speed of the interface was also studied.

Research on Movement of Fluid Stratification Interface in Density Lock

YU Pei, YAN Chang-qi, GU Hai-feng

2010, 31(S1): 39-43.

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The cause for the formation of fluid stratification interface was analyzed,and a simple mechanical model was founded to interpret the principle of interface movement. The effects of the grid shape and size,the disturbance magnitude and temperature difference on the interface removal speed was validated experimentally. The results indicated that only the grid size and temperature difference had effect on the interface movement speed. The smaller the channel sectional area was and the larger temperature difference was,the more slowly the interface moved.

Effect of Fluctuation Caused by Rolling Motion on Single-Phase Natural Circulation Heat Transfer

JIA Hui, TAN Si-chao, GAO Pu-zhen, YAN Chang-qi

2010, 31(S1): 44-48.

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Experimental studies of single-phase natural circulation flow and heat transfer under a rolling motion condition are performed. Experiments with and without rolling motions are conducted,and the effects of rolling motion on natural circulation heat transfer are obtained. The experimental results show that the flow fluctuation caused by rolling motion enhances the heat transfer,and the heat transfer coefficient of natural circulation flow increases with the rolling amplitude and frequency. Analysis of data shows that the effect of heat capacity of heated tube can not be neglected and Reynolds number of acceleration is introduced to employ the effect of heat capacity. Based on the experimental data,an empirical equation for the heat transfer coefficient under rolling motion is achieved. The calculated results agree well with experimental data.

Study on Pool Boiling and Flow Boiling with Artificial Neural Networks

CHEN Rong-hua, SU Guang-hui, QIU Sui-zheng

2010, 31(S1): 49-52.

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In this paper,two artificial neural networks （ANNs） are trained successfully to predict the CHF of thermosyphon and heat transfer coefficient of pool nucleate boiling respectively. The root mean square of predicated value are 16.43% and 19.57%,respectively. The analysis results indicate that CHF would be improved by inserting an inner tube in the thermosyphon. CHF increases initially as inner tube diameter increases and then decreases with the further increase of inner tube diameter. The heat transfer coefficient of pool nucleate boiling increases linearly as pressure increases,and when the pressure is close to the critical pressure,the increasing rate increases.

Non-Heating Experimental Study on External Reactor Vessel Cooling

LI Yong-chun, YANG Yan-hua, KUANG Bo, CHENG Xu

2010, 31(S1): 53-56.

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As a key severe accident management strategy,External Reactor Vessel Cooling （ERVC） could terminate the physical process in the reactor pressure vessel and achieve the in-vessel molten debris retention. However,the strategy still needs to be investigated before its application in nuclear power plants. A slice geometry,full scale simulation experiment facility is designed to observe and investigate the two phase natural circulation of ERVC in CPR1000. The study is focused on the effects of the flow path geometry and size on the natural circulation. The experiment results reveal that the inlet/outlet area,penetration tubes and insulation influence the flow process at different levels and the inlet/outlet areas shows a dominating effect on the mass flow rate. The effect of penetration tubes on the boiling heat transfer need to be further evaluated. The maximum measured errors of air injection flowrate and circulation flowrate are 12.9% and 3.4% respectively.

Experiments of Two-Region Pebble Dynamics of Two-Dimension Core in HTR

ZHANG Xiu-zhi, YANG Xing-tuan, ZHANG You-jie, LIU Zhi-yong, JIANG Sheng-yao

2010, 31(S1): 57-59,64.

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According to the similarity law,the experiment system of pebble dynamics of two-dimension core in High Temperature Gas-cooled Reactor （HTGR） is established to carry out the study on the rules of pebble dynamics. The issues closely related to the reactor core physics and structure design of HTGR is experimentally studied. The experiment results indicate that the pebble dynamics is random and episodic in microcosmic aspect,but is certain and steady in macrocosmic aspect. The stable state is feasible in this experiment condition,the sizes of two regions change regularly with the variation of the proportion of black balls to the white ones,the steady two-region can be obtained in different proportions of black balls to the white ones,and the proportion in the outlet tube is the same as that in the inlet tube.

Experimental Study on Wall Temperature Distribution of Flat Tube

HE Shi-jing, YAN Chang-qi, SUN Zhong-ning, FAN Guang-ming

2010, 31(S1): 60-64.

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Research on the distribution of outside wall temperature,horizontally and vertically laid flattube in the casing pipe,is carried out by experimental method. The present study indicates that the outside wall temperature,which is measured by the thermocouple in middle of straight wall surface of flat tube,is lower than the temperature near the arc part,and regular methods can not be adopted to measure the flat tube outside wall temperature,like the circular tube.

Experimental Research on Effect of Rolling on Natural Circulation Flow and Heat Transfer

LIU Yang, XING Dian-chuan, GAO Pu-zhen

2010, 31(S1): 65-68.

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Single-phase natural circulation flow and heat transfer characteristics under rolling conditions are studied experimentally. The experiment results show that compared with the results under non-rolling condition,the natural circulation flow rate decreases under rolling condition. As the rolling frequency and amplitude increases,the flow rate reduces but the heat exchange increases. The empirical correlations for natural circulation flow rate and heat exchange are obtained by fitting the experimental data under rolling mo-tions.

Experimental Study on Single-Phase Heat Transfer in Circular Tube under Oscillatory Flow Conditions

ZHAO Da-wei, SU Guang-hui, ZHANG You-jia, LIANG Zhen-hui, QIU Sui-zheng

2010, 31(S1): 69-72.

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The experimental study on the heat transfer in circular tube under oscillatory flow conditions was conducted. The sinusoidal fluid oscillations under 0.1～1 Hz were made by the oscillator in the text loop to analyze the effect of flow instability on two-phase flow boiling heat transfer. The traces of inlet oscillatory flow in 0.4 and 1.37 MPa and traces under Frequency Modulation Conditions were analyzed,respectively. The oscillations of Wall Temperature and Outlet Temperature were obtained. And,the oscillation of Outlet Temperature which was lag phase about 1/4 cycle,and oscillations of Wall Temperature were with same frequency.

Numerical Investigation of Maximum Charging Flowrates under Accident Condition in PWR Nuclear Power Plants

WANG Zhi-gang, WANG Xiao-jiang, LI Li-juan, LI Jun

2010, 31(S1): 73-76,82.

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RCV System and RIS System are simulated with CFD software Flowmaster7.5,and the model is validated with design data.The relationship between maximum charging flowrates and Reactor Coolant System pressure is predicted by this model,when the safety injection mode switches to the charging mode after SGTR,and the relationship between the maximum charging flowrates and the corresponding Reactor Coolant Pump Seal Injection and the Pressure of Reactor Coolant System are quantitatively compared,when the charging pump is pressurized by various low voltage safety injection pumps.

Numerical Analysis of Magnetohydrodynamic Flow in Rectangular Duct of Chinese Liquid Metal LiPb Experiment Loop DRAGON-Ⅳ

YANG Zhi-yi, ZHOU Tao, CHEN Hong-li, WU Yi-can

2010, 31(S1): 77-82.

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The Magnetohydrodynamic （MHD） experiment,which will be carried out in Chinese liquid metal LiPb experimental loop DRAGON-Ⅳ,is an important approach to study the flow characteristics under the high magnetic field of the magnetic confinement fusion reactors. An accurate numerical analysis is neces-sary not only for the design and optimization of MHD experiment loop,but also for the measurement of pres-sure and velocity. In this paper,a conservative scheme and consistent projection method were adopted to simulate the MHD flow at high Hartmann numbers. The code developed by the FDS Team was used to ana-lyze the MHD effect of LiPb MHD loop under fringing and uniform magnetic fields. The current density dis-tribution was shown to illustrate the effect of Lorentz force on the velocity and pressure fields.

Numerical Study on Horizontal Natural Circulation in Fuel Transfer Channel

GUO Qiang, LEI Ning-bo, LIU Jian-ping

2010, 31(S1): 83-87.

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The cooling process of the spent fuel assembly stuck in the transfer channel under horizontal natural circulation is studied by FLUENT. The effects of the stuck position of fuel assembly,the length of transfer channel,and the mechanical structure of the fuel transfer system on the cooling effectiveness are numerically studied. The comparison of the simulation results reveals the key factors with effects on the cooling effect of horizontal natural circulation,and provides the necessary basis for the thermal hydraulics analysis for nuclear safety research. Based on the analysis results for the thermal-hydraulics process,an optimized mechanical design for the fuel transfer system is suggested,and the numerical calculation indicates that it can effectively improve the cooling effect,and enhances the reliability and redundancy of the fuel assembly.

Geometry Model and Mesh Division Technology of Fuel Assembly Spacer Grids

CHEN Jie, CHEN Bing-de, ZHANG Hong

2010, 31(S1): 88-92.

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In order to analyze the mixing performance of spacer grids with Computational Fluid Dynamics （CFD） method,the geometry model and mesh division of fuel assembly spacer grids are investigated systemically. Compared with the results of different geometry models,the spacer grids can be simplified as the structures of ribbon and mixing vanes only,without dimple or spring. And the exit length and entrance length are set as 250 mm and 230 mm,respectively. And the fuel assembly with only one pacer grid is to be researched. Through the effect of mesh on the calculation results,the method of mesh division and mesh quantity are confirmed. The entrance part and exit part are for the structured mesh,while the part of space grids is for the unstructured mesh. The total nodes number of whole research object is 1032258,and total cells number is 2601614,of which the spacer grids part is 75.8%. The following research of mixing performance of fuel assembly spacer grids with CFD method,will be performed on the base of geometry model and mesh size,which are confirmed in the research of this paper.

Numerical Analysis of Supercritical Water Vertical Upward Flow Heat Transfer in Tube with Different Obstacles

ZHANG Bo, DAN Jian-qiang

2010, 31(S1): 93-96,108.

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The heat transfer coefficient is relative low when the bulk temperature is above the pseudo-critical point due to the properties of vapor-like fluid. Numerical simulation on the heat transfer phe-nomena downstream of an obstacle in a vertical tube is carried out with commercial CFD code Fluent 6.1 us-ing adaptive grid. The results show that the RNG k-ε model with enhanced wall treatment can obtain reliable results; and the higher of the blockage ratio and the local temperature can achieve the better heat transfer en-hancement. The influence region and decay trend are also given.

Numerical Study and Sensitivity Analysis of PRHR HX Transient Heat Transfer Performance

PAN Xin-xin

2010, 31(S1): 97-102.

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Natural convection in AP600 （American Passive Advanced Pressurized Water Reactor） passive residual heat removal heat exchanger （PRHR HX） is numerically studied with FLUENT. Numerical investigation using standard k-ε model and Boussinesq model is performed. The transient numerical results show that the heat transfer performance of the PRHR HX and IRWST are consistent with AP600 experimental results. The sensitivity analysis has the following conclusions:a. the effect of baffle position and inlet construction on heat transfer performance is very small and b. with the increase of the number of heat exchanger pipes or the initial temperature of water in IRWST,the heat transfer will be enhanced.

Three-Dimensional Numerical Study on Forced Convection Heat Transfer in Structured Packed Porous Media

YANG Jian, CENG Min, YAN Xiao, WANG Qiu-wang

2010, 31(S1): 103-108.

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A three-dimensional forced convection heat transfer inside small pores of structured porous media with spherical particles is numerically studied in this paper. The Navier-Stokes equations and RNG k-ε turbulent model with scalable wall function are adopted for the computations. The effects of Reynolds number and packing form on the performances of forced convection heat transfer in porous media are investigated in detail. The results show that,with the same physical parameters,the pressure drops in porous media with proper packing form can be greatly reduced and the overall heat transfer efficiencies are significantly im-proved. The traditional empirical correlations are unavailable for the structured packed porous media and some modifications are required. Furthermore,it is found that,with different packing forms,the overall heat transfer performance of SC packing is the best,and with the same packing form,the overall heat transfer per-formance of uniform packing is much better than that of the non-uniform packing.

An Analytical Model for Pool Boiling Critical Heat Flux in Inclined Confined Spaces

WEN Qing-long, CHEN Jun, ZHAO Hua

2010, 31(S1): 109-113,118.

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Based on the bubble behaviors observed during experiments of pool boiling critical heat flux in inclined confined spaces,an analytical model of critical heat flux is presented,including four major sub-models:a separated flow model,a stability analysis,a heater surface energy balance and a lift-off criterion. Calculation results of critical heat flux agree well with experiment results of 5mm and 8mm gap sizes both qualitatively and quantitatively. Though calculation results show the varying trend of experiment results of 3 mm gap size qualitatively,to some extent the discrepancies do exist quantitatively. In addition,the calculation results predict well the experimental results of Yang,Kim,Howard and Gribov,but the discrepancies between the calculation results and Guo’s correlation are great.

Numerical Model of Secondary Loop of Marine Nuclear Power Plant Based on RELAP5 Code

WANG Shao-wu, PENG Min-jun, DAI Shou-bao, CHENG Shou-yu, SUN Ying-jie

2010, 31(S1): 114-118.

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This paper presents a new secondary loop of nuclear power plant system,including the main com-ponents of turbine,condenser,feed pump,and heater,based on RELAP5/MOD3.4 code. The research of the single component model adaptive authentication analysis and the local computing ability of the code have been studied. The results show that the steady-state calculation results of the RELAP5/MOD3.4 code are basically consistent with the design value,and the dynamic calculations can also meet the calculation accuracy requirement of the main components in secondary loop.

Numerical Simulation of Flow Pressure Drops for Complex Pipe Segment

QI Zhan-fei, TONG Li-li, CAO Xue-wu

2010, 31(S1): 119-122.

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To study the causes for the local pressure drop,the three-dimension single-phase flow in the complex pipe segment with holes in nuclear power plants was studied with the computational fluid dynamics method under two flow conditions. The hybrid grids technique and standard k-ε turbulence model were applied. The results show that the different forms of pressure drop were caused by different flow directions in the pipe; the local stream line resulting in eddy and pressure changed greatly near the porous liner tube; how and where the cross-sections of flow channel change corresponding to the certain flow direction cause mainly different local pressure drop.

Numerical Simulation of Rayleigh-Taylor Instability with MPS Method

CHENG Hui-fang, DUAN Ri-qiang, JIANG Sheng-yao

2010, 31(S1): 123-126.

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As a meshless method,the Moving Particle Semi-Implicit Method is used to calculate the in-compressible fluid flow. The fluid is represented by a set of moving particles with full Lagrange description,and the differential operators in governing equations are discretized to the interactions of particles. With this method,two-dimensional Rayleigh-Taylor Instability （RTI） is simulated. It is demonstrated that,when the amplitude is much smaller than the wave,the perturbation of interface grows exponentially at early stage of the instability,which is consistent with the linear theory. During non-linear stage,the characteristics of bub-bles and strikes obtained in the calculation agrees with the previous experiment to some extent.

2010 Vol. 31, No. S1