核动力工程

Numerical Analysis of Stress Intensity Factor of Surface Inclined Cracks at Intersection of Cylindrical Vessels and Nozzles

HE Jia-sheng, ZHU Wei-wei, LI Shu-rong, WU Yuan-xin, XUAN Ai-guo, ZHU Xiao-ming, LU Yuan-ming, YANG Feng

2010, 31(5): 1-4,8.

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A fracture mechanics finite element analysis model of cylindrical pressure vessel with an opening nozzle is established through setting singular elements in the forefront of cracks.A stress intensity factor（SIF） automatically-calculated software of surface inclined cracks at the intersection of cylindrical vessels and nozzles which can be communicated with ANSYS program is programmed with Visual Basic.Based on the calculated SIF values of the cracks of different geometrics subjected to internal pressure,the SIF curves varying with relative depths of cracks,crack relative radians,diameter ratio of vessels and nozzles and crack inclined angle have been obtained.

Test Research on Spectral Characteristics of Acoustic Emission from Pressure Pipe Defect

LIU Zhi-zhong, WANG Yao, AI Qiong, HE Pan, SONG Jian

2010, 31(5): 5-8.

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A coustic emission signals were monitored incessantly in the hydraulic pressure fatigue test on pressure piping with a prefabricated crack,and this paper focused on the characteristics analysis of acoustic emission signals in the time domain and its corresponding frequency spectrum.It is found that the acoustic emission signals caused by crack extending are mostly in the low frequency,and comparatively other AE signals caused by mechanical noise and friction are in the high frequency.The signals can be distinguished by frequency.

Site Adaptability Study on Nuclear Island Buildings Design for 1000 MW Pressurized Water Reactor Plant

LI Zhong-cheng, DONG Zhan-fa, LI Xiang-ye, ZHANG Tao, XU Hai-tao, DAI Wei-ying, CHEN Jin-feng

2010, 31(5): 9-13.

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The site adaptability of one kind of standard design of the 1000MW pressurized water reactor plant is investigated in this paper,and the enveloped parameters scope are acquired as an important results employing the universal finite element software ANSYS.These parameters package,which show the site adaptability boundary conditions for the current design,mainly concern about the structural design and floor response spectra of the buildings in nuclear island.The analysis results demonstrate that the site adaptation range for the current structural design is wide,but that for the seismic design is narrow.

Virtual Age Model for Equipment Aging Plant Based on Operation Environment and Service State

ZHANG Li-ming, CAI Qi, ZHAO Xin-wen, CHEN Ling

2010, 31(5): 14-17,23.

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The accelerated life model based on the operation environment and service state was established by taking the virtual age as the equipment aging indices.The effect of different operation environments and service states on the reliability and virtual age under the continuum operation conditions and cycle operation conditions were analyzed,and the sensitivities of virtual age on operational environments and service states were studied.The results of the example application show that the effect of NPP equipment lifetime and the key parameters related to the reliability can be quantified by this model,and the result is in accordance with the reality.

Research on Professional Adaptability Psychological Selection Indices of Nuclear Power Plant Operators

LIU Jing-quan, LI Zhe, LI Mao-you

2010, 31(5): 18-23.

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Based on the analysis of the work characteristics of nuclear power plant operators and the comparison of professional psychological selection indices for different occupations,the indices of psychological selection system which is applicable to nuclear power plant operators are proposed in this paper,using the method named ‘taking classes,cross-comparison’.The index results of the suggested psychological selection system reflects all the professional requirements on the nuclear power plant operators,which can also be used for the recruitment,training and the retraining programs for operators.

Research on Faults Data Processing and Attributes Reduction Arithmetic Based on Data Mining for Nuclear Power Plants

LIU Yong-kuo, XIE Chun-li, XIA Hong

2010, 31(5): 24-27,38.

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A distance standardized method is proposed considering the features of fault data of nuclear power plants and based on the knowledge discovery function of the data mining method,to standardize the data under different condition and of different order magnitudes.According to the characteristics of parameters,the parameters are dispersed using the alarm values,as a reference for the choice of the break points of the discrete data.The data are reduced using the concept lattice attribute reduction method,and thus the core attributes,relative necessary attributes and unnecessary attributes for fault diagnosis are obtained.The data in literature are calculated and the attributes are classified accurately.When the formal context is affirmatory,the fault can be diagnosed exactly using the core attributes.

Study on Corrosion of Alloy 800 in PWR Primary Water

QIAO Pei-peng, ZHANG Le-fu, LIU Rui-qin, ZHU Fa-wen

2010, 31(5): 28-31.

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The corrosion behavior of Alloy 800 was investigated in the high-oxygen solution containing 600 mg/kg boron and 2 mg/kg lithium for 1500 h at 320℃.The results showed that the general corrosion rate of Alloy 800 after 1500 h was 4.03×10^-4 mm/a,and the oxide film layer was very thin.The TiN precipitates in the matrix could result in pitting.Intergranular attack could be found in the inner side of the tubular specimens after exposure for 1500 h.

Theoretical Modeling and Experimental Study on Fatigue Initiation Life of 16MnR Notched Components

WANG Xiao-gui, GAO Zeng-liang, QIU Bao-xiang, JIANG Yan-yao

2010, 31(5): 32-38.

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In order to investigate the effects of notch geometry and loading conditions on the fatigue initiation life and fatigue fracture life of 16MnR material,fatigue experiments were conducted for both smooth rod specimens and notched rod specimens.The detailed elastic-plastic stress and strain responses were computed by the finite element software（ABAQUS） incorporating a robust cyclic plasticity model via a user subroutine UMAT.The obtained stresses and strains were applied to the multiaxial fatigue damage criterion to compute the fatigue damage induced by a loading cycle on the critical material plane.The fatigue initiation life was then obtained by the proposed theoretical model.The well agreement between the predicted results and the experiment data indicated that the fatigue initiation of notched components in the multiaxial stress state related to all the nonzero stress and strain quantities.

Perpendicularity Automatic Detection System for Nuclear Fuel Pellets

GUO Yong-cai, LI Ming-xing, GAO Chao, LIN Xiao-gang, ZHU Hong-jun

2010, 31(5): 39-41,47.

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In view of the characteristics of pellets,the relation between perpendicularity and end circular run-out error was analyzed,and the perpendicularity detection system for nuclear fuel pellets was designed.Using the grating displacement transducer for measuring end circular run-out error during the pellet rotation course droved by a rotation stage,the system realized the pellet automatic detection.It obtained the resolution over 0.5μm,the accuracy over 3.0μm and the speed 3 pellets/min,and the functions such as data displaying in real time,rapid storage,alarm,and automatic report generation were endowed.The actual operation results indicated that the system was secure,stable and reliable.

Study on High Temperature Fretting Wear of Ti-Al-Zr Alloy

XU Xiao-jun, LIU Han-wei, ZHU Min-hao, QIU Shao-yu, ZHOU Zhong-rong

2010, 31(5): 42-47.

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This paper focuses on the fretting wear behavior of Ti-Al-Zr alloy carried out on a high precision hydraulic fretting wear test rig respectively at ambient and high temperature（400℃）.An experimental layout was designed to perform analysis of the friction coefficient,wear volume and morphology of the worn surface and cross-sections,and investigate the mechanism of fretting wear and wear debris.Experimental testing shows that friction coefficient at 400℃,compared with the ambient,was higher in initial stage,but both were little difference observed in steady stage.The wear volume at 400℃,under the same normal loads and displacement amplitudes,was larger than that at room temperature,and the wear increased with the increase of normal loads at the same temperature.At ambient,the damage mechanism of Ti-Al-Zr alloy was mainly delamination and abrasive wear,while the damage mechanism at 400℃ was primarily the combination of delamination,oxidative and adhesive wear.Based on the results,tribologically transformed structure played an important role in the fretting wear models.

Sensitivity Analysis of Thermal Correlations on Coupling Calculation of Reactor Physics and Thermal-Hydraulics

MA Ting-wei, CAO Xin-rong

2010, 31(5): 48-52,56.

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For analyzing the sensitivity on reactivity feedbacks of thermal–hydraulics,Qinshan-I reactor coolant system is simulated by the coupling codes of nuclear reactor core physics simulation model REMARK and thermal–hydraulic model THEATRe on SIMEXEC simulation platform.Different correlations are applied in the coulping codes corresponding to the suitable thermal–hydraulic conditions.The sensitivity on reactivity feedbacks of the different correlations are analyzed with the simulation results.The result shows that MИxeeB correlation and Shah correlation are most sensitive in the heat transfer correlations,and the correaltions of computing flow friction are insensitive on reactive feedback.

Numerical Simulation on Flow Field of Nuclear Safety Grade 2 Single-Seat Pneumatic Diaphragm Control Valve

ZHONG Yun, ZHANG Ji-ge, WANG De-zhong, SHI Jian-zhong

2010, 31(5): 53-56.

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The Computational Fluid Dynamics（CFD） method is employed to simulate numerically the steady flow and transient flow under variable openings of the nuclear safety grade 2 single-seat pneumatic diaphragm control valve,which is a sleeve valve.The steady simulations under rated condition tells that there is a large amount of vortex in the valve seat necking and around the valve cone,which leads to a much greater flow impact on the head of the valve cone and uneven pressure distribution on spool face.More consideration should be taken on the characteristics of the valve cone accordingly,when designing a valve of this kind.Then the transient flow under 100% and 40% openings is simulated numerically on the basis of steady simulations.The pulsation of the pressure magnitude at the points with large vorticity,in the valve seat necking and around the valve cone,is monitored.The main pulsation frequencies differ from the low natural frequencies of the model,which means that it is safe from leading to structural resonance.

Critical Flow Velocity and Propagation Characteristics of Pressure Disturbance in Gas-Liquid Two-Phase Flow

CHEN Er-feng, LI Yan-zhong, YING Yuan-yuan

2010, 31(5): 57-62.

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According to the small perturbation theory and solvable conditions of one-order linear equations,the dispersion equation of two-phase pressure wave propagation with phase change was derived via the ensemble average two-fluid model.The relationship between critical flow velocity and propagation speed of pressure wave and pressure pulse was analyzed,and the propagation characteristics of pressure wave in the bubbly and slug flow were studied.The results indicate that the pressure wave in gas-liquid two-phase flow has the dispersion property,and the propagation speed of pressure wave with indefinite angular frequency is the propagation speed of the pressure pulse.Moreover,the two-phase critical flow velocity is equal to the propagation speed of the pressure pulse.

Prediction Model for Pressure Drop of Single-Phase Flows in Porous Media Channel

YU Li-zhang, SUN Li-cheng, SUN Zhong-ning

2010, 31(5): 63-66,88.

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Compared with that in the conventional pipes,the fluid flow in porous media channels is more complicated and the flow resistance increases dramatically,which makes it difficult to predict the pressure drop exactly.The pressure drop can be predicted by developing the porous media channel model and solving N-S equation.But it can not be used widely,because of a great quantity meshes.In this paper,based on the similarity theory,a prediction model for pressure drop in porous media channels was developed using Flu-ent6.3.The adiabatic flow was numerically simulated by solving the three dimensional N-S equation.The comparison with the experiment results showed that the predictions agreed well with the measurement of various working conditions with an error less than 5 %.

Laminar Flow Condensation Heat Transfer Characteristics in Vertical Tube under Natural Circulation Condition

LI Yong, YAN Chang-qi, SUN Li-cheng, LIU Jia

2010, 31(5): 67-71.

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Considering the effect of vapor velocity,the Nusselt theoretical formula for laminar film condensation is modified.By using the Nusselt formula and the modified model,the laminar flow condensation heat transfer characteristics inside a vertical tube are studied in a natural circulation system.The result shows that the effect of vapor velocity can not be neglected.The vapor shear decreases the liquid-film thickness and results in a increasing of condensation heat transfer compared to that calculated from Nusselt formula.The modified model considers the interface shear stress appropriately,so the calculated values agree well with the experiment results.

Correlations of CO₂ at Supercritical Pressures in a Vertical Circular Tube

LI Zhi-hui, JIANG Pei-xue

2010, 31(5): 72-75.

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The experiment results of convection heat transfer of CO₂ at supercritical pressures in a 2 mm diameter vertical circular tube for upward flow and downward flow were analyzed for pressures ranging from 78 to 95 bar,inlet temperatures from to 25 to 40℃,and inlet Re numbers from 3000 to 20000.The results were compared with some well known empirical correlations for the heat transfer without buoyancy effects and the heat transfer with strong buoyancy effects.It is found that there is a big deviation between the ex-periment results and empirical correlations.Based on the experiment data,correlations are developed for the local Nusselt correlations of CO₂ at supercritical pressures in vertical circular tubes.

Study on Dry-out CHF Characteristics of R134a Flow Boiling in Horizontal Helically-coiled Tubes

CHEN Chang-nian, HAN Ji-tian, SHAO Li, CHEN Wen-wen, CHEN Bin

2010, 31(5): 76-80.

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An experimental study of critical heat flux（CHF） using R134a as work fluid in horizontal helically-coiled tubes was performed to investigate the CHF characteristics.The stainless steel test section was directly heated uniformly by high-power source.The experiment parameters were: a.outlet pressures of 0.40～1.05 MPa,b.mass fluxes of 51～257 kg/m²s,and c.inlet qualities of-0.18～0.43.And the structure parameters of the test section were: inner diameter of 7.6mm,helix diameter of 300 mm,helix pitch of 40 mm and heated length of 7.07 m.The characteristics of wall temperature and the effects of pressure,mass flux and vapor qualities on CHF were investigated and discussed in this paper.Experiment data were compared with the calculated results by Bowring correlation and Shah correlation,respectively.It was found that the two correlations were not suitable for predicting CHF under the present experiment conditions.

Analysis of Function of Stratified Interface in Density Lock

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

2010, 31(5): 81-83.

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The anti-disturbance mechanism of the density lock is analyzed by cold and hot state experi-ments with theoretical analysis.The result shows that in cold state,the pressure difference which caused by the fluid flow will produce the large circulation of fluid movement in channels of density lock.And in hot state,the formation of fluid stratified interface in the density lock will stop the large circulation of fluid movement.The further study shows that as a result of pressure difference,the geopotential difference is pro-duced by the slope of fluid stratified interface,and when the pressure difference is equal to the geopotential difference,the large circulation of fluid movement is stopped.

Characteristics Analysis for Passive Residual Heat Removal System Based on Density Lock

CHEN Wei, YAN Chang-qi, GU Hai-feng, ZHANG Nan

2010, 31(5): 84-88.

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Based on the passive residual heat removal system（PRHRS）,the closure of the density lock on normal condition and the startup of the density lock on accidents are analyzed by experiments,and the iN-Stantaneous characteristics of PRHRS after accident are simulated by RELAP/MOD3.2 code.The calculated results are compared with the experimental data,and they agree well.It is shown that on normal condition the density lock can separate the heat removal loop from the main loop,and the PRHRS does not work during this condition,while when the accident happens the PRHRS can operate instantly by different gravities in two loops and it will set up the steady natural circulation between the heat removal loop and the main loop to re-move the residual heat gradually.

Theoretical Research on Laminar Friction Resistance in Tubes in Rolling Motion

YAN Bing-huo, YU Lei, YANG Yan-hua

2010, 31(5): 89-92.

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The model of laminar flow in tubes in rolling motion is established.The dimensionless correlation of velocity is derived,and the correlation of frictional resistance coefficient is also obtained.Of all the additional forces,only the tangential force effects on the flow.The effect of centrifugal and Coriolis forces on the flow is counteracted.The correlation of average frictional resistance coefficient is the same with that of no rolling motion.The effect of rolling motion on frictional resistance coefficient of laminar flow diminishes with the increase of Reynolds number.

Theoretical Research on Flowing Characteristics in Rolling Motion with Fractional Maxwell Model

YAN Bing-huo, YU Lei, YANG Yan-hua

2010, 31(5): 93-96,107.

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The flowing characteristics of the flow with no initial velocity in a tube in rolling motion are investigated theoretically,with the fractional Maxwell method.The velocity in rolling motion is derived.The effect of rolling motion on the flowing velocity of Newtonian fluids and Non-Newtonian fluids is analyzed.The variation of peak velocity and velocity gradient in the wall is also investigated.Compared with the non-Newtonian fluid,the effect of rolling motion on Newtonian fluid is more averaged.The oscillation of velocity gradient at the wall is similar to that of velocity.And their oscillation period is half of the rolling period.

Research on Transient Flux Distribution in Parallel Channels

LI Lei, ZHANG Zhi-jian

2010, 31(5): 97-101.

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When multi-channels model is used for thermal-hydraulic analysis in reactors with closed fuel lattice,the mass flux distribution must be solved at first.Three methods for mass flux distribution are proposed.Transient flux distribution program is developed to calculate the mass flux distribution in parallel channels and the results are compared.The results show that the first method is only proper for slow transients but the second and the third methods are also proper for fast transients.However,the stability of the third method is higher than the second method.So the third method can be used for mass flux distribution calculation in plate type fuel parallel channels.

Analysis of Process for Core Melt and Debris Pool Formation in IVR Evaluation for CPR1000

FU Xiao-liang, YANG Yan-hua, ZHOU Wei-hua, YANG Xiao

2010, 31(5): 102-107.

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In-Vessel Retention（IVR） is an important and crucial method for preventing and mitigating the consequences of severe accidents.In this paper,a lot of severe accident sequences are selected under investigation according to the related laws and regulations,and then calculated by an integral severe accident analysis computer code to get the details of debris pool in lower head and their effects on the heat focusing effect.The results show that: there is a marked heat focusing effect on the height of metal layer in debris pool and the vessel can get its integrality under external reactor vessel cooling.

Study on Radiological Consequence Induced by SGTR Accident with Alternative Source Term

ZHENG Xiao-yu, HUANG Gao-feng, CAO Xue-wu

2010, 31(5): 108-112.

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By introducing and implementing the alternative source term and accident methodology defined in RG 1.183 and using integral safety analysis code,this paper evaluates the radiological consequence induced by（SGTR） accident for the domestic nuclear power plant of 90 MW.Compared with the dose criteria of RG 1.183,the dose values of the control room,exclusion area boundary and outer boundary of low population zone are acceptable.

Development of Simulation and Verification Platform for Nuclear Plant Digital I＆C System

ZHU Li-zhi, LIN Meng, YANG Zong-wei, LIU Peng-fei, YANG Yan-hua

2010, 31(5): 113-117.

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The emulation functional test platform of the digital instrument and control system（DCS） in Ling’ao Nuclear Power Station Phase II was developed.Upon the verified simulation system（including the thermal object and simulation control logic）,based on the virtual instruments LabVIEW software and data acquisition card,the emulation functional test platform was established for the data communication between the simulation system and DCS system,and the emulation test can be conducted on the automatic DCS system before the commissioning.

Uncertainty Analysis of Failure Probability for Nuclear Reactor Coolant Pumps Group System Based on GO-FLOW Methodology

SHANG Yan-long, CAI Qi, CHEN Li-sheng, ZHAO Xin-wen, YAN Can-bin

2010, 31(5): 118-123.

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Nuclear Reactor Coolant Pumps Group（NRCPG） is a complex redundancy system which contains more than one common causes and multiple common cause component groups（CCCG）,and it is complex and difficult to conduct the uncertainty analysis for the system failure probability.In this paper,GO-FLOW combined with Monte-Carlo is applied to solve this problem,and the uncertainty of the system failure probability considering the common cause failure（CCF） is quantitatively calculated.The results show that the mean value of system failure probability is obviously increased by CCF,and the occurrence of the multiple common cause failure makes the standard deviation,x₀₅ and x₉₅ which contain 90% range of uncertainties larger,but makes the error factor E_F smaller;The probability density function（PDF） curve of the system failure probability describes the integrated distribution,so it can make up the imperfection caused by the point estimate value used to evaluate the system reliability.

Comparison of Performances of Full-Speed Turbine and Half-Speed Turbine for Nuclear Power Plants

WANG Hu, ZHANG Wei-hong, ZHANG Qiang, LI Shao-hua

2010, 31(5): 124-126,130.

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The steam turbines of nuclear power plants can be divided into the full-speed turbine and half-speed turbine.Different speed leads to differences in many aspects.Therefore,the rational speed is the key point in the selection of steam turbines.This paper contrasts the economy between the half-speed turbine and full-speed turbine,by calculateing the relative internal efficiency of half-speed and full-speed steam turbines with the typical level of 1000 megawatt.At the same time,this paper also calculate the relative speed of high speed water drops in the last stage blade of half-speed turbine and full-speed turbine,to contrast the water erosion between the half-speed turbine and full-speed turbine.The results show that the relative internal efficiency of half-speed turbine is higher than that of the full-speed turbine,and that the security especially the ability of preventing water erosion of half-speed turbine is better than that of the full-speed turbine.

Modular Design on Qualification Test System of Pumps of Nuclear Safety Grade

YAN Jian-hua, OU Ming-xiong, TENG Guo-rong, GENG Wei-hao, SUN Xiao-ming

2010, 31(5): 127-130.

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As a dynamic equipment in the PWR units,the reliability of the pumps is key to the security of whole nuclear power station.In the development of the pumps of nuclear grade 1 and 2,various strict qualification tests shall be carried out,to make sure that the design satisfies the system requirement.The research on the tests of nuclear grade pumps is not much in China.Based on the requirements on the qualification tests for pumps of nuclear grade 2 used in current nuclear power stations,this paper studies the test system design with modular design theory,and discusses the design idea and problem,and then proposes a general design scheme for the test system.

Calculation of Effective Coefficient of Spontaneity Fission Neutron Source in CFBR-Ⅱ Reactor

DU Jin-feng, FAN Xiao-qiang

2010, 31(5): 131-133.

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In order to calculate the effective intensity of spontaneity fission neutron source in CFBR-II reactor,Monte Carlo calculation method of effective coefficient is built.The naissance and transport process of spontaneity fission neutron and eigen distribution neutron are simulated.The effective coefficient of spontaneity fission neutron source equals to the ratio of leakage neutron number which come from two kinds of neutron sources.Considering the special structure of CFBR-II reactor,for the spontaneity fission neutron sources,the effective coefficient of different positions in upper hemisphere and lower hemisphere are calculated separately using cell-rejection technique.These results can be used as the basis to calculate the total effective intensity of spontaneity fission neutron source.

Core Physics Calculation and Analysis for SNRE

JIE Jia-chun, ZHAO Shou-zhi, JIA Bao-shan

2010, 31(5): 134-138.

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Five different precise calculation models have been set up for Small Nuclear Rocket Engine（SNRE） core based on MCNP code,and then the effective multiplication constant,drum control worth and power distribution were calculated.The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation,but a detailed description of elements have to be used in the element internal power distribution calculation.The results of physics parameters show that the basic characteristics of SNRE are reasonable.The drum control worth is sufficient.The power distribution is symmetrical and reasonable.All of the parameters can satisfy the design requirement.

Neutron Nondestructive Essay of the Plutonium Metal Parts

LIU Xiao-bo, XIAO Jian-guo, CHEN Xuan-yong

2010, 31(5): 139-142.

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Based on the principle of neutron multiplicity and data unfolding mathematics,this paper developed the software which can execute the neutron multiplicity analysis,neutron attenuation analysis,parameter calibration,Pu mass solution with the neutron pulse sequence acquisition method.The measurement system consisted of detector,nuclear electronic apparatus,pulsed sequence acquisition and analysis software was tested and calibrated by californium source.Three mental plutonium components with different mass were used for experimental assay and validation,which showed that the assay bias was within 15% against the nominal value of the samples.

2010 Vol. 31, No. 5