核动力工程

Irregular Region Transformation Method for Hexagonal Assembly Homogenization and Its Numerical Validation

XIA Bangyang, LI Qing, LI Xiang, LI Manchang

2011, 32(5): 1-4.

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In order to simplify the calculation process of hexagonal assembly homogenization,and enhance the adaptability of homogenization method for hexagonal geometry,a simple irregular region transformation method is proposed in this paper,by which these irregular regions such as assembly wrapper and water gap can be transformed into regular hexagon with the constraint condition that material mass remains constantly.The numerical results show that it is a very accurate and efficient method for the prediction of assembly reactivity and normalization power distribution versus burnup.

Monte Carlo Simulations of Reactor α Eigen-Values

LI Zeguang, LI Tianya, WANG Kan, YU Ganglin

2011, 32(5): 5-8,28.

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In Monte Carlo simulation,the α eigen values are usually calculated using the（k,α） iteration method.This method can give good results for supercritical and light subcritical systems,but due to the limitation of this method,it can not give the right results of α eigen values in deep subcritical systems.To calculate the α eigen values in deep subcritical systems using Monte Carlo method,a method named Transient Curve Fitting method which is based on the MC transient code TMCC is put forward.The Transient Curve Fitting method can solve the difficulties of simulating the α eigen functions and calculate the α eigen values effectively.In this paper,the results of this method are compared with MCNP results by calculating super-critical and subcritical problems,and they agree well with each other.And then the results of this method are compared with the experimental results of Subcritical Facility in Tsinghua University.And the results represent that the transient curve fitting method can calculate the α eigen values of reactor systems,and can give accurate results in deep subcritical situations.

Measurement of Dynamic Parameter of Sub-Critical Facility with Rossi-α Method

YANG Xin, YU Ganglin, WANG Kan

2011, 32(5): 9-12.

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This paper presents the Rossi-α method for the micro-noise analysis in the reactor physics ex-periments,which is applied in the study of measurement of the kinetic parameter,α eigen value,of Tsinghua University’s Sub-Critical Facility.The measurement is performed under extremely low flux condition,and α_c,the critical α eigenvalue,is obtained by the way of extrapolation.Meanwhile,the theoretical calculation has been carried out based on MCNP and TMCC codes.Results from measurement and calculation agree well with each other.

Analysis of Capability of Homogeneous Spent Fuel Transmutation Reactor Core to Burn Actinides

WANG Lianjie, WANG Liangzi, YAO Dong, WU Yinghua

2011, 32(5): 13-17.

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Seven models of homogeneous spent fuel transmutation reactor（HSTR） core with different spent fuel composition and core power have been constructed to burn the actinides.The burnup variation of plutonium,neptunium and other long-life actinides is calculated by FMCHR to analyze its capability on burning actinides.The result shows that the HSTR can be effectively used for burning ²³⁹Pu and transmuting ²³⁷NP.For the transmutation of ²⁴¹Am and ²⁴³Am,an enhanced initial content of these two nuclides in the plutonyl nitrate or an adoption of uranyl nitrate are required.

Control Analysis for Failure Transient Accident of Feed-Water Control System in Supercritical Water-Cooling Fast Reactor

WANG Handing, ZHOU Tao, CHEN Juan, LI Jingjing, LIU Mengying

2011, 32(5): 18-22.

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Supercritical cooling fast reactor is the once-through cooling system,and its transient safety characteristics is different from that of the existing LWRs.Under the failure transient accident of the feedwater control system,the reactor pressure,power,coolant temperature,coolant mass flow and the surface clad temperature changing with the time are studied with three different control modes（the control rods,steam turbine valves,and reactor coolant pump）.The results show that with the combined control by the turbine valve and control rod,all reactor parameters are stable and the clad temperature is within the safety range,thus the transient accident of feedwater control failure can be mitigated to ensure the safety of the supercritical fast reactors.

Analysis of Effect of Safety Classification on DCS Design in Nuclear Power Plants

GOU Guokai, LI Guomin, WANG Qunfeng

2011, 32(5): 23-28.

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By analyzing the safety classification for the systems and functions of nuclear power plants based on the general design requirements for nuclear power plants,especially the requirement of availability and reliability of I&C systems,the characteristics of modern DCS technology and I&C products currently applied in nuclear power field are interpreted.According to the requirements on the safety operation of nu-clear power plants and the regulations for safety audit,the effect of different safety classifications on DCS design in nuclear power plants is analyzed,by considering the actual design process of different DCS solutions in the nuclear power plants under construction.

Identification Method for Transfer Function of Steam Generator Level Control System of Ling’ao Phase Ⅱ NPP

CHEN Zhi, ZHANG Ying, WANG Huajin, ZHANG Rui

2011, 32(5): 29-32.

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The identification of the unknown transfer function about controlled object and physical process of the steam generator level control system of Ling’ao Phase II NPP was studied and the identification method was given.The basic data used for the identification were obtained by analyzing the feature curves of the device provided by the supplier and the transient results calculated by transient analyzing program CA-TIA2 and extracting the related data.The unknown transfer functions were finally gained using the parameter estimation models built in MATLAB software after getting the basic data.The identification results meet the requirements on the general design and performance verification of the control system.

Analysis Method for LOFT L2-5 LBLOCA Experiment

ZHANG Zhongwei, LIANG Guoxing, KUANG Bo

2011, 32(5): 33-37,48.

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The methodology for Large Break Loss of Coolant Accident（LBLOCA） licensing analysis with conservative Appendix K models plus statistical plant parameters analysis is presented in this paper.With the Appendix K version of RELAP5,which has been modified in compliance with the requirements of 10CFR50 Appendix K evaluation model,the quantification of the effect of plant status uncertainty on the safety parameter is carried out,based on a non-parametric statistical method.To demonstrate the new LOCA methodology,the LOFT L2-5 double-ended cold leg guillotine large break loss of coolant accident integral test is analyzed.It is revealed that this deterministic-realistic hybrid methodology can provide more PCT margin than the traditional Appendix K LOCA analysis by 88.7 K.

Research on Real-Time Simulation of Steam Dump Control System in PWR Nuclear Power Plants

WANG Baosheng, WANG Dongqing, ZHANG Jianmin, QIU Jianwen, WANG Jian

2011, 32(5): 38-44.

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In order to analyze the dynamic response performance of steam dump control system,the real-time simulation model of steam dump control system in NPPs is established.Based on this model,a real-time calculation code is programmed to dynamically simulate the behavior of the control system.The visual and real-time simulation for closed loop condition,such as steady state and transient state,is performed with this code.The agreement between simulation results and the practical response of the control system is reasonable.The real-time property of the model can be verified for the steam dump control system.

Study on Fault Diagnosis Technology for Nuclear Power Plants Based on Time Series Data Mining

MU Yu, XIA Hong, LIU Yongkuo

2011, 32(5): 45-48.

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Time series data mining is applied to the fault diagnosis for nuclear power plants.dow method is used to convert the problem to a standard supervised learning problem.Simulation experiment is carried out by LOCA.The simulation results show that the diagnostic accuracy has certain improvement when the sliding-window method is applied.Furthermore,extracting the feature of the time series data in the sliding-window,the diagnostic accuracy is improved greatly.Some problems which can not be solved by classical algorithm can be diagnosed by time series data mining method.

Effect of Doped Titanium on Microstructure of UO₂ Microspheres

WANG Hui, LIU Jinhong, YIN Rongcai, REN Meng, MENG Qingyang

2011, 32(5): 49-53.

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This paper studies the effect of 0.7% of doped titanium on the microstructure of UO₂ microspheres.The results show that the increasing of the sintering temperature reduces the number of pores,but the volume of each single pore increases instead,which might be due to a fusion between the small pores.In the presence of the doped titanium,the growth rates of grain are inconsistent at a temperature of 1250℃ to 1450℃.Some grains begin to grow rapidly and the distribution of grain sizes shows a "double high" characteristics.However,at the temperature of 1550℃,the grain sizes are distributed in a normal way with the average grain size of about 42 μm.At the temperature of 1250℃ to 1450℃,UO₂ microspheres with doped titanium are sintered with the majority of titanium dioxide uniformly distributed inside the grains in the microspheres and a small amount of titanium dioxide enriching in the grain boundary.While at the sintering temperature of 1550℃,titanium dioxide is most enriched in the grain boundary.

Recovery of UMo Alloy from UMo/Al Dispersion Fuel Plates by Dissolution

REN Meng, LI Jia, LIU Jinhong, ZHU Changgui

2011, 32(5): 54-58.

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Methods for dissolving UMo/Al dispersion fuel plates in the compounded mixed basic aqueous（NaOH and NaNO₃） are studied on laboratory scale.After removing the clad and the matrix of the sub-standard UMo/Al dispersion fuel elements,the U loss ratios are calculated and the granularity distributions of the recovered UMo alloy powder are analyzed by the metallurgical microscope.Besides,the phase structure and the composition of the recovered UMo alloy powder are analyzed by the XRD.The results indicate that as the concentration of NaOH increases,uranium loss ratio increases;but as the concentration of NaNO₃ in-creases,U loss ration increases firstly and then decreases subsequently;generally,the U recovery ratios are more than 99.3%.The granularity of recovered UMo powders are very small and most parts of U have been oxidated to UO₂.Therefore,further study is required to determined whether the recovered UMo alloy could be returned to the product line.

Contrast of Aseismic Design for NPP Pressure Pipelines of Class 2

BAI Wenting, DAI Junwu, FENG Guozhong, RONG Feng

2011, 32(5): 59-63,68.

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At present,the RCC-M of French,ASME（2007） of U.S.A and GB50267-97 of China are the primary nuclear technical codes for the design of facilities,systems,and components,which are similar in the classification of nuclear facilities in nuclear power plants,but are not exactly the same in the piping design clauses of the class 2 pressure pipeline.For the earthquake input methods,GB,ASME and RCC-M are basically similar.The hard soil standard response spectrum of GB is relatively safer.RCC-M emphasizes on the impact of the pressure,the GB50267-97 and ASME emphasize more on the weight and other occasional loads such as earthquakes effects.RCC-M is safer than GB and ASME in level D criteria.Case analysis shows that in the conditions of lower pressure and the same stress intensification factor,GB and ASME criteria are safer than RCC-M.

Load Analysis of Hypothetical Ruptures and Optimum Design of H-Type Anti-Whip Restraint Devices for High Energy Pipes Based on Bilinearity Method

ZHANG Xingtian, CAO Feng, DING Youyuan, WANG Jianjun, FANG Jiang

2011, 32(5): 64-68.

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In order to deal with the dynamic effects induced by the hypothetical ruptures of the high energy pipes in nuclear power plants,the physical isolation and anti-whip restraint devices should be adopted based on the mechanical analysis during the NPP design.This paper not only introduces the calculation methods of jet forces,jet energy and characteristic lengths of circumferential rupture of high energy pipes based on the optimum mechanic mode,but also presents one new bilinearity method which could be used for the design and verification of the H type anti-whip restraint devices.

Thermal Hydraulic Behavior of SCWR Sliding Pressure Startup

FU Shengwei, ZHOU Chong, XU Zhihong, YANG Yanhua

2011, 32(5): 69-74.

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The modification to ATHLET-SC code is introduced in this paper,which realizes the simulation of trans-critical transients using two-phase model.With the modified code,the thermal-hydraulic dynamic behavior of the mixed SCWR core during the startup process is simulated.The startup process is similar to the design of SCLWR-H sliding pressure startup.The results show that maximum temperature of clad-ding-surface does not exceed 650℃ in the whole startup process,and the sudden change of water properties in the trans-critical transients will not cause harmful influence to the heat transfer of the fuel cladding.

Experimental Research on Evolvement of Density Wave Oscillation in Parallel Narrow Rectangular Channels

WANG Yanlin, CHEN Bingde, HUANG Yanping, WANG Junfeng

2011, 32(5): 75-79.

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Evolvement of flowrate oscillation in parallel narrow rectangular channels was investigated based on the visual experiment with a pair of 3-side visual rectangular channels.The experimental parameter ranges were: system pressure 0.2~0.8 MPa,mass flowrate 60.0~300.0 kg/h,and inlet subcooling 20.0~60.0℃.The evolvement procedure of bubbles and flow pattern transitions in rectangular channel（with a cross section of 40 mm×3 mm） was recorded by the high speed digital camera,while the heat flux of parallel rectangular channels increasing until DWO happened.The experimental results showed that the mass flux of the channels became unsteady as void fraction increasing.Flow transition boundaries,characteristic points of bubble behavior and DWO boundary were associated.

Improved Algorithm Based on Equivalent Enthalpy Drop Method of Pressurized Water Reactor Nuclear Steam Turbine

WANG Hu, QI Guangcai, LI Shaohua, LI Changjian

2011, 32(5): 80-82,88.

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Because it is difficulty to accurately determine the extraction steam turbine enthalpy and the exhaust enthalpy,the calculated result from the conventional equivalent enthalpy drop method of PWR nuclear steam turbine is not accurate.This paper presents the improved algorithm on the equivalent enthalpy drop method of PWR nuclear steam turbine to solve this problem and takes the secondary circuit thermal system calculation of 1000MW PWR as an example.The results show that,comparing with the design value,the error of actual thermal efficiency of the steam turbine cycle obtained by the improved algorithm is within the allowable range.Since the improved method is based on the is entropic expansion process,the extraction steam turbine enthalpy and the exhaust enthalpy can be determined accurately,which is more reasonable and accurate compared to the traditional equivalent enthalpy drop method.

Numerical Investigation of Heat Transfer Enhancement of Parallel-Plate Channel with Different Porous Media Insert Modes

ZHANG Hongjun, ZOU Zhengping, YE Jian, LIU Huoxing

2011, 32(5): 83-88.

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Based on the SIMPLE scheme on structured/unstructured collocated meshes,the code used for the prediction of the flow and heat transfer characteristics in porous media domain was achieved,where the momentum equations are formulated by Darcy-Brinkman-Forchheimer model;and the thermal equilibrium model is adopted for energy equation.The heat transfer characteristics of parallel-plate channel with porous inserts along the channel center and wall region were analyzed respectively.The results indicate that compared with the clear channel,the porous inserts can enhance the heat transfer,but cause large pressure drop as well.The performance has a great relationship with the porous media insert modes and filling ratio.Under the conditions of this paper,at low Re number,the channel with porous insert along the wall region has better comprehensive heat transfer performance.

Effect of Ocean Environment on Turbulent Heat Transfer in Lattices

YAN Binghuo, GU Hanyang, YU Lei

2011, 32(5): 89-95.

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The flowing and heat transfer characteristics of turbulent flow in channels with 7 and 4 fuel bundles in ocean environment is investigated theoretically.In the ocean environment,there is periodic heat accumulation on the channel wall due to the periodic velocity.The effect of rolling motion can be restricted greatly by the wall.In the channel with 7 fuel bundles,the effect of the additional force perpendicular to flowing direction on the flow is weak due to the restriction of the wall.The flowing and heat transfer characteristics is mainly determined by the axial turbulent intensity and inlet velocity.In the channel with 4 fuel bundles,since the restriction of the wall is limited,the effect of the additional force perpendicular to flowing direction on the flowing and heat transfer characteristics is significant.And the additional force perpendicular to flowing direction also has effect on the Reynolds stress.

Numerical Analysis of Bubble Rising Behavior in a Liquid Metal Using MPS

CHEN Ronghua, TIAN Wenxi, ZUO Juanli, SU Guanghui, QIU Suizheng, XU Jianhui

2011, 32(5): 96-99.

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Moving Particle Semi-Implicit（MPS） Method has advantages over the traditional mesh-based methods in the accurate capture of the vapor-liquid interface.In the present study,the numerical simulation of single bubble rising behavior in the liquid Pb-Bi alloy had been performed.The numerical results are provided for bubble shape deformation and rising velocity.Numerical simulation results indicate that as the bubble rises,the bubble exhibits in turn spherical,dimpled ellipsoidal,spherical-cap shapes.Terminal velocity of the bubble predicted by MPS agrees well with that predicted by Grace and increases with the initial bubble diameter.

Study on Multi-Scale and Double Fractal Characteristics of Gray Pulse Signals in Gas-Liquid Two-Phase Flow Images

LI Hongwei, ZHOU Yunlong

2011, 32(5): 100-104.

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Dynamic image videos of three typical flow patterns have been taken by a high speed camera in the horizontal tube of the gas-liquid two-phase flow.The average gray scale of each frame of the pulse signals is extracted,which decomposed by the multi-scale intrinsic mode function.Combined with R/S analysis method,the HURST exponent of each scale and double fractal characteristics is extracted.Two forms of the dynamic behavior of the bubbles and the single bubble are analyzed,which belong to the three typical gas-liquid two-phase flow patterns.The results are checked by extracting the kurtosis coefficient,and the variation of HURST exponent with the increase of the gas apparent velocity is discussed.The results show that the EMD decomposition combined with the R/S analysis can commendably reveal the nonlinear kinetic characteristics of gas-liquid two-phase flow.

Modification Design of Inlet Duct for a Helium Circulator

YANG Wei, WANG Hong, ZHANG Qinzhao

2011, 32(5): 105-108.

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A three dimensional steady turbulence numerical simulation technique was applied to simulate the internal flow of the inlet duct.A shear-stress transport k-ω turbulent model based on Reynolds averaged Navier-Stokes equations was used.Based on the simulation results,the origins for the internal flow losses were analyzed,and then a series of modification designs on the geometry of the inlet duct were carried out.The results showed that the flow losses in the inlet duct were mainly caused by the flow separation and vortex flow.After the modification,a best result of 56.3% loss decrease is achieved,which meets the engineering requirement.

Design of Experiment Apparatus for Testing Effects of Counter-Current Air-Water Flow on Containment Water Film Formation

LU Yanghui, LI Weihua

2011, 32(5): 109-112.

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This paper introduces the working principle of Passive Containment Cooling System（PCS） following a design basis accident.A flat-plate test bench and the accompanying measurement system are designed to study how the counter-current air affects the water film on the containment surface.Two key experiment parameters,flat-plate spacing and probe positions,are calculated and analyzed through fluid mechanics,experimental characteristic and experience correlation.The analysis results show that the experiment plan is reasonable and feasible,and it is proper when the flat-plate spacing is 150 mm and the probe position of the first rank is 1 m away from the flat-plate upside.

Analysis of Reliability Centered Maintenance for Service Water System in Nuclear Power Plants

WANG Zhenya, XIE Shenghua, TANG Guoxiang

2011, 32(5): 113-116,120.

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Reliability centered maintenance（RCM） methodology has been applied to a service water system in AP1000 nuclear power plant.Using the functional failure modes and effect analysis（FMEA） and logic tree analysis（LTA）,the optimized maintenance strategy is established based on a better understanding of the relevant information on this system and related equipments（including the functional failures,failure modes and effects）conclude.Compared with the current maintenance strategy,this maintenance strategy optimized by RCM conducts the condition monitoring and periodical maintenance for dominant failures but conducts periodical test for the recessive failures.

Study on Engineering Design for Reconstruction of Low-Level Liquid Radwaste Draining Systems

WANG Jinming, RONG Feng, XIE Hongyu, WANG Xin, LI Jinyan

2011, 32(5): 117-120.

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Potential safety hazards occur on the low-level liquid radwaste draining systems of nuclear facilities constructed during 60~70 s of the last century and reconstruction is demanded.By analyzing the present situation of the draining systems,the design principle and important factors for the design of the reconstruction engineering are discussed.Taking the reconstruction of the low-level radwaste draining system in the old building area as an example,the technical difficulties are studied and analyzed,and a design scheme is developed.

Installation and Management of RCP for CPR1000 Nuclear Power Station

WANG Yuxu, HUO Yabang

2011, 32(5): 121-124.

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The deposal of the typical cases and the technology innovation in the construction of CPR1000 reactor cooling pump are explored and summarized,and some suggestions to the future construction of the RCP in the coming CPR1000 projects are provided.

Installation and Commissioning of RCP Cartridge for Nuclear Power Stations

HUO Yabang, WANG Yuxu

2011, 32(5): 125-127,132.

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The installation process and method for the cooling pump sealing system of CPR1000 nuclear power plant reactor is studied,and various cases occurred during the installation and commisioning of RCP sealing systems are analyzed.The scheme for flushing of the nuclear loop in the various cases is developed to provide a reference for the installation and commissioning of RCP sealing system in the coming CPR1000 project.

Visualization Research of 3D Radiation Field Based on Delaunay Triangulation

XIE Zhangji, ZHU Bo, CHEN Yuqing, LI Shiting

2011, 32(5): 128-132.

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Based on the characteristics of the three dimensional partition,the triangulation of discrete date sets is improved by the method of point-by-point insertion.The discrete data for the radiation field by theoretical calculation or actual measurement is restructured,and the continuous distribution of the radiation field data is obtained.Finally,the 3D virtual scene of the nuclear facilities is built with the VR simulation techniques,and the visualization of the 3D radiation field is also achieved by the visualization mapping techniques.It is shown that the method combined VR and Delaunay triangulation could greatly improve the quality and efficiency of 3D radiation field visualization.

Effect of Change of Reheat Steam Temperature on Operation Economy of Nuclear Steam Turbine

LI Yong, WANG Chao

2011, 32(5): 133-137.

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Firstly,the reasons for the decreasing of the reheat steam temperature are analyzed.Secondly,the methods to calculate the effect of the changing of reheat steam temperature on the relative internal efficiency and ideal cycle thermal efficiency of the nuclear steam turbine are given,and the independence of relative internal efficiency and ideal cycle thermal efficiency of nuclear steam turbine is analyzed.Finally,the method to calculate the effect of the change of reheat steam temperature on the heat rate is given.The calculation results,by taking a 900 MW nuclear steam turbine as an example,show that when the reheat steam temperature decreases per 10 centigrade,the ideal cycle thermal efficiency decreases 0.0423%,the relative internal efficiency decreases 0.078%,and the heat rate of the turbine increases 0.1203%.

Combined Prediction Model for Supply Risk in Nuclear Power Equipment Manufacturing Industry Based on Support Vector Machine and Decision Tree

SHI Chunsheng, MENG Dapeng

2011, 32(5): 138-142.

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The prediction index for supply risk is developed based on the factor identifying of nuclear equipment manufacturing industry.The supply risk prediction model is established with the method of support vector machine and decision tree,based on the investigation on 3 important nuclear power equipment manufacturing enterprises and 60 suppliers.Final case study demonstrates that the combination model is better than the single prediction model,and demonstrates the feasibility and reliability of this model,which provides a method to evaluate the suppliers and measure the supply risk.

2011 Vol. 32, No. 5