核动力工程

Characteristic Statistic Algorithm （CSA） for In-Core Loading Pattern Optimization

LIU Zhi-hong, HU Yong-ming, SHI Gong

2007, 28(2): 1-4,18.

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To solve the problem of PWR in-core loading pattern optimization,a more suitable global optimization algorithm,i.e.,Characteristic statistic algorithm（CSA）,is used.The searching process of this algorithm and how to apply it to this problem are presented.Loading pattern optimization code SCYCLE is developed.Two different problems on real PWR models are calculated and the results are compared with other algorithms.It is shown that SCYCLE has high efficiency and good global performance on this problem.

Forecast of Fuel Cycle Shutdown Date and Evaluation of Fuel Cycle Length in Nuclear Power Station

CAI Guang-ming

2007, 28(2): 5-7,37.

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In nuclear Power Station,the purpose of the forecast of shutdown date for reactor fuel cycle and evaluation of fuel cycle length is to provide the input for the fuel management.Two methods for the fore-cast of shutdown date for fuel cycle,including their scope of application,are introduced in this paper.Calibration method of cycle length is also introduced in this paper.Several cycle lengths are evaluated by using the calibration method.At the end of this paper,error of calibration is analyzed..

Solving Point Reactor Kinetic Equations by Time Step-Size Adaptable Numerical Methods

LIAO Cha-qing

2007, 28(2): 8-12.

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Based on the analysis of effects of time step-size on numerical solutions,this paper showed the necessity of step-size adaptation.Based on the relationship between error and step-size,two-step adaptation methods for solving initial value problems（IVPs） were introduced.They are Two-Step Method and Em-bedded Runge-Kutta Method.PRKEs were solved by implicit Euler method with step-sizes optimized by using Two-Step Method.It was observed that the control error has important influence on the step-size and the accuracy of solutions.With suitable control errors,the solutions of PRKEs computed by the above mentioned method are accurate reasonably.The accuracy and usage of MATLAB built-in ODE solvers ode23 and ode45,both of which adopt Runge-Kutta-Fehlberg method,were also studied and discussed.

Study on Steady Natural Circulation Capacity of CARR

TIAN Wen-xi, QIU Sui-zheng, WANG Jia-qiang, SU Guang-hui, JIA Dou-nan, ZHANG Jian-wei

2007, 28(2): 13-18.

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An analysis code was developed to calculate the natural circulation capacity of China Advanced Research Reactor（CARR） under the conditions of different pool water temperatures.The influence of the pool water temperature on natural circulation characteristics was analyzed,and it was showed that the natural circulation flow rate increased however,the natural circulation capacity decreased with the increasing pool water temperature.Based on the computational result and the theoretical deduction,a correlation was proposed to predicting the relationship between the natural circulation mass flow and the core power under different conditions.The correlation prediction agreed well with the computational result,and the maximal error was less than ±10%.

Research for Rolling Effects on Flow Pattern of Gas-Water Flow in Horizontal Tubes

LUAN Feng, YAN Chang-qi

2007, 28(2): 19-23.

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The flow pattern transition of two-phase flow is caused by the inertial force resulted from rolling and incline of horizontal tubes under rolling state.An experimental study on the flow patterns of gas-water flow was carried out in horizontal tubes under rolling state,which rolling period is 15 second and rolling angle is 10 degrees,and a pattern flow picture is shown.It was found that there are two flow patterns in one rolling period under some gas flux and water flux.

Calculation of Frictional Pressure Drop of Annular Flow in Rolling Vertical Tubes

CAO Xia-xin, YAN Chang-qi, SUN Zhong-ning

2007, 28(2): 24-27.

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The calculation of the pressure drop of annular flow in rolling vertical tubes with three different diameters was studied.Based on separated flow model,it was found that calculations of annular flow frictional pressure drop with conventional Chisholm C-value were higher than the experimental data.Through the analysis,the Chisholm C-value under rolling condition was shown to decline exponent with the increase of slip ratio,then the correlation for C-value under rolling condition was proposed on the basis of the experimental data.Applying the new proposed correlation for C-value to separated flow model,it was shown that the calculations of annular flow frictional pressure drop correlated well with experimental data.

Experimental Research on Onset of Flow Instability in Narrow Rectangular Channel

WANG Yan-lin, HUANG Yan-ping, LU Dong-hua

2007, 28(2): 28-32,46.

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The onset of flow instability（OFI） was investigated under the condition of: P=1~15MPa,G=500~2000kg/（m².s）,ΔT_sat=20~100℃,q=40~1000kW/m² in a narrow rectangular channel（1000×25× 2mm） with upflow deionized water.The experimental results showed the effects of the system pressure,heat flux,inlet subcooling,and the exit equilibrium void fraction on OFI.A relationship between the Stanton number and Peclet number at OFI was proposed:St=0.542Pe^-0.438（Pe＜70000）,St=0.0042（Pe＞70000）.The deviation is within ±30% when heat flux is below 400kW/m² and within ±10% when heat flux is above 400kW/m².A new purely empirical correlation was developed by comparing the heat flux leading to OFI with that leading to saturation: q_OFI=1.95 q_ast^0.87.The deviation is within ±15% when heat flux is below 400kW/m² and within ±5% when heat flux is above 400kW/m².

J-Integral Study of Elbow with CircumferentialThrough-Wall Crack

HUANG Qing, ZANG Feng-gang

2007, 28(2): 33-37.

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The elbow is an important component in nuclear piping system,and also is the location where crack easily occurrs.To ensure the structural integrity of the nuclear piping system,a fracture mechanics analysis for the cracked elbows will be necessarily performed.In this paper,the 3-D elastic-plastic fracture mechanics methodology is used to study the elbows with a through-wall circumferential crack by using of ABAQUS software,and some numerical results for J-integral are obtained.The calculated results show that the influence of bending radius on J-integral is not associated with locations of the cracks.It is,however,intimate to sizes of the cracks.The effects on the J-integral for elbows with small size of cracks are insignificant.As to the J-integral for elbows with large size of cracks,the value of the J-integral is reduced as the R_b/R ratio of bending radii increases.In the case that the end rotating angle is kept unchanged,the relation between R/t ratio and J-integral is approximately linear,i.e.,a linear interpolation can be used.

Artificial Ground Motions Compatible with Both Specified Peak Displacement and Target Response Spectrum

ZHAO Feng-xin, ZHANG Yu-shan

2007, 28(2): 38-41.

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This paper proposed an artificial ground motion simulating method,which may be used to generate the artificial seismic ground shaking time history that satisfies both the target absolute acceleration response spectrum and the target peak ground displacement.Firstly,the method utilizes the traditional method,which modifies the Fourier amplitude spectrum in the frequency domain,to generate the initial acceleration time history,a（t）,with the given peak ground acceleration（A_PG）,response spectrum S_T（ω,ζ）,and intensity envelope being prescribed as targets;and then through superimposing the narrow-band time history in the time domain,the method further modulates a（t） to make its peak displacement approach the target peak ground displacement,i.e.,D_PG^T,as well as to improve its fitting precision to the target response spectrum.The numerical examples demonstrates that the results obtained by this proposed algorithm possess are with very high fitting precision.

Method of Power Self-Regulation of CFBR-II Reactor Based on DSP

BAI Zhong-xiong, ZHOU Wen-xiang

2007, 28(2): 42-46.

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To the control system of Power Self-regulation of CFBR-II Reactor,a new digital control scheme based on DSP has been brought forward.The TMS320F2812 DSP chip is adopted as the core controller to realize Power self-regulation of CFBR-II Reactor.In this paper,the successful program of DSP control system is introduced in both hardware and software technology in detail.

Design of Double-Fuzzy-Integral Intelligent Water Level Controller for Steam Generator

CAI Meng, ZHANG Da-fa, ZHANG Yu-sheng, ZHANG Guang-fu

2007, 28(2): 47-51.

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In order to effectively control the water level of steam generator,a double-fuzzy-integral-water level controller is designed in this paper.A “false water level” distinguishing fuzzy controller is added-before normal fuzzy controller,to control the water level effectively when SG is in “false water level” phase.In order to have an optimal water level control,this paper gives the result of fuzzy controller and integral-controller export in phase.By the analyzing of control curve,it has validated that the designed double-fuzzy-integral water level controller may overcome the disadvantageous influence given by “false-water level” phenomena to the water level control,and that the control excess adjustment quantity is small and the steady time is short.

Analysis on Features for Propagation of Zircaloy-4 Crack Growth

REN Wen-wei, Yin Guo-fu, LI Cong, CHEN Ling, QIU Shao-yu

2007, 28(2): 52-55.

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In this paper,the fractal properties and entropy properties of the low-cycle fatigue fracture of Zircaloy-4 are primarily investigated.By analyzing the trend of fractal dimension and entropy in the growth direction,obvious segmented zone are discovered in crack growth period.Along the growth direction of crack,the complexity falls gradually in low frequency and vertical high frequency.However,level high frequency and diagonally high frequency are the most complicated in the metaphase of crack growth period.Moreover,in despite of discrepancy of physical meanings and solving methods,fractal dimension and entropy have the accordant rules in characteristic curves in crack growth direction.It means that the fractal dimension and entropy are correlative.Finally,the crack growth period could be plotted out into some stages based on variate curve of the low frequency direction.The subsections in the other directions are analyzed separately as singular point of crack growth.

Study on Mechanism of UO₂ Freezing in Tube

WANG Zhe, CAO Xue-wu

2007, 28(2): 56-61.

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To study the mechanism of UO₂ freezing in fast reactor,the mechanism of models for UO₂ freezing in steel tubes are introduced in this paper,including conduction-limited freezing,bulk freezing and Fuel Caps freezing model.As an example,the GERSER tests are simulated and compared under these models.The conclusion is that Fuel Caps freezing model is a better mechanism model for interpretation of liquid UO₂ freezing in steel tubes,however,further experiments and model improvement are needed to study the freezing mechanism,including the effects of the pressure difference and flow regimes of melt leading edge.

Study on Irradiation-Assisted Stress Corrosion Cracking of Baffle-Former Bolts

DUAN Yuan-gang, XU Bin, TANG Chuan-bao

2007, 28(2): 62-65.

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Baffle-Former Bolts in PWR are made of austenitic stainless,locating in the area of strong Neutron irradiation for a long time.When Neutron irradiation exceeds the limit,even if the stress level is very low,the plasticity of the material will lose step by step because of the defects,phase transformation and grain boundary segregation,and then Irradiation-Assisted Stress Corrosion Cracking takes place.Techniques for mitigating IASCC is put forward: using high purity materials,quality control of water,and improvement of structure.

Design of Research Reactor Emergency Power Supply System

WANG Ming-shan, ZU Shi-lei, LIU Chuang, XIA Ming, ZHANG Yang

2007, 28(2): 66-68.

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This paper has introduced the design principle,system equipments,capabilities and operating working conditions,including normal running,prediction running events and accident conditions.ways of the emergency power supply system of research reactor,which consisted of single-phase 40kVA UPS,three phase 160kVA UPS,storage battery and power distribution system.The results show that the system equipments and main operational parameters have met the requirements of research reactors in different

New Techniques for Prototype Manufacturing of Great Conical Ring and Its Numerical Simulation

WANG Ze-wu, YUAN Hai-lun, ZENG Qing, WANG Cheng

2007, 28(2): 69-72.

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Aiming at solving the difficulties of prototype manufacturing of conical ring for the steam generator shell of 600MW Nuclear Reactor,this paper developed a new techniques for the prototype manufacturing of the great conical ring utilizing ring rolling technology based on RAW200/160-5 ring rolling.And based on the common dynamical explicit FEM software Ansys/Ls-dyna,a 3D model of Radial–axial Ring Rolling Mill is created to simulate the dynamic process of ring rolling within a conical ring producing period so as to prove the reliability of the new techniques.The result of the simulation also shows that the method can observe the real time dynamical process of ring enlarging and fleck generating;optimizing the billet structure and the rolling techniques,and at the same time,the repeated tests can be avoided.

Development of Conductor Feedthrough Module of LV Electrical Penetration Assembly for Research Reactors

LUO Zhi-yuan, WANG Guang-jin, ZHOU Bin

2007, 28(2): 73-76.

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A LV electrical penetration assembly with perfusion sealing conductor feedthrough module was developed,which can be used for the connection of internal and external cables through the wall of the research reactor workshop.The LV electrical penetration assembly was combined with several independent modules.The maintenance and replacement of the assembly can be easily done in service.The sealing of conductor feedthrough module was achieved with the perfusion of self-extinguishing epoxy.The leakage be-tween the conductor feedthrough module and the end plate module was blocked with rubber rings.The result of the leakage test and the electrical performance test for the samples of conductor feedthrough module satis-fied the requirement of research reactor.The structure of the new electrical penetration assembly is simple and compact.It can be manufactured with mature technology and cost low price.The performance of the assembly is steady.It can be used widely in research reactors.

Coupling Analysis of Frictional Heat of Fluid Film and Thermal Deformation of Mechanical Seal End Faces

ZHOU Jian-feng, GU Bo-qin

2007, 28(2): 77-81.

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The heat transfer model of the rotating ring and the stationary ring of mechanical seal was built.The method to calculate the frictional heat that transferred by the rings was given.The coupling analysis of the frictional heat of fluid film and thermal deformation of end faces was carried out by using FEA and BP ANN,and the relationship among the rotational speed ω,the fluid film thickness h_i on the inner diameter of sealing face and the radial separation angle β of deformed end faces was obtained.Corresponding to a given ω,h_i and β can be obtained by the equilibrium condition between the closing force and the bearing force of fluid film.The relationship between the leakage rate and the closing force was analyzed,and the fundamental of controlling the leakage rate by regulating the closing force was also discussed.

Analysis on Spurious Turbine Runback of the Unit 2 at Qinshan Ⅱ NPP

RUAN Liang-cheng, YU Zhong-de, CHEN Zhong-wu, HONG Yuan-ping, CHEN Yi-feng, WU Jun-yi, DING Jian-yang

2007, 28(2): 82-86.

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This paper described the turbine runback transient resulted by spurious signal of low-selected reactor coolant average temperature,which happened at Unit 2 of Qinshan Ⅱ NPP.Changes of key parameter during the transient were provided,and responds of the unit control systems（especially Rod Control System and Steam Bypass Dumping System） were analyzed.Since high-selected reactor coolant average temperature was adopted as the input variable for Rod Control System and Steam Bypass Dumping System,during the overall transient,the control systems could cooperate well according to the difference value between average temperature and reference temperature,meanwhile the other main control systems such as the steam generator level,the pressurizer level and pressure could respond correctly and timely,finally the unit was stabilized at lower power state.The data from the transient will be used as a base for the plant modification.

Application of Passive Residual Heat Removal System under Blackout Accident of Chinese Advanced Nuclear Power Plant

SHEN-Jin, JIANG Guang-min, TANG-Gang, YU Hong-Xin

2007, 28(2): 87-90.

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The passive residual heat removal（PRHR） system provides emergency core decay heat removal during the transient accidents such as blackout or loss of forced reactor coolant flow.Referenced to the design of AP1000,Chinese advanced nuclear power plant applied PRHR to mitigate the accident consequence only by the natural circulation.The blackout accident was calculated to identify the capacity of PRHR of Chinese advanced nuclear power plant by using RELAP5/MOD3 code.The analysis results show that PRHR of HLS can build the natural circulation and remove the core residual heat after the blackout accident successfully.

Reliability Analysis of Safety Injection System of Nuclear Reactor in GO Methodology

GUO Qiang, ZHAO Xin-wen, CAI Qi

2007, 28(2): 91-94.

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This paper analyzes a representative operation process of Safety Injection System of nuclear reactor in GO Methodology,gives the dynamic formulas of failure probability for phases in the operation process of the system,and presents a practical calculating example.Its result shows that GO Methodology directly reflects the change trend of failure probability in the operation process of the system;and GO Methodology is an effective methodology of reliability analysis in the possessing-liquid,multi-state,related-time,complicate system.

Research on Method of Nuclear Power Plant Operation Fault Diagnosis Based on a Combined Artificial Neural Network

LIU Feng, YU Ren, LI Feng-yu, ZHANG Meng

2007, 28(2): 95-100.

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To solve the online real-time diagnosis problem of the nuclear power plant in operating condition,a method based on a combined artificial neural network is put forward in the paper.Its main principle is: using the BP neural network for the fast group diagnosis,and then using the RBF neural network for distinguishing and verifying the diagnostic result.The accuracy of the method is verified using the simulation values of the key parameters in normal status and malfunction status of a nuclear power plant.The results show that the method combining the advantages of the two neural networks can not only diagnose the learned faults in similar power level of the nuclear power plant quickly and accurately,but also can identify the faults in different power status,as well as the unlearned faults.The outputs of the diagnosis system are in form of the reliability of the faults,and are changing with the lasting of the operation time of the plant.This makes the diagnosis results be more acceptable to operators.

Quality Assurance and Quality Control of Nuclear Engineering during Construction Phase

ZHANG Zhi-hua, DENG Yue, LIU Yao-guang, XU Xian-qi, ZHOU Shan, QIAN Da-zhi, ZHANG Yang

2007, 28(2): 101-104.

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The quality assurance（QA） and quality control（QC） is a very important work in the nuclear engineering.This paper starts with how to establish quality assurance system of nuclear engineering construction phase,then introduces several experiments and techniques such as the implementation of quality assurance program,the quality assurance and quality control of contractors,the quality surveillance and control of supervisory companies,quality assurance audit and surveillance of builders.

Research of Nuclear Engineering Project Management Based on Split Package Approach & Multiple Package Approach

CHEN Chang-bing, LI Hui-qiang, CHENG Ping-dong

2007, 28(2): 105-110.

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On the basis of the exploration of different models of nuclear engineering project management containing the different construction patterns of split package approach & multiple package approach in China,aiming at the construction patterns,this paper introduces a kind of typical construction of project management model by analyzing the relationship of commission and agency between the proprietor of the nuclear plants and the specialized nuclear engineering company,.According to the specific characteristics of nuclear engineering,this paper designs the organization system of project management and illustrates the various responsibilities of the proprietors,nuclear engineering companies and other major partners.

Discussion of Manage Mode for Nuclear Power Construcation in China

GAO Ming-shi, CHEN Hua

2007, 28(2): 111-114.

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This paper analyzed the development status of management mode for NPP construction and nuclear power engineering companies.Considering the national development plan of nuclear power,and making reference of the experiences of the successful construction of NPPs,the management mode for NPP construction in which the nuclear engineering companies are the main factors have been discussed.This paper proposed that EPC/TurnKey as the management mode for the nuclear power construction,led by the owner,and constructed by engineering companies according to the contracts,so as to establish a construction group with expertise knowledge.

Modeling on a PWR Power Conversion System with System Program

GAO Rui, YANG Yan-hua, LIN Meng

2007, 28(2): 115-118,123.

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Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station,this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5.To simulate the full-scope power conversion system,not only the traditional basic system models of nuclear island,but also the major system models of conventional island are all considered and modeled.A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station,and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time.

Design of In-Core Refueling Management Subsystem for Tianwan Nuclear Power Plant

YAN Jing, CHI Hu-qing, JIANG Ping, LI Guo, LI Yuan-wen

2007, 28(2): 119-123.

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The core of Tianwan NPP is refueled by in-core refueling technology.A set of refueling management subsystem was developed,which is one important part of the fuel management system.It can provide assistance to the management staff by the graphics,to accomplish the step design and optimization for in-core refueling.This subsystem is based on B/S structure and developed with ASP.NET.,which adopts the methods of vector plotting,hidden committing and picture buffer.This paper describes the design idea of the subsystem and mainly introduces the resolution of graphics technology.

Development of a Beam Transient Code for ADS

YU Tao, LI Ji-gen, LIN Qiu, SHI Yong-qian, LUO Zhang-lin, RONG Yong-hua

2007, 28(2): 124-127.

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Accelerator Driven System（ADS） will be a transition between the current nuclear system and the complete application of fusion energy.Because the maintain of power level in ADS relies on the neutron source generated through intensive proton beam hitting spallation target,the instability of proton beam as well as trips in any form will influence the reactor power level and further the safety of ADS.Currently,the research of trip problem is becoming internationally a hot issue in ADS research.Beam transient change accident in ADS is extensively studied in this paper,and the physical and mathematical simulation models is established.Finally,the special software-SIMULINK-ADS dedicating to ADS beam transient change accident is designed and developed.Through the analysis of typical beam transient change accident,simulation results are obtained and recorded,which tally satisfying with the results of OECD/NEA and FZK Karlsruhe research institutions.The results show that SIMULINK-ADS can effectively compute and analyze the sub-critical physics and transient thermal-hydraulic characteristics of ADS.

2007 Vol. 28, No. 2