核动力工程

2009, 30(S1): .

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Numerical Simulation of Flow Behavior in Tight Lattice Rod Bundle

YU Yi-qi, YANG Yan-hua, GU Han-yang, CHENG Xu, SONG Xiao-ming, WANG Xiao-jun

2009, 30(S1): 1-5.

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The Numerical investigation is performed on the air turbulent flow in triangular rod bundle array.Based on the experimental data,the eddy viscosity turbulent model and the Reynold stress turbulent model are evaluated to simulate the flow behavior in the tight lattice.The results show that SSG Reynolds Stress Model has shown superior predictive performance than other Reynolds-stress models,which indicates that the simulation of the anisotropy of the turbulence is significant in the tight lattice.The result with different Reynolds number and geometry shows that the magnitude of the secondary flow is almost independent of the Reynolds number,but it increases with the decrease of the P/D.

Low-Temperature Release Mechanism of Kr and Xe in Fuel Pellet

ZHOU Yi

2009, 30(S1): 6-8,13.

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The low-temperature release mechanism of fission gas Kr and Xe has been studied.A model has been developed to calculate the pellet chamfer surface average release efficiency,based on the recoil micro-mechanism model of Wise,by considering the effect of pellet chamfer.After that,a more detailed micro-mechanism model of fission gas low-temperature release has been developed,using the knockout micro-mechanism model of Olander for reference,which can be applied to the final fission gas low-temperature release model.

Angular Discretization of Neutron Transport Equation Based on Daubechies Wavelets

ZHENG You-qi, WU Hong-chun, CAO Liang-zhi, YU Ying-rui

2009, 30(S1): 9-13.

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The recent studies on the MOX fuel cell calculation indicated that the angular flux is a bumpy and irregular function,which is difficult to be represented by using traditional methods.The Daubechies wavelets,possessing the properties of orthonormality and compact support,are applied in the angular discretization of the neutron transport equation to solve this problem.In order to improve the computational stabilization,a coupled S_N and wavelet expansion scheme is applied instead of the common two-dimensional wavelet expansion.Numerical results demonstrated that this method is accurate and applicable for both traditional and challenging problems.

Research on and Code Development of Direct Simulation Method for Neutron Space-Time Kinetics

SHEN Hua-yun, WANG Kan

2009, 30(S1): 14-17.

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DSM （Direct Simulation Method） is a new method for the transient,three-dimensional neutron transport problems.In this method,the transient problems are solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process.It is with universal use since it eliminates various approximations.In this paper,the method is studied in detail and the relevant transient analysis code is developed.Finally,some numerical results are presented.

Application of Wavelets Scaling Function Expansion to Continuous-Energy Neutron Transport Calculation

YANG Wei-yan, WU Hong-chun, CAO Liang-zhi, ZHENG You-qi

2009, 30(S1): 18-22,38.

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Wavelets expansion method can simulate the functions oscillating violently.It has been applied to simulate the violently variety of the angular flux,and has obtained satisfactory results.Neutron spectrum has violent oscillations within the resonant energy range.This paper introduces a method to discretize the energy variable of neutron transport equation by coupling the multi-group method and wavelets expansion method,which simulates the energy-dependent angular flux with wavelets scaling function within the resonant energy range and uses multi-group method within fast and thermal neutron energy ranges.The preliminary calculation results indicate that this method can obtain the accurate k-inf,as well as determine the fine-structure energy distribution of the angular flux,which is significant to resonance self-shielding calculation.

Validation of Coupled RELAP5-TDNK Code Using OECD/NEA Core Transient Benchmark

LI Feng, ZHANG Yu, JIANG Guang-ming

2009, 30(S1): 23-27.

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In order to verify and validate the coupled RELAP5-TDNK code,analysis of the rod ejection accident framed by OECD/NEA was conducted.Three dimensions neutronics model is introduced to RELAP5 in RELAP5-TDNK which advances the ability to analyze the transients with strong feedback phenomenon.The models of the rod ejection are established and two problems are processed.Compared to other coupled codes,RELAP5-TDNK gets a good accuracy and has the ability to analyze the transients with strong feedback phenomenon.

International Research Progress of CFD Application in Analysis of Nuclear Power System

LI Lin-sen, WANG Kan, SONG Xiao-ming

2009, 30(S1): 28-33.

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This paper introduces the latest international research progress of CFD application in nuclear reactor system analysis.CFD method has been applied to a few 3-D single phase transient simulations,including flow field modeling of the reactor cores,assemblies,and vessel plenums.On the other hand,CFD method applied to reactor system still needs further validation and benchmarking,meanwhile,the application of CFD also needs to be studied,including the setup of the Best Practice Guidelines （BPG）.Furthermore,CFD codes are used to couple with thermal-hydraulic system codes or neutronic codes.Eventually,in two phase field and turbulence modeling,CFD codes are still being developed.

Research on Application of CFD Method in Thermal-Hydraulic Performance Analysis of Rod Bundle Grid

CHEN Wei-hong, ZHANG Hong, ZHU Li, XIONG Wan-yu

2009, 30(S1): 34-38.

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UG code and Computational Fluid Dynamics method are used to generate the 5×5 rod bundle grid of AFA-2G geometry and simulate the single phase three flow field,including the qualitative and quantitative flow field analysis under specific velocity,and the resistance characteristics under different velocity.The computational results and experimental results are compared.The comparison shows that CFX code can simulate the three flow fields of single phase coolant in the rod bundle grid channel using reasonable geometry,appropriate physical model and arithmetic.The simulation results are well corresponding with the experiments.

CFD Research on Flow and Heat Transfer Characteristics of Coolant in Pebble Bed Reactor Core

LI Hua, QIU Sui-zheng, SU Guang-hui, SONG Xiao-ming

2009, 30(S1): 39-43.

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The 3-D physical and geometrical model of the water cooled pebble bed reactor were built using the commercial software ProE and CFX,in which the of the coolant were studied based on the computational fluid dynamics （CFD）.The velocity field,temperature field and pressure distribution of the coolant are obtained and analyzed in the cases of inter-pebble gap and direct contact,and its effects on the core safety are also analyzed.

Application of Nuclear Safety Principle for Reactor Protection System in QinShan Phase II NPP Extension Project

XIAO Peng, XU Dong-fang, FENG Wei

2009, 30(S1): 44-47.

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The design of reactor protection system in Qinshan phase II NPP extension project（3# and 4#） is optimized based on that of QinShan phase II NPP project（1# and 2#）.This paper describes two aspects,one is a realized improvement based on the research,and the other is a NPP owner’s suggestion that is not accepted.The application of nuclear safety principle for reactor protection system is introduced in this paper.

Research on Engineering Simulator for Function Validating of DCS in Nuclear Power Plant

LIU Peng-fei, LIN Meng, HOU Dong, CHEN Zhi, YANG Yan-hua

2009, 30(S1): 48-51,64.

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An engineering simulator for the function validating of Distributed Control System in Nuclear Power Plant（NPP） was developed in this paper.In the engineering simulator,the thermal-hydraulics was modeled by Relap5,the main control system of the NPP was modeled by Matlab/Simulink,the database was built by MySQL,and the control panel was developed by the Visual Studio.NET.Data acquisition system was used to realize the real-time communication between the simulator and the real Distributed Control System in the NPP.The validating results show that the simulator can meet the requirements of validating the hardware and logic control system of DCS in NPP.

Automatic Start-up System of Nuclear Reactor Based on Sequence Control Technology

ZHANG Yao, PENG Hua-qing, ZHANG Da-fa

2009, 30(S1): 52-55.

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A conceptive design of an automatic start-up system based on the sequence control for the nuclear reactors is given in this paper,so as to solve the problems during the start-up process,such as the long operation time,low automatic control level and high accident rate.The start-up process and its requirements are analyzed in detail at first.Then,the principle,the architecture,the key technologies of the automatic start-up system of nuclear reactors are designed and discussed.With the designed system,the automatic start-up of the nuclear reactor can be realized,the work load of the operator can be reduced,and the safety and efficiency of the nuclear power plant during its start-up can be improved.

Research on Flow Distribution in UTSG under Low Flow Rate Condition

ZHANG Yong, SONG Xiao-ming, HUANG Wei

2009, 30(S1): 56-59,64.

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The CFD method is used to study the flow characteristics in the primary side of a vertical,inverse U-bend tube steam generator in this paper.The study result shows that the reversal flow occurrs in the inner short tubes,with the structure data of the U tubes of the natural circulation simulation test facilities.With the low flow rate,when the inlet flow rate is increased,the reversal flow rate will be decreased,and the number of reversal flow tubes will be decreased as well,till it disappears.When the temperature in the second side is increased,the reversal flow rate will be decreased,and the number of reversal flow tubes will be decreased as well,till it disappears.

Theoretical Study on Effect of Ocean Conditions on Single-Phase Forced Circulation

DU Si-jia, ZHANG Hong

2009, 30(S1): 60-64.

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Based on the basic governing equations,the effect of the ocean conditions on the flow characteristics of the forced circulation is analyzed,and is simulated and verified by CFD calculation method.The calculation results showed that under a given velocity,the gravitational force,centripetal force and angular acceleration force caused little effect on the flow configuration and frictional resistance.But the Coriolis force caused by rolling effect,can influence the flow configuration,and with the larger palstance,it will influence the frictional resistance.

Numerical Simulation of Flow and Heat Transfer Characteristics Focused on Gas-Liquid Two-Phase Flow Case in Core Simulator of MIPR

NIE Hua-gang, SONG Xiao-ming, NIU Wen-hua

2009, 30(S1): 65-69.

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In this paper,the CFD technique is used to conduct the numerical simulation study which aims at a body of 1/20 of the core simulator of the Medical Isotopes Production Reactor （MIPR）,and non-fabric gridding is adopted to compartmentalize the model to study the gas-liquid two-phase flow case.The flow field,temperature field and heat transfer characteristics are analyzed as well.The comparison results show a good agreement between the computation and the experimental results.

Study on Generation of Artificial Accelerogram from Shock Response

WU Wan-jun, LIU Wen-jin, ZHANG Yixiong

2009, 30(S1): 70-71.

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This paper transforms the shock response spectrum to the artificial random accelerogram based on that from the seismic response spectrum and the shock characteristics.An example is presented and comparison between the artificial accelerogram and the target spectrum is also performed,then the conclusion can be reached that the artificial accelerogram is well compatible with the target spectrum.

Virtual Maintenance Technology for Reactor System Based on PPR Technology

WU Ya-xiang, MA Bai-yong

2009, 30(S1): 72-76.

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Based on the Product,Process and Resources （PPR） technology,the establishing technology of virtual maintenance environment for the reactor system and the process structure tree for virtual maintenance is studied,and the flow for the maintainability design and simulation for reactor system is put forward.Based on the subsection simulation of maintenance process and layered design of maintenance actions,the leveled structure of the reactor system virtual maintenance task is studied.The relation for the data of product,process and resource is described by Plan Evaluation and Review Technology （PERT） diagram to define the maintenance operation.

Study on Categorization of Components Based on Binary Importance Decision

CENG Wei, YU Hong-xing, SUN Yu-fa

2009, 30(S1): 77-81.

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There are some problems for the current probabilistic component categorization in terms of the selection and the calculation of the importance measures,and the confirmation of the categorization threshold.The combination of binary importance decision,the additivity transferring model and the risk evaluation standard based on the aftermath can solve the problems.The binary importance categorization decision is validated to be reasonable based on the probabilistic safety analysis model for Qinshan phase II nuclear power station.

Application of Computer Aided Tolerance Ananlysis in Product Design

DU Hua

2009, 30(S1): 82-85.

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This paper introduces the shortage of the traditional tolerance design method and the strongpoint of the computer aided tolerancing （CAT） method,compares the shortage and the strongpoint among the three tolerance analysis methods,which are Worst Case Analysis,Statistical Analysis and Monte-Carlo Simulation Analysis,and offers the basic courses and correlative details for CAT.As the study objects,the reactor pressure vessel,the core barrel,the hold-down barrel and the support plate are used to upbuild the tolerance simulation model,based on their 3D design models.Then the tolerance simulation analysis has been conducted and the scheme of the tolerance distribution is optimized based on the analysis results.

2009 Vol. 30, No. S1