核动力工程

In-Core Fuel Management Design for CNP1500

LI Dong-sheng

2006, 27(4): 1-4.

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Abstract:
CNP 1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies.The active core height at cold is 426.4cm and equivalent diameter is 347.0cm.The reactor thermal output is 4250MW,and the average linear power density is 179.5W/cm.Calculation results show that the cycle length of equilibrium cycle core reaches to 470 equivalent full power days,and for all cycles,the moderator temperature coefficients at all conditions are negative values,the nuclear enthalpy rise factors F_△H at hot full power,all control rods out and equilibrium xenon are less than the limit value,the maximum discharge assembly burnup is less 55000MW·d/t（U）,and the shutdown margin values at the end of life meet design criteria.The low-leakage core loading reduces ra- diation damage on pressure vessel and is beneficial to prolong use lifetime of it.This paper presents the in-core fuel management design scheme and main calculation results:for CNP 1500 nuclear power plant.

Fuel Array Project for the Sub-critical Reactor of Accelerator Driven System Verification Facility

YU Tao, SHI Yong-qian, XIA Pu, LIAO Yi-xiang, XIAN Wen-ping, DU Hai-dong

2006, 27(4): 5-7.

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As the sub-critical reactor of the verification facility“Venus”of Accelerator Driven System, the fast-thermal coupling project is introduced in sub-critical core design.The fast neutron energy spectrum area is composed of natural metal uranium components;the thermal neutron energy spectrum area is com- posed of low-density uranium components.The critical calculation is carried out with Universe card and Fill card of 3-D Monte Carlo code（MCNP）.Based on the request of the verification facility（k_eff=0.90～1.00）,the core fuel grid of thermal area is schemed out.

Fuel Assembly Few Group Parameter Calculation Code AFGPB and Benchmarking

YAO Dong, LI Da-tu, YU Ying-rui, MA Yong-qiang, YIN Qiang

2006, 27(4): 8-12.

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This paper briefly introduced theoretical model and methodology of the reactor fuel assembly few group parameter calculation code AFGPB.The verification and validation of the code have been made by calculations of the IAEA plate-type fuel assembly benchmark problem and comparisons are made with the power distribution for CE company fuel assembly between the calculation values of the AFGPB, TPFAP and EPRI-CPM codes.The numerical results indicated that the calculated values of AFGPB are in good agreement with those of other international organizations.

Feasibility Study for Monte Carlo Method and Its Application on HFETRγ-Heat Calculation

QIU Li-qing, JIA Dou-nan, DENG Cai-yu, FU Rong, HU Yue-chun, WANG Zhen-dong

2006, 27(4): 13-15.

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In this paper the Monte Carlo Method is used to calculate theγ-heat of the HFETR core. Comparison of MCNP and experimental results ofγ-heat in the core of HFETR is made.The Comparison shows that it is feasible to use MCNP to calculate theγ-heat of HFETR with satisfied precise.MCNP can be used for calculation ofγ-heat in any location of the HFETR core.

Analysis of Forced-Convection Flow Instability in Steam Generator

WU Ge-ping, QIU Sui-zheng, SU Guang-hui, JIA Dou-nan

2006, 27(4): 16-20,25.

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A numerical method is developed to predict the steady-state and transient behaviors of forced-convection flow in steam generator.From the basic differential conservation equations,an intact model is constructed fully considering the structure of the steam generator,its possible flow and heat transfer.Gear method is adopted,The curves of flow instabilities and unstable regions are acquired with computation.The effects of pressure,flow flux,inlet subcooling and inlet/outlet throttle on the system behaviour are discussed.

Analysis on Radiation Heat Transfer between Pressure Tube and Calandria Tube of TACR

XU Liang-wang, JIA Bao-shan

2006, 27(4): 21-25.

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Heat radiation is the primary way for the heat transfer between pressure tube and calandria tube of TACR,and its accurate analysis is very important for design and safety analysis of TACR.A calculation model of radiation heat transfer between the pressure tube and calandria tube of TACR,which is based on grey body radiation model and electronic network method,has been established.The cases with temperature boundary condition and heat flux boundary condition are discussed,and the results prove that this model can be applied to the analysis of radiation heat transfer between the pressure tube and calandria tube.

Processing of CHF Experimental Data Base

DAN Jian-qiang, D.C.Groeneveld, S.C.Cheng

2006, 27(4): 26-29,39.

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Abstract:
The world vertical upward flow round critical heat flux experimental database is analyzed based on the slice method,similarity measure,heat balance and inlet temperature verification.Finally,a re- vised CHF data is obtained after abandoning 522 bad database points,326 outliers,1640 duplicated data,619 heat unbalance data and 10 inlet temperature violation data.The present work is the base for the development of new CHF models,new corrations and new CHF look up table.

Structure Analysis and Sealing Design of Reactor Building for China Advanced Research Reactor

LI Zhong-xian, RONG Feng, DONG Zhan-fa, FU Ji-yang

2006, 27(4): 30-34,43.

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Abstract:
CARR building is a typical structure of short-term cycle,which is a complex consisted of various structural types and structural cells of different materials,integrated internal force of the partial struc- ture is analyzed by Algor code,local stress of the building structure is analyzed by ANAYS code, and calcula- tion is carried out for the pre-stressed concrete structure by PREC code.The analysis and calculation show that the maximum displacement of the structure occurs in the middle of the side wall,and the displacement under hit is about 2.28mm.The parts with larger stress are the interfaces between the top plate and side wall, the side wall and the floor plate,and the side walls.The maximum stress is 2.7MPa.Calculated deflection of the girder is 13.5mm,camber is 7.5mm,and pre-stress is 0.745.In order to control the crack of the reinforced concrete components,partial roof is made of pre-stressed concrete structure.The inner liner for sealing of the building is expoxy and glass fabric.

New Finite Element Method to Analyze Crack Tip of Plate Structure

ZHANG Ji-ping, CHEN Qiu

2006, 27(4): 35-39.

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In this paper a new finite element method to analyze the crack tip is proposed.Firstly the stiffness matrixes of the singular elements are completed on the base of the displacement field near the crack tip of the Reissner plate.Then the stiffness matrix of the singular element corresponding with the broad sense displacements is transformed into the stiffness matrix corresponding with the node displacements,so the latter can join the total stiffness directly.Finally the relativity between the displacements of the two singular ele- ments is relieved,and the final stiffness is advanced.The new method,employing the transforming of the dis- placement and the conservation of the strain energy,focuses the calculation on the singular elements.So it not only decreases the quantity of the elements in the filed of the fracture,but also simulates the singularity of the crack tip,and the calculation near the crack tip is simplified not only in the theory but also in the calculation.

Effect of Temperature on Iodine-Induced Stress Corrosion Cracking of Zr-Sn-Nb Alloy

PENG Qian, ZHAO Wen-jin, LI Wei-jun, TANG Zheng-hua, CUI Xu-mei, HENG Xue-mei

2006, 27(4): 40-43.

Abstract(15) PDF(0)

Abstract:
The behaviors of iodine-induced stress corrosion cracking of N 18 and N36 alloys at different temperatures have been investigated.The characterizations of the SCC fractures have been examined by SEM.The results show that,for recrystallized zirconium alloys,the critical stress intensity factor K_ISCCand stress decrease,as well as critical time shortens with test temperature increasing;for stress relieved zirconium alloys,the critical stress intensity factor K_ISCC almost keeps unchanged,but the critical stress decreases and critical time shortens when temperature changes from 300℃to 350℃.The higher temperature is,the more the corrosion product is.

On Cyclic Deformation of 304 Stainless Steel under Strain-Controlled Nonproportionally Multiaxial Loading at High Temperatures

ZHANG Juan, GAO Qing, KANG Guo-zheng, LIU Yu-jie

2006, 27(4): 44-49.

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In order to accurately describe the strain-controlled cyclic deformation of material at high temperatures,an experimental research was carried out on the cyclic deformation of 304 stainless steel sub- jected to uni-axial and non-proportionally multi-axial cyclic straining at 350℃C and 700℃.The cyclic hard- ening behavior under different loading paths and different load conditions was studied.It is shown from the research that the non-proportionally multi-axial strain-controlled cyclic deformation of 304 stainless steel at high temperatures has distinct dependence on temperatures and loading paths.This result is useful to estab- lish constitutive model for the cyclic deformation of the material.

A Mechanism Investigation of Core Melting in Nuclear Reactor Using Molecular Dynamics Method

CHEN Shuo, SHANG Zhi, ZHAO Jun

2006, 27(4): 50-53.

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In this paper,the molecular dynamics method is employed to simulate the physical process under the severe accident of core melting in nuclear reactors.Based on the numerical simulation results,some mechanics mechanisms are discovered.The whole motion process of the melted metal particle from the core melting to dropping onto the low pipe sheet has been captured by current simulation results.Using molecular dynamics,a further research for the severe accident of core melting in nuclear reactor could be carried out.

Quantification Method of Cognitive Error and Its Application in PSA

WANG Yao, HE Xu-hong, SHEN Zu-pei, HUANG Xiang-rui

2006, 27(4): 54-58.

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Abstract:
Cognitive Reliability and Error Analysis Method （CREAM） is one of the representative second-generation human reliability analysis methods.The CREAM quantitfication analysis is improved based on the core thought of CREAM,and the performance influence index of the common performance condition （CPC） on human reliability and the context influence index which quantifies the effect of the con- text on human error are presented.A quantification method of the error probabilities for the human failure events in probability safety assess （PSA） is provided,and it is applied in the error probability analysis for the human actions in the identification and isolation of the ruptured steam generator tube of nuclear power plants.

Theoretical Design of an Epithermal Neutron Radiation Field

ZHANG Xiao-min, ZHANG Wen-zhong, LUO Yi-sheng

2006, 27(4): 59-63.

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In order to get an Epithermal neutron radiation field for the using of Boron neutron capture therapy（BNCT）,based on the Tsinghua University experimental reactor neutron source,the research pro- posed two engineering schemes through MC simulating method.In the end the differences between the two schemes were analyzed and contrasted,and the result showed that Scheme 1 is superior to Scheme 2,which was consequently selected for the BNCT system.

Status of Boiler Effect for Safety-Related Gate Valves and Its Improvement

WANG An, XIANG Wen-yuan, LI Shu-zhou

2006, 27(4): 64-67.

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Based on international experimental and research results for Boiler Effect,this paper describes the analysis principle and method for Daya Bay ＆ Ling’ao nuclear power stations to confirm which gate valves are affected by boiler effect.Further analysis with PSA and detailed safety functions is to define finally RIS063/064VP and PTR022VB need to be modified.The solution avoiding Boiler Effect for RIS063/064VP is adding a bypass line with double-directional check valve,and drilling a hole on the upstream disk of PTR022VB is adopted.

Design of Secondary Cooling Water System for China Advanced Research Reactor

RONG Feng, WANG Jian-yong

2006, 27(4): 68-70,74.

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Abstract:
The function of the secondary cooling water system of China Advanced Research Reactor is to transmit the heat of systems,such as coolant system,to the final heat sink.The article introduces the func- tion of the secondary cooling water system,the operating conditions,the composition and process of the sys- tem.It also analyzes the parameters for system design,the water treatment for secondary cooling water,the control and monitoring over the system.The design of this system is proper,and it accords with correspond- ing nuclear safety criterion and standard requirements.

Study on Virtual Simulation Technology for Operation and Control of PWR

FANG Bao-guo, ZHANG Da-fa, LIN Ya-jun

2006, 27(4): 71-74.

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Abstract:
The way to build graphical models of PWR with MultiGen Creator is discussed,and the three-dimensional model used in the virtual simulation is built.The mathematical simulation model for PWR based on the platform of MFC and Vega is built through the analysis of the mathematical simulation of PWR. The way to perform the virtual effect is introduced associating with the Pressurizer.And,all above parts are connected in one with VC++ to perform the whole virtual simulation of PWR.

Computer Simulation for Sodium-Concrete Reactions

ZHANG Bin, ZHU Ji-zhou

2006, 27(4): 75-82,89.

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Abstract:
In the liquid metal cooled fast breeder reactors （LMFBRs）,direct contacts between sodium and concrete is unavoidable.Due to sodium’s high chemical reactivity,sodium would react with concrete violently. Lots of hydrogen gas and heat would be released then.This would harm the integrality of the containment. This paper developed a program to simulate sodium-concrete reactions across-the-board.It could give the re- action zone temperature,pool temperature,penetration depth,penetration rate,hydrogen flux and reaction heat and so on.Concrete was considered to be composed of silica and water only in this paper.The variable, the quotient of sodium hydroxide,was introduced in the continuity equation to simulate the chemical reactions more realistically.The product of the net gas flux and boundary depth was ably transformed to that of pene- tration rate and boundary depth.The complex chemical kinetics equations was simplified under some hy- pothesises.All the technique applied above simplified the computer simulation consumedly,In other words, they made the computer simulation feasible. Theoretics models that applied in the program and the calculation procedure were expatiated in detail. Good agreements of an overall transient behavior were obtained in a series of sodium-concrete reaction ex- periment analysis.The comparison between the analytical and experimental results showed the program pre- sented in this paper was creditable and reasonable for simulating the sodium-concrete reactions.This pro- gram could be used for nuclear safety judgement.

Establishment of Professional Nuclear Power Architectural Engineering Company

GUO Dong-li, CHEN Hua

2006, 27(4): 83-85.

Abstract(9) PDF(0)

Abstract:
The rapid development of nuclear powe industry in China requires specialized management for the nuclear power engineering projects.It is necessary to establish the nuclear power architecturalengi- neering company to meet the increasing market needs by providing the owner with specialized nuclear engi- neering project management and overall contracting services.It is imperative that the purpose of establishing the corporation and enterprise core competitiveness should be clearly identified when it is established.Its organizational structure should be geared to the enterprise operation management and development to facili- tate the intensified project management and control,and improve its risk-proof ability.

Procurement and Quality Control of Components Important to Safety in Nuclear Engineering Projects

ZHANG Zhi-hua, ZHANG Yi-yun, XU Xian-qi, QIAN Da-zhi, DENG Yue

2006, 27(4): 86-89.

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The procurement and quality control of components is a very important work in the nuclear engineering.This paper introduces the project management techniques,such as how to make a plan of com- ponents purchase in nuclear engineering.This paper discussed the classification of components,evaluation of the potential suppliers,invitation of bids,exchange of design details with the suppliers,quality assurance and quality assurance audit,and the equipment checks before acceptance and some engineering experiences.

Deposition of Draphite Dust in Heat Gas Duct in HTR-10

LUO Xiao-wei, YU Su-yuan, TANG Hui

2006, 27(4): 90-92,96.

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Abstract:
The deposition of graphite dust in heat gas duct of 10MW high temperature gas-cooled reactor （HTR-10） was analyzed.The deposition efficiency of graphite dust in heat gas duct was obtained,which was necessary to estimate the radioactivity activity at the surface of heat gas duct.In the analysis,the turbulent deposition and thermophoretic deposition of graphite dust were involved.The calculation results revealed that the deposition amount of graphite dust in heat gas duct was small due to the large flow velocity of helium in heat gas duct.

Design of Data Acquisition System in Nuclear Power Station

XU Hui, HUANG Wen-jun, JIANG Zhu-xuan, SU Guo-quan

2006, 27(4): 93-96.

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The performance of Instrumentation and Control System is critical for nuclear power plants. Now the Instrument and Control System of domestic nuclear power plants mostly uses the analog control,in which the precision of data displaying and automation are poor.This paper introduces a data acquisition sys- tem applying in Qinshan unitⅢreactor shutdown system,which incorporates the technology of DCS to realize the function of quick data acquisition,save,alarm,query,displaying and improve the safety and econ- omy of nuclear power plants.

Hydrodynamic Model of Slug Flow in Vertical Pipes Considering Variable Thickness of Liquid Film

WANG Wei-yang, CHEN Ting-kuan, LUO Yu-shan, GAO Feng

2006, 27(4): 97-100.

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Abstract:
Based on analysis of flow mechanism,a hydrodynamic model of slug flow in vertical pipes was developed considering the variable thickness of liquid film.Effects of film thickness on results were dis- cussed and the model was evaluated with experimental data available from the literature,It was found that if the variation of thickness was neglected,and Taylor bubble was considered as a cylindrical bubble,the length of Taylor bubble would be under predicted.The simplification can also result in an over prediction of pressure gradient,and the deviation will increase as gas superficial velocity increases for a fixed liquid superficial ve- locity.The improved model takes into account the flow characteristic of liquid film,and can predict the data obtained from different sources with a fairly well accuracy.

2006 Vol. 27, No. 4