核动力工程

Calculation and Application of Discontinuity Factors in Control Rod Region for High Temperature Gas-Cooled Reactor

ZHOU Xu-hua, LI Fu, WANG Deng-ying, YAN Jian-qiu, LV Wei-feng, HAN Ren-yu

2008, 29(6): 1-5,9.

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For the strong absorber and void in the control rod region located in the reflector area of the High Temperature Gas-Cooled Reactor（HTGR）,this paper presents the method of diffusion equation cor-rected by discontinuity factor based on the homogenization on the local area.Facing the difficulties such as no fission source in the homogenous area and strong neutron current in the boundary,a special skill is required for the homogenization and calculation of discontinuity factor.Taking the fine mesh transport solution as the reference,the verification results demonstrates that the control rod region can be homogenized individually with approximate boundary condition,and the diffusion theory corrected by discontinuity factors can accu-rately model the control rod.Hence,applying the diffusion calculation with discontinuity factor correction to calculate the control rod in HTGR is excellent not only in accuracy but also in consumed calculation time.

Calculation of Fission Rate in CFBR-Ⅱ Reactor

DU Jin-feng

2008, 29(6): 6-9.

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In order to study the fission rate distribution in CFBR-II reactor,the calculation method of fis-sion rate based on MCNP is built and checking computations are conducted using the high enriched bare ura-nium sphere in foreign countries.The fission rate distribution in the top and bottom hemispheres in CFBR-Ⅱ reactor is calculated along virtual path setting at the direction of 45o.The results indicate that the maximum fission rate lies in neither the system center nor the hemisphere center but some place in the enriched uranium core.The total fission rate of center normalization in the reactor is calculated to be 3 150.3 by distribution integral method,in which the contribution of deficient uranium reflector is 5%.

A General Formula Considering one Group Delayed Neutron under Nonequilibrium Condition

LI Hao-feng, CHEN Wen-zhen, ZHU Qian, LUO Lei

2008, 29(6): 10-13,29.

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A general neutron breeder formula is developed when the reactor does not reach the steady state and the reactivity changes in phase.This formula can be used to calculate the results of six groups de-layed neutron model through a way of amending λ in one group delayed neutron model.The analysis shows that the solution of amended single group delayed neutron model is approximately equal to that of six-group delayed neutron model,and the amended model meets the engineering accuracy.

Measurement of Prompt Neutron Decay Constant on Uranium-Plutonium Metal Assembly

LIU Xiao-bo, FAN Xiao-qiang, YANG Cheng-de, DENG Men-cai

2008, 29(6): 14-16.

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The prompt neutron decay constant α of the Uranium-Plutonium metal assembly was meas-ured with Rossi-α technique at delay critical and sub-critical conditions,and then α_c is obtained.The relation-ship between α with reactivity and α is analyzed with detector counts reciprocal.The comparison shows that the direct measured result under delay critical condition is 0.835±0.005/μs,which error is 1/6 of that from Rossi-a and extrapolation methods.

Simulation Calculation of Response to Linear Positive Reactivity Inserted in Reactor with Temperature Feedback

SHANG Xue-li, ZHANG Fan, CHEN Wen-zhen, ZHAO Lei

2008, 29(6): 17-20.

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In this paper,the model is founded with point-reactor neutron kinetics equation,and main parameters of a small reactor are calculated under six typical operation conditions,when a linear positive reactivity is inserted.The results are compared with those obtained by 3-D real-time simulation software of reactors.It is shown that point-reactor model can describe the maximum changes and steady values of main parameters in the reactor when a linear positive reactivity is inserted,although improvements are also needed in calculating the response time and duration of the fluctuation.

Active Nucleation Site Density during Subcooled Flow Boiling at Vertical Narrow Rectangular Channel

PAN Liang-ming, ZHANG Jun-qi, YUAN De-wen, CHEN De-qi, WANG Xiao-jun

2008, 29(6): 21-24.

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At present work,visualized subcooled flow boiling experiments were conducted using a flat stainless steel surface heater.Results indicate that the active nucleation site density of narrow channel is with the same tendency as the regular channels,which is proportioned to the heat flux.Mass flux is with little ef-fect on the active nucleation site density of channel,but with effect on the onset of nucleate boiling.With the same heat flux,the narrower the channel,the highter the active nucleation site density.The effect of channel width on the active nucleation site density increases with the increasing of the heat flux.

Analysis of Pipeline Virbation Induced by Water Hammer

LI Song, MA Jian-zhong, GAO Li-xia, HU Yong-tao

2008, 29(6): 25-29.

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Water hammer often happens in liquid-pipelines,and impulsive forces of water hammer often damage pipelines and affect the safety of pipeline systems.In this paper,one spring-pipeline is studied.The ANSYS/LS-DYNA,a finite-element codes,is used to simulate the response of pipelines under impulsive force.The response of pipes within the air and the water are compared.The results show that the velocity and displacement are quickly declined within the water.Though importing the same water hammer force in inlet and outlet,the results show that the system responsed from the flexible pipe are stronger and the vibration time is also longer.The results are the same as the previous experimental ones.

Effect of Inlet Parameters on Maximum Discharge Pressure of Boost Installation for Steam-Water Two-Phase Flow

LI Gang, YUAN Yi-chao, LIU Yu-zheng

2008, 29(6): 30-34.

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Taking into account the errors of discharge pressure between experimental and computational values are off limits by the empirical models,one-dimensional theory model of steam-water two-phase lifting pressure facility is constructed and solved by the results based on direct contact condensation,and some key issues,such as inter-phase mass transfer and phase volume fraction,are discussed in detail.In order to vali-date the model,the maximum discharge pressures are computed and the results are relatively precise to the corresponding experimental data.The maximum discharge pressures at various inlet parameters are also con-cerned,and the results show the maximum discharge pressure is descended by improving the water tempera-ture,but improved by improving the inlet water mass flow or steam pressure.

Numerical Analysis of Crosswind Effect on Wet Cooling Tower Ae rodynamic Field

ZHAO Yuan-bin, SUN Feng-zhong, WANG Kai, GAO Ming, YAN Sheng-ping, GAO Tao

2008, 29(6): 35-40.

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Based on CFD code FLUENT,three-dimensional numerical analyses were carried out for natural draft wet cooling tower under crosswind conditions.Sensitivity analyses to parameters such as ambi-ent crosswind velocity profile and water droplet equivalent diameter validated the adopted numerical model.The effect of crosswind on wet cooling tower inner and outer aerodynamic field and tower internal heat and mass transfer performance were investigated numerically.The results show that crosswind causes the in-crease of air inflow relative departure degree and induces horizontal air mass flow rate which improves rain zone heat and mass transfer but reduces tower vertical air mass flow rate,and then produces an unfavorable effect on fill zone and increases outflow water temperature.The analyses about air inflow relative departure degree show that improving the air inflow aerodynamic field can reduce the unfavorable effect of crosswind on the circumference distribution of air inlet air radial velocity and then improve the total cooling perform-ance of wet cooling tower under crosswind conditons.

Experimental Investigation of Conevection Heat Transfer of CO₂ at Supercritical Pressures in a Vertical Circular Tube at High Re

LI Zhi-hui, JIANG Pei-xue

2008, 29(6): 41-45.

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Convection heat transfer during the upward flow of CO₂ at supercritical pressures in a vertical circular tube（d_in=2 mm） at high Reynolds numbers was investigated experimentally,and the effects of heat fluxes,mass fluxes,inlet temperatures,pressures,buoyancy and thermal acceleration on the convection heat transfer was analyzed.The results show that the tube wall temperature occurs abnormally distribution for high heat-fluxes with upward flow.The degree of deteriorated heat transfer increases with increasing heat flux.Increasing of the mass flux delays the occurrence of the deterioration of heat transfer and weakens the deterioration of heat transfer down-stream section.The inlet temperature strongly influences the heat transfer.The deterioration degree of heat transfer decreases with increasing pressure.

Study on Axial Surface Cracks inside Optimal Autofrettaged Thick Cylinder

ZHENG Xiao-tao, YU Jiu-yang, LIU Yu-hua, LU Xia, WANG Wei

2008, 29(6): 46-49,57.

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This paper analyzed the optimum autofrettage pressure of a thick wall cylinder under the in-ternal pressure and mode I stress intensity factor（SIF） for axial semi-elliptical cracks on the surface of the autofrettaged thick wall cylinder based on ANSYS.The SIF along the crack front are directly computed by 3D finite element method for a wide range of variations of the crack geometry.Also cracks on both surfaces of the autofrettaged thick wall cylinder were considered.The results show that the optimum autofrettage pressure of cylinder changes with the internal pressure,the SIF of the axial crack on the inner suface decrease after autofrettage,and the differences decrease while the crack depth increases,the axial crack outside the cylinder causes a decrease of about 1% of the SIF of the crack inside.

Signal Process and Profile Reconstruction of Stress Corrosion Crack by Eddy Current Test

ZHANG Si-quan, CHEN Tie-qun, LIU Gui-xiong

2008, 29(6): 50-53,65.

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The reconstruction of crack profiles is very important in the NDE（nondestructive evaluation） of critical structures,such as pressure vessel and tubes in heat exchanges.First a wavelet transform signal processing technique is used to reduce noise and other nondefect signals from the signals of crack,and then based on an artificial neural network method,the crack profiles are reconstructed.Although the results reveal that this method is with many advantages such as a short CPU time and precision for reconstruction,it does have some drawbacks,for example,the database generation and network training is a much time consuming work.Moreover,this approach does not expressly reconstruct the distribution of conductivity inside a crack,so the reliability of a reconstructed crack shape is unknown.But in practical application,if we do not consider the multiple cracks,this method can be used to reconstruct the natural crack.

Numerical Simulation of Surface Crack Growth on Overlap of Hot-wall Hydrogenation Reactor

LIU Bin, SHEN Shi-ming

2008, 29(6): 54-57.

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The principle of the three dimensional fracture analysis software FRANCE3D（Fracture Analysis Code in 3Dimensions） on fatigue crack growth is introduced,and the actual surface crack growth on overlay of the hot-wall hydrogenation reactor was analyzed by FRANCE3D.The results of numerical simula-tion were compared with that of the experiment data.The results show that the surface crack growth path ob-tained by numerical simulation is same to that by experiment basically;and compared with the result of ex-periment,the crack growth rate got by FRANC3D is more conservative than the experimental data in the ser-vice-life of hydrogenation reactor.

Study on Pull-out Force in Tube-and-Shell Heat Exchangers with Finite Element Method

WANG Jian-ping, JIN Wei-ya, WANG Xiu-min, GAO Zeng-liang

2008, 29(6): 58-61.

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Using the finite element method to shell-and-tube multiple fixed-tube plate heat exchanger in three kinds of cases,the pull-out force for a heat exchanger and tube-sheets under the different temperature distribution are calculated.The results show that the pull-out force have a bigger difference in the different locations of a heat exchanger tube for multiple exchangers.According to the allowable value of ASME,the pull-out force of all heat exchanger tubes and tube-sheet joint for case 3 can meet the requirements,but the pull-out force of the joints with 56 heat exchanger tubes and tube-sheets for case 1 and with 22 heat ex-changer tubes and tube-sheets for case 2 are over the allowable values.Therefore,under the co-action of vi-bration,corrosion and so on caused by the flow of the medium,it is easy for these joints to rupture and leak.

Calculation of Displacement Damage Based on Average Displacement Kerma Factor Method

ZOU De-hui, QIU Dong

2008, 29(6): 62-65.

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In view of the present weak points of the conventional average energy kerma factor method,the average displacement kerma factor method calculating the displacement damage in the different energy grouping structure is presented based on Si displacement kerma factor function table.Displacement damage of several radiation sources are calculated with this method.The results show that the discrepancy of dis-placement damage using the average energy method is about 17%,and it is about 4% using the average dis-placement kerma factor method.The results also show that the displacement damage for the different neutron distributions in same energy group do not significantly change with the average displacement kerma factor method.

Design of Scaler Protector for Reactor

LU Yi, LI Meng, RONG Ru

2008, 29(6): 66-69.

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In order to satisfy the special requirements of reactor physical experiment,the design method for a new scaler protector is presented in this paper.For this method,Field Programmable Gate Array（FPGA） is adopted in structure design,and combined with the techniques of signal disposal,driving output and dy-namical display,the functions for counting and protection are realized,while the communication interface with PC is expanded,and thus a more intelligent scaler protector is achieved.

Temperature Correction for Stress Measurement in Fast Burst Reactor

QIU Dong, YIN Yan-peng

2008, 29(6): 70-73,90.

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This paper briefly describes the originating mechanism for stress in fast burst reactor under pulsed operation.The formula,which transforms strain into stress,has been deduced based on a spherical shell model and according to this,the model of temperature correction for measured strain is put forward.The results of validation experiments on CFBR-II indicate that the requirements on instantaneous tempera-ture-rise measurement and synchronization can be avoided effectively by introducing the temperature correc-tion factor,and that the results of stress from measurement and calculation agrees well within about 25% er-ror range.

Design of Period Measurement Instrument for Fast Neutron Criticality Facility Based on Virtual Instrument Technology

LI Meng, HU Qian, HU Jin-quan

2008, 29(6): 74-78.

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Based on the data acquisition curve fitting methold,the period measurement instrument for fast critical assembly has been designed with virtual instrument technology.The configuration and design of the instrument is introduced in this paper.Prelimilary experiment results prove that the period measurement instrument has improved the uncertainty,interference immunity,precision and consistency compared with the traditional window-mode period measurement instrument,and can satisfy the expected reqirements.

Design of Multi-Channel Amplitude Analyzer Based on LonWorks

ZHANG Ying, ZHAO Li-hong, CHEN Ai-hua

2008, 29(6): 79-82,105.

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The paper introduces the multi-channel analyzer which adopts LonWorks technology.The sys-tem detects the pulse peak by hardware circuits and controls data acquisition and network communication by Micro Controller and Unitand Neuron chip.SCM is programmed by Keil C51;the communication between SCM and nerve cell is reallized by Neron C language,and the computer program is written by VB language.Test results show that this analyzer is with fast conversion speed and low power consumption.

Radiation Embrittlement Surveillance of ReactorPressure Vessel

XIAO Bing-shan, ZHANG Le-fu

2008, 29(6): 83-86.

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This paper describes the vessel radiation surveillance program for Qinshan second nuclear power plant 600MWe pressurized water reactor.Surveillance program,test,evaluation method and origins for larger lead factor were discussed in detail.The effect of radiation embrittle for reactor pressure vessel for Qinshan second nuclear power plant was evaluated on the basis of the results of two surveillance capsules.

One Method for Assessing Internal Dose Due to Tritium

MAO Yong, WANG Xiao-dong

2008, 29(6): 87-90.

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In order to assess the internal exposure dose for radiation protection,reference biokinetic models and its parameters for radionuclide are recommended by International Commission on Radiological Protection（ICRP）.As an example in this paper,an individual biokinetic model for a tritiated water intake in an accident are developed based on the monitoring data of urine samples.The results show 25.5mSv of committed effective dose by developed individual biokinetic model,and 38.6mSv by reference model rec-ommended by ICRP.The deviations with reference value given by IAEA are 1% and 48%,respectively.

Research of Pneumatic Control Transmission System for Small Irradiation Samples

BAI Zhong-xiong, ZHANG Hai-bin, RONG Ru, ZHANG Tao

2008, 29(6): 91-94.

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In order to reduce the absorbed dose damage for the operator,pneumatic control has been adopted to realize the rapid transmission of small irradiation samples.On/off of pneumatic circuit and direc-tions for the rapid transmission system are controlled by the electrical control part.The main program initial-izes the system and detectes the location of the manual / automatic change-over switch,and call for the corre-sponding subprogram to achieve the automatic or manual operation.Automatic subprogram achieves the automatic sample transmission;Manual subprogram complets the deflation,and back and forth movement of the radiation samples.This paper introduces in detail the implementation of the system,in terms of both hard-ware and software design.

Development of Augmentation Mechanism for Large Reactivity Step in CFBR-II Reactor

YE Cen-ming, ZHANG Yi, RONG Ru, HU Qian

2008, 29(6): 95-97,101.

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Linear motor,as the implementation mechanism for augmentation of large reactivity step in the CFBR-II reactor,is used to step increase the reactivity during the pulse burst.The whole system includes linear motor,Driving amplifier,Siemens PLC,ACE shock Absorber,and etc.The augmentation mechanism driven by linear motor can effectively overcome the disadvantages of pneumatic drive,and is featured by small volume,light weight,high speed movement,and high driving accuracy,and at the meantime,the linear motor can operate according to the preset rate curve.

Preliminary Design of Sub-Critical Verification Facility for Accelerator Driven System with Auxiliary Shut-Down System

YU Tao, LI Xiao-hua, ZHOU Cheng-long, XIE Qin, MA Zhi-yuan, LIU Ping

2008, 29(6): 98-101.

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Considering the potential safety hazards in the current concept design of accelerator driven nuclear system,sub-critical verification facility with auxiliary shut-down system is designed in this paper by using MCNP code.The core parameters can be adjusted by changing thermal zone pitch;the thickness of thermal area and the inserting depth of control rod,and simulation calculation and comparative analysis of multiplication factor is carried out,and k_eff can be modulated under various parameters.Finally,the shut-down margin and control rode effectiveness for different thermal pitches and fuel numbers are calculated,and the calculation results show that auxiliary shut-down system improves the safety of the accelerator driven system.

Design of Regulator for n/γ in Complicated Radiation Field

LI Jun-jie, QIU Dong, DU Jin-feng, WANG Qiang

2008, 29(6): 102-105.

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In oder to satisfy the spcial demand of some users on the radiation field,it is necessaary to further increase the n/γvalue（neutron-fluence-to-gamma-ray dose ratio） in CFBR-II.The regulator with Pb as the main material is designed,considering its well shielding capability to gamma-ray.The n/γvalues with different lead thickness have been given from theory calculation.Considering the effect of regulator on the neutron spectrum and the load capacity of the installation platform,the regulator consisted of 4 layers（1cm for each layer） is worked out,and the experimental measurements have been carried out.The results indicate that the experimental results accord with the calculational one within allowable error range.The n/γvalue in CFBR-II radiation field can be regulated within 1.07×10¹²cm^-2.Gy^-1 to 4.87×10¹²cm^-2.Gy^-1.

First Ten-Yearly Modification Study for Daya Bay Nuclear Power Plant

LU Xiu-sheng

2008, 29(6): 106-109,124.

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This paper analyzed the features for various operation phases for nuclear power plant,and provides the period and goals for long-term modifications and the necessity for 10 yearly modifications.The introduction of classification,contents of main items and implementation results of the first ten-yearly modi-fications for Daya Bay Nuclear Power Plant indicates the importance of periodical modifications to the im-provement of safety,economy and reliability of Daya Bay Nuclear Power Plant.

Life Management of Reactor Pressure Vessel at Qinshan Nuclear Power Plant

KONG De-ping, LI Hua, ZHENG Hong-lian

2008, 29(6): 110-114.

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This paper introduces the Ageing and Life management of Reactor Pressure Vessel（RPV） at Qinshan Nuclear Power Plant（QNPP）.The analysis of requirements in the codes,criterion and guides applica-ble for Ageing and Life management of RPV at QNPP illustrates that the methods adopted and works carried out in the ageing and life management of RPV at QNPP is rational and practical.

Aging Management of Steam Generator in Qinshan Nuclear Power Plant

TAO Jun, WEI Wen-bin, LI Shi-wei

2008, 29(6): 115-118.

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According to the principle of PDCA cycle in SG aging management,this paper explains the establishment of aging management system,the operation and control（mainly water chemistry control）,and the check,inspection,evaluation and maintenance of steam generator（SG） in Qinshan Nuclear Power Plant（QNPP）.With the implementation of these measures,the aging and degradation of SG in QNPP are managed effectively.The SG of QNPP are still in good condition after 16-years of operation.

Stress Relaxation Performance and Prediction Models for Bolt Material of 1Cr10NiMoW2VNbN

GUO Jin-quan, XUAN Fu-zhen, HE Lei

2008, 29(6): 119-124.

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Stress relaxation tests of the newly developed bolt alloy 1Cr10NiMoW2VNbN were carried out in order to investigate and predict its long-term relaxation performance.Effects of the environment tem-perature and primary stress on the relaxation performance were discussed.Based on the experimental results,the applicability of Hook-Norton model,generalized Maxwell relaxation model,logistic model and polyno-mial model for the extrapolation of residual stress was reexamined.Results indicate that temperature is the dominating contributor for the relaxation performance of bolt material.According to the experimental results in this paper,stress relaxation limit of the newly developed bolt alloy 1Cr10NiMoW2VNbN is still unavail-able.Compared to the experimental data,the developed polynomial relaxation model can lead to an appro-priate extrapolation of the remaining stress.

Loose Part Monitoring System for Nuclear Power Plants Based on Virtual Instrument

PAN Shi-biao, LIU Cai-xue, HE Shao-qun, HU Jian-rong, LI Xiang, ZHENG Wu-yuan, WANG Cheng-yuan

2008, 29(6): 125-127.

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Taking LabVIEW as the development platform,online real-time monitoring for loose parts of reactor pressure vessel and steam generator is carried out,by integrating the technologies of virtual instrument,the equipment configuration graphing and the database management.Experiment validation was carried out on the test device for loose part simulation of nuclear reactor.The basic functions for loose parts monitoring is realized.

Design and Development of Virtual TXP Control System Software

WANG Yun-wei, LENG Shan, LIU Zhi-sheng, WANG Qiang, SHANG Yan-xia

2008, 29(6): 128-131.

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Taking distributed control system（DCS） of Siemens TELEPERM_XP（TXP） as the simulation object,Virtual TXP（VTXP） control system based on Virtual DCS with high fidelity and reliability was designed and developed on the platform of Windows.In the process of development,the method of object-oriented modeling and modularization program design are adopted,C++ language and technologies such as multithreading,ActiveX control,Socket network communication are used,to realize the wide range dynamic simulation and recreate the functions of the hardware and software of real TXP.This paper puts emphasis on the design and realization of Control server and Communication server.The development of Virtual TXP control system software is with great effect on the construction of simulation system and the design,commission,verification and maintenance of control system in large-scale power plants,nuclear power plants and combined cycle power plants.

Nuclear Power Engineering 2008 Index

2008, 29(6): 132-144.

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2008 Vol. 29, No. 6