核动力工程

2010, 31(6): .

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Analysis and Control of Welding Deformation in Qinshan Nuclear Power Phase Ⅱ Extension Project

WANG Qing-tian, XU Bin, HE Da-ming, LI Yan

2010, 31(6): 1-4,9.

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The paper analyzes the severe deformation in the welding of core barrel in Qinshan Nuclear Power Phase II Extension Project Reactor 3_# unit,which nearly induces the loss of the function of the core barrel.Measures such as improving the welding fixture,process and parameter,and loading the counterweight is taken for the 4_# unit to minimize the deformation,and the result shows that the weld of the 4_# core barrel satisfies the design requirements.

Static Status and Treatment Solution Study on Sub-Grade with Non-Homogeneous Constituents in Local Area of Nuclear Island Buildings

LI Zhong-cheng, ZHANG Yang, DONG Zhan-fa, WANG Yu-xia, LI Xiang-ye, FAN Hong, DAI Wei-ying

2010, 31(6): 5-9.

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Since there are some weathered accidental inclusion in the sub-grade of a nuclear project in China,an integrative model including the sub-grade,foundation and super-structure were generated using the finite element software ANSYS to assess the influence of these weak rocks.Based on the geological data,the non-homogeneous rock conditions were accurately simulated and the sensitivity analyses to the parameters of weathered accidental inclusion were also carried out.Then the non-homogeneous settlement of the sub-grade rock,the internal force distribution of the raft and super-structure and the structural reinforced design were analyzed in detail,and the results provided a theoretical basis for obtaining the treatment solution of weak rocks.It demonstrated that the weak rocks in the specified limited scale have slight impact on the static capacity of the sub-grade since the surrounding good rocks can support the loads transmitted from the super-structure.The engineering practice also showed the relevant treatment solution is appropriate.

Seismic Analysis for SEC Drum-Shaped Filter of 1000 MW Nuclear Power Plant

FU Qiang, YUAN Shou-qi, ZHU Rong-sheng, OU Ming-xiong, YANG Ai-ling

2010, 31(6): 10-14,28.

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SEC drum-shaped filter structure in 1000 MW nuclear power plant was seismic analyzed by the seismic analysis method of response spectrum and the natural modes of each structure order combined using SRSS and the curves,and taking the amplify coefficient of several seismic waves as the input seismic load.The calculation results show that the maximal response for drum-shaped filter stress was 69.75 MPa at the connection between the main shaft and the A main spoke;the maximal response for displacement was 12.50 mm at the end of the main cross-beam.The strength meets the seismic requirement when checked by the third strength theory,but the limit of side sealing gap was 12 mm that did not meet the seismic requirement when the response value was bigger than the limit.To improve Seismic level of the drum-shaped,the side sealing gap should be increased.

New Method of Individual Modal Response Combination in Nuclear Power Plants

LIU Sheng-hua, XIA Zu-feng

2010, 31(6): 15-19.

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This paper introduced the recommended combination methods of individual modal response for nuclear power plant systems,structures and components（SSCs） defined in RG1.92 Rev.2.Also,this paper developed a new APDL macro of ANSYS based on the recommended combination method,and performed a piping example with this macro and compared the results with those obtained from CQC and DSUM methods.The results indicated that the macro this paper developed according to RG1.92 Rev.2 was reasonable.

Study on Model of Continuous Rotor under Steam Seal Force

YANG Yi, LI Yu, CHEN Jian-hong, SHENG De-ren

2010, 31(6): 20-23.

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In order to study the impact of steam seal force,due to the steam turbine seals clearance,on the rotor vibration and stability,a complete structural model of single-disc and single-span continuous rotor system with double bearings and shaft seals is set up.Besides the nonlinear factors such as bending,shearing effect and shaft asymmetry,this paper takes the combined influence of both steam seal force,described by Musznyska model,and short bearing unsteady oil-film force into consideration.A nonlinear dynamic model of rotor system is built based on the Hamilton principle and is solved.

Application of Containment Leakage Test Code ANSI/ANS56.8-1994 in Tianwan Nuclear Power Station

OUYANG Qin, CHU Ying-jie

2010, 31(6): 24-28.

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For measuring the containment integrated leakage rate effectively,a series of calculation program was developed according to American national standard for containment system leakage test requirements ANSI/ANS-56.8-1994 by Tianwan Nuclear Power Station（TNPS）,and this computing program was applied well in the containment system leakage tests for Unit 1 and Unit 2 of TNPS.The detailed calculation process is introduced by this paper to give an example in other containments integrated leakage tests.

Application of Internal Model Adaptive Control In Water System of Nuclear Power Plant Steam Generator

MI Ke-song, GU Jun-jie, XU Pei-pei

2010, 31(6): 29-32.

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The internal model control scheme is proposed for the control of the water level of U-type steam generator.The parameters for the internal model controller are obtained by math model.The internal model control system for U-type steam generator is established by simulation tools.The result shows that the control efficiency of this scheme is better than that of the PID control,which can simplify the parameters for the controller,and is convenient for real-time control and improve the robustness.

Comparison and Analysis of Methods for Containment Leakage Rate Calculation

CHU Ying-jie, OUYANG Qin

2010, 31(6): 33-37.

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Three international commonly used methods for containment leakage rate calculation are introduced in detail,and are applied to calculate the leakage rate based on the actual containment integrated leakage tests in Tian’wan nuclear power station.The principle difference among these methods and their effects on the results are analyzed,which shows that the calculated leakage rates are almost the same.

Design of Modular Power Source for Ultrasonic Control Rods Position Measuring System of 5 MW Nuclear Heating Reactor

LI Zhi-hui, ZHU Jiang, SU Sheng-min, ZHONG Wei

2010, 31(6): 38-40,46.

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The power supply for the control rods position measuring system of 5 MW nuclear heating reactor became unstable because of the circuit ageing.A new modular design for the power system has been proposed.The new system is composed of the standard switch power supply modules and three terminal regulators.While this new design can power the control rods position measuring system as effective as the former separate element design,its stability and maintainability are enhanced.

Interactive Analysis of Human Error Factors in NPP Operation Events

ZHANG Li, ZOU Yan-hua, HUANG Wei-gang

2010, 31(6): 41-46.

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Interactive of human error factors in NPP operation events were introduced,and 645 WANO operation event reports from 1999 to 2008 were analyzed,among which 432 were found relative to human errors.After classifying these errors with the Root Causes or Causal Factors,and then applying SPSS for correlation analysis,we concluded:（1） Personnel work practices are restricted by many factors.Forming a good personnel work practices is a systematic work which need supports in many aspects.（2）Verbal communications,personnel work practices,man-machine interface and written procedures and documents play great roles.They are four interactional factors which often come in bundle.If some improvements need to be made on one of them,synchronous measures are also necessary for the others.（3） Management direction and decision process,which are related to management,have a significant interaction with personnel factors.

Study on Heat Transfer Behavior of Molten Pool in Lower Head Introduced by SBO Severe Accident

ZHOU Wei-hua, YANG Yan-hua, FU Xiao-liang, YANG Xiao

2010, 31(6): 47-51.

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In this paper,the severe accident induced by Station Blackout（SBO） of CPR1000 is simulated by SCDAP/RELAP5 code.The progression from the initiation of core uncovering and the full uncovering of the core to the core melt relocation into the lower head are analyzed.The heat transfer behavior of molten pool in the lower head is also discussed in this paper.The results achieved indicate that the damage location is at the 30°~40° side surface of the lower head（taking the bottom of PRV as 0°）.

Development of MCATHAS System of Coupled Neutronics/ Thermal-Hydraulics in Supercritical Water Reactor

AN Ping, YAO Dong

2010, 31(6): 52-55,74.

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The MCATHAS system of coupled neutronics/thermal-hydraulics in the supercritical water reactor is described,which considers the interaction between the obvious axial evolution of material temperature and density and the power distribution.This code is coupled externally.The MCNP code with the library of continuous cross section is used for neutronics analysis.The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN code for burn-up analysis.The calculation results for the assembly of HPLWR show that the results from this code is reliable.

Research and Development of Real-Time Thermal-Hydraulic Simulation Code for Plate Type Fuel Reactors

ZHANG Zhi-jian, LI Lei, GUO Yun

2010, 31(6): 56-63.

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Based on a three-equation model,the real-time thermal-hydraulic simulation code for plate type fuel reactors introduced proper constitutive equations which calculate the heat transfer coefficient,friction pressure drop and other parameters of plate type fuel reactors’ rectangle coolant channels.A transient flow distribution for multi-channel model and a solution of core flow among coolant channels was established,looking up table method is used to solve water-steam corporality parameters.The Reactivity Insertion Accident（RIA） and Loss of Flow Accident（LOFA） which had been defined in the IAEA 10MW MTR benchmark program is analyzed by the code programmed and executed with the real-time simulation support system SimExec.The result shows that the code is precise and proper for real-time thermal-hydraulic simulation of plate type fuel reactors.

Comparison of CO₂ Power Cycle to He Cycle

DUAN Cheng-jie, WANG Jie, YANG Xiao-yong, DING Ming, CAO Jian-hua

2010, 31(6): 64-69.

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Brayton cycle is the key technology for the fourth-generation nuclear reactor power cycle.This paper analyzed the carbon dioxide power cycle,and compared it with the most widely studied helium Brayton cycle.The result demonstrates that the unit mass of CO₂ absorbing less heat than Helium,resulting in larger cycle mass flow rate than Helium cycle under the same power;however,higher density of CO₂ makes smaller cycle volume flow rate.Because of the special physical properties of carbon dioxide,the CO₂ Brayton cycle,which can reach a higher level of efficiency at a lower temperature（650℃） and reduce the volume of heat exchanger and turbo machinery,has a great potential.

Mathematical Analysis of Heat Transfer Characteristics of Laminar Flow in a Rectangular Channel in Rolling Motion

YAN Bing-huo, YU Lei, YANG Yan-hua

2010, 31(6): 70-74.

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The laminar flow characteristics in a rectangular channel in rolling motion is investigated theoretically.The correlations of velocity and temperature in rolling motion are derived.And the heat transfer correlation is also obtained.The Nusselt number is a function of the position,while the oscillating amplitude of transient Nusselt number increases with the increase of Prandtl number.There is an initial phase difference of 45 degree between the transient Nusselt number and the rolling motion.

Research on Invalidation of Density Lock by Oscillation in Small Channels

WANG Sheng-fei, YAN Chang-qi, GU Hai-feng, YU Pei

2010, 31(6): 75-79.

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Experiment study on small channels oscillation in density lock is carried out.The invalidation of density lock is due to the stratified interface of one small channel going beyond during the process of oscil-lation,and a circulation is formed among small channels.Three solutions are proposed through the theoretical analysis: variability of diameter channels,reduction of pressure among channels and disturbance of homog-enization.According to these solutions,the density lock structures are designed and verified through experi-ment.The result shows that the variability of diameter channels would effect only when oscillation passed;the way of reducing pressure among channels would increase the length of density lock and the disturbance of homogenization could reduce the oscillation effectively.

Visualized Experimental Observation on Flow Patterns in a Single-Side Heated Narrow Rectangular Channel

WANG Jun-feng, HUANG Yan-ping, WANG Yan-lin

2010, 31(6): 80-84,92.

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Visualized experimental observation on flow patterns during the flow boiling of water under single-side heated and fluid-inlet sub-cooled conditions in a vertical narrow rectangular channel with the section of 40×3mm² were carried out.Four discernible flow patterns which were dispersed bubbly,and the coalesced bubbly,churn flow and annular flow were observed.Flow visualization in two dimensions of two-phase flow patterns for narrow rectangular channel,which provided intuitionistic evidence to distinguish flow patterns,were performed.Based on the experiment results,a flow pattern map for the single-side heated narrow rectangular channel was developed and then compared with some existing flow pattern maps.The comparison result showed that the flow patterns and its transition mechanisms are obviously different between the heated steam-water flows and the adiabatic air-water flows.

Friction and Wear Behaviors of C/C-SiC under Water Lubrication

GAO Wen, TANG Rui, LONG Chong-sheng, WANG Ji-ping

2010, 31(6): 85-88.

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C/C-SiC composites,prepared by a modified thermal gradient chemical vapor infiltration（preparing C/C） combined with reactive melt infiltration（infiltrating Si）,were studied for the friction and wear behaviors in this paper.The results show that during block-on-ring tests under water lubrication,the friction coefficient and specific wear rate of C/C-SiC samples increase with the increasing of load,but decrease with the increasing of sliding velocity,and the sliding velocity is with greater effect on the friction and wear behaviors than the load does.The C/C-SiC samples mainly subject to the grain wear and fracture wear in the process of friction.

Analysis of Dispersion-Type Nuclear Fuel Using Eshelby’s Solution

WAN Yuan-fu, DING Shu-rong, HUO Yong-zhong

2010, 31(6): 89-92.

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Eshelby’s solution was used for the analysis of dispersion-type nuclear fuel,considering the thermal field and irradiation-induced swellings.The impact of particle shape and irradiation-induced swellings on stress field was discussed.The maximum of the first principal stress and Von Mises equivalent stress can be found on the interface between the matrix and the particle,whose value is related with the particle shape and other factors.And the irradiation-induced swellings enhance the first principal stress and Von Mises equivalent stress very strongly.

Extrapolaion Modification of Physical Start-up of Reactors

DAI Qian-jin, ZHANG Jiao, ZHAN Yong-jie, PAN Ze-fei, YE Guo-dong

2010, 31(6): 93-95,101.

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The result of extrapolation is very important to the physical start-up,and it is used to control the speed and quantity of the reactivity.Using the neutron multiplication formula for active and sub-critical condition,the non-balance situation of neutron counts,the dilution,and the lifting of banks are analyzed.The results show that the delayed effect of dilution,the non-linearity of the integral worth of the control rods,and the non-linear increase of the flux have great effects on the accuracy of the extrapolation.We consider those effects and give some advices on the modification of the extrapolation.

Forecast Dilution or Boration Value during Power Changing Process in Qinshan Project Ⅱ

SHI Wei-hua, PAN Ze-fei, YE Guo-dong

2010, 31(6): 96-101.

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During the power changing process in the nuclear power plant,in order to control the power distribution along the axial direction and to compensate the dynamic reactivity,the boron concentration has to be adjusted,besides the adjustment of the control rods.This paper established a dilution and boration model by using SIMULINK to forecast the speed and the amount of dilution or boronation during power changing process.It also compared the forecast value with the actual value in Qinshan phase II and maximum relative error is percent 5.2.

Research on Load-Following Characteristics of Pressurized Water Reactor Nuclear Power Plants

SHI Xi, WU Ping, ZHAO Jie, LIU Di-chen

2010, 31(6): 102-105,112.

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Pressurized Water Reactor（PWR） Nuclear Power Plants（NPPs） realize the load following through the Power Control System（PCS）.In this paper,the fuzzy controller was introduced into PCS to make the optimal rod control and boron concentration adjustment.A power compensation channel was proposed to accelerate the response speed.The simulation results in Matlab illustrated the excellent load-following capability of NPP with PCS.Through the user-defined program of PSASP,PCS module was embedded with the whole NPP model.The simulation results in PSASP showed that the load-following capability of NPP meets the industrial technical requirements,and demonstrated the capability of NPP for power system daily peak load regulation.

Preventative Maintenance Cycle of Contact Switches for Nuclear Power Plants Based on Lifetime Assessment and Economic Analysis

SHI Jie

2010, 31(6): 106-112.

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An approach to determine the preventive maintenance cycle was proposed in consideration of the lifetime,optimal cost and economy.Two parameters Weibull distribution was used to calculate the lifetime of contact switch.The block replacement model and age replacement model were built with the objective of optimal cost,and the preventive replacement cycle was accounted.Eight proposals for preventive replacement cycle were given.Economy model was applied to assess those proposals and the optimal proposal was confirmed.

Alarm Method for Loose Part of Nuclear Reactor Based on Fractal Theory

FANG Li-xian, ZENG Fu, CAO Yan-long, YANG Jiang-xin, WANG Chi-hu, XIE Yong-cheng, SHEN Xiao-yao

2010, 31(6): 113-116,122.

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To improve the alarm veracity rate for Loose Part Monitoring System（LPMS） of nuclear power stations,a new alarm method based on the nonlinear fractal theory is presented in this paper.Correlation dimensions for background noise,loosen part impact signal and loosen part impact signal under background noise are studied respectively.It is found that the change of correlation dimension reflects the situation of loosen part well during nuclear reactor operation.Correlation dimensions for steel balls with different mass have approximate values,and the difference between the correlation dimensions of steel ball impact signal and background noise is significant,even if the steel ball impact signal is under background noise.It has been proven by experiments that the impact signal can be detected efficiently even if signal-to-noise ratio is poor.

Evaluation of Emergency Planning Zone for EPR Nuclear Power Unit in TSNPP

LUO Hai-ying, WANG Jian-hua, LI Wen-hui, GUO Jing-ren

2010, 31(6): 117-122.

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The Emergency Planning Zone（EPZ） of EPR unit was evaluated using the MELCORE Accident Consequence Code System 2（MACCS2） developed by Sandia National Laboratory.Hourly meteorological data were collected from Sep 2007 to Aug 2008 by the site meteorological monitoring system.The accident source term including the core inventory and release characteristics were based on EPR PSA Level 2.Based on NUREG-0396 and GB/T 17680.1-2008,this paper put forward the quantitative safety criteria.The effective dose and thyroid dose were calculated.By comparing the result to the protective guide and safety criteria,a radius of 0.5 km was evaluated for EPR unit in TSNPP.We suggested that a radius of 4 km is the reasonable conservative value of interior EPZ for TSNPP and for exterior EPZ the value is 7km.

Discussion on Design Scheme of AP1000 ’s Radioactive Waste Treatment

LU Jun

2010, 31(6): 123-126.

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Based on the minimization of radioactive waste and the localization of the nuclear power plant,this paper discusses the proper mode of AP1000’s radioactive waste disposal,and gives a design scheme about the radioactive waste treatment of AP1000,which can control the volume of radioactive waste package of a single reactor below the 50m³/a,satisfying the principle of radioactive waste minimization.

Research on Application of New Type of Dissipater in Flood Control Design at Yang River Nuclear Power Station

XU Xi-rong, LIAO Ze-qiu

2010, 31(6): 127-130.

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Considering the design condition at Yang River nuclear power station,the flood control dissipater is studied by the model experimentation method.The result shows that,the flood control dissipation design within the site of Yang River nuclear power station,i.e.to use the bridle path at mountain slope excavation to get the multi-level dissipation of energy,and to set up the stilling baffle outside the bridle path and permeable dike inside the bridge path as the dissipater,could obtain the satisfactory result of dissipation of energy.

Index

2010, 31(6): 131-142.

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2010 Vol. 31, No. 6