核动力工程

Calculation and Discussion of Pressure-Temperature Limit Curves Based on RCCM Code

LU Feng, CHEN Mingya, WU Hong, HUANG Ping

2012, 33(6): 1-5.

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Abstract:
Comparing with RCCM code Ver.2000,some changes for pressure-temperature（P-T） limit curve calculation in the Ver.2007 are introduced,and two P-T limit curve calculation procedures are proposed based on the Ver.2007.In the numerical example,the P-T limit curves are given by the RCCM code Ver.2000 and Ver.2007,respectively.Furthermore,the effect of some parameters,such as the cooling rate and the cladding,on the P-T limit curves,is studied.

Assessment and Reducer of Pipe Vibration of Tianwan NPP

YUAN Shaobo, YU Danping, ZHOU Zhengping, XI Zhide, ZHAO Hui, HE Chao

2012, 33(6): 6-8.

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This paper presents a method of reducing pipe vibration of Tianwan NPP.Several modification designs have been tried for the pipeline system,and its vibration amplitude corresponding to each modification is briefly stated and compared.According to the requirements of the related codes,assessment on the vibration is carried out.It is showed that the best way of reducing vibration is to eliminate the source of excitation of the pipeline system.

Numerical Simulation Research on C-ring of CPR1000 Reactor Pressure Vessel

XIONG Guangming, DUAN Yuangang, DENG Xiaoyun, TAN Yu, JIN Ting, YANG Nengren

2012, 33(6): 9-12.

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The seal element of CPR1000 reactor pressure vessel is C-ring,in this paper,five finite element models are established,including practical model,ring model,equivalent cylinder model.Using linear elastic analysis and elastoplastic analysis,each model’s load-displacement curve of bearing radial load is obtained,which contains applying loading and unloading process.The result show that,equivalent cylinder model based on the medium-diameter can reflect the characteristics of C-ring.At the same time,it can effectively reduce the cost of calculation,and can be used for seal analysis of reactor pressure vessel bolt-flange connecting structure.Analysis also indicates that,the alloy cladding and silver have a great influence on the results.

Analysis on One underground Nuclear Waste Repository Rock Mass in USA

HA Qiuling, ZHANG Tiantian

2012, 33(6): 13-16,34.

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When analyzing the rock mass of a underground nuclear waste repository,the current studies are all based on the loading mechanical condition,and the unloading damage of rock mass is unconsidered.According to the different mechanical condition of actual engineering rock mass of loading and unloading,this paper implements a comprehensive analysis on the rock mass deformation of underground nuclear waste repository through the combination of present loading and unloading rock mass mechanics.It is found that the results of comprehensive analysis and actual measured data on the rock mass deformation of underground nuclear waste repository are basically the same,which provide supporting data for the underground nuclear waste repository.

Study on Effect of Water Hammer in Space Pipe with CFD

XI Zhide, MA Jianzhong, SUN Lei

2012, 33(6): 17-20.

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The typical water hammer theory is introduced at first,and the shortcoming of typical water hammer to calculate water hammer in the space pipe is analyzed,and then the CFD method to study the effect of water hammer in space pipe is applied.If the coupling effect of the structure and the fluid is not considered,the fluid govern equation in the space pipe can use the basic fluid control equation.The result can be obtained by CFD technology,when the govern equation and the fluid state equation and the initial condition,boundary condition is obtained.The numerical effect numerical dissipation and the frequency dissipation is avoid when the case include wave process.The CFD is applied to simulate the model in related reference,and discuss the effect of the discrete format and the grid.Finally the progress of the water hammer in the pipe is discussed.The method in this paper can be used in more complex pipe system to simulate the water hammer effect.

Study on CPR1000 Protection System Design

ZHENG Weizhi, LI Xiangjian

2012, 33(6): 21-26.

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In order to optimize and improve the protection system architecture of current CPR1000 power plant,protection system of Ling’ao II and Hongyanhe as representative of the power stations are studied,and both the advantages and disadvantages are analyzed.By combining the advantages of both,a new protection system architecture is designed to enhance the reliability of the design.This architecture has many advantages compared with Ling’ao II and Hongyanhe,such as the separation between operation of safety equipments and non-safety system,and the separation between automatic control station and manual control station of engineered safety feature actuation system,and the countermeasures against common cause failure of DCS.

Risk Analysis on Safety Injection Test of PWR Nuclear Power Plant

XU Yonghua

2012, 33(6): 27-29.

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During the safety injection（SI） test in the PWR nuclear power plant,full water in pressurizer and high-high level,low-low level in SG may take place.This paper analyzes the risks and response measures in the SI test.Focusing on the full scale problem of pressurizer thermal calibration level gauge in the SI test of nuclear power plant,the test process is analyzed and the problems that should be noted in the SI test are summed up.

Intercurrent Fault Diagnosis of Nuclear Power Plants Based on Hybrid Artificial Neural Network

PENG Qiao, GUO Lifeng, MA Jie

2012, 33(6): 30-34.

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Based on the analysis of the structure of ART-2 and parallel BP neural network,a hybrid artificial neural network is proposed aiming at the intercurrent faults diagnosis of nuclear power plants.Firstly the ART-2 net is used to identify the single fault,then the parallel BP net is used to distinguish intercurrent faults from new fault.The simulation shows that,the hybrid artificial neural network resolves the problem of single neural network in distinguishing intercurrent faults from new fault,and can diagnose the intercurrent fault and new fault efficiently.

Robust PID Design Based on Static H_∞ Loop-Shaping Method for Steam Generator Water Level Control

ZHOU Shiliang, LIU Yuyan, DUAN Feng

2012, 33(6): 35-41.

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The problem for control of steam generator water level is complex because its complicated shrink and parameters of its model vary with operation conditions.Static H∞ loop shaping is used to design the robust PID controller for the steam generator.Simulation results show that the control efficiency of this scheme is better than that of the other three H∞ loop shaping based methods,and the derived controller has good reference tracking capability,distribution rejection ability and acceptable control performance at different operation points.

Simulation Study on Correction Factor for Changing of Power Range Neutron Flux Rate of NPP Nuclear Instrumentation System

ZHANG Ying, CHEN Zhi, WANG Shu, SUN Jian

2012, 33(6): 42-45,60.

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The change rate of neutron flux in nuclear power plant（NPP） nuclear instrumentation system is an important protection parameter.If the parameter is corrected properly,the reactor trip function not only can be carried out normally in rod ejection accident and rod drop accident,but also the economic effect can be assured by avoiding reactor trip at the houseload rejection operation condition.The basic correction theory is explained in this paper.Furthermore,the paper uses the Ling’ao phase II NPP project and Qinshan phase II NPP extension project as examples and discusses the basic simulation study method of correction factor for the change rate of the neutron flux by the simulation analysis tools.The study result is applied successfully in tests on site.

Study on Application of CHF Models on Reactor Pressure Vessel Lower Head for Severe Accident Condition

YU Hongxing, SU Guanghui, GUAN Zhonghua, HUANG Daishun

2012, 33(6): 46-50.

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According to the in-vessel retention（IVR） mitigation and the heat transfer characteristics,two theoretical models for predicting the critical heat flux（CHF）,the interfacial lift-off theoretical model and the theoretical model considering the hydrodynamic behavior of the vapor-liquid interface of a bubble,are modified,respectively,and a general CHF model is developed to analyze the lower head CHF.The results indicate that the CHF model could well predict the experimental data on large scale curved surface.

Study on Two Phase Pressure Drop of Saturated Boiling in Narrow Gap Channel under High Pressure

QIN Shengjie, CHEN Bingde, YAN Xiao, HONG Gang, XIAO Zejun, HUANG Yanping

2012, 33(6): 51-54.

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An experiment on two phase pressure drop in narrow rectangular channel was carried out.The experiment shows that the current classical correlations and the prediction methods based on air-water and organic coolant are with large deviation in the prediction of the pressure drop in narrow rectangular channels.The dimensionless Nconf number was used as an important parameter to predict the pressure drop in narrow rectangular channel.Prediction is conducted by the correlations for saturated two-phase resistance in narrow rectangular channels based on Chisholm’s B coefficient method.The deviation of predicted result with the experimental data is within ±10%.

Study on Two-Phase Heat Transfer Enhancement by Longitudinal Vortex

HUANG Jun, HUANG Yanping, MA Jian, WANG Yanlin, WANG Qiuwang

2012, 33(6): 55-60.

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Experiment of two-phase heat transfer enhancement has been conducted in narrow rectangular channel with longitudinal vortex generators.The results show that coefficient of boiling heat transfer is more effected by mass flow flux than quality and inlet pressure.The mechanism of boiling heat transfer is changed by longitudinal vortex generators in narrow rectangular channel.Within vapor-liquid two-phase condition,the boiling heat transfer coefficient is increased by 1.1%-25.8% in narrow rectangular channel with longitudinal vortex generators.

Study on Heat Transfer Prediction Models of Precursory Cooling Region in Reflooding Phase of Tight Lattice

WU Dan, YU Hongxing

2012, 33(6): 61-64,71.

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On the basis of the reflooding experiment NEPTUN-LWHCR performed in Switzerland,this paper summarized the characteristics of the reflooding phase of tight lattice,and analyzed the reasons for the high peak clad temperature.A new heat transfer model of precursory cooling is established.It is proposed that the steam cooling is relatively poor in the precursory cooling to make the peak clad temperature higher.Through validating this mew model by calculating the peak clad temperature,it can be concluded that the new model is reasonable.

VOF Simulation of Bubble Characteristics of Subcooled Flow Boiling

WEI Jinghua, PAN Liangming, YUAN Dewen, YAN Xiao, HUANG Yanping

2012, 33(6): 65-71.

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Subcooled flow boiling in vertical rectangular channel under different pressure and heat flux was numerically investigated based on CFD.The phase change process was accomplished through the User Defined Function（UDF） to describe the mass and energy transportation,and the interface of liquid and vapor was captured by the Volume of Fluid（VOF） method.The results reveal that the interaction of evaporation and condensation in the cross section perpendicular to the flow direction forms the secondary flow,and enhances the natural convection near the wall.The bubble grows up in the sliding process,and merges with the adjacent ones to form bigger one,which has greater deformation.The bubble sliding enhances the heat transfer in the down stream and restrains the nucleation.With the higher pressure and lower heat flux,the bubble size,growth rate and the average void fraction at the outlet will be reduced.The simulation results of bubble growth and the wall temperature near the Onset of Boiling（ONB） agree well with the correlations in the literature.

Analysis of Influencer for Secondary-Side Passive Residual Heat Removal System Based on RELAP5 Code

ZHOU Lei, XI Zhao, XIONG Wanyu, YAN Xiao, XIAO Zejun

2012, 33(6): 72-75,87.

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Calculation and analysis are carried out for the secondary side of secondary-side passive residual heat removal system（SPRHRS）based on RELAP5 code.Effects of several factors such as the initial height of water column,initial temperature of water column,resistance coefficient,heating power and initial water volume on the characteristics of natural circulation are studied.

Research on Axial Total Pressure Distributions of Sonic Steam Jetin Subcooled Water

WU Xinzhuang, LI Wenjun, YAN Junjie

2012, 33(6): 76-80.

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The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters（6.0mm,8.0mm and 10.0mm）.The inlet steam pressure,and pool subcooling subcooled water temperature were in the range of 0.2-0.6MPa and 420-860℃,respectively.The effect of steam pressure,subcooling water temperature and nozzle size on the axial pressure distributions were obtained,and also the characteristics of the maximum pressure and its position were studied.The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure,but the influence could be ignored for high steam pressure.Moreover,a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature,and the discrepancies of predictions and experiments are within ±15%.

Study on Radiation Shielding Function Aggregate and its Calcination Mechanism

HUANG Xiulin, DING Qingjun, SUN Hua, YANG Kun

2012, 33(6): 81-87.

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Using the preparation method of artificial lightweight aggregate,the waste sludge containing heavy metal Ba from barium salt factory is roasted to the radiation function aggregate（RSFA）,which has the ability of radiation shielding,and the effect of different sludge concentrations and different n values on its density,mechanical properties and shielding properties is studied,and also the roasting mechanism by XRD and FTIR is analyzed.The study shows that the appropriate BaO content is 8%-12%,and n values is between 3.16 to 4.55.Under this condition,compressive strength of RSFA can be up to 60 MPa,and the apparent density is above 2 g/cm³,and linear attenuation coefficient to γ ray is more than 0.19 cm-1.

Effect of Annealing Temperatures on Corrosion Resistance of a Zr-0.85Sn-0.16Nb-0.38Fe-0.18Cr Alloy

YAO Meiyi, ZHANG Weipeng, ZHOU Jun, ZHOU Bangxin, LI Qiang

2012, 33(6): 88-92.

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To investigate the effect of annealing temperatures on the corrosion resistance of a Zr-0.85Sn-0.16Nb-0.38Fe-0.18Cr alloy,different annealing temperatures（740-820℃） before and after the final cold rolling were employed to prepare the specimens for corrosion tests.The specimens were corroded in deionized water at 360℃/18.6 MPa,in super heated steam at 400℃/10.3 MPa and at 500℃/10.3 MPa,respectively by autoclave test.The microstructure including the size and composition of second phase particles（SPPs） was examined by TEM and EDS.Results show that the SPPs in the Zr-0.85Sn-0.16Nb-0.38Fe-0.18Cr alloy are Zr（Fe,Cr,Nb）2 with a hcp structure;With the increasing of annealing temperatures,the size of the SPPs increases and the Nb content in the SPPs decreases;Under the three test conditions,the specimens annealed at different temperatures between 740℃ and 800℃ are all with corrosion resistance behavior as good as the specimen prepared by conventional procedures,and their corrosion resistance is superior to Zircaloy-4.This indicates that the corrosion resistance of this alloy is insensitive to the annealing temperatures.

Research on Process Pipe Vibration Improvements in NPP Conventional Island

LI Gang, LIANG Bingbing, YIN Haifeng

2012, 33(6): 93-95,100.

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Process pipe in NPP conventional island may vibrate by unstable fluid disturbance,since most of the pipes are non-seismic and flexible designed.Use a specific NPP conventional island pipe as an example,this paper studied the vibration amplitude,the natural frequency,the vibration limit,the vibration reason,and proposed and implemented the improved methods without changing the piping layout.The result shows that after improvement,the natural frequency increased significantly and the vibration amplitude qualified the vibration limit of ASME code OM3.

Preventive Measures to Avoid Core Uncovering in Cold Shutdown for Pressurized Water Reactor Nuclear Power Plants

XIONG Guohua, QIN Yuxin, ZHANG Qiang, ZHANG Tao

2012, 33(6): 96-100.

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During cold shutdown,the loss of residual removal function（unavailable of the RHR pump） due to the vortex suction or cavitations distinctly increases the core uncovering possibility,which might lead to core meltdown.In order to prevent the core uncovering accident during cold shutdown,the water motion in the reactor primary circuit and the RHR degrade incidents occurred in nuclear power plants in the world are analyzed and the three following preventive measurement are proposed: RCP water level measurement,RHR pump vortex detection and water makeup after loss of RHR function.The measures are based on experiences of Daya Bay nuclear power plant.The Probability Risk Analysis indicates that these three modifications decrease core uncovering efficiently.

Numerical Analysis for Causes of Cavitation Fracture Working Condition on Centrifugal Pump

WANG Xiuli, YUAN Shouqi, ZHU Rongsheng, YU Zhijun

2012, 33(6): 101-104,114.

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In order to research the flow-head curve plunge caused by the cavitation of centrifugal pump,the standard k-ε turbulence model,homogeneous multiphase model and Rayleigh-Plesset equation were applied to simulate the cavitation characteristics in a centrifugal pump with specific speed of 59 under different conditions based on ANSYS CFX software.The results show that the numerical simulation result has the same trend with experiment result,and absolute error is 0.02%.The analysis of flow field shows that: the steep fall of flow-head curve is not only caused by the traditional cativation,but also mainly caused by the Vortex loss.As the empty bubble in the passageway increases to some degree,the liquid boundary layer separation happened,then vortex appears and vortex losses.While the vortex appears originally,it has an impact on the flow-head curve.When the bubble becomes more and the whole passageway is full of vortex,cavitation fault condition happens.It reveals the vapor-liquid tow-phase flow distribution within the centrifugal pump.

Research on Fault Diagnosis Method for Rotating Machinery Vibration Based on Wavelet Transformation and Probabilistic Neural Network

WU Wenjie, HUANG Dagui

2012, 33(6): 105-109.

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Based on wavelet transformation and Neural Network Data Fusion,a Fault Diagnosis Technology is proposed.Fault feature extraction is carried out using wavelet decomposition,probabilistic neural network fault diagnosis technologies by optimizing the selection,and by the MATLAB Simulation.The simulation and results verify that using wavelet decomposition extract fault characteristics of the energy vector,which has strong generalization ability and anti-noise ability to adapt to Wide dynamic range and small sample,and building the adaptive probabilistic neural network is a good anti-noise capability,classification advantage of the high rate of diagnostic accuracy.Integration of the wavele and neural network application will provide a better classification of diagnosis results,and better reliability and accuracy.

Exploration on Parallel Operation of LOCA Surveillance System in Daya Bay NPP

ZHANG Yilin, WANG Yuan, WANG Min

2012, 33(6): 110-114.

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LOCA Surveillance System（LSS） in Daya Bay NPP is facing the problems of aging parts,unavailable spare parts leading to unstable work and the frequent failures,and the modification and upgrading of the whole LOCAL system is demanded.To verify the basic function and performance of the new system and at the same time to avoid any effect on the normal operation of the old system,the new and old systems of LSS for Unit 2 are operated in parallel in Daya Bay NPP.This paper describes the design solution of the new and old LSS parallel operation,then the required function tests for the parallel operation are conducted to verify the function and performance of the new system.The expected requirement of the parallel operation is satisfied.

Discussion on Spent Fuel Bay Purification Resin Selection

SHEN Zhaogen, LUO Shihua

2012, 33(6): 115-117.

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The performance of spent fuel bay（SFB） clean up system is important for both radiation and radwaste reduction.Demineralizer resin is operated adverse conditions facing high levels of hydrogen peroxide generated from radiolysis of water and also direct radiation attack from radioactive colloids in the pool.Resin cross-linked structure will be easily destroyed,because of sulfate ion and purification efficiency rise and decline.This paper,through nuclear resin oxidation stability tests and based on the nuclear power plant application experience,discusses the failure mechanism of resin,and give the purification resin selection principle,in order to improve the purification efficiency and reduce the amount of radioactive spent resin.

Research on Radioactive Waste Volume Reduction Technology in Qinshan Nuclear Power Base

KANG Yunding

2012, 33(6): 118-120.

Abstract(16) PDF(1)

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The paper analyzes the present situation of radioactive wastes in Qinshan nuclear power base and its processing and storage,and focuses on the volume reduction technology for large number of solid wastes including steamed residual liquid condensates,compressible waste and waste resin.Research shows that there is great potential for volume reduction of radioactive wastes in Qinshan nuclear power base,and the volume reduction scheme is provided.With the proposed scheme,the storage capacity for solid wastes in Qinshan nuclear power base can be increased by more than 5 years,which will effectively relieve the pressure for waste storage capacity.

Research on Several Abnormities in Vaporization Treatment of Radioactive Waste Liquid with Complicated Composition

KONG Jingsong, GUO Weiqun

2012, 33(6): 121-123.

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Based on the engineering practice of the vaporization treatment of radioactive waste liquid,a detailed analysis of the causes for the problems during the vaporization treatment of the radioactive waste liquid with complicated composition is described in the paper,such as the treatment capability that can not satisfy the design criteria,the violent fluctuation of the level in evaporator,and the blockage of pipeline by crystalization of the condensed liquid.The corresponding solutions of these problems are given in the paper also.

Research on Rejection Performance of Reverse Osmosis to Manganese in Simulated Radioactive Wastewater

KONG Jingsong, WANG Xiaowei

2012, 33(6): 124-126.

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In order to reveal the performance of reverse osmosis applied in the radioactive wastewater treatment,treatment experiments are carried out on a pilot RO equipment using wastewater containing manganese nuclide.Results show that the rejection ratio of RO to manganese is almost not influenced by the operation pressure and the ration of reclaiming,and has no direct relation with the salt rejection ratio.The ratio of manganese rejection is more than 95% and can meet the requirement on the disposal of radioactive wastewater produced by pressurized water reactors.

Analysis on Zero Power Experiment of High Flux Engineering Test Reactor with Three-Dimensional Continuous Energy Monte Carlo Code

PENG Gang

2012, 33(6): 127-131,138.

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Three-dimensional,continuous energy Monte Carlo code（MCNP） is adopted to carry out the analysis of zero power experiment of HFETR.From the results,the impurity,density of Beryllium block and the ¹⁰B concentration in control rod transition part should be carefully determined in the analysis of zero power experiment.While in the experiment of HFETR,the Beryllium block is considered as pure metal and the ¹⁰B concentration in control rod transition part is different from that of zero power experiment.From the calculation results,these parameters（effective neutron multiplication factor k_eff,relative distribution of neutron flux density,γ dose rate distribution and component reactivity） are quite fit with the experiment.The difference of small reactivity between calculation and experiment is quite large,and may be related to the deficiency of MCNP model.

Research Progress of SCWR Water Chemistry and Related Technologies

GONG Bin, HUANG Yanping, JIANG E, LIU Jinhua, XIA Xiaojiao, QIU Tian, HUO Songmin

2012, 33(6): 132-138.

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The progress of ongoing research on three sub-fields including change of water chemistry under radiation,effect of water chemistry on corrosion of candidate materials,corrosion product behaviors and corresponding monitoring and control methods,is reviewed in this paper.The recent experiment results obtained by SCW test facility are introduced;the research progress in Nuclear Power Institute of China（NPIC） on corrosion of candidate materials is also included.The future challenges in water chemistry field of SCWR are summarized.

Primary Study on Breeding Property of Improved Supercritical Water Cooled Fast Reactor

LIU Zijing, YU Tao, XIE Jinsen

2012, 33(6): 139-143.

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In this paper,the core mode of improved supercritical water cooled fast reactor is established.At first,reasonable fuel assembly design is obtained by studying the influences of seed fuel pin diameter and blanket coolant channel diameter to conversion ratio（CR）.Then,viod reactivity coefficient and CR of six different core arrangements are calculated.Finaly,the influences of fuel components to CR and void reactivity coefficient are analysed.The results show that negative void reactivity coefficient can be satisfied and CR can be increased by reducing Hydrogen to Heavy-metal ratio（H/HM）,increasing blanket assembly numbers by proper distribution.CR is substantially increased and more negative void reactivity coefficient can be met by reducing PuO₂ mass ratio in fuel.when PuO₂ mass ratio reach 20.8% in MOX fuel and ²³⁵U enriched at 0.2% in UO₂ fuel have been adopted as seed and blanket assmbly respectively,the sixth core program reaches CR=1.04395 and give negative void reactivity coefficient,which meets the primary requirements for SCFR breeding.

Research on Quality Assurance Classification Methodology for Domestic AP1000 Nuclear Power Projects

BAI Jinhua, JIANG Huixia, LI Jingyan

2012, 33(6): 144-146.

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To meet the quality assurance classification requirements of domestic nuclear safety codes and standards,this paper analyzes the quality assurance classification methodology of domestic AP1000 nuclear power projects at present,and proposes the quality assurance classification methodology for subsequent AP1000 nuclear power projects.

Design of the Ancillary Equipment for Spent Fuel Transport of PWR Nuclear Power Plant

WENG Songfeng

2012, 33(6): 147-150.

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In order to load and unload spent fuel assemblies safely in NPP,the ancillary equipment for spent fuel transport used to perform fluid transfer activities such as water feeding,air charging,and cask drying and heat rejection.This paper introduces the requirements,the scheme and the principle of the ancillary equipment,which adopts centralized-control and modular design,and have the performances such as higher work efficiency,higher safety,less radioactive substance accumulation and friendly man-machine conversation.

2012 Vol. 33, No. 6