核动力工程

Research and Development of Modern Large-Scale Reactor Pressure Vessel Materials

ZHU Zheng-qing

2011, 32(S2): 1-4,8.

Abstract(15) PDF(1)

Abstract:
The purpose of the study is to improve the internal quality,overall performance and safety margin of the large-scale reactor pressure vessel materials,to meet the requirements of the reactor pressure vessel macro-sized,integrated and long-lifetime on the special property of its structural materials.This paper covers the followings: the structure,function,service conditions,and manufacturing process characteristics of modern reactor pressure vessel and the special requirements on its material,and the research and development of domestic and foreign modern large-scale reactor pressure vessels.This paper mainly summarizes the experimental study on nuclear application,manufacturing process and the achievement of the performance improvement of domestic and foreign modern large-scale reactor pressure vessel materials,analyzes the engineering value of the above research results,and gives the guidance and suggestion on the improvement of the materials of the domestic large-scale reactor pressure vessels,the operation of manufacturing processes and the overall performance improvement.

Application of Multiple Spectrum Analysis Method in Nuclear Pipe Seismic Analysis

TANG Feng, ZHANG Zhou-hong

2011, 32(S2): 5-8.

Abstract(28) PDF(0)

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Multiple spectrum analysis method and single spectrum analysis method are two stand analysis methods which are used for the pipe seismic analysis in nuclear power plants.Multi-support excitation is applied to pipe supports in multiple spectrum analysis method,and uniform excitation is applied to pipe supports in the single spectrum analysis method.Variation of seismic motion according to space position is considered in the multiple spectrum analysis method.In this paper,multiple spectrum analysis method and single spectrum analysis method are used in the seismic analysis of VVP line after damper-cutting,to compare the effect on damper under these two analysis methods.The results reveals that,the conservatism of two analysis methods is different,and for the auxiliary pipes coupling with primary pipes,multiple spectrum analysis method can be applied in the seismic analysis;On the other hand,for the small scale pipes decoupling with primary pipe,single spectrum analysis method is conservative in the seismic analysis.

Analysis on Fatigue Strength Reduction Factor of Reactor Pressure Vessel Flange Thread

LIU Pan, JIN Ting, MAO Qing

2011, 32(S2): 9-12.

Abstract(19) PDF(1)

Abstract:
Considering the model type,element type,numbers of grid and contact type,the mechanics property is analyzed and the theoretic fatigue strength reduction factor is obtained.The results show that the fatigue strength reduction factor peaks under the tension-compression bearing way which can envelope the bending bearing way and torsion bearing way.Meanwhile,it shows that the contact type has the greatest effect on the fatigue strength reduction factor.Under bonded contact type,the effect of mesh refinement is ignorable.With the increasing of grid numbers for hexahedral element,the fatigue strength reduction factor increases,and vice versa for the tetrahedron element.

Analysis of Thermal and Hydro Characteristics of Steam Generator Blowdown System

XIE En-fei, LIU Xi-chao, MING Qian

2011, 32(S2): 13-17,22.

Abstract(19) PDF(0)

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In this paper,the thermal and hydro characteristics of steam generator blowdown system in CPR1000 nuclear power plant is analyzed by flowmaster software.The set model is changed,and the parameters which affect the system severity are greatly considered.From the analyzing we know that,fast change of SG blowdown water temperature will not affect the regenerative heat exchanger because of "thermal inertia function",and the temperature downstream regenerative heat exchanger is not sensitive to the change of SG blowdown water but sensitive to the entrance temperature of cooling water,and quick close of treatment equipment（demineralizer and filter） may lead high-press alarm and safe-valve release.

Experimental Calculation and Analysis of Thermal Performance for Steam Turbine of Nuclear Power Plants

WANG Shi-yong, BU Yu-bing, XU Zong-fu

2011, 32(S2): 18-22.

Abstract(22) PDF(0)

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Small deviation heat balance method can solve the calculation problems of steam turbine wet steam zone,and characteristic flow area of steam turbine can be used for the performance calculations and analysis of steam turbine flow area.In this paper,both methods are used in the experimental calculation and analysis of thermal performance of the nuclear power steam turbine,and the results are verified by the performance test on No.3 turbine in Ling’ao Phase II,and test calculation method are proved to be accurate and reliable.

Reference and Analysis of Design Criteria for Chilled Water System and Water Chiller of Nuclear Island Electrical Workshop

WANG Zhi-sheng, LIU Xiao-zhong

2011, 32(S2): 23-26,30.

Abstract(21) PDF(0)

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This paper introduces the safety classification of the chilled water system and water chiller for nuclear island electrical workshop of Generation II plus PWR NPP project.By analyzing the correlative design criteria,some questions about safety classification of the DEL system and DEL chiller is asked.Meanwhile,some constructive advice on the operability of the regulation on the supervision and management of the civil nuclear safety equipment is proposed.

Correction Calculation of Circulating Cooling Water Pipeline Resistance

MIAO Hui, SHANG Xian-you, HOU Ping-li

2011, 32(S2): 27-30.

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The relevant factors of circulating cooling water pipeline resistance are analyzed and a correction calculation method of resistance is provided.When the condenser parameters change,corresponding to different cooling water flow,the circulating cooling water pipeline resistance can be modified according to the reference power plant condenser water resistance curve.This method can be used in the optimization of steam turbine cold end.

Study on Heat Recovery Technology of Closed Cooling Water System for Nuclear Power Plants

SHI Jian-zhong, HOU Ping-li, HU You-qing, WANG Shi-yong

2011, 32(S2): 31-33.

Abstract(23) PDF(0)

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A new technology of energy saving and emission reducing which can recover the heat of closed cooling water was proposed.The condensate is used to absorb the heat from the closed cooling water when the condensate temperature is not greater than 34℃,instead of sea water.An application case for Hongyanhe 1000 MW class nuclear power plant is given to show that with this method,the unit output would be increased about 1.5 MW,the warm water emission would be reduced about 2.1×10⁷ t /a and the economic benefit would be increased about 3.58×10⁶ RMB each year.

Selection and Design of Spring Foundation of Half Speed Turbine Generator for Ling’ao Phase Ⅱ NPP

SONG Yuan-qi, ZHOU Xiang-qun, WU Jian-jun, FU Yu

2011, 32(S2): 34-37.

Abstract(18) PDF(0)

Abstract:
Half-speed steam turbine generator for Ling’ao phase II nuclear power plant uses the spring vibration isolation foundation at first.From the type selection,design standards and vibration analysis and comparison with the measured data,the design process of spring foundation is elaborated.Exerting fully the advantages of the spring foundation,such as "low frequency",the earthquake-resistance and the adjustability,the safe operation of the half-speed steam turbine unit is ensured.

Application of GPS-RTK in BOP Axis Check of Binhai Power Station Project

LIU Qi-rui, YE Rui, FENG Ke

2011, 32(S2): 38-43.

Abstract(23) PDF(0)

Abstract:
RTK is a real time kinematic GPS measuring technique,the application of which is a milestone in the development of measuring techniques.The paper simply introduces the principle of GPS-RTK and analyzes the advantages and drawbacks of this technique,as well as the factors which can influence the precision of RTK.Besides,the paper also discusses the feasibility of BOP axis check and proposes some relevant countermeasures.

Research and Application of Modularized Scheme of Containment Liner of Yangjiang NPP Reactor Building

LU Qin-wu, HAO Jian, ZHANG Shu-xia, GUO Jun-ying, CAI Hui

2011, 32(S2): 44-46,57.

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Research of the modularized scheme in the design,spatial grid structure spreader and crane,and its successful application in Yangjiang NPP project 3# unit provided a reference for the modularized construction of containment liner in the nuclear construction projects.

Preparation and Construction Technology of Nuclear Power Plant High-Performance Mass Concrete

WANG Hui-cheng, CHEN Li-sheng, LV Guangye

2011, 32(S2): 47-50.

Abstract(22) PDF(0)

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The paper introduces the technical specifications and requirements on the mass concrete for the reactor plant of CPR1000 + nuclear power plant in detail.Taking the A,B,and C layer overall placing of Unit 3 reactor plant（3RX） in Yangjiang nuclear power plant as an example,the reasonable concrete mix through experimental studies significantly improve the concrete performance and solve the durability problems of CPR1000+ mass concrete.Through the quality control of mass concrete such as mold-into temperature control and conservation measures,the temperature cracks and other issues in mass concrete construction are solved.

Review on Loss of Power Supply Incident Analysis for Ling’ ao Phase Ⅱ NPP

WANG Zhen-ying, ZHENG Wen-bo, LI Hong-lin, LI Shuang, LI Min

2011, 32(S2): 51-57.

Abstract(17) PDF(0)

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The loss of power supply incident analysis for Ling’ ao nuclear power plant phase 2 include the background information on loss of power supply incident analysis,the analysis scope,the analysis approaches,the applications and the validation.Some achievements,such as the determination of management strategy on loss of power supply incident,the analysis approach of DCS loss of power and the analysis of DCS fallback value,are innovative,and have being applied to new nuclear power projects.

Partial Discharge Test for 1000 MW Nuclear Power Generator

ZHOU Li, LI Gang, TAO Hai-quan, CAO Bang-wei, PAN Kai

2011, 32(S2): 58-62.

Abstract(28) PDF(0)

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Partial discharge test is an effective method for monitoring and preventing winding insulation defects of generator stators.This paper,through the comparison of partial discharge tests on generators for Unit 3 & 4 of Ling’ao Phase II and the generator in DFEM for unit 1 of Hongyanhe Phase I,analyzes and studies the problem of partial discharge value beyond national standard for the first domestic 1000MW nuclear power inner water cooling generators.The cause for the abnormal partial discharge value is found out,and it is concluded that the partial discharge value is an inherent property of this type of generator,which provides a theoretical foundation for the safe operation and partial discharge test evaluation of same models.

Investigation on Relationship between Steady Speed Droop and Periodic Test Mode of Emergency Diesel Generator Set in Nuclear Power Plants

CAO Yu

2011, 32(S2): 63-68.

Abstract(22) PDF(0)

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Relationship between steady speed droop and periodic test of emergency diesel generator in nuclear power plants is investigated.With respect to the speed governing characteristics of emergency diesel generator set used in Ling’ao Phase II nuclear power plant,modes selection are proposed under detailed analysis in different speed droop and periodic test modes which is referring to relevant international standard and power plant regulation.In normal operation of the power plant,the emergency diesel generator sets are in hot standby which is under the normal pre-lubrication and preheating.In case an exceptional event or accident happens to the power plant,power loss of emergency bus-bar will automatically start the emergency diesel generator sets promptly within required safe time,so that the safety equipments in the power plant recover the electricity power.To testify the reliability of generator sets and system,a periodic test mode is stipulated in plant regulation for the power source equipments.Meanwhile,because of the property of power source,its quality of output should fulfill relevant requirements to make downstream consumers operate normally and steadily.It can demonstrate through the investigation and research that the periodic test mode can be considered and selected combining the equipment characteristic and maintenance strategy so as to heighten the reliability of equipments and further ensure the security of power plant.

Development and Implementation of NPP Function Test System Based on Virtual Instrument

ZHANG Xu-feng, HUANG Tie-ming, LI Xian-min, YANG Zong-wei, LUAN Zhen-hua

2011, 32(S2): 69-72.

Abstract(23) PDF(0)

Abstract:
This paper described the virtual instrument process test platform based on Agilent VEE visual engineering environment virtual instrument graphical programming language and PC-DAQ data acquisition technology,which was connected with DCS control cabinets I/O modules and performed nuclear power plant control logic function test to realize the consistency checking between DCS configuration and system design manual.The implementation result in Ling’ao phase Ⅱ has proved that this technology could correspond to DCS control system commissioning periods and characteristics,optimize the commissioning schedule and reduce the commissioning risks.

Start-up Test of Feed-Water Control System in Ling’ao Phase Ⅱ NPP

SHI Xiao-long, LUAN Zhen-hua, YANG Zong-wei, FENG Guang-yu, LIU Yang

2011, 32(S2): 73-77.

Abstract(21) PDF(0)

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During the commissioning of Ling’ao Phase II NPP,a full-scale test of the logical configuration in DCS was conducted on the feed-water control system.Based on RRC simulator,the feed-water control system had been tested in normal operation and transient conditions.And lots of logical errors and design nonconformity were found.After the modification,normal operation and fine control can be realized on the feed-water control system in the automatic mode.The double fault happened in the feed-water pump simulation result provided the guarantee for ATO50 test,and also had great importance on the actual transient test.

Research on Effect of Film Color Difference in Core Sheets on Core Performance in Large Transformer

DU Ya-ping, LU Zhen-guo, CHEN Qi, JIANG Xin

2011, 32(S2): 78-80,84.

Abstract(26) PDF(0)

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The paint film defects of magnetic sheet occurred during the manufacturing process of high-capacity electrical transformer is studied and its origin is analyzed.Tests and analysis find that there is no effect on the performance of magnetic core.

Research on TXP Controller PID Parameter Matching of Ling’ao Phase Ⅱ NPP

LUAN Zhen-hua, YU Hang, CHOU Shao-shuai, SHI Xiao-long, LI Xian-min, YANG Zong-wei, ZHANG Xu-feng

2011, 32(S2): 81-84.

Abstract(28) PDF(0)

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The reason for the mismatching of the PID parameter and the design value in TXP controller of DCS system in Ling’ao Phase II NPP is analyzed.Taking the pressurizer pressure control system as an example,the effect of the parameter mismatching is described,and the PID parameter matching method is studied based on the open-loop response method.The hot function test and transient test in Ling’ao Phase II NPP shows that this method can match PID parameter with the design value,to get better control performance,shorter startup time of control system and huge economic benefits.

Analysis of Solutions for Cable Out-of-Tolerance in DCS Control Room of Nuclear Power Plants

YANG Lin, JIANG Zhi-wei, LIU Jian-yi, CHEN Wen

2011, 32(S2): 85-89.

Abstract(17) PDF(0)

Abstract:
The reasons and potential risks of cable out-of-tolerance,which happened in CPR1000 DCS control room,are analyzed in this paper.And optimization of the civil structure,I/O list distribution,equipment structure and layout,cable character selection,and trunk cable design are also introduced,and the advantages and disadvantages of the above mentioned solutions are analyzed.

Improvement of RCV Letdown Temperature Control in Daya Bay Nuclear Power Plant

ZHANG Ben, HUANG Wei-jun, YANG Guo-jun

2011, 32(S2): 90-96.

Abstract(26) PDF(0)

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The letdown temperature adjustment and control circuit in the Chemical and Volume control system RCV takes on serious time-delay and feeble repellence characteristics when the temperature acutely fluctuates or the outside disturbs both in the Daya Bay nuclear power station and Ling’ao nuclear power station,which cause lies in the value and the inherent pure time-delay characteristics of the heat exchanger in this circuit,as well as the controller used in the circuit is a simple PI controller.By comparison,a more advanced technique,combining the Smith predictive control,adaptive fuzzy control and PID control,is put forward based on theoretical derivation and simulation experimental verification.This technique gives the control system a fine dynamic stability and robustness,to overcome the disadvantage of the original circuit.

Cause Analysis on Condenser Corrosion based on Accident Tree

ZHANG Wei-ming, ZHANG Hong-liu

2011, 32(S2): 97-100.

Abstract(25) PDF(0)

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A great deal of corrosion was found on the condenser during the commissioning.The analysis revealed that the rust of the condenser was caused by the followings: the broken supporting-rod at water room entrance,welding quality,expanding tube quality,and etc.Firstly,the middle-event is analyzed from the top-event;then step by step,the underlying causes of the middle-event are analyzed;finally,the ultimate cause for the quality-events（end-event） is obtained,and suggestion for the improvement of the condenser manufacturing is given.

Optimization of Design Logic by Linear Programming

WANG Hao, MA Zhi-hua

2011, 32(S2): 101-104.

Abstract(20) PDF(0)

Abstract:
During the schedule management progress for the detailed design of the nuclear power project,the model structure of the design logic is optimized by the operational research and linear programming,to optimize the scheduled construction.Of all the work,the design logic model for the electrical cable tray is modified and optimized.

CGNPC Commissioning Organization and Strategy Transformation Based on Multi-Projects,Multi-Bases and Multi-Courses

YU Yang, TIAN Xi-feng, WANG Can

2011, 32(S2): 105-108.

Abstract(23) PDF(0)

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With the rapid development of nuclear power commissioning and startup industry,a lot of commissioning organizations will confront both new opportunities and challenges.This paper introduces briefly the running and management of CGNPC commissioning organization and its development status.Based on the analysis of the developmend trend of the nuclear power industry,the strategy transformation and development goal for CGNPC commissioning organization is propsed,considering the new situation of multi-projects,multi-bases and multi-courses,and several suggestions on the sound development of nuclear power commissioning and startup industry is proposed.

Research on Quality Risk of Nuclear Power Engineering Construction Project

LIU Wei, WU Zhi-jian, YIN Xiang-ping, HE Hong-jun

2011, 32(S2): 109-113.

Abstract(28) PDF(0)

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Applying the method of system and the engineering practice,the paper studies the nuclear power engineering construction project process of quality risk.From the point of view of risk management,the quality risk is analyzed and evaluated,and the table of quality risk and the model of risk evaluation are obtained.Based on the classification and considering the different risk features,the model of quality risk forewarning and reliability is established.

Study on Detailed Cost Control Scheme for Nuclear Power Plant Projects

CHEN Jian-bo

2011, 32(S2): 114-118,123.

Abstract(21) PDF(0)

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The paper analyzes the significance and main contents of project cost control,and provides the detailed cost control scheme for nuclear power plant projects including the establishment of project cost control organization,project cost planning and assessing mechanism,processing management,earned value management,inner-trade mechanism and expertise evaluation mechanism.It also anatomizes the keystones of process supervision for project cost.

Research on Methodology for CPR1000 NPP Fire Functional Analysis

ZHU He, REN Zhao-ying, JIN Xiao-hong

2011, 32(S2): 119-123.

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Fire common mode analysis is the safety conformity analysis of fire protection design of nuclear power plants.It is also an important guarantee for the safety of the nuclear power plant.Fire functional analysis is the most important step in the fire common mode analysis,and the basis of defining the fire common modes.This paper elaborates the methodology of the fire functional analysis based on the engineering practice,and then sorts the analysis methods for different equipments,giving the examples and explanations.This methodology has been applied in several projects effectively.

Study and Optimization of Debris Transport in Containment Sump

TANG Hui, HAN Zhi-hang, WANG Qing-li, ZHAO Yan-hui, YANG Jing-long

2011, 32(S2): 124-126,132.

Abstract(28) PDF(0)

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The process of debris transport in containment sump of CPR1000 nuclear power plant is studied,and different methods to decrease the debris load on the filter are analyzed and calculated.It is found that by setting some internal barriers into the containment sump,the flow in the sump will be more steady,and the debris load that can be transported to the filter will be reduced.Therefore,safer margin for sump filter design can be provided.

Research and Development of Module 3D Design System Based on PDMS

WU Xiang-yong, LI Yi, LU Qin-wu, LUO Ya-lin, MAO Jie, LI Zhen

2011, 32(S2): 127-132.

Abstract(22) PDF(0)

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Modularization construction technology for the construction of nuclear power plant is an effective way to shorten the construction period and reduce the project cost.CPR1000,at present with proven construction technology,has been developed for batch construction in China.However,it still has the potentiality for further improvement with respect to the construction period and project cost,and therefore,it has great significance to promote the research on the development and application of the CPR1000 Modularization technology.Relying on CPR1000 nuclear power projects,this study aims to explore the scheme of the module 3D design based on the PDMS,and develop the self-reliant module 3D design system,so as to offer a convenient and effective design tool for the Modularization design in CPR1000 nuclear power projects.The research will provide good reference to AP1000 nuclear power projects.

2011 Vol. 32, No. S2