核动力工程

Ergonomics 3D Simulation Design and Evaluation of Nuclear Power Plant Control Room

Yan Shengyuan, Chen Yu, Wang Yanshu, Chen Wenlong

2014, 35(S1): 1-4.

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Ergonomics design and evaluation of nuclear power plant control room is still mainly based on the designers experience currently.This method is easily influenced by personal experience and preferences that make it difficult to get a complete and scientific evaluation results.By establishment of nuclear power plant control room ergonomics design and evaluation index system,and develop nuclear power plant control room 3D simulation design and evaluation software,the interactive and visual of ergonomics design and evaluation platform is constructed.Case studies shown that the developed software can complete ergonomics 3D design and evaluation of nuclear power plant control room accurately and efficiently.

Analysis on Amendment of “Code of Quality Assurance for China Nuclear Power Plant Safety”

Zhang Yan, Liu Pengfei, Zhao Yan

2014, 35(S1): 5-7.

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Based on the studying of the Code and Guidance,learning correlative literature,and combined the practical work,it is analyzed and concluded that Code of Quality Assurance for Nuclear Power Plant Safety was not applicable as the development and advancement of quality management for nuclear power plants.This paper gave some suggestion to the amendment of HAF003 and HAD003 from safety culture,leadership,staff participation,QA grade,nonconformance,continually improve and process approach.Through comprehending and utilizing these management ideas,we could promote the organization management ability,have a better development and guarantee the nuclear safety.

Design Strategies to Improve the Human Factor Features in Control Room of HTR-PM

Jia Qianqian, Zhang Liangju

2014, 35(S1): 8-11.

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The application of the digital I＆C in NPPs improves the human machine interfaces in the control room,which also brings many new challenges in human machine interaction.In order to reduce the impact of the computerized HMI on operators,some strategies are presented based on the characteristics of the HTR-PM from the point view of a HMI designer.In the control room,besides the computerized HMIs,the spatially-dedicated continuously visible（SDCV） displays,such as traditional mimic displays and alarm tiles are included,which helps the operators keep big picture of the plant in mind.According to the preliminary results of the verification on the V＆V platform,the traditional SDCV displays can help reduce the impact of the computerized HMI on operators,such as key-hole effect,the interface management issue,etc.,which will support the operator better.

M310 Nuclear Power Plant Situation Calculation Design Used for Computerized Alarm Inhibition

Li Li, Xu Zhao

2014, 35(S1): 12-14.

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Operation condition of nuclear power units in the nuclear power plants with digital control system and advanced main control room shall be calculated in real-time by digital instrument controlled system（DCS） and presented to the operator in the main control room by user-friendly digital man-machine interface.The calculated results can be used for the computerized alarm inhibition process.This paper gives the calculation method and its implementation for the design operation condition.

Research and Design of Mobile Electricity Generators Added for CPR1000 NPP Station Blackout

Zhang Shuxing, Cao Yu

2014, 35(S1): 15-17.

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Long-drawn station blackout may occur from the Fukushima nuclear accident experience feedback,so it is necessary to add mobile electricity generators for NPP.This paper firstly analyzes CPR1000 NPP’s power configuration,then discusses the station blackout risk due to extreme natural disasters from the Fukushima nuclear accident experience feedback,whereafter fixes the functional orientation and essential loads of the mobile electricity generators,and then determines the reasonable power.This paper also describes the classification,advantages and disadvantages of mobile electricity generators,and recommends the appropriate type of power according to the specific needs of the NPP.Finally,it points out the key elements and advices for the design of mobile electricity generators.

Design of Display Function for Computerized Nuclear Power Plant General Operation

Xu Zhao, Li Li

2014, 35(S1): 18-20.

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In order to fulfill the Nuclear Power Plant（NPP） general operation requirement,the purpose of NPP general operation display function design is to integrate the DCS system information reasonably.A NPP general operation display function design method and related design example are provided in this paper,from the point of view of operation.And this paper also gives a preliminary analysis on the advantage of the design method.With this design method,the operator can get the DCS information flexibly,quickly and completely,and the safety and availability of NPP can be improved by this method.

Modeling and Simulation of Shared Condensation

Xue Ruojun, Chen Zhilong, Li Yanrui

2014, 35(S1): 21-25.

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Shared condenser have two steam imports,and it is consisted of two steam turbine.A dynamic mathematical model of the condensation system in shared condensation is presented based on the working principle of the system and structural.The operation characteristics of the shared condenser is simulated and analyzed and different dynamic characteristics of each subsection is obtained.

Validation Method for Computerized Alarm Procedure

Liu Yong, Yang Qingming

2014, 35(S1): 26-28.

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The Computerized Alarm Procedure is one of most important operating procedures in the NPP,and it will be used in the Digital Control System（DCS system）.The alarm will be trigged when the NPP operation parameters overstep the normal operation,the equipments statuses are different with the currently operating status,equipments failures and system equipments are inoperable,and so that the unit degraded status could be avoided.Based on the feedback from building NPP computerized alarm procedures validation process and validation results,this paper describes the general validation process in the computerized emergency procedures.By using this validation process,the correctness of alarm procedures drafting and computerized alarm procedures design can be ensured,and the requirements on NPP safe operation can be satisfied.

Influence Analyses of CPR1000 PWR SGTR with Full Scope Simulator Based on RELAP5-3D

Jiang Xialan, Qin Zhiguo

2014, 35(S1): 29-32.

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This paper researched the behavior of CRP1000 nuclear power plant steam generator tube rupture（SGTR） accident with full scope simulator（FSS）.Two different conditions of SGTR,without intervention in 1.5 hours after SGTR and with operator intervention according to operation specification are calculated.Results have been compared with the reference accident analysis report to verify this simulator.Tendency and reasons for the change of important parameters have been studied and analyzed,and the entire incident sequence and operation intervention are given.

Design and Implementation of Full Scope Simulator DIO Interface Card and Offline Detection System

SHi Xiaowei, Lu Mingxian, Wang Yu, Qiu Jianwen, Zhu Yuandong

2014, 35(S1): 33-36.

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A new RTP DIO interface card and offline test platform for the full scope simulation system in Qinshan 300 MWe nuclear power plant are designed,the DIO card is mainly used in translating system collected analog signal to digital signal,after logic operation,the digital signal is sent back to master machine through the RTP BUS,and at the same time the status signal is sent to the LED lamp for real-time status display,so that the real-time monitoring on the whole system running come true.

Comprehensive Measurement of Physical Properties of Floating Liquid Film on Large Surface in Test Facility

Zhang Ziyang, Lu Yanghui, Wang Huan, an Xu, Song Yihao

2014, 35(S1): 37-39.

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The object to be measured is the facility for water distribution experiments for CAP1400 which surface area is more than 200m².The high-accuracy capacitance spacing probes are used to get the thickness of the film.Many tanks are located under the facility to measure the levels of these tanks which show the rate of flow along the surface.CCD（Charge-coupled Device） industrial cameras is used to take image of the surface of the test object,and the video is treated by the especial software of machine vision.The coverage rate is obtained by calculating the whole proportion of wet area and dry area.By these new measurement methods,the law of floating liquid film in large space can be understood.

Analysis on System Commissioning Problems in Ling’ao II Nuclear Power Plant

Liu Jianwei, Fan Sui

2014, 35(S1): 40-43.

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To resolve the plant field problems during the test and start-up,the problem cause shall be found out firstly.To make up the optimized method with the field environment,the relation of design,system operation and test method shall be clear.Some problems for plant test and start-up,such as the valve control problem of boron recycle system（REA）,the signal block of residual heat removal system（RRA） and the pump false start-up of auxiliary feed water system（ASG） are stated.With deep understanding to the plant field environment and communicating with each one,the problem solutions are addressed in detail.

Improvement of Generator Loss-of-Excitation Protection in Tianwan Nuclear Power Plant

Jiang Xiaopeng, Li Cong, Liu Yang

2014, 35(S1): 44-46.

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This paper mainly presents how to improve the safety and reliability of generator loss-of-excitation protection of Tianwan nuclear power station,and the backward technology and poor reliability generator loss-of-excitation protection which was made in Russian had been transformed to microcomputer protection.A set of maintain effective experiences of designing,installation,commissioning and maintenance about generator loss-of-excitation protection is summarized.Practice has proved that it is of great advantage to enhance the operation reliability and safety of the generator,and shorten the maintenance period after the transformation of the generator loss-of-excitation protection.

Reconstruction of Second Highest/Lowest Selection in 600 MW PWR Control System

Liu Jun, Zhu Changrong

2014, 35(S1): 47-49.

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During Instrumentation Channel Calibration Test（T1 test） or under the condition of instrumentation malfunction,mistakes due to man-made intervention are observed now and then,so that the steady state operation of the unit is not guaranteed.In order to solve this problem,control logic of Second Highest Selection,Second Lowest Selection has been introduced to the control system of four units in the Second Plant of CNNP Nuclear Power Operations Management Co.,LTD.This paper describes the operation principle,method of work and detailed alteration of Second Highest Selection,Second Lowest Selection.Besides,based on the analysis of test data,the feasibility of this alteration project has been validated.At the same time,this paper introduces the detailed application of this alteration.

Analysis of Debugging Problems of Emergency Diesel Generator Excited System in Tianwan Nuclear Plant

Huang Changjun

2014, 35(S1): 50-53.

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This paper gives a brief introduction on the principle and the construction of the emergency diesel generator excited system,and analyzes the problems and the solutions to the problems during the course of the debugging in Tianwan Nuclear Power Plant.These experiences can provide a reference and practical value to the operation and maintenance.

Study on Quantitative Software Reliability for Digital Control System

Yang Ming, Song Mengchu

2014, 35(S1): 54-58.

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This paper presents a methodology for modeling and quantitative reliability assessment of nuclear safety-level digital I＆C system software based on Multilevel Flow Models（MFM）.By developing a MFM graphical modeling and analysis platform,this paper presents a MFM model for a PID control software of DCS and performs software reliability assessment based on test results.Using the proposed method of this paper,the reliability of main goal and all sub-goals of software can be obtained by only one calculation and the weak points in the software design are very convenient to be identified.In addition,the software models are easy to build and modify.

Relay Protection Orientation and CT/PT Polarity Static Energization Test Approach

Wu Bingchen

2014, 35(S1): 59-63.

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The CT/PT polarity and relay protection orientation are usually tested through the dynamic tests as no-load/under load unit tests.This paper proposes that the static energization before plant energization and unit start-up will avoid the potential breakdown of the on-going test,malfunction protection or equipment damage,and will reduce the workload and the test difficulty to ensure safety risks of dynamic tests are under control.

Analysis and Design of RCP Seal Water Flow Measurement

Shao Wen, Tan Yue

2014, 35(S1): 64-66.

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The control solution of HAINAN nuclear project’s RCP is changed compared with the reference plant,thus the RCP shaft seal water flow measurement is influenced,and the relevant I＆C design of Chemical and Volume Control System（RCV） has to be changed.A detailed analysis of several measure solution and related problems is expatiated here,and a feasible solution has been found to solve the problems,to make sure the system function is performed safely.

Optimization of Controls Layout Based on Simulated Annealing Algorithm

Yan Shengyuan, CHeN Yu, Liang Longyuan

2014, 35(S1): 67-70.

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In order to solve the problem of controls layout scheme varies from person to person and the non-traceability of layout evidences by using experience method,the simulated annealing algorithm based controls layout optimization method was proposed.The multiple ergonomics principles,such as the importance,operation frequency and operation sequence are used as the optimization varies,and the reciprocal of the multiple weights of controls as objective function,a mathematical model for optimizing the layout of the controls was established.An instance of controls layout was conducted,and the certain layout scheme was obtained.The instance shows that the layout method is scientific,the layout process is traceable,and the layout scheme is the best.

Numerical Simulation of Rising Bubble Behavior in Sloshing Condition

Song Yulin, Tan Sichao, Fu Xuekuan

2014, 35(S1): 71-74.

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The storm and other factors under the ocean conditions could cause the container which exist the free surface on the ship produced the intense sloshing motion.In order to analyze the rising bubble behavior in the liquid under the sloshing conditions.This paper uses the VOF（Volume of Fluid） model to simulate the rising bubbles in the liquid at the fluent platform.Calculation results show that the bubble can cyclical swing up and accompany by aggregation and separation when it is in sloshing conditions.The results also show that the violent sloshing motion can affect the bubble’s departure size,and accelerate the rise of bubbles which along with generating the separation bubbles,it can also hinder the movement of the bubble which can produce the aggregation phenomena,and the bubble will be deviated relatively large displacements in the horizontal direction,it cannot be ignored that the sloshing motion can make a serious impact to the behavioral characteristics of bubbles.

Review on Core Design of AP1000

Yao Zenghua

2014, 35(S1): 75-79.

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A comparison of some core design parameters of AP600 and AP1000 are performed based on the URD.It is suggested that the MSHIM Baseload Operation Mode shall not be used in Sanmen NPP and Haiyang NPP.

Comparison of Fire Protection Design at Pressurized Water Reactor Nuclear Power Plant

Liang Bo, CHeN Liping, Wang Shuai

2014, 35(S1): 80-84.

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The fire protection design of the 3rd generation passive pressurized nuclear power plant（AP1000 for short） in U.S.and the 3rd generation innovation pressurized nuclear power plant（EPR for short） in France and the 2nd plus generation pressurized nuclear power plant（CPR1000 for short） in China are introduced and compared.In respect of the purpose and principle of fire protection,fire prevention,fire detection and alarm,the smoke control of these three PWRs are basically the same.Due to the passive design,AP1000 fire water supply system is designed on the basis of "Different Solutions for Different Cases,Critical Defenses for Critical Places",comparing with the other two types,with more complexity in the classification of FPS,increasing the functions and diversification of the system and decreasing the fire load.

Study on Fuel Assembly Selection for Long Fuel Cycle Management in Nuclear Power Stations

Wang Congmei, Ye Guodong, Zhan Yongjie, Xiang Junjun, He Zishuai, Dai Qianjin, Li Yanrong

2014, 35(S1): 85-88.

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Based on the analysis of the performance of different fuel assemblies,and the domestic and overseas feedback,and on the actual instance in Qinshan Phase II,the fuel assembly for long fuel cycle management is AFA3G fuel assembly,with fuel length as 4060.2 mm,using the improved grid with three mixing vane and M5 structure material.

Improvement for Control Rod Driving Mechanism of Unit3＆4 of TNPS

Cheg Xu, Jiang Baiwen, Li Zhangshun

2014, 35(S1): 89-92.

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In view of the importance of the reliable operation and convenience maintenance of the control rod actuator to the safety operation and economic effectiveness of the reactor,through analyzing the deficiency of the control rod actuator of Unit 1＆2,the design improvement plan for the control rod actuator of Unit 3＆4 is formulated,in which the calculation and test method are used to demonstrate the possibility of the improvement.This may provide the effective reliability of the control rod actuator of Unit 3＆4 and lay the foundation for station operation economic index and enhance the market competitiveness.

Analysis and Improvement of Auxiliary Feedwater System Excessive Flow Based on 1D-3D Models

Zeng Chang, Lai Jianyong, Duan Yongqiang

2014, 35(S1): 93-96.

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This paper uses the k-εturbulence model to simulate the three-dimensional flow condition of auxiliary feedwater system（ASG） orifice,and obtains the characteristics of the velocity,pressure distribution and the relationship between flowrate and pressure drop.Subsequently,combined with the calculated orifice characteristics,the one-dimensional simulate model of ASG is established to analyze the alarm problem caused by the gas stripper excessive flow during the pre-service test,whereafter an improvement method for alarm signal delay is presented.The presented method can solve the problems effectively.

Design and Use of Short Handle Grab Tool for Irradiated Sample Plug

Zhou Wei, Qin Xiaoguang, Liu Pengqi, Zhao Qiang

2014, 35(S1): 97-99.

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Limitation and operation risk of the existing grab tool for grabbing the irradiated sample plugs in PWR power plant is analyzed,and the short handle grab tool for irradiated sample plug is developed.Its operation advantages and the supplementary to the function of the existing grab are introduced,to enhance the level of safety for extracting the irradiated sample plug during outage.

Passive Hydrogen System Used at Chashma Nuclear Power Plant Phase Ⅱ Project

LinG Xing

2014, 35(S1): 100-102.

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This paper briefly introduces the usage of passive hydrogen recombiner equipment at severe accident status in Chashma Nuclear Power Plant PhaseⅡ.It is combined with system design description and equipment fabrication to describe its items of work principle,design principle,site layout and reference and so on.In the meantime,system safety analysis as well as operation and maintenance have been introduced.

Analysis and Solution of Overpressure Issue of Main Steam Isolation Valve of Expansion Project

Mou Yang, Dong Juncheng, Lu Qi

2014, 35(S1): 103-106.

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The overpressure problem frequently occurred in the commissioning and operation of main steam isolation valve（MSIV） actuator of Qinshan Phase II expansion project.The analysis of the change of the nitrogen with the temperature found that the fundamental cause of the problem is as follows:there is no temperature compensation space for the accumulator plunger,and the hydraulic oil expanded with the change of temperature,and there is no space for accumulator to expand.A temporary scheme for overpressure problem is proposed,and a pressure relief device is designed and applied in real situation.The overpressure problem of MSIV actuator is solved successfully.

Research on Flexible Sludge Lancing Device for Steam Generator

Zhu Lihui, Liu Yulong, ZhOu Zheng, Liu Yan, Li Shiwei

2014, 35(S1): 107-109.

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The sludge deposited on the secondary side of stream generator（SG） tube sheet will become hard as the process of time,and it will degrade the tube.In order to lance the sludge directly and decrease the harm to the tube,a sort of SG sludge flexible lancing was researched.This paper describes the structure,principle,technology index and difficulty of the sludge flexible lancing.The application result shows that the lancing satisfies the flexible lancing requirement and ensures the cleanness of SG secondary side.

Study on Welding Technology for Domestic Manufacture of Fuel Assembly Top Nozzle

Li Sha

2014, 35(S1): 110-112.

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The welding performance and welding process of top nozzle in native is intruduced.Base on the analysis of a large number of the weldability test on the special material,such as deposited performance and uniformity test,welding base metal penetration performance test,and nozzle entity welding,the welding parameter satisfying the welding technology requirement and the welding technology qualification is obtained.The form of butt welded joint is without groove.Firstly the assembled nozzle tack welding is carried out manually,and then the weld is conducted without filler metal by robot.Repair welding of nozzle can be conducted by filler metal if necessary.The method of variable speed welding can be adopted to increase the welding speed in the plate spring stage and the ending stage.The results from the process appraisal using the new welding procedure satisfy the requirements of the technical conditions.

Improvement of Overpressure Protection in Cold Solid Condition for 1000MWe PWR NPP

OuYang Yong, Jiang Xiaohua, Lu Xianghui

2014, 35(S1): 113-116.

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An improvement project of overpressure protection in cold solid condition for 1000MWe PWR NPP is proposed,i.e.combined with the RRA safety valves,PSVs are used to provide dual overpressure protection by lowering the pressure setpoints for opening and closing.Normally,the overpressure protection is implemented by the RRA safety valves,while in case of RRA isolated,PSVs are used to provide back-up protection in place of the RRA safety valves.The analysis results show that PSVs can provide effective overpressure protection in cold solid condition,consequently ensure the integrity of RCP pressure boundary.

Analysis of Bolt Fracture Caused by Bypass Valve Vibration in Nuclear Power Plant

Zhang Guiying, Li Zhikai, Zhao Fuqiang

2014, 35(S1): 117-119.

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During the turbine startup process of one Nuclear Power Plant,the bypass valves vibrated severely,even to break all of the connecting bolts between the valve bodies and the oil-driven actuator.In this paper,a careful analysis of the unit load during severe vibration and the motion curve of the oil-driven actuator,together with the condition of the joint surfaces of the valve bodies and the spools after the split of the valves have been made.Moreover,the authors analyze the structures of the valves and the relative positions between spools and sleeves under different load conditions,and give the origin for the break of the connecting bolts,i.e.,the valve spool of 300kg weight moved back and forth in the pre-inlet travel length and hit the spool,and after 23 min.,the bolts reached the abruption limit and caused four connecting bolts to be broken.In this paper,the recommended modifications are provided.

Experience Feedback from Commissioning of AP1000 Project Nuclear Island Polar Crane

Yan Peifu, FeNG Lei, Li Kun, Zhang Hua, NA Hongwei, Tong Qiusheng, Sun Jingyi, Liu Jiahe, Gao Zhiqing

2014, 35(S1): 120-123.

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The paper introduces the work content,test method and the problems with the solutions of the nuclear island polar crane during the test in Zhejiang Sanmen and Shandong Haiyang nuclear projects,then make a summary of the advice on the improvement of the domestic manufacture of the advanced passive pressurized water reactor（AP1000） polar crane.It can be a reference for the domestic manufacture and commissioning of the future AP1000 project polar cranes.

Steam Hammer Calculation and Analysis in Main Steam System of PWR Nuclear Power Plants

Yu Pei, Li Changyue

2014, 35(S1): 124-127.

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Steam hammer pressure is solved though the simplified calculation.A mathematical model of main steam system in a million-kilowatt nuclear power plant is established using PIPENET software,and the model is from steam generator to main steam header.Steam hammer occurrence and reduction is simulated through transient calculating function.Parameters are calculated including maximum steam hammer pressure,maximum steam hammer load,the time of occurrence of the maximum steam hammer load and the position of occurrence of the maximum steam hammer load.The effect of straight pipe length and valve closing time on steam hammer is analyzed.The conclusion is that the shorter straight pipe is or the longer valve closing time is,the lower energy of steam hammer is.

Mathematical Model of 95/95 Criterion and Its Application in Thermal Design Procedures

Xu Liangwang, Lu Bo, Ye Jie

2014, 35(S1): 128-132.

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According to the requirement of regulations,the statistics model was established based on the 95/95 criterion and the calculation method of Owen’s factor（k） is derived as well,which was employed in developing the mathematical models of STDP/ITDP/RTDP,respectively.The procedures were also proved to be conservative and consistent with regulations through theoretical mathematical derivation.

Nuclear Safety Valve Components Analysis and Implementation of Seismic Qualification Test

Zhang Liqin, Qiu Jianwen, Wang Jiangbo

2014, 35(S1): 133-135.

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A series of safety-classified qualification tests are carried out on nuclear safety level K2,K3 class solenoid valve and limit switch developed by domestic manufacturers.Test condition and test requirements for anti-seismic experiments on the equipment to be qualified are introduced,and the universal device spectrum is selected for the anti-earthquake test based on the actual situation of the equipment to be qualified.In order to ensure the safety and reliability of the equipment operation during the earthquake,the load circuit of solenoid valve and the contact bounce monitoring devices of limit switch are developed,and operation test is conducted on the solenoid valve and the switch contacts chatter during the seismic tests is monitored.Test conditions and results fully meet the requirements.

Contrast Analysis of Welding Technical Between ASME and RCC

Wang Yong

2014, 35(S1): 136-137.

Abstract(12) PDF(1)

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ASME B＆PV Code Section Ⅸ Welding is contrasted and analyzed with RCC-M S.In welding technique evaluation,every welding method in ASME Section Ⅸ displays essential variables and nonessential variables.In general,the number of essential variables to be controlled in RCC-M welding technique evaluation is more than that in ASME.In fact,some specifications are beyond the range of welding technique evaluation.

Calculation Study on Stress Indices of Main Pipe Oblique Safety Injection Nozzle

Lu Xifeng, ai Honglei

2014, 35(S1): 138-141.

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A finite element calculation method for stress Indices based on stress Indices definition in RCC-M standard is presented.A finite element model for vertical branch pipe connection is established by the finite element analysis software ANSYS,to calculate the stress indices of branch pipe connection.The effectiveness of the calculation method for stress indices is verified by comparing with the stress indices of branch pipe connection specified in RCC-M.

Seismic Capability Walkdown in Seismic Margin Assessment of Nuclear Power Plant

Gong Zhenbang, CHeN Li

2014, 35(S1): 142-144.

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The seismic capability walkdown is very important in the seismic margin assessment of nuclear power plants.This paper discussed the problem that the walkdown can solve directly.Also this paper described the flow of walkdown,including the seismic capability preparatory work and walkdown process,and described the walkdown guidelines for several kinds of equipment.Lastly,the author gave the envision of achieving the walkdown with high efficiency.

Application of Equivalent Static Method and Spectrum Analysis Method in Equipment Seismic Analysis

Lan Qi, Hu Wenting

2014, 35(S1): 145-148.

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This paper presents how to use the equivalent static method and the spectrum analysis method to solve the seismic problem of the equipment,and analyzes the advantage and disadvantage of each method by comparing the results of the calculation,validates the feasibility of the two methods in the seismic analysis of pressure vessel,and provides convictive proof for how to choose the appropriate method to solve different seismic problems in the future.

Application of Solid Combining Shell Element in Residual Heat Removal Heat Exchanger Seismic Analysis

Liu Jiayi, Yu Shunli

2014, 35(S1): 149-151.

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Residual heat removal system is an important safety concerning system in PWR,and the detailed seismic analysis need to be performed for the major equipments in this system.For the residual heat removal heat exchanger in residual heat removal system,solid element and shell element was used for the model,multi-point spectral analysis was used for detailed stress analysis,and RCC-M code for results assessment,which showed that the method above is convenient and reasonable for analyzing the residual heat removal heat exchanger,and it can reduce the calculation time.

Application of Limit Analysis in Pipe Radial Stops of Nuclear Power Station

ZheNG Xiupeng, SHi Ji, Wang Yanping, SHeNG Feng

2014, 35(S1): 152-155.

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As a new technology in safety analysis of equipment,the limit analysis is an important plasticity failure analysis method in the area of engineering intensity design.The limit bearing capacity of pipe radial stops in nuclear power station is evaluated by adopting the limit analysis.The application of limit analysis in evaluations of equipment is illustrated in details by using of the above-mentioned example.The definitions and methods can be extended to elastic-plastically finite element analysis of other equipment in nuclear power stations.

Common Cause Failure Analysis of RHRS at PWR Based on GO-FLOW Methodology

Yang Jun, Yang Ming

2014, 35(S1): 156-160.

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In this paper,GO-FLOW method is applied to analyze the reliability of residual heat removal system at pressurized water reactor NPPs.The contribution to system unavailability caused by common cause failures is further calculated with β model,MGL model and α model.The results show that common cause failure has a significant effect on system reliability and GO-FLOW method can be effectively applied to system reliability analysis with common cause failures.

ATWS Accident Analysis for AP1000 Nuclear Power Plant

CHeN Wenhu, Cai Wei, Ge Zhenzhen

2014, 35(S1): 161-165.

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According to the analysis of all the anticipated transients without scram（ATWS） for AP1000 plant,it was determined that the loss of normal feedwater ATWS event is the limiting ATWS.Based on several sensitivity studies,the diverse actuation system（DAS） was improved as follow:low SG wide range water level setpoint implement the steam dump isolation and CMTs actuation,and then trip the RCP immediately.The final cases have been analyzed assuming the new DAS ATWS protection logic,and the results of analysis show that:considering the limiting MTC for the whole AP1000 plant life,the RCS peak pressure of LNFW ATWS meets the acceptance criteria with considerable margin.

Sensitivity Analysis of Physical Models of Reflood through RELAP5

Li Dong, Liu Xiaojing, Yang Yanhua

2014, 35(S1): 166-169.

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In order to achieve better understanding of reflood phenomenon and make improvement on the mainstream system code,this paper use RELAP5/MOD3.2 code to simulate the FEBA（Flooding Experiments with Blocked Arrays） facility from Germany.firstly,the calculation is compared with the experiment data.results show that both the peak cladding temperature and the bundle quenched time are under predicted.Then sensitivity analysis is done to the important physical model parameters related to the reflood phenomenon.The results prove that the interfacial friction of bubbles and droplets,heat coefficient between gas and interface of dispersed flow,wall to liquid drag coefficient,heat transfer of film boiling flow regime and the minimum droplet diameter are parameters with relatively large influence.More efforts should be made on the research of these parameters and models in the next step of work.

Seismic Margin Assessment for Refueling Water Storage Tank

Xu Xiaogang, Yu Shunli, Zhang Shuangwang

2014, 35(S1): 170-173.

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Seismic Margin Assessment was used to analyze the safety of nuclear power plant under the seismic of beyond design basis accident.The major failure modal of the refueling water storage tank is the failure of the anchor bolt in the method of EPRI SMA.Firstly,the seismic response was analyzed for the tank and the water.Secondly,the capacity is evaluated for the compressive buckling of the tank shell and the bolt hold-Down.Lastly,the Seismic Margin Earthquake was achieved with the assessment of overturning moment capacity and the sliding capacity.

Calculation Methods for Foundation Load for Earthquake

Yu Shunli, Liu Jiayi

2014, 35(S1): 174-176.

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Abstract:
In seismic analysis of equipment,the foundation load is an important parameter for the design of civil work.According to the different project fact,two methods for the calculation of the foundation load for earthquake are presented,i.e.,the simple method and the rigid region method.Calculation example is also presented.

Some Default OILs at Emergency Situation in AP1000 NPP

Zhang Pengfei, Zhang Lei, Xu Yanfeng, Yan Lili, Huang Jing, Li Zhangli

2014, 35(S1): 177-180.

Abstract(9) PDF(0)

Abstract:
Based on the formulae presented in IAEA-TECDOC-955 and the related article for operational intervention levels（OILs） at the emergency situation in nuclear power plants（NPPs）,and by RASCAL 4.2 computer code,this paper calculates the default OIL1 and OIL2 for all kinds of postulated severe accidents of an AP1000 NPP.The effects on OIL1 and OIL2 calculation results of related AP1000 design features are also discussed.

Quality Control for On-Site Welding of Self-Cleaning Type RCP Canopy Seals in AP1000 Nuclear Power Plant

Yang Qingyun, Wu Yong, Liu Xianwen

2014, 35(S1): 181-185.

Abstract(9) PDF(0)

Abstract:
AP1000 reactor coolant pump uses the shielding pump with high inertia flywheel,and its cleaning,lubrication,sealing and cooling function are realized by auxiliary impeller to circulate the cooling water.No repair is required in the design life of 60 years.Canopy seals weld,as the RCP installation welds between the moving parts and the pump shell,acts as the seal of the internal circulating cooling water of the RCP.The Canopy seals qualification of welding procedure,type tests,and welding process control of site installation,nondestructive testing of quality control are discussed,in order to ensure the safety,quality and progress of AP1000 RCP installation welding on site.

Design and Development of Special Tool for Installation of Seal Ring in Reactor Vessel

Yang Qingsong, Hao Zhonghang, Tang Chao

2014, 35(S1): 186-188.

Abstract(10) PDF(0)

Abstract:
The C-ring seal of reactor vessel is replaced manually during the refueling outage of Qinshan Nuclear Power Plant at the present stage.This paper analyzes the current process problems and risks.Special tools are designed to install the C-ring,and to improve the C-ring installation process.

Technology Improvement for Pipeline Anticorrosion of Sea Water Cooling System in Qinshan Nuclear Power Plant

Yang Guodong

2014, 35(S1): 189-191.

Abstract(7) PDF(0)

Abstract:
On the background of reconstruction of sea water cooling system pipeline,this paper analyzes the status of the anticorrosion of the sea water cooling water system pipe.Taking the generator air cooler water pipe with most serious corrosion as an example,the cause for the serious corrosion is analyzed,and a new technology for anticorrosion is developed,the improvement measures is proposed,through continuous improvement,and a good anticorrosion result is obtained.

Developing Features of Sandy Coast in Rizhao and Stability Analysis

Gu Hongqin, Yan Liwen, Ni Heng, Huang Haijun

2014, 35(S1): 192-195.

Abstract(6) PDF(0)

Abstract:
Based on the high-precision repeated observations on the representative cross-section and grain size analysis on surficial sediment,the short-term shoreline evolution dynamics in the winter in 2010 were studied.Combined with the topographic maps and remote sensing images,long-term shoreline evolution was also discussed.The results show that beach ridge-tidal creek is representative topographical feature.Beach ridges commonly develop between high tidal zone and middle tidal zone.Tidal creeks commonly develop between middle tidal zone and low tidal zone.The phenomenon that beach ridges swing landward is a typical signal of coastal erosion,which can be attributed to the strong impulse wave overtopping.The overall tendency that high tidal zone was eroded and low tidal zone is filled up can be clearly revealed in the winter in 2010.The coast is slowly eroded in short term and erosion pattern alters from drawing back to scouring vertically.

Enterprise Risk Management of Nuclear Power Enterprise

Xu Han, Zhang Zhihui

2014, 35(S1): 196-199.

Abstract(9) PDF(0)

Abstract:
At present,there is no such publication which has systematically discussed the nuclear power enterprise risk management in China,and there is no relatively mature and perfect system can be used for reference in the field of nuclear power in China.This paper analyzed the risk management problems of the Fukushima accident.Based on the actual condition of the nuclear power,this paper introduced the main work of enterprise risk management to carry out and the content of risk management in the nuclear power company,and also summarized the experience and gave some suggestions to the future work.

Necessity of Developing Nuclear Energy in China

Yang Chen, Fang Chao, Tong Jiejuan

2014, 35(S1): 200-202.

Abstract(7) PDF(0)

Abstract:
The study on nuclear energy development is summarized in the following four conclusions:the development of nuclear power can offset the electricity deficit of economic development;it is necessary to develop nuclear energy which is a kind of stable low-carbon energy considering CO₂ emission reduction;with a good foundation of industrial chain of steady progressing,nuclear development cannot be easily forsaken from sustainable development perspective;and the nuclear fuel is an important way of safeguarding energy security,promoting self-sufficiency of energy and replacing fossil energy import.Therefore it is very important to improve nuclear safety and the public acceptance to ensure that nuclear energy can be developed steadily.

Comprehensive Control of TOC in Demineralized Water at AP1000 Nuclear Power Plant

Liu Yongfeng, Zhang Huazheng

2014, 35(S1): 203-205.

Abstract(5) PDF(0)

Abstract:
AP1000 as the third generation of nuclear power units,requires its operation water to be below 50ppm of TOC as described in its Chemistry Manual.TOC contained in Demineralized Water（DW） is mainly the residual of the process of DW raw water pre-treatment.Another source of TOC is owing to decomposition of the organic materials in the system.When exposed to high temperature or radiation,TOC yields organic acid which will corrode metallic materials.TOC contained in DW can be effectively removed by operations of mixed condensing,reverse osmosis（RO）,ultra-violet ray radiation,and etc.

Study on Vacuum Exhaust Method and Device for Reactor Primary Loop of Pressurized Water Reactor Nuclear Power Plant

Qin Yuxin, Xiang Wenyuan, Dong Yachao, Yuan Jie, Liu Qingsong

2014, 35(S1): 206-210.

Abstract(11) PDF(0)

Abstract:
This paper introduces the vacuum exhaust method for the reactor primary loop of Pressurized Water Reactor（PWR） nuclear power plant.It describes the scheme design and operation process of the vacuum exhaust device that is constituted by the temporary cover with sealing ring and vacuum releaser.This is the first time in China for reactor primary loop to be vacuum exhausted during the low-low water level in reactor plant overhaul.It makes the dynamic draining air process not needed,and effectively shortens 10 hours at least in the critical path time of the overhaul,reduces the damage risk of RCP pump,improves the economy and security of reactor power plant operation.

2014 Vol. 35, No. S1