核动力工程

Requirement Description for R.G.1.20 Addenda ‘Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Test’

YAO Wei-da, ZHANG Ming, XIE Yong-cheng, LIANG Xing-yun, ZHANG Kai

2011, 32(S1): 1-3.

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Abstract:
The new revision of R.G.1.20 ’Comprehensive Vibration Assessment Program for Reactor Internals during Preoperational and Initial Startup Test’ was issued by NRC in 2007.In this paper,the back-ground for the new revision of R.G.1.20 is demonstrated.Major notes of addenda are stated in contrast with the former revision,including CFD analysis and supplementary analysis in flow-induced vibration analysis,and evaluation requirement for acoustic resonance problem of main steam system（pipe line,relief valve and dryer）.Some typical examples are presented for flow-induced vibration failure cause.It’s shown that the new comprehensive vibration assessment program is crucial for the safety operation and economics in nuclear power plants.

On Regulatory Requirements upon Seismic Qualification of Nuclear Safety Equipment

SUN Zao-zhan, LU Yan, ZHU Xiu-yun

2011, 32(S1): 4-8.

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U.S.NRC new regulatory requirements on equipment seismic qualification,mainly the more stringent restrictions on the use of empirical methods is analyzed.Suggestions on the establishment of relative norms of nuclear safety equipment qualification and nuclear safety regulatory requirements in China are proposed.The norm,standards and regulatory requirements should be amended timely with the development of technologies and the build-up of experiences,but fully considering the implementablility of the proposals during the amending.

Status and Development of R5 Procedure for Structural Integrity Assessment of High Temperature Components

LEI Yue-bao, GAO Zeng-liang

2011, 32(S1): 9-12.

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In order to support the continue and reliable operation and the equipment life extension of AGR nuclear power stations,the structural integrity assessment technologies for high temperature components have been studied and developed by British Energy since 1980s and compiled in the internal code R5 Assessment Procedure for the High Temperature Response of Structure,.This paper provides a brief introduction to R5 procedure and its history,technology adopted,current status and further development plans.

Study on Pipe Whip Analysis of High Energy Pipe Break

DING Kai, LI Gang, LIANG Bing-bing

2011, 32(S1): 13-17.

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Take the pipe whip analysis of main steam pipe and feed water pipes in AP1000 first bay as an example,the pipe layout is optimized to reduce the breaks and provide the necessary reference for the design of anti-whip limit.By deriving of the finite-difference theory based on energy balance and the dynamic analysis of the finite element,the methods for analyzing of the kinetic energy from the pipe whip is studied.Finally,the impacted structure is analyzed to decide if an anti-whip device shall be installed,and to demonstrate that the pipe whip will not cause unacceptable damage to the important structures such as the building of nuclear island.

Research on Creep and Fatigue Characteristics of MSIV at Elevated Temperature in HTR

WANG Hai-tao, WU Xin-xin

2011, 32(S1): 18-21.

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The main stream isolation valve（MSIV） is a key component at the secondary loop of the high temperature gas-cooled reactor（HTR）,and its structural integrity is essential to the safety of the reactor.In this paper,the creep and fatigue characteristics of the MSIV at elevated temperature is studied in accordance with Subsection NH of ASME BPV code Section III,Division 1.Both stresses and deformation of the MSIV subjected to a combination of the elevated temperature,high pressure and seismic load are analyzed,and the damage arising from both creep and fatigue at several critical locations of the MSIV is calculated.The sensitivity of the service life of MSIV to both stress types and levels is also explored.The results demonstrate that high temperature is a key factor that affect the structural integrity of MSIV,and that creep at high temperature dominates high temperature structural damages.

Fracture Mechanics Analysis of Embedded Cracks in Plates

WANG Wen-hua, LI Rong-sheng, GAO Zeng-liang, LEI Yue-bao

2011, 32(S1): 22-27.

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In this paper,the fracture behavior of plates with embedded off-set elliptical cracks under combined tension and bending is analyzed using 3-D elastic-plastic finite element（FE） method.The J-integral values obtained from the FE analyses for the crack tip at the smallest ligament are then compared with those estimated using the reference stress method.The results show that the reference stress method may underestimate the FE J when the global limit load is used.The reason could be that the plastic zone correction in the reference stress method is not sufficient to cover the plastic zone in the small ligament due to the ligament yielding.

Shakedown Analysis of Pressurizer Surge Line Based on Elasto-Plastic Theory

ZHANG Shi-wei, CHEN Xue-de

2011, 32(S1): 28-30.

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According to the RCC-M rule,1/4 scale model of the nozzle of pressurizer surge lines in Qinshan Phase II project is established,and the elastoplastic analysis is conducted.10 cycles of loads is calculated firstly,and then the response of pipe after which endured 100 cycles is calculated by extrapolation.The assessment results show that the plastic instability of pressurizer surge line will not occur under the most severe cyclic load of load case II.

Study on Fracture Toughness Prediction for A508-Ⅲ Steel Based on Master Curve Approach

FANG Ying, LI Hui, HUI Hu, HE Yin-biao, LI Pei-ning

2011, 32(S1): 31-34.

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In this study,tensile test and Charpy impact test of domestic A508-III steel are carried out in the ductile-brittle transition range,in which obtained the mechanical properties and temperature dependence of Charpy impact energy.Fracture toughness test temperature is determined through the correlation with characteristic temperature,at which Charpy impact energy is 28 J or 41 J.The Master Curve of 12.7 mm SE（B） is respectively from the single-temperature and multi-temperature method,and measured T₀ was valid.The result shows that the reference temperature T₀ derived from two methods is basically consistent.And the reference temperature T₀ of domestic A508-Ⅲ steel is about-63℃.

Overview of LWR Environment Assisted Fatigue Issues for Primary Components and Piping of NPPs

HE Yin-biao, CAO Ming, YAO Wei-da

2011, 32(S1): 35-39,97.

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The latest researches on LWR Environment Assisted Fatigue Issues are summarized in this paper.The requirements on EAF analysis for new NPPs of RG 1.207 published by USNRC are also introduced to discuss how to take the consideration of EAF for new NPPs.At the same time,new advice is put forward on incorporating the EAF into the fatigue analysis method of Section III of ASME B&PV Code.

Design Rules for Nuclear Components in Elevated Temperature

LI Xiao-tian, LUO Xiao-wei, HE Shu-yan

2011, 32(S1): 40-43.

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High-temperature is the common problem for Generation IV reactors.This paper reviews and assesses the currently available codes in the design of high-temperature components,such as ASME-NH,RCC-MR and procedure R5.ASME code is evaluated in-depth by comparing ASME-NB and ASME-NH,Also the application procedures of the ASME-NH rules are described in detail.

Comparative Analysis of Two Different Finite Models for SG Steam Outlet Nozzle Assembly

XIONG Guang-ming, DENG Xiao-yun, JIN Ting

2011, 32(S1): 44-47.

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The flow restrictor is simplified to the equivalent solid structure by the method described in ASME Appendix,and then the axisymmetry model and 3-D model are used for its solution.This paper com-pares the results from the two methods and their advantages and disadvantages in detail.It has been solved the SG steam outlet nozzle assembly,and furthermore demonstrated that the feasibility of simplifying the perforated structures to equivalent solid structures,and the reasonableness of 3-D model.

Study on Numerical Solution to Soil Dynamic Impedance for Nuclear Power Plants Based on SBFEM

ZHU Xiu-yun, LI Jian-bo, LIN Gao, HU Zhi-qiang

2011, 32(S1): 48-53.

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Based on the outstanding advantages of the scaled boundary finite element method（SBFEM） for the problem of infinite domain,an improved numerical solution to the soil dynamic impedance of the nu-clear power plants is presented in this paper,by utilizing a continued-fraction solution of SBFEM analysis,which is especially fit for the numerical simulation of non-homogeneous layered soil region,and also can give the soil dynamic impedance varying with the excitation frequencies.Through the numerical example of a rectangular base,the rationality of the continued-fraction solution algorithm was shown.Finally,mainly aimed at the solution to half-space homogeneous/layered site problems,some numerical results are given to validate its accuracy and applicability by comparing with the harmonic response analysis of viscoelastic soil dynamic model.

FE Simulation of Bolt-up Process for Press-Retraining Bolts of HTR-PM Pressure Vessel

ZHANG Tian-yi, WAN Li, SHI Li, ZHANG Chuan-yong

2011, 32(S1): 54-56.

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The reactor pressure vessel FE model with 42 M88 bolts is built in this study by ADVEN-TURECluster.These 42 bolts are divided into seven groups,and there are two steps in the bolt-up process.The simulation results show that every bolt-up step will make different influence to different groups of bolts.

Comparative Study on Probabilistic and Deterministic Methods for Stress Assessment of Reflector Graphite Bricks in HTR

ZHANG Zhen-sheng, SUN Li-bin, WANG Hai-tao

2011, 32(S1): 57-60.

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Stress assessment of graphite components in high-temperature reactors should be carried out in order to estimate their lifetime.The assessment is affected greatly by dispersion of graphite strength,which can be described by Weibull distribution.The relationship was studied between failure probability of probabilistic method and stress limit of deterministic method,while shape parameter m of 2-parameter Weibull distribution is set as 5,10 and 15 respectively.

Study on Simplified Stress Analysis Model of Steam Generator Tube in 10 MW High Temperature Gas Cooled Reactor

XU Yu, DONG Jian-ling

2011, 32(S1): 61-64.

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The steam generator tube of 10 MW high temperature gas cooled reactor in Tsinghua University adopts the spiral tube module structure with single-end and small bending radius,and the structure stress shall be ensured below the threshold under normal operation,accident and other operation conditions,to prevent the structure from failure caused by large deformation.This paper uses the finite element software ABAQUS to analyze the stress distribution of such tube under specific operation condition to obtain the stress distribution of spiral tubes of different cycles,to establish the simplified stress analysis model of such type tube,and prepare for the fracture analysis of vulnerable area.

Rigid Mode Correction and Program Development for Spectrum Analysis

WANG Yan-ping, ZHANG Shuang-wang, GONG Zhen-bang, WANG Chun-ming, LIU Shu-bin, TIAN Jin-mei

2011, 32(S1): 65-68.

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If the sum of equivalent masses of natural Frequencies below the cut-off frequency is lower than 90 % of the total mass,the rigid mode response needs to be taken into account.This paper studies the rigid mode correction theory.The second develop program to ANSYS10.0 is created by APDL and UIDL following this theory.The results from the verification example show the correction of the program.

Study on Numerical Simulation of Bursting Pressure of Steam Generator Tubes with Local Wall-Thinning

HUI Hu, LI Zhi-qiang, ZHANG Li-yan, JIAO Ming, LI Pei-ning

2011, 32(S1): 69-72.

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The local wall thinning of steam generator tubes due to corrosion and mechanical wear is the main reason decreasing the bearing capacity and the following rupture of SG tubes.Based on the theory of limit load,the numerical simulation method is used in this paper to estimate the bursting pressure of steam generator tubes.The effect of the volume defect dimensions in all directions on the burst pressure of the pipe is studied.

Study on Computational Method of Allowable Nozzle Loads for Nuclear Vessels

HUANG Qing, CHEN Meng, ZHAO Fei-yun, ZHANG Li-yan

2011, 32(S1): 73-75.

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The vessel-based method,with respect of the piping-based method,is one of the two methods generally used to obtain the nozzle loads for nuclear vessels.This paper studies the calculation and evaluation criteria of allowable nozzle loads for nuclear vessels of nuclear safety class 1 with this method.An example of a specific vessel has provided.

Analysis of Structural Seismic Response of Nuclear Power Plant Concrete Containment Vessel with Isolating Devices

HOU Gang-ling, CHEN Shu-hua, LI Dong-mei

2011, 32(S1): 76-79.

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In order to reduce the effect of seismic disaster on the nuclear power plant reinforced Concrete Containment Vessel（CCV）,the feasibility of the application of isolating technology in CCV is analyzed based on the CCV dynamic response characteristics and the principle of the structural isolating devices technology.Taking a nuclear power station as an example,the seismic responses of normal CCV with isolating CCV is compared and analyzed,and isolating design is conducted by optimized technology.The results show that the the isolating CCV can greatly improve anti-seismic capacity.

Seismic Analysis for Plant Structure of 5 MW Nuclear Heating Reactor

WEI Wei, ZHANG Zheng-ming

2011, 32(S1): 80-82,106.

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Seismic analysis of was carried out for the plant of 5 MW Nuclear heating reactor in Tsinghua University in order to check its safety properties.Finite analysis model was established using SAP2000 pro-gram,and the effect of soil on the structure was simplified.The internal forces of structural elements under seismic condition were calculated by the linear elastic method,and the actual bearing capacities of the structural elements were calculated using the method in the standards.Then the internal forces and the bearing capacities were compared to evaluate the plant safety.The results show that the plant frame structure is relatively weak,and some framing members would be disabled and part of the structure would collapse in a strong earthquake.These weak components need to be reinforced.

Safety Analysis of Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident

ZHOU Yun-qing, LIU Jia-zheng, ZHU Li-bing

2011, 32(S1): 83-86.

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Under seismic or LOCA condition,dynamic collision and impact will occur between the the adjacent fuel assemblies located in the reactor core.The object of safety analysis of fuel assembly for combined seismic and loss-of-coolant accident is to evaluate the performance of fuel assembly under limiting condition,and to demonstrate the satisfaction of related design criteria.The safety analysis is based on several fuel assembly models,including the detailed model,lumped mass model and impact model.According to the analysis results for a certain plant,the fuel assembly in this plant can meet all the design criteria related to fuel assembly performance under combined seismic and LOCA condition.The research of safety analysis method for fuel assembly under seismic and LOCA condition can provide theoretic guidance for the setup of fuel assembly safety evaluation system.

Dynamic Time-History Analysis of Relay Frame’s Reconstructed Support of Daya Bay Nuclear Power Plant

ZHANG Xiao-ling, SUN Lei, LI Tian-yong, LI Xi-hua

2011, 32(S1): 87-89.

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For obtaining the stress response accurately of the Relay Frame’s Reconstructed Support of Daya Bay Nuclear Power Plant（NNP） under the horizontal and vertical seismic load,dynamic time-history method is used.The method considers the relative displacement between the top support and the bottom sup-port,which is applied by the worst-case combination.From the analysis results,the output and input transfer function is obtained,which can verify the analysis results and supply as a reference for the improvement of the relay frame support.

Research of Seismic Analysis for Pressurizer of Nuclear Power Plants

YANG Neng-ren, QIN Jun-wei, LIU Pan

2011, 32(S1): 90-92,102.

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The pressurizer is a primary equipment of the reactor coolant system.Its main function is to control the pressure of the reactor coolant system.The pressurizer is classified Safety class 1,RCC-M class 1,Seismic class 1.The aseismatic capability of the pressurizer should be evaluated through two facets.One is that the nature vibration frequencies should be sufficiently removed from the excitation frequencies for which accelerations are significant.The other is that the stress of pressurizer under seismic load is lower than allowable stress.Seismic analysis of pressurizer is performed in normal operation condition and initial reactor start-up condition via finite element methods.

Simulation and Analysis of Main Steam Pipe Whip Based on LS-DYNA

WANG Chun-lin, SHE Jing-ce, CHU Jin-hua

2011, 32(S1): 93-97.

Abstract(25) PDF(1)

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In order to accurately simulate the collision process and obtain the overall distribution,the finite element model of main steam pipe and the U-bolt is established by LS-DYNA software,the collision process is simulated,and the U-bolt stress-strain distribution and time history curves is analyzed.The variation of the velocity and acceleration of the pipe is discussed and the detailed calculations of the U-bolt deformation and rejection are conducted.The result shows that the rejection of the finite element method is only about 36% of the rejection by kinetic analysis.

Nonlinear Dynamic Analysis of LOCA Based on Secondary Development of ANSYS

QI Huan-huan, CENG Zhong-xiu, ZHANG Yi-xiong, LIU Wen-jin, WANG Wei

2011, 32(S1): 98-102.

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In this paper,nonlinear dynamic response under loss of coolant accident（LOCA） transient in the nuclear reactor coolant system（RCS） is investigated with ANSYS program.Many nonlinear factors,such as different structural stiffness in tension and compression,gap and plastic strain caused by main pipes’ break etc.are considered.The Secondary development of ANSYS is performed to form the customized module for implementing effective parameterized and modular modeling and LOCA nonlinear analysis.Comparison of the results calculated by ANSYS and program-specific shows that the results are generally consistent,but there are some differences locally.According to the experience of transient dynamic analysis,and there are so many non-linear factors in the reactor coolant system,the difference is acceptable.The efficiency of this engineering analysis is improved remarkably with the convenience of input and modeling,viewable layout and automatic creating of reports,when using the customized module based on ANSYS.

Study on Assessment Method for Typical Plate and Shell-Support in Nuclear Power Plants

LIAO Kuai, LI Peng-zhou, CHEN Xue-de

2011, 32(S1): 103-106.

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The work of shock test of the support-model including model design,layout of the sensor,the test method and process,experimental results and discussion is presented in this paper.The experimental and the numerical results show that the nominal elastic stresses are much higher than those stipulated in the existing shock assessment codes of supports,which are supposed to be conservative.This preliminary study will provide some experimental and numerical data for modifying the present shock assessment criterions of supports.

Analysis of Evaluation Methods for NPP Piping System Vibration

HE Chao, YUAN Shao-bo, YU Dan-ping

2011, 32(S1): 107-109,124.

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According to the requirements of ASME OM-S/G-2000 Part 3,this paper presents the common principles of selecting the measuring points on the piping system by theoretical analysis and relevant experiences.Moreover for some specific piping systems,the allowable values of vibration displacement and velocity are derived,which can be used in assessing pipe vibration.It is helpful in the safety monitoring of the stresses caused by vibration.

Strength Analysis for Equipment Gate of Nuclear Power Plants

DU Kun, WU Gao-feng

2011, 32(S1): 110-112.

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In the operation condition and accident condition,the equipment gate in the nuclear power plants is closed and acts as the third shielding together with the reactor containment,to prevent the release of radioactive materials,and so it is one important part of containment pressure boundary.Based on the FEA method of ANSYS and RCC-M normative theory,a detailed stress calculation analysis is conducted for the equipment gate and its support components in nuclear power plants,and the result is in accordance with the RCC-M rule.The analysis method is reasonable and accepted by NNSA.

Simplification of Thermal Model for Control Rod Design Mechanism and Its Application in Fatigue Analysis

JIN Ting, ZHANG Qing-hong, XIONG Guang-ming

2011, 32(S1): 113-116.

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Thermal situation of control rod design mechanism is complex,and it is difficult to directly analyze the effect of various heat transfer mechanism on the thermal distribution.Taking one CRDM as an example,and based on 3 kinds of heat transfer theory,transient analysis is conducted by taking the equivalent convective heat transfer as the external heat transfer of CRDM and the equivalent heat conductivity as the internal heat transfer,to simplify the heat transfer mechanism,and then fatigue analysis is conducted by the rain-flow method.This method finds an effective way to solve the fatigue problem of complex thermal situation.

Stress Analysis and Evaluation of Heavy Double-Hole Pipe Clip

LI Hai-long, GAO Fu-hai, LI Nan

2011, 32(S1): 117-119.

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The material and structure of the heavy double-hole pipe clip used widely in the nuclear power plants are briefly introduced,and the finite elment model of the heavy double-hole pipe clip is built using ANSYS,to calculate the classified stress strength of double-hole pipe clip,the stress of bolt for pipe clip and the stress of the bolt joint.Using the computed results,the stress of the pipe clip is strictly evaluated according to the RCC-M codes and standards.The results show that the stress of pipe clip under all loads satisfies the code.

Analysis of Plastic Limit on Pipe Elbow under High Temperature and Pressure Based on ANSYS

LI Xing-hua, CAO Lei-sheng, NIE Lin-cheng

2011, 32(S1): 120-124.

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The plastic limit load of pipe elbows,not matter the failure model of the pipe elbows is plastic collapse or elastic-plastic fracture,is one necessary important parameter in the safety assessment.In the design of piping systems for nuclear power stations,the pressure and temperature is the work loads for pipe el-bows and also the main loads considered in the structural design of pipe elbows.Assuming the elbow material with ideal elastic-plastic behavior and considering the geometric nonlinearity,the distribution pattern of the plastic limit load in the pipe elbow under high temperature and pressure in the safety injection system is obtained using ANSYS.

Analysis of Outer Pressure Stability of Shell Part for Excess Letdown Heat Exchanger

LU Ai-min, NIE Lin-cheng, WANG Ming-yu, LIANG Ming-bang

2011, 32(S1): 125-126,133.

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The heat exchanger is nuclear equipment,and its shell part is a thin shell,which will endure the outer pressure 0.5 MPa in accident conditions.The calculation results of the structure buckling analysis by RCCM section ZIV is conservative.In this paper,the outer pressure buckling for the shell part of the heat exchanger is analyzed using ANSYS.The analysis results show that,comparing with the methods used in RCCM,the structure buckling analysis by finite element method is more accurate,and close to the project practice.

Optimal Design for Vibration Control of Cooling Pipe of Emergency Diesel Generator

ZHANG Kun, CUI Cheng-xin, QIAO Hong-wei, LIN Song, SUN Lei

2011, 32(S1): 127-130,144.

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During the operation of the emergency diesel generator in the Daya Bay nuclear power station,the vibration of the generator will cause strong vibration of the cooling pipes connected to it,which brings potential safety hazard to the normal operation of the generator.In order to discover the feasibility of the method to reduce the vibration level of the cooling pipe by attaching Dynamic Vibration Absorber（DVA） on it,the simulation and optimal design of the method was carried out.At first,the computational model of the cooling pipe was set up by finite element method.The response,which was measured from the experiment on site,was applied on the model of the cooling pipe as an excitation to compute the total response of the whole cooling pipe.Based on the result of the computation,the parameters of the DVA used to reduce the vibration of the pipe were determined.Then the optimal procedure was carried out to decide the best mount position of the DVA on the pipe,during which the vibration response of the key position of the cooling pipe was chosen as the optimal goal.The result of the simulation verified the validity of the vibration control method.

Stress Analysis of Reactor Pressure Vessel Support Rings

ZHANG Qing-hong, LIU Pan, SHANG Er-tao

2011, 32(S1): 131-133.

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The finite element software ANSYS11.0,a finite element methods codes,is used to simulate the mechanical characteristics of the Reactor Pressure Vessel（RPV） support rings under serious conditions.According to RCC-M code,the stress intensity is evaluated for the each part of the RPV support rings.The result shows that the RPV support rings satisfy the correlative criterion.This model is a more realistic simulation of the actual structure,the boundary conditions and applied loadings,effectively reducing the errors caused by model simplification.

Temperature Analysis Study on Core Shroud Based on Thermal-Fluid-Structure Coupling Effect

ZHAO Fei-yun, HUANG Qing, ZHU Kun, ZHANG Ming

2011, 32(S1): 134-136.

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In this paper,3-D FEM core shroud and core barrel are modeled based on thermal-fluid-structure coupling effect.The temperature analysis method are studied with 3-D thermal solid element,3-D thermal surface effect element and 3-D coupled thermal-fluid pipe element,which is expected to be useful for key technique study on reactor internals.

Equilibrium Iteration Algorithm for Analyzing Fluid-Structure Problems Using ALE Finite Element Method

LIU Li-ling, YI Li-qing, WEI Yong-tao

2011, 32(S1): 137-140.

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Based on the Galerkin/Least squares finite element method under arbitrary Lagrange-Euler description,the equilibrium iteration algorithm in which forces applied by fluids onto rigid bodies are treated as converging parameter is established for analyzing the fluid-solid interaction problems,and a new algorithm of mesh updating on the basis of pseudo elasticity method is proposed which can maintain fluid elements sides straight with mid-nodes coinciding with the center of element sides during mesh motion.The numerical example is about a mass-spring system vibrating freely in closed fluids which is analyzed with the method proposed,and the equivalent damping ratio,added-mass introduced by fluid-solid interaction are obtained as well as the streamlines of flow and pressure contours.

Analysis of Liquid Sloshing Characteristics in Two-Dimensional Horizontal Cylindrical Tank

LU Jun, YANG Yi-ren, LI Peng-zhou

2011, 32(S1): 141-144.

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Liquid sloshing of different proportional liquid-filled cylindrical tanks is investigated in this paper based on the Potential Flow Theory.The boundary value problem is solved by the Galekin method.Sloshing frequencies changed along with the liquid-filled proportions and radiuses are discussed in the free vibration,and the calculation results are compared with the experimental data,which have a good agreement.The result demonstrates that the method in this paper is right and effective.

Preliminary Analysis of Flow Induced Vibration for AP1000 Secondary Core Support Assembly

QIAN Hao, XIE Yong-cheng, ZHANG Ke-feng

2011, 32(S1): 145-148,158.

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The secondary core support assembly of Advanced Power Plant（AP1000） is modeled using the finite element methods（FEA）.Modal analysis and flow induced vibration（FIV） analysis without flow skirt for secondary core support assembly are performed.The natural frequency of AP1000 secondary core support assembly is smaller than that for Qinshan Phase I NPP with the same modal shape,which is produced by the combined effect of smaller stiffness and weight.Furthermore,the method of FIV analysis with fluid-structure coupling effect is studied.Three schemes that are full model,simplified shell model and single beam model for FIV analysis with fluid-structure coupling effect are proposed.

CFD Research on Normally Stagnant Non-Isolable Branch Lines Attached to RCS Loop Piping

QIN Jie

2011, 32(S1): 149-151,161.

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In this paper,ANSYS CFX program is used to perform Computational Fluid Dynamic analyses of normally stagnant non-isolable reactor coolant system branch lines.Take Safety Injection,out-of-service charging and Residual heat removal suction lines as example to make models for calculation.The potential exists for thermal stratification is discussed in the paper.The effects of location of valve leakage,the in-leakage rate,the temperature of leakage,the length of horizontal or vertical section and other factors are studied.The results verify that the different generic configurations have different mechanisms for thermal stratification.Thermal stratification in up-horizontal/horizontal configurations is caused by the interaction of swirl flow with in-leakage of cold water from a leaking normally closed valve;in down-horizontal configurations,the thermal stratification occurs due to the cyclic penetration and retreat of the swirl flow in the branch line,combined with heat transfer to the environment.

Cross Mobility Computation of Supporting Structures for Pumps

LIN Song, ZHANG Kun, SUN Lei

2011, 32(S1): 152-154.

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Mobility is one important topic in the research of structure vibration and noise.In this paper,cross mobility of supporting structures for pumps is calculated with finite element method（FEM） and statistical energy analysis（SEA）.The validity of predicting the mobility of supporting structures with FEM and SEA is evaluated by comparing the values from experiments,and it shows that this method is effective in the prediction of mobility characteristics.

Discussion on Flow Induced Vibration Analysis Method for Shell-and-Tube Heat Exchanger with Nuclear Safety Class 2 and 3

LANG Hong-fang, YE Quan-liu, CHEN Fu-long

2011, 32(S1): 155-158.

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Corresponding to the real operation condition of the shell-and-tube heat exchanger with nu-clear safety class 2 and 3,i.e.the water-water heat transfer without phase changing,and according to TEMA standard,the possibility and criteria of the flow induced vibration by vortex shedding,turbulent buffeting,fluid-elastic instability and acoustic vibration are discussed,and the flow induced vibration analysis method for this issue,as well as the solution for the design with the potentially existing dangers of the flow induced vibration damage are presented.

AP1000 Equipment Qualification Requirements and Comparisons to Existing Domestic Qualification Capabilities

XIE Yong-cheng, WANG Chi-hu, DOU Yi-kang

2011, 32(S1): 159-161.

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The AP1000 Equipment Qualification（EQ） requirements are briefed.Among them,the seismic and Loss of Coolant Accident（LOCA） qualification requirements are emphasized.The comparisons be-tween the AP1000 EQ requirements and capabilities of primary domestic test laboratories are addressed,focusing on the gaps of existing test capabilities to the AP1000 EQ requirements.

Application of Metallographic Analysis Techniques in Nuclear Components Failure Analysis

SUN Hai-tao

2011, 32(S1): 162-165.

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Failure analysis of nuclear mechanical components is an important means to deal with the in-service defects and guarantee safety and operating functions,in which metallographic analysis techniques including optical microscope,microhardness,metallurgic replica technique,scanning electron microscope,transmission electron miscroscope and electron probe have been widely applied.Mechanism of these techniques in combination with the application in nuclear cases study has been discussed in this paper.

Thermal Output Test of High Temperature Strain Gauge

LIU Zi-cai, YU Dan-ping, LU Yan-yan, CONG Bin, LI Xi-hua

2011, 32(S1): 166-168.

Abstract(14) PDF(0)

Abstract:
The aim of this test is to acquire accurate thermal output of high temperature strain gauges,to stabilize the thermal output performance of high temperature strain gauge,to improve measurement precision of thermal stresses with resistive high temperature strain gauge and to suggest a method of clamping instead of welding for fixture of high temperature strain gauge for thermal output measurement.Thermal output parameters including parameters like thermal output of temperature rise and drop,thermal divergence,and thermal hysteresis are measured for resistive type strain gauge with an apparatus for measurement of parameters of high temperature strain gauges.Special clamp which uses clamping instead of welding was utilized to test the thermal output so that each strain gauge could be used again with a more stable performance,better precision and lower cost for engineering projects.The conclusion was that the thermal output of high temperature strain gauge was obtained,and the method of clamping instead of welding for thermal output test is feasible.

Study on Failure Assessment Curves for Safe End of Nuclear Pressure Vessels

LIU Zhi-wei, WANG Guo-zhen, XUAN Fu-zhen, LIU Zhang-jun, TU Shan-dong

2011, 32(S1): 169-172,178.

Abstract(18) PDF(0)

Abstract:
In this paper,three-dimensional finite element analysis models are built for the welded safe end of typical nuclear pressure vessels.Through finite element analysis（FEA） of fracture mechanics for the circumferential part-through surface cracks with different sizes in the weld metal,the accurate failure assessment curves（FAC） are constructed according to R6 option 3.The FACs are related to the complex structure of the safe end,the geometry of welding groove,the complex materials at weld joint region of the dissimilar metal and the crack sizes.The FEA results show that with the increase of depth and length of the cracks,the FACs gradually shift down and are close to the R6 option 1 curve,and the FAC of the largest size crack shift down below the option 1 curve.It is concluded that the option 1 curve is overly conservative for shallow and short cracks and non conservative for deep and long cracks.It is suggested that the accurate FACs which are related to crack sizes should be selected for the defect assessment of the safe end.The mechanism that how the crack sizes affect the FACs is also analyzed in this paper.

Study on Mechanical Behavior of Expanding Tube-to-Tubesheet and Joint Performance for Steam Generator of Nuclear Power

YAN Zong-bao, WANG Guo-zhen, XUAN Fu-zhen, LIU Zhang-jun, TU Shan-dong

2011, 32(S1): 173-178.

Abstract(24) PDF(0)

Abstract:
The expansion and pull-out mechanics processes of the tube-to-tubesheet joints are numerically simulated by using three-dimensional finite element method（FEM） at different expansion pressures.Distributions of the residual contact pressures along axial and hoop directions,and the effect of the expanding pressure on residual contact pressure,pull-out force and wall reduction are obtained.The results show that there are two high residual contact pressure bands（sealing circular-band pressures） near the two surfaces of the tubesheet,and there is a steady residual pressure distribution along the axial direction at central part.The maximum tensile residual stress existed on the inner surface of tube at the transition zone of expansion joints,and which is assumed to be as the main factor for causing the stress corrosion cracks.With increasing the expansion pressure from 240 MPa to 330 MPa,the residual contact pressure,pull-out force and wall reduction increase.The formulas for relating the residual contact pressure and expansion pressure,and the pull-out force and expansion pressure are obtained by analyzing the FEM calculation results.The relationships between the expansion pressure and the joint performance and service reliability are also discussed.

Measurement and Evaluation of Nuclear-Class Pipe Vibration in Nuclear Power Plant Commissioning

YAN Jun-ming, YUAN Shao-bo, SHI Qing-feng, CHEN Can, XU Wei-zu

2011, 32(S1): 179-181.

Abstract(16) PDF(0)

Abstract:
According to the requirement of American Society of Mechanical Engineers（ASME） criterion,the vibration of nuclear-class pipes in nuclear power plants shall be measured in preoperational and initial start-up stage（during decommissioning）.This paper takes the unit three of Qinshan Phase II nuclear power plant as an example,and describes the nuclear-class pipe vibration test flow,test objects,test condition and test result evaluation in detail.

Experimental Study on Dynamic Collision of Graphite Bricks

WANG Hong-tao, SUN Li-bin, WANG Hai-tao, SHI Li, MA Shao-peng

2011, 32(S1): 182-184.

Abstract(19) PDF(0)

Abstract:
Under the seismic load or other loads,the collision between graphite bricks which are used as structural components in high temperature gas-cooled reactors（HTGRs） will happen.In order to study the collision behavior of graphite bricks,a collision test apparatus of two large specimens was achieved by a track-type device,and the velocity,the coefficient of restitution and contact time were measured by a high-speed image capturing system and the optical experiment measurement.The relations between kinetic parameters and collision velocities were obtained.The experimental results show that the coefficient of restitution was increased with the rise of the collision velocity,but the contact time was decreased with the rise of the collision velocity.

Experimental Study on Fracture Toughness of Nuclear Graphite

SHI Li, WANG Hong-tao, WANG Hai-tao, SUN Li-bin, HU Yu-qin, YAO Xue-feng, XIONG Chao

2011, 32(S1): 185-188.

Abstract(20) PDF(0)

Abstract:
Nuclear graphite is a brittle,polygranular material,which brings difficulties to the research of its fracture toughness.In this paper,an experimental work on measuring the fracture toughness of graphite IG11 is described.Three-point bending tests are employed.The effects of graphite specimen size,thickness,and the relative crack length a/W are studied in the tests.The deformation and stain at the crack tips are also measured.The results indicate that the fracture toughness of IG11 graphite is in a range of 0.82~1.27 MPa·m^1/2.The fracture toughness increases with W of the graphite specimens,but decreases with S of the graphite specimens.The B of the graphite specimens has little effect on the fracture toughness.

Overview of LBB Analysis Technologies for High Power Lines in Nuclear Power Plants

LI Qiang, CEN Peng, ZHEN Hong-dong

2011, 32(S1): 189-191.

Abstract(17) PDF(0)

Abstract:
The LBB（leak-before-break） technology has been obtained extensive application in designing the high power lines in nuclear power plants since early 1970s.It plays an important role in simplifying the design of nuclear power plants,making them less expensive,convenient for in-service inspection and decreasing the total radioactive exposure.The background and current status is reviewed,and the processes to apply the LBB technology are summarized.Suggestion on the future research is presented.

Research on Mechanical Properties of ODS Ferritic Steel by Small Punch Test

QIAN Xin, LU Dao-gang, MA Yan

2011, 32(S1): 192-196,200.

Abstract(21) PDF(0)

Abstract:
Test of the mechanical properties of ODS ferritic steel is conducted by small punch test（SPT）,which is one of candidate materials for supercritical water reactor（SCWR）.Effect of the sample thickness,loading rate and test temperature on the load-displacement curves is analyzed,and the suitable test parameters for ODS ferritic steel is identified.The yield load and yield strength are fitted,which are tested respectively by SPT and conventional test.The result shows that SPT can test the strength of ODS ferritic steel.

Acoustic Analysis of Pump Induced Pressure Pulsations on Primary Loop

CAI Kun, ZHOU Ying, HUA Yu-chao

2011, 32(S1): 197-200.

Abstract(14) PDF(0)

Abstract:
The pulsations in the primary loop have been studied by acoustic analysis.The results calculated by acoustic analysis code and theories have been compared to verify the feasibility of acoustic analysis code.Primary loop is simulated by ANSYS using simplified model.According to the operation of reactor coolant pump,a harmonic analysis has been finished.The result provides the pump-induced loads for the fatigue analysis of main components.It can be used to optimize the design of primary loop.

Effect of Fabrication Error of Reactor Pressure Vessel on Its Load-Bearing Ability

TAN Xiao-hui

2011, 32(S1): 201-204.

Abstract(23) PDF(0)

Abstract:
Using reactor pressure vessel（RPV） bottom junction with core shell as an example,the effect of fabrication dimension error on RPV load-bearing ability is analyzed.The load-bearing ability of the RPV can be evaluated by applying ANSYS and elastic analysis method to carry out stress analysis on RPV bottom junction with core shell and then comparing the stress analysis results and corresponding RCC-M rules.Fol-lowing the above-mentioned procedure,the stress analysis and evaluation on RPV bottom junction with core shell are carried out for 6 cases.In the first case,the dimension of RPV is the same as that given in the design.In other 5 cases,the thickness of the thinnest transition part of the RPV is reduced by 3%,6%,10%,20%,and 30%,respectively.Analysis results indicate that the load-bearing ability of RPV bottom junction with core shell meets the standards of RCC-M rules if the reduction of the thickness does not exceed 30% of its original dimension.

2011 Vol. 32, No. S1