最新刊期
卷 0 ， 期 0 ， 2020
- 摘要：Background: Spent fuel racks are important facilities to store the spent fuel which are freestanding in the spent fuel pool.The response of racks to seismic load is highly nonlinear and involves a complex combination of motions: sliding,impact,twisting and turning.Purpose: An analysis method should be built to accurately replicate these nonlinear responses.Methods: The whole pool multi-rack FEA model was developed and time history analysis was performed which contains the consideration of effect of sliding,impact and friction and the fluid structure interaction effect.Results: The analysis results such as displacement and force under seismic loads were obtained.Conclusions: The method can be used to the seismic analysis for spent fuel racks.关键词：Spent fuel rack;Nonlinear;Time history analysis
- 12
- |
- 195
- |
- 7
发布时间：2021-06-03 - 摘要：Background: For the nuclear power plant main component under design phase,the anti-seismic capability analysis should be evaluation used response spectrum analysis or time history analysis usually.Purpose: This paper attempts to get the non-linear influence because of gaps.Methods: Based on ANSYS FEM software,get the CRDM seismic result used improve spectrum analysis,which compare with the time history analysis.Results: the bending moments and shear force on each sections of CRDM housing was showed in this paper.Conclusions: The result shows that the improve spectrum analysis can get the structure dynamic characteristics,the calculation results between the improve spectrum analysis and the time history analysis are consistent,it can provide guidance on the subsequent equipment design.关键词：Response spectrum analysis;Time history analysis;ANSYS
- 14
- |
- 233
- |
- 5
发布时间：2021-06-03 - 摘要：Background: The steam generator（SG） serves as the primary means for removing the heat generated within the reactor core and is part of the reactor coolant system（RCS） pressure boundary.Purpose: Seismic analysis is required for SG,whose seismic category is Cat.I.Methods: The analysis model of SG is created with moisture separator assembly and tube bundle assembly herein.The seismic analysis is performed with RCS pipe and Reactor Pressure Vessel（RPV）.Results: The seismic stress results of SG are obtained.In addition,parameter sensitivities of seismic analysis results are studied,such as the effect of another SG,support,anti-vibration bars（AVBs）,and so on.Our results show that seismic results are sensitive to support and AVBs setting.Conclusions: The guidance and comments on these parameters are summarized for equipment design and analysis,which should be focused on in future new type NPP SG’s research and design.关键词：Steam generator;Seismic analysis;Parameter sensitivity
- 14
- |
- 96
- |
- 2
发布时间：2021-06-03 - 摘要：Background: Reactors with graphite core structure include production reactor,water-cooled graphite reactor,gas-cooled reactor,high-temperature gas-cooled reactor and so on.Multi-body graphite core structure has nonlinear response under seismic excitation,which is different from the response of general civil structure,metal connection structure or bolted structure.Purpose: In order to provide references for the designing and construction of HTR-PM.This paper reviews the history of reactor seismic research evaluation from certain countries,and summarizes the research methods and research results.Methods: By comparing the methods adopted in different gas-cooled reactor cores,inspiration for our own HTR seismic research was achieved.Results & Conclusions: In this paper,the research ideas of graphite core seismic during the process of designing,constructing and operating HTR-10 are expounded.Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced.关键词：Graphite core;HTGR;HTR-PM;Seismic research
- 17
- |
- 231
- |
- 2
发布时间：2021-06-03 - 摘要：Background: The electric heaters can regulate the pressure of the reactor coolant system,and are pressure boundary of the reactor coolant.The clad and sleeve of the heater are classified RCC-M class 1 and Seismic class 1I.Purpose: Model simplification and seismic analysis method of the heater are studied to support the structure design of the heater.Methods: The water around the electric heater is processed equivalently based on related theory.According to features of structure and support the finite element model of the heater is made by beam elements through finite element software ANSYS.Modal analysis and spectrum analysis for the heater are performed.Results: Analysis results are:（1） The nature vibration frequencies of the heater are respectively beyond of the frequency range which give significant values of acceleration.（2） During the SSE earthquake condition the structural strength of the heater meets the requirement.Conclusions: The seismic capability of the electric heater can be evaluated through analyzing its frequencies and stress intensity.关键词：Electric heaters;Model simplifying;Modal analysis;Spectrum analysis
- 11
- |
- 75
- |
- 2
发布时间：2021-06-03 - 摘要：Background: Pressurized water Reactor Pressure Vessel（RPV） sealing capability is one of the key factors to ensure the safety operation of nuclear power plants.Purpose: In order to explore the numerical simulation technology of the RPV sealing behaviors.Methods: In this paper,the 3-D thermal elastoplastic finite model is established,which based on the CPR1000 project.Results: The load and its combination of equipment during operation are considered,then the separation of RPV upper and lower flanges,radial slip and bolt loads etc.are also obtained.Conclusions: The results show the 3-D thermal elastoplastic finite model can simulate the sealing performance of the structure.关键词：Reactor pressure vessel;Sealing Behavior;Numerical simulation;Analysis technology
- 13
- |
- 267
- |
- 17
发布时间：2021-06-03 - 摘要：Background: There is no clear requirement on the rod drop performance of Control Rod Drive Mechanism（CRDM） in seismic condition.Purpose: Acceptance criteria of AP1000 CRDM seismic test need to be determined.Methods: Related regulations and the safety function of AP1000 CRDM are investigated,as well as the conclusions drawn from the CRDM seismic tests worldwide.Results: Acceptance criteria of this test should be in accordance with the limit is in AP1000 Nuclear Plant Safety Analysis Report.Conclusions: Drop time of control rods in AP1000 CRDM seismic test at the room temperature without flow is 2.7 s before and after Safe Shutdown Earthquake（SSE）.关键词：AP1000;CRDM;Seismic test;Acceptance criteria
- 13
- |
- 169
- |
- 3
发布时间：2021-06-03 - 摘要：Background: Hold down spring is a structural component that supports the reactor pressure vessel.It is located between the flange and barrel components,and has a function that transfers the force caused by reactor pressure vessel cap to barrel flange.Purpose: This paper attempts to have an analysis calculation and comparison of AP1000 and LPP Hold down spring.Methods: This article inherited Qinshan nuclear power hold down spring analysis method,use analytical method and FEM method to solve the problem.Results: The stress and deflection of HDS are calculated.Since the friction coefficient may have a large effect,the results of different friction coefficients are also analyzed.Conclusions: This paper compared the Qinshan,AP1000 and LPP HDS design advantages and disadvantages and proposed two kinds of new design,expected to have a reference value for follow-up engineers.关键词：Hold down spring;Reactor internal;Transform;Strength;Optimization
- 11
- |
- 108
- |
- 3
发布时间：2021-06-03 - 摘要：Background: Seismic test analysis of a cabinet in 18-5 citical device building is requried for safety inspection of zero power reactor,however,the floor response spectra at the height of 3.5 m where the cabinet is located are lacking.Purpose: For seismic test analysis of the cabinet in 18-5 citical device building,dynamic analysis is performed for the building,with the purpose to obtain the floor response spectra at the height of 3.5 m where the cabinet is located.Methods: The displacement time history and acceleration time history on the given height floor are calculated by using transient dynamic time history method with the software ANSYS under the load of foundation displacement time history,then acceleration response spectrum on the floor that the cabinet locates is calculated with the corresponding acceleration time history obtained on the cabinet floor.Results: The design floor response spectra under load OBE and SSE at the height of 3.5 m are offered with damping ratios of 2%,4%,5% and 7%,respectively.Conclusions: The design floor response spectra are credible and have been adopted by test analysis of the cabinet.关键词：Floor response spectra;Displacement time history;Acceleration response spectrum;Operation basis earthquake;Safe shutdown earthquake
- 12
- |
- 71
- |
- 1
发布时间：2021-06-03 - 摘要：Background: In the framework of lumped-parameter numerical models,the traditional soil dynamic impedance models,recommended by the main international seismic design codes of the nuclear power plant,are only expressed by a single parallel connection system of spring and damper.Purpose: In order to combine the recent development of soil-structure interaction analysis.Methods: Comparative study of both the direct method and substructure method history analysis is carried out based on the lumped parameter models recommended by seismic design codes of ASCE4-98,RCCG which are both applicable to the homogeneous site and also massless foundation model and viscous-spring artificial boundary model of especially fit for the numerical simulation of non-homogeneous site.Results: Finally,by taking the analysis of FRS for a certain CPR1000 reactor plant as an example,comparison analyses of homogeneous and inhomogeneous layer site conditions for various soil dynamic numerical models above-mentioned are carried out.In addition,the calculated results are compared to that of SASSI program.It can be concluded that the calculation results are similar based on different methods.Conclusions: This paper validates the applicability of different soil dynamic numerical models mentioned above and will provide some reference and guidance significance in the aspect of evaluation the seismic suitable for the site of nuclear power plant.关键词：Floor response spectra（FRS）;Soil-structure interaction（SSI）;Viscous-spring artificial boundary model;Lumped parameter models;SASSI program
- 13
- |
- 332
- |
- 27
发布时间：2021-06-03 - 摘要：Background: The seismic test is one of Plant Equipment Seismic Qualification Methodologies.Purpose: To satisfy the special requirements of AP1000 seismic qualification,this test is not same to traditional seismic test.Methods: AP1000 nuclear plant equipment qualification has new methods and requirements.The special requirements as response spectrum,accelerometer installation and functional test are described for AP1000 MCR seismic test.Results: The test results are demonstrated that MCR assemblies are satisfied as AP1000 seismic qualification requirements.Conclusions: These requirements are beneficial to the qualification for structural integrity and functional safety.They are also used to find design margin.关键词：AP1000;Equipment qualification;Seismic test;Specific requirements
- 15
- |
- 118
- |
- 1
发布时间：2021-06-03 - 摘要：Background: Piping is an important part in nuclear power plants.Purpose: In order to make piping meet the specification requirements,it is necessary to analyze and evaluate piping stress.Methods: This paper deals with the stress analysis and evaluation of the class 1 and class 2 piping of improved second-generation nuclear power plants,and discusses the calculation methods of piping stress due to various loads.Also,the effects of changing the RCC-M code edition on the calculation of piping stress are summarized.Results: Taking the LingAo Nuclear Power Plant Phase II engineering safety injection system as an example,the piping stress analysis and evaluation are performed,and the results meet the RCC-M code requirements.Besides,the interface parameters including the support loads,nozzle loads,displacements of the piping and so on are obtained.Conclusions: This paper will provide technical support for stress analysis and evaluation of improved second-generation nuclear power plant piping systems.关键词：Piping;Stress analysis and evaluation;SYSPIPE
- 11
- |
- 113
- |
- 5
发布时间：2021-06-03 - 摘要：Background: The main control room consoles have functional requirements,so the models must be validated to simulate the real physical properties,especially the dynamic properties.Purpose: In order to described the typical verification process used in the main control room consoles seismic qualification.Methods: By comparing with the test results,parameters of the model should be modified.Considering the verification of 1E electronic equipment in consoles,the time histories from analysis should be close to those from test.Results: By means of the modification,the model can match the test in mass,both integral and local modals,and the spectrum around electric component.Conclusions: The verified SRO model is able to represent the test unit in dynamic properties and it is reliable to be used for the further analysis.关键词：Equipment qualification;Test and analysis;Model verification
- 11
- |
- 59
- |
- 0
发布时间：2021-06-03 - 摘要：Background: AP1000 is a advanced passive PWR which designed by the Westinghouse.Purpose: The purpose of the seismic qualification test is to obtain test data that will demonstrate structural integrity and functionality of AP1000 MCR equipment during and after exposure to the dynamic effect of seismic load.Methods: The test was performed at the Nuclear Equipment Qualification Center（NEQ） of Nuclear Power Institute of China（NPIC）.The dynamic characteristics of specimens are detected by sine sweep and random input respectively.Multi-frequency method is applied along three orthogonal axis of the specimen in seismic qualification tests.And the seismic qualification tests are taken by using the acceleration on surface of the shake table as the control signal.Results: The structural integrity and the operability of the specimen are perfect during and after the seismic qualification test.Conclusions: Therefore,the AP1000 MCR equipments（PDSP and SDSP） pass the seismic qualification test.关键词：AP1000;MCR equipment;Shake table;Seismic qualification test
- 14
- |
- 143
- |
- 6
发布时间：2021-06-03 - 摘要：Background: It is very important to ensure the integrity of the reactor pressure vessel under core melt down accident.The high-temperature creep failure is the main failure mode of the reactor pressure vessel under core melt down accident.Purpose: This paper is to present an overview of research status and progress on high-temperature creep behavior of reactor pressure vessel considering the hypothetical core melt down scenario.Methods: Emphasis is placed on accomplished achievements in creep tests,scale model experiments and numerical simulation,and the domestic newly research productions on high-temperature creep behavior of reactor pressure vessel structure integrity.Conclusions: This paper also discusses the limitations of existing researches and indicates future research directions,such as multi-axis tensile tests,analysis of three-dimensional coupling temperature field,scaled model tests,and so on.关键词：Reactor pressure vessel;Integrity;Core melt down;High-temperature creep
- 14
- |
- 330
- |
- 15
发布时间：2021-06-03 - 摘要：Background: The drop time of control rod assembly is one of the most important parameters to ensure the safe operation of nuclear power plants.Due to the fluid-structure interaction（FSI）,the elastic structures,such as control rods and guide tubes,will vibrate in the dropping of control rod assembly.The impact and friction between the control rod and guide tubes caused by large transverse vibration will influence the drop time calculation.Purpose: To study in detail the flow-induced vibration and the friction,this paper focus on the fluid-structure interaction in the control rod drop process.Methods: Firstly,the vibration equations of control rod and guide tubes considering the fluid-structure interaction are established.Then the various fluid forces are analyzed in accordance with their qualities,and the influences of different guide tubes in a guide tubes array are also considered.Results: The friction between control rod and guide tube is not zero,and the friction under seismic condition is larger.Conclusions: The analysis on the fluid and structure interaction presented in this paper is reasonable and can improve current analytical models of control rod drop time calculation.关键词：Control rod assembly;Dropping process;Impact;FSI;Guide tubes array
- 16
- |
- 211
- |
- 5
发布时间：2021-06-03 - 摘要：Background: With the development of fracture mechanics,Leak-Before-Break（LBB） is widely used in nuclear power plant piping design.Purpose: In order to support the application of LBB,leak rate through crack need to be calculated.Methods: In this text,an analytical flow model is developed based on homogeneous non-equilibrium model,a computer program is also developed based on this analytical model.Results: Comparison between the results from the above program and test results shows that deviations of calc.results to test results are within ±50%.Conclusions: Conclusions can be got that this analytical model and computer program meet the requirement of engineering application.关键词：Leak-Before-Break;Homogeneous non-equilibrium model;Computer program
- 13
- |
- 124
- |
- 5
发布时间：2021-06-03 - 摘要：Background: During the process of Leak Before Break（LBB） design for the pipes in nuclear power plant,a through-wall crack is hypothesized in each selected location of the candidate pipe line,then the Crack-Opening Displacement（COD） and leak rate of the crack under normal operation condition is predicted,finally the applicability of the LBB design criterion is ascertained.Purpose: Discuss the disadvantages of the COD and leak rate calculation methodologies,develop a COD and leak rate calculation program.Methods: Through compare the results from the program developed by ourselves with PICEP,the program developed by ourselves is verified.Results: The results from the program developed by ourselves tally with PICEP.Conclusions: The disadvantages of each COD and leak rate methodologies are discussed;then a program was successfully developed.Both can provide some reference for the application of the LBB approach to the nuclear power stations of China.关键词：LBB;COD;Leak rate;Program development
- 12
- |
- 201
- |
- 6
发布时间：2021-06-03 - 摘要：Background: In the design of AP1000,Reactor Coolant Pump（RCP） is welded directly to the Steam Generator（SG） channel head.But this design makes the flow field in the SG channel head and RCP inlet complex and there may be vortex in this flow field for which reason the SG outlet resistance will increase and affect the long-term steady operation.Purpose: For this issue,some companies made tests on it.But the cost of test is high and the applicability of the test result is limited.If the parameters or components size changed a little,the test result will be no longer applicable.This article aims to find a way to solve this problem.Methods: This article considers using PumpLinx,a 3-D CFD flow field analysis software to analyse the SG and RCP coupled flow field.First,Pro/ENGINEER is applied to establish the 3D model of SG channel head with two RCPs welded to it and then establish mesh in PumpLinx and then analyze the problem in PumpLinx with Dynamic Mesh method.Results: The flow filed varying condition of SG-RCP coupled part from RCP rotation beginning to steady operation and the vortex condition after steady operation are obtained so as to support plant design.Conclusions: PumpLinx is suitable for AP1000 reactor coolant pump flow field CFD analysis.关键词：Centrifugal pump;PumpLinx
- 14
- |
- 349
- |
- 9
发布时间：2021-06-03 - 摘要：Background: In recent years some reactors have experienced significant steam dryer cracking.Studies indicate that the fatigue is due to flow-induced acoustic vibration.Purpose: In this study,sound source behaviors of a single valve on the main steam line are investigated.Methods: Mode and CFD analysis were carried to investigate the acoustic mode and resonance behaviors.Results: The following results can be concluded from this study: various categories of acoustic modes exist in the structure;geometries directly influence acoustic mode;peak excitation occurs around St of some specifically range.Conclusions: The results match theory and experimental studies well,and provide reference for the related researches.关键词：Valve resonance cavity;Acoustic resonance;Acoustic mode;Large eddy simulation model（LES Model）;Strouhal number（St）
- 13
- |
- 104
- |
- 4
发布时间：2021-06-03 - 摘要：Background: When upgrading the reactor power to promote the power efficiency of nuclear power plants,the BWR steam dryer is easier to be damaged by high cycle pulsating pressure,which is called acoustic fatigue problem.Purpose: An acoustic fatigue experimental schemes in steam generators is designed to study the acoustic load on the steam dryer.Methods: The acoustical computational software ACTRAN has been applied to simulate the acoustic pressure distribution inside scaled model of the upper chamber of steam generators.Results: Sound pressure level contour distributions at different frequencies are predicted.Conclusions: Sound pressure level contour distribution at mid-frequency is heterogeneous which is different with the low frequency one.关键词：Nuclear power;Steam dryer;Acoustic fatigue;ACTRAN
- 11
- |
- 105
- |
- 0
发布时间：2021-06-03 - 摘要：Background: High speed rotate of the flow in gaps around upper flywheel of the reactor coolant pump may create the phenomenon of Taylor vortex,which causes the fluid to be well-mixed.Purpose: Provide the input of thermal analysis for the whole structure.Methods: Make model of the flow around the lower flywheel,then use CFX program to perform Computational Fluid Dynamic（CFD） analyses of the flow and thermal fields.The influence of speed of flywheel and the size of radial gap are also discussed for the flow of Taylor vortex in the report.Results: Effective thermal conductivity and heat source can be calculated.Conclusions: Rotate speed and size of radial gap can be used to control the temperature distribution of upper flywheel.关键词：Taylor vortex;Effective thermal conductivity;Heat source
- 12
- |
- 180
- |
- 4
发布时间：2021-06-03 - 摘要：Background: The Integrated Head Package（IHP） is required to provide CRDM cooling.Purpose: In order to evaluate the CRDM ventilation system to ensure that there is adequate flow to cool the CRDMs.Methods: The CFD method is used to calculate the fluid field of CAP1000 reactor IHP airflow system.Results: This analysis demonstrates that the IHP design meets or exceeds the requirements and will therefore provide the required cooling air flow to the CRDMs.Conclusions: The results will provide the technical support for the choice and design of cooling fan.关键词：CAP1000;Integrated head package;Fluid field analysis;CFD
- 14
- |
- 141
- |
- 11
发布时间：2021-06-03 - 摘要：Background: The crack will grow gradually under alternating load for a pressurized pipe,whereas the load is less than the fatigue strength limit.Purpose: Both calculation and evaluation methodology for a flawed pipe that have been detected during in-service inspection is elaborated here base on the Elastic Plastic Fracture Mechanics（EPFM） criteria.Methods: In the compute,the depth and length interaction of a flaw has been considered and a compute program is developed per Visual C++.Results: The fluctuating load of the Reactor Coolant System transients,the initial flaw shape,the initial flaw orientation are all accounted here.Conclusions: The calculation and evaluation methodology here is an important basis for continue working or not.关键词：Flaw growth rate;Evaluation;Fluctuating load;Program developing
- 14
- |
- 123
- |
- 1
发布时间：2021-06-03 - 摘要：Background: A flaw is discovered by ultrasonic test at reactor pressure vessel during the service.Purpose: In order to insure the structural integrality of RPV,it is necessary to perform the fracture analysis for RPV flaw.Methods: According to ASME rule,fracture analysis is performed,which the flaw is assumed as a circumferential surface crack and crack depth is 10.1 mm.The analysis work contains the calculation of fatigue crack growth and the assessment of stress intensity factor under several category conditions.The loads are the temperature fluctuation,pressure and weld residual stress.The concerned category conditions include normal and upset conditions,emergency conditions,and faulted conditions.Results: The analysis results show that normal and upset conditions transient loading has little effect on the fatigue crack growth of low inner surface crack.The fatigue crack growth is about 0.228 mm at the end of 40 years service life.Conclusions: The stress intensity factor results of all conditions satisfy the requirement of ASME Rule.The RPV with flaw can continue service without repair.关键词：Stress intensity factor;Pressure vessel;Fracture;Assessment
- 13
- |
- 119
- |
- 1
发布时间：2021-06-03 - 摘要：Background: Leak-before-break technique is widely used in operating and abuilding nuclear plants,for which net-section-criteria is adopted by ASME to assess flawed pipes under full yielding condition.Purpose: However,net-section-criteria has been proved to overestimate the loading capacity of structure.Methods: For nuclear pipeline with internal surface crack,finite element method is used to simulate its deformation process under internal pressure,and initial plastic failure load is determined based on J integral curves of crack front.Results: Then,calculated value of initial plastic load is compared with theoretical one from ASME code,and the results show that the theoretical value overestimates the capacity load of structure.Conclusions: Finally,applicability of allowable membrance stress for level-A service restriction set by ASME-BPVC-XI code is evaluated.关键词：Nuclear pipeline;Surface crack;Initial plastic failure load;ASME-BPVC-XI code
- 12
- |
- 115
- |
- 0
发布时间：2021-06-03 - 摘要：Background: Dissimilar metal welded joints（DMWJs） were indicated to be vulnerable components in the primary systems,an accurate structural integrity assessment for such DMWJ is needed.Purpose: Understand the fracture performance of the welded joint containing defects,determine if appropriate to assess the defects in the weld joint with the present codes.Methods: The fracture tests of the initial crack located in the different positions of Alloy 52M DMWJ specimens were carried out by using single-edge notched bend（SENB） method.Results: Different crack resistance and crack propagation path were obtained for the SENB specimens,and the cracks always deviate to the materials with lower strength.Conclusions: The crack path deviation is mainly controlled by strength mismatch,rather than toughness mismatch.The J-resistance curves with larger crack path deviation only reflect the apparent fracture resistance along the crack growth region,rather than the intrinsic fracture resistance of the initial crack-tip region material.Without considering the local fracture resistance properties of heat affected zone（HAZ）,interface region and internal affected region（IAZ）,following present assessment codes will unavoidably produce non-conservative or excessive conservative assessment results.In most cases,the assessment results will be potentially non-conservative.关键词：Dissimilar metal welded joint;J-resistance curve;Crack growth path;Integrity assessment
- 12
- |
- 372
- |
- 15
发布时间：2021-06-03 - 摘要：Background: Many investigations on the structural integrity assessment of reactor pressure vessel（RPV） under the pressure thermal stress（PTS） conditions have been performed.Probabilistic fracture mechanics（PFM） technique has been used for RPV PTS analysis since 1980.However,research on the structure integrity assessment techniques using the PFM method is insufficient in China to support the application of the method to be used for assessment of RPV under PTS conditions.Purpose: To gain the effective right assessment approach applied to Chinese RPV against PTS.Methods: Thermal stress distribution along the cylinder wall-thickness under PTS is firstly evaluated by FE in order to get the corresponding stress intensity factors（SIF） and to validate them with some available SIF solutions.The PFM analyses based on crack initiation are then performed under conditions of the crack size,fracture toughness and material yield stress.Both linear elastic fracture mechanics and R6 method are used in the analyses,and the sensitivity of uncertainties as well as the effect of the assessment method is discussed.Monte Carlo simulation is used to calculate the failure probability.Results: preliminary software is developed to perform the assessment for PTS analysis.The failure probability with R6 method is a little higher than with the linear elastic fracture mechanics approach.The effect of reference nil-ductility transition temperature RTNDT on the failure probability is more significant than other factors.Conclusions: Many uncertainties should be considered in RPV PTS analysis with PFM approach in the further,including the input parameters and failure models.Low RTNDT can extend the time of failure,and reduce the failure probability.关键词：Probabilistic fracture mechanics;Pressurized thermal shock;Reactor pressure vessel;Monte Carlo simulation
- 12
- |
- 288
- |
- 11
发布时间：2021-06-03 - 摘要：Background: The structure integrity of Reactor Pressure Vessel（RPV） is one of the main concerns related to the safe operation and plant-life extension issues for nuclear power plant.It is very important for the operation safety of nuclear power plant to verify the structure integrity of RPV under Pressurized Thermal Shock（PTS）.Purpose: To verify the structure integrity of AP1000 RPV,the necessary of structure integrity evaluation with deterministic method is discussed in this article and the evaluations for AP1000 RPV are performed.Methods: FAVLoad,which is the fracture mechanical analysis module in FAVOR,is applied in the evaluations.The benchmarks from IAEA-TECDOC-1627 are analyzed for verification.At last,the deterministic structure integrity analysis and evaluation of AP1000 RPV are performed.Results: The RTPTS for AP1000 RPV is lower than the limit under the assuming PTS transients obtained in the analysis.Conclusions: The results show that the integrity of AP1000 RPV can be maintained,and the PTS analysis method with FAVLoad for RPV is feasible.关键词：AP1000;RPV;PTS;Structure integrity
- 14
- |
- 429
- |
- 9
发布时间：2021-06-03 - 摘要：Background: With the development and wide application of LBB technology,LBB can be applied on the pipe with known defects through aborative crack propagation analysis and deterministic LBB analysis.Purpose: Under this circumstance,some additional analysis are needed besides the traditional procedures required by SRP 3.6.3 and NUREG 1061.Methods: One of them is the analysis of the influence of off-center TWC on pipe LBB applicability.Results: In this text,three types of NPP high energy piping,whose nominal diameter is 6in,14in and 22in respectively,are chosen to be analyzed for the research of off-center TWC on pipe LBB applicability.First,the influence on COA is analyzed.Then the influences both on leakage crack size and on the maximum crack stable loads are analyzed.Conclusions: And finally the conclusions which can aid the LBB application on pipe with defect are presented.关键词：LBB;Off-center crack;COA;TWC;Leakage crack size
- 10
- |
- 87
- |
- 0
发布时间：2021-06-03 - 摘要：Background: There is a discrepancy between off-center crack and center-crack.Purpose: Crack grow rule and path need to be investigated under different load conditions.Methods: 3D elastic finite element method is used to create an off-center inner-surface crack model based on Paris-law crack grow equation by ABAQUS.Results: Crack grow rule and path are obtained under different load conditions.The influence of the crack rotation angle to crack extension has been researched.At the same time,centered circumferential crack model has been compared with off-centered circumferential crack model,and the discrepancy has been found.Conclusions: In LBB analysis,off-center influence should be considered in crack leakage analysis,but not necessary in crack stability analysis.关键词：LBB;Off-centered crack;Crack grow
- 11
- |
- 164
- |
- 4
发布时间：2021-06-03 - 摘要：Background: Elbows are important constituent parts of piping systems in nuclear power plant.The pressure boundary integrity of elbows direct influences reliability of nuclear power plant operation.Purpose: Therefore,it is necessary to carry out analysis of the cracked elbows to evaluate whether the safety requirement is satisfied.Methods: The principal fracture mechanics parameters required in analysis and evaluation are stress intensity factor K and J integral.The most important procedure for calculating fracture parameters K/J is to build up correct and reasonable mesh using finite element method.Results: In this paper,a fast programmed and partitioned three dimensional modeling procedure was proposed,which is useful to model the complex finite element model in fracture mechanics calculation,and calculation case of KJ of cracked elbows which subjected to inner pressure and in-plan moment was given based on the modeling method.Conclusions: The results prove that the programmed and partitioned modeling method developed in this paper could solve the complex modeling issue in fracture mechanics calculation of cracked elbow,as well as offer experiences for modeling other cracked components and calculating fracture mechanics parameters KJ.关键词：Elbows;Cracks;Fracture mechanics parameters;Finite element
- 13
- |
- 180
- |
- 6
发布时间：2021-06-03 - 摘要：Background: The uncertainties of input parameters in an deterministic structural integrity assessment of pressure vessels may affect the assessment results.This can be improved by performing probability facture mechanics（PFM） analysis.Purpose: This work investigates the effect of uncertainties of load,defect size,fracture toughness and failure criteria on the failure probability of semi-elliptical surface cracks in plates under combined tension and bending.Methods: The correction factor method provided by EPRI is used to estimate the stress intensity factor（SIF）.The J-integral values at the deepest point of the surface crack tip are evaluated using the reference stress method and the globe limit load solution developed by Goodall & Webster and Lei.PFM analysis is performed with considering the uncertainty of crack size,yield strength and fracture toughness and Monte-Carlo（MC） simulation is used to calculate the failure probability.Results: Failure probability increases with increase of load level,Lr,for all load ratio values considered in this work for a given failure criterion.However,the failure probability based on the elastic-plastic fracture criterion is higher than that based on the linear elastic fracture criterion for a given load lever,Lr.Conclusions: The load level and the failure criteria have significant effect on the failure probability.However,the load ratio makes a little contribution to the failure probability for a given failure criterion.关键词：Probabilistic fracture mechanics;Monte-Carlo simulation;Surface crack;Fracture;Combined loading
- 12
- |
- 253
- |
- 1
发布时间：2021-06-03 - 摘要：Background:The wear type defect is very common volumetric flaw in steam generator tube,which is most important aspect to affect strength of steam generator tube.Thus,a specific assessment method shall be formulated aiming to this thin-wall tube.Purpose: The aim is to establish a formula of remaining strength factor based on burst pressure experiment results of steam generator tube specimens,which is applicable to safety assessment of cylinderical tubing containing local wall thinning defect.Methods: Therefore,the burst pressure test was performed in order to calculate the remaining strength of steam generator tube containing local thinning wall.Concerning assessment methodology of local wall thinning of cylindrical tubing,several assessment methodologies of the structural integrity were recommended within the codes,such as China code of NB20013,BS7910,API579 and Janelle’s dissertation.In the recent year,the burst pressure experiment was carried out on cylindrical tubing with local wall thinning defect,and the methods mentioned above were used to calculate the burst pressure and compared with the experiment result.Results: However,the calculated burst pressure result does not always match well the experiments.In the case that defect depth by wall thickness is over than or equal 70%,the calculated burst pressure is far more than the experiment’s,thus this case is not conservative as far as safety concerned.Conclusions: Finally the modified RSF formula is developed on the basis of the experiment’s burst pressure,and which is most accurately applicable to the result of burst pressure experiment.关键词：Remaining strength;Defects;Tube;Pipe;Assessment methodology
- 11
- |
- 82
- |
- 0
发布时间：2021-06-03 - 摘要：Background: Thermal output of high temperature strain gauges can be measured by different methods.Purpose: This paper is to discover measurement methods’ effects on thermal output of 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.Methods: By analysis of impacting factors of thermal output,thermal outputs were explored for mismatching thermal expansion coefficients between material of components to be tested and that of strain gauge base.Thermal outputs of all kinds by different fixture methods were obtained and compared for hanging,clamping and spot welding,respectively.Results: Thermal output obtained by way of hanging strain gauge can be used in case of installation on material of the same thermal expansion coefficient as the strain gauge base.And this hanging measurement method provides access to thermal output that strain gauge undergoes different temperature changing ratio from that one of provision offered by strain gauge fabrication factory.Thermal output obtained by way of clamping measurement method can substitute for the one obtained by spot welding.This clamping measurement method can not only realize the effect of spot welding,reducing calibration cost due to spot welding of strain gauges,but also realize the aim of calibration of strain gauges one by one with the strain gauges be intact after calibration.Differences between thermal outputs by two measurement ways of both hanging and clamping explain that large divergence can be made when thermal expansion coefficients between the material of strain gauge base and the one of the component to be installed on.Conclusions: Thermal output can be measured among all methods by fixture of clamping instead of welding,with an advantage of realizing strain gauge calibration one by one and also with a high precision.关键词：High temperature strain gauge;Thermal output;Measurement method;Clamping
- 12
- |
- 277
- |
- 8
发布时间：2021-06-03 - 摘要：Background: Thermal stress will be occurred in the Sodium Cooled Fast Reactor when it starts or stops,after a number of cycles,thermal fatigue damages will be emerged,especially in the connection areas of tees.Purpose: To study the relations between the angles and the fatigue life in tees.Methods: The thermal fatigue life and fatigue damage coefficient of the tees of different angles are computed by the ANSYS software.Results: This paper gives a function charactering the relations between the fatigue allowable intensity and the tee angle.Conclusions: The fatigue strength is reducing with the decreasing of angle,which can provide certain valuable reference for designing tee nuclear reactor.关键词：Thermal fatigue;Fatigue strength;Tee;Stress concentration factor
- 13
- |
- 232
- |
- 6
发布时间：2021-06-03 - 摘要：Background: There are many uncertainties in the fatigue design of the equipments,the probability method can be used to assign the fatigue design.Purpose: To provide the structure fatigue reliability of the area of the plane head interface.Methods: The fatigue reliability analysis has been done by using the reliability method which considering the geometry dimensions uncertainty and loads uncertainty.Results: The dimensions uncertainty has great effect on the uncertainty of the products’ life.Meanwhile,this article provides one method,by using 6σ concept,to solve the fatigue reliability analysis considering the loads uncertainty.Conclusions: The loads uncertainty results can be combined with the dimensions results to get the overall reliability results.关键词：Fatigue;Reliability;Monte Carlo simulations;Geometry/load uncertainty
- 11
- |
- 118
- |
- 2
发布时间：2021-06-03 - 摘要：Background: Reactor internals are important equipments in the reactor coolant system,its structure design needs high reliability in the entire lifetime.Reactor internals have occurred breakdown and the damage event due to flow induced vibrations in the domestic and foreign nuclear power plants,which make immediate influence on reactor safe operation and economic efficiency.Purpose: In this work,the dynamic response of reactor internals-vortex suppression plate and secondary core support structure（SCSS） under the loading from pump induced vibrations and flow induced vibrations are studied.Methods: Based on the finite element model of SCSS,Spectrum analysis and the harmonious analysis are performed,in order to get the response of the structure under flow induced vibrations.Then,the high fatigue of the structure is assessed according to the ASME B&PV Code.Results: The results indicate that alternate stresses of all the components satisfy the limiting value in the correlative requirements.Conclusions: The structure of SCSS could bear the vibration induced from the flow and the pump,and the method used in this article provides the reference for other reactor internals structure analysis like this.关键词：Reactor internals;Flow induced vibrations;Pump induced vibrations;High cycle fatigue
- 11
- |
- 69
- |
- 0
发布时间：2021-06-03 - 摘要：Background: Considering the EAF（Environment Assistant Fatigue） effect during the design status is a key issue for generation III nuclear power plant.Purpose: The calculation methodology of Fen（Environmentally Assistant Fatigue Factor） and strain ratio of ASME Code Case N-792 has been introduced in this paper.Methods: By creating 2D symmetry model of reactor pressure vessel nozzle,the environmentally assistant fatigue factor Fen and fatigue usage factor has been analyzed according Code Case N-792 with simplified and detailed integral methods.Results: The fatigue usage factor results of ASME-III which does not consider environmental effect and Code Case N-792 which consider the environmental effect have been compared and discussed.Conclusions: And the fatigue life considering Fen effect of SCL1 section will reduce to about 1/3 of the life in air condition while the fatigue life of SCL2 decrease 3/5.
- 12
- |
- 85
- |
- 0
发布时间：2021-06-03 - 摘要：Background:Fatigue is an important ageing mechanism in RPV and it must be contained to ageing management working range.Purpose:In order to ensure the safety operation of nuclear power plants,as extension of RPV service time,it is necessary to assess the fatigue damage caused by actual operation transient.Methods:Based on monitoring data of actual operation during the past eleven years,refer to design transient,the statistic analysis for types and occurrence times of actual transient is carried out,at the same time,every transients are combined as different operation cycles and the temperature field and stress field of typical components are analyzed by FEM.Results:Based on these information,fatigue analysis and assessment are finished,if later-actual transients are similar with the previous transients,the calculation result shows that the ratio between maximum of cumulative usage factors and design calculation value is 0.4967 the design transients is conservative.Conclusions:Fatigue status of RPV could be assessed and traced quickly through fatigue status assessment method in this paper based on actual operational transient and assessment result would be a good reference for RPV aging management.关键词：Reactor pressure vessel;Actual operational transient;Fatigue damage;Aging management
- 13
- |
- 129
- |
- 5
发布时间：2021-06-03 - 摘要：Background:In super-critical water-cooled reactor,creep-fatigue under high working temperature will affect life of various equipments.Purpose:Understanding characteristics of creep-fatigue interaction correlation is important before evaluating its damage by ASME code.Methods:In order to study creep-fatigue interaction of 316 stainless steel（316SS）,three types of creep-fatigue interaction have been taken into consideration in our experiments,which include traditional creep-fatigue interaction subjected to cycling load with hold times at its peak and also creep-fatigue interaction by decoupling creep and fatigue load.Based on the experimental data,the different types of creep-fatigue interaction have been compared and discussed.Moreover,suitability of ASME NH rules in assessing various creep-fatigue interactions of 316SS produced in China is investigated.Results:Creep-fatigue damage envelop of 316SS from ASME can cover the experimental results from the three types of experiments mentioned above.Conclusions:This experimental study is useful as a reference in creep-fatigue interaction experiments,application of ASME code and study of mechanical characteristic of high temperature structural materials in generation IV nuclear power plant.关键词：316SS;Creep;Fatigue;Creep-fatigue interaction;Loading history
- 14
- |
- 509
- |
- 6
发布时间：2021-06-03 - 摘要：Background: Constraint can significantly alter the material’s fracture toughness.Purpose: In order to increase accuracy of the structural integrity assessment.It needs to consider the effect of constraint on the fracture toughness of nuclear power materials and structures.A unified measure which can reflect both in-plane and out-of-plane constraint is needed.Methods: In this paper,the finite element numerical simulation method was used,a unified measure and characterization parameter of in-plane and out-of-plane constraint based on crack-tip equivalent plastic strain have been investigated.Results: The results show that the area surrounded by ε p isoline has a good relevance with the material’s fracture toughness on different constraint conditions,so it may be a suitable parameter.Based on the area APEEQ,a unified constraint characterization parameter Ap is defined.It was found that there exists a sole linear relation between the normalized fracture toughness JIC/Jref and Ap regardless of the in-plane,out-of-plane constraint and the selection of the ε p isolines.The sole JIC/Jref-Ap line exists for a certain material.For different materials,the slope of JIC/Jref-Ap reference line is different.The material whose slope is larger has a higher JIC/Jref and is more sensitive to constraint at the same magnitude of normalized unified parameter.Conclusions: The unified JIC/Jref-Ap reference line may be used to assess the safety of a cracked component with any constraint levels regardless of in-plane or out-of-plane constraint or both.关键词：In-plane constraint;Out-of-plane constraint;Crack-tip equivalent plastic strain;GTN damage model;Fracture toughness
- 12
- |
- 198
- |
- 3
发布时间：2021-06-03 - 摘要：Background: A connection between a low alloy steel pressure vessel nozzle and a stainless steel main pipe safe-end in a pressurized water reactor（PWR） nuclear power plant is a dissimilar metal weld,which is at key location of pressure boundary and serves in high temperature primary water environments.Since 1990,cracking failure incidents happened on the welds in the primary water coolant of several PWRs in the countries such as USA,Sweden and Japan.Purpose: This paper is to briefly review the typical cases involving this problem in engineering and relevant researches.Methods: Methods such as nondestructive testing,crack and microstructure observation,finite element analysis were used in the failure analysis.Results: The cause of the failure cases was primary water stress corrosion cracking（PWSCC）.Conclusions: It is concluded that the welds of 600 Alloy type such as 82/182 are susceptible to PWSCC but those of 690 Alloy Type such as 52/152 have excellent resistance to the cracking,and more researches are necessary.关键词：Pressurized water reactor nuclear power plant;Pressure boundary;Dissimilar metal weld;High temperature water environment;Failure;Stress corrosion cracking
- 14
- |
- 331
- |
- 24
发布时间：2021-06-03 - 摘要：Background: The high temperature creep resistance is an important indicator of the performance evaluation of reactor material.Purpose: This paper studies the high temperature creep behavior of cladding candidate material Ni-based alloy C276 of Supercritical Water Reactor（SCWR） under argon?s protection,analyzes the creep mechanism and evaluates creep properties of C276.Methods: In order to reduce the experiment cost,radiation dose and strengthen the penetration of irradiation experiments,using the unconventional small sample has become the trend of experimental study.In this paper,small sheet sample will be used at high temperature creep experiments.In order to avoid the impact of oxidation at high temperature on the data of Material properties,this article will use independently designed and improved high temperature creep machine with argon?s protection.Results: The creep curves of C276 have been obtained at the high temperature,when the samples of C276 are respectively applied at the different constant stress.Conclusions: The results prove that using small sample under the protection of argon can more truly reflect the mechanical properties of materials.关键词：Creep;Ni-based alloy C276;SCWR;Small sheet sample
- 14
- |
- 132
- |
- 1
发布时间：2021-06-03 - 摘要：Background: It is difficult to accurately simulate the non-continuous structure in reactor equipment system.Purpose: In order to investigate and simulate the non-linear characteristic of core barrel alignment plate.Methods: In this paper,a finite element model of core barrel alignment plate interface is established to investigate the non-linear analysis technology.Results: The elastoplastic deformation characters,which are obtained using finite element analysis by ANSYS software,are prepared as inputs for reactor equipment system model.Conclusions: The method used in this paper has certain referential significance for simulating the contact between non-continuous structures.关键词：Core barrel alignment plate;Non-continuous structure;Non-linear contact analysis;ANSYS
- 11
- |
- 39
- |
- 0
发布时间：2021-06-03 - 摘要：Background: Passive hydrogen autocatalytic recombiner is a device for eliminating hydrogen in the containment of the nuclear power plant when severe accident occurs,avoiding hydrogen explosion.After the Fukushima nuclear accident,the nuclear power plants pay more attention to the role of Passive hydrogen autocatalytic recombiner.Purpose: This paper studies the stresses of passive hydrogen autocatalytic recombiner under the seismic and LOCA conditions.Methods: Modeling by using the finite element software ANSYS,the impacts of airflow load under the LOCA conditions are considered reasonably and the strength of passive hydrogen autocatalytic recombiner is also evaluated according RCC-M.Results: The results show that the model can meet the requirement of the standard document.Conclusions: This paper will provide technical support for stress analysis and evaluation of passive hydrogen autocatalytic recombiner.关键词：Passive hydrogen autocatalytic recombiner;ANSYS;Stress analysis;RCC-M
- 12
- |
- 74
- |
- 2
发布时间：2021-06-03 - 摘要：Background: In the AP1000 Reactor system,many devices sustain the pressure.Purpose: The sealing function of these devices is essential to ensure the structural integrity of the reactor system.Methods: In the ASME code,Waters method is adopted for the design of sealing structure.Although Waters code is very reliable,it can’t show the details of the deflection and stress distribution of the sealing structure under pressure.In this paper,a finite element method is developed to evaluate the typical sealing structure.The preload of the bolt and the spring back behavior of the gasket are also simulated by a proposed equivalent method.Results: With the finite element model,the spring back of the gasket,the deflection of the flange and the stress distribution of the sealing structure are predicted.Conclusions: Finally,The quantitative correlation between gasket spring back and inner pressure of the device is obtained,which provide a useful reference for the design of the sealing structure.关键词：Sealing structure;Finite element analysis;Flange;Gasket
- 12
- |
- 134
- |
- 2
发布时间：2021-06-03 - 摘要：Background: Reactor safety is an important component of developing and designing any nuclear energy systems.The Freeze-Valve is one of the core technologies of the molten salt reactor,which thermal-structural property is directly related to the inherent safety of the molten salt reactor nuclear system.Purpose: The purpose of this paper is to improve the inherent safety of the fourth-generation nuclear energy systems,by researching and optimizing the thermodynamic properties of Freeze-Valve,and exploring reliable safety design.Methods: A 3D Finite Element model to simulate the thermal-structural coupling behaviors was established by the ANSYS software to analyze the performances and the properties of a Freeze-Valve.Results: The analysis result shows that the flat part of the Freeze-Valve plays a very important role in engineering applications.The effects of different factors（heat transfer coefficient,heating power,and insulation size and so on） on the temperature and the stress field of the Freeze-Valve during operation are obtained,which provide some basis for optimization of design and safe operation.The thermal stress of the initial model of the Freeze-Valve is so large that it is easy to cause fatigue failure,owing to the unreasonable initial design（insulation size and heating power,for example）.Conclusions: After the relative analysis and structural optimization,the maximum stress of the Freeze-Valve is significantly reduced,and the performance has been greatly improved.关键词：Molten salt reactor;Freeze-Valve;Thermal-structural coupled;Finite element method
- 15
- |
- 235
- |
- 6
发布时间：2021-06-03 - 摘要：Background: Hydraulic Control Rod Drive Technology（HCRDT） is a newly invented patent and Institute of Nuclear and New Energy Technology Tsinghua University owns HCRDT’s independent intellectual property rights.Purpose: Integrated valve which is made up of three direct action solenoid valves is the key part of this technology,so the performance of the solenoid valve directly affects the function of the integrated valve and the HCRDT.The performance of the coil affects the service behavior of the solenoid valve,so the thermal performance of the coil is researched.Methods: The temperature field of the coil of the direct action solenoid valve was analyzed using the ANSYS software with various currents,the results of which were validated by experiments.Results: The result shows that the temperature of the coil of the solenoid valve increases with the current increasing firstly.Second,the temperature of the inner wall of the coil is higher than that of the exterior wall.The temperature of the middle coil is higher than that of the edge of the coil.Third,the coefficient of thermal conductivity of the coil is obtained.Fourth,the highest temperature of the coil is under the limiting temperature.Furthermore,the comparison of the temperature of the coil achieved by calculation with the one achieved by experiment under different condition shows that the calculation results are rational if the experiment errors are considered.Conclusions: The results can be provided for the parameter design of the solenoid valve.关键词：Control rod hydraulic drive system;Solenoid valve;Coil;Temperature field
- 14
- |
- 278
- |
- 9
发布时间：2021-06-03 - 摘要：Background: The module construction technology is widely used in the AP1000 nuclear power plant.The CA structural module in the containment building is important for the safety of the nuclear power plant,the corner wall of the CA structural is critical under load,so it is important to do research on bearing capacity of CA corner wall.Purpose: Finite element analysis model of the CA corner wall is conducted in the ABAQUS software to study the bearing capacity and the stress/strain state of the CA corner wall under different load cases.Methods: Finite element analysis model considering the material nonlinearity,contact nonlinearity and geometric nonlinearity conducted in the ABAQUS software is used for the analysis of the CA corner wall.Results: The ultimate bearing capacities of the CA corner wall in different loads are studied,and the study also gets the load-displacement curves.The bearing capacities of CA corner wall under different axial loads are also studied in this paper.The influence of ultimate bearing capacities of CA corner wall is founded out if the steel and concrete work together.Conclusions: When the CA corner wall subjects to the bending load,the failure part of the CA corner wall is the corner core concrete.When the CA corner wall subjects to the shear load,the failure part of the CA corner wall is the straight module wall.The axial load has notable influence on the capacity of CA corner wall.关键词：AP1000;CA corner wall;Ultimate bearing capacity;States of damage;Axial load;Work together
- 12
- |
- 45
- |
- 1
发布时间：2021-06-03 - 摘要：Background: Modularization is one of the main characteristics for AP1000 nuclear power plant building.The steel-concrete-steel module wall is used instead of reinforced concrete structure wall.Usually,lots of Overlay Plate Embedments will be installed on the module wall to connect and fasten other structures,such as pipes,equipments and operation platforms.As for many supports taking less design loads,the safety margin is too big when using OLP embedment.Purpose: An improvement design will make sense that the supports with less deisn loads can be welded directly to the module wall instead of embedments.Methods: A finite element analysis based on nuclear-related concrete code is carried out.Results: Through analysis,the equations for the allowable design loads of supports to be welded directly to module wall are provided in this paper.Conclusions: The improvement design is proved feasible.In this way,the strength for steel face plate and studs will be utilized fully and this method will facilitate and simplify the design and construction with considerable engineering application value.关键词：Module wall;Supports;Stud;Capacity
- 15
- |
- 91
- |
- 3
发布时间：2021-06-03 - 摘要：Background: The supporting capacity of embedment plates in AP1000 nuclear power plant is obtained by finite analysis method,and the effect of concrete should be taken into consideration.Purpose: The interaction coefficients between embedment and concrete will be studied in this paper.Methods: The theory of elasticity for load distributed over a part of the boundary of a semi-infinite solid and method for anchoring design in nuclear safety concrete structures code ACI 349 are used to proceed with the study.Results: The formulas for interaction coefficients between embedment and concrete were deduced.Conclusions: The study findings can be used for other similar situations and it makes a theoretical foundation for AP1000 nuclear power plant embedment finite element analysis.关键词：AP1000;Embedment;Concrete;Interaction coefficients
- 12
- |
- 88
- |
- 0
发布时间：2021-06-03 - 摘要：Background: Several aspects will influence the weld residual stress analysis of CRDM penetration,such as formation and distribution of heat source,heating and cooling processing,finite elements meshing,phase changing,etc.Purpose: Develop an effective method for weld residual stress analysis,research the influence of interval between heating and cooling step during multipass beads procedure.Methods: Finite elements with a sequentially coupled approach are applied,including a transient thermal analysis simulating the heating and coduring and their interval between multipass welding.Results: An appropriate finite method is developed for weld residual stress simulation.Conclusions: Apropriate interval time must be choosed between multipass weld beads to avoid poor weld residual stress,and contact pressure between RPV shell and CRDM adapter tube also should be check for strcture integrity.关键词：Control rod drive mechanism;Pre-stress fields;Weld residual stress
- 11
- |
- 143
- |
- 3
发布时间：2021-06-03 - 摘要：Background: Flow Accelerated Corrosion（FAC） leads to wall thinning（metal lose） and strength reduction of steel piping.Piping degradation due to FAC can affect plant reliability and may cause serious accident.Purpose: This paper sets up the analysis and assessment method for the wall thinning due to FAC.Methods: This method mainly bases on local metal loss Level 3 of API 579.Using FEA software ANSYS develop parameterized models,including straight pipe,elbow,reducer,etc.The models can be expediently defined geometry and material parameter,load and constraint condition.Furthermore,the real thickness model of the pipe can be established by using the data of in-service inspection.Results: With the limit-load analysis method to assessment the strength of piping,the method achieved local metal loss Level 3 assessment effectively.Conclusions: Using this method to establish the true model of pipe wall thickness,can improve the accuracy of analysis of pipe wall thickness thinning evaluation.关键词：Wall thinning;Parameterized model;Level 3 assessment;Limit-load analysis
- 12
- |
- 97
- |
- 1
发布时间：2021-06-03 - 摘要：Background: China experimental fast reactor（CEFR） is the first fast reactor in China,the piping designs are very important for the fast reactor.Purpose: The purpose is to improve the economy of pipeline design for the fast reactor.Methods: According to the optimization principles,redesign the support system of emergency decay heat removal system.Results: The piping of emergency decay heat removal system meets the ASME standard stress limit value under various prospective loads,and reduces the number of dampers,spring hangers.Conclusions: We accumulate layout experience for the support design of high temperature piping by the optimization design for emergency decay heat removal system,lay the foundation for the fast reactor piping design in the future.关键词：Emergency decay heat removal system;Support;ASME;High temperature piping
- 12
- |
- 66
- |
- 1
发布时间：2021-06-03 - 摘要：Background: The definition of hard rock shear wave velocity is one of the most critical issues in the work of site selection.Purpose: To make a definition of hard rock site on which the model can be assumed as fixed-base condition,a series of research had been done.Several possible hard rock site soil models were developed.Methods: Shear wave velocity of hard rock had been assumed from 1100 m/s to 3200 m/s.For each case,free field analysis and soil structure analysis had been performed.And responses in soil and key nodes of structure were compared.Results: In free field analysis,responses of models that shear wave velocity below 2400 m/s decreased a lot.In SSI analysis,structure responses didn’t change much when shear wave velocity was above 2400 m/s.Conclusions: 2400 m/s was the lowest shear wave velocity for hard rock site for fixed-base assumption.关键词：Nuclear power plant;SSI;Hard rock;Shear wave velocity
- 12
- |
- 102
- |
- 1
发布时间：2021-06-03 - 摘要：Background: Self-compacting concrete that has good working property is the prerequisite of steel-concrete structure.The early age mechanical property of self-compacting concrete is the important parameter when design steel-concrete structure.Purpose: This paper attempts to research the working property and early age mechanical property of self-compacting concrete.Methods: Test is used to research the working property and early age mechanical property of self-compacting concrete.Results: Self-compacting concrete that could meet the requirement of steel-concrete structure has been mixed and parameters of early age mechanical property of self-compacting concrete which is necessary for design of steel-concrete structure have been presented.Conclusions: Base on the results,this paper can guide the construction of self-compacting concrete in steel-concrete structure and the design and construction of steel-concrete structure.关键词：Self-compacting concrete;Working property;Early age mechanical property
- 10
- |
- 76
- |
- 1
发布时间：2021-06-03 - 摘要：Background: The core support pads were used to limit circumferential rotation of core barrel,and the structure integrity of the core support pads was an important factor to the safe operation of nuclear power plant.Purpose: To ensure the structure integrity of the core support pads.Methods: Three-dimensional FEA model for bottom head,core support pads and part cylinder of CAP1000 RPV was established.Thermal analysis,static analysis,fatigue analysis and fracture analysis were performed.The analysis results were evaluated according to ASME B&PVC-III-NB-3200 and ASME B&PVC-III-1-Appendix G.Results: The evaluation indicated that the core support pads and surround bottom head could satisfy related requirements of above code.Conclusions: The analysis methodology used in this paper could also be applied to the core support pads of RPV for above 1000 MW nuclear power plant.关键词：Reactor pressure vessel;Core support pads;Structural analysis;Fatigue analysis;Fracture analysis
- 12
- |
- 158
- |
- 2
发布时间：2021-06-03 - 摘要：Background: Steel Containment Vessel（SCV） penetrations are special parts in nuclear plant.The analysis method and acceptance criteria of SCV penetration are different from piping.The analysis method and acceptance criteria for SCV penetration of CAP1000 plant are required.Purpose: To set the analysis method and acceptance criteria for SCV penetration.And base on these calculation and evaluation method,make a program for the penetration calculation and evaluation that is automatic,rapid and easy to use.Methods: The development of this program is based on the research and contrast.The axisymmetric penetration model with nonaxisymmetric loading is established in this program.The two sides loads combination of penetration is considered in this program.The stress is calculated by ANSYS 11.0;preprocess is developed by VB and post-process is done by Excel.Results: The analysis method and acceptance criteria of SCV penetration are established,and a program for the penetration calculation and evaluation is developed which is automatic,rapid and easy to use.Conclusions: The analysis method and acceptance criteria of SCV penetration and the program for the penetration calculation and evaluation in this paper are reliable.By this program,the high speed and accurate penetration analysis will be expected.关键词：Penetration;Axisymmetric model;Nonaxisymmetric loading;Stress analysis;Fatigue analysis
- 15
- |
- 96
- |
- 0
发布时间：2021-06-03 - 摘要：Background: The calculation of irradiation-induced dilation of the fuel subassembly duct in LMFBR is important for fast reactor core design.Purpose: To investigate how to calculate the dilation by using finite element method（FEM）.Methods: First,irradiation-induced creep and swelling material models are introduced.Then,a theoretical solution based on a simplified bending plate model is briefly given.Finally,a stress update scheme for the adopted material models is presented and furthermore embedded into ABAQUS user interface UMAT to conduct finite element analysis.Both solutions are compared and discussed.Results: FEM successfully predicts the duct dilation and its solution agrees well with theoretical one in small deformation.Conclusions: The proposed stress update scheme is effective.The accuracy of the theory solution declines when dilation becomes larger.The maximum stress occurs at the duct corner point,and the location has stress relaxation effect.关键词：Irradiation-induced creep;Swelling;Finite element method;Fuel subassembly duct;Dilation
- 12
- |
- 141
- |
- 3
发布时间：2021-06-03 - 摘要：Background: Steel plate reinforced concrete structure has small thickness,high capacity and high ductility,which is mainly used in nuclear power reactor facilities.Due to test condition,the mechanical properties and ultimate capacity of steel concrete structure under high temperature is little studied.The properties of steel concrete structure under high temperature have significant effect on application in nuclear power.It is necessary to study the properties of steel concrete structure under high temperature.Purpose: This article is to study the temperature and stress field distribution of steel concrete wall under high temperature.Methods: Based on the determination of temperature curve and material thermal properties,a finite element model of steel concrete wall was established using 3-D solid element,transient temperature and thermal stress analysis were carried out.Results: Temperature gradient between stainless steel plate and concrete and between stud and concrete increases with time,then declines,and increment of thermal stress of wall section become smaller gradually with time.The maximum first principle stress of concrete is larger than cracking stress.Conclusions: Thermal stress has significant effect on mechanical properties of steel concrete wall,and shall be considered in designing the steel concrete structure.关键词：Steel concrete;Finite element model;Temperature gradient;Thermal stress
- 14
- |
- 104
- |
- 1
发布时间：2021-06-03 - 摘要：Background: One of the most advanced nuclear power technology,the first AP1000 reactor is under construction in China.Modularization is one of the main characteristics for AP1000 nuclear plant building.Module wall with steel face plate is used instead of reinforced concrete structure wall.A number of OLP embedments need to be installed into the module wall to connect other structures such as pipes,equipments,operation platforms and any other component attached to the module wall.Therefore,the design of embedment is very important in AP1000 structural design.Purpose: A finite element analysis method and tool for embedment design is needed for convenience.Methods: This paper applies the self-developed GTStrudl command template and VBA macro program for embedment capacity calculation and evaluation based on Microsoft Excel to the embedment design.Results: A Microsoft Excel template for embedment design is developed.Conclusions: The analysis method and template brings reasonable results and may provide some help and use for reference for the engineering practice.关键词：AP1000;OLP embedment;GTStrudl finite element analysis software;VBA macro program
- 13
- |
- 107
- |
- 2
发布时间：2021-06-03 - 摘要：Background: Modularization is one of characteristics of AP1000 nuclear power plant,however,the challenge of large module lifting should be met.Purpose: An analysis method for large module lifting has been studied in this paper.Methods: Based on ANSYS software and steel design code of America,the finite element analysis of CA01 structure module in AP1000 nuclear power plant was conducted.Results: The analysis results show that stress and deformation meet the requirements of codes and construction when the temporary brace is properly set in CA01 module.The maximum stress can be found in the juncture of wall,and maximum deformation occurs at the corner of CA01 module.Conclusions: Proper temporary brace is important to the lifting of structure and maximum stress or deformation usually occurs at irregular locations of structure.And it is necessary to perform detail analysis before structure lifted.关键词：AP1000;Structure module;Lifting;Finite element analysis
- 12
- |
- 135
- |
- 2
发布时间：2021-06-03 - 摘要：Background: It is important for the safety of the radioactive material containers during transportation.Purpose: In the procedure of reviewing radioactive material containers transportation,it is very important factor to carry a drop impact analysis of container.Methods: This paper presents a drop impact analysis method of radioactive material container.First,do the calculation of several drop cases of the container such as horizontal drop,vertical drip and gradient drop with the famous transient dynamic analysis program LS-DYNA.Second,do the stress evaluation according to the rules in the ASME Section III Division I Appendices which are about the fatigue analysis.Results: With this method,we can do the judgment that whether the container’s strength is good enough or not.Conclusions: The radioactive material container’s strength is good enough by analysis.关键词：Radioactive material;Container;Drop impact;Stress intensity evaluation
- 15
- |
- 203
- |
- 16
发布时间：2021-06-03 - 摘要：Background: Control rod drive mechanism（CRDM） is one of the important safety-related equipment for nuclear power plants.Purpose: The operating parameters of stepping movement,including lifting loads,step distance and step velocity,are all critical design targets.Methods: FEA and numerical simulation are used to analyze stepping movement separately.Results: The motion equations of the movable magnet in stepping movement are established by load analysis.Gravitation,magnetic force,fluid resistance and spring force are all in consideration in the load analysis.The operating parameters of stepping movement are given.Conclusions: The results,including time history curves of force,speed and etc,can positively used in the design of CRDM.关键词：CRDM;Stepping movement;Numerical simulation
- 11
- |
- 106
- |
- 4
发布时间：2021-06-03 - 摘要：Background: The reactor structure may be destroyed by steam explosion,which could cause serious accident.Purpose: To evaluate the damage caused by the steam explosion outside the reactor.Methods: The method to evaluate the structural response of reactor cavity is established.Firstly,the flexural capacity and shear resistance are assessed.Secondly,based on the elastic-plastic FEM established in this paper,the dynamical response under different soil stiffness is analyzed.Results: The time displacement curve of the reactor structure is obtained.Conclusions: The structural integrality could be estimated and the possibility of leak accident can be judged.关键词：Steam explosion;Elastic-plastic FEM;Dynamics analysis
- 14
- |
- 54
- |
- 0
发布时间：2021-06-03 - 摘要：Background: DN350 squib valve is introduced into the third generation passive nuclear power plant.Purpose: In squib valve design,the required piston kinetic energy for valve opening is need to be calculated,and used to determine the right propellant quantity.Methods: LS-DYNA is used for the dynamic simulation of the piston impacting the shear cap.A multi-parts FEM model is created for calculation.The effects of strain rate and failure criteria are considered in the material models.Dynamic relaxation method is used to apply inner pressure and screw pre-tension loads.An equivalent material model for the crush tube assembly is set up with appropriate contact algorithm applied.Results: Calculation results such as kinetic energy,velocity and valve body loads are obtained.The calculated piston velocity is in good agreement with the test data.Conclusions: The simulation methods and models can be applied in valve structure design and type approval.关键词：Squib valve;Dynamic;LS-DYNA;Elasto-plastic
- 12
- |
- 132
- |
- 2
发布时间：2021-06-03 - 摘要：Background: In recent years some reactors have experienced significant steam dryer cracking.In some cases,this cracking has necessitated unplanned outages to implement steam dryer repairs and has also resulted in de-rated operation of the affected units.Initial inspection showed it was likely that steam dryer had been damaged by high cycle fatigue due to flow-induced acoustic resonance in the main steam lines.Because of the complexity and acoustic-structure coupling effect,scale model test is adopted to research the mechanism of acoustic fatigue.Purpose: This paper describes the derivations of scaling laws observed to control the system response for phenomena considered to be significant in the real plants.Methods: Basic governing equations of elasticity and acoustics are written in non-dimensional form,non-dimensional groups are defined and derived.Results: Using the reference values in the real plants,the scaling laws and scaling relationships are derived and recognized to enable conversion of model data into real plant predictions.Conclusions: Successful model testing can be achieved if these significant parameters are preserved in the model scale.关键词：Steam dryer;Acoustic resonance;Scale model;Scaling law;Acoustic-structure coupling
- 11
- |
- 122
- |
- 0
发布时间：2021-06-03 - 摘要：Background: A core drop accident causes the core barrel and supporting structure to drop and impact on the bottom of the reactor vessel.Purpose: To ensure the scram functionality is maintained.Methods: The stress-strain curves applicable to the material are approximated by Ludwik’s expression.Considering the core deadweight,buoyancy force and heat expansion,material train ratios for both cold and hot conditions are calculated.Results: Vessel impact load（8294482N in cold condition & 6064537N in hot condition） is determined,which remains within the vessel design specification limits.The calculated drop length（47.44 mm in cold condition & 27.63 mm in hot condition） is less than the fuel pin full-diameter engagement length,so the fuel assembly top nozzle will remain engaged during a core drop.Finally the compressive assemblies are evaluated,which would not buckle due to the core drop accident loads.Conclusions: A core drop will not affect the scram function.关键词：Core drop;Ludwik expand;Curve fitting;Energy absorption
- 12
- |
- 74
- |
- 1
发布时间：2021-06-03
- Address：2019 Jialuo Road, Jiading, Shanghai, China Postal code：201800
- Tel：021-39194977 Email：LHB@sinap.ac.cn
- Technical support is provided by Beijing Founder electronics co., LTD 沪ICP备05005479号-1
京公网安备11010802024621 - It is recommended to read the content of this site in Chrome&IE9+. Please switch to extreme mode in browser 360.
- Cookies We use cookies to help provide and enhance our service and tailor content. By continuing, you agree to the use of cookies.
- Total Visits：675265 Visits Today：341
0