Limit Load Solution of Non-Aligned Multiple Flaws

2018 ◽  
Author(s):  
Fuminori Iwamatsu ◽  
Katsumasa Miyazaki ◽  
Koichi Saito
Keyword(s):  
Power Plants ◽  
Limit Load ◽  
Pressure Vessels ◽  
Ductile Material ◽  
Evaluation Procedure ◽  
Test Results ◽  
Collapse Load ◽  
Flat Plates ◽  
Fracture Tests ◽  
Single Flaw

Limit load solutions have been applied to estimate the collapse load of a component made of ductile material. Worldwide maintenance codes for power plants, such as ASME Boiler and Pressure Vessels Code, Section XI, and JSME fitness-for-service code, describe limit load solutions under the assumption of a single flaw. Detected flaws are, however, not always a single flaw, and adjacent flaws due to stress corrosion cracking have been detected in power plants. Thus, development of a limit load solution to estimate the collapse load in the case of multiple flaws remains an issue of structural integrity evaluation. Under the aim of developing a method for evaluating the effect of multiple flaws on collapse load as a part of a limit load solution, fracture tests of flat plates and pipes with multiple flaws were conducted. Although experimental approaches have been attempted to establish the evaluation method, further efforts are required to incorporate the evaluation procedure into a code rule. Effective parameters for considering reduction of collapse load on the basis of test results for specimens with multiple flaws were identified. Test results clearly show a correlation between collapse load and ratios of net-section areas. This correlation leads to the conclusion that distance parameters and flaw length of a smaller flaw determine the existence of an effect on the collapse load by multiple flaws. To investigate the physical sense of the correlation, finite element analysis (FEA) was performed. The FEA results show that strain distributions at the flaw tip under several conditions correspond at the time of maximum load of the fracture tests regardless of the effect of multiple flaws. Also according to the FEA results, the extent of the strain field is linearly proportional to flaw length. These FEA results are consistent with the correlation obtained by the test results.

Load Bearing Capacity of Through Wall Cracked Elbows: Comparison of Test Results With Calculation

2004 ◽  
Author(s):  
J. Chattopadhyay ◽  
B. K. Dutta ◽  
H. S. Kushwaha
Keyword(s):  
Power Plants ◽  
Limit Load ◽  
Research Centre ◽  
Test Results ◽  
Element Analysis ◽  
Bhabha Atomic Research Centre ◽  
Integrity Assessment ◽  
Theoretical Predictions ◽  
Fracture Tests ◽  
Set Up

Integrity assessment of piping components with postulated cracks is very important for safe and reliable operation of power plants. Pipe bends or elbows are one of the very important piping components in any power plant. The existing equations of limit load of elbows have various shortcomings. Additionally, the test data on elbows are not so abundant in the literature. Against this backdrop, a comprehensive experimental and analytical program has been undertaken at Reactor Safety Division (RSD) of Bhabha Atomic Research Centre (BARC) to carry out fracture tests on through wall cracked elbows and also to propose new limit load formulas of through wall cracked elbow. The present paper describes the elbow test specimens, test set-up, test results, brief description of elastic-plastic finite element analysis, newly proposed collapse moment equations for through wall circumferentially cracked elbows and the comparison of test results with theoretical predictions.

Prediction of Collapse Load Reduction due to Non-Aligned Multiple Flaws

2020 ◽  
Author(s):  
Fuminori Iwamatsu ◽  
Katsumasa Miyazaki ◽  
Koichi Saito
Keyword(s):  
Plastic Strain ◽  
Fracture Strain ◽  
Equivalent Plastic Strain ◽  
Plastic Collapse ◽  
Load Reduction ◽  
Test Results ◽  
Collapse Load ◽  
Flat Plates ◽  
Fracture Tests ◽  
Flaw Characterization

Abstract Fitness-for-Service (FFS) codes, such as ASME Boiler and Pressure Vessels Code, Section XI, have flaw characterization rules for evaluation of structural integrity. Since stress corrosion cracking (SCC) and thermal fatigue frequently cause multiple flaws, FFS codes should have proximity rules as a part of flaw characterization rules. The flaw characterization rules should consider fracture modes, such as brittle fracture, ductile fracture, and plastic collapse. Those in the current codes are not divided by the fracture modes. Especially, application of the current proximity rules to plastic collapse of non-aligned multiple flaws should be validated because there are few studies for this issue. Thus, fracture tests of flat plates with through-wall flaws and finite element analysis (FEA) were conducted for predicting collapse loads due to plastic collapse. A series of the fracture tests of flat plates with non-aligned two flaws has been conducted, and a trend between the load reduction and the flaw locations was shown from the results. This trend shows that the defined net-section for non-aligned multiple flaw dominate the collapse load. For the validation the trend shown by the fracture tests, FEA was performed for predicting the measured collapse load. Equivalent plastic strain around a flaw tip dominates a collapse behavior, and an equivalent plastic strain at collapse called as fracture strain was determined for FEA. The collapse loads predicted by the fracture strain are correspond with the test results for any flaw locations. FEA conditions can interpolate and cover a wide range of flaw locations conducted by the tests. The load ratios which represent effect of flaw interaction on a collapse load were estimated by parametric FEA. The ratios were mapped to investigate the trend of the effect on a collapse load. The mapped results show that the load ratio depends on a shorter flaw length of two flaws. This trend shown by the analysis results is corresponds with the fracture test results. These results are fundamental idea to make a flaw characterization rule in the FFS codes, such as ASME BPVC Section XI, for ductile fracture evaluation.

Evaluation Procedure of Limit Load for Non-Aligned Multiple Flaws

2013 ◽  
Author(s):  
Fuminori Iwamatsu ◽  
Katsumasa Miyazaki ◽  
Hidekazu Takazawa ◽  
Koichi Saito
Keyword(s):  
Correlation Coefficients ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Collapse Load ◽  
Evaluation Procedures ◽  
Direct Distance ◽  
Fracture Tests ◽  
Almost All ◽  
Technical Basis

The fitness-for-service codes such as the ASME Boiler and Pressure Vessel Code Section XI require the characterization of non-aligned multiple flaws for flaw evaluation, which is performed using a flaw alignment rule. Worldwide, almost all such codes provide their own alignment rule, often with an unclear technical basis regarding the application of the rule to plastic collapse due to ductile fracture as prescribed by limit load analysis based on a net-section approach. Therefore, evaluation procedures to calculate collapse load for non-aligned multiple flaws have been proposed in various experimental and analytical studies. In these proposals, a collapse load for non-aligned multiple flaws is evaluated using the net-section stress approach in consideration of the ratio of a distance between flaws to a flaw length parameter. However, because each study proposes its own appropriate length and distance parameters, which are based on a few experimental results limited to that study, the applicability of the proposed parameters to evaluation of collapse load for arbitrary flaw sizes and locations is unclear. In this study, we performed fracture test result on a flat plate with two through-wall flaws in order to determine appropriate parameters for the evaluation procedure of the collapse load for non-aligned multiple flaws. Appropriate parameters were determined by correlation coefficients obtained by comparison of maximum loads of fracture tests and collapse loads of evaluation with various parameters. We found that the appropriate parameters to apply the alignment rule with equations to evaluate collapse load for non-aligned flaws were the ratio of the vertical or direct distance between flaws to the maximum or average flaw length.

Validity of Assessment Procedure in p-M Method for Multiple Volumetric Flaws

2008 ◽  
Author(s):  
Shinji Konosu ◽  
Masato Kano ◽  
Norihiko Mukaimachi ◽  
Shinichiro Kanamaru
Keyword(s):  
Internal Pressure ◽  
Yield Stress ◽  
Bending Moment ◽  
Pressure Vessels ◽  
Assessment Procedure ◽  
Plastic Collapse ◽  
Storage Tanks ◽  
Collapse Load ◽  
The Status ◽  
Single Flaw

General components such as pressure vessels, piping, storage tanks and so on are designed in accordance with the construction codes based on the assumption that there are no flaws in such components. There are, however, numerous instances in which in-service single or multiple volumetric flaws (local thin areas; volumetric flaws) are found in the equipment concerned. Therefore, it is necessary to establish a Fitness for Service (FFS) rule, which is capable of judging these flaws. The procedure for a single flaw or multiple flaws has recently been proposed by Konosu for assessing the flaws in the p–M (pressure-moment) Diagram, which is an easy way to visualize the status of the component with flaws simultaneously subjected to internal pressure, p and external bending moment, M due to earthquake, etc. If the assessment point (Mr, pr) lies inside the p–M line, the component with flaws is judged to be safe. In this paper, numerous experiments and FEAs for a cylinder with external multiple volumetric flaws were conducted under (1) pure internal pressure, (2) pure external bending moment, and (3) subjected simultaneously to both internal pressure and external bending moment, in order to determine the plastic collapse load at volumetric flaws by applying the twice-elastic slope (TES) as recommended by ASME. It has been clarified that the collapse (TES) loads are much the same as those calculated under the proposed p–M line based on the measured yield stress.

Introduction of CASS Pipe Flaw Evaluation of JSME Rules on Fitness for Service

2019 ◽  
Author(s):  
Kiminobu Hojo
Keyword(s):  
Stainless Steel ◽  
Evaluation Method ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Screening Criteria ◽  
Steel Pipes ◽  
Evaluation Procedures ◽  
Degradation Models ◽  
Fracture Tests ◽  
Cast Stainless Steel

Abstract This paper summarizes the revised flaw evaluation procedures for cast austenitic stainless steel (CASS) pipe of the Japan Society of Mechanical Engineers (JSME) rules on fitness for service (FFS) in 2018 addenda. The revision includes the introduction of thermal aging degradation models for stressstrain curve and fracture resistance (J-R) curve, application of a screening criteria for the fracture evaluation procedure of cast stainless steel pipes, and introduction of a new critical stress parameter for the limit load evaluation method of a shallow flaw with a flaw depth to thickness ratio of less than or equal to 0.5. These revisions are based on a large database of specimen tests and several fracture tests of flat plate and large pipe models using thermally aged material, which have already been published.

Limit Load and Reference Stress for Curved Wide Plates

2010 ◽  
Author(s):  
Stijn Hertele´ ◽  
Wim De Waele ◽  
Rudi Denys ◽  
Jeroen Van Wittenberghe ◽  
Matthias Verstraete
Keyword(s):  
Experimental Validation ◽  
Limit Load ◽  
Plastic Collapse ◽  
Test Results ◽  
Finite Element Analyses ◽  
Flat Plates ◽  
Stress Equation ◽  
Reference Stress ◽  
Plate Curvature ◽  
Girth Welds

Curved wide plates are a valuable tool in the assessment of defective pipeline girth welds under tension. Throughout the years, Laboratory Soete collected an extensive database of curved wide plate test results. In an effort to investigate these results through FAD analysis, the authors recently developed a reference stress equation for curved plates. The approach followed is similar to the development of the Goodall and Webster equation for flat plates. This paper elaborates finite element analyses of the equation’s capability to predict plastic collapse. It is found that, although overestimated, the influence of plate curvature is correctly predicted in a qualitative way. For all simulations, the curved plate reference stress equation produced conservative estimations. This indicates that the proposed equation is suited to safely predict the plastic collapse of defective pipeline girth welds. An experimental validation is underway.

Prediction of Radiation Embrittlement for Highly Irradiated Reactor Pressure Vessel Steels

2009 ◽  
Author(s):  
Hiroshi Matsuzawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Test Results ◽  
Prediction Formula ◽  
Irradiation Embrittlement ◽  
Nuclear Plants ◽  
Domestic Plant ◽  
Test Reactor ◽  
Reactor Pressure

There are 53 (fifty-three) nuclear power plants (both PWR and BWR type) are now under operating in Japan, and the oldest plant has been operating more than thirty years. These plants will be operated until sixty years for operation periods, and will be verified the integrity for assessment of nuclear plants for every ten years in Japan. Reactor Pressure Vessels (RPVs) are required to evaluate the reduction of fracture toughness and the increase of the reference temperature in the transition region. As the operating period will be longer, the prediction for these material properties will be more important. Recently the domestic prediction formula of embrittlement was revised based on the database of domestic plant surveillance test results for thirty years olds as the JEAC4201-2007 [7]. The adequacy for this prediction formula using for sixty year periods is verified by use of the results of the material test reactors (MTRs), but the effects of the accelerated irradiation on embrittlement has not been clear now. So, JNES started the national project, called as “PRE” project on 2005 in order to investigate how flux influences on the ΔRTNDT. In this project the RPV materials irradiated in the actual PWR plant have been re-irradiated in the OECD/Halden test reactor by several different fluxes up to the high fluence region, and the microstructual change for these materials will be investigated in order to make clear the cause of the irradiation embrittlement. In this paper the overall scheme of this project and the summary of the updated results will be presented.

Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using a Newly Developed Statistical Thermal Aging Model

2019 ◽  
Author(s):  
M. Uddin ◽  
C. Sallaberry ◽  
G. Wilkowski
Keyword(s):  
Statistical Model ◽  
Failure Modes ◽  
Austenitic Stainless Steels ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Compositional Variation ◽  
Chemical Compositions ◽  
Test Results ◽  
Thermal Embrittlement ◽  
Reactor Operating

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures can lead to a reduction in the fracture toughness and increase in strength. Some aged CASS materials have the potential to have exceedingly low toughness and also show high variability due to the nature of their microstructure or compositional variation within the casting. Because of their low aged toughness with the variability, flaw evaluations of CASS material need to be done with an understanding of the materials aged condition, especially since most US PWR nuclear plants have been given plant life extensions for 60-year operation, and consideration of further extension to 80 years is underway. In this paper, a flaw evaluation procedure for CASS materials is presented using a new statistical model developed to predict the toughness of fully aged CASS using the material’s chemical composition. The new statistical model was developed based on the experimental toughness using standard 1T CT specimens (generally in the L-C orientation) at 288C to 320C and chemical compositions of the CF8m CASS materials. While the detail development of the model is beyond the scope of this paper, a brief validation of predicted toughness using chemical compositions is presented in this paper. Using the predicted toughness, a flaw evaluation procedure was developed using the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) that needs to be applied to the limit-load solution to predict the failure stress for CASS pipe and fittings with a circumferential surface crack. Variability within a single casting was also determined from available test results which was included in the procedure to determine Z-factor. The procedure was then validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes (i.e., limit load or EPFM), etc. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in US PWR plants (whose chemical compositions were known) — essentially solving 1096 sample problems to understand what range of Z-factors might exists in US PWR plants (for CF8m CASS materials) considering all variations in pipe dimensions, ferrite contents, materials’ toughness, etc. Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also investigated by comparing the predictions with available test results.

Development of an Integrity Evaluation System for Steam Generator Tubes in a Nuclear Power Plant

2007 ◽  
Vol 120 ◽  
pp. 157-162
Author(s):  
J.C. Kim ◽  
Sang Min Lee ◽  
Yoon Suk Chang ◽  
Jae Boong Choi ◽  
Young Jin Kim ◽  
...  
Keyword(s):  
Steam Generator ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Efficient System ◽  
Reference Pressure ◽  
Steam Generator Tubes

Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an efficient system named as STiES was developed employing three representative integrity evaluation methods: fracture mechanics analysis (crack driving force diagram and J-integral/Tearing modulus method) and limit load method. Exemplary analyses for steam generator tubes with various types of flaws showed good applicability of the proposed integrity evaluation system. So, it is anticipated that the system can be used for the calculation of reference pressure to decide either the continued operation or repair until next outage.

Development of Evaluation Method on Integrity of RPV at the Upper Shelf Region

2002 ◽  
Author(s):  
Kazunobu Sakamoto ◽  
Shunichi Hatano
Keyword(s):  
Fracture Toughness ◽  
Nuclear Power ◽  
Power Plants ◽  
Energy Use ◽  
Evaluation Method ◽  
Integrated Management ◽  
Pressure Vessels ◽  
Test Results ◽  
Irradiation Embrittlement ◽  
Use Values

In order to develop the integrity evaluation technology for aged major components of nuclear power plants, the Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project named “Nuclear Power Plant Integrated Management Technology (PLIM)” entrusted by Japanese Ministry of Economy, Trade and Industry (METI) since 1996Fiscal Year (FY). One of the objectives of this project is to establish the method for integrity evaluation of aged domestic reactor pressure vessels (RPV) by developing the prediction equations of reduction of Charpy V-notch Upper Shelf Energy (USE) due to neutron irradiation embrittlement and the correlation equations between USE and fracture toughness. Because the tests are now in progress, this paper presents the following preliminary results as of the end of 2000FY, using irradiated Charpy and Compact Tension specimens of RPV materials. • Study on the effect of accelerated irradiation for test specimens. • Comparison of USE values between test results and U.S. surveillance data. • Comparison of USE values between test results and published predictions. • Correlation between USE value and fracture toughness.

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