Experimental and Analytical Studies on the Effect of Excessive Loading on Fatigue Crack Propagation in Piping Materials

2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
Yinsheng Li ◽  
Genki Yagawa
Keyword(s):  
Fatigue Crack ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Structural Integrity ◽  
Large Earthquake ◽  
Retardation Effect ◽  
Crack Growth Behavior ◽  
Fatigue Crack Growth Test

The seismic design review guide in Japan was revised in September 2006 to address the occurrence of a large earthquake beyond the design basis. In addition, Japanese nuclear power plants (NPPs) experienced multiple large earthquakes, such as Niigata-ken Chuetsu-Oki Earthquake in 2007 and the Great East Japan Earthquake in 2011. Therefore, it is very important to assess the structural integrity of reactor piping under such a large earthquake when a crack exists in the piping. In this work, crack growth behavior after excessive loading during the large-scale earthquake were experimentally and analytically evaluated for carbon steel and austenitic stainless steel. Some cyclic loading patterns with increasing and decreasing load amplitudes and maximum loads were applied to fatigue crack growth test specimens. From the results, the retardation of crack growth rate was clearly observed after excessive loading. In addition, the applicability to the retardation effect of the modified Wheeler model was confirmed. It is also concluded that the retardation effect has little influence on the failure probability due to seismic loading using probabilistic fracture mechanics (PFM) analyses with the modified Wheeler model.

Study on Banded Structure in Low Carbon Vessel Plate for Nuclear Power Plant

2010 ◽  
Vol 44-47 ◽  
pp. 1763-1766
Author(s):  
Fei Xue ◽  
Zhi Feng Luo ◽  
Wei Wei Yu ◽  
Zhao Xi Wang ◽  
Lu Zhang
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Tests ◽  
Crack Growth Behavior ◽  
Low Carbon ◽  
Banded Structure ◽  
Crack Retardation

In this paper, the role of the pearlite-banded structure on fatigue crack growth behavior was investigated on carbon vessel plate material SA516, which is commonly used in the nuclear power plants. Along pearlite-banded orientation, in situ fatigue tests indicate that the crack initiated and propagated in the ferrite and then extended along the ferrite-pearlite interface when it met pearlitic colony. For comparison, the cyclic loading was also carried out perpendicular to the banding direction of the microstructure, and an intense crack branching was observed which led to fatigue crack retardation. Besides, the orientation perpendicular to banded pearlite in the investigated ferrite-pearlite steel was found to have a lower fatigue crack growth rate.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs: Fatigue Crack Growth Analysis of Hydrogen Flakes

2017 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Simplified Approach ◽  
Integrity Assessment ◽  
The Core

During the 2012 outages at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the Reactor Pressure Vessels (RPVs). The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated. The Structural Integrity Assessment of the RPVs, through the Flaw Acceptability Assessment, aimed at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This demonstration has been done on the basis of a specific methodology inspired by the ASME B&PV Code Section XI procedure but adapted to the nature and the number of indications found in the Doel 3 and Tihange 2 RPVs. As requested by Article IWB-3610(a) of ASME B&PV Code Section XI, one of the parts that have to be addressed through the Flaw Acceptability Assessment is the Fatigue Crack Growth (FCG) Analysis of the flaws in the core shells until the end-of-service lifetime of the RPVs. Due to the large number of flaws in the core shells, a specific methodology has been developed in order not to perform the FCG Analysis of each flaw separately. The paper describes this simplified approach aiming at distributing the flaws according to their inclination and at defining envelope flaws covering the actual flaws to carry out FCG Analysis. Furthermore, the paper highlights and quantifies the conservatisms of this analysis leading finally to demonstrate that the FCG of hydrogen flakes is not a concern in Doel 3 and Tihange 2 RPVs.

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Experimental and Analytical Studies on the Effect of Excessive Loading on Fatigue Crack Propagation in Piping Materials

2009 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
Hideharu Sugino ◽  
Yinsheng Li ◽  
...  
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Large Scale ◽  
Fatigue Cracks ◽  
Plastic Region ◽  
Retardation Effect ◽  
Seismic Motion ◽  
Loading Patterns ◽  
Piping Systems

The seismic regulatory guide was revised in September 2006 and the Niigata-ken Chuetsu-oki earthquake, whose magnitude was beyond the design base seismic motion, occurred in July 2007. Due to these events, attention is being drawn to the evaluation of the effects of large scale earthquakes for some piping systems in which SCC and/or fatigue cracks may potentially occur. Many papers have been already published about the retardation effect that excessive loading has on fatigue crack growth. The retardation effect is treated qualitatively in regard to the plastic strain generated by excessive loading. In this work, crack growth after excessive loading is evaluated for carbon steel and austenitic stainless steel. Some cyclic excessive loading patterns such as stepwise increases or decreases were applied to fatigue crack growth experiments. The FEM analyses were conducted to evaluate the plastic region size during these loading conditions. PFM analyses were performed to evaluate the extent to which the retardation of crack growth influences the probability of failure.

scholarly journals An Analytical Model for Fatigue Crack Propagation Prediction with Overload Effect

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Shan Jiang ◽  
Wei Zhang ◽  
Xiaoyang Li ◽  
Fuqiang Sun
Keyword(s):  
Fatigue Crack ◽  
Plastic Zone ◽  
Crack Growth ◽  
Crack Closure ◽  
Retardation Effect ◽  
Crack Growth Behavior ◽  
Zone Model ◽  
Constant Amplitude ◽  
Proposed Model ◽  
Single Overload

In this paper a theoretical model was developed to predict the fatigue crack growth behavior under the constant amplitude loading with single overload. In the proposed model, crack growth retardation was accounted for by using crack closure and plastic zone. The virtual crack annealing model modified by Bauschinger effect was used to calculate the crack closure level in the outside of retardation effect region. And the Dugdale plastic zone model was employed to estimate the size of retardation effect region. A sophisticated equation was developed to calculate the crack closure variation during the retardation area. Model validation was performed in D16 aluminum alloy and 350WT steel specimens subjected to constant amplitude load with single or multiple overloads. The predictions of the proposed model were contrasted with experimental data, and fairly good agreements were observed.

Fatigue Fracture Mechanics Modeling and Structural Integrity Assessment of Offshore Welded Tubular Joints

1989 ◽  
Vol 111 (3) ◽  
pp. 170-176 ◽  
Author(s):  
J. C. P. Kam ◽  
D. A. Topp ◽  
W. D. Dover
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Fracture Mechanics ◽  
Nondestructive Testing ◽  
Crack Growth ◽  
Large Scale ◽  
Structural Integrity ◽  
Offshore Structures ◽  
Integrity Assessment ◽  
Tubular Joints

Evaluation of the structural integrity of offshore structures requires information on the reliability of nondestructive testing, the accuracy of fatigue crack growth modeling and other data. The University College London Underwater NDE Centre has been set up to provide information on the effectiveness and reliability of different nondestructive testing methods. To achieve this aim, a large library of cracked specimens will be assembled. In the preliminary phase of producing this library, a series of large-scale welded tubular joints were fatigue tested and the crack growth was fully monitored with the ACPD technique. This paper will describe briefly the background to the crack library and present the data obtained from fatigue tests. It will also describe a new model for fatigue crack growth prediction in tubular joints using fracture mechanics. This model allows the prediction of the size effect noted previously in the stress/life curves for tubular joints.

Initiation and Growth of Small Fatigue Cracks of Steels Used for Nuclear Power Plants Under Low Cycle Regime

2014 ◽  
Author(s):  
Shota Hasunuma ◽  
Takeshi Ogawa
Keyword(s):  
Fatigue Crack ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Fatigue Cracks ◽  
Cycle Fatigue

Low cycle fatigue tests were conducted for carbon steel, STS410, low alloy steel, SFVQ1A, and austenitic stainless steel, SUS316NG, which were used for nuclear power plants, in order to investigate the mechanism of fatigue damage when the plants were subjected to huge seismic loads. In these tests, the surface behavior of fatigue crack initiation and growth was observed in detail using cellulose acetate replicas, while the interior behavior was detected in terms of fracture surface morphology developed by multiple two-step strain amplitude variations with periodical surface removals. Fatigue crack growth rates were evaluated by elasto-plastic fracture mechanics approach. For SFVQ1A and SUS316NG, the fracture mechanics approach is available in order to predict the crack growth life from the metallurgical crack initiation size to the final crack length of the specimens. For STS410, numerous small cracks initiated, grew and coalesced each other on the specimen surface under low cycle fatigue regime.

Structural Evaluation for Repaired J-Weld Portion of Reactor Vessel Head Penetration

2006 ◽  
Author(s):  
Seiji Asada ◽  
Kiminobu Hojo ◽  
Mayumi Ochi ◽  
Itaru Muroya ◽  
Hajime Ito
Keyword(s):  
Fracture Mechanics ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Nuclear Power Generation ◽  
Head Penetration ◽  
Structural Evaluation ◽  
Non Destructive

Leakage was found in a Reactor Vessel (RV) Head Penetration of Ohi unit 3 of the Kansai Electric Power Co., Inc. in May 4, 2004. Non-destructive examinations identified flaws in a J-weld portion of the Head Penetration. The J-weld portion was repaired by using Embedded Flaw Repair Technique [1] that performs welding of 52 weld metal on the J-weld surface remaining the flaws. In order to show the structural integrity of the J-weld portion, a fracture mechanics evaluation was performed in accordance with the Rules on Fitness-for-Service for Nuclear Power Plants of the JSME Codes for Nuclear Power Generation Facilities, JSME S NA1-2002 [2] (hereafter, the JSME Fitness-for-Service Rules) and literatures related. The flaw was characterized as both case of an embedded flaw and a surface flaw and KI for each flaw was directly calculated by using FE analysis. Fatigue crack growth analysis using KI calculated showed the amount of the crack growth was quite small. The fracture mechanics evaluation followed confirmed that the result satisfied the criteria. This paper explains the method and results for evaluation of the structural integrity of the J-weld portion.

STYLE: Project Overview

2010 ◽  
Author(s):  
Tomas Nicak ◽  
Elisabeth Keim
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Best Practice ◽  
Structural Integrity ◽  
Life Time ◽  
Assessment Tools ◽  
Integrity Assessment ◽  
University Of Bristol

The purpose of this paper is to introduce a new EUROATOM project focusing on the structural integrity assessment of reactor coolant pressure boundary components (RCPB) relevant to ageing and life time management. The project started in January 2010 and will last 4 years. The project is coordinated by AREVA NP GmbH with 20 partner organizations from Europe, one collaborator from USA and one collaborator from Russia: AEKI, Hungary; AREVA NP GmbH, Germany (coordination, WP2 leader); AREVA NP SAS, France; Bay Zoltan, Hungary; British Energy Generation Ltd., UK (WP7 leader); CEA, France (WP1 leader); EDF, France; IdS, France; INR, Romania; IWM, Germany; JRC, Netherlands (WP4 leader); NRI, Czech Republik; NRG, Netherlands; SCK-CEN, Belgium; Serco Assurance Technical Services, UK (WP3 and WP5 leader); University of Bristol, UK; University of Manchester, UK; Technatom, Spain; Vattenfall, Sweden (WP6 leader); VTT, Finland. Within STYLE (Structural integrity for lifetime management – non-RPV components) realistic failure models for some of the key components will be identified. The range of assessment tools considered will include those for assessment of component failure by advanced fracture mechanics analyses validated on small and large scale experiments, quantification of weld residual stresses by numerical analysis and by measurements, stress corrosion crack initiation/ growth effects and assessment of RCPB components (excluding the reactor pressure vessel) under dynamic and seismic loading. Based on theoretical and experimental results, performance assessment and further development of simplified engineering assessment methods (EAM) will be carried out considering both deterministic and probabilistic approaches. Integrity assessment case studies and large scale demonstration experiments will be performed on Mock-ups of safety-relevant components. These will include a repair weld in an aged butt-welded austenitic pipe, a dissimilar narrow gap TIG weld (following the EPR design) and a cladded ferritic pipe. Moreover experiments on specimens and feature test pieces will be carried out to support the large scale Mock-up analyses. The end product of the project (“STYLE TOOLS”) will comprise best practice guidelines on the use of advanced tools, on improvement and qualification of EAM as a part of European Leak-before-break (LBB) procedures and on life time management of the integrity of RCPB components in European nuclear power plants. The project will interact with the European Network of Excellence NULIFE.

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