Probabilistic Leak-Before-Break Assessment of a Main Coolant Line

2010 ◽  
Author(s):  
Igor Varfolomeev ◽  
Denis Ivanov ◽  
Dieter Siegele ◽  
Gerhard Nagel
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Power Plants ◽  
Statistical Treatment ◽  
Crack Size ◽  
Additional Treatment ◽  
Pressurized Water ◽  
Crack Penetration ◽  
Leak Before Break

The paper presents results of a probabilistic leak-before-break (LBB) assessment of a ferritic main coolant line of a pressurized water reactor representative for German nuclear power plants. The analysis approach is based on the elastic-plastic fracture mechanics methodology, incorporating the failure assessment diagram to calculate the critical through-thickness crack size, as well as fatigue crack growth calculations to determine the flaw length at wall penetration. An essential part of this study is the collation and statistical treatment of material data, such as strength properties, crack resistance and fatigue crack growth curves, and their incorporation in the probabilistic assessment. The analysis yields negligible break probabilities, thus demonstrating the LBB behavior of the piping. This conclusion is validated by results of a sensitivity study with additional conservative assumptions and simplifications with respect to the initial crack size and material state. For the sake of simplicity, the blocking effect of the intact austenitic cladding on the crack extension, as well as the conditional probability of crack penetration through the wall during the service life are neglected in this investigation. Through an additional treatment of these issues a more realistic assessment is achieved, resulting in even smaller failure probabilities.

Comparison of Leak-Before-Break Assessment of Main Loop Piping Lines Fabricated of Different Materials

2016 ◽  
Author(s):  
Linwei Ma ◽  
Xiaotao Zheng ◽  
Yan Wang ◽  
Jiasheng He ◽  
Anqing Shu
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Material Properties ◽  
Plastic Instability ◽  
Crack Size ◽  
Critical Crack ◽  
Main Loop ◽  
Critical Crack Size ◽  
Leak Before Break

Leak-Before-Break (LBB) assessment is used for the design of nuclear reactor coolant system main loop piping to lower the cost of construction and operation in China. In these applications, the materials of main loop piping lines are cast austenitic stainless steel (CASS) or wrought stainless steel (WSS) due to the different type of reactor design. According to US.NRC SRP3.6.3, LBB assessment includes two major calculations, such as critical crack size calculation and leakage flaw size calculation. The elastic-plastic instability analysis or plastic instability analysis is chosen to perform critical size calculation depending on material properties, especially fracture toughness. In this paper, LBB assessment in the guidance of SRP 3.6.3 was performed to evaluate main loop piping lines of CASS and WSS. The JR curve tests and the adjustment due to thermal aging are performed to achieve reasonable material properties. J integral/tearing modulus approach is used to determine critical crack size of CASS pipe and net section collapse (NSC) approach is used to determine critical crack size of WSS pipe. Leakage flaw size under 1gpm leakage detection capability is determined based on Henry’s homogeneous nonequilibrium critical flow model. In order to demonstrate that fatigue crack growth is not a potential source of pipe rupture for the evaluated piping lines, the fatigue crack growth of a postulated circumferential part-through-wall crack under nuclear power plant full life time operating transients and the fatigue crack growth of a circumferential through-wall crack under one time safe shutdown seismic are analyzed. And the LBB assessment procedure and results of CASS pipe and WSS pipe are compared.

Characterization of Crack Tip Damage Zone Formation on Alloy 625 During Fatigue Crack Growth at 750°C by Transmission EBSD Method

2017 ◽  
Author(s):  
Yuji Ozawa ◽  
Tatsuya Ishikawa ◽  
Yoichi Takeda
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Power Plants ◽  
Crack Path ◽  
Damage Zone ◽  
Crack Tip ◽  
Loading Frequency ◽  
Zone Formation ◽  
Alloy 625

In order to clarify the mechanism of fatigue crack growth in alloy 625, which is a candidate material for use in advanced ultra supercritical power plants, the crack tip damage zone formation after a crack growth test conducted in high temperature steam was investigated. It was observed that the oxide thickness at the crack tip tended to increase with decreasing cyclic loading frequency. The crack path was a mix of transgranular and intergranular fractures. According to the grain reference orientation deviation (GROD) maps, it was revealed that the density of geometrically necessary dislocations (GNDs) in the matrix along the crack path and ahead of crack tip increased with an increase in the fatigue crack growth rate (FCGR) due to environmental effects. It was observed that (1) mobile dislocations at the crack surface were blocked due to the thick oxide layer, resulting in an increase in the density of GNDs, and (2) an increase in the density of GNDs might induce stress concentration at the crack tip, deformation twinning, and the acceleration of FCGRs.

Comparison of International Codes for a Fatigue Crack Growth Flaw Tolerance Sample Problem

2021 ◽  
Author(s):  
Gary L. Stevens
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Large Diameter ◽  
Sample Problem ◽  
Pressurized Water ◽  
Typical Application ◽  
Flaw Tolerance ◽  
Asme Code ◽  
The Impact

Abstract As part of the development of American Society of Mechanical Engineers Code Case N-809 [1], a series of sample calculations were performed to gain experience in using the Code Case methods and to determine the impact on a typical application. Specifically, the application of N-809 in a fatigue crack growth analysis was evaluated for a large diameter austenitic pipe in a pressurized water reactor coolant system main loop using the current analytical evaluation procedures in Appendix C of Section XI of the ASME Code [2]. The same example problem was previously used to evaluate the reference fatigue crack growth curves during the development of N-809, as well as to compare N-809 methods to similar methods adopted by the Japan Society of Mechanical Engineers. The previous example problem used to evaluate N-809 during its development was embellished and has been used to evaluate additional proposed ASME Code changes. For example, the Electric Power Research Institute investigated possible improvements to ASME Code, Section XI, Nonmandatory Appendix L [3], and the previous N-809 example problem formed the basis for flaw tolerance calculations to evaluate those proposed improvements [4]. In addition, the ASME Code Section XI, Working Group on Flaw Evaluation Reference Curves continues to evaluate additional research data and related improvements to N-809 and other fatigue crack growth rate methods. As a part of these Code investigations, EPRI performed calculations for the Appendix L flaw tolerance sample problem using three international codes and standards to evaluate fatigue crack growth (da/dN) curves for PWR environments: (1) ASME Code Case N-809, (2) JSME Code methods [5], and (3) the French RSE-M method [6]. The results of these comparative calculations are presented and discussed in this paper.

Fatigue Crack Growth Curve for Austenitic Stainless Steels in PWR Environment

2004 ◽  
Author(s):  
Yuichiro Nomura ◽  
Kazuya Tsutsumi ◽  
Hiroshi Kanasaki ◽  
Naoki Chigusa ◽  
Kazuhiro Jotaki ◽  
...  
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Growth Curve ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Stress Intensity Factor Range ◽  
Reference Curve ◽  
Pressurized Water ◽  
Crack Growth Curve

Although reference fatigue crack growth curves for austenitic stainless steels in air environments and boiling water reactor (BWR) environments were prescribed in JSME S NA1-2002, similar curves for pressurized water reactors (PWR) were not prescribed. In order to propose the reference curve in PWR environment, fatigue tests of austenitic stainless steels in simulated PWR primary water environment were carried out. According to the procedure to determine the reference fatigue crack growth curve of BWR, which of PWR is proposed. The reference fatigue crack growth curve in PWR environment have been determines as a function of stress intensity factor range, Temperature, load rising time and stress ratio.

A Probabilistic Micromechanical Code for Predicting Fatigue Life Variability: Model Development and Application

2005 ◽  
Vol 128 (4) ◽  
pp. 889-895 ◽  
Author(s):  
K. S. Chan ◽  
M. P. Enright
Keyword(s):  
Fatigue Crack ◽  
Fatigue Life ◽  
Fatigue Crack Growth ◽  
Crack Initiation ◽  
Crack Growth ◽  
Model Development ◽  
Fatigue Crack Initiation ◽  
Crack Size ◽  
Crack Initiation Life ◽  
Variability Model

This paper summarizes the development of a probabilistic micromechanical code for treating fatigue life variability resulting from material variations. Dubbed MICROFAVA (micromechanical fatigue variability), the code is based on a set of physics-based fatigue models that predict fatigue crack initiation life, fatigue crack growth life, fatigue limit, fatigue crack growth threshold, crack size at initiation, and fracture toughness. Using microstructure information as material input, the code is capable of predicting the average behavior and the confidence limits of the crack initiation and crack growth lives of structural alloys under LCF or HCF loading. This paper presents a summary of the development of the code and highlights applications of the model to predicting the effects of microstructure on the fatigue crack growth response and life variability of the α+β Ti-alloy Ti-6Al-4V.

Study on Evaluation Accuracy Required for Stress Intensity Factor Solution

2013 ◽  
Author(s):  
Naoki Miura ◽  
Masaki Nagai ◽  
Masaki Shiratori
Keyword(s):  
Stress Intensity Factor ◽  
Fatigue Crack ◽  
Stress Intensity ◽  
Fatigue Crack Growth ◽  
Intensity Factor ◽  
Crack Growth ◽  
Power Plants ◽  
Growth Analysis ◽  
Correction Factors ◽  
Evaluation Accuracy

Stress intensity factor solutions are often used as a dominant fracture mechanics parameter for fatigue crack growth analysis. In ASME Boiler and Pressure Vessel Code Section XI as well as JSME Rules on Fitness-for-Service for Nuclear Power Plants, fatigue crack growth is predicted on the basis of the stress intensity factor range. Stress intensity factor solutions are frequently provided by the correction factors, which are tabulated as the functions of structure and/or crack sizes. In this study, the effect of the variation of the correction factors on the crack growth analysis results was investigated for pipes with surface cracks. The evaluation accuracy required for the correction factors of the stress intensity factor solutions was then examined and recommended from the comparison with the necessary accuracy of the parameters used for the fatigue crack growth analysis.

Technical Basis for Revision of Code Case N-809 on Reference Fatigue Crack Growth Curves for Austenitic Stainless Steels in Pressurized Water Reactor Environments

2021 ◽  
Author(s):  
Russell C. Cipolla ◽  
Warren H. Bamford ◽  
Kiminobu Hojo ◽  
Yuichiro Nomura
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Stainless Steels ◽  
Growth Curves ◽  
Austenitic Stainless Steels ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
R Ratio ◽  
Technical Basis

Abstract Reference fatigue crack growth curves for austenitic stainless steels exposed to pressurized water reactor environments have been available in the ASME Code, Section XI in their present form with the publication of Code Case N-809 in Supplement 2 to the 2015 Code Edition. The reference curves are dependent on temperature, loading rate (loading rise time), mean stress (R-ratio), and cyclic stress intensity factor range (ΔK), which are all contained in the model. Since the first implementation of this Code Case, additional data have become available, and the purpose of this paper is to provide the technical basis for revision of the Code Case. Changes have been made in three areas: R-ratio behavior, threshold for crack growth (ΔKth), and crack growth rate dependence on ΔK. In addition, the temperature model was revisited to study the temperature effects for T < 150°C, where the current model predicts an increase in da/dN based on limited test data at about 100°C (200°F). At this point, the current temperature model is considered conservative and no change is proposed in this revision to N-809. The R-ratio model has been revised for both high and low carbon stainless steels, a significant improvement over the original procedures. Perhaps the most important revision is in the area of the threshold for the initiation of fatigue crack growth; such data are difficult to obtain, and the previous model was very conservative. Finally, the crack growth exponent was revised slightly to make it consistent with the regression analysis of the original data.

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