Environmental Crack Growth Predictions for Piping Components

2007 ◽  
Author(s):  
Sam Ranganath ◽  
Guy DeBoo
Keyword(s):  
Stainless Steel ◽  
Crack Growth ◽  
Test Data ◽  
Structural Integrity ◽  
Reference Curve ◽  
Pressurized Water ◽  
Water Reactor ◽  
Integrity Assessment ◽  
Asme Code ◽  
Reference Curves

Structural integrity assessment of reactor components requires consideration of crack growth. A key input to this is the development of reference stress corrosion crack (SCC) growth rate curves for use in the structural evaluation. The ASME Section XI Task Group on SCC Reference Curve is looking into available SCC data for stainless steel and nickel based alloys and associated weldment in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. The test data show significant data scatter in crack growth rates (CGR). The conservative approach is to develop reference curves that bound all available data so that upper bound crack growth predictions. While this approach may be conservative, it may lead to excessive estimates of crack growth and result in unrealistic (and often meaningless) structural margin predictions. Selection of the appropriate SCC reference curves requires realistic interpretation of test data so that the predictions are consistent with field behavior and provide reasonable, but conservative assessment. This paper describes crack growth assessment for stainless steel piping and Alloy 600 safe end components with Alloy 182/82 welds in BWR environment. The results from the crack growth analysis for piping can be used to determine whether a proposed reference curve provides reasonable results. The objective is to use the piping and safe end crack growth predictions to develop optimal SCC Reference Curves for use in ASME Code evaluations.

The Background to Light Water Reactor Environment Fatigue Evaluation Methods for Austenitic Stainless Steel and Their Application to Sample Pressurised Water Reactor Components

2011 ◽  
Author(s):  
C. T. Watson
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Test Data ◽  
Cylindrical Specimen ◽  
Fatigue Curve ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Specimen Test ◽  
Asme Code

There is now a large amount of small cylindrical specimen test data, which indicates that in a Light Water Reactor (LWR) environment, compared to that in air, the fatigue life of stainless steel is significantly reduced. The current ASME III design fatigue curve does not explicitly include factors to account for a LWR environment. Using the available cylindrical specimen test data, methods for accounting for a LWR environment in fatigue assessments have been proposed in NuReg/CR-6909 and in two American Society of Mechanical Engineers (ASME) code cases. One of the code cases (N-792) uses a penalty factor (Fen) approach, similar to that in NuReg/CR-6909, another (N-761) utilizes a set of environmentally corrected fatigue curves. A third code case, which is still under development, uses a flaw tolerance approach. In this paper the background to the methods for correcting for a LWR environment in fatigue calculations is presented. The safety margin present in the ASME fatigue design methodology is discussed and a short review of civil nuclear plant operating and geometrical features testing experience provided. The NuReg/CR-6909 and ASME code case N-761 methods are applied to a number of ASME III Class 1 austenitic stainless steel components, and the cumulative usage factors calculated compared with those obtained using the ASME 2007 design fatigue curve. An objective of the paper is to highlight some of the issues arising out of applying the newly proposed methods to reactor plant components.

ASME Section XI Appendix L Flaw Tolerance Evaluation of Pressurized Water Reactor Piping Systems to Support Second License Renewal (80-Years Operation)

2018 ◽  
Author(s):  
Anees Udyawar ◽  
Charles Tomes ◽  
Alexandria Carolan ◽  
Steve Marlette ◽  
Thomas Meikle ◽  
...  
Keyword(s):  
Crack Growth ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Pressurized Water Reactor ◽  
Safe Operation ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
Piping Systems ◽  
Operating Plant

One of the goals of ASME Section XI is to ensure that systems and components remain in safe operation throughout the service life, which can include plant license extensions and renewals. This goal is maintained through requirements on periodic inspections and operating plant criteria as contained in Section XI IWB-2500 and IWB-3700, respectively. Operating plant fatigue concerns can be caused from operating conditions or specific transients not considered in the original design transients. ASME Section XI IWB-3740, Operating Plant Fatigue Assessments, provides guidance on analytical evaluation procedures that can be used when the calculated fatigue usage exceeds the fatigue usage limit defined in the original Construction Code. One of the options provided in Section XI Appendix L is through the use of a flaw tolerance analysis. The flaw tolerance evaluation involves postulation of a flaw and predicting its future growth, and thereby establishing the period of service for which it would remain acceptable to the structural integrity requirements of Section XI. The flaw tolerance approach has the advantage of not requiring knowledge of the cyclic service history, tracking future cycles, or installing systems to monitor transients and cycles. Furthermore, the flaw tolerance can also justify an inservice inspection period of 10 years, which would match a plant’s typical Section XI in-service inspection interval. The goal of this paper is to demonstrate a flaw tolerance evaluation based on ASME Section XI Appendix L for several auxiliary piping systems for a typical PWR (Pressurized Water Reactor) nuclear power plant. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth mechanism, such as fatigue crack growth, and the inspection detection capabilities. The purpose of the Section XI Appendix L evaluation is to demonstrate that a reactor coolant piping system continues to maintain its structural integrity and ensures safe operation of the plant.

Elastic-Plastic Fracture Mechanics Analysis of Cast Stainless Steel Pipes (Influence of Axial Load on Z-Factor of Pipes With Circumferential Crack Under Bending Load)

2013 ◽  
Author(s):  
Masayuki Kamaya ◽  
Kiminobu Hojo
Keyword(s):  
Stainless Steel ◽  
Axial Load ◽  
Power Plants ◽  
Structural Integrity ◽  
Term Operation ◽  
Steel Pipes ◽  
Integrity Assessment ◽  
Plastic Failure ◽  
Elastic Plastic ◽  
Bending Load

Since the ductility of cast austenitic stainless steel pipes decreases due to thermal aging embrittlement after long term operation, not only plastic collapse failure but also unstable ductile crack propagation (elastic-plastic failure) should be taken into account for the structural integrity assessment of cracked pipes. In the ASME Section XI, the load multiplier (Z-factor) is used to derive the elastic-plastic failure of the cracked components. The Z-factor of cracked pipes under bending load has been obtained without considering the axial load. In this study, the influence of the axial load on Z-factor was quantified through elastic-plastic failure analyses under various conditions. It was concluded that the axial load increased the Z-factor; however, the magnitude of the increase was not significant, particularly for the main coolant pipes of PWR nuclear power plants.

Rupture Hardware Minimization in Pressurized Water Reactor Piping

1989 ◽  
Vol 111 (1) ◽  
pp. 64-71 ◽  
Author(s):  
S. K. Mukherjee ◽  
J. J. Szy Slow Ski ◽  
V. Chexal ◽  
D. M. Norris ◽  
N. A. Goldstein ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Ferritic Steel ◽  
High Energy ◽  
Pressurized Water Reactor ◽  
Pipe Failure ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Water Reactor ◽  
Leak Before Break ◽  
Generic Sense

For much of the high-energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station—Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.

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.

Crack Growth Behavior of Nickel Alloy Welds in a PWR Environment

2007 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Crack Growth ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Crack Growth Behavior ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Ni Alloy ◽  
The Individual

A program is under way at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated Light Water Reactor (LWR) coolant environments. This paper focuses on the cracking behavior of Ni-alloy welds in simulated pressurized water reactor (PWR) environment at 290–350°C. Crack growth tests have been conducted on both field- and laboratory-produced welds. The results are compared with the existing crack-growth-rate (CGR) data for Ni-alloy welds to determine the relative susceptibility of specific Ni-alloy welds to environmentally enhanced cracking. To analyze the CGRs, a superposition model was used to establish the individual contributions of mechanical fatigue, corrosion fatigue, and stress corrosion cracking.

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