Mechanistic Studies on Environmentally-Assisted Fatigue Crack Growth in Light Water Reactor Environments

2016 ◽  
Author(s):  
Alexandra Panteli ◽  
Norman Platts ◽  
David R. Tice
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
High Temperature ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
The Impact

Fatigue crack growth of austenitic stainless steels can be enhanced significantly in high temperature light water reactor coolant environments and an ASME Code Case, N-809, has recently been developed to provide fatigue crack growth rate curves for these alloys in pressurized water reactor environments. However, under some conditions, the enhanced rates can decrease to rates close to those in air at long rise times, a process referred to as retardation which is not taken account of in the Code Case. An improved understanding of the mechanisms of both enhancement and retardation would be beneficial to determining whether advantage could be taken of these retarded rates in plant assessment. A number of studies have been undertaken to evaluate fatigue crack growth behavior in both air and water environments in order to provide mechanistic insight. Progress on this work will be described. The data from air and inert environments support the proposed mechanism of environmentally enhanced fatigue by environmental enhancement of planar slip, although it is not yet possible to differentiate between the impact of oxidation and corrosion hydrogen on the level of enhancement in aqueous environments. Testing in high temperature water environments suggests that both corrosive blunting and/or oxide-induced closure mechanisms may contribute to crack growth rate retardation under specific circumstances.

Effect of Loading Waveform and Spectrum Loading on the Fatigue Crack Growth Rate in Simulated Light Water Reactor Environments

2016 ◽  
Author(s):  
Norman Platts ◽  
Keith Rigby ◽  
David R. Tice ◽  
David I. Swan
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Rise Time ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Rate Retardation

High temperature water environments, typical of light water reactor primary coolant, are known to lead to significant environmental enhancement of fatigue crack growth of austenitic stainless steels. For PWR environments. these effects have recently been codified in ASME Code Case N-809. However, just as for the detrimental effect of these environments on fatigue endurance, plant experience indicates that crack growth rates must be significantly lower than predictions based on laboratory data using simple sawtooth waveforms. In order to explain this discrepancy, a significant amount of research has been conducted to quantify factors leading to crack growth rate retardation with sulfur content having been identified as significant in promoting crack growth rate retardation. However, the inherent conservatisms in current analysis techniques may be just as significant in generating the perceived over-conservatism of environmental fatigue crack growth laws such as Code Case N-809. The current work looks at the impact of waveform shape and spectrum loading on the level of environmental enhancement for a given stress intensity factor range and total rise time by considering simplified transients and loading spectra. The observations suggest that simplified definitions of total rise time used in fatigue assessments can lead to large over-estimation of actual fatigue damage. These data form the basis of an analytical methodology being developed by RollsRoyce (presented in a separate paper at this conference) aimed at partitioning damage across the loading cycle in order to remove over-conservatisms in current analytical methodologies.

Application of Corrosion Fatigue Crack Growth Rate Data to Integrity Analyses of Nuclear Reactor Vessels

1979 ◽  
Vol 101 (3) ◽  
pp. 182-190 ◽  
Author(s):  
W. H. Bamford
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Water Environment ◽  
Reference Curve ◽  
Reactor Water ◽  
Reference Curves ◽  
The Impact

The methodology of fatigue crack growth analysis in evaluating structural integrity of nuclear components has been well established over the years, even to the point where a recommended practice has been incorporated in Appendix A to Section XI of the ASME Code. The present reference curve for crack growth rates of pressure vessel steels in reactor water environment was developed in 1973, and since that time a great deal of data have become available. The original curve was meant to be a bounding curve, and some recent data have exceeded it, so an important question to address is which reference curve to use for these calculations. The important features of fatigue crack growth behavior in a reactor water environment are reviewed, along with some suggested explanation for the observed environmental enhancement and overall trends. The variables which must be accounted for in any reference crack growth rate curve are delineated and various methods for accomplishing this task are discussed. Improvements to the present reference curve are suggested, and evaluated as to their accuracy relative to the present curve. The impact of the alternative curves is also evaluated through solution of an example problem. A discussion of the conservatisms included in the alternative reference curves as compared with the present reference curve is included. Also research work is identified which could lead to further improvement in the reference curves.

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