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.

Technical Basis of Fatigue Crack Growth Rate Curve for Ni-Base Alloy in BWR Water Environment

2014 ◽  
Author(s):  
Takuya Ogawa ◽  
Masao Itatani ◽  
Toshiyuki Saito ◽  
Hiroshi Nagase ◽  
Satoru Aoike ◽  
...  
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Fatigue Crack Growth Rate ◽  
Base Alloy ◽  
Water Environment ◽  
Reference Curve ◽  
Rising Time

When the flaws are detected in Japanese nuclear power components by in-service inspection, structural integrity assessment are performed in the technical judgment on continuous service. If cyclic loading is assumed, fatigue crack growth analysis should be conducted based on the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code). However, fatigue crack growth analysis for BWR components consisting of Ni-base alloy is currently impossible, since the reference curve of fatigue crack growth rate for Ni-base alloy in BWR water environment is not yet prescribed in the JSME FFS Code. In this study, fatigue crack growth behavior of Ni-base alloy used for Japanese BWR plants in BWR water environment was investigated. Based on the experimental data, the fatigue crack growth rate curve was evaluated. Four test parameters of material, corrosion potential, stress ratio and load rising time were considered. As a result of fatigue crack growth tests, the effects of all test parameters on the fatigue crack growth behavior were found. A Mean curve of fatigue crack growth rate in Paris law format, which was a function of stress ratio and rising time, was formulated based on crack growth data in normal water chemistry (corrosion potential was over 150 mVSHE) for weld metal and heat affected zone (HAZ), respectively. A reference curve of fatigue crack growth rate was also formulated by the statistical treatment considering the scatter of crack growth rate. Further, in order to determine the threshold stress intensity factor range ΔKth of reference curve of fatigue crack growth, ΔK decreasing tests were conducted under the test condition of 1 second of rising time. As a result, the threshold value of ΔK was evaluated based on the ASTM E 647, and the ΔKth of the reference curve was conservatively determined considering the margin.

Influence of reactor water of nominal parameters on fatigue crack growth rate in 15Kh2NMFA steel

1985 ◽  
Vol 21 (2) ◽  
pp. 130-133
Author(s):  
V. I. Pokhmurskii ◽  
A. S. Zubchenko ◽  
A. A. Popov ◽  
I. P. Gnyp ◽  
V. M. Timonin ◽  
...  
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Fatigue Crack Growth Rate ◽  
Reactor Water ◽  
15Kh2nmfa Steel

A Review of Fatigue and SCC Crack Growth Rate Relationships for Ferritic and Stainless Steels and Ni-Cr-Fe Materials in BWR Environment

2006 ◽  
Author(s):  
Hardayal S. Mehta
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Fatigue Crack Growth Rate ◽  
Corrosion Fatigue ◽  
Full Range ◽  
Water Environment ◽  
Plant Component

When in-service inspection of a nuclear plant component reveals the presence of cracking, an engineering evaluation (typically called a justification for continued operation, or JCO) is required to demonstrate the structural suitability for continued operation. A key element in such a flaw evaluation is the projected crack growth over the period when the cracked component will be reinspected. The crack growth is expected to be a combination of stress corrosion cracking (SCC) and corrosion fatigue. The ASME Section XI Code is in the process of developing a full range of SCC and corrosion fatigue crack growth rate relationships (CGRs) for stainless steel and Ni-Cr-Fe materials. The objective of this paper is to summarize several available SCC and fatigue crack growth rate relationships for these materials exposed to boiling water reactor (BWR) water environments. For completeness, low alloy steel SCC and corrosion fatigue CGRs in BWR water environment are also briefly reviewed. Two example evaluations are provided that used some of these CGRs in developing the JCOs for BWR components. A detailed comparison of these CGRs along with a review of the underlying data will be part of a future effort undertaken by the ASME Section XI Task Group.

Fatigue Crack Growth Curve for Austenitic Stainless Steels in BWR Environment

2000 ◽  
Vol 123 (2) ◽  
pp. 166-172 ◽  
Author(s):  
M. Itatani ◽  
M. Asano ◽  
M. Kikuchi ◽  
S. Suzuki ◽  
K. Iida,
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Fatigue Crack Growth Rate ◽  
Stainless Steels ◽  
Stress Ratio ◽  
Austenitic Stainless Steels ◽  
Reference Curve

Fatigue crack growth data obtained in the simulated BWR water environment were analyzed to establish a formula for reference fatigue crack growth rate (FCGR) of austenitic stainless steels in BWR water. The effects of material, mechanical and environmental factors were taken into the reference curve, which was expressed as: da/dN=8.17×10−12s˙Tr0.5s˙ΔK3.0/1−R2.121≦ΔK≦50 MPam where da/dN is fatigue crack growth rate in m/cycle, Tr is load rising time in seconds, ΔK is range (double amplitude) of K–value in MPam, and R is stress ratio. Tr=1 s if Tr<1 s, and Tr=1000 s if Tr cannot be defined. ΔK=Kmax−Kmin if R≧0.ΔK=Kmax if R<0.R=Kmin/Kmax. The proposed formula provides conservative FCGR at low stress ratio. Although only a few data show higher FCGR than that by proposed formula at high R, these data are located in a wide scatter range of FCGR and are regarded to be invalid. The proposed formula is going to be introduced in the Japanese Plant Operation and Maintenance Standard.

Fatigue Crack Growth Rate for Nickel-Based Alloys in PWR Environment

2007 ◽  
Author(s):  
Yuichiro Nomura ◽  
Hiroshi Kanasaki
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Fatigue Crack Growth Rate ◽  
Stress Ratio ◽  
Water Environment ◽  
Primary Water ◽  
Water Reactors

Reference fatigue crack growth rate (FCGR) curves for ferrite and austenitic stainless steels in light water reactors environments are prescribed in JSME S NA1-2004 in Japan. The reference FCGR curves in the environment in pressurized water reactors (PWR) are determined as functions of the stress intensity factor range, temperature, load rising time and stress ratio. However, similar reference FCGR curve for nickel-based alloys for PWR environment are not prescribed. In order to propose reference curve in PWR environment, fatigue tests of nickel-based alloys in a simulated PWR primary water environment were conducted. The results of the study show that FCGR in a PWR primary water environment increases with decreasing cyclic loading frequency f, increasing stress ratio R, and increasing temperature Tc.

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