Influence of Mean Strain on Fatigue Life of Stainless Steel in Pressurized Water Reactor Water Environment

被引:1
|
作者
Kamaya, Masayuki [1 ]
机构
[1] Inst Nucl Safety Syst Inc, Nucl Power Plant Aging Res Ctr, 64 Sata, Mihama, Fukui 9191205, Japan
关键词
CRACK-GROWTH; STRENGTH;
D O I
10.1115/1.4050089
中图分类号
TH [机械、仪表工业];
学科分类号
0802 ;
摘要
The Influence of application of the mean strain on the fatigue life was investigated for Type 316 stainless steel in the simulated pressurized water reactor (PWR) primary water environment. Low-cycle fatigue tests were conducted for a constant mean strain by controlling the strain range to be 1.2%. The applied strain rates were 0.4%/s, 0.004%/s, or 0.001%/s. The applied mean strain was 15% in nominal strain. In addition, cold worked specimens were also used for the tests without applying the mean strain. The cold working simulated the application of mean strain without an increase in surface roughness due to the application of plastic deformation. By inducing the cold working at low temperature, the influence of martensitic phase on the fatigue life was also examined. The PWR water environment reduced the fatigue life and the degree of the fatigue life reduction was consistent with the prediction model of the code issued by the Japan Society of Mechanical Engineers (JSME) and NUREG/CR-6909 Rev. 1. Increases in the maximum peak stress and stress range caused by cold working did not cause any apparent change in the fatigue life. It was revealed that the 10.5 wt% martensitic phase and the increase in the surface roughness caused by the application of 15% mean strain did not bring about further fatigue life reduction. The current JSME and NUREG/CR-6909 models were applicable for predicting the fatigue in the PWR water environment even when the mean strain or cold working was applied.
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页数:8
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