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

Author:

Kamaya Masayuki1

Affiliation:

1. Nuclear Power Plant Aging Research Center, Institute of Nuclear Safety System, Inc, 64 Sata, Mihama-cho, Fukui 919-1205, Japan

Abstract

Abstract 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.

Publisher

ASME International

Subject

Mechanical Engineering,Mechanics of Materials,Safety, Risk, Reliability and Quality

Reference13 articles.

1. Cold Work and Temperature Dependence of Stress Corrosion Crack Growth of Austenitic Stainless Steels in Hydrogenated and Oxygenated High-Temperature Water;Corrosion,2007

2. Comparison of Environmental Fatigue Evaluation Methods in LWR Water,2008

3. Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,2017

4. Influence of PWR Primary Water on LCF Behavior of Type 304 L Austenitic Stainless Steel at 300 Degrees C: Comparison With Results Obtained in Vacuum or in Air,2012

5. Codes for Nuclear Power Generation Facilities: Environmental Fatigue Evaluation Method for Nuclear Power Plants;Japan Society Mechanical Engineers,2009

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