Effects of LWR Coolant Environments on Fatigue Lives of Austenitic Stainless Steels

Chopra, O.K.; Gavenda, D. J.

doi:10.1115/1.2842228

Cited by 35 publications

(106 citation statements)

References 9 publications

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“…3,34 The cyclic strain hardening of Type 316NG tested in air at room temperature and 288°C is shown in Fig. 5; that of Type 304 SS in air at 288°C is shown in Fig.…”

Section: Cyclic Hardening Behaviormentioning

confidence: 99%

“…For these tests, the strain Based on cyclic stress-vs.-strain correlations for Type 316 SS, 3 actual strain amplitudes for these tests should be 0.23-0.32%. These data were excluded from the analysis in NUREG/CR-5704 3 for updating the statistical model for estimating the fatigue life of austenitic SSs in air. Figure 2 also indicates that the ASME mean curve is not consistent with the existing fatigue e-N data for austenitic SSs.…”

Section: Fatigue Lifementioning

confidence: 99%

“…18 In water, the existing fatigue e-N data include the tests performed by General Electric Co. (GE) in a test loop at the Dresden 1 reactor, 19 the JNUFAD database, studies at Mitsubishi Heavy Industries, Ltd. (MHI), [20][21][22][23][24][25] Ishikawajima-Harima Heavy Industries Co. (IHI), 26,27 and Hitachi 28,29 in Japan, and the present work at ANL. 3,6,[30][31][32][33] In these studies, various criteria are used to define fatigue life of a test specimen. In the present study, fatigue life N for strain-controlled tests is defined as the number of cycles for tensile stress to decrease by 25% from its peak or steady-state value.…”

Section: Overview Of Fatigue E E E E-n Datamentioning

confidence: 99%

“…Existing fatigue e-N data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of carbon and low-alloy steels, 5,6 as well as of austenitic SSs 3,6 (Fig. 1).…”

Section: Introductionmentioning

confidence: 99%

“…However, the best-fit fatigue curve used to develop the current Code fatigue design curve for austenitic stainless steels (SSs) does not accurately represent the available experimental data. 2,3 The current Code design curve for SSs includes a reduction of only ª1.5 and 15 from the mean curve for the SS data, not the 2 and 20 originally intended. Also, because, for the current Code mean curve, the fatigue strength at 10 6 cycles is greater than the monotonic yield strength of austenitic SSs, the current Code design curve for austenitic SSs does not include a mean stress correction.…”

Section: Introductionmentioning

confidence: 99%

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Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Chopra¹

2002

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The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life (e-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue e-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue e-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed. Executive SummarySection III, Subsection NB, of the ASME Boiler and Pressure Vessel Code contains rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III specify the Code design fatigue curves for applicable structural materials. However, Section III, Subsection NB-3121, of the Code states that effects of the coolant environment on fatigue resistance of a material were not intended to be addressed in these design curves. Therefore, the effects of environment on fatigue resistance of materials used in operating pressurized water reactor (PWR) and boiling water reactor (BWR) plants, whose primary-coolant pressure boundary components were designed in accordance with the Code, are uncertain.The current Section-III design fatigue curves of the ASME Code were based primarily on strain-controlled fatigue tests of small polished specimens at room temperature in air. Best-fit curves to the experimental test data were first adjusted to account for the effects of mean stress and then lowered by a factor of 2 on stress and 20 on cycles (whichever was more conservative) to obtain the design fatigue curves. These factors are not safety margins but rather adjustment factors that must be applied to experimental data to obtain estimates of the lives of components. They were not intended to address the effects of the coolant environment on fatigue life. Recent fatigue-strain-vs.-life (e-N) data obtained in the U.S. and Japan demonstrate that light water reactor (LWR) environments can have potentially significant effects on the fatigue resistance of materials. Specimen lives obtained from tests in simulated LWR environments can be much shorte...

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“…3,34 The cyclic strain hardening of Type 316NG tested in air at room temperature and 288°C is shown in Fig. 5; that of Type 304 SS in air at 288°C is shown in Fig.…”

Section: Cyclic Hardening Behaviormentioning

confidence: 99%

Section: Fatigue Lifementioning

confidence: 99%