Fatigue Design Criteria for Pressure Vessel Alloys

Jaske, Carl E.; O’Donnell, W. J.

doi:10.1115/1.3454577

Cited by 73 publications

(34 citation statements)

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“…In air, the fatigue e-N behavior of Alloy 600 is comparable to that of austenitic SSs. 2 Fatigue CGR data indicate that the enhancement of CGRs of Alloy 600 and austenitic SSs in LWR environments is also comparable. 51 However, the fatigue e-N behaviors of Alloy 600 and austenitic SSs in water differ significantly; only moderate effects of environment are observed for Alloy 600 and its weld both in low-DO * and high-DO 9 water.…”

Section: Equation 13mentioning

confidence: 97%

“…2, where the designated three curves are based on the current ASME mean curve, the best-fit curve developed by Jaske and O'Donnell, 2 and the updated statistical model that is discussed later in this report. The results indicate that the fatigue lives of Types 304 and 316 SS are comparable; those of Type 316NG are slightly higher at high strain amplitudes.…”

Section: Fatigue Lifementioning

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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Section: Equation 13mentioning

confidence: 97%

Section: Fatigue Lifementioning

confidence: 99%

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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“…．とくに，疲労寿命が数百回程度のいわゆる極低サイクル疲労では，1 回目のサイクルからき裂が発生しているという報告もある (17) ．したがって，疲労寿命のほとんどはき裂の成長に費やされていることになり，疲労寿命はき裂が限界サイズに成長するまでの繰返し数とほぼ等しくなる．村上ら (18)- (20) や小茂鳥と清水 (21) はき裂成長予測により疲労寿命が予測できると指摘している．これまでも，研究レベルにおいてはき裂成長と疲労寿命を対応づける試みが数多く報告されている (22)- (26) ．また，実用においても米国機械学会の圧力容器の設計規格では，溶接部の疲労評価に対して，き裂成長を想定した損傷駆動力（Structural Stress）を用いた設計手法が規定されている (27)- (29) ．しかし，一般的な機器設計においては，先に述べたように，評価上の損傷量である UF を用いて疲労損傷が評価されている．合理的な評価のためには，実際の損傷過程，つまりき裂の発生と成長を想定した損傷評価を行うことが望ましい．疲労設計において対象となる低サイクル領域の疲労寿命は，ひずみ範囲（または，塑性ひずみ範囲）に依存し (30) ，応力変動幅にはほとんど影響を受けない ( より，アスペクト比（深さ/表面長さ）を 0.5 とし，表面長さを深さに換算した．図 3(a) (a) Correlation with strain intensity factor range (b) Correlation with effective strain intensity factor range…”

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Damage Assessment of Low-Cycle Fatigue by Crack Growth Prediction (Development of Growth Prediction Model and Its Application)

Kamaya¹,

Kawakubo²

2012

Trans.JSME, Ser.A

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In this study, the fatigue damage was assumed to be equivalent to the crack initiation and its growth, and fatigue life was assessed by predicting the crack growth. First, a low-cycle fatigue test was conducted in air at room temperature under constant cyclic strain range of 1.2%. The crack initiation and change in crack size during the test were examined by replica investigation. It was found that a crack of 41.2 m length was initiated almost at the beginning of the test. The identified crack growth rate was shown to correlate well with the strain intensity factor, whose physical meaning was discussed in this study. The fatigue life prediction model (equation) under constant strain range was derived by integrating the crack growth equation defined using the strain intensity factor, and the predicted fatigue lives were almost identical to those obtained by low-cycle fatigue tests. The change in crack depth predicted by the equation also agreed well with the experimental results. Based on the crack growth prediction model, it was shown that the crack size would be less than 0.1 mm even when the estimated fatigue damage exceeded the critical value of the design fatigue curve, in which a twenty-fold safety margin was used for the assessment. It was revealed that the effect of component size and surface roughness, which have been investigated empirically by fatigue tests, could be reasonably explained by considering the crack initiation and growth. Furthermore, the environmental effect on the fatigue life was shown to be brought about by the acceleration of crack growth.

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“…where B and Se are constants determined using linear, least-squares regression analyses to the data (Jaske and O'Donnell, 1977).…”

Section: Present Asme Codementioning

confidence: 99%

Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

Ware¹,

Morton²,

Nitzel

1995

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Fatigue Design Criteria for Pressure Vessel Alloys

Cited by 73 publications

References 0 publications

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

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

Damage Assessment of Low-Cycle Fatigue by Crack Growth Prediction (Development of Growth Prediction Model and Its Application)

Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

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