scholarly journals Fatigue Damage Management According to Performance Based Maintenance (PBM) Concept

Author(s):  
Masayuki Kamaya

Abstract A maintenance concept of performance based maintenance (PBM) has been proposed by the current author. According to the PBM concept, inspection results are considered in determining the next inspection schedule. In this study, this concept was applied to fatigue degradation for stainless steel components in the pressurized water reactor (PWR) primary water environment. It is possible to estimate the fatigue life for the PWR water environment from that obtained in an air environment and the parameter Fen, which represents the ratio of the fatigue life in the air and PWR water environments. It was shown that the fatigue life prediction using Fen can be replaced by the crack growth analysis using the growth rate for the PWR water environment. Then, the crack growth was predicted for a thermal loading assuming the growth occurred in the PWR water environment. It was shown that the duration until the next inspection could be optimized based on the inspection results together with the crack growth curve. A long term operation before the inspection resulted in a longer duration until the next inspection.

Author(s):  
N. Platts ◽  
P. Gill ◽  
S. Cruchley ◽  
E. Grieveson ◽  
M. Twite

The Pressurized Water Reactor (PWR) primary coolant environment is known both to significantly reduce the fatigue life of austenitic stainless steels and to lead to enhanced fatigue crack propagation rates. Relationships for the impact of the PWR coolant environment on fatigue life have been presented in NUREG/CR-6909 using an environmental fatigue correction factor (Fen), which is a function of temperature. Fatigue crack growth behavior has been codified in ASME Code Case N-809 in terms of parameters such as rise time, stress intensity factor, load ratio and temperature. However, plant performance suggests that the application of these predicted environmental effects using current assessment procedures for fatigue for plant transient loading may be unduly pessimistic. One potential reason for this over-conservatism is thought to be that, although the majority of plant design transients result from variations in thermal loading, most available data are derived from isothermal testing. For the calculation of fatigue initiation life, NUREG/CR-6909 gives guidance on the effective temperature to be used in assessments of thermal transients. Recent results from Thermo-Mechanical Fatigue (TMF) testing on stainless steels in PWR coolant show that this guidance is conservative for out-of-phase cycling of temperature and loading, and potentially non-conservative for in-phase thermal loading. In contrast, code case N-809 gives no guidance on the effective temperature for fatigue crack growth assessments, resulting in maximum temperatures frequently being adopted for assessments of thermal transients. There is therefore a need for a clearer understanding of the impact of variable temperatures during transients on the predicted levels of environmental fatigue. This paper describes test facilities developed to permit measurements of both thermo-mechanical fatigue life and fatigue crack growth rates in pressurized water reactor environments. Initial test results obtained using these facilities are presented. The fatigue life data have been generated for a range of applied strain amplitudes, 0.45% to 1%, using temperature cycling between 100°C and 300°C. These data, for both in- and out-of-phase temperature and loading, are compared to the predictions of the “weighted Fen” model which is detailed in a separate paper, PVP2017-66030. Similarly, crack growth rate data generated for cycles between 140°C and 280°C are presented and comparisons made against the predictions of the “weighted K rate” (WKR) method detailed in paper PVP2017-65645. In both cases, the test results suggest that the weighted models are able to provide good predictions of an effective temperature to be used in fatigue assessment methods, which offer a significant improvement in the treatment of variable temperatures compared to current assessment practice.


Author(s):  
Kyoung Joon Choi ◽  
Seung Chang Yoo ◽  
Taeho Kim ◽  
Seong Sik Hwang ◽  
Min Jae Choi ◽  
...  

With the extension of pressurized water reactor’s design life or continued operation, more careful study on the integrity of the internal structures needs to be pursued. In this study, warm-rolling and heat-treatment were applied to 316L stainless steel, in order to simulate the effect of radiation damage such as hardening and radiation-induced grain boundary segregation. And, the crack growth rate testing under constant load condition was performed in the primary water conditions of a pressurized water reactor. Also, in order to investigate the effect of dissolved hydrogen on the crack growth, the dissolved hydrogen concentration was varied between 30 to 50 cc/kg in simulated primary water condition of a pressurized water reactor. The warm-rolled specimens showed the higher crack growth rate than as-received one. Also, the crack growth rate increased as the dissolved hydrogen concentration increases.


2019 ◽  
Vol 141 (5) ◽  
Author(s):  
Masayuki Kamaya

The mean stress effect on the fatigue life of type 316 stainless steel was investigated in simulated pressurized water reactor (PWR) primary water and air at 325 °C. The tests in air environment have revealed that the fatigue life was increased with application of the positive mean stress for the same stress amplitude because the strain range was decreased by hardening of material caused by increased maximum peak stress. On the other hand, it has been shown that the fatigue life obtained in simulated PWR primary water was decreased compared with that obtained in air environment even without the mean stress. In this study, type 316 stainless steel specimens were subjected to the fatigue test with and without application of the positive mean stress in high-temperature air and PWR water environments. First, the mean stress effect was discussed for high-temperature air environment. Then, the change in fatigue life in the PWR water environment was evaluated. It was revealed that the change in the fatigue life due to application of the mean stress in the PWR water environment could be explained in the same way as for the air environment. No additional factor was induced by applying the mean stress in the PWR water environment.


CORROSION ◽  
10.5006/2572 ◽  
2017 ◽  
Vol 74 (1) ◽  
pp. 24-36 ◽  
Author(s):  
Koji Arioka ◽  
Roger W. Staehle ◽  
Robert L. Tapping ◽  
Takuyo Yamada ◽  
Tomoki Miyamoto

The primary purpose of this research is to examine the stress corrosion cracking (SCC) resistance of Alloy 800NG in pressurized water reactor (PWR) primary water and pressurized heavy water reactor (PHWR) primary water. Rates of SCC growth of 20% cold-worked (CW) Alloy 800NG measured over the temperature range between 270°C and 360°C were compared with previously reported results for 20% CW Alloy TT690 and 20% CW Alloy 600 in order to consider which material is the most SCC resistant among materials presently being used for steam generator (SG) tubing worldwide. The secondary purpose is to examine the effect of chromium addition on SCC growth in PWR primary water of a series of alloys based on the Alloy 800 composition. SCC growth measurements were performed in PWR primary water over the chromium concentration range from 16% to 27% to obtain fundamental knowledge useful for considering a future alternative SCC-resistant material for SG tubing in extended life PWRs and PHWRs. The third objective is to examine the rate of cavity formation of 20% CW Alloy 800NG to obtain basic knowledge of one possible mechanism for SCC initiation after long-term operation. Measured rates of cavity formation in 20% CW Alloy 800NG were compared with previously reported results of 20% CW Alloy TT690 to compare the rate of SCC initiation caused by cavity formation. Four important patterns were observed. First, excellent SCC growth resistance was observed for 20% CW Alloy 800NG compared to 20% CW Alloy TT690 at 320°C, 340°C, and 360°C. Second, an inverse temperature dependence on SCC growth was observed in Alloy 800NG. The rate of SCC growth increased with decreasing temperature which was completely different from the trend for Alloy 600. Third, a significant beneficial effect by chromium addition in 800 series alloys on SCC growth resistance was observed in PWR primary water in the operating temperature range of PWRs and PHWRs. The rate of SCC growth decreased with increasing chromium concentration in the chromium concentration range between 16% and 27% chromium at 270°C, 290°C, and 320°C. However, no beneficial effect of chromium addition in these alloys was observed at 340°C and 360°C. Finally, a more than 10 times slower rate of cavity formation was observed in 20% CW Alloy 800NG than for 20% CW Alloy TT690. Results suggested that because of cavity formation, a more than 10-fold faster crack initiation occurred in Alloy TT690 than in Alloy 800NG. Further, carbide coverage and grain size significantly affected the rate of cavity formation. Detailed and comprehensive studies of long-term SCC initiation are necessary to ensure the future reliability of life-extended PWRs and PHWRs.


Author(s):  
Ernest D. Eason ◽  
Edward E. Nelson ◽  
Graham B. Heys

Models of fatigue crack growth rates for medium and low sulfur ferritic pressure vessel steels in pressurized water reactor (PWR) primary environments are developed based on a recent collection of UK data and the EPRI Database for Environmentally Assisted Cracking (EDEAC). The combined UK and EDEAC database contains a broader range of experimental conditions specific to PWRs than either database by itself. Both probabilistic and conventional crack growth rate models are developed that reduce unnecessary conservatism for medium and low sulfur PWR primary water applications and eliminate the explicit dependence on rise time that caused difficulties applying prior models.


2018 ◽  
Vol 85 (13) ◽  
pp. 625-634
Author(s):  
Yibo Jia ◽  
Sichun Ling ◽  
Kun Zhang ◽  
Jiarong Ma ◽  
Tongming Cui ◽  
...  

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