Progress in studying the fretting wear/corrosion of nuclear steam generator tubes

Guo, Xianglong; Lai, Ping; Li, Ling; Tang, Lichen; Zhang, Lefu

doi:10.1016/j.anucene.2020.107556

Cited by 42 publications

(7 citation statements)

References 91 publications

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“…Thus, the reactor vessel materials of the PWR design are the components most affected by neutron radiation, thereby reducing its mechanical properties [59]. On the other hand, the SG is the component of PWR design suffering the most corrosion damage [28,60], being both stress corrosion cracking (SCC) and particle deposition the main degradation mechanisms. The degradation of steam generators tubes has involved costly modifications, inspections and even the replacement of SG in the most severe cases [61,62].…”

Section: Simulation Of the Life Extension Of Operational Pwrmentioning

confidence: 99%

A Global Review of PWR Nuclear Power Plants

2020

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Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%.

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Section: Simulation Of the Life Extension Of Operational Pwrmentioning

confidence: 99%

A Global Review of PWR Nuclear Power Plants

2020

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“…The 2.25Cr1Mo steel tube has been widely used in fourth-generation sodium-cooled experimental fast reactors, primarily owing to its high resistance to hydrogen attacks and excellent high-temperature strength [6]. However, owing to the combined effects of high flow rate, high fluid pressure and temperature, and narrow clearance between tubes and anti-vibration bars, various wear behaviors were observed on their surfaces, which significantly degraded their operating performance and service life [7][8][9].…”

Section: Introductionmentioning

confidence: 99%

Impact-sliding wear response of 2.25Cr1Mo steel tubes: Experimental and semi-analytical method

et al. 2021

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The impact-sliding wear behavior of steam generator tubes in nuclear power plants is complex owing to the dynamic nature of the mechanical response and self-induced tribological changes. In this study, the effects of impact and sliding velocity on the impact-sliding wear behavior of a 2.25Cr1Mo steel tube are investigated experimentally and numerically. In the experimental study, a wear test rig that can measure changes in the impact and friction forces as well as the compressive displacement over different wear cycles, both in real time, is designed. A semi-analytical model based on the Archard wear law and Hertz contact theory is used to predict wear. The results indicate that the impact dynamic effect by the impact velocity is more significant than that of the sliding velocity, and that both velocities affect the friction force and wear degree. The experimental results for the wear depth evolution agree well with the corresponding simulation predictions.

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“…The problem of fretting wear for the fuel rods in nuclear light water reactor, which occurs in their contact with the spacer grids, has a great importance for the safety of nuclear power plants (Blau 2014;Guo et al 2020;Zhang et al 2018). It is essential that this type of fretting wear, which is caused by the interface slippage due to flow induced vibrations of the fuel rods, cannot be get rid of extensively, because of the design requirements for the contact force between the fuel rods and the grid supports (Kim et al 2008).…”

Section: Introductionmentioning

confidence: 99%

Optimal design of the functional grading in elastic wear-resisting bearings: a simple analytical model

Argatov

Chai

2021

Int J Mech Mater Des

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A one-dimensional wear contact problem for a functionally-graded elastic coating attached to a rigid base is considered under the condition of a constant pre-strain with the effect of the coating thickness evolution taken into account. It is assumed that the wear coefficient is inversely proportional to the material stiffness at the bearing surface. The presented analysis is primarily given in the dimensionless form in order to illustrate qualitative features of the model behavior. It is shown that the wearing-out period essentially depends on the wear coefficient variation with depth. It is found that the optimal design of FGM bearings is constrained by the maximum of the grading gradient of material properties.

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Progress in studying the fretting wear/corrosion of nuclear steam generator tubes

Cited by 42 publications

References 91 publications

A Global Review of PWR Nuclear Power Plants

A Global Review of PWR Nuclear Power Plants

Impact-sliding wear response of 2.25Cr1Mo steel tubes: Experimental and semi-analytical method

Optimal design of the functional grading in elastic wear-resisting bearings: a simple analytical model

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