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

Meigui Yin,Liwen Wang,Zhenbing Cai,Daniel Nélias,Chaise Thibaut,Minhao Zhu

doi:10.1007/s40544-021-0538-9

Abstract

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.

Highlights

In 1957, the world’s first commercial nuclear power plant (Shipping Port, USA) began generating electricity, indicating that humans had entered the era of peaceful nuclear energy usage [1]
The steam generator tube is an important component of nuclear power plants that operate under harsh operating conditions
Based on Newton’s second law and the Hertz elastic theory, at each instant during the impact contact process, the rate of change of the impact velocity depends on the interaction force and compressive displacement, which are expressed in Eqs. (1) and (2)

Summary

Introduction

In 1957, the world’s first commercial nuclear power plant (Shipping Port, USA) began generating electricity, indicating that humans had entered the era of peaceful nuclear energy usage [1]. Second- and third-generation nuclear power reactors are operated normally worldwide. In the early 21st century, to improve the thermal efficiency and operational safety of existing nuclear power reactors, researchers began to study and develop fourth-generation nuclear reactors, whose commercial operations are expected to begin after the 2030s [2, 3]. The steam generator tube is an important component of nuclear power plants that operate under harsh operating conditions. Inconel 600, 690, and 800 steam generator tubes are installed in most active nuclear power plants [4, 5]. 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]. 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]

Methods

Results

Conclusion