In the core industrial equipment of nuclear power plant steam generators, heat transfer tubes play a crucial role. They are not only critical components for heat transfer, but also have a significant impact on the safety and efficiency of the entire steam generation system. The steam generator operates in high temperature and irradiation environments, withstanding dynamic loads caused by water flow, earthquakes, etc. The Flow Induced Vibration often leads to wear and looseness of the connecting components, thereby affecting the integrity of the Pressurized Water Reactor (PWR) internal components. Therefore, the optimization and modification of heat transfer tube wear in the steam generator is a very noteworthy issue. Aiming at the optimization and renovation plan for the wear of heat transfer tubes in a certain third-generation nuclear power plant steam generator, the riveted bolt anti-loosening structural components is proposed. The method of determining equivalent vibration assessment test parameters based on the flow induced vibration test results of the internal component model is explained. High temperature rise and drop assessment tests, accelerated vibration tests, and bolt damage detection are carried out for the riveted bolt anti-loosening structural components, and conclusions that have reference significance for practical engineering are obtained.