Flow-induced vibrations in nuclear steam generators

Hassan, Marwan

doi:10.1016/b978-0-08-100894-2.00015-7

Cited by 7 publications

(1 citation statement)

References 40 publications

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“…Reported failures of NSG tubes due to mechanically-aided damage (excluding fatigue) in order of relevance are as follows: fretting wear at AVBs, impact wear damage by loose parts and fretting wear in the economiser region. Fretted tubes are more susceptible to fatigue or SCC crack initiation (Hassan 2017;IAEA 2011a) Fatigue is a progressive and localised type of damage that occurs on materials subjected to cycling loading. If the applied stress amplitude is lower than the material yield strength but higher than its fatigue limit high-cycle fatigue takes place.…”

Section: Mechanically-aided Degradation Processesmentioning

confidence: 99%

Corrosion control of nuclear steam generators under normal operation and plant-outage conditions: a review

Rodríguez

2020

Corrosion Reviews

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AbstractNuclear steam generators (NSGs) are key components of nuclear power plants (NPPs) since their reliability affects the overall plant performance. Corrosion control of steam generators under operation and during plant outages is attained by proper design, adequate material selection and control of water chemistry in the entire secondary cooling circuit. This article reviews the types of steam generators, the materials used for their construction (with a special focus on tubing alloys), the corrosion processes that may occur under operation and the strategies to preserve steam generators and other components of the secondary cooling system from corrosion during plant outages.

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Section: Mechanically-aided Degradation Processesmentioning

confidence: 99%

Corrosion control of nuclear steam generators under normal operation and plant-outage conditions: a review

Rodríguez

2020

Corrosion Reviews

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Impact-sliding fretting tribocorrosion behavior of 316L stainless steel in solution with different halide concentrations

Tan

Bonzom

et al. 2023

Friction

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Impact–sliding caused by random vibrations between tubes and supports can affect the operation of heat exchangers. In addition, a corrosive environment can cause damage, accelerating the synergism of corrosion and wear. Therefore, the focus of this work was the impact–sliding fretting tribocorrosion behavior of 316L heat exchanger tubes at different halide concentrations. A device system incorporating the in situ electrochemical measurements of impact–sliding fretting corrosion wear was constructed, and experiments on 316L heat exchanger tubes in sodium chloride (NaCl) solution with different concentrations (0.0, 0.1, 0.5, 1.0, 3.5, and 5.0 wt%) were carried out. The synergism between wear and corrosion was also calculated and analyzed. The wear and damage mechanisms were elucidated by correlating the corrosion–wear synergism, morphologies, and material loss rates. The results indicated that the stable wear stage occurred at approximately 9–12 h, after which the corrosion current increased with the expansion of the wear area. As the halide concentration increased, the scale of damage on the wear scars gradually decreased, changing from being dominated by cracks, delaminations, and grooves to being dominated by scratches, microgrooves, and holes. There was an obvious positive synergism between wear and corrosion. The material loss was dominated by pure mechanical wear and wear enhanced by corrosion, but corrosion enhanced by wear contributed more than tangential sliding fretting corrosion. The total mass loss increased gradually in the range of 0.0–0.5 wt% and decreased in the range of 0.5–5.0 wt%. Large-scale damage enhanced by corrosivity and small-scale damage reduced by lubricity dominated the material loss at low and high concentrations, respectively.

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

Guo

Lai

Li³

et al. 2020

Annals of Nuclear Energy

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Flow-induced vibrations in nuclear steam generators

Cited by 7 publications

References 40 publications

Corrosion control of nuclear steam generators under normal operation and plant-outage conditions: a review

Corrosion control of nuclear steam generators under normal operation and plant-outage conditions: a review

Impact-sliding fretting tribocorrosion behavior of 316L stainless steel in solution with different halide concentrations

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

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