Results 21 to 30 of about 1,567 (242)
Atomic simulations of nanoscale friction behavior in polycrystalline alloy 690
Materials Research Express, 2022 Fretting wear is one of the most important failure forms of alloy 690 heat exchanger tubes in nuclear power plants. The key to understanding the fretting wear of alloys lies in the friction process, especially at the atomic scale.
Ai-Long Zhou +4 more
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Nuclear Engineering and Technology, 2021 The dislocation density in strain-hardened Alloy 690 was analyzed using scanning transmission electron microscopy (STEM) to study the relationship between the local plastic strain and susceptibility to primary water stress corrosion cracking (PWSCC) in ...
Sung-Woo Kim +2 more
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SUSCEPTIBILITY OF ALLOY 690 TO STRESS CORROSION CRACKING IN CAUSTIC AQUEOUS SOLUTIONS
Nuclear Engineering and Technology, 2013 Stress corrosion cracking (SCC) behaviors of Alloy 690 were studied in lead-containing aqueous alkaline solutions using the slow strain rate tension (SSRT) tests in 0.1M and 2.5M NaOH with and without PbO at 315°C.
DONG-JIN KIM +2 more
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Archives of Metallurgy and Materials, 2017 This work aims to study the stress corrosion crack growth behavior of cold-rolled Alloy 690 in the primary water of a pressurized water reactor. Compared with Alloy 600, which shows typical intergranular cracking along high angle grain boundaries, the ...
Kim S.-W. +3 more
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Nuclear Engineering and Technology, 2019 The crack growth responses of as-received and as-welded Alloy 600/182 and Alloy 690/152 welds to constant loading were measured by a direct current potential drop method using compact tension specimens in primary water at 325 °C simulating the normal ...
Yun Soo Lim +3 more
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Fatigue Life Prediction of Steam Generator Tubes by Tube Specimens with Circular Holes
Metals, 2019 Heat exchangers manufactured from Inconel 690 tubes are widely used for steam generators in nuclear power plants. Inconel 690 tubes have suffered failures of fatigue fracture due to flow induced vibration.
Qiwei Wang +4 more
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Advances in Materials Science and Engineering, 2015 This work clarifies the effect of surface modifications on the corrosion rate of Alloy 690, a nickel-based alloy for steam generator tubes, under the simulated test conditions of the primary water chemistry in nuclear power plants. The surface stress was
Kyung Mo Kim +3 more
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Nuclear Engineering and Technology, 2019 This paper provides further validation of the burst pressure estimation equations for multiple axial surface cracked steam generator tubes, recently proposed by the authors based on analytical local collapse load approach against systematic FE damage ...
Ji-Seok Kim +3 more
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PWSCC growth rate model of alloy 690 for head penetration nozzles of Korean PWRs
Nuclear Engineering and Technology, 2019 This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors.
Sung-Woo Kim +3 more
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Impact of MgO Additions on the Oxidation Resistance and Compression Behavior of Vanadium Alloys
Advanced Engineering Materials, EarlyView.An increase in oxidation resistance at 600 °C is achieved for alloys containing more than 2.5 wt.% MgO. It has been established that, along with V2O5, MgO6V2, and Mg2O7V2 are formed, which are presumed to enhance oxidative stability. Furthermore, the possible transformation of Mg‐vanadates through reaction with CO2, which presumably occurs in this ...
Ievgen Solodkyi +5 more
wiley +1 more source