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Microstructure dependence on mechanical behavior of Alloy 690 [PDF]
AIP Conference Proceedings, 2016 The influence of microstructure on the mechanical behavior of Alloy 690 was studied in this work. The microstructures with and without carbide precipitation were obtained through solution annealing and/or thermal treatment. The mechanical impact tests on different microstructure at a high strain rate of 5.3×103/s were performed by compressive split ...
X. M. Pan +3 more
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Journal of Materials Research and Technology, 2022 In this study, in response to the requirement for a higher heat transfer efficiency of the small-scale nuclear power reactors, a novel material treatment method in the form of electromagnetic coupling treatments (EMCT) was studied and applied to improve ...
Qianwen Zhang +6 more
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Nuclear Engineering and Technology, 2019 Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s.
Junjie Chen +5 more
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EFFECTS OF IRRADIATION ON THERMAL CONDUCTIVITY OF ALLOY 690 AT LOW NEUTRON FLUENCE
Nuclear Engineering and Technology, 2013 Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor
WOO SEOG RYU +4 more
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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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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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