Results 11 to 20 of about 1,326,999 (268)
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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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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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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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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Corrosion Control of Alloy 690 by Shot Peening and Electropolishing under Simulated Primary Water Condition of PWRs [PDF]
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 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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The microstructure and precipitation effects in Inconel alloy 690. [PDF]
Failure of Alloy 600 steam generator tubing in Pressurised Water Reactors (PWRs) has prompted the investigation of alloy 690 as an alternative material. Six commercially produced tubes and ten experimentally produced alloys have been examined with varying amounts of carbon, aluminium and titanium.
Smith, Alan J.
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Trustworthy Multimodal Attention Framework for Creep Rupture Life Prediction Under Data-Scarce Conditions: A Case Study on IN718. [PDF]
Adv Sci (Weinh)An attention‐based multimodal deep learning framework is developed to predict the creep life of Ni‐based superalloys by fusing processing parameters with microstructural micrographs. The model achieves high accuracy (R2 = 0.92), aligns with metallurgical principles by capturing δ‐phase influence, and incorporates uncertainty quantification, offering a ...
Lv H +10 more
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Effects of the weld thermal cycle on the microstructure of alloy 690. [PDF]
Alloy 690 has been introduced as a material for use as the heat exchanger tubes in the steam generators (SGs) of pressurised water reactor (PWR) nuclear power plant. Its immediate predecessor, alloy 600, suffered from a number of degradation modes and another alternative, alloy 800, has also had in-service problems.
Tuttle, James R.
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Magnetite Deposition Behavior on Alloy 600 and Alloy 690 Tubes in Simulated PWR Secondary Water
, 2022 Fouling due to magnetite deposition has been a major concern for steam generator (SG) tubing of pressurized water reactors (PWRs). Alloy 690 SG tubes are now used for new plants or are scheduled to replace old Alloy 600 tubes of operating plants.
Soon-Hyeok Jeon +3 more
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