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contributor authorYu, Mingda
contributor authorDu, Juan
contributor authorShao, Xuejiao
contributor authorPu, Zhuo
date accessioned2025-04-21T10:15:30Z
date available2025-04-21T10:15:30Z
date copyright10/23/2024 12:00:00 AM
date issued2024
identifier issn0094-9930
identifier otherpvt_146_06_061505.pdf
identifier urihttp://yetl.yabesh.ir/yetl1/handle/yetl/4305815
description abstractThe structural safety of the cylinder in liquid-cooled fast reactors is significantly affected by high-temperature thermal striping on the cylinder wall due to the free level fluctuation. Based on the elevated-temperature creep-fatigue evaluation theory of the RCC-MRx rules, combined with the finite element and numerical heat transfer analysis methods, a thermal analysis method appropriate for the fast reactor cylinder containing hypothetical cracks after being subjected to normal transient loads was established. The proposed model is validated compared with the test reactors, FAENA and SUPERSOMITE, and the calculated thermal striping limits agree with the experimental data. Furthermore, the involved model was implemented to predict the thermal striping limits for several ideal normal transients. Simultaneously, the effects of some key parameters, including the frequency, the heat transfer coefficient, wall thickness, and the mean temperature of the fluid on the thermal striping limit of the SS316 stainless steel cylinder were studied.
publisherThe American Society of Mechanical Engineers (ASME)
titleThermal Striping Limits Analysis on the Cylinder With a Defect of Liquid-Cooled Nuclear Fast Reactors
typeJournal Paper
journal volume146
journal issue6
journal titleJournal of Pressure Vessel Technology
identifier doi10.1115/1.4066677
journal fristpage61505-1
journal lastpage61505-8
page8
treeJournal of Pressure Vessel Technology:;2024:;volume( 146 ):;issue: 006
contenttypeFulltext


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