Abstract
In this study, we conducted a simulation investigation of a hypothetical core barrel assembly drop accident based on the dynamic mesh model. The investigation included three separate circumstances: high-temperature dynamic water, high-temperature still water, and low-temperature still water. Results showed that dynamic parameters (e.g. velocity and resultant force) of the drop assembly in high-temperature dynamic water are smaller than those in still water due to the buffer effect of the flowing water. The velocity and hydraulic force in the dynamic water situation increased quickly at first and then remained stable throughout the remainder of the experiment; due to turbulence effect, the hydraulic force fluctuated slightly, first increased, then decreased, and then reached as stable a level in still water as that in dynamic water. The velocity of the drop assembly showed parabolic growth.
Highlights
Safety is of the absolute utmost priority during the design and commercial operation of a nuclear power plant
We conducted simulations to determine the dynamic parameters of a drop assembly based on the dynamic mesh model
Results showed that the velocity and resultant force of the drop assembly in dynamic water were far lower than those in still water, the impact on the support keys was smaller in dynamic water, suggesting that the reactor would have a higher safety margin
Summary
Safety is of the absolute utmost priority during the design and commercial operation of a nuclear power plant. The reactor, which lies at the heart of the plant, is readily affected by flow-induced vibration that can cause the component to fail due to phenomena such as nut loosening or core barrel fracture. Weld cracking in the core barrel, for example, can cause the core barrel assembly (including the core, thermal shield, flow distribution device, core barrel, and other parts) to drop— the falling assembly hits the secondary support assembly, threatening the integrity of the entire reactor. Flow-induced vibration has garnered considerable attention from researchers and developers in terms of ensuring nuclear safety. Paidoussis,[1] for example, reviewed the flow-induced vibration in reactors and reactor components and established four categories into which flow-induced vibration phenomena can be divided: cross flow, internal axial flow, external axial flow, and leakage flow. Hassan and Hayder,[3] Sigrist and Broc,[4] Kuehlert et al.,[5]
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