Coupling analysis and calculation method of flow-induced vibration and fretting wear of u-shaped heat transfer tubes in nuclear power plants
A technology of fretting wear and coupling analysis, applied in nuclear power generation, nuclear engineering, nuclear reactor monitoring, etc., can solve problems such as stray bullet instability and disregard
Active Publication Date: 2020-10-23
SHANGHAI NUCLEAR ENG RES & DESIGN INST CO LTD
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However, these analyzes of stray bullet instability are limited to the calculation and analysis of the initial structure, and do not consider whether stray bullet instability will occur in the structure after several years of operation.
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[0018] In order to make the above-mentioned objects, features and advantages of the present invention more obvious and understandable, the following takes a U-shaped heat transfer tube of a nuclear power plant steam generator as an example and further describes the present invention in detail with reference to the drawings and specific embodiments.
[0019] In the steam generator of a nuclear power plant, there are generally thousands to tens of thousands of heat transfer tubes. For any of the heat transfer tubes, set up as figure 1 The finite element model shown. Among them, the connection between the heat transfer tube and the tube sheet is simplified as a fixed-support boundary condition; the contact with the support plate is simplified as a simply-supported boundary condition; the contact with the anti-vibration strip is simplified as the boundary condition of a simply supported in the Z direction and a grounding spring in the X-Y plane. The stiffness setting of the spring bo...
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The invention provides a flow-induced vibration and fretting wear coupling analysis calculating method for a U-shaped heat transfer tube of nuclear power station steam generators. The method is characterized by comprising heat transfer tube-antivibration strip fretting wear test, flow-induced vibration analysis and fretting wear calculation analysis. By means of the flow-induced vibration and fretting wear coupling analysis calculating method for a U-shaped heat transfer tube of nuclear power station steam generators of the invention, the calculation and analysis of flow-induced vibration and fretting wear of heat transfer tubes are coupled, determining and considering whether fluid elasticity instability will happen if tube walls become thin and antivibration strip support turns weak. At the same time, the wearing condition of heat transfer tubes for a long time can be predicted under the changing flow-induced vibration condition.
Description
Technical field [0001] The invention relates to the field of nuclear power plant steam generator design analysis, in particular to a method for coupling analysis and calculation of flow-induced vibration and fretting wear of a U-shaped heat transfer tube of a nuclear power plant steam generator. Background technique [0002] Flow-induced vibration and fretting abrasion are the main factors for the failure of heat transfer tubes of steam generators in nuclear power plants. In the design stage, it is very important to accurately predict the flow-induced vibration and fretting wear of the steam generator heat transfer tube during the life of the steam generator, and it is an important guarantee for the normal operation of the steam generator during the life. For the steam generator of the third-generation nuclear power plant, the design life of the heat transfer tube of the steam generator should be the same as the equipment itself, that is, 60 years. [0003] In traditional calculat...
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IPC IPC(8): G06F30/20G21C17/00
CPCG21C17/00G06F30/00Y02E30/30
Inventor 唐力晨钱浩谢永诚景益
Owner SHANGHAI NUCLEAR ENG RES & DESIGN INST CO LTD