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Flow-induced vibration calculation method for nuclear reactor steam generator

A technology of steam generators and nuclear reactors, which is applied in the fields of instruments, design optimization/simulation, and special data processing applications, etc. It can solve problems such as error-prone, limited development, and slow data exchange speed on the coupling surface

Active Publication Date: 2020-10-30
XI AN JIAOTONG UNIV
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Problems solved by technology

This method is accompanied by the disadvantages of large amount of calculation, slow data exchange on the coupling surface and error-prone, and has certain applicability requirements and version requirements for the calculation program, especially when the geometric structure of the calculation object is complex, the two-way fluid-solid coupling work The development will be limited by computing resources

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  • Flow-induced vibration calculation method for nuclear reactor steam generator
  • Flow-induced vibration calculation method for nuclear reactor steam generator
  • Flow-induced vibration calculation method for nuclear reactor steam generator

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Embodiment Construction

[0037]Below in conjunction with accompanying drawing and specific embodiment the present invention is described in further detail:

[0038] The present invention provides a figure 1 The calculation method for the flow-induced vibration of the nuclear reactor steam generator shown is as follows:

[0039] Step 1: Obtain the geometric parameters of the U-shaped heat transfer tube of the nuclear reactor steam generator, and considering the symmetry and repeatability of the U-shaped tube arrangement, establish a single U-shaped tube solid domain, primary side fluid domain inside the tube, and secondary side fluid outside the tube A simplified geometry model assembly of the domain, such as figure 2 shown. In the finite element analysis program, the physical properties of the heat transfer tube Inconel690 are given, and the simple support and fixed support methods are used to impose fixed support constraints on the roots of both ends of the U-shaped heat transfer tube. , 120°, an...

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Abstract

The invention discloses a flow-induced vibration calculation method for a nuclear reactor steam generator, and the method comprises the following steps: 1, building a steam generator heat transfer tube wet modal analysis model, carrying out the wet modal analysis, and extracting a mass matrix [M], a rigidity matrix [K] and a damping matrix [C] of a heat transfer tube; 2, writing a Newmark-beta method for solving a transient kinetic equation and a dynamic grid model into a user-defined function file; 3, establishing fluid dynamic calculation models of primary and secondary side fluid domains inside and outside the heat transfer tube, and carrying out first time step iterative calculation; 4, calling and executing the user-defined function file in the step 2, updating the fluid domain grid,and performing iterative computation of the next time step on the fluid domain after the grid is updated; and 5, circularly executing the step 4 until the set calculation termination time is calculated, and stopping the calculation. The flow-induced vibration characteristics of the nuclear reactor steam generator are obtained through calculation, and the method has important significance in designand safety analysis of the nuclear reactor steam generator.

Description

technical field [0001] The invention belongs to the technical field of flow-induced vibration of a nuclear reactor steam generator, and in particular relates to a calculation method for flow-induced vibration of a nuclear reactor steam generator. Background technique [0002] The steam generator is the core component connecting the primary and secondary circuits of the pressurized water reactor nuclear power plant, and its effectiveness is related to the safe and stable operation of the nuclear reactor. The steam generators currently used in PWR nuclear power plants mainly include vertical U-shaped heat transfer tube natural circulation steam generator (UTSG) and tubular once-through steam generator (OTSG). For the vertical U-shaped heat transfer tube natural circulation steam generator, the high-pressure thermal fluid of the primary circuit flows into the U-shaped heat transfer tube from the hot leg through the inlet chamber, and transfers heat to the secondary circuit in t...

Claims

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): G06F30/23G06F119/14
CPCG06F30/23G06F2119/14
Inventor 王明军王莹杰张大林田文喜苏光辉秋穗正
Owner XI AN JIAOTONG UNIV
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