Method for evaluating flow impact at reactor core outlet of sodium-cooled fast reactor

Abstract

The invention discloses a method for evaluating flow impact of a reactor core outlet of a sodium-cooled fast reactor, which comprises the following steps of: carrying out geometric modeling by using Solidworks according to an actual structure of the reactor core outlet of the sodium-cooled fast reactor, and carrying out fluid domain and solid domain subdivision; performing grid division on a fluiddomain in a grid division function of computational fluid mechanics software, setting physical properties and boundary conditions of liquid metal sodium in a fluid computational domain, and performing transient flow field calculation by adopting a high-precision large eddy simulation method; performing condition setting and mesh generation on the solid domain in finite element analysis software;using a multi-physical field coupling platform MpCCI for transmitting flow field parameters obtained through computational fluid mechanics software to finite element analysis software, and making solid structure mechanics calculation in the finite element analysis software; and finally, evaluating the flow impact influence of the reactor core outlet of the sodium-cooled fast reactor through mechanical analysis of a flow field and a solid structure.

Description

technical field [0001] The invention belongs to the technical field of sodium-cooled fast reactor core coolant, and in particular relates to a method for evaluating the flow shock at the outlet of the sodium-cooled fast reactor core. Background technique [0002] With the rapid growth of my country's economy, the demand for energy is increasing, and nuclear energy has been vigorously developed because of its clean and efficient characteristics. Due to the continuous development of nuclear power technology and the continuous accumulation of reactor operating experience, the international nuclear engineering community has put forward the idea of ​​developing the fourth generation nuclear power system. The standards of the fourth-generation reactors mainly include: ① Economically competitive; ② Inherent safety; ③ Minimize the generation of nuclear waste; ④ Prevent nuclear proliferation; ⑤ Good social benefits. In 2002, the Fourth Generation Nuclear Energy Science and Technolog...

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Problems solved by technology

However, in previous studies, the numerical simulation method mainly adopts the Reynolds time-average model, so the calculation results obtained are time-averaged thermal-hydraulic parameters, which cannot capture the real thermal-hydraulic parameters in the process of impinging jet flow in the local area of ​​the core outlet. In addition, because the Prandtl number of liquid metal sodium is much smaller than the Prandtl number of liquid water, the calculation error of Reynolds number Acacia assumption for liquid metal sodium in the traditional turbulence model is relatively large, so , the present invention will use the high-precision large eddy simulation method to calculate and study the impingement jet flow of liquid metal sodium, so as to obtain the accurate three-dimensional flow and the anisotropic turbulent pulsation and temperature oscillation characteristics existing in the mixing phenomenon in the impingement jet process

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