Method for analyzing thermal-hydraulic characteristics of reactor core of lead-bismuth fast reactor

Abstract

The invention discloses a method for analyzing thermal-hydraulic characteristics of a lead-bismuth fast reactor core. The method comprises the following steps: 1, establishing a coolant flow heat exchange characteristic analysis model; 2, establishing an inter-box flow heat exchange characteristic analysis model; 3, establishing a thermal characteristic analysis model of the fuel rod; 4, establishing a multi-flow-field fluid-structure interaction heat exchange model; and 5, establishing a multi-flow-field fluid-solid coupling solving method. According to the method, a positioning grid and a rod cluster structure in a lead-bismuth fast reactor core are simplified, and the influence of component boxes and inter-box flow on the physical quantity of a coolant in a component is considered in the calculation process; through simplification of inter-box flow and wire winding, inter-box flow temperature distribution, fuel rod temperature distribution and a coolant flow field and temperature field can be accurately calculated, consumption of calculation resources is greatly reduced, the calculation speed is increased, and the calculation efficiency is improved. And an efficient and accurate novel computational fluid mechanics numerical simulation method is provided for analysis of the thermotechnical hydraulic characteristics of the whole reactor core of the lead-bismuth fast reactor in practical engineering application.

Description

technical field [0001] The invention belongs to the technical field of safety analysis of nuclear reactors, and in particular relates to a method for analyzing thermal and hydraulic characteristics of a reactor core of a lead-bismuth fast reactor under normal operation and accident conditions. Background technique [0002] The reactor core is the core component of the nuclear power plant system, and the temperature distribution of the internal coolant and fuel rods is crucial to the safety of the reactor. The lead-bismuth fast reactor core is different from traditional light water reactors. The lead-bismuth alloy is used as the coolant. The rod bundles are usually arranged in a hexagonal component box in the form of a regular triangle or a rectangle. There are small gaps through which coolant flows. The inter-box flow plays an important role in accident conditions, especially in the case of passive decay heat removal and lack of coolant due to blockage. Excess heat inside ...

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Problems solved by technology

Aiming at the complex rod bundle structure in the reactor core and the problem that the flow between the boxes does not flow in the core and the heat transfer characteristics are complicated, the present invention is based on computational fluid dynamics to establish a coolant flow heat transfer characteristic analysis model, fuel rod thermal engineering The characteristic analysis model and the flow and heat transfer characteristic model of the box-to-box flow are established, and the fluid-solid coupling heat transfer model is established. The temperature distribution of the core fuel rods and the flow field and temperature field of the coolant are calculated by the fluid-solid coupling solution method, and the thermal engineering of the reactor core is realized. Analysis of hydraulic properties

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