Visual experiment device and method suitable for rod bundle channel flow boiling heat transfer

A rod bundle channel and boiling heat transfer technology, which is applied in the field of nuclear power, can solve the problems of inability to obtain two-phase flow characteristics in the inner region of rod bundles, and achieve the effects of visualization, high heat resistance and small thermal expansion coefficient.

Pending Publication Date: 2021-10-15
SHANGHAI JIAO TONG UNIV
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Problems solved by technology

Although this existing technology has preliminarily realized the visualization experiment of two-phase flow, the characteristics of two-phase flow in the inner region of the rod bundle cannot be obtained

Method used

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  • Visual experiment device and method suitable for rod bundle channel flow boiling heat transfer
  • Visual experiment device and method suitable for rod bundle channel flow boiling heat transfer
  • Visual experiment device and method suitable for rod bundle channel flow boiling heat transfer

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Embodiment Construction

[0037] The present invention will be described in detail below in conjunction with specific embodiments. The following examples will help those skilled in the art to further understand the present invention, but do not limit the present invention in any form. It should be noted that those skilled in the art can make several changes and improvements without departing from the concept of the present invention. These all belong to the protection scope of the present invention.

[0038] Such as figure 1 As shown, the present invention provides a visual experiment device suitable for flow boiling heat transfer in a rod bundle channel, including: an experimental circuit part and a data acquisition part, the experimental circuit part includes a visual experiment body and a high-voltage power supply 12, and the visual experiment body includes an experimental Rod bundle; after the deionized water in the experimental circuit enters the visualization experiment body, it is heated by th...

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Abstract

The invention provides a visual experiment device suitable for rod bundle channel flow boiling heat transfer, the device comprises an experiment loop part and a data acquisition part, the experiment loop part comprises a visual experiment body and a high-voltage power supply, and the visual experiment body comprises an experiment rod bundle; deionized water in the experiment loop part enters the visual experiment body and then is heated by the experiment rod bundle, and water at the outlet of the visual experiment body is cooled by a cooling agent and then enters the visual experiment body again; an ITO electroplated layer is arranged on the experimental rod bundle, the ITO electroplated layer is connected with the high-voltage power supply, and a coolant in the rod bundle channel absorbs heat generated by the ITO film and then generates bubbles; and the data acquisition part acquires transient bubble characteristics of two-phase flow in the internal area of the rod bundle channel through the visual experiment body. A set of closed experimental loop is designed, the visualization requirement of the rod bundle channel is met, and the flow, the water temperature, the wall temperature and the heating power in the rod bundle channel are obtained through the data acquisition system part while the transient steam bubbles of the rod bundle channel are obtained.

Description

technical field [0001] The invention relates to the technical field of nuclear power, in particular to a visualization experimental device and method suitable for flow boiling heat transfer in rod bundle channels. Background technique [0002] The power density of nuclear reactors is much higher than that of conventional active devices, and the reactor core is in a high-temperature and high-pressure environment. In order to increase the heat exchange area of ​​nuclear fuel, fuel bundles are usually arranged in a cell bundle structure. The coolant in the primary circuit of the reactor flows through the rod bundle assembly and takes away the heat generated by the core. It is of great significance to study the flow and heat transfer characteristics in the rod bundle channel to ensure the safety of the reactor core. [0003] For the pressurized water reactor core, there is a single-phase flow in the core under normal operating conditions. When a small breach accident occurs in...

Claims

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): G21C17/00G01N25/20
CPCG21C17/00G01N25/20Y02E30/30
Inventor 顾汉洋张琦刘莉刘利民丛腾龙
Owner SHANGHAI JIAO TONG UNIV
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