Simulated heat exchange test system and use method thereof

Through the simulated heat exchange test system, the cooling process of core melt is simulated using different cycle modes, the problem of the inability to obtain the long-term cooling heat exchange performance parameters of the core melt is solved, and the effect of improving the retention success rate and cooling efficiency of core melt is achieved, providing data support for the design after the nuclear power plant accident.

CN115188510BActive Publication Date: 2025-03-28CHINA NUCLEAR POWER ENGINEERING CO LTD
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Patent Information

Application Number
CN202210833500.8
Authority / Receiving Office
CN · China
Patent Type
Patents(China)
Current Assignee / Owner
Filing Date
2022-07-15
Publication Date
2025-03-28
Estimated Expiration
2042-07-15

AI Technical Summary

Technical Problem

The prior art cannot effectively obtain the heat exchange performance parameters of the core melt in the long-term cooling stage of the core trap, especially in dealing with serious accidents in nuclear power plants above 1000MWe.

Method used

It provides a simulated heat exchange test system, including low-level water tanks, high-level water tanks, cooling channels, thermal hydraulic adjustment devices and a variety of valves. It simulates the cooling process of core melt through different cycle modes (forced circulation, natural circulation, etc.), and collects and analyzes a large amount of test data.

Benefits of technology

Through the simulation test of this system, the success rate of core melt retention and cooling efficiency can be effectively improved, and basic data can be provided for engineering design, helping to improve the safety and reliability of nuclear power plants after serious accidents.

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Abstract

The invention discloses a simulated heat exchange test system and a method for using the same. The system comprises: a low-level water tank; a high-level water tank; a cooling channel comprises: a first pipeline, a test section pipeline, a second pipeline, a third pipeline, and a fourth pipeline connected in sequence; a sixth pipeline is connected to the low-level water tank and the fourth pipeline; a seventh pipeline is connected to the low-level water tank and the first pipeline; a thermal hydraulic regulating device is arranged at any position on the fourth pipeline and the first pipeline; an eighth pipeline is connected to the third pipeline and the low-level water tank; a ninth pipeline is connected to the high-level water tank and the first pipeline; a tenth pipeline is connected to the high-level water tank and the ninth pipeline; an eleventh pipeline is connected to the third pipeline and the high-level water tank; a first heat source is in contact with the test section pipeline, and is used to provide a first preset heat flux density for the core melt inside the inner crucible to be transferred to the inside of the test section pipeline after a serious accident in a nuclear power plant. The system effectively improves the success rate of core melt retention and cooling efficiency, and provides basic data for engineering design.
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