Fast reactor carbon dioxide power generation system and method

Abstract

The invention discloses a fast reactor carbon dioxide power generation system and method. The system comprises a fast reactor core, a sodium pump and a sodium helium heat exchanger, the sodium-helium heat exchanger is a shell-and-tube heat exchanger, sodium is located on the shell side of the heat exchanger, helium is located on the tube side of the heat exchanger, a fast reactor core, a sodium pump, the shell side of the sodium-helium heat exchanger and a pipeline connected with the shell side of the sodium-helium heat exchanger form a primary loop of the fast reactor carbon dioxide power generation system, an electric heating device is arranged in the primary loop, and the medium of the primary loop is sodium. The electric heating device can heat the medium to 200 DEG C or above. According to the method, helium belongs to inert gas and does not chemically react with sodium in a primary loop, and after the sodium-helium heat exchanger leaks, even if helium enters the sodium in the primary loop, chemical reaction does not occur, and nuclear safety accidents cannot be caused; under the condition of high temperature parameters, the supercritical carbon dioxide power generation cycle is more remarkable in efficiency advantage and better matched with fast reactor positioning, and the net efficiency can be improved by 2-4% compared with that of water working medium cycle at the 500 DEG C.

Description

technical field [0001] The invention belongs to the technical field of nuclear power, and in particular relates to a system and method for generating electricity from carbon dioxide from a fast reactor. Background technique [0002] At present, the fast reactor demonstration power station uses the once-through evaporator to heat the water with the heat generated by the reactor, generate steam, and drive the steam turbine to generate electricity. [0003] The system has the following problems: [0004] (1) The structure of the system is complex, especially the manufacture and operation of the once-through evaporator are difficult. [0005] (2) For the demonstration project of the fast reactor nuclear power plant currently under construction, the heat exchange tube of the evaporator is steam at 14MPa and 485°C, and the outside of the heat exchange tube is liquid sodium at 0.5MPa and 520°C. After the heat exchange tube leaks during operation, the sodium and Water and water va...

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Problems solved by technology

[0004] (1) The structure of the system is complex, especially the manufacture and operation of the once-through evaporator is difficult

[0005] (2) In the fast reactor nuclear power plant demonstration project currently under construction, steam at 14MPa and 485°C is inside the heat exchange tube of the evaporator, and liquid sodium at 0.5MPa and 520°C is outside the heat exchange tube. After the heat exchange tube leaks during operation, the sodium and Water and water vapor react with sodium water to produce hydrogen gas and explode, which will cause safety accidents and affect the safety of nuclear power units

[0006] (3) During the start-stop stage and the dry-wet state transition stage, the feedwater flow temperature and flow rate at the evaporator inlet are difficult to control, the evaporator outlet pressure fluctuates greatly, and the main steam temperature changes drastically, which affects the operation safety of the evaporator and steam turbine

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