CONVECTIVE HEAT TRANSFER IN THE BREST-OD-300 NUCLEAR REACTOR FUEL ROD

被引：0

作者：

Fedorovich, D. ^{[1
]}

Paramonova, I. ^{[1
]}

机构：

[1] Peter Great St Petersburg Polytech Univ, Inst Energy, St Petersburg 194064, Russia

来源：

PROCEEDINGS OF CONV-22: INT SYMP ON CONVECTIVE HEAT AND MASS TRANSFER, 2022 | 2022年

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中图分类号：

O414.1 [热力学];

学科分类号：

摘要：

The given work is devoted to the study of the heat transfer between the fuel rod's shell and coolant in the nuclear reactor BREST-OD-300. This fast neutron reactor now is building to realize a closed fuel cycle. Fuel rod in BREST-OD-300, according to the draft, consists of Mixed Nitride Uranium Plutonium (MNUP) fuel pellets, gap filled with helium and ferritic-martensitic steel core. The fuel element is bordered by pure lead coolant while it's forced circulation through the reactor core. Using all these materials provides a high level of safety, but the main problem is the leak of empirical data for the calculation of the heat transfer coefficient between fuel rod and lead coolant in the fuel assembly. Nusselt number is necessary for the calculation of the temperature fields in fuel elements and justification of the thermal reliability of the reactor, as it close the system of equations. The work was implemented based on data provided in scientific articles and other literature. Analytical calculations of the fuel rod in the center of the reactor core were carried out using many empirical relationships for Nusselt number. Finally, there were performed numerical simulations in the ANSYS package using one of the calculated heat transfer coefficients. Results of numerical simulations shows the temperature field in the fuel rod and the great value of temperature gradient caused by the using of helium in the gap.

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页数：8