Two-dimensional transient thermal analysis of a fuel rod by finite volume method

Rhayanne Yalle Negreiros Costa; Mário Augusto Bezerra da Silva; Carlos Alberto Brayner de Oliveira Lira

doi:10.15392/bjrs.v8i3B.617

Autores/as

Rhayanne Yalle Negreiros Costa ,
Mário Augusto Bezerra da Silva Universidade Federal de Pernambuco
Carlos Alberto Brayner de Oliveira Lira Universidade Federal de Pernambuco

DOI:

https://doi.org/10.15392/bjrs.v8i3B.617

Palabras clave:

Finite Volume Method, Thermal analysis, AP1000 nuclear reactor.

Resumen

One of the greatest concerns when studying a nuclear reactor is the warranty of safe temperature limits all over the system at all times. The preservation of core structure along with the constraint of radioactive material into a controlled system are the main focus during the operation of a reactor. The purpose of this paper is to present the temperature distribution for a nominal channel of the AP1000 reactor developed by Westinghouse Co. during steady-state and transient operations. In the analysis, the system was subjected to normal operation conditions and then to blockages of the coolant flow. The time necessary to achieve a new safe stationary stage (when it was possible) was presented. The methodology applied in this analysis was based on a two-dimensional survey accomplished by the application of Finite Volume Method (FVM). A steady solution is obtained and compared with an analytical approach that disregards axial heat transport to determine its relevance. The results show the importance of considering axial heat transport in this type of study. A transient analysis shows the behavior of the system when submitted to coolant blockage at channel’s entrance. Three blockages were simulated (10%, 20% and 30%) and the results show that, for a nominal channel, the system can still be considerate safe (there’s no bubble formation until that point).

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Biografía del autor/a

Mário Augusto Bezerra da Silva, Universidade Federal de Pernambuco

Nuclear Energy Engineering (DEN)
Carlos Alberto Brayner de Oliveira Lira, Universidade Federal de Pernambuco

Nuclear Energy Engineering (DEN)

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