Loading…

Heat transfer coefficients for bubbly molten salt nuclear reactors

Modeling the heat transfer process in a molten salt nuclear reactor is challenging due to the non-uniform distribution of fuel material and cooling fluid. Developing theoretical tools to calculate effective parameters and temperature distribution is essential for ensuring the safety of reactor desig...

Full description

Saved in:
Bibliographic Details
Published in:Nuclear engineering and design 2023-12, Vol.414, p.112549, Article 112549
Main Authors: Herrera-Hernández, E.C., Pérez-Valseca, A.D., Aguilar-Madera, C.G., Vázquez-Rodríguez, A.
Format: Article
Language:English
Subjects:
Citations: Items that this one cites
Items that cite this one
Online Access:Get full text
Tags: Add Tag
No Tags, Be the first to tag this record!
Description
Summary:Modeling the heat transfer process in a molten salt nuclear reactor is challenging due to the non-uniform distribution of fuel material and cooling fluid. Developing theoretical tools to calculate effective parameters and temperature distribution is essential for ensuring the safety of reactor design. In this study, we utilized the theory of volume averaging to upscale the local model for heat transport in a fluid fuel reactor, known as the Molten Salt Fast Reactor (MSFR). This upscaled model, which is valid for the bulk of the reactor core, combines the heat transport physics that occurs at the microscale through mathematically derived effective coefficients. These coefficients can be obtained by solving related boundary-value problems. The standalone core of the MSFR with inert gas bubbles was studied and the local thermal equilibrium (non-equilibrium) condition between the fluid and bubbles was found to be achievable for small (large) bubbles volume fractions. We present the numerically calculated values of the effective coefficients in terms of the volume fraction of bubbles under operating conditions and provide correlations to facilitate their use in computational codes.
ISSN:0029-5493
1872-759X
DOI:10.1016/j.nucengdes.2023.112549