This report describes the set up of a simplified three-dimension numerical model for the thermo-fluid dynamic analysis of an open square lattice reactor core lead cooled on the basis of a three-dimension CFD computer code developed by DIENCA UniBo. A preliminary assessment of the model has been performed by comparing its results with the T/H reactor core behavior predicted with a one-dimension independent channel model based on RELAP5 code. To this purpose the conceptual design of the LFR core developed in the framework of the ELSY EU collaborative project has been adopted as a reference. Moreover, in order to support the set-up of the model, some results of the CFD analysis for fuel bundle of liquid metal reactors have been considered.

Set up and Preliminary Assessment of a 3D Numerical Model for the Thermo-Fluid Dynamics Analysis of an Open Square Lattice Core of a Lead Cooled Reactor

Polidori, M.
2009

Abstract

This report describes the set up of a simplified three-dimension numerical model for the thermo-fluid dynamic analysis of an open square lattice reactor core lead cooled on the basis of a three-dimension CFD computer code developed by DIENCA UniBo. A preliminary assessment of the model has been performed by comparing its results with the T/H reactor core behavior predicted with a one-dimension independent channel model based on RELAP5 code. To this purpose the conceptual design of the LFR core developed in the framework of the ELSY EU collaborative project has been adopted as a reference. Moreover, in order to support the set-up of the model, some results of the CFD analysis for fuel bundle of liquid metal reactors have been considered.
Rapporto tecnico;Generation IV reactors;Reattori nucleari veloci;Termoidraulica dei reattori nucleari
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/20.500.12079/5355
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