In the context of GEN-IV Heavy Liquid Metal-cooled reactors safety studies, the flow blockage in a Fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR Fuel Assembly. The present document is a first step towards a detailed analysis of such phenomena, and a CFD model and approach is presented to have a detailed thermo-fluid dynamic picture in the case of blockage. In particular the closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED has been modeled and computed. At this stage, the details of the spacer grids have not been included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, have not been included in this analysis. Results indicate that critical conditions, with clad temperatures exceeding~1000°C, are reached with blockage larger than 30% in terms of area fraction.

Analisi di uno scenario di flow-blockage

Di Piazza, Ivan
2013

Abstract

In the context of GEN-IV Heavy Liquid Metal-cooled reactors safety studies, the flow blockage in a Fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR Fuel Assembly. The present document is a first step towards a detailed analysis of such phenomena, and a CFD model and approach is presented to have a detailed thermo-fluid dynamic picture in the case of blockage. In particular the closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED has been modeled and computed. At this stage, the details of the spacer grids have not been included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, have not been included in this analysis. Results indicate that critical conditions, with clad temperatures exceeding~1000°C, are reached with blockage larger than 30% in terms of area fraction.
Generation IV reactors;Sicurezza nucleare;Analisi incidentale;Termoidraulica dei reattori nucleari
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/20.500.12079/7642
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