This report, carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP) of the University of Pisa in collaboration with ENEA Bologna Research Centre, illustrates preliminary numerical results and the model for fuel dispersion in a LFR reactor types, i.e. the MYRRHA FASTEF reactor. The first part of this work shows the generalities and the historical background of the SIMMER-III code and some relevant integral applications. In this context the SIMMER-III code is able to be implemented in safety analysis of sodium-cooled fast reactors, lead and LBE-cooled fast reactors and also with some limitations in molten salt reactors and light water reactors. The second part focuses on MYRRHA-FASTEF modelled by SIMMER-III, highlighting how each component was represented. The reactor was simulated by a two-dimensional R-Z geometry with 38x89 cell mesh. After this introductive part concerning the description of the set-up model, steady state and transient results are reported. Steady state condition occurs about twenty seconds after the start of the simulation and this report shows the most relevant results obtained for temperature trends and profiles, both in the core and in the PHX, and for velocity and mass flow rate. Preliminary transient results were analyzed, i.e. during an Unprotected Loss Of Flow (ULOF) accident, and results were compared with those obtained with a RELAP5 code. A fuel dispersion transient was also simulated, comparing the effect of fuel porosity on the fuel dispersion inside the pool

Loss of core integrity in a LFR system: models and preliminary numerical analysis

Bandini, Giacomino
2012

Abstract

This report, carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP) of the University of Pisa in collaboration with ENEA Bologna Research Centre, illustrates preliminary numerical results and the model for fuel dispersion in a LFR reactor types, i.e. the MYRRHA FASTEF reactor. The first part of this work shows the generalities and the historical background of the SIMMER-III code and some relevant integral applications. In this context the SIMMER-III code is able to be implemented in safety analysis of sodium-cooled fast reactors, lead and LBE-cooled fast reactors and also with some limitations in molten salt reactors and light water reactors. The second part focuses on MYRRHA-FASTEF modelled by SIMMER-III, highlighting how each component was represented. The reactor was simulated by a two-dimensional R-Z geometry with 38x89 cell mesh. After this introductive part concerning the description of the set-up model, steady state and transient results are reported. Steady state condition occurs about twenty seconds after the start of the simulation and this report shows the most relevant results obtained for temperature trends and profiles, both in the core and in the PHX, and for velocity and mass flow rate. Preliminary transient results were analyzed, i.e. during an Unprotected Loss Of Flow (ULOF) accident, and results were compared with those obtained with a RELAP5 code. A fuel dispersion transient was also simulated, comparing the effect of fuel porosity on the fuel dispersion inside the pool
Rapporto tecnico;Generation IV reactors;Termoidraulica dei reattori nucleari;Analisi di sicurezza
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/20.500.12079/7478
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