This report deals with the validation of a thermo-fluid dynamic model for the lead-cooled reactor ELSY implemented on the finite element 3D FEM-LCORE code and developed by the DIENCA department of the University of Bologna. This model has been used for the investigation of pressure, velocity and temperature distributions inside the open square ELSY core. The main purpose of this work is to compare the FEM-LCORE results obtained by the University of Bologna with the results obtained by ENEA using a more validated code, under the same initial and boundary conditions. The 2D SIMMER-III code, which is widely used at international level for fast reactor thermo-fluid dynamic analysis, has been selected by ENEA for code to code result benchmarking. In spite of the less detailed 2D simulation of the SIMMER-III code, a substantial agreement has been found with the FEM-LCORE results regarding pressure, velocity and temperature profiles over the whole core region.

Benchmark on Thermo-Fluid Dynamics of an Open Square Lattice Core of a Lead Cooled Reactor with SIMMER-III and FEM-LCORE Codes

Bandini, G.
2010-09-15

Abstract

This report deals with the validation of a thermo-fluid dynamic model for the lead-cooled reactor ELSY implemented on the finite element 3D FEM-LCORE code and developed by the DIENCA department of the University of Bologna. This model has been used for the investigation of pressure, velocity and temperature distributions inside the open square ELSY core. The main purpose of this work is to compare the FEM-LCORE results obtained by the University of Bologna with the results obtained by ENEA using a more validated code, under the same initial and boundary conditions. The 2D SIMMER-III code, which is widely used at international level for fast reactor thermo-fluid dynamic analysis, has been selected by ENEA for code to code result benchmarking. In spite of the less detailed 2D simulation of the SIMMER-III code, a substantial agreement has been found with the FEM-LCORE results regarding pressure, velocity and temperature profiles over the whole core region.
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: http://hdl.handle.net/20.500.12079/5468
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