Fabio Giannetti

Pubblicazioni

Titolo Pubblicato in Anno
A novel ansys CFX – RELAP5 coupling tool for the transient thermal-hydraulic analysis of liquid metal systems PROGRESS IN NUCLEAR ENERGY 2025
Uncertainty quantification for severe-accident reactor modelling: Results and conclusions of the MUSA reactor applications work package ANNALS OF NUCLEAR ENERGY 2025
Pre-test analysis and thermal-hydraulic characterization of the versatile loop facility NUCLEAR ENGINEERING AND DESIGN 2025
Main outcomes of the Phebus FPT1 uncertainty and sensitivity analysis in the EU-MUSA project ANNALS OF NUCLEAR ENERGY 2024
A systematic approach for the adequacy analysis of a set of experimental databases. Application in the framework of the ATRIUM activity NUCLEAR ENGINEERING AND DESIGN 2024
RELAP5/Mod3.3 thermal-hydraulics characterization of the steam generator mock-up during operational transients in STEAM facility in support of the design of the DEMO WCLL BoP FUSION ENGINEERING AND DESIGN 2024
Pre-test analysis of low power operations of STEAM, the EU-DEMO steam generator mock-up facility FUSION ENGINEERING AND DESIGN 2024
An OpenFOAM multi-region solver for tritium transport modeling in fusion systems FUSION ENGINEERING AND DESIGN 2024
RELAP5-based thermal-hydraulic assessment of the STEAM facility for DEMO WCLL balance of plant analysis FUSION ENGINEERING AND DESIGN 2024
System code simulation of DEMO WCLL central outboard blanket equatorial cell operational transients FUSION ENGINEERING AND DESIGN 2024
Analysis of Steady State and Operational Transients of the Versatile Loop Facility in support to Lead Fast Reactor development Proceeding of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) 2024
Investigation of the fast flux test facility transient behavior during a loss of flow without scram test NUCLEAR ENGINEERING AND DESIGN 2024
Assessment of the relevancy of ENEA water loop facility with respect to ITER WCLL TBS water cooling system by considering their thermal-hydraulic performances FUSION ENGINEERING AND DESIGN 2024
Development and Optimization Criteria of Cementitious Mortars Used for the Immobilization of ILW Radioactive Waste 2024 31st International Conference on Nuclear Engineering - Volume 8: Decontamination and Decommissioning, Radiation Protection, and Waste Management 2024
Analysis of the experimental tests performed at NACIE-UP facility through a novel CFX-RELAP5 codes coupling NUCLEAR ENGINEERING AND DESIGN 2024
Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility ANNALS OF NUCLEAR ENERGY 2024
Performance evaluation of a coupled CFX-RELAP5 tool adopting experimental data from the TALL-3D facility NUCLEAR ENGINEERING AND DESIGN 2024
Re-scaling and Pre-test Analysis of SIRIO facility 41th UIT International Heat Transfer Conference 2024
Coupled CFD-STH modelling of the ATHENA lead-cooled pool-type facility 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-14) 2024
Loss-of-Heat-Sink Transient Simulation with RELAP5/Mod3.3 Code for the ATHENA Facility 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-14) 2024

ERC

  • PE8_6

KET

  • Big data & computing
  • Sustainable technologies & development

Interessi di ricerca

The research activity is focused on two-phase thermal-hydraulic transient analysis based on system TH computer programs. He acquired capability mainly in the safety analysis and TH best-estimate transient calculations, with the aid of RELAP5/mod3.3, RELAP5-3D©, TRACE and MELCOR computer programs, to enhance the safety performances and new system/component design for nuclear reactors (GEN III, GEN IV and fusion) and relative sensitivity analysis, as well through RAVEN, developed at INL. He is involved, in collaboration with ENEA, in the validation of such codes in liquid metals and the developing of a fusion version of RELAP5/mod 3.3 (for liquid metals and helical coil steam generators) and is a member of EU DEMO WCLL Breeding Blanket and Balance of Plant design team and ITER WCLL Water Cooling System for the Test Blanket System design team. He is also involved, in collaboration of Idaho National Laboratory (USA), in the validation of PHISICS/RELAP5-3D NK-TH coupled code for fast reactors mainly through the IAEA CRP Benchmark Analysis of FFTF Loss Of Flow Without Scram Test.

Keywords

nuclear thermal hydraulics
Thermal-hydraulic transient analysis
SEVERE ACCIDENT
nuclear fusion reactor
Fast neutron reactors
fusion reactor blanket
balance of plant
3D coupled dynamic analysis
Gen-IV lead cooled fast reactors
nuclear accidents

Gruppi di ricerca

Laboratori di ricerca

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