Fabio Giannetti

Pubblicazioni

Titolo Pubblicato in Anno
Dynamic event tree analysis of a severe accident sequence in a boiling water reactor experiencing a cyberattack scenario ANNALS OF NUCLEAR ENERGY 2023
Cutting-edge R&D activities of CIRTEN in support of the Technology Park annexed to the Italian National Repository of radioactive waste IL NUOVO CIMENTO C 2023
Subchannel Analysis of LFR Wire-Wrapped Fuel Bundle with RELAP5-3D NUCLEAR TECHNOLOGY 2023
Estimation of Cesium Penetration Into Concrete Structures After an Experimental Decontamination Activity Conference proceedings - International Conference on Radioactive Waste Management and Environmental Remediation 2023
Post-Test Analysis of SIRIO Facility Data by System Thermal-Hydraulic Codes for LFR Application Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) 2023
CFD - STH Code Coupling for the Thermal Hydraulic Analysis of NACIE-UP Experimental Facility Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) 2023
The STEAM Facility: Design and Analysis Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) 2023
Status, Design and Thermal-Hydraulic Analyses of the Steam Facility for DEMO WCLL Balance of Plant Systems Proceedings of the 2023 30th International Conference on Nuclear Engineering ICONE30 2023
STEAM experimental facility. A step forward for the development of the EU DEMO BoP Water Coolant Technology ENERGIES 2023
The design of water loop facility for supporting the WCLL breeding blanket technology and safety ENERGIES 2023
PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project ENERGIES 2023
IAEA'S COORDINATED RESEARCH PROJECTS ON THERMAL HYDRAULICS OF FAST REACTORS International Conference on Nuclear Engineering, Proceedings, ICONE 2023
Preliminary MHD pressure drop analysis for the prototypical WCLL TBM with RELAP5/MOD3.3 FUSION ENGINEERING AND DESIGN 2022
Tokamak cooling systems and power conversion system options FUSION ENGINEERING AND DESIGN 2022
Maturation of critical technologies for the DEMO balance of plant systems FUSION ENGINEERING AND DESIGN 2022
Dynamic event tree analysis as a tool for risk assessment in nuclear fusion plants using RAVEN and MELCOR IEEE TRANSACTIONS ON PLASMA SCIENCE 2022
Analysis of EU-DEMO WCLL Power Conversion System in Two Relevant Balance of Plant Configurations: Direct Coupling with Auxiliary Boiler and Indirect Coupling SUSTAINABILITY 2022
Transient analysis of OSU-MASLWR with RELAP5 JOURNAL OF PHYSICS. CONFERENCE SERIES 2022
Evaluation of the thermal-hydraulic performances of a once-through steam generator in nuclear fusion applications JOURNAL OF PHYSICS. CONFERENCE SERIES 2022
Analysis of Fukushima Daiichi unit 4 spent fuel pool using MELCOR JOURNAL OF PHYSICS. CONFERENCE SERIES 2022

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

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