Fabio Giannetti


Titolo Pubblicato in Anno
Severe accident sensitivity and uncertainty estimation using MELCOR and RAVEN JOURNAL OF PHYSICS. CONFERENCE SERIES 2022
Decontamination and remediation of underground holes and testing of cleaning techniques based on the use of liquid cold decontaminant SUSTAINABILITY 2022
CFD Analysis and Optimization of the DEMO WCLL Central Outboard Segment Bottom-Cap Elementary Cell JOURNAL OF NUCLEAR ENGINEERING 2022
Transient analysis of a locked rotor/shaft seizure accident involving the EU-DEMO WCLL Breeding Blanket primary cooling circuits FUSION ENGINEERING AND DESIGN 2022
STEAM: a novel experimental infrastructure for the development of the DEMO BOP water coolant technology 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 19) 2022
Analysis of the thermal-hydraulic behavior of the EU-DEMO WCLL breeding blanket cooling systems during a loss of flow accident FUSION ENGINEERING AND DESIGN 2021
Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System FUSION ENGINEERING AND DESIGN 2021
Total loss of flow benchmark in CIRCE-HERO integral test facility NUCLEAR ENGINEERING AND DESIGN 2021
Pre-conceptual design of EU DEMO balance of plant systems: Objectives and challenges FUSION ENGINEERING AND DESIGN 2021
Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment FUSION ENGINEERING AND DESIGN 2021
Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger FUSION ENGINEERING AND DESIGN 2021
Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis FUSION ENGINEERING AND DESIGN 2021
Design and optimization of the secondary circuit for the WCLL BB option of the EU-DEMO power plant FUSION ENGINEERING AND DESIGN 2021
Study of the EU-DEMO WCLL breeding blanket primary cooling circuits thermal-hydraulic performances during transients belonging to LOFA category ENERGIES 2021
Experimental investigation on circe-hero for the eu demo pbli/water heat exchanger development ENERGIES 2021
Preliminary uncertainty quantification of the core degradation models in predicting the Fukushima Daiichi unit 3 severe accident NUCLEAR ENGINEERING AND DESIGN 2021
Thermal-hydraulic transient analysis of the FFTF LOFWOS Test #13 NUCLEAR ENGINEERING AND DESIGN 2021
Development of a RELAP5/MOD3.3 module for MHD pressure drop analysis in liquid metals loops. Verification and validation ENERGIES 2021
Improvements of RELAP5/Mod3.3 heat transfer capabilities for simulation of in-pool passive power removal systems ANNALS OF NUCLEAR ENERGY 2021


  • PE8_6


  • 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.


nuclear thermal hydraulics
Thermal-hydraulic transient analysis
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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