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Marco Caramello

Observed names: CARAMELLO, MARCO, Caramello M., Caramello, Marco, M Caramello, M. Caramello, Marco CARAMELLO, Marco Caramello (Ansaldo Nucleare S. p. A. )
Institution: Ansaldo (Italy)
ORCID: 0000-0001-6562-1324
Observed ORCID: 0000-0001-6562-1324
OpenAlex ID: A5031568770
ALFRED reactor coolant system design
2020 · Alessandro Alemberti, Marco Caramello, et al. · Nuclear Engineering and Design · 77
Maturation of critical technologies for the DEMO balance of plant systems
2022 · L. Barucca, W. Hering, et al. · Fusion Engineering and Design · 50
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Thermal hydraulic analysis of a passively controlled DHR system
2017 · M. Caramello, M. Gregorini, et al. · Progress in Nuclear Energy · 26
Preliminary evaluation of ALFRED revised concept under station blackout
2020 · Vincenzo Narcisi, Fabio Giannetti, et al. · Nuclear Engineering and Design · 20
Helical coil thermal-hydraulic model for supercritical lead cooled fast reactor steam generators
2016 · Marco Caramello, Cristina Bertani, et al. · Applied Thermal Engineering · 17
Improvement of ALFRED thermal hydraulics through experiments and numerical studies
2023 · M. Caramello, M. Frignani, et al. · Nuclear Engineering and Design · 11
Application of a non-intrusive reduced order modeling approach to magnetohydrodynamics
2024 · M. Lo Verso, S. Riva, et al. · Physics of Fluids · 9
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SIRIO: An Experimental Facility for a New Heat Removal System Passively Controlled by Non-Condensable Gases
2018 · R. Marinari, M. Tarantino, et al. · The Proceedings of the International Conference on Nuclear Engineering (ICONE) · 9
Helical coil thermal hydraulic model
2014 · M Caramello, C Bertani, et al. · Journal of Physics Conference Series · 9
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Non-modal stability analysis of magneto-hydrodynamic flow in a single pipe
2024 · Matteo Lo Verso, Carolina Introini, et al. · Fundamental Plasma Physics · 7
The development of the research infrastructure in support of Alfred demonstrator implementation in Romania
2021 · Marin CONSTANTIN, Giacomo GRASSO, et al. · EMERG - Energy Environment Efficiency Resources Globalization · 6
Verification of RELAP5-3D code in natural circulation loop as function of the initial water inventory
2017 · C Bertani, N Falcone, et al. · Journal of Physics Conference Series · 5
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ATHENA Main Heat Exchanger Conceptual Design and Thermal-Hydraulic Assessment With RELAP5 Code
2023 · Tommaso Del Moro, Pierdomenico Lorusso, et al. · Journal of Nuclear Engineering and Radiation Science · 4
Lead Fast Reactor Thermal Hydraulic Testing Facilities in Support of the UK Advanced Modular Reactor Program
2023 · J. Liao, C. Stansbury, et al. · International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH) · 4
Lead Fast Reactor Thermal-Hydraulic Testing Facilities in Support of the UK Advanced Modular Reactor Program
2024 · J. Liao, C. A. Stansbury, et al. · Nuclear Technology · 3
Thermal-Hydraulic Characterization and Numerical Modeling with RELAP5 Code of ATHENA Secondary Loop
2023 · Tommaso Del Moro, Fabio Giannetti, et al. · Nuclear Technology · 3
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Numerical Analysis and Codes Qualification Based on SIRIO Experiments Within the PIACE Project
2023 · P. Lorusso, M. Tarantino, et al. · International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH) · 3
Preliminary numerical studies of an experimental facility for heat removal in natural circulation
2017 · C Bertani, M De Salve, et al. · Journal of Physics Conference Series · 3
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Enhancing computational efficiency in nuclear fusion through reduced order modelling: Applications in magnetohydrodynamics
2025 · Matteo Lo Verso, Stefano Riva, et al. · Fusion Engineering and Design · 2
The Passive Heat Removal Facility: Testing and Demonstrating an Innovative Decay Heat Removal System
2023 · M. Caramello, M. Frignani, et al. · Nuclear Technology · 2
The Versatile Loop Facility: A New Infrastructure for Testing Components and Systems of Lead Fast Reactor Technology
2023 · M. Caramello, M. Frignani, et al. · Nuclear Technology · 2
CFD Analysis of the Passive Heat Removal System in the Conceptual Design Phase of the Westinghouse Lead-Cooled Fast Reactor Under Air Cooling Conditions
2023 · Aidan Wimshurst, Graham Macpherson, et al. · International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH) · 2
Post-Test Analysis of SIRIO Facility Data by System Thermal-Hydraulic Codes for LFR Application
2023 · T. Moro, F. Giannetti, et al. · International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH) · 2
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Testing Innovative Decay Heat Removal System Through the Passive Heat Removal Facility (PHRF): Facility Description and Early Testing in Support to Lead Fast Reactor Development
2023 · M. Caramello, M. Frignani, et al. · International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH) · 2
GOTHIC Thermal Hydraulic Analysis of the Passive Heat Removal Facility to Support the Westinghouse Lead Fast Reactor Development
2023 · M.E. Durse, D.L. Wise, et al. · International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH) · 2
Works count: 51
Citation count: 289
H-index: 9
I10-index: 6
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