Georges REPETTO
- Service de l'Ingénierie et des Projets de Recherche (IRSN/PSN-RES/SIPR)
Publications
Publications
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VIKTORIA experiments in the frame of R&D project on sump filtration during a Loss Of Coolant Accident: effect of the type of filter, the mass of fiber and the presence of zinc21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21), Thermal-Hydraulics Division of the Korean Nuclear Society (KNS); American Nuclear Society, Aug 2025, Busan, South Korea |
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On the progress made in safety assessment and severe accident management as part of the french Fukushima post-accident research programmeThe 11th European Review Meeting on Severe Accident Research (ERMSAR2024), Königlich Technische Hochschule Stockholm (KTH), May 2024, Stockholm, Sweden |
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COAL reflooding experiments during a Loss Of coolant Accident: Effect of the water flow rate and the residual power with ballooned rods14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-14), Canadian Nuclear Society (CNS), Aug 2024, Vancouver, Canada |
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COAL experiments investigating the reflooding of a 7 X 7 rod bundle during a Loss Of coolant Accident: Effect of a partially blocked area with ballooned rods20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), American Nuclear Society, Aug 2023, Washington, United States. pp.1910-1923, ⟨10.13182/NURETH20-40021⟩ |
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Coal experiments investigating the reflooding of a 7 x 7 rods bundle during a loss of coolant accident thermalhydraulics resultsThe 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), ANS - American Nuclear Society, Mar 2022, Bruxelles (ON LINE), Belgium |
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Simulation de l'écoulement dispersé vapeur/gouttes dans des conditions d'APRP29e Congrès français de thermique, Société Française de Thermique, Jun 2021, Belfort, France. ⟨10.25855/SFT2021-037⟩ |
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Operation Extension of 900MWe NPPs: French TSO Main Conclusions regarding Long Term Sump Performance after a Loss of Coolant AccidentRCCS - 2021 - Specialist Workshop on Reactor core and containment cooling systems – long term management and reliability, OECD/NEA, Oct 2021, Virtual (on line), Slovakia. 5 p |
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VIKTORIA experiments on sump filtration during a Loss Of Coolant Accident (DBA)Specialist Workshop on Reactor core and containment cooling systems – long term management and reliability, OECD NEA, Oct 2021, Virtual Worshop, Slovakia |
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VIKTORIA experiments investigating the filtering system in the sump of a pwr after a Loss Of Coolant Accident : Part II Downstream effects18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society, Aug 2019, Portland, United States. pp.ISBN 9781510893450 |
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Viktoria experiments investigating the filtering system in the sump of a pwr after a loss of coolant accident : Part I Physical /chemical effects on strainer head loss evolutionNURETH 2019 - 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2019, PORTLAND, United States. pp.407-420 |
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Analyse des transferts thermiques d'un écoulement vertical vapeur/gouttes dans une géométrie tubulaire27e Congrès français de thermique, Jun 2019, Nantes, France |
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Identification des transferts thermiques d'un écoulement vertical vapeur/gouttes dans une géométrie tubulaireCongrès français de thermique, May 2018, Pau, France |
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CoreSOAR Core Degradation SOAR Update Status May 20178th European Review Meeting on Severe Accident Research (ERMSAR - 2017), May 2017, Varsovie, Poland |
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CORE COOLABILITY IN LOSS OF COOLANT ACCIDENT: THE COAL EXPERIMENTS INVESTIGATING THE THERMAL HYDRAULICS OF A ROD BUNDLE WITH BLOCKED AREA DURING THE REFLOODING2017 International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), American Nuclear Society, Sep 2017, Xi’an, China |
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The PERFROI project investigating the thermal mechanical behaviour of a fuel rod assembly during a loss of coolant accidentWater Reactor Fuel Performance Meeting 2017 (TopFuel 2017), Korean Nuclear Society (KNS), Sep 2017, Jeju City, Jeju Island, South Korea |
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Heat transfer inside a blocked region of a rod bundle during a LOCA-simulations and experiments9th World Conference on Experimental Heat Transfer, Fluid Mechanics and Thermodynamics (EXHFT-9), Jun 2017, Iguazu, Brazil |
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Study on Heat Transfer Inside a Blocked Region of a Rod Bundle During a LOCA - Model and Experimental Approach17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics NURETH-17, Sep 2017, Xian, China |
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Thermal hydraulics behavior of a rod bundle with partially blocked area during the reflooding phase of the Loss Of Cooling AccidentNUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, American Nuclear Society and Korean Nuclear Society, Oct 2016, Gyeongju, South Korea |
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Experimental study on heat transfer inside a blocked region of a rod bundle during a LOCA11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operations and Safety - NUTHOS 11, Oct 2016, Gyeongju, South Korea |
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CORE COOLABILITY IN LOSS OF COOLANT ACCIDENT: THE COAL EXPERIMENTS16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2016, American Nuclear Society, Aug 2015, CHICAGO, United States |
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Core coolability in loss of coolant accident: the PERFROI projectTop Fuel 14, American Nuclear Society, Sep 2014, Sendai (Japan), Japan |
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New insights in the thermal hydraulics behaviour during reflooding of a high temperature debris bed15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-15), American Nuclear Society (ANS), May 2013, Pisa, Italy |
Experimental program on debris reflooding (PEARL) – Results on PRELUDE facilityThe 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), Sep 2011, Toronto, Canada. pp.176-186, ⟨10.1016/j.nucengdes.2012.11.024⟩ |
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Quench front progression in a superheated porous medium: experimental analysis and model developmentNURETH 14 2011 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Sep 2011, Toronto, Canada. pp.0 |
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DRACCAR, a new 3D-thermal mechanical computer code to simulate LOCA transient on Nuclear Power Plants Status of the development and the validationInternational Congress on Advances in Nuclear Power Plants 2009 (ICAPP 2009), American Nuclear Society (ANS), May 2009, Tokyo, Japan |
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Preliminary Analysis of Phebus FPT3 Experiment with the Severe Accident ICARE2 Code14th International Conference on Nuclear Engineering (ICONE-14), ASME – American Society of Mechanical Engineers, Jul 2006, Miami, United States. pp.85-94, ⟨10.1115/ICONE14-89191⟩ |
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Application of the Porous Medium Heat transfer model of ICARE/CATHARE code against debris bed and "bundle" " experimentsEUROTHERM Seminar 81 – Computational Thermal Radiation in Participating Media 2007, École des Mines d’Albi-Carmaux (EMAC), Jun 2007, Albi, France |
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Analysis of the FPT-0 and FPT-2 PHEBUS FP experiments using porous medium geometry with the ICARE2 CODE11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11), American Nuclear Society (ANS), Oct 2005, Avignon, France |
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Analyses of the PHEBUS-LOCA tests 215-R and 218 using FRETA-F V1.1 CODE11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11), American Nuclear Society (ANS); Société Française d’Énergie Nucléaire (SFEN), Oct 2005, Avignon, France |
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Analysis of the FPT0, FPT1 and FPT2 experiments of the PHÉBUS FP program investigating in vessel phenomena during a LWR accident10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), American Nuclear Society (ANS), Oct 2003, Séoul, South Korea |
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MRI as a tool for nuclear safetyItalian-French International Conference on Magnetic Resonance, Sep 2022, Milan, Italy |