
Philippe Fillion
79%
Libre accès
42
Documents
Publications
Publications
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Non-uniform flows through a 8x34 PWR-type rod bundle: PIV and pressure drop measurementsSWINTH 2024 - Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal Hydraulics and Severe Accidents, Jun 2024, Dresde, Germany. pp.017
Communication dans un congrès
cea-04750812
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Adequacy analysis of critical flow experiments in the framework of the OECD-NEA ATRIUM project: CEA's contributionBEPU 2024 - Best Estimate Plus Uncertainty International Conference, May 2024, Lucca, Italy. pp.BEPU2024-353
Communication dans un congrès
cea-04706836
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Experimental qualification of a PIV and two-dye LIF temperature field measurement system with the "filling box" experiment in liquid flowSWINTH 2024, Jun 2024, Dresden, Germany. pp.056
Communication dans un congrès
cea-04675570
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Validation of the system code CATHARE3 on critical flow experiments in the framework of the OECD-NEA ATRIUM projectNUTHOS-14 - The 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, Aug 2024, Vancouver, Canada
Communication dans un congrès
cea-04711360
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METERO-V and PRIUS experimental programs: complementary separate effects tests for core mixing validationNURETH-20 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2023, Washington DC, United States. pp.1848-1861, ⟨10.13182/NURETH20-41365⟩
Communication dans un congrès
cea-04371973
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Application of SAPIUM guidelines to Input Uncertainty Quantification: the ATRIUM project20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 20), American Nuclear Society, Aug 2023, Washington DC, United States. pp.5734-5747, ⟨10.13182/NURETH20-40380⟩
Communication dans un congrès
cea-04191079
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Recommendations for new Experiments able to Better Characterize Flashing Flows in Nozzles for Improving Critical Flow Modelling in System CodesNURETH-20, Aug 2023, Washington, D.C., United States. pp.1502-1515, ⟨10.13182/NURETH20-40676⟩
Communication dans un congrès
cea-04390515
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Uniform and non-uniform flows through a PWR-type rod bundle with mixing grids at Reynolds number ranging from 800 up to 70000: pressure loss measurementsNURETH-20 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2023, Washington, D.C., United States. pp.1946-1956
Communication dans un congrès
cea-04371975
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METERO-V experimental set-up: presentation and first results. Uniform flows through a 8x34 PWR-type rod bundle with mixing grids at Reynolds number ranging from 800 to 70000: pressure loss measurementsICAPP 2023 - 2023 International Congress on Advances in Nuclear Power Plants, in conjunction with 38th Korea Atomic Power Annual Conference, Apr 2023, Gyeongju, South Korea. pp.233
Communication dans un congrès
cea-04464968
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Validation of the CATHARE 3 code based on Rod Bundle Heat Transfer (RBHT) benchmark data using two-fluid and three-field modelsNURETH-19 - The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, ANS; IAEA; NRG, Mar 2022, Brussels (online), Belgium
Communication dans un congrès
cea-04256708
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Qualification expérimentale d'un système de thermographie LIF-2DCongrès Francophone de Techniques Laser, Sep 2022, Leuven, Belgique
Communication dans un congrès
cea-04675537
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TPTF horizontal flow prediction by SYS-TH codes - Recent analyses made within the FONESYS networkThe 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Mar 2022, Brussels (online), Belgium. pp.35792
Communication dans un congrès
cea-04098083
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Horizontal stratification criteria of 8 system codes and direct confrontation to TPTF and Mantilla dataThe 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Mar 2022, Bruxelles (Virtual), Belgium. pp.35791
Communication dans un congrès
cea-04371870
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Droplet entrainment prediction in horizontal flow by SYS-TH codes - Mantilla experiments - Recent analyses made with the FONESYS networkThe 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Mar 2022, Brussels (Virtual), Belgium. pp.35793
Communication dans un congrès
cea-04371871
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SAPIUM - Follow-up activities, perspectives for future developmentNURETH-19 - 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Mar 2022, Bruxelles, Belgium
Communication dans un congrès
cea-04417233
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Assessment of stratification and entrainment models in CATHARE 3 code against TPTF-H and Mantilla experimentsNURETH-19 - The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Mar 2022, Bruxelles (Conference virtuelle), Belgium
Communication dans un congrès
cea-03985842
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SAPIUM - Selection of adequate test database for IUQ19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Mar 2022, Brussels (virtual), Belgium
Communication dans un congrès
cea-04440551
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Some progresses in system code modelling validation and application including 3D modellingATH'2020 - Advances in Thermal Hydraulics 2020, Oct 2020, Palaiseau, France. pp.661-674
Communication dans un congrès
cea-04220353
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Some Progress Made in Multiscale Analysis, Multiscale Simulation and Multi-Scale Validation in Reactor ThermalhydraulicsAdvances in Thermal Hydraulics 2020, Oct 2020, Palaiseau, France. pp.1105-1116
Communication dans un congrès
cea-04221795
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CATHARE-3 V2.1: The new industrial version of the CATHARE codeATH'20 - Advances in Thermal Hydraulics 2020, Oct 2020, Palaiseau, France. https://www.ans.org/pubs/proceedings/article-49142/
Communication dans un congrès
cea-04087378
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Development of the CATHARE 3 three-field model for simulations in large diameter horizontal pipesNURETH 2019 - 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2019, Portland, United States. pp.6271-6282
Communication dans un congrès
cea-04382534
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3-D SYS-TH: an OECD/NEA activity on multi-dimensional capabilities of thermalhydraulic system codesICAPP 2019 – International Congress on Advances in Nuclear Power Plants, May 2019, Juan-les-pins, France
Communication dans un congrès
cea-04221858
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Analysis of flow in rod bundles with velocity and temperature radial dissymmetry using CATHARE-3ICAPP 2019 - International Congress on Advances in Nuclear Power Plants, May 2019, Juan-les-Pins, France. pp.1-8
Communication dans un congrès
hal-02380403
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A mechanistic flashing model for critical flowrate prediction using CATHARE 3 system codeICAPP 2019 - 2019 International Congress on Advances in Nuclear Power Plants, May 2019, Juan-les-pins, France
Communication dans un congrès
cea-04217103
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Development of a systematic approach for quantification of thermal hydraulic code model input uncertaintyNURETH2019 - 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2019, Portland (OR), United States. pp.3054-3067
Communication dans un congrès
cea-04378696
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Revisiting the PIRT and scaling analysis within the frame of 3D system code modelling18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2019, Portland (OR), United States. pp.4109-4122
Communication dans un congrès
hal-04380272
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SAPIUM a systematic approach for input uncertainty quantificationBEPU-2018, May 2018, Lucca, Italy
Communication dans un congrès
cea-02338977
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Improved PWR LOCA simulations through refined core 3D simulations - An advanced 3D modelling and the associated METERO validation programNUTHOS-12 - 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety, Oct 2018, Qingdao, China
Communication dans un congrès
cea-04084545
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Assessing the accuracy of energy turbulent diffusion dispersion correlation in a porous two-fluid model dedicated to PWR core simulations14emes Journees d'etudes des Milieux Poreux - JEMP 2018, Oct 2018, Nantes, France
Communication dans un congrès
cea-02339457
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Prediction of annular flows in vertical pipes with new correlations for the CATHARE-3 three-field model17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), Sep 2017, Xi'An, China
Communication dans un congrès
cea-02434037
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3d core thermalhydraulic phenomena in pwr sblocas and iblocasNURETH-17, Sep 2017, Xian, China
Communication dans un congrès
hal-02433889
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Assessment of correlations for low pressure subcooled boiling flows studies using the drift flux model of FLICA codesNURETH-15 - The 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, May 2013, Pise, Italy
Communication dans un congrès
cea-05128940
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