A5064979144- Nuclear Engineering Thermal-Hydraulics
- Nuclear reactor physics and engineering
- Nuclear Materials and Properties
- Heat Transfer and Boiling Studies
- Probabilistic and Robust Engineering Design
- Heat transfer and supercritical fluids
- Real-time simulation and control systems
- Computational Fluid Dynamics and Aerodynamics
- Fluid Dynamics and Mixing
- Nuclear Physics and Applications
- Simulation Techniques and Applications
- Nuclear and radioactivity studies
- Embedded Systems Design Techniques
- Fault Detection and Control Systems
- Reservoir Engineering and Simulation Methods
- Superconducting Materials and Applications
- Seismic Waves and Analysis
- Drilling and Well Engineering
- Heat Transfer Mechanisms
- Aerodynamics and Acoustics in Jet Flows
- Fluid Dynamics and Turbulent Flows
- Fluid Dynamics and Heat Transfer
Commissariat à l'Énergie Atomique et aux Énergies Alternatives
2007-2024
CEA Paris-Saclay
2007-2024
Université Paris-Saclay
2019-2024
Korea Atomic Energy Research Institute
2023
Becker Technologies (Germany)
2023
Lappeenranta-Lahti University of Technology
2023
Helmholtz-Zentrum Dresden-Rossendorf
2023
Direction des énergies
2011-2020
Laboratoire de Dynamique des Fluides
2019
The NEPTUNE project constitutes the thermal-hydraulic part of long-term Electricité de France and Commissariat à l'Energie Atomique joint research development program for next generation nuclear reactor simulation tools. This is also financially supported by Institut Radioprotection et Sûreté Nucléaire AREVA NP. aims at developing a new software platform advanced two-phase flow thermal hydraulics covering whole range modeling scales allowing easy multiscale multidisciplinary calculations....
Uncertainty analysis is a key element in nuclear power plant deterministic safety using best-estimate thermal-hydraulic codes and best-estimate-plus-uncertainty methodologies. If forward uncertainty propagation methods have now become mature for industrial applications, the input quantification (IUQ) on physical models still requires further investigations. The Organisation Economic Co-operation Development/Nuclear Energy Agency PREMIUM project attempted to benchmark available IUQ methods,...
A benchmark activity on two-fluid simulations of high-pressure boiling upward flows in a pipe is performed by 12 participants using different MCFD (Multiphase Computational Fluid Dynamics) codes and closure relationships. More than 30 conditions from DEBORA experiment conducted CEA are considered. Each case characterised the flow rate, inlet temperature, wall heat flux outlet pressure. High-pressure Freon (R12) at 14 bar 26 boiled 19.2mm heated over 3.5m. Flow rates range 2000 kg m−2 s−1 to...
This paper discusses the results of a computational activity devoted to prediction two-phase flows in subchannels and rod bundles. The capabilities FLICA-OVAP code have been tested against an extensive experimental database made available by Japanese Nuclear Power Energy Corporation (NUPEC) frame PWR subchannel bundle tests (PSBT) international benchmark promoted OECD NRC. herein addressed involve void fraction distributions boiling crisis phenomena bundles with uniform nonuniform heat flux...
<title>Abstract</title> Laser-induced fluorescence (LIF) visualization techniques are used to quantify a wide range of aqueous flows, whether monophasic or two-phase. Depending on the resolution cameras used, these generate large amount data that can be further study behavior flow. The use laser sheet stimulate fluorescent dyes seeded in flow, resulting fields local measurements, depending thickness sheet, and offering variety possibilities for comparison with different digital tools....
The FONESYS network of system code developers made a benchmark eight codes against Two-Phase Critical Flow (TPCF) experiments, which updated the state art in TPCF modelling.0-D and 1-D models still have large prediction errors particularly for slightly sub-cooled inlet conditions.Three main reasons explain these errors: (1) 3D effects complex geometry, (2) nucleation delay, depends on non-well-controlled parameters such as purity water or status metallic surface, (3) interfacial transfers...
The analysis of Large Break-Loss Coolant Accidents (LB-LOCAs) required specific experimental programs to investigate large-scale 3D effects, particularly during downcomer refill and core reflooding.SB-LOCAs (Small Break) IB-LOCAs (Intermediate also encounter significant effects in due the radial power profile, with crossflows diffusion-dispersion.Other transients such as steam line break are sensitive all mixing phenomena core.Then a need exists for more precise validation each processes,...