TAILIEUCHUNG - Hydraulic and statistical study of metastable phenomena in PWR rod bundles

The present work aims at improving our understanding of nefarious reorganisation phenomena observed by numerous studies in the flow large-scale structures. 3D simulations allowed identifying two distinct reorganisations consisting in a sign change for either a transverse velocity in rod-to-rod gaps or for a subchannel vortex. | Hydraulic and statistical study of metastable phenomena in PWR rod bundles EPJ Nuclear Sci. Technol. 6 6 2020 Nuclear Sciences F. Muller published by EDP Sciences 2020 amp Technologies https epjn 2019057 Available online at https REGULAR ARTICLE Hydraulic and statistical study of metastable phenomena in PWR rod bundles Florian Muller CEA Saclay DEN DANS DM2S STMF LMSF 91191 Gif-sur-Yvette France Received 30 October 2019 Accepted 15 November 2019 Abstract. The analysis of fuel rod bundle flows constitutes a key element of Pressurized-Water Reactors PWR safety studies. The present work aims at improving our understanding of nefarious reorganisation phenomena observed by numerous studies in the flow large-scale structures. 3D simulations allowed identifying two distinct reorganisations consisting in a sign change for either a transverse velocity in rod-to-rod gaps or for a subchannel vortex. A Taylor frozen turbulence hypothesis was adopted to model the evolution of large-scale 3D structures as transported-2D. A statistical method was applied to the 2D field to determine its thermodynamically stable states through an optimization problem. Similarities were obtained between the PWR coherent structures and the stable states in a simplified 2D geometry. Further 2D simulations allowed identifying two possible flow bifurcations each related to one of the reorganisations observed in 3D simulations laying the foundations for a physical explanation of this phenomenon. 1 Introduction both for enhancing their characterization and for identifying their origin with the long-term goal of developing small-scales Insufficient flow thermal mixing in the rod bundles within a models suited for this type of flow. Pressurized-Water Reactor PWR can lead to a boiling Little concrete information can be found in the crisis which is nefarious for the reactor operational safety. literature on the reorganisation phenomenon. This is due Mixing grids are typically used to

• Năng lượng
• Nefarious reorganisation phenomena
• Large scale structure
• Rod to rod gap
• PWR rod bundles
• Pressurized Water Reactor
• Nuclear engineering and technology
• Fatigue crack growth characteristics
• Nitrogen alloyed type 347 stainless steel
• Operating conditions of a pressurized water reactor
• Different fatigue crack growth rates
• Nuclear science and engineering
• Validation of computational fluid dynamics calculation
• Rossendorf coolant mixing model flow measurements
• Primary loop of coolant in a pressurized water reactor model
• Pressurized water reactor model
• The FLUENT code
• Hydrogen mitigation system during large break loss of coolant accident
• Two loop pressurized water reactor
• Core Concrete Interaction
• Loss of Coolant Accident
• Passive Autocatalytic Recombiner
• The reactor vessel
• The steam generators
• The reactor coolant pumps
• Pressurized Water Reactor Systems
• BWR reactor plant design
• Reactor coolant system
• Nonlinear control for core power
• Pressurized water nuclear reactors using constant axial offset strategy
• Pressurized water nuclear reactors
• Constant axial offset strategy
• The given region
• Two point nuclear reactor model
• Communications in Physics
• Simulation for neutron transport in pwr reactor moderator
• Evaluation for proper thickness of light water reflector
• Pwr reactor moderator
• Monte Carlo method
• Fuel design
• Long life core
• Neutronic analysis
• SRAC code system
• Nuclear energy
• Supercritical water reactor
• Safety performance
• System code development
• Pressurized water reactors
• Small modular reactor
• AP1000 reactor
• Thermal hydraulics analysis
• AP1000 fuel assembly design
• Reactor pressure vessel
• Structural integrity
• Probabilistic assessment
• Pressurized thermal shock
• Light water reactor
• Simulation of reactivity initiated accident transients
• UO2 M5® fuel rods
• ALCYONE V1
• 4 fuel performance code
• Pressurized Water ReactorPressurized Water Reactor
• Reactivity initiated Accident
• Numerical prediction of a flashing flow
• Saturated water at high pressure
• Feedwater line break
• Uncertainty analysis of ROSA LSTF test
• PKL counterpart test concerning PWR hot leg break LOCAs
• Large scale test facility
• Primarkreislaufe Versuchsanlage
• Eddy current testing
• Bottom mounted instrumentation
• Developed ECT system
• BMI mock up
• Comparison of oxide layers formed
• The low cycle fatigue crack surfaces
• Alloy 690 and 316 SS tested in a simulated PWR environment
• Low cycle fatigue
• Simulated pressurized water reactor
• Loading pattern
• Nuclear Power Plant
• Neutron delayed fraction
• Moderator temperature feedbacks
• FP chemical behaviour
• Effect of the oxygen potential
• Severe accident condition
• Three highly reactive FP
• PWR MOX assembly
• Group constants
• Benchmark data
• Transient analyses