TAILIEUCHUNG - Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. | Contribution to the study of fission products release from nuclear fuels in severe accident conditions effect of the pO2 on Cs Mo and Ba speciation EPJ Nuclear Sci. Technol. 6 2 2020 Nuclear Sciences C. Le Gall et al. published by EDP Sciences 2020 amp Technologies https epjn 2019058 Available online at https REGULAR ARTICLE Contribution to the study of fission products release from nuclear fuels in severe accident conditions effect of the pO2 on Cs Mo and Ba speciation Claire Le Gall1 Fabienne Audubert1 Jacques Léchelle1 Yves Pontillon1 and Jean-Louis Hazemann2 1 CEA DEN DEC Cadarache 13108 Saint-Paul-Lez-Durance France 2 Institut Néel CNRS UGA amp CRG FAME ESRF 38041 Grenoble cedex 9 France Received 30 October 2019 Accepted 15 November 2019 Abstract. The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically the speciation of Cs Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields providing experimental data on Pressurized Water Reactor PWR MOX fuel behaviour submitted to severe accident conditions and related FP speciation going further in the understanding of FP speciation mechanisms at different stages of a severe accident developing a method to study volatile FP behaviour involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during .

• Năng lượng
• FP chemical behaviour
• Effect of the oxygen potential
• Severe accident condition
• Three highly reactive FP
• 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
• Nefarious reorganisation phenomena
• Large scale structure
• Rod to rod gap
• PWR rod bundles
• PWR MOX assembly
• Group constants
• Benchmark data
• Transient analyses