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TEM Examination of a swelling-resistant Austenitic steel irradiated at high temperature (>600Dc) in the Phenix Fast Reactor

Abstract : In the framework of the Sodium Fast-Reactor (SFR) project ASTRID, an optimized austenitic steel 15%Cr-15%Ni, stabilized with Ti, was chosen for the fuel-pin cladding. Besides irradiation swelling, one of the main issue of irradiation is the embrittlement at high temperature (>600°C). To investigate the long-term behavior of this alloy irradiated at temperatures greater than 600°C, fuel pins cladded with optimized Si-enriched 15-15Ti, were irradiated in several experimental assemblies of the Phenix reactor to reach a cumulative irradiation time of 941 days (EFPD Equivalent Full Power Day). Thin foils suitable for Transmission Electron Microscope (TEM) were taken from the upper part of the fuel pin where the irradiation temperature is maximum and exceeds 600°C. The final dose of the irradiated thin foil is about 40 dpa NRT. TEM observations were carried out to characterize irradiation defects, dislocation network and precipitations in order to identify the possible embrittlement mechanisms. TEM examinations show significant dislocation recovery but no sign of recrystallization was detected. Ti-rich MC nanoprecipitates, Faulted Frank loops, nanometric bubbles and possibly SFT were identified. An abundant coarse precipitation is observed inside grains and in grain boundaries.
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  • HAL Id : cea-02434539, version 1

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A. Courcelle, M. Jublot, E. Piozin, P. Gavoille, C. Bisor, et al.. TEM Examination of a swelling-resistant Austenitic steel irradiated at high temperature (>600Dc) in the Phenix Fast Reactor. FR17 International Conference on Fast Reactors and Related Fuel Cycles Next Generation Nuclear Systems for Sustainable Development, Jun 2017, Yekaterinburg, Russia. ⟨cea-02434539⟩

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