Mechanical behavior at high temperature of highly oxygen- or hydrogen-enriched α and (prior-) $\beta$ phases of zirconium alloys

I. Turque (1) , R. Chosson (1) , Matthieu Le Saux (2) , J.-C. Brachet (1) , V. Vandenberghe (1) , J. Crepin (3) , Af, Gourgues-Lorenzon (3)

1 CEA-DES (ex-DEN) - CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire)
2 SRMA - Service des Recherches Métallurgiques Appliquées
3 Mines Paris - PSL (École nationale supérieure des mines de Paris)

During a hypothetical loss-of-coolant accident (LOCA), zirconium alloy fuel claddings can be loaded by internal pressure and exposed to steam at high temperature (HT, potentially up to 1200°C) until they are cooled and water quenched. A significant fraction of the oxygen reacting with the cladding during oxidation at HT diffuses beneath the oxide through the metallic substrate. This diffusion of oxygen induces a progressive transformation of the metallic $\beta _{Zr}$ phase layer into an intermediate layer of $\alpha _{Zr}$(O) phase containing up to 7 wt.% of oxygen. Furthermore, in some specific conditions, the cladding may rapidly absorb a significant amount of hydrogen during steam exposition at high temperature. Then, hydrogen, as a $\beta _{Zr}$-stabilizer, would mainly diffuse and concentrate up to several thousands of wt.ppm into the inner$\beta _{Zr}$ phase layer

Mots clés: Zircaloy-4; M5; oxygen; hydrogen; LOCA; mechanical behavior; high temperature

Domaines: Physique Nucléaire Expérimentale [nucl-ex]; Physique Nucléaire Théorique [nucl-th]

I. Turque, R. Chosson, Matthieu Le Saux, J.-C. Brachet, V. Vandenberghe, et al.. Mechanical behavior at high temperature of highly oxygen- or hydrogen-enriched α and (prior-) $\beta$ phases of zirconium alloys. 18th International Symposium on Zirconium in the Nuclear Industry, May 2016, Hilton Head, United States. ⟨cea-02442337⟩