Influence of the transient conditions on release of corrosion products and oxidation of alloy 690 tubes during pressurized water reactor restart after steam generators replacement

Julie Flambard (1) , Florence Carrette (1) , Carole Monchy-Leroy (1) , Eric Andrieu (2) , Lydia Laffont (2)

1 EDF R&D - EDF R&D
2 CIRIMAT - Centre interuniversitaire de recherche et d'ingenierie des matériaux

The radioactivity of the Reactor Coolant System (RCS) of Pressurized Water Reactors (PWRs) mainly comes from the release of corrosion products of Steam Generator (SG) tubes made of Ni-base alloys. In order to reduce this activity and thus the radiation exposure of PWR operators during maintenance operations, it is necessary to minimize the release. That requires prior understanding of the various mechanisms involved. EDF R&D constructed a loop, BOREAL, to specifically measure rates of release of SG tubes in various conditions of primary environment. Tubes were usually tested at high temperature, under con- stant conditions of primary chemistry. So, it is necessary to carefully investigate the impact of transient conditions during a PWR restart after SG replacement. Tests were performed on the industrial material with curvature, roughness, defects and heterogeneities, regularly observed on this type of component. Characterisations of the inner surface were done on as-received and corroded specimens of SG tubes and were correlated with the obtained release kinetics. The native oxide layer is formed of a very thin layer (1-2 nm) of oxidised matrix, without specific en- richment. During the restart, the most critical step for the release phenomenon is revealed from 170 °C to 297 °C. The majority of the metal is indeed released into the fluid during this step. The characterisa- tions after release tests have made it possible to propose oxidation and release mechanism during a PWR restart after SG replacement. Up to 170 °C, a thin layer of amorphous chromium oxide is formed by selec- tive dissolution of iron and nickel. When the temperature rises, this chromium oxide layer is not stable enough to be protective and the diffusion phenomena are activated. At 325 °C, the oxide does not exhibit any particular enrichment and corresponds to an oxidised metal layer; an equilibrium is established and the rate of release reaches a pseudo-stationary regime.

Mots clés: Alloy 690- Steam generator tube; Primary circuit (PWR); Release; Oxidation

Domaines: Matériaux

Julie Flambard, Florence Carrette, Carole Monchy-Leroy, Eric Andrieu, Lydia Laffont. Influence of the transient conditions on release of corrosion products and oxidation of alloy 690 tubes during pressurized water reactor restart after steam generators replacement. Journal of Nuclear Materials, 2021, 543, pp.152562. ⟨10.1016/j.jnucmat.2020.152562⟩. ⟨hal-03033098⟩