Impact of Irradiation Damage Recovery During Transportation on the Subsequent Room Temperature Tensile Behavior of Irradiated Zirconium Alloys

Abstract

Zirconium alloys are commonly used in Pressurized Water Reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the Spent Nuclear Fuel Assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on Nuclear Power Plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1%Nb and stress relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400°C and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, Transmission Electron Microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the asirradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The Transmission Electron Microscopy examinations, especially conducted on recrystallized Zr-1%Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that, as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channel was indeed observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may occur during dry transportation or at the beginning of dry storage.

Mechanical Properties of Zircaloy-2 Neutron Irradiated to High Fluence

“…Dislocation channels had been observed in irradiated zirconium alloys in the 1960s-1980s [1][2][3][4][5], but understanding of the deformation channelling mechanism developed many years later [6][7][8][9][10][11][12][13]. Deformation of unirradiated zirconium alloys for low plastic strains was predominantly by prismatic slip of a-dislocations, because the critical resolved shear stress (CRSS) of the prismatic slip system was lower than that for the other slip systems [14].…”

“…On the contrary, the junction between a-loops and a-dislocations gliding on the prismatic plane are sessile in 2 out of 3 a-type loops, thus leading to a lower ability and a higher CRSS for prismatic channelling. Dislocation channelling is dependent on factors such as mechanical testing direction relative to texture, testing temperature, irradiation temperature, and fluence [8,9,13]. For example, a change in channel plane has also been attributed to the oxygen content in Zircaloy [12].…”

“…For example, a change in channel plane has also been attributed to the oxygen content in Zircaloy [12]. To explain the occurrence of basal channels in Zr-1Nb for deformation at 350°C, but both prism and basal channels at 20°C, Bourdiliau et al [13] proposed the schematic shown Figure 16, where the CRSS for basal and prism slips would be equally probable at 20°C, but basal slip would have a lower CRSS than prismatic slip.…”

“…Plastic deformation of neutron-irradiated zirconium alloys occurs in a highly inhomogeneous manner, forming bands or channels confined to localized regions [1][2][3][4][5][6][7][8][9][10][11][12][13]. This strain localization is produced by the annihilation of irradiation defects into channels that are dislocation free and can thus accommodate further deformation.…”

“…FIGURE 16. Schematic of the normalized prismatic and basal CRSS as a function of temperature before and after irradiation [13]. For personal use only.…”