Abstract
The effect of irradiation temperature and alloying elements on defect clustering behaviour directly from the cascade collapse in Zircaloy-2 is examined. The in-situ ioWn irradiation technique was employed to study the formation of <a>-type dislocation loops by Kr ion irradiation at 573 K and 773 K, while the dependence of dislocation loop formationon the presence of alloying elements was investigated by comparing with the defect microstructures of pure Zr irradiated under similar conditions. The experimentally observed temperature dependence of defect clustering was further investigated using molecular dynamics (MD) simulations near the experimental irradiation temperatures. We particularly concentrate on yield and morphology of small defect clusters formed directly from cascade collapse at very low ion doses. Smaller loop size and higher defect yield (DY) in Zircaloy-2 as compared to pure Zr suggests that the presence of the major alloying element Sn increases the number of nucleation sites for the defect clusters but suppresses the point defect recombination. MD simulations at 600 and 800 K revealed that the production of both vacancy and interstitial clusters drops significantly with an increase of irradiation temperature, which is reflected in experimentally collected DY data.
Highlights
Owing to their low neutron absorption, good mechanical strength, and corrosion resistance, Zr alloys are widely used in nuclear power reactors
It has been reported that the point-defect clusters formed in Zircaloy-2 at temperatures between 250 and 400 ◦ C and for irradiation doses lower than 5 × 1025 nm−2 (E > 1 MeV), which can be observed by using transmission electron microscopy (TEM) (>2 nm), and consist of perfect dislocation loops, either of vacancy or interstitial nature, with Burgers vector of 13 1120, lying in the prismatic planes (1010)
TEM (Company, FEI, Hillsboro, OR, USA) analysis, and scanning transmission electron microscopy associated with energy-dispersive X-ray spectroscopy (STEM-EDS) (FEI, Hillsboro, OR, USA) mapping done on an FEI Technai Osiris (FEI, Hillsboro, OR, USA) are presented in Figure 1, which shows the distribution of alloying elements in the secondary phase precipitates present in Zircaloy-2 used in the irradiation experiments
Summary
Owing to their low neutron absorption, good mechanical strength, and corrosion resistance, Zr alloys are widely used in nuclear power reactors. The fundamental mechanism of these degradations is neutron induced atomic displacement in alloys. In Zr alloys, prismatic (-type) dislocation loops are the basic and most well-known component of radiation damage structures and are observed in a wide range of irradiation temperatures (80–500 ◦ C) [2]. These loops are visible at lower to higher neutron doses, with densities saturating at relatively low doses, and have been extensively studied by neutron, charged particle irradiation, and MD simulations [2,3,4,5,6,7,8]. It has been reported that the point-defect clusters formed in Zircaloy-2 at temperatures between 250 and 400 ◦ C and for irradiation doses lower than 5 × 1025 nm−2 (E > 1 MeV), which can be observed by using transmission electron microscopy (TEM) (>2 nm), and consist of perfect dislocation loops, either of vacancy or interstitial nature, with Burgers vector of 13 1120 , lying in the prismatic planes (1010)
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