Strain localization in copper canister FSW welds for spent nuclear fuel disposal

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A1 Alkuperäisartikkeli tieteellisessä aikakauslehdessä
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Journal of Nuclear Materials, Volume 523
Spent nuclear fuel disposal in copper canisters in a deep geologic repository is planned in Finland and Sweden. The purpose of the copper shell is to perform as a ductile corrosion barrier to prevent radioactive substances from leaking into the environment. Therefore, the most important property of the copper shell, besides the good corrosion resistance, is its ductility. The copper canisters are sealed by friction stir welding (FSW), which results in strong welds when compared to the base materials, but the microstructural heterogeneity introduced by the welding may also lead to strain localization. Thus, the strain localization behavior of two different copper canister welds was studied by tensile tests in combination with digital image correlation (DIC). The main difference between the welds is the utilization of shielding gas to reduce oxidation during welding. The shielding gas improves the stability of the welding process, as well as reduces the number of oxide particles in the welds. It is known, that oxide particles are detrimental in copper in the presence of hydrogen. Therefore, the two welds were also thermally hydrogen charged to study hydrogen trapping in the weld material. Thermal desorption measurements (TDS) show that considerable hydrogen uptake occurs in the weld oxide zone, but it did not compromise the ductility of the copper welds in these tests. However, the DIC tests indicate considerably earlier strain localization on the retreating side of the new weld, welded with the shielding gas. This is attributed to differences in the initial state of the lid materials.
Copper, Digital image correlation, Friction stir welding, Hydrogen uptake, Strain localization
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Forsström, A, Bossuyt, S, Yagodzinskyy, Y, Tsuzaki, K & Hänninen, H 2019, ' Strain localization in copper canister FSW welds for spent nuclear fuel disposal ', Journal of Nuclear Materials, vol. 523, pp. 347-359 .