Effects of hydride morphology on the embrittlement of Zircaloy-4 cladding

Ju Seong Kim, Tae Hoon Kim, Dong Hak Kook, Yong Soo Kim

Research output: Contribution to journalArticle

26 Scopus citations

Abstract

Spent nuclear fuel claddings discharged from water reactors contain hydrogen up to 800 wppm depending on the burn-up and power history. During long-term dry storage, the cladding temperature slowly decreases with diminishing decay heat and absorbed hydrogen atoms are precipitated in Zr-matrix according to the terminal solid solubility of hydrogen. Under these conditions, hydrides can significantly reduce cladding ductility and impact resistance, especially when the radial hydrides are massively present in the material. In this study, the effects of hydride morphology on the embrittlement of Zircaloy-4 cladding were investigated using a ring compression test. The results show that circumferentially hydrided Zircaloy-4 cladding is brittle at room temperature but its ductility is regained substantially as the temperature goes above 150 °C. On the other hand, radially hydrided cladding remains brittle at 150°C and micro-cracks developed in the radial hydrides can act as crack propagation paths. Fracture energy analysis shows that ductile to brittle transition temperature is low in between 25°C and 100°C in the former case, whereas it lies in between 200°C and 250°C in the latter case.

Original languageEnglish
Pages (from-to)235-245
Number of pages11
JournalJournal of Nuclear Materials
Volume456
DOIs
StatePublished - 2015 Jan

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