Mechanical behavior at high temperatures of highly oxygen- or hydrogen-enriched α and (Prior-) β phases of zirconium alloys

Abstract : Mechanical behavior at high temperature of highly oxygen-or hydrogen-enriched α and (prior-) β phases of zirconium alloys ABSTRACT: During a hypothetical loss-of-coolant accident (LOCA), zirconium alloy fuel claddings can be loaded by internal pressure and exposed to steam at high temperature (HT, potentially up to 1200°C), then cooled and water quenched. A significant fraction of the oxygen reacting with the cladding during HT oxidation diffuses beneath the oxide through the metallic substrate. This induces a progressive transformation of the metallic βZr phase layer into an intermediate layer of αZr(O) phase containing up to 7 wt.% of oxygen. Furthermore, in some specific conditions, the cladding may rapidly absorb a significant amount of hydrogen during steam exposition at HT. Being a βZr-stabilizer, hydrogen would mainly diffuse and concentrate up to several thousands of wt.ppm into the inner βZr phase layer. Oxygen and hydrogen are known to modify the metallurgical and mechanical properties of zirconium alloys but data are scarce for high contents, especially at HT. However, such data are important basic components to improve the assessment of the oxidized cladding mechanical behavior and integrity during and after LOCA-like thermal-mechanical transients. This study intended to provide new, more comprehensive data on the HT mechanical behavior of the αZr(O) and the (prior-) βZr phases containing high contents of oxygen and hydrogen, respectively. Model samples, produced from M5® 5 and Zircaloy-4 cladding tubes, homogeneously charged in oxygen (≤6 wt.%) and in hydrogen (≤3000 wt.ppm) respectively, were prepared. Their mechanical behavior was determined under vacuum between 800 and 1100°C for the oxygen-enriched αZr phase, and in air between 700 and 20°C, after cooling from the βZr temperature domain, for the hydrogen-enriched (prior-) βZr phase. The αZr phase is substantially strengthened and embrittled by oxygen. Power-law and nearly linear creep regimes are observed and were modelled for stress levels beyond and below 15 MPa, respectively. The model αZr(O) material experiences a ductile-to-brittle transition at 1000-1100°C for oxygen contents between 3.4 and 4.3 wt.%. The viscoplastic behavior of the αZr(O) phase was used to evaluate the contribution of the αZr(O) layer to the HT creep behavior of an oxidized fuel cladding tube subjected to internal pressure. The model (prior-) βZr phase becomes macroscopically brittle at temperatures ≤135°C and ≤350-400°C for average hydrogen contents
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R.J. Comstock, A.T. Motta. 18th International symposium on zirconium in the nuclear industry, May 2016, Hilton Head Island, SC, United States. ASTM, pp.240-280, Zirconium in the nuclear industry. 〈https://www.astm.org/DIGITAL_LIBRARY/STP/PAGES/STP159720160063.htm〉. 〈10.1520/STP159720160063〉
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Isabelle Turque, Raphaël Chosson, Matthieu Le Saux, Jean-Christophe Brachet, Valérie Vandenberghe, et al.. Mechanical behavior at high temperatures of highly oxygen- or hydrogen-enriched α and (Prior-) β phases of zirconium alloys. R.J. Comstock, A.T. Motta. 18th International symposium on zirconium in the nuclear industry, May 2016, Hilton Head Island, SC, United States. ASTM, pp.240-280, Zirconium in the nuclear industry. 〈https://www.astm.org/DIGITAL_LIBRARY/STP/PAGES/STP159720160063.htm〉. 〈10.1520/STP159720160063〉. 〈hal-01702107〉

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