PSI - Issue 80

Vaclav Sklenicka et al. / Procedia Structural Integrity 80 (2026) 493–500 Author name / Structural Integrity Procedia 00 (2019) 000 – 000

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to preventing fission products from being released from the fuel to the coolant in water-cooled reactors. Structural materials for nuclear fuel claddings should have low neutron absorption, good-in-service hydrothermal corrosion and oxidation resistance, excellent high-temperature mechanical properties, and microstructural and dimensional stability. Zirconium alloys have been nuclear fuel cladding in light water reactors (LWRs) for several decades. There have been reviews of zirconium cladding alloys, notably Murty and Charit (2008), Adamson and Rudling (2013), Yagnik and Garde (2019), and others. Zircaloy-4 (Zry-4), containing Zr-Sn-Fe-Cr, enjoys widespread use for Western types of light water and heavy water reactors (Garzarolli et al. (1996)). Zircaloy-4 is a reliable alloy that fulfills all requirements for regular operation and expectant accidental situations. The optimization of Zircaloy-4 during recent decades includes the optimization of microstructure and its chemical composition. Although still in use in some water reactors, Zircaloy-4 is being gradually replaced by the advanced zirconium cladding alloys containing niobium in addition to tin and increased oxygen and iron levels, e.g. ZIRLO (Pan et al. 2015)). Nevertheless, without regard to the individual chemical compositions of zirconium cladding alloys, they all undergo various and similar degradation processes that limit the lives of claddings under reactor operation and/or interim dry storage conditions. In addition to thermal aging and instability of microstructure, irradiation, and corrosion/oxidation, the limiting degradation process can be the time-dependent plastic deformation – thermal creep (Čadek 1988, Kassner 2009, Sandström 2024, Murty et al. 2013, Hong et al. 2018, Sklenicka et al. 2023), which deforms the cladding tubes and thus can degrade their integrity at elevated temperatures leading to disastrous conditions as the LOCA (loss-of-coolant) accident (Erbacher and Leistikow 1987, Suman et al. 2016, Thiermel et al. 2019). Consequently, limited creep test data and models were suggested to predict the creep behavior of Zircaloy-4 alloy during thermal creep under various stress and temperature conditions (Sagar et al. 2024). However, due to the small number of such works published in the open literature, the rate-controlling creep and damage deformation mechanisms remain insufficiently understood. This was the reason for the present study: to provide further relevant information on thermal creep behavior and damage processes in Zircaloy-4 alloy. 2. Experimental 2.1. Zircaloy-4 cladding tube The test material used in this work was commercial-grade Zircaloy-4 alloy (Zry-4). Zircaloy-4 alloy was received as a real thin-walled nuclear cladding tube produced by Sandvick (Sweden) with an outer diameter of 10.72 mm and a wall thickness of 0.725 mm. The chemical composition of the tested alloy was as follows: 1.46 wt.% Sn, 0.21 wt.% Fe, 0.09 wt.% Cr and 1090 wppm O, 10 wppm H, 30 wppm N, 120 wppm C, 70 wppm Hf, < 0.3 wppm Cd, < 0.4 wppm B, balance Zr. 2.2. Creep testing The tubular specimens for creep testing were directly cut from a cladding tube so the tensile axis was parallel with the rolling direction. Each creep specimen had a gauge length of 50 mm, which was delimited using two end plugs pasted onto the tubular specimen. Uniaxial constant-stress creep tests in tension were carried out in a purified and dry inert argon atmosphere using the creep testing machines that permitted the applied tensile stress to be kept constant up to the strain of ε ~ 40%. The testing temperature was stabilized within ± 1 °C and kept constant along the gauge length of the creep specimen. The creep specimen was loaded after soaking for 1h at the testing temperature. The creep elongation of the tested specimen was continuously measured in situ during a creep exposure with a sensitivity of 5 x 10 -6 using a linear variable displacement transducer extensometer W2K from Hottinger-Baldwin Co. (Germany). The creep elongation was digitally recorded and processed with a computer. 2.3. Microstructural and fractographic investigations Most metallographic and fractographic investigations were performed using scanning electron microscopy (SEM TESCAN LYRA 3, Tescan s.r.o., Brno, Czech Republic) equipped with a NordlysNano detector operating at an accelerating voltage of 20 kV with the specimen tilted at 70°. The EBSD data were analyzed using HKL Channel 5