Zirconium alloy is frequently used in the nuclear industry as a reactor fuel cladding material. Uniform corrosion of zirconium alloys has received much attention because it is one of the material's life-limited properties. To study their long-term uniform corrosion behavior, two types of Zr-Sn-Nb-Fe-V alloy claddings, one with a high niobium content and one with a low niobium content, were exposed to 400oC and 10.3 MPa of superheated steam for 800 d. Both alloys clearly exhibit periodic oxide densification-transition. Different quantitative methods were used to study the evolution and mechanism of multiple oxide features. The periodic layers of oxide morphologies are still observed even after two times of transitions. Periodically, columnar grains and the defect layer appear. The undulation intensity (related to the “cauliflower” morphologies) and the equivalent thickness of tetragonal zirconia (t-ZrO2) at the metal/oxide interface increase with corrosion time in pretransition oxides and decrease at transition time. The evolution of both features in the subsequent oxide densification-transition period is the same as initial densificationtransition cycle. The critical value of the interface undulation intensity is approximately 0.25 for both alloys. The critical thickness of t-ZrO2 is approximately 450 nm for low-niobium alloy and approximately 200 nm for high-niobium alloy. Both features periodically reach critical conditions. The evolution of undulation intensity provides an excellent explanation for the production of isolated and interconnected lateral cracks. Additionally, the transformation of t-ZrO2 to monoclinic zirconia (m-ZrO2) results in cracking of the oxide and produces interconnected, tiny equiaxed defects at the interface. Both lateral cracks and equiaxed defects are both important components of the defect layer. It is hypothesized that the synergetic effects of interface undulation and t-ZrO2 transformation affect the transition. The occurrence of periodic transitions is strongly correlated with the periodic behavior of oxide features reaching critical conditions. The volume of oxygen vacancy in t-ZrO2 was presumably evaluated by studying the Raman shift of the characteristic t-ZrO2 peak of 280 cm-1. As a result, the amount does not change considerably during oxide densification but decreases during the transition period. The Raman shift of t-ZrO2 in low-niobium alloys is approximately -1.2 cm-1 less than that in high-niobium alloys, indicating that the low-niobium alloys have a greater volume of oxygen vacancies. It is proposed that the difference in the corrosion behavior of two alloys is derived from the difference in the volume of oxygen vacancies.
Fig.1 Mass gain curves of Zr-0.5Sn-0.15Nb-0.5Fe-0.25V (N2) and Zr-0.2Sn-1.3Nb-0.1Fe-0.05V (N3) alloys corroded in 400oC and 10.3 MPa steam
Alloy
tdx d
Mass gain at transition point mg·dm-2
Kinetics law in the pre-transition
Quasi-linear relationship in the post-transition
N2
130
45.04
6.76t0.42
0.44t - 0.26
N3
80
51.75
7.19t0.46
0.60t + 16.30
Table 1 Fitted results of corrosion kinetics law
Fig.2 Cross-section SEM images of N2 alloy corroded for 40 d (T2 = 0.31) (a), 80 d (T2 = 0.62) (b), 120 d (T2 = 0.92) (c), 160 d (T2 = 1.23) (d), and 200 d (T2 = 1.54) (e) (T2 is the corrosion state parameter of N2 alloy. T2 = 0.92 indicates that corrosion is in pre-transition stage, and T2 = 1.54 indicates that corrosion is in re-densification stage after primary transition)
Fig.3 Cross-section SEM images of N3 alloy corroded for 20 d (T3 = 0.25) (a), 60 d (T3 = 0.75) (b), 100 d (T3 = 1.25) (c), and 200 d (T3 = 2.50) (d) (T3 is the corrosion state parameter of N3 alloy. T3 = 1.25 indicates that corrosion is in re-densification stage after primary transition, and T3 = 2.50 indicates that corrosion is in re-densification stage after secondary transition)
Fig.4 TEM images of oxide foil of N3 alloy corroded for 200 d (a) and locally magnified images of areas 1 (b), 2 (c), 3 (d), 4 (e), 5 (f), and 6 (g) in Fig.4a (O/M—oxide film/metal)
Fig.5 SEM images (a-d) and CLSM 3D images (e-h) of inner-face of N2 alloy corroded for 20 d (T2 = 0.15) (a, e), 60 d (T2 = 0.46) (b, f), 120 d (T2 = 0.92) (c, g), and 140 d (T2 = 1.08) (d, h)
Fig.6 Evolutions of undulation intensity (Sdq) versus corrosion time of the O/M interface of N2 and N3 alloys
Fig.7 Raman spectra of inner face of N2 (a) and N3 (b) alloys
Fig.8 Evolutions of equivalent tetragonal thickness versus corrosion stage parameter (a) and t-ZrO2 peaks versus corrosion time (b)
1
Motta A T, Couet A, Comstock R J. Corrosion of zirconium alloys used for nuclear fuel cladding[J]. Annu. Rev. Mater. Res., 2015, 45: 311
doi: 10.1146/annurev-matsci-070214-020951
2
Hillner E, Franklin D G, Smee J D. Long-term corrosion of Zircaloy before and after irradiation[J]. J. Nucl. Mater., 2000, 278: 334
doi: 10.1016/S0022-3115(99)00230-5
3
Yang Z B, Zhao W J, Cheng Z Q, et al. Effect of Nb content on the corrosion resistance of Zr-xNb-0.4Sn-0.3Fe alloys[J]. Acta Metall. Sin., 2017, 53: 47
Likhanskii V V, Evdokimov I A. Effect of additives on the susceptibility of zirconium alloys to nodular corrosion[J]. J. Nucl. Mater., 2009, 392: 447
doi: 10.1016/j.jnucmat.2009.04.003
6
Zhang H X, Li Z K, Zhou L, et al. Effects of structure and internal stresses in oxide films on corrosion mechanism of new zirconium alloy[J]. Acta Metall. Sin., 2014, 50: 1529
doi: 10.11900/0412.1961.2014.00261
Polatidis E, Frankel P, Wei J, et al. Residual stresses and tetragonal phase fraction characterisation of corrosion tested Zircaloy-4 using energy dispersive synchrotron X-ray diffraction[J]. J. Nucl. Mater., 2013, 432: 102
doi: 10.1016/j.jnucmat.2012.07.025
8
Preuss M, Frankel P, Lozano-Perez S, et al. Studies regarding corrosion mechanisms in zirconium alloys[J]. J. ASTM Int., 2011, 8: 103246
doi: 10.1520/JAI103246
9
Ni N, Lozano-Perez S, Sykes J M, et al. Focussed ion beam sectioning for the 3D characterisation of cracking in oxide scales formed on commercial ZIRLO™ alloys during corrosion in high temperature pressurised water[J]. Corros. Sci., 2011, 53: 4073
doi: 10.1016/j.corsci.2011.08.013
10
Platt P, Wedge S, Frankel P, et al. A study into the impact of interface roughness development on mechanical degradation of oxides formed on zirconium alloys[J]. J. Nucl. Mater., 2015, 459: 166
doi: 10.1016/j.jnucmat.2015.01.028
11
Liao J J, Yang Z B, Qiu S Y, et al. Corrosion of new zirconium claddings in 500oC/10.3 MPa steam: Effects of alloying and metallography[J]. Acta Metall. Sin. (Engl. Lett.), 2019, 32: 981
doi: 10.1007/s40195-018-0857-7
12
Liao J J, Yang Z B, Qiu S Y, et al. The correlation between tetragonal phase and the undulated metal/oxide interface in the oxide films of zirconium alloys[J]. J. Nucl. Mater., 2019, 524: 101
doi: 10.1016/j.jnucmat.2019.06.039
13
Qu J W, Tian H, Shi M H, et al. Effect of V addition on the mechanical properties and corrosion resistance of high temperature water vapor of Zr-1Nb-0.1Fe alloy[J]. Nonferrous Met. Eng., 2020, 10(1): 15
Liao J J, Qiu S Y, Zhang J S, et al. Research on laterally cracking, vertically cracking and transition mechanism in oxide film of zirconium alloy[J]. Nucl. Power Eng., 2020, 41(): 164
Yardley S S, Moore K L, Ni N, et al. An investigation of the oxidation behaviour of zirconium alloys using isotopic tracers and high resolution SIMS[J]. J. Nucl. Mater., 2013, 443: 436
doi: 10.1016/j.jnucmat.2013.07.053
16
Platt P, Frankel P, Gass M, et al. Critical assessment of finite element analysis applied to metal-eoxide interface roughness in oxidising zirconium alloys[J]. J. Nucl. Mater., 2015, 464: 313
doi: 10.1016/j.jnucmat.2015.05.002
17
Zhang J S, Lyu J N, Long C S, et al. Calculation of internal stress in oxide films of zirconium alloy[J]. Nucl. Power Eng., 2021, 42(4): 101
Yao M Y, Zhang X W, Hou K K, et al. The initial corrosion behavior of Zr-0.75Sn-0.35Fe-0.15Cr alloy in deionized water at 250oC[J]. Acta Metall. Sin., 2020, 56: 221
Guo X. Property degradation of tetragonal zirconia induced by low-temperature defect reaction with water molecules[J]. Chem. Mater., 2004, 16: 3988
doi: 10.1021/cm040167h
20
Liao J J, Xu F, Peng Q, et al. Research on the existence and stability of interfacial tetragonal zirconia formed on zirconium alloys[J]. J. Nucl. Mater., 2019, 528: 151846
doi: 10.1016/j.jnucmat.2019.151846
21
Qin W, Nam C, Li H L, et al. Tetragonal phase stability in ZrO2 film formed on zirconium alloys and its effects on corrosion resistance[J]. Acta Mater., 2007, 55: 1695
doi: 10.1016/j.actamat.2006.10.030
22
Bouvier P, Godlewski J, Lucazeau G. A Raman study of the nanocrystallite size effect on the pressure-temperature phase diagram of zirconia grown by zirconium-based alloys oxidation[J]. J. Nucl. Mater., 2002, 300: 118
doi: 10.1016/S0022-3115(01)00756-5
23
Liao J J, Zhang J S, Zhang W, et al. Critical behavior of interfacial t-ZrO2 and other oxide features of zirconium alloy reaching critical transition condition[J]. J. Nucl. Mater., 2021, 543: 152474
doi: 10.1016/j.jnucmat.2020.152474
24
Sundell G, Thuvander M, Andrén H O. Barrier oxide chemistry and hydrogen pick-up mechanisms in zirconium alloys[J]. Corros. Sci., 2016, 102: 490
doi: 10.1016/j.corsci.2015.11.002
25
Wang Z, Zhou B X, Wang B Y, et al. Second phase particles and their corrosion behavior of Zr-0.72Sn-0.32Fe-0.15Cr-0.97Nb alloy[J]. Acta Metall. Sin., 2016, 52: 78
doi: 10.11900/0412.1961.2015.00260
