Effect of isothermal annealing on the corrosion behavior of Zr–xNb alloys

Kim, Hyun Gil; Jeong, Yong Hawn; Kim, To Hoon

doi:10.1016/j.jnucmat.2004.01.015

Cited by 66 publications

(31 citation statements)

References 6 publications

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“…5a. It can be seen that the higher the Nb content, the higher the oxidation rate in pre-transition (which is in agreement with the literature data) [16,[24][25][26]. The post-transition stage was not reached, in the experiment, with Zr0.4%NbO alloy.…”

Section: Resultssupporting

confidence: 89%

Oxidation kinetics of ZrNbO in steam: Differences between the pre- and post-transition stages

Tupin

Pijolat

Valdivieso

et al. 2005

Journal of Nuclear Materials

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Zirconium based alloys (and particularly Zircaloy-4) are widely used as cladding material of fuel rods in water-cooled nuclear reactors. However, advanced alloys are now used for longer operation time in more severe conditions. One of these new alloys is a ZrNbO, which contains 1% of Nb as alloying element (instead of Sn in the case of Zircaloy-4).This ZrNbO alloy, like many zirconium alloys, undergoes a kinetic transition, which is a sharp increase in the oxidation rate when the oxide thickness exceeds a critical value. In the pre-transition stage, the kinetic oxidation curves are approximately parabolic (which is generally interpreted by a diffusion controlling step), and tend towards a quasi-linear behaviour after the transition. Recent studies have shown that the oxygen pressure has an accelerating effect on the oxidation rate of ZrNbO; a platinum layer deposited on the oxide surface has the same effect, both in oxygen and in water vapour. Thus it has been suggested that the oxidation in the pre-transition stage could be described by a mixed reaction-diffusion kinetics. Due to the lack of consistent data on the effect of water vapour on the oxidation rate of ZrNbO before and after the transition, and to confirm (or not) these interpretations, we have studied the oxidation of ZrNbO by isothermal thermogravimetry at 550°C, under controlled partial pressures of water vapour (13 to 80 hPa) and hydrogen (10 hPa). The aim is to verify if the oxidation proceeds in a steady state and if there is a rate-limiting step in one (or both) stages, and to clearly evidence the differences between the pre-and post transition stages from the kinetic point of view.TG-DSC experiments have shown that the system is in a steady state from the beginning of the oxidation, in the pre-and post-transition stages. Then, the existence of a ratelimiting step was verified using an experimental method based on temperature or pressure jumps: it has been concluded that this assumption is valid in both stages, which is not compatible with the assumption of a mixed reaction-diffusion controlled rate). Moreover, it comes out that the kinetic behaviour is different before and after the transition: the influence of temperature jumps is greater in pre-transition, whereas the effect of water vapour partial pressure is more pronounced in post-transition (nevertheless, an accelerating effect is also observed before the transition). No influence of hydrogen partial pressure has been observed. Besides, the higher the Nb content in the alloy, the higher the oxidation rate (in pre-transition). A mechanism has been proposed to account for the results obtained in pretransition, involving the diffusion of adsorbed species in the porous part of the oxide layer as rate-determining step.Keywords: ZrNbO ; oxidation ; kinetics ; water vapour ; steady state ; rate-limiting step. 1 IntroductionDespite numerous works dedicated to the oxidation kinetics of zirconium alloys by pressurized water, water vapour or oxygen, the mechanisms and the rate-limiting steps...

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Section: Resultssupporting

confidence: 89%

Oxidation kinetics of ZrNbO in steam: Differences between the pre- and post-transition stages

Tupin

Pijolat

Valdivieso

et al. 2005

Journal of Nuclear Materials

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“…Generally, a finer size and more homogeneous distribution of precipitates benefits the corrosion resistance in Zircaloys [8,9]. In Zr-Nb alloys, the super-saturation of Nb in the matrix was demonstrated to degrade corrosion resistance; thus a full precipitation of β-Nb either from α-Zr or β-Zr is recommended [10][11][12]. It is reported that accelerated irradiation growth is related to the appearance of <c> component vacancy dislocation loops on the basal plane and Fe concentration in the matrix [13,14].…”

Section: Introductionmentioning

confidence: 99%

Precipitate Stability in a Zr–2.5Nb–0.5Cu Alloy under Heavy Ion Irradiation

Dong

Yao

Wang

et al. 2017

Metals

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Abstract:The stability of precipitates in Zr-2.5Nb-0.5Cu alloy under heavy ion irradiation from 100 • C to 500 • C was investigated by quantitative Chemi-STEM EDS analysis. Irradiation results in the crystalline to amorphous transformation of Zr 2 Cu between 200 • C and 300 • C, but the β-Nb remains crystalline at all temperatures. The precipitates are found to be more stable in starting structures with multiple boundaries than in coarse grain structures. There is an apparent increase of the precipitate size and a redistribution of the alloying element in certain starting microstructures, while a similar size change or alloying element redistribution is not detected or only detected at a much higher temperature in other starting microstructures after irradiation.

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“…Such alloys with high performance were the main focus in Zr-Sn and Zr-Nb series [4][5][6][7][8][9][10], e.g., Zr-1.0Sn-1.0Nb-0.1Fe (ZIRLO alloy, the number before each element represents the weight percent, wt %), Zr-1.0Nb-O (M5), Zr-1.2Sn-1.0Nb-0.35Fe (E635), and Zr-1.0Sn-1.0Nb-0.25Fe (N36). They were generally minor-alloyed with multiple solute elements to achieve the optimum comprehensive properties.…”

Section: Introductionmentioning

confidence: 99%

New Low-Sn Zr Cladding Alloys with Excellent Autoclave Corrosion Resistance and High Strength

Zhang

Jiang

Pang

et al. 2017

Metals

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Abstract:It is expected that low-Sn Zr alloys are a good candidate to improve the corrosion resistance of Zr cladding alloys in nuclear reactors, presenting excellent corrosion resistance and high strength. The present work developed a new alloy series of Zr-0.25Sn-0.36Fe-0.11Cr-xNb (x = 0.4~1.2 wt %) to investigate the effect of Nb on autoclave corrosion resistance. Alloy ingots were prepared by non-consumable arc-melting, solid-solutioned, and then rolled into thin plates with a thickness of 0.7 mm. It was found that the designed low-Sn Zr alloys exhibit excellent corrosion resistances in three out of pile autoclave environments (distilled water at 633 K/18.6 MPa, 70 ppm LiOH solution at 633 K/18.6 MPa, and superheated water steam at 673 K/10.3 MPa), as demonstrated by the fact of the Zr-0.25Sn-0.36Fe-0.11Cr-0.6Nb alloy shows a corrosion weight gain ∆G = 46.3 mg/dm 2 and a tensile strength of σ UTS = 461 MPa following 100 days of exposure in water steam. The strength of the low-Sn Zr alloy with a higher Nb content (x = 1.2 wt %) is enhanced up to 499 MPa, comparable to that of the reference high-Sn N36 alloy (Zr-1.0Sn-1.0Nb-0.25Fe, wt %). Although the strength improvement is at a slight expense of corrosion resistance with the increase of Nb, the corrosion resistance of the high-Nb alloy with x = 1.2 (∆G = 90.4 mg/dm 2 for 100-day exposure in the water steam) is still better than that of N36 (∆G = 103.4 mg/dm 2 ).

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Effect of isothermal annealing on the corrosion behavior of Zr–xNb alloys

Cited by 66 publications

References 6 publications

Oxidation kinetics of ZrNbO in steam: Differences between the pre- and post-transition stages

Oxidation kinetics of ZrNbO in steam: Differences between the pre- and post-transition stages

Precipitate Stability in a Zr–2.5Nb–0.5Cu Alloy under Heavy Ion Irradiation

New Low-Sn Zr Cladding Alloys with Excellent Autoclave Corrosion Resistance and High Strength

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