Effect of Water Chemistry and Composition on Microstructural Evolution of Oxide on Zr Alloys

Zhou, Bangxin; Li, Q.; Yao, Meiyi; Liu, W. Q.; Chu, Yuliang

doi:10.1520/jai101121

Cited by 4 publications

(7 citation statements)

References 24 publications

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“…TEM and scanning electric microscope (SEM) techniques were employed to examine the microstructure and the fracture surface morphology of the oxide film covered on the Zircaloy-4 specimens corroded in lithiated water and deionized water at 633 K/18.6 MPa by Zhou et al [11,12]. Their results showed that more pores and microcracks formed along the grain boundaries in the oxide film covered on the specimens corroded in lithiated water than that corroded in deionized water.…”

Section: Resultsmentioning

confidence: 99%

“…So, the gathering of defects in the grain boundaries of the ZrO2 formed on the Zircaloy-4 specimen corroded in lithiated water could have been easier than that corroded in deionized water. Scanning probe microscopy (SPM) was employed to examine the surface morphology of the oxide covered on the specimens corroded in lithiated water in the early stage of pretransition [11]. Based Based on the discussion in the introduction, we know that the mechanism of the Li ions' effect on the surface energies of the grain boundaries of the oxide is more reasonable than others.…”

Section: Resultsmentioning

confidence: 99%

“…However, up to now, there have been no reported results showing this kind of substance. Zhou et al [11,12] held that the surface free energy of zirconium oxide could be decreased when Li ions were incorporated in the oxide. Therefore, the formation of defects in the oxide film could be enhanced, and the corrosion resistance was deteriorated.…”

Section: Introductionmentioning

confidence: 99%

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The Distribution of Li Ions in the Oxide Film Formed on Zircaloy-4 Corroded in Lithiated Water at 633 K

Xie

Zhou

et al. 2020

Materials

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Transmission electron microscopy (TEM), second ion mass spectrum (SIMS) and atom probe tomography (APT) techniques are used to study the Li ion distribution in the oxide formed on the rolling surface (SN) of Zircaloy-4 corroded in lithiated water with 0.01 M LiOH at 633 K/18.6 MPa. The results showed that the Li ions segregated in the grain boundaries and subgrain boundaries in the oxide film, but nearly no Li ions were found in the oxide around the interface between the oxide and matrix. Finally, we discussed the mechanism of the LiOH influence on the corrosion resistance of Zircaloy-4.

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Section: Resultsmentioning

confidence: 99%

Section: Resultsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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The Distribution of Li Ions in the Oxide Film Formed on Zircaloy-4 Corroded in Lithiated Water at 633 K

Xie

Zhou

et al. 2020

Materials

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“…The oxidation of alloying elements dissolved in α-Zr on the corrosion resistance of alloys has dispersive effect than that of SPPs; Moreover, Bi dissolved in α-Zr can be more likely to dissolve in ZrO 2 than the Bi in SPPs, which can reduce the decrease of surface free energy due to the infiltration of oxygen ions and hydroxide ions. Therefore, the vacancy in the oxide film cannot easily diffuse and condense to form the pore-clusters and cracks, and the corrosion resistance of zirconium alloys can be improved [21].…”

Section: Discussionmentioning

confidence: 99%

The Influence of Different Contents of Bi Addition on the Corrosion Behavior of Various Zirconium-Based Alloys

Yao¹,

Wu²,

Duan³

et al. 2016

JEP

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Zr-4(Zr-1.5Sn-0.2Fe-0.1Cr, wt%), S5(Zr-0.8Sn-0.34Nb-0.39Fe-0.1Cr), T5(Zr-0.7Sn-1.07Nb-0.32Fe-0.08Cr) and Zr-1Nb were adopted to prepare Bi-containing zirconium alloys for systematically investigating the effect of Bi addition on the corrosion resistance of zirconium alloys. The specimens were corroded in superheated steam at 400˚C/10.3 MPa, and in lithiated water with 0.01 M LiOH or in deionized water at 360˚C/18.6 MPa by autoclave testing. Results show that the corrosion resistance increases with the increasing of Bi content dissolved in α-Zr. But the presence of Bi-containing second phase particles (SPPs) is unfavorable for the enhancement of corrosion resistance. This indicates that the Bi dissolved in α-Zr matrix plays an important role in improving the corrosion resistance, while the precipitation of the Bi-containing SPPs does harm to the corrosion resistance.

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“…Experimental and computed corrosion in an autoclave according to the data in[10] (E),[18] (B),[19] (C),[20] (a).…”

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confidence: 99%

Systematic approach to predicting corrosion of zirconium alloys in the water coolant of nuclear reactors

Kritskii¹,

Berezina²

2011

At Energy

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A method of semiempirical prediction of corrosion of cladding zirconium alloys as a function of the operating conditions and composition is presented. The laws of thermodynamics and chemical kinetics of the oxidation reactions of a multicomponent zirconium alloy form the physicochemical basis of the computational method. The method is based on a model developed at the All-Russia Research and Design Institute of Integrated Power Technology for the corrosion of commercial and experimental zirconium alloys in water media under autoclave and reactor conditions taking account of the composition of the alloy and the water chemistry. The model is verified on the basis of independent tests performed on a series of zirconium alloys under autoclave and reactor conditions. The method developed makes it possible to predict the corrosion of fuel-element cladding made from zirconium alloys with fuel burnup to 80 MW·days/kg under the conditions of one-and two-phase VVER and RBMK coolant.To validate any alloy for use as fuel-element cladding, it is necessary to know the properties of the alloy including its corrosion resistance. The alloy can be chosen in three ways: reactor tests, extra-reactor studies, and verification of the computational models on the basis of experimental data. When studying the corrosion of zirconium alloys, a large number of factors must be taken into account, including the properties of the coolant and the cladding material as well as the operating parameters.The empirical models of the corrosion of zirconium alloys treat oxidation as a thermally activated process. The models are limited to the systems for which the corrosion rate is determined mainly by the corrosion aggressiveness of one participant in the reaction -water, i.e., the models are developed for strictly determined operating conditions and one chosen alloy [1,2].The objective of the present work was to develop a single methodological approach to predicting the corrosion of commercial and experimental zirconium alloys in water media under autoclave and reactor conditions.The present work proposes a method of predicting corrosion of cladding zirconium alloys as a function of their composition and service conditions. The physical-chemical basis of the method is based on the laws of thermodynamics and chemical kinetics of the oxidation reactions of a multicomponent zirconium alloy. Model of Corrosion of Zirconium AlloysEffect of Water Chemistry. A model giving a unified description of the corrosion mechanism for cladding zirconium alloys for nuclear power plants with RBMK and VVER throughout the entire operating cycle of fuel assemblies -in the core at high temperature and during storage of spent nuclear in water at low temperature -has been developed at the All-Russia Research and Design Institute of Integrated Power Technology (VNIPIET) on the basis of an analysis of operating and exper-

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Effect of Water Chemistry and Composition on Microstructural Evolution of Oxide on Zr Alloys

Cited by 4 publications

References 24 publications

The Distribution of Li Ions in the Oxide Film Formed on Zircaloy-4 Corroded in Lithiated Water at 633 K

The Distribution of Li Ions in the Oxide Film Formed on Zircaloy-4 Corroded in Lithiated Water at 633 K

The Influence of Different Contents of Bi Addition on the Corrosion Behavior of Various Zirconium-Based Alloys

Systematic approach to predicting corrosion of zirconium alloys in the water coolant of nuclear reactors

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