Hydrogen diffusion process in the oxides formed on zirconium alloys during corrosion in pressurized water reactor conditions

Tupin, Marc; Martin, Frantz; Bisor, Caroline; Verlet, Romain; Bossis, Philippe; Chêne, J.; Jomard, François; Berger, Pascal; Pascal, Serge; Nuns, Nicolás

doi:10.1016/j.corsci.2016.10.027

Cited by 39 publications

(36 citation statements)

References 18 publications

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“…The outer sublayer seems easily damaged, and the inner sublayer stays still well-adjacent to the substrate. This corresponds to the previous research [71] in which the oxide layer has been shown to be composed of two sublayers: the external one extremely permeable to hydrogen even at over 300 • C, and the inner one, much more impermeable to hydrogen.…”

Section: Barrier Effect Of the Oxide Layersupporting

confidence: 90%

“…All these factors depend on the oxidation temperature and time, composition and fugacity of a medium, and chemical and phase composition of an alloy [23]. The obtained results here correspond to the previous reports regarding the temperature dependence of the oxide layer growth [25,[61][62][63][64][65][66][67], the complex multilayer and multiphase structure [68][69][70][71], growth of the oxide layer in columnar grains [68,72], and the appearance of porosity [25,33,[73][74][75] and cracks [76][77][78][79]. Interestingly, it was also noted that a severe descaling of the oxide layer appeared in these investigations at a relatively high temperature.…”

Section: Characteristics Of Oxide Layerssupporting

confidence: 89%

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Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

et al. 2020

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The present paper is aimed at determining the less investigated effects of hydrogen uptake on the microstructure and the mechanical behavior of the oxidized Zircaloy-2 alloy. The specimens were oxidized and charged with hydrogen. The different oxidation temperatures and cathodic current densities were applied. The scanning electron microscopy, X-ray electron diffraction spectroscopy, hydrogen absorption assessment, tensile, and nanoindentation tests were performed. At low oxidation temperatures, an appearance of numerous hydrides and cracks, and a slight change of mechanical properties were noticed. At high-temperature oxidation, the oxide layer prevented the hydrogen deterioration of the alloy. For nonoxidized samples, charged at different current density, nanoindentation tests showed that both hardness and Young’s modulus revealed the minims at specific current value and the stepwise decrease in hardness during hydrogen desorption. The obtained results are explained by the barrier effect of the oxide layer against hydrogen uptake, softening due to the interaction of hydrogen and dislocations nucleated by indentation test, and hardening caused by the decomposition of hydrides. The last phenomena may appear together and result in hydrogen embrittlement in forms of simultaneous hydrogen-enhanced localized plasticity and delayed hydride cracking.

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Section: Barrier Effect Of the Oxide Layersupporting

confidence: 90%

Section: Characteristics Of Oxide Layerssupporting

confidence: 89%

Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

et al. 2020

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“…where D 0 is the exponential pre-factor, e m is the migration energy, k is Boltzmann's constant, and the superscript α can refer to the oxide ('ox') or the metal ('m'). The diffusivity of H in Zircaloy-4 oxides has been recently measured by Tupin et al [80], which give values of 2.5 × 10 −14 m·s −1 and 0.41 eV for D ox 0 and e ox m , respectively. Alloy composition, however, has been shown to have a significant impact on diffusion parameters.…”

Section: H Atom Diffusionmentioning

confidence: 91%

Kinetic Model of Incipient Hydride Formation in Zr Clad under Dynamic Oxide Growth Conditions

Reyes

Shah

et al. 2020

Materials

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The formation of elongated zirconium hydride platelets during corrosion of nuclear fuel clad is linked to its premature failure due to embrittlement and delayed hydride cracking. Despite their importance, however, most existing models of hydride nucleation and growth in Zr alloys are phenomenological and lack sufficient physical detail to become predictive under the variety of conditions found in nuclear reactors during operation. Moreover, most models ignore the dynamic nature of clad oxidation, which requires that hydrogen transport and precipitation be considered in a scenario where the oxide layer is continuously growing at the expense of the metal substrate. In this paper, we perform simulations of hydride formation in Zr clads with a moving oxide/metal boundary using a stochastic kinetic diffusion/reaction model parameterized with state-of-the-art defect and solute energetics. Our model uses the solutions of the hydrogen diffusion problem across an increasingly-coarse oxide layer to define boundary conditions for the kinetic simulations of hydrogen penetration, precipitation, and dissolution in the metal clad. Our method captures the spatial dependence of the problem by discretizing all spatial derivatives using a stochastic finite difference scheme. Our results include hydride number densities and size distributions along the radial coordinate of the clad for the first 1.6 h of evolution, providing a quantitative picture of hydride incipient nucleation and growth under clad service conditions.

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“…-the pre-transition one consists in the growth of a protective oxide on the metal and the oxidation kinetics follows a cubic law [7,23] ;…”

Section: The Present Work Is Then Focused On the Effect Of Ion Irradimentioning

confidence: 99%

“…The main assumptions suggested in the literature to explain this acceleration are the precipitation of zirconium hydrides under the metal/oxide interface resulting from the hydrogen uptake by the matrix during the corrosion process, the tin distribution evolution, the amorphization of the Second Phases Particles (SPPs) and their dissolution under irradiation [3][4][5][6][7][8][9][10][11][12][13][14]. This high burnup acceleration could also be partially caused by the accumulation of irradiation damage in the cladding material or the radiolysis of the water.…”

Section: Introductionmentioning

confidence: 99%

Influence of light ion irradiation of the oxide layer on the oxidation rate of zircaloy-4: Irradiation of the post-transition oxide layer

2019

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The oxidation rate of Zircaloy-4 fuel cladding in nuclear reactor strongly increases at high burnups. This kinetics acceleration could be in large part due to the irradiation damage. The irradiation effect of the post-transition oxide layer on the Zircaloy-4 corrosion rate has been investigated using protons. As previously observed on pre-transition oxide layer, irradiation defects increases the oxidation rate of the alloy up to around 30 days in autoclave in agreement with the irradiation defect annealing characterized by Raman spectroscopy. The model proposed in previous works is able to describe the oxidation rate after irradiation of post-transition oxide layers.

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Hydrogen diffusion process in the oxides formed on zirconium alloys during corrosion in pressurized water reactor conditions

Cited by 39 publications

References 18 publications

Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

Kinetic Model of Incipient Hydride Formation in Zr Clad under Dynamic Oxide Growth Conditions

Influence of light ion irradiation of the oxide layer on the oxidation rate of zircaloy-4: Irradiation of the post-transition oxide layer

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