Precipitate growth kinetics in Zircaloy-4

Gros, J.P.; Wadier, Jean-Francois

doi:10.1016/0022-3115(90)90012-c

Cited by 59 publications

(17 citation statements)

References 4 publications

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“…Various cumulative annealing parameters (CAP) have been proposed for this purpose in the literature, based on recrystallization activation energy for corrosion behavior. It is believed that the growth of these precipitates during the various stages of alloy processing sequence will involve Ostwald ripening due to the rapid precipitation of iron and chromium in Zircaloy-4 [34]. The first stage of Ostwald ripening involves a nucleation period, followed by growth of the nuclei as the matrix supersaturation decreases.…”

Section: Characterization Of Materialsmentioning

confidence: 99%

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Macdonald¹,

Urquidi‐Macdonald²,

Chen³

et al. 2006

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Our work involves the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with the emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the concentration of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order that we can accurately describe the environment that is contact with the Zircaloy cladding.In the current report, we have derived the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing the oxidation. The models are "deterministic" because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday's law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is some evidence that they control the rate of the overall film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film.Controlled hydrodynamic methods have been used to measure important parameters in the Point Defect Model, which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing alloy in water-cooled nuclear reactors. Firstly, polarization studies have been carried to determine the relevant potential ranges for BWR and PWR coolant systems. Impedance and Mott-Schottky analyses have been carried out on steady state passive films at different formation potentials in 0.1 M B(OH) 3 + 0.001 M LiOH solution of pH = 6.94 at 250 o C. The pressure was 62 bar (900 psi) to maintain the electrolyte in the liquid state. The relationships between current density and film thickness and the film formation potential satisfy the diagnostic criteria of the PDM for an n-type passive film. The kinetic parameters are being extracted by optimization of the PDM on the experimental impedance data, which are validated as conforming to the constraints of linear system theory by using the Kramers-Kronig (K-K) transforms. Mott-Schottky analysis has shown that the passive film is n-type in electronic character corresponding to a preponderance of oxygen/hydrogen vacancies or metal interstitials in the barrier layer. The donor density, N D , calculated from the slope of Mott-Schottky plots is in the range of 10 17~1 0 18 cm -3 , which decreases with increasing formation potential, demonstrating that the film is only lo...

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Section: Characterization Of Materialsmentioning

confidence: 99%

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Macdonald¹,

Urquidi‐Macdonald²,

Chen³

et al. 2006

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“…These are the interfacial energy of the SPPs, and the nucleation site density. Values for these parameters were taken directly from those used in previous work to model SPP formation in Zircaloy-2 [11], which themselves were derived from experiments on Zircaloy-4 [2]. No further tuning of these values was undertaken since the purpose of the present model was not to provide a high level of quantitative accuracy (which would not be justified given the inherent approximations) but rather to demonstrate qualitatively how the SPPs should evolve under combined thermal and irradiation exposure.…”

Section: Kinetic Modelmentioning

confidence: 99%

“…In addition, a validated kinetic model has been developed previously for SPPs in Zircaloys can provide many of the input parameters needed in the present work [11]. Only the dominant SPP phase present under the conditions of interest is considered, which is face-centred cubic Zr(Cr,Fe) 2 [2].…”

Section: The Modelmentioning

confidence: 99%

“…Fe and Cr in Zircaloy-4) precipitate as SPPs. In Zircaloy-4, the dominant precipitate phases are intermetallics of composition Zr(Fe,Cr) 2 and can take either hexagonal or face centred cubic allotropic forms [2].…”

Section: Introductionmentioning

confidence: 99%

“…Corrosion behaviour depends critically on the volume fraction, size, and distribution of the SPPs, which must be optimised to obtain the best performance. For example, in Zircaloy-4 it has been demonstrated that the optimum precipitate size distribution must be in a range where the particles are sufficiently fine to give good nodular corrosion resistance whilst remaining above a critical minimum size that is required for good uniform corrosion resistance [1,2]. SPPs and the solute from which they are formed also have a key role in determining other properties such as strength and susceptibility to irradiation growth [3][4][5].…”

Section: Introductionmentioning

confidence: 99%

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Modeling precipitate evolution in zirconium alloys during irradiation

Robson

2016

Journal of Nuclear Materials

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The second phase precipitates (SPPs) in zirconium alloys are critical in controlling their performance. During service, SPPs are subject to both thermal and irradiation effects that influence volume fraction, number, and size. In this paper, a model has been developed to capture the combined effect of thermal and irradiation exposure on the Zr(Fe,Cr) 2 precipitates in Zircaloy. The model includes irradiation induced precipitate destabilisation integrated into a classical size class model for nucleation, growth and coarsening. The model has been applied to predict the effect of temperature and irradiation on SPP evolution. Increasing irradiation displacement rate is predicted to strongly enhance the loss of particles that arises from coarsening alone. The effect of temperature is complex due to competition between coarsening and irradiation damage. As temperature increases, coarsening is predicted to become increasingly important compared to irradiation induced dissolution and may increase resistance to irradiation induced dissolution by increasing particle size.

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The rolling schedule of Zircaloy-4 strip during multi-schedule and multi-pass hot rolling process

Cao,

Gao

et al. 2023

Int J Adv Manuf Technol

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Precipitate growth kinetics in Zircaloy-4

Cited by 59 publications

References 4 publications

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Modeling precipitate evolution in zirconium alloys during irradiation

The rolling schedule of Zircaloy-4 strip during multi-schedule and multi-pass hot rolling process

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