Hydrogen in Zircaloy-4: effects of the neutron irradiation on the hydride formation

Vizcaı́no, P.; Banchik, A.D.; Abriata, J. P.

doi:10.1007/s10853-007-1525-x

Cited by 29 publications

(26 citation statements)

References 4 publications

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“…They reported optimum size distributions for SPPs in the range of approximately 25-175 nm for BWR conditions. Optimum SPP Fe/Cr ratios of 0.6-1.2 and Fe/Ni ratios of 0.9-1.0 have also been found [69].…”

Section: Oxidationmentioning

confidence: 71%

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Glazoff¹

2013

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Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel.This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

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Section: Oxidationmentioning

confidence: 71%

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Glazoff¹

2013

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“…The terminal solid solubility of hydrogen in Zircaloy-4 was reported to be approximately 250 ppm and 70 ppm at 673 K and 573 K, respectively. 25) The terminal solubility of hydrogen below 473 K was found to 15 ppm. 1,25) Since the solubility of hydrogen in Zr-Nb-Sn-Fe is not available, it is assumed that the solubility of hydrogen in Zr-Nb-Sn-Fe is not much different from that in Zircaloy-4.…”

Section: )mentioning

confidence: 99%

“…25) The terminal solubility of hydrogen below 473 K was found to 15 ppm. 1,25) Since the solubility of hydrogen in Zr-Nb-Sn-Fe is not available, it is assumed that the solubility of hydrogen in Zr-Nb-Sn-Fe is not much different from that in Zircaloy-4. The hydrogen content in this study was measured to be 150 ppm and hydrides are supposedly completely dissolved at 673 K for RS43AC sample.…”

Section: )mentioning

confidence: 99%

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Reorientation of Hydrides and Its Effect on the Mechanical Properties of Zr-Nb-Sn-Fe Cladding Tubes

Lee¹,

Choi

Joo

et al. 2005

Journal of Nuclear Science and Technology

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The reorientation of hydrides and its effect on the fracture of Zr-Nb-Sn-Fe cladding tubes were investigated. The reorientation of hydrides along the radial direction was most pronounced if the tube was cooled from 573 to 473 K under circumferential loading. Reorientation occurred much less frequently at either higher (from 673 to 573 K) or lower (from 473 to 373 K) temperature range. The strength decreased to 250 MPa and the ductility decreased to zero in the tube which was reorientation-treated from 573 to 473 K (RS32AC). Fracture surface of RS32AC sample exhibited flat cleavage fracture along the radial hydride platelet. The reorientation of hydrides was also found to increase with increase of loading time, suggesting time dependent stress-aided dissolution of circumferential hydrides and reprecipitation of radial hydrides.

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“…The effect of SPP on metallurgical properties and corrosion has been also reported [10]. The properties of these intermetallics under irradiation environment in reactor up to a high burn-up have been the subject of several investigations [15][16][17][18][19]. However, structural aspects concerning lattice defects and their preferential location in these intermetallic phases have never been investigated.…”

Section: Introductionmentioning

confidence: 99%

Micro-focussed XAFS spectroscopy to study Ni-bearing precipitates in the metal of corroded Zircaloy-2

et al. 2009

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The present work concerns an investigation of the local atomic environment of Ni-containing secondary phase precipitates (SPP) present in the metal-part of Zircaloy-2 cladding tubes. An unirradiated Zircaloy-2 and two specimens irradiated in a commercial nuclear power plant are characterized using μ-focussed synchrotron radiation, and by x-ray absorption fine structure (XAFS) spectroscopy. The patterns of Ni K-edge XANES and EXAFS of SPP in unirradiated and irradiated cladding are found different. Considering the fact that Ni-bearing SPP in the unirradiated samples are mainly Zintl phase Zr 2 (Fe, Ni) type, a detailed EX-AFS analysis of near-neighbor Ni atoms has been made. The result of a curve fit for the first two shells shows that about 2 Ni(Fe) and 8 Zr atoms are coordinated at 2.68 and 2.77 Å, respectively, around a central Ni atom in the SPP. XANES data analysis provides total electronic density of states at the Fermi level of unirradiated Zr 2 (Fe, Ni). At the Ni K-edge EXAFS spectra of irradiated SPP, however, only a single scattering peak is observed demonstrating the structural disorder introduced by the neutron irradiation. The coordination number of the Ni neighboring shells is reduced markedly due to the formation of point and extended defects in the damaged SPP lattice. Dissolution of Ni from the SPP is also evident from the data. The results of this study provide a further basis for the description of both crystallographic and electronic structures of intermetallic secondphase precipitates found in Zr-based alloys.

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Hydrogen in Zircaloy-4: effects of the neutron irradiation on the hydride formation

Cited by 29 publications

References 4 publications

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Reorientation of Hydrides and Its Effect on the Mechanical Properties of Zr-Nb-Sn-Fe Cladding Tubes

Micro-focussed XAFS spectroscopy to study Ni-bearing precipitates in the metal of corroded Zircaloy-2

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