Oxygen and nitrogen interstitial ordering in hcp Ti, Zr, and Hf: An<i>ab initio</i>study

Ruban, Andrei V.; Baykov, Vitaly; Johansson, Börje; Дмитриев, В. В.; Blanter, M. S.

doi:10.1103/physrevb.82.134110

Cited by 46 publications

(30 citation statements)

References 41 publications

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“…S2). The stoichiometries Hf 6 N and Hf 3 N were never reported by experiments, while previous MonteCarlo simulations [36] predicted the existence of Hf 6 N and Hf 3 N, and also indicated that the groundsate of Hf 6 N was P31c-Hf 6 N while the Hf 3 N belonged to R3c or P 6 3 22-Hf 3 N. Our calculations, with or without considering the zero point energy (ZPE), both not only showed that P31c-Hf 6 N is the most stable (note the free energy P31c-Hf 6 N is slightly lower than that of R3-Hf 6 N), but also indicated P 6 3 22-Hf 3 N is more stable that R3c-Hf 3 N at low pressures. Unlike P 4 2 /mnm-Ti 2 N (the stable phase for Ti 2 N at 0 GPa), Hf 2 N is more energetically favorable in an orthorhombic phase (space group: Pnnm); this Pnnm-Hf 2 N transforms to Cmcm-Hf 2 N at 12.5 GPa and Cmcm-Hf 2 N continues to be stable up to 55 GPa ( Fig.…”

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confidence: 78%

Pressure-stabilized hafnium nitrides and their properties

et al. 2017

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confidence: 78%

Pressure-stabilized hafnium nitrides and their properties

et al. 2017

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“…At temperatures between (about) 300°C and 900°C various ordered phases have been reported [5]. Octahedral interstitial ordering of oxygen increases the microhardness [58] and embrittlement [74], and therefore, promotes stress corrosion cracking. However, as it will be demonstrated later, oxygen can be very beneficial for reducing the hydrogen uptake under certain conditions, [71].…”

Section: Suboxides Of Zirconiummentioning

confidence: 99%

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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“…It has been suggested that the solubility of oxygen in the octahedral interstitial sites of α-Hf (hcp-Hf) can be as high as 20 at.% [8], while solubility of oxygen in β-Hf (bcc-Hf) is only 3 at.% [9]. Several experimental [10,11] and theoretical studies [12,13] have investigated the interstitial oxygen in hcp-Hf. Now it is well established that three stoichiometric compositions Hf 6 O, Hf 3 O and Hf 2 O can be formed with increasing occupation of the octahedral-interstitial positions in hcp-Hf by oxygen atoms.…”

Section: Introductionmentioning

confidence: 99%