Chromia doped UO2 fuel: Investigation of the lattice parameter

Cardinaels, Thomas; Govers, Kevin; Vos, B; Berghe, S. Van den; Verwerft, Marc; Tollenaere, L de; Maier, Georg; Delafoy, C.

doi:10.1016/j.jnucmat.2012.02.025

Cited by 50 publications

(35 citation statements)

References 25 publications

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“…At the same time, only small amounts of Cr could be incorporated into UO 2 , with maximum solubility at high temperature (T~2300 K) of up to 0.1 wt%∕UO 2 [3][4][5]9 . The formed diluted solid solution shows a linear decrease in the matrix lattice parameter with content of Cr 1,9 , which however, is significantly smaller than the predictions by most theoretical models 1 .…”

mentioning

confidence: 58%

“…uclear reactors use mainly UO 2 -based fuel elements. Additions of other trace elements, such as chromium, are considered to improve the microstructural and mechanical properties of fuel matrices 1,2 . Among other desired effects, chromium is known to significantly increase the grain sizes, which due to the elongation of bulk diffusion path limits the release of fission gas products such as xenon or krypton 2 , and increases the fission products retention capability by the fuel elements 3 .…”

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confidence: 99%

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The +2 oxidation state of Cr incorporated into the crystal lattice of UO2

2020

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Doping by Cr is used to improve the performance of uranium dioxide (UO 2)-based nuclear fuel. However, the mechanism of structural incorporation of Cr remains unclear. Here, in order to understand this process on the atomic scale and the redox state of Cr in UO 2-based nuclear fuel, we performed intensive ab initio atomistic simulations of the Cr doped UO 2 matrix. We unexpectedly found that Cr in UO 2 exists as Cr 2+ species and not as the widely claimed Cr 3+. We re-evaluated previously published x-ray absorption near edge structure spectroscopy data and confirmed the computed redox state of Cr. Thermodynamic consideration shows that the favorable structural arrangement of Cr in UO 2 is given by a pair of associated Cr 2+ and oxygen vacancy. The realism of this doping mechanism is further demonstrated by a match to the measured maximum Cr solubility and small lattice contraction.

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confidence: 58%

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confidence: 99%

The +2 oxidation state of Cr incorporated into the crystal lattice of UO2

2020

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“…The chromium content in solid solution for each ramped rodlet was measured in the vicinity of 0.1 wt% Cr/UO 2 regardless of temperature, in the whole range [20°C -1400°C]. The chromium soluble content was equivalent to that of the initial un-irradiated fuel after sintering, measured to 0.102 ± 0.004% 11,17,18 .…”

Section: Low Temperature Range (<1460°c): Cold Section Of the Pelletmentioning

confidence: 99%

Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

Riglet-Martial

Sercombe

Lamontagne

et al. 2016

Journal of Nuclear Materials

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The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO 2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that sub-stoichiometric irradiated fuels under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U 4 O 9 type in its cold section, of lower temperature. To our knowledge, it is the first time that such a phenomenon, likely to greatly impact the corrosive fission gas release, is experimentally attested for PWR fuels. The parameters governing the oxidation states of UO 2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO 2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

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“…Indeed, Riglet-Martial et al [2] have shown by X-ray absorption near edge structure (XANES) that the oxidation state of soluble Cr is 3+ only in UO 2 , creating obvious charge defects. According to the experimental and calculation work of Cardinaels et al [16], the most favorable site for Cr satisfying the observed variation of lattice parameter of doped UO 2 is the substitution of uranium combined with a bonding with a [9], are formally expected to be positively charged, and should therefore in principle be invisible to PAS. However, Vathonne et al [15] has shown by DFT + U method that V O charged À2 could also be stable for Fermi levels lying close to the middle of the band gap, so that the presence and detection of this very-stable defect should not be absolutely excluded.…”

Section: Pas Componentsmentioning

confidence: 98%

Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels

Chollet

Kršjak

Cozzo

et al. 2017

EPJ Nuclear Sci. Technol.

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Abstract. Fission gas behavior within the fuel structure plays a major role for the safety of nuclear fuels during operation in the nuclear power plant. Fission gas distribution and retention is determined by both, micro-and lattice-structure of the fuel matrix. The ADOPT (Advanced Doped Pellet Technology) fuel, containing chromium and aluminum additives, shows larger grain sizes than standard (undoped) UO 2 fuel, enhancing the fission gas retention properties of the matrix. However, the additions of such trivalent cations shall also induce defects in the lattice. In this study, we investigated the microstructure of such doped fuels as well as a reference standard UO 2 by positron annihilation spectroscopy (PAS). Although this technique is particularly sensitive to lattice point defects in materials, a wider application in the UO 2 research is still missing. The PAS-lifetime components were measured in the hotlab facility of PSI using a 22 Na source sandwiched between two 500-mmthin sample discs. The values of lifetime at the center and the rim of both samples, examined to check at the radial homogeneity of the pellets, are not significantly different. The mean lifetimes were found to be longer in the ADOPT material, 220 ps, than in standard UO 2 , 190 ps, which indicates a larger presence of additional defects, presumably generated by the dopants. While two-component decomposition (bulk + one defect component) could be performed for the standard material, only one lifetime component was found in the doped material. The absence of the bulk component in the ADOPT sample refers to a saturated positron trapping (i.e., all positrons are trapped at defects). In order to associate a type of lattice defect to each PAS component, interpretation of the PAS experimental observations was conducted with respect to existing experimental and modeling studies. This work has shown the efficiency of PAS to detect lattice point defects in UO 2 produced by Cr and Al oxides. These additives create lattice irregularities, which are acting as sinks for fission products on one hand and trapping positrons on the other hand. Fitting of the obtained experimental data with a suitable theoretical model can provide a valuable qualitative assessment of these defects. At this stage of the research, some of the existing models were used for this purpose.

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Chromia doped UO2 fuel: Investigation of the lattice parameter

Cited by 50 publications

References 25 publications

The +2 oxidation state of Cr incorporated into the crystal lattice of UO2

The +2 oxidation state of Cr incorporated into the crystal lattice of UO2

Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels

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