Comparative study of simulated zirconia inert matrix fuel stabilized with yttrium, lanthanum or praseodymium: Synthesis and leaching tests

Kamel, N.; Aït-Amar, H.; Taouinet, M.; Benazzouz, C.; Kamel, Ziane; Fodil-Cherif, H.; Telmoune, S.; Slimani, Rachid; Zahri, A.; Sahel, D.

doi:10.1016/j.pnucene.2005.05.082

Cited by 24 publications

(21 citation statements)

References 37 publications

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“…It is highly radiation tolerant and compatible with reactor materials [2,3,[8][9][10][11][12]. Zirconia is suited to structurally incorporate fissile material and burnable poisons within desired compositional ranges.…”

Section: Introductionmentioning

confidence: 99%

Synthesis and characterization of zirconia–magnesia inert matrix fuel: Ce homolog studies

Holliday

Hartmann

Czerwinski

2009

Journal of Nuclear Materials

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a b s t r a c t X-ray powder diffraction, microscopy, thermal analysis and electron probe microanalysis were used to characterize a ZrO 2 -MgO inert matrix containing CeO 2 as a homolog for PuO 2 and Er 2 O 3 as a burnable poison. The synthesis was carried out using a precipitation method. A large composition range of MgO to ZrO 2 was evaluated to determine phases present, phase mixing, phase composition, microstructure and thermal properties. It was found that most compositions of the material consist of two phases: MgO (periclase) and ZrO 2 (cubic zirconia). The zirconia phase incorporates 5% (wt/wt) MgO and up to 14% and 12% (wt/wt) CeO 2 and Er 2 O 3 , respectively. The MgO phase remains pure, which will enable it to retain its heat transfer and solubility properties and will improve the overall thermal conductivity and reprocessing component of the inert matrix fuel. The results with Ce will be used as the basis of future studies with actinides.Published by Elsevier B.V.

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Section: Introductionmentioning

confidence: 99%

Synthesis and characterization of zirconia–magnesia inert matrix fuel: Ce homolog studies

Holliday

Hartmann

Czerwinski

2009

Journal of Nuclear Materials

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“…Cubic zirconia (ZrO 2 ), which adopts the fluorite structure, has often been proposed for use as IMFs due to its low neutron cross section, chemical durability, and resistance to radiation damage. [13][14][15][16] Further, cubic zirconia has the added benefit of being a viable long-term storage material, which would allow for direct storage of the spent nuclear fuel without the need for further reprocessing. 13,17,18 However, pure cubic zirconia is not stable at temperatures below 2370°C, adopting a tetragonal structure between 1170 and 2370°C and a monoclinic structure below 1170°C.…”

Section: Introductionmentioning

confidence: 99%

“…19,31 The depth of knowledge regarding yttria-stabilized zirconia (YSZ) is large, and YSZ materials have been studied extensively for use in a wide variety of applications, including solid-oxide fuel cells, refractory material applications, and structural wear components. 14,20,[32][33][34][35][36] However, research to determine how the material properties are impacted by adding TRU species to YSZ remains relatively rare. It is necessary to understand how stability of the fluorite structure of YSZ is impacted by the addition of TRU elements, as there is an increase in volume and a decrease in thermal conductivity accompanying the transition from fluorite to a lower symmetry phase.…”

Section: Introductionmentioning

confidence: 99%

Investigation of Nd_xY_0.25–_xZr_0.75O_1.88 inert matrix fuel materials made by a co-precipitation synthetic route

Hayes

Grosvenor

2016

Can. J. Chem.

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Yttria-stabilized zirconia (YSZ) is a material that is being considered for use as an inert matrix fuel in nuclear reactors, but a complete characterization of these materials is required for them to be licensed for use. A series of Nd x Y 0.25-x Zr 0.75 O 1.88 materials have been synthesized using a co-precipitation method, and the thermal stability of these materials has been studied by annealing them at 1400 and 1500°C. (Nd was used as surrogate for Am.) The long-range and local structures of the materials were characterized via powder X-ray diffraction, scanning electron microscopy, wavelength dispersive spectroscopy, and X-ray absorption spectroscopy at the Zr K-and Y K-edges. These results were compared with the previous characterization of Nd-YSZ materials synthesized using a ceramic method. The results indicated that the ordering in the local metal-oxygen polyhedral remains relatively unaffected by the synthetic method, but there was increased long-range disorder in the materials prepared by the co-precipitation method. Further, it was found that the materials produced by the co-precipitation method were unexpectedly unstable when annealed at high temperature. This study highlights the importance of determining the effect of synthetic method on material properties and demonstrates how the co-precipitation route could be used to produce inert matrix fuels.Key words: XAS, ceramics, stabilized zirconia, nuclear energy, XRD.Résumé : La zircone stabilisée par oxyde d'yttrium (YSZ) constitue un matériau dont l'utilisation est envisagée comme combustible à matrice inerte dans les réacteurs nucléaires, mais une caractérisation complète de tels matériaux est nécessaire pour que leur utilisation puisse être autorisée. Nous avons synthétisé une série de matériaux de formule Nd x Y 0,25-x Zr 0,75 O 1,88 par une méthode de coprécipitation et étudié la stabilité thermique de ces matériaux en les soumettant à un recuit à 1 400 et 1 500°C. (Le Nd a été employé comme substitut de l'Am.) Par la suite, nous avons caractérisé les matériaux en effectuant des analyses structurales d'ordre à longue et à courte distance au moyen de techniques de diffraction des rayons X sur poudre, de microscopie électronique à balayage, de spectroscopie de dispersion des longueurs d'onde et de spectroscopie d'absorption des rayons X au seuil K du Zr et de l'Y. Nous avons comparé ces résultats avec les analyses antérieures de caractérisation de matériaux de Nd-YSZ préparés à l'aide d'une méthode de synthèse de céramiques. Les résultats ont montré que la méthode de synthèse a relativement peu d'influence sur l'ordre à courte distance dans le polyèdre métal-oxygène, tandis qu'une augmentation du désordre à longue distance a été observée dans les matériaux synthétisés par la méthode de coprécipitation. De plus, nous avons observé que les matériaux produits par la méthode de coprécipitation présentaient une instabilité imprévue lorsqu'ils étaient soumis au recuit à haute température. La présente étude souligne l'importance de déterminer l...

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“…The lack of uranium in the fuel allows for theconsumption of transuranics without any additional production, hence the term "inert". Because of its mechanical properties and radiation hardness, zirconium dioxide based fuel has received considerable attention in the past decade and have gone through initial testing at high flux test facilities with good results [4,[13][14][15][16][17][18].…”

Section: Introductionmentioning

confidence: 99%

Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

Recktenwald

Deinert

2013

Energies

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Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

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Comparative study of simulated zirconia inert matrix fuel stabilized with yttrium, lanthanum or praseodymium: Synthesis and leaching tests

Cited by 24 publications

References 37 publications

Synthesis and characterization of zirconia–magnesia inert matrix fuel: Ce homolog studies

Synthesis and characterization of zirconia–magnesia inert matrix fuel: Ce homolog studies

Investigation of Nd_xY_0.25–_xZr_0.75O_1.88 inert matrix fuel materials made by a co-precipitation synthetic route

Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

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Comparative study of simulated zirconia inert matrix fuel stabilized with yttrium, lanthanum or praseodymium: Synthesis and leaching tests

Cited by 24 publications

References 37 publications

Synthesis and characterization of zirconia–magnesia inert matrix fuel: Ce homolog studies

Synthesis and characterization of zirconia–magnesia inert matrix fuel: Ce homolog studies

Investigation of NdxY0.25–xZr0.75O1.88 inert matrix fuel materials made by a co-precipitation synthetic route

Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

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Investigation of Nd_xY_0.25–_xZr_0.75O_1.88 inert matrix fuel materials made by a co-precipitation synthetic route