Nonoxide Ceramic Nuclear Fuels

Blank, H.

doi:10.1002/9783527603978.mst0108

Cited by 9 publications

(7 citation statements)

References 168 publications

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“…Nitrogen gas that was generated owing to the oxidation of UN could be re-dissolved in the UO 2-x lattice and form a UO 2 based nonstoichiometric oxynitride phase [27]. Upon cooling, the oxynitride phase decomposed into a UO 2 and nitride phase [28]. Therefore, the U 2 N 3 phase in the sintered samples might originate from the U(ON) 2-x type oxynitride phase.…”

Section: Resultsmentioning

confidence: 99%

UO2–UN composites with enhanced uranium density and thermal conductivity

Yang

Kim

et al. 2015

Journal of Nuclear Materials

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Section: Resultsmentioning

confidence: 99%

UO2–UN composites with enhanced uranium density and thermal conductivity

Yang

Kim

et al. 2015

Journal of Nuclear Materials

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“…Each 3DOT simulation model consisted of a fuel cube with an implanted sphere of xenon. Owing to the small computation expense of BCA and the wide spread of bubble sizes present in uranium carbide [19], the simulated bubble radii ranged from 0.5 nm to 500 nm with periodic boundary conditions enforced on all box sides. As the current study was focused on intra-granular bubbles, grain boundary effects were not considered and the theoretical density of 13.63 g/cm 3 was assumed for uranium carbide.…”

Section: Methodsmentioning

confidence: 99%

“…The ability of the heterogenous re-solution model to accurately estimate resolution in oxide fuels, while homogenous re-solution seems more applicable to uranium carbide fuel, lies in the inherent differences between the two materials; uranium carbide is bonded by a mix of covalent and metallic contributions [19], resulting in different energy transport properties such as higher thermal conductivity and diffusivity when compared to UO 2 . Fission fragments are well known to create displacement spikes in materials [20].…”

Section: Introductionmentioning

confidence: 99%

Radiation re-solution of fission gas in non-oxide nuclear fuel

Matthews¹,

Schwen²,

Klein³

2015

Journal of Nuclear Materials

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Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called "homogenous" atom-by-atom resolution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the resolution parameter as bubble radius increases until about 50 nm, at which the resolution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated resolution parameter can be used to find a resolution rate for future bubble simulations.

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“…Sintering is performed at high temperature (T>2; 000 K) in order to achieve the 15-20% porosity recommended for nitride fuels. 17) However, metal nitrides tend to dissociate at the temperatures used, and volatile elements may evaporate:…”

Section: Thermal Stability: Methodsmentioning

confidence: 99%

Thermo-chemical Modelling of Uranium-free Nitride Fuels

Jolkkonen

Streit²,

Wallenius³

2004

Journal of Nuclear Science and Technology

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A production process for americium-bearing, uranium-free nitride fuels was modelled using the newly developed ALCHYMY thermochemical database. The results suggested that the practical difficulties with yield and purity are of a kinetic rather than a thermodynamical nature. We predict that the immediate product of the typical decarburisation step is not methane, but hydrogen cyanide. HCN may then undergo further reactions upon cooling, explaining the difficulty in observing any carbophoric molecules in the gaseous off stream. The thermal stability of nitride fuels in different environments was also estimated. We show that sintering of nitride compounds containing americium should be performed under nitrogen atmosphere in order to the avoid the excessive losses of americium reported from sintering under inert gas. Addition of nitrogen in small amounts to fuel pin filling gas also appears to significantly improve the in-pile stability of transuranium nitride fuels.

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Nonoxide Ceramic Nuclear Fuels

Cited by 9 publications

References 168 publications

UO2–UN composites with enhanced uranium density and thermal conductivity

UO2–UN composites with enhanced uranium density and thermal conductivity

Radiation re-solution of fission gas in non-oxide nuclear fuel

Thermo-chemical Modelling of Uranium-free Nitride Fuels

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