Use of a Centrifugal Atomization Process in the Development of Research Reactor Fuel

Kim, Chang Kyu; Park, Jong Man; Ryu, Ho Jin

doi:10.5516/net.2007.39.5.617

Cited by 41 publications

(14 citation statements)

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“…U-(6-10 wt.%)Mo alloy powders were produced by a centrifugal atomization technique [7]. Depleted uranium lump (99.9 wt.%) and molybdenum lump (99.7 wt.%) were induction melted in a graphite crucible coated with yttria stabilized zirconia and a total weight of charges per each run was 4 kg.…”

Section: Methodsmentioning

confidence: 99%

Neutron diffraction analyses of U–(6–10wt.%)Mo alloy powders fabricated by centrifugal atomization

Park

Ryu

Kim

et al. 2010

Journal of Nuclear Materials

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Section: Methodsmentioning

confidence: 99%

Neutron diffraction analyses of U–(6–10wt.%)Mo alloy powders fabricated by centrifugal atomization

Park

Ryu

Kim

et al. 2010

Journal of Nuclear Materials

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“…U-7 wt.% Mo particles (enrichment <20 wt.% 235 U) produced by centrifugal atomisation were selected [20,21]. They were http://dx.doi.org/10.1016/j.jnucmat.2014.05.052 subsequently oxidised after the initial particle manufacturing process using a thermal treatment under air at 200-250°C during a few hours.…”

Section: Iris 4 Fuel Plate Manufacturing and Fresh Fuel Plate Charactmentioning

confidence: 99%

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

Zweifel¹,

Valot²,

Pontillon³

et al. 2014

Journal of Nuclear Materials

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is an open access repository that collects the work of Arts et Métiers ParisTech researchers and makes it freely available over the web where possible. a b s t r a c t U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

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“…While the use of large U-Mo particles with diameters of 200-500 µm has been proposed as a solution to mitigate the interaction problems [6,7], the addition of Si to the Al matrix has also been found to be a promising remedy to reduce the interactions [8]. Irradiation tests using Al-Si matrices instead of the Al matrix have shown reductions in the interaction layer growth [9].…”

Section: Introductionmentioning

confidence: 99%

The Effect of Si-Rich Layer Coating on U-Mo vs. Al Interdiffusion

Ryu

Park²,

Park³

et al. 2011

Nuclear Engineering and Technology

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Use of a Centrifugal Atomization Process in the Development of Research Reactor Fuel

Cited by 41 publications

References 1 publication

Neutron diffraction analyses of U–(6–10wt.%)Mo alloy powders fabricated by centrifugal atomization

Neutron diffraction analyses of U–(6–10wt.%)Mo alloy powders fabricated by centrifugal atomization

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

The Effect of Si-Rich Layer Coating on U-Mo vs. Al Interdiffusion

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