Results of Atr Sample Fuel Plate Irradiation Experiment

Graber, M.J.; Gibson, G.W.; Walker, V.; Francis, W.C.

doi:10.2172/4093899

Cited by 5 publications

(2 citation statements)

References 2 publications

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“…These oxide-hydrates are soluble in water, especially in flowing water, meaning that even the hydrated oxide becomes less protective, further enhancing film growth. As tests of pre- filmed samples showed [15], because boehmite has a higher solubility than bayerite, dissolution is selective in the oxide. Dissolution may take place not exclusively at the oxide-water interface, but within the oxide if cracks and fissures are available.…”

Section: Model Developmentmentioning

confidence: 99%

Oxidation of aluminum alloy cladding for research and test reactor fuel

Kim

Hofman

Robinson

et al. 2008

Journal of Nuclear Materials

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Section: Model Developmentmentioning

confidence: 99%

Oxidation of aluminum alloy cladding for research and test reactor fuel

Kim

Hofman

Robinson

et al. 2008

Journal of Nuclear Materials

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“…This work focuses on the Mark VII fuel design. Information related to legacy fuel forms is considered only as needed to supplement the Mark VII fuel literature [13][14][15]. An ongoing effort is under way to convert the ATR core to low-enriched uranium (LEU) [16,17].…”

Section: Advanced Test Reactor Fuelmentioning

confidence: 99%

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Hirschhorn

McFarlane²

2023

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Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys.A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

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The Use of Uranium Aluminide Powders in Nuclear Reactor Fuel Elements

Gibson¹,

deBoisblanc²

1966

Modern Developments in Powder Metallurgy

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Results of Atr Sample Fuel Plate Irradiation Experiment

Cited by 5 publications

References 2 publications

Oxidation of aluminum alloy cladding for research and test reactor fuel

Oxidation of aluminum alloy cladding for research and test reactor fuel

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

The Use of Uranium Aluminide Powders in Nuclear Reactor Fuel Elements

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