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Corrosion of irradiated MOX fuel in presence of dissolved H-2

P. Carbol ; Patrik Fors (Institutionen för kemi- och bioteknik, Kärnkemi) ; S. Van Winckel ; Kastriot Spahiu (Institutionen för kemi- och bioteknik, Kärnkemi)
Journal of Nuclear Materials (0022-3115). Vol. 392 (2009), 1, p. 45-54.
[Artikel, refereegranskad vetenskaplig]

The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO3 solution in presence of dissolved H-2 for 2100 days. The results show that dissolved H-2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 x 10(-10) and 5 x 10(-11) M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible. (C) 2009 Elsevier B.V. All rights reserved.

Nyckelord: spent nuclear-fuel, near-field hydrogen, radionuclide release, product, deposits, fission-gas, dissolution, uo2, groundwater, radiolysis, repository

Denna post skapades 2010-02-24.
CPL Pubid: 114636


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Institutioner (Chalmers)

Institutionen för kemi- och bioteknik, Kärnkemi (2005-2014)



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