Corrosion of irradiated MOX fuel in presence of dissolved H-2
Journal article, 2009

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.

deposits

groundwater

radiolysis

uo2

product

fission-gas

repository

near-field hydrogen

dissolution

spent nuclear-fuel

radionuclide release

Author

P. Carbol

Joint Research Centre (JRC), European Commission

Patrik Fors

Chalmers, Chemical and Biological Engineering, Nuclear Chemistry

S. Van Winckel

Joint Research Centre (JRC), European Commission

Kastriot Spahiu

Chalmers, Chemical and Biological Engineering, Nuclear Chemistry

Journal of Nuclear Materials

0022-3115 (ISSN)

Vol. 392 1 45-54

DOI

10.1016/j.jnucmat.2009.03.044

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