Corrosion of irradiated MOX fuel in presence of dissolved H-2
Artikel i vetenskaplig tidskrift, 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.

radionuclide release

near-field hydrogen

fission-gas

uo2

spent nuclear-fuel

dissolution

deposits

product

radiolysis

groundwater

repository

Författare

P. Carbol

European Commission Joint Research Centre, Institute for Transuranium Elements Karlsruhe

Patrik Fors

Chalmers, Kemi- och bioteknik, Kärnkemi

S. Van Winckel

European Commission Joint Research Centre, Institute for Transuranium Elements Karlsruhe

Kastriot Spahiu

Chalmers, Kemi- och bioteknik, Kärnkemi

Journal of Nuclear Materials

0022-3115 (ISSN)

Vol. 392 1 45-54

DOI

10.1016/j.jnucmat.2009.03.044

Mer information

Skapat

2017-10-07