In order to assess the impact of a-radiolysis of water on the oxidative dissolution of spent fuel, an unirradiated, annealed MOX fuel pellet with high content of Pu (~24 wt%), and a specific a-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the a-decays emitted from the surface are expected to produce ~3.6 107 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1$10-6M after 1 h to ~7 10-5M after 58 days. Leaching under an atmosphere starting at 1MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the a-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ~3 10-8M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.