Identification of the mechanism limiting the alteration of clad spent fuel segments in aerated carbonated groundwater
Leaching experiments were performed with five spent fuel samples (20 mm segments of clad fuel rods) from French power reactors (four UO 2 fuel samples with burnup ratings of 22, 37, 47 and 60 GW d t HM-1 and a MOX fuel sample irradiated to 47 GW d t HM-1) to determine the release kinetics of the matrix containing most of the radionuclides. The experiments were carried out with carbonated groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25 °C) in a hot cell. Until 313 days of leaching and below uranium saturation, the Sr/U congruence ratios for all the UO 2 fuel samples ranged from 1 to 2; allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO 2 fuel matrix. The daily strontium release factor was approximately 2.7 × 10 -8 d -1 for UO 2 fuel after 706 days of leaching, and seven to eight times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings. All the available experimental data (characterization of secondary phases and leaching data) indicate that the mechanism limiting the spent fuel alteration kinetics, for the conditions studied, is likely based on the transport and accessibility of oxidizing species and/or water within the segment.