A preliminary consideration for underground autocatalytic criticality by vitrified high-level waste in water-saturated geologic repository

A preliminary assessment is conducted for the autocatalytic criticality phenomena for the Japanese repository concept, where the vitrified high-level waste (HLW) from reprocessed fuel of commercial light-water reactors is assumed to be disposed of in fractured granitic rock saturated with water. By a static neutronic analysis, the mass of U-235 in 12%-enriched uranium required to make the transition from an under-moderated to an over-moderated critical system in water-saturated granite of a porosity of 0.3 is obtained 35 kg in a spherical core with a radius of 90 cm. The radionuclide transport analysis indicates that in any case the only fissile nuclide of concern for the autocatalytic criticality scenario is 235U. Other fissile nuclides are too short-lived, or are accompanied with a strong neutron absorber. This preliminary study indicates that the transport model can be simplified for uranium isotopes, with which a statistical uncertainty analysis can be performed to finally determine whether or not the autocatalytic criticality event can be screened out from the scenarios for the generic performance assessment of the conceptual repository.