Potential methods to reduce the thermal impact of vitrified waste from spent mixed oxide fuel on geological repositories

Abstract

Compared to the vitrified, high-level radioactive waste (HLW) from spent UO2 fuel (UO2-HLW glass), that of mixed oxide (MOX) fuel (MOX-HLW glass) exhibits high heat generation over a long period of time. Thus, to reduce the associated waste volume and repository footprint, it is essential to prevent bentonite buffer alteration by mitigating heat generation. In this study, we evaluated five potential methods of decreasing the bentonite buffer temperature in a geological repository for MOX-HLW glass. Of these five methods, the separation of minor actinides (MA) from high-level liquid waste (HLLW) was the most effective. This is because the main nuclide involved in heat generation in MOX-HLW glass is Am-241. An MA separation rate of 81% was required to maintain the bentonite buffer temperature below the Japanese upper disposal limit of 100 °C. Thus, it is important to measure the amount of Am-241 when disposing of MOX-HLW glass. Assuming that only Am-241 was disposed of, the maximum Am-241 waste loading under a bentonite buffer temperature of 100 °C was 0.33 wt%, equating to the disposal of only 1.32 kg of Am-241 per 400 kg of MOX-HLW glass. Thus, there is a need for new methods of treating and disposing of Am-241.Graphical abstractRelationship between bentonite buffer temperature and (a) minor actinide (MA) separation ratio from high-level liquid waste (HLLW), (b) mixed oxide (MOX) high-level radioactive waste (HLW) glass and UO2 HLLW mixing ratio, (c) waste loading in MOX-HLW glass, (d) MOX-HLW glass storage time, and (e) canister footprint, representing five methods of mitigating heat generation from disposed MOX-HLW glass.