The Methodologies to Reduce the Thermal Impact of Vitrified Waste from Spent MOX Fuel on Geological Repository

In Japan, the storage of spent MOX fuels discharged from some commercial nuclear power plants has started. In the future, reprocessing, vitrification and final disposal of the spent MOX fuel will be needed in order to reduction of waste volume and radiotoxicity in a potential repository.<br/>Compared to the vitrified waste from spent UO₂ fuels (UO₂-HLW glass), the MOX-HLW glass has high heat generation for a long period, and thus it takes about 300 years to cool down to the same heat generation. So, it is essential to decrease heat generation in order to prevent the bentonite buffer alteration, otherwise the waste volume and repository footprint will be huge.<br/><br/>In this study, 5 options are considered to decrease the bentonite buffer temperature in geological repository for the MOX-HLW glass. One option is minor actinides (MA) separation from high-level liquid waste (HLLW). MA separation is greatly effective to decrease thermal impact because main nuclide of heat generation of MOX-HLW glass is Am-241. MA separation ratio from HLLW requires 80% to lower the bentonite buffer temperature limit of 100 ° C. So, the development of MA separation technology is important.<br/><br/>The other options such as long-term storage or low waste loading of MOX-HLW glass, expansion of repository footprint, and mixing with UO₂ and MOX reprocessing liquid waste will be discussed in detail.<br/><br/>This work was carried out as a part of the basic research programs of vitrification technology for waste volume reduction supported (JPJ010599) by the Ministry of Economy, Trade and Industry, Japan.