Spent nuclear fuels are temporarily stored in nuclear power plant site. When a problem such as cracking of spent nuclear fuel assembly or cladding occurs or uranium that has not been separated during the reprocessing remains, it is necessary to treat it. The borosilicate glasses have been considered to vitrify whole spent nuclear fuel assembly. However, a large amount of Pb addition was necessary to oxidize metals in assembly to make them suitable for oxide glass vitrifcation. Furthermore, these borosilicate glasses need to be melted at high temperatures (> 1,400°C) when UO2 content is more than 20wt%. Iron phosphate glasses can be melted at a relatively low temperature (< 1,300°C) even with a similar UO2 addition. A composition of iron phosphate glass for immobilization of uranium oxide has been developed. The glasses have glass transition temperatures of ~555°C that are high enough to maintain its phase stability in geological repositories. The waste loading of UO2 in the glass is ~33.73wt%. Normalized elemental releases from the product consistency test were well below the US regulation of 2 g/m2. Nuclear criticality safety and heat generation in deep geological repositories were calculated using MCNP and computational fluid dynamics simulation, respectively. The glass had effective neutron multiplication factor (keff) of 0.755, which is smaller than the nuclear- criticality safety regulation of 0.95. Surface temperature of the disposal canister is expected to lower than the limit temperature (< 100°C). Most of the U in the glass is in the 4+state, which is more chemically durable than the 6+state. As a result of long-term dissolution experiment, chemically-durable uranium pyrophosphate (UP2O7) crystals were formed.