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.

Magnesium potassium phosphate cements (MKPCs) are prepared by the acid-base reaction of dead burned magnesia (MgO) and monopotassium phosphate (KH2PO4). Low-pH cementitious materials such as MKPCs are currently of interest for the geological disposal of nuclear waste. MKPCs have advantages such as high early strength, high bonding strength, small drying shrinkage, low permeability, and high sulfate resistance. According to the results of previous studies, it is known that cesium, strontium, and cobalt are immobilized in the form of MgCsxK1−xPO4·6H2O, MgxSr1−xKPO4·6H2O, and Co3(PO4)2, respectively, in MKPCs. However, these results were predicted based on thermodynamic data, not directly observed precipitates to clearly show the evidence. Therefore, in this study, we directly analyzed the immobilized forms of Cs, Sr, and Co, respectively. CsNO3, Sr(NO3)2, and Co(NO3)2·6H2O powders (0.3 mol each) were mixed individually in each of the MKPC suspensions. The suspensions in which KH2PO4 was dissolved were pH 4.3 and the dissolution of MgO decreased the H+ concentration, raising the pH close to 11. The hydration products according to pH evolution in the MKPC suspensions were analyzed, and the change in the concentration of ions in the aqueous solution was also measured. An aqueous solution was obtained using a syringe filter (0.45 μm) to analyze the ion concentrations in the solution of the suspension. The collected solutions were diluted with nitric acid and analyzed using inductively coupled plasma mass spectrometry. To characterize the solid phases, the suspensions were obtained with a pipette at specific times and filtered under a vacuum in a Buchner funnel. Because the amounts of hydration products including Cs, Sr, and Co were small, it was not observed by XRD and TGA analysis, but their components could be analyzed by SEM-EDS. The final precipitate forms of Cs, Sr, and Co in the MKPC matrix are MgCsPO4·6H2O, SrHPO4, and Co3(PO4)2·8H2O, respectively.

The decommissioning of a nuclear power plant generates large amounts of radioactive waste, which is of several types. Radioactive concrete powder is classified as low-level waste, which can be disposed of in a landfill. However, its safe disposal in a landfill requires that it be immobilized by solidification using cement. Herein, a safety assessment on the disposal of solidified radioactive concrete powder waste in a conceptual landfill site is performed using RESRAD. Furthermore, sensitivity analyses of certain selected input parameters are conducted to investigate their impact on exposure doses. The exposure doses are estimated, and the relative impact of each pathway on them during the disposal of this waste is assessed. The results of this study can be used to obtain information for designing a landfill site for the safe disposal of low-level radioactive waste generated from the decommissioning of a nuclear power plant.

Silver tellurite glasses with melting temperature of approximately 700°C were developed to immobilize 129I wastes. Longterm dissolution tests in 0.1 M acetic acid and disposability assessment were conducted to evaluate sustainability of the glasses. Leaching rate of Te, Bi and I from the glasses decreased for up to 16 d, then remained stable afterwards. On the contrary, tens to tens of thousands of times more of Ag was leached in comparison to the other elements; additionally, Ag leached continuously for all 128 d of the test owing to the exchange of Ag+ and H+ ions between the glasses and solution. The I leached much lower than those of other elements even though it leached ~10 times more in 0.1 M acetic acid than in deionized water. Some TeO4 units in the glass network were transformed to TeO3 by ion exchange and hydrolysis. These silver tellurite glasses met all waste acceptance criteria for disposal in Korea.