Radioactive cesium is a heat generated and semi-volitile nuclide in spent nuclear fuel (SNF). It is released gasous phase by head-end treatment which is a pretreatment of pyroprocessing. One of the capturing methods of gasous radioactive cesium is using zeolite. After ion-exchanged zeolite, it is transformed to ceramic waste form which is durable ceramic structure by heat treatment. Various ceramic wasteforms for Cs immobilization have been researched such as cesium aluminosilicate (CsAlSi2O6), cesium zirconium phosphate (CsZr2(PO4)3), cesium titanate (CsxAlxTi8-xO16, Cs2TiNb6O18) and CsZr0.5W1.5O6. The cesium pollucite is composed to aluminosilicate framework and cesium ion incorporated in matrix materials lattices. Many researchers are reported that the pollucite have high chemical durability. In this study, the Cesium pollucite was fabricated using mixtures of aluminosilicate denoted Absorbent product (AP) and Cs2CO3 by calcination and pelletized by cold pressing. The characterization of fabricated pollucite powder and pellets was analyzed by XRD, TGA, SEM, SEMEDS and XRF. The chemical durability of pollucite powder was evaulated by PCT-A and ICP-MS and OES. Thus, the optimal pressure condition without breaking the pellets which is low Cs2O/AP ratio and pelletizing pressure was selected. The long-term leaching test was performed using MCC-1 method for 28 days with the fabricated pollucite pellets. The leachate of leaching test was allard groundwaster and Deionized water and replaced 5 contact periods which is 3 hours, 3 days, 7 days, 14 days and 28 days and analyzed by ICPMS. The leaching rate was shown two stages. The first stage was rapid and relatively large amount of nuclides were leached. The leaching rate was decreased in the second stage. The fractional release rate of this study was shown same trend. These results were similar to previous studies.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

Radioactive Cesium is fission products of spent nuclear fuelwith high heat generating nuclide, having a 30 years half-life. Particularly, it is important to make stable waste form because Cs-137 have high solubility and mobility at ground water. The ceramic waste form has higher thermal and structural stability and lower solubility than glass and cement waste form. Various ceramic waste forms for Cs immobilization have been researched such as aluminosilicate (CsAlSi2O6), phosphate (CsZr2(PO4)3), titanate (CsxAlxTi8-XO16) and CsZr0.4W1.5O6. Cs pollucite is incorporated radio-Cesium to aluminosilicate framework by inorganic ion-exchange with zeolite. Therefore, it is an extremely stable structure. In previous study, we are prepared Cs pollucite pellet with various ratio of Cs precursor/matrix materials, and attempted to evaluate applicability as ceramic waste form. Cs pollucite is produced by mixing Mullite and SiO2 obtained by heat treatment Kaolinite with Cs2CO3 in ratios of 0.5, 0.6, 0.7, 0.8. Optimized ratio was 0.5 revealed single pollucite phase and the others exhibited CsAlSiO4 phase with pollucite. Cs pollucite of ratio 0.5 was pelletized under various conditions and evaluated performance as waste form. herein, the pellets were cracked on surface and edges broken. Therefore, Cs pollucite having high ratio of matrix materials contained Si and Al was prepared and pelletized, and then waste form was evaluated. The Cs pollucite powder is ratio of Cs precursor/matrix materials were 0.1, 0.2, 0.3, 0.4. Pollucite powder was mixed with 1.5, 2.0wt% Polyvinyl alcohol as binder, and dried at 70°C for overnight. Afterward, these powders obtained were pressed using punch-die apparatus at 50, 100 bar for 1 hour and the pellets with about dia. 25 mm and height 10 mm was acquired. These pellets were sintered at 1,400°C for 5 hours. Subsequently, the waste forms were evaluated physicochemical test such as compression strength, thermal conductivity, thermal expansion and leaching properties analysis.

Garnet is one of the promising ceramic waste forms for immobilizing radioactive wastes. It has an A3 [VIII]B2 [VI]T3 [IV]O12 structure, so it can accommodate various cations of different sizes and coordination. Silicon usually occupies the centers of the tetrahedron structural site (T[IV]O4) in natural garnet. However, substitution of the T-site with iron, which has a relatively large ionic radius, causes the expansion of a unit cell volume of garnet and allows the incorporation of large cations such as actinides at other sites. Relatively few leaching data have been reported for ferrite garnet waste forms to date. In this study, we synthesized gadolinium-iron-garnet and evaluated the leaching property using cerium as a surrogate for actinide elements. The test specimens were made by cold pressing and sintering process. Three different standard leaching tests were performed as follows. The PCT-A (ASTM C1285) was performed for 7 days at 90°C to the crushed sample (0.149 to 0.074 mm). The ANSI/ANS-16.1 standard leach test was performed at ambient conditions for 5 days with constant replacement of leachate. Finally, the MCC-1 (ASTM C1220) test was performed for 28 days at 90°C with different types of leachants such as ultrapure water, brine, and silicate water. The last two leaching tests were conducted on monolithic specimens. After the end of the test, leachate was analyzed by inductively coupled plasma mass spectroscopy (Agilent, ICP-MS 7700S).

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.

An important property of glass and ceramic solid waste forms is processability. Tellurite materials with low melting temperatures and high halite solubilities have potential as solid waste forms. Crystalline TiTe3O8 was synthesized through a solid-state reaction between stoichiometric amounts of TiO2 and TeO2 powder. The resultant TiTe3O8 crystal had a three-dimensional (3D) structure consisting of TiO6 octahedra and asymmetric TeO4 seesaw moiety groups. The melting temperature of the TiTe3O8 powder was 820℃, and the constituent TeO2 began to evaporate selectively from TiTe3O8 above around 840℃. The leaching rate, as determined using the modified American Society of Testing and Materials static leach test method, of Ti in the TiTe3O8 crystal was less than the order of 10-4 g·m-2·d-1 at 90℃ for durations of 14 d over a pH range of 2-12. The chemical durability of the TiTe3O8 crystal, even under highly acidic and alkaline conditions, was comparable to that of other well-known Ti-based solid waste forms.