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.

The fabrication of waste forms with high thermal and structural stability is an essential technology for the safe management and disposal of radioactive wastes. In particular, the thermal characteristics of waste forms containing high heat-generating nuclides such as Cs and Sr can be used for the optimized design of the waste form to secure its thermal safety, and they also provide basic design data for the safe management of canisters, storage systems, and repositories. The Korea Atomic Energy Research Institute is actively developing processes and equipment for fabricating various types of high-level wastes into a stable glass or ceramic waste form. In previous research related to the thermal analysis of the waste form, a relatively simple analysis was performed by using the analytic solution of the one-dimensional steady-state heat conduction equation considering the decay heat properties of the waste. As a specific application study, the optimized diameter of the cylindrical glass waste form was proposed by evaluating the centerline temperature of the waste form. In this study, we extended previous research by introducing a more complicated model, and the main results are summarized as follows. First, an analytical solution was derived by applying the temperaturedependent thermal conductivity expressed in the general form of polynomial function to the onedimensional heat conduction problem previously studied. Second, the two-dimensional axisymmetric steady-state heat conduction problem with a more realistic cylinder model with finite length was modeled and solved by using the finite element method via Matlab’s PDE (partial differential equation) toolbox. Third, thermal analysis was performed on the SrTiO3 waste form, selected as a stable form of strontium nuclide, using the developed analytical and numerical methods. The differences in the temperature distribution and computation time were evaluated through a comparative study of both solutions. Although the problem considered in this study could easily be solved by using commercial CFD software such as ANSYS or SolidWorks, a code-based program was developed to facilitate parametric design study in conjunction with optimization algorithms. The analysis results could be used to evaluate the thermal stability of waste form and to optimize the shape and size of the waste form in consideration of the design constraints of storage systems or repositories.

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.