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.

In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

During the treatment of spent nuclear fuel, radioactive iodine is generated in a liquefied or gaseous form in a specific process. In the case of iodine 129, it is a long-lived nuclide with a very long halflife and has high groundwater mobility under repository conditions. Despite showing a low radioactivity value, research on the management of radioactive iodine from a long-term perspective is continuously being performed. Although research has been conducted using borosilicate glass as a medium for solidifying iodine, compatibility of I in borosilicate glass is very small and the volatility is high in the solidification process. So it is not suitable as a solidified substance of iodine. Therefore, studies on other solidification media to replace them are continuously being conducted. Our research team tried to develop a new medium that can contain iodine in a solidified body stably through a simple heat treatment process and can improve problems such as volatility and waste loading. Iodine is captured as AgI in the Ag ion-exchanged zeolite. So, TeO2, Ag2O, and Bi2O3 having a high AgI loading rate were used as main components. It was named TAB after taking the first letter of each element. In previous studies, the physical properties, structure, and chemical stability of TAB materials were confirmed. PCT (Product Consistent test) was performed to confirm chemical stability. It is mainly used to compare the chemical stability of glass materials with other glass materials, but there are limitations in evaluating the long-term chemical stability of materials. In this experiment, we tried to evaluate the long-term stability of TAB and compare it with borosilicate, which is conventionally used to treat radioactive waste. In addition, we tried to understand the leaching behavior inside the TAB medium. For this purpose, ASTM C1308 test was performed for 365 days, and distilled water and KURT groundwater were used as leachates to examine the effect of ions in the groundwater on the solidified body. To analyze the leaching behavior, ICP-MS and ICP-OES analyses were performed, and the cross-section of the sample after leaching was observed through SEM.

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.