Currently, non-volatile nuclides such as 94Nb, 99Tc, 90Sr, 55Fe, and 59/63Ni are used a sequential separation. In this study, we developed a separation for 99Tc and 90Sr by a carbonate precipitation. Sodium Carbonate (Na2CO3) was inserted in the aqueous sample from a Dry Active Waste (DAW) and a carbonate precipitation was produced. The precipitate is composed of di- or tri-valent element such as Co, Sr, Fe, Ni and the supernatant is composed of mono-valent element (Cs) and anion materials (ReO4 -, TcO4 -). In DAW, it was confirmed that the recovery of 90Sr (precipitate) and 99Tc (supernatant) were > 90%, respectively. The precipitate and supernatant separated by using a Sr-resin and an anion-exchange resin, respectively. The separated samples were measured by a Liquide Scintillation Counter (LSC, 90Sr) and Induced-Coupled Plasma-Mass Spectroscopy (ICPMS, 99Tc).

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

Engineered barriers (concrete and grout) in Low- and Intermediate-Level Waste (L/ILW) disposal facilities tend to degrade by groundwater or rainfall water over a long period of time. During the degradation process, radionuclides stored in the disposal facility might be released into the pore water, which can pass through the natural rock barriers (granite and sedimentary rock) and may reach the near-field and far-field. In this transportation, radionuclide might be sorbed onto the engineered and natural rock barriers. In addition, the organic complexing agent such as ethylenediaminetetraacetic acid (EDTA) and α-isosaccharinic acid (ISA), is also present in pore water, which may affect the sorption and mobility of radionuclide. In this study, the sorption and mobility of 90Sr under different conditions such as two pHs (7 and 13), different initial concentrations of organic complexing agents (from 10-5 M to 10-2 M), and solutions (groundwater, pore water, and rainfall water) were investigated in a batch system. The groundwater was collected at the L/ILW disposal facility located at Gyeongju in South Korea. The pore water and rainfall water were artificially made in the laboratory. The concrete, grout, granite, and sedimentary rock samples were collected from the same study sites from where the groundwater was collected. The rock samples were crushed to 53-150 micrometers and were characterized by XRD, XRF, SEM-EDS, BET, and zeta potential analyzer. 90Sr concentration was determined using liquid scintillation counting. The sorption of 90Sr was described by distribution coefficients (Kd) and sorption reduction factor (SRF). In the case of EDTA, the Kd values of 90Sr remained constant from 10-5 M to 10-3 M and tended to decrease at 10-2 M, while in case of ISA the Kd values decreased steadily as the concentration of ISA was increased from 10-5 M to 10-3 M; However, a sudden reduction in the Kd values were observed above 10-2 M. In comparison to EDTA, ISA gave a higher SRF of 90Sr. Therefore, from the above results, it can be concluded that the presence of ISA has a greater effect on the sorption and mobility of radionuclide in the solutions than EDTA, and the radionuclide may reach near- and far-field of the L/ILW disposal facility.

Concrete is one of the largest wastes, by volume, generated during the decommissioning of nuclear facilities, which significantly influences the projected costs for the disposal of decommissioning wastes. Concrete consists of aggregates and a cement binder. In radioactive concrete, the radioisotopes are mainly associated with the cement component. If the radioactive isotope can be separated from the concrete to below the clearance criteria, the volume of radioactive concrete waste could be reduced effectively. We were studied to separate the radioactive materials from the concrete by using the thermomechanical and chemical treatment processes, sequentially. From the study, separated aggregate could be treated to achieve the clearance level. However, these processes generate a large volume of secondary acidic radioactive wastewater, which might be a critical problem to reduce the volume of radioactive concrete waste. In this research, separating the 137Cs and 90Sr from dissolved concrete wastewater to below the discharge criteria by precipitation method, it would be released to the environment under industrial waste guidelines. The experiments were conducted to using a simulated radioactive wastewater, formed by the dissolution of concrete within HCl, which was spiking the 137Cs and 90Sr, respectively. In addition, we applied the chemical precipitation methods with wastewater, using ferrocyanide for 137Cs and BaSO4 coprecipitation for 90Sr. As a result, targeted radionuclides could be removed to the discharge level (137Cs: 0.05 Bq·ml−1, 90Sr: 0.02 Bq·ml−1) by precipitation method. Therefore, it could reduce the secondary wastewater effectively by precipitation method and enhance the additional volume reduction for radioactive concrete waste.