Various dry active wastes (DAWs) have been accumulated in nuclear power plants since the DAWs are mostly combustible. KAERI has developed a thermochemical treatment process for the used decontamination paper as an operational waste to substitute for incineration process and to decontaminate radionuclides from the DAWs. The thermochemical process is composed of thermal decomposition in a closed vessel, chlorination of carbonated DAWs, separation of soluble chlorides captured in water by hydroxide precipitation, and immobilization of the precipitate. This study examined the third and fourth steps in the process to immobilize Co-60 by fabricating a stable wasteform. Precipitation behaviors were investigated in the chloride solution by adding 10 M KOH. It was shown that the precipitates were composed of Mg(OH)2 and Al(OH)3. Then, the glass-ceramic wasteform for the precipitates were produced by adding additive mixtures in which silica and boron oxide were blended with various ratios. The wasteform was evaluated in terms of volume reduction ratio, bulk density, compressive strength, and leachability.

Carbon 14 (14C) is radioactive isotope of carbon which emits beta ray with long half-life (5730±30 years). Since the 14C is significantly hazardous for human being, the appropriate process to treat 14C is necessary. From the nuclear power plant, the ion exchange resin, graphite, and activated carbon are the main source of 14C. During the effort to reduce the volume of those wastes, the 14C is inevitably occurred as carbon dioxide (CO2) form, so called 14CO2. Thus, the development of technology to permanently capture and safely dispose 14CO2 is required. In this presentation, we introduce the decommissioning technology ranging from 14CO2 capture to solidification. First, the new class of glass adsorbent is developed which can irreversibly capture CO2 even under mild conditions. This material promotes the dissolution of alkaline earth ions due to the unstable glass structure. Then, the physical and chemical optimization of glass adsorbent enhances the performance of CO2 capture. Further, room temperature geopolymeric solidification is also performed to safely dispose 14C without any potential release.

Radionuclides in low- and intermediate-level radioactive wastes from the decommissioning process of nuclear power plants were generally immobilized by cementation methods. Ethylenediaminetetraacetic acid (EDTA), which is extensively used as a decontamination agent, can affect the behaviors of radionuclides immobilized in cement waste forms. In this study, the effects of EDTA contained in simulated radioactive decommissioning wastes on the leaching characteristics of immobilized Co and Cs and the microstructure evolution of cement waste form. Co leaching was accelerated by the formation of Co–EDTA complexes with high mobility and solubility. Cs leaching was hindered by the ion competition with other metal–EDTA complexes for releasing from the cement waste form. Cs leaching was also retarded by carbonated layer at edge of the cement waste form, which process is facilitated by the presence of EDTA. Finally, the effects of EDTA on the leaching characteristics of immobilized Cs and Co and the microstructure evolution of the cement waste form should be considered to ensure the safety of disposal for lowand intermediate-level radioactive wastes.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

Various disposal methods for spent nuclear fuels (SNFs) are being researched, and one of these methods involves separating high heat-generating nuclear isotopes such as Strontium-90 (90Sr) and Cesium-137 (137Cs) for deep disposal. These isotopes has relatively short half-lives and substantial decay energies. Especially, 90Sr undergoes decay through Yttrium-90 to Zirconium-90, emitting intense heat with beta radiation. Therefore, the removal of these high heat-generating isotopes will significantly contribute to reducing disposal site area. To remove 90Sr from SNFs, molten salt was utilized in KAERI. During this process, it was discovered that 90Sr dissolves in the molten salt in the form of SrCl2 and/or Sr4OCl6. Afterwards, it is crucial to recover 90Sr in the form of oxide from the salt to create immobilized forms for disposal. This can be achieved by reactive distillation with K2CO3. However, the amount of 90Sr within the SNFs is only 0.121wt%, and even if all the 90Sr in the SNFs were to leach into the molten salt, the quantity of 90Sr in the molten slat would still be very small. Therefore, adding K2CO3 to the molten salt for reactive distillation could result in significant possibilities of side reactions occurring. In this study, a two-step process was employed to mitigate the side reactions: the 1st step involves evaporating the all molten salts and the 2nd step includes adding K2CO3 to make oxides through solid-solid reaction. Eutectic LiCl-KCl, which is the most commonly used salt, was employed. The eutectic LiCl-KCl with SrCl2 was heated at 850°C for 2 h to evaporate the salts under a vacuum (> 0.02 torr). However, after examining the distillation product before the solid-solid reaction, it was observed that SrCl2 reacted with KCl in the salt, resulting in the formation of KSr2Cl5. It means that salts containing KCl are not suitable candidates for reactive distillation aimed at producing immobilized forms. As an alternative, MgCl2 could be a highly promising candidate because it is inert to SrCl2 and according to a recent study in KAERI, MgCl2 exhibited the most efficient separation of Sr among various salts. Therefore, we plan to proceed with the two-step reactive distillation using MgCl2 for the future work.

Sulfur and nitrogen co-doped carbon dots (NSCDs) were quickly synthesized by the microwave-assisted method from triammonium citrate and thiourea. NSCDs showed a quantum yield of 11.5% with excitation and emission bands at 355 and 432 nm, respectively. Also, a fluorescence quenching was observed in the presence of Pb(II) ions, and the as-synthesized CDs were used as a sensitive probe for detecting Pb(II) in water and food samples. The results showed the optimal conditions for Pb(II) determination were CDs concentration of 0.02 mg mL− 1 at pH 6.0–7.0 and an incubation time of 20 min. The relative fluorescence intensity of NSCDs was proportional to Pb(II) concentrations in the range of 0.029–2.40 and 2.40–14.4 μmol L− 1 with a correlation coefficient (R2) of 0.998 and 0.955, respectively, and a detection limit of 9.2 × 10– 3 μmol L− 1. Responses were highly repeatable, with a standard deviation below 3.5%. The suggested method demonstrates the potential of a green, fast, and low-cost approach for Pb(II) determination in water, tea, and rice samples with satisfying results.

Nitrogen and phosphorous dual-doped carbon nanotubes (N,P/CNT) have been grown in a single-step direct synthesis process by CVD method using iron-loaded mesoporous SBA-15 support, as an electrode material for the energy storage device. For comparison, pristine nanotubes, nitrogen and phosphorous individually doped nanotubes were also prepared. The basic characterization studies clarify the formation of nanotubes and the elemental mapping tells about the presence of the dopant. Under three-electrode investigations, N,P/CNT produced a maximum specific capacitance of about 358.2 F/g at 0.5 A/g current density. The electrochemical performance of N,P/CNT was further extended by fabricating as a symmetric supercapacitor device, which delivers 108.6 F/g of specific capacitance for 0.5 A/g with 15 Wh/kg energy density and 250 W/kg power density. The observed energy efficiency of the device was 92.3%. The capacitance retention and coulombic efficiency were 96.2% and 90.6%, respectively, calculated over 5000 charge–discharge cycles.

A novel kind of self-assembled graphene quantum dots-Co3O4 (GQDs-Co3O4) nanocomposite was successfully manufactured through a hydrothermal approach and used as an extremely effectual oxygen evolution reaction (OER) electrocatalyst. The characterization of morphology with scanning electron microscopy (SEM) and transmission electron microscopy (TEM) showed that Co3O4 nanosheets combined with graphene quantum dots (GQDs) had a new type of hexagonal lamellar selfassembly structure. The GQDs-Co3O4 electrocatalyst showed enhanced electrochemical catalytic properties in an alkaline solution. The start potential of the OER was 0.543 V (vs SCE) in 1 M KOH solution, and 0.577 V (vs SCE) in 0.1 M KOH solution correspondingly. The current density of 10 mA cm− 2 had been attained at the overpotential of 321 mV in 1 M KOH solution and 450 mV in 0.1 M KOH solution. Furthermore, the current density can reach 171 mA cm− 2 in 1 M KOH solution and 21.4 mA cm− 2 in 0.1 M KOH solution at 0.8 V. Moreover, the GQDs-Co3O4 nanocomposite also maintained an ideal constancy in an alkaline solution with only a small deterioration of the activity (7%) compared with the original value after repeating potential cycling for 1000 cycles.