This study aimed to develop a solid self-nanoemulsifying drug delivery system (solid-SNEDDS) to enhance the formulation of ketoconazole (KTZ), a BCS Class II drug with poor solubility. Ketoconazole, which is insoluble above pH 3, requires solubilization for effective delivery. This SNEDDS comprises oil, surfactant, and co-surfactant, which spontaneously emulsify in the gastrointestinal tract environment to form nanoemulsions with droplet sizes less than 100 nm. The optimal SNE-vehicle composition of oleic acid, TPGS, and PEG 400 at a 10:80:10 weight ratio was determined based on the smallest droplet size achieved. This composition was used to prepare liquid SNEDDS containing ketoconazole. The droplet size and polydispersity index (PDI) of the resulting liquid SNEDDS were analyzed. Subsequently, solid-SNEDDS was fabricated using a spray-drying method with solidifying carriers such as silicon dioxide, crospovidone, and magnesium alumetasilicate. The physicochemical properties of the solid-SNEDDS were characterized by scanning electron microscopy and powder X-ray diffraction, and its solubility, droplet size, and PDI were evaluated. In particular, the solid-SNEDDS containing ketoconazole and crospovidone in a 2:1 weight ratio exhibited significantly enhanced solubility, highlighting its potential for improved medication adherence and dissolution rates.

The need for the development of sustainable, efficient, and green radioactive waste disposal methods is emerging with the saturation of spent nuclear waste storage facilities in the Republic of Korea. Conventional radioactive waste management methods like using cement or glass have drawbacks such as high porosity, less chemical stability, high energy consumption, carbon dioxide production, and the generation of secondary wastes, etc. To address this gigantic issue of the planet, we have designed a study to explore the potential of alternative materials having easy processability, low carbon emissions and more chemical stability such as ceramic (hydroxyapatite, HAP) and alkali-activated materials (geopolymers, GP) to capture the simulated radioactive cobalt ions from the contaminated water and directly solidify them at low temperatures. Physical and mechanical properties of HAP alone and 15wt% GP incorporated HAP (HAP-GP- 15) composite were studied and compared. The surface of both materials was fully sorbed with an excess amount of Co(II) ions in the aqueous system. Co(II) sorbed powders were separated from aqueous media using a centrifuge machine operating at 5,000 RPM for 10 minutes and dried at 100°C for 8 hours. The dried powders were then placed in stainless steel molds, and shaped into cylindrical pellets using a uniaxial press at a pressure of 1 metric ton for 1 minute. The pellets were sintered at 1,100°C for 2 hours at a heating rate of 10°C/min. Following this, the water absorption, density, porosity, and compressive strength of the polished pellets were measured using standard methods. Results showed that HAP has a greater potential for decontamination and solidification of Co(II) due to its higher density (2.65 g/cm3 > 1.90 g/cm3), less open porosity (16.2±2.9% < 42.4 ±0.9%) and high compressive strength (82.1±10.2 MPa > 6.9±0.8 MPa) values at 1,100°C compared to that of HAP-GP-15. Nevertheless, further study with different constituent ratio of HAP and GP at various temperatures is required to fully optimize the HAP-GP matrix for waste solidifications.

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.

Various types of solidifying materials are used to stabilize and solidify low and intermediatelevel radioactive dispersible waste. Portland cement is generally used to solidify various radioactive wastes because its facilities and processes are simple, less dangerous, and it has excellent compressive strength after curing compared to other materials. However, it is difficult to use Portland cement in radioactive waste containing highly water-soluble harmful substances such as sodium fluoride because it is prone to leaching harmful ingredients in immersion tests due to its low water resistance. In this study, solidification was achieved using an organic-inorganic hybrid solidifying binders consisting of inorganic binders such as Portland cement, blast furnace slag powder, silica fume, and organic binders such as epoxy resin. This material was then compared with a solidification material made of Portland cement alone. The mixing ratio of inorganic binders, water, and organic binders to simulated waste is 35%, 20%, and 25%, respectively. The mixing ratio of inorganic binders and water when using only Portland cement for simulated waste is 100% and 80%, respectively. The mixed paste was poured into a cylinder mold (Φ 5 × 10 cm) to seal the upper part, cured at room temperature for 28 days to produce a solidification specimen, and then subjected to various tests were performed, including compressive strength, immersion compressive strength, hydration peak temperature, length change, and immersion weight change. The compressive strength of the organic-inorganic hybrid solidification test was 13-17 MPa, the immersion compressive strength was 15-18 MPa, the hydration peak temperature was 33-36°C, the length change rate was -0.086%, and the immersion weight change rate was –2.359%. The compressive strength of the Inorganic solidification test using only Portland cement was 16-18 MPa, the immersion compressive strength was 20-21 MPa, the hydration peak temperature was 23-25°C, the length change rate was -0.150%, and the immersion weight change rate was -5.213%. The compressive strength and immersion compressive strength of the organic-inorganic hybrid solidification materials were slightly lower compared to those of Portland cement solidification materials, they still met the compressive strength standard of 7-12 MPa, taking into consideration the strength reduce and economic feasibility of the core drill process. Furthermore, it indicates that the rates of change in length and immersion weight decreased to about 1% and 5%, suggesting an improvement in water resistance. The above results suggest that applying the organic-inorganic hybrid solidification method to radioactive waste treatment can effectively improve water resistance and help secure long-term stability.

Radioactive liquid waste generated during the operation of domestic nuclear power plants is treated through a somewhat different liquid radwaste system (LRS) for each plant. Prior to the introduction of standard nuclear power plants, LRS used a concentrated water dry system (CWDS) to evaporate liquid waste and manage it in the form of dry powder. The boron-containing radioactive liquid waste dry powder was solidified using paraffin from 1995 to 2010, and about 3,650 drums (based on 200 L) of paraffin solidified drums are currently stored in nuclear power plants. Paraffin solidification drums do not meet the acceptance criteria for radioactive waste repositories because it is difficult to secure the homogeneity of the solidified body and there are concerns about leaching of radioactive waste due to the low melting point of paraffin. In order to solve this problem and safely permanently dispose of paraffin solidification drums, the characteristics of dry powder paraffin solidification drums containing boron-containing radioactive liquid waste must be analyzed and appropriate treatment technology utilizing the results must be introduced. This study analyzes the physical properties of paraffin, the chemical properties of boron-containing radioactive waste dry powder, and the physicochemical properties of paraffin solidification powder, and proposes an appropriate alternative technology for treating boron-containing radioactive waste dry drum. When disposing of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder, the solidification body must be effectively withdrawn from the drum and the paraffin must be completely separated from the solidification body. When disposing the drum, the solidified material must be effectively extracted from the drum and the paraffin must be completely separated from the solidified material. Afterwards, the paraffin must be self-disposed, and the radioactive waste must be disposed of in accordance with acceptance criteria of repository. We looked at how each characteristic of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder can be utilized in each of the above treatment processes.

The development of existing radioactive waste (RI waste) management technologies has been limited to processing techniques for volume reduction. However, this approach has limitations as it does not address issues that compromise the safety of RI waste management, such as the leakage of radioactive liquid, radiation exposure, fire hazards, and off-gas generation. RI waste comes in various forms of radioactive contamination levels, and the sources of waste generation are not fixed, making it challenging to apply conventional decommissioning and disposal techniques from nuclear power plants. This necessitates the development of new disposal facilities suitable for domestic use. Various methods have been considered for the solidification of RI waste, including cement solidification, paraffin solidification, and polymer solidification. Among these, the polymer solidification method is currently regarded as the most suitable material for RI waste immobilization, aiming to overcome the limitations of cement and paraffin solidification methods. Therefore, in this study, a conceptual design for a solidification system using polymer solidification was developed. Taking into account industrial applicability and process costs, a solidification system using epoxy resin was designed. The developed solidification system consists of a pre-treatment system (fine crush), solidification system, cladding system, and packing system. Each process is automated to enhance safety by minimizing user exposure to radioactive waste. The cladding system was designed to minimize defects in the solidified material. Based on the proposed conceptual design in this paper, we plan to proceed with the specific design phase and manufacture performance testing equipment based on the basic design.

The sustainability of the nuclear power industry hinges increasingly on the safe, long-term disposal of radioactive waste. Despite significant innovations and advancements in nuclear fuel and reactor design, the quest for a permanent solution to handle accumulating radioactive waste has received comparatively less attention. Conventionally, two widely recognized solidification methods, namely cementation for low and intermediate-level waste and vitrification for high-level waste, have been favored due to their simplicity, affordability, and availability. Recently, geopolymers have emerged as an appealing alternative, gaining attention for their minimal carbon footprint, robust chemical and mechanical properties, cost-effectiveness, and capacity to immobilize a broad spectrum of radionuclides, including radioactive organic compounds. This study delves into the synthesis of metakaolin-based geopolymers tailored for the immobilization of fission products like cesium (Cs) and molybdenum (Mo). The investigation unfolded in two key steps. In the initial step, we optimized the alkali content to prevent the occurrence of efflorescence, a potential issue. Remarkably, as the Na2O/Al2O3 ratio increased from 0.82 to 1.54, we observed significant enhancements in both compressive strength (11.45 to 27.07 MPa) and density (up to 2.23 g/cm3). This suggests the importance of careful adjustment in achieving the desired geopolymer characteristics. The second phase involved the incorporation of 2wt% of Cs and Mo, both individually and as a mixture, into the geopolymer matrix. We prepared the GP paste, which was poured into cylindrical molds and cured at 60°C for one week. To scrutinize the crystallinity, phase purity, and bonding type of the developed matrix, we employed XRD and FTIR techniques. Additionally, we conducted standard compressive strength tests (following ASTM C39/C39M-17b) to assess the stacking durability and robustness of the developed waste form, vital for storage, handling, transportation, and disposal in a deep geological repository. Furthermore, to evaluate the chemical durability, diffusivity and leaching of the GP waste matrix, we employed the ASTM standard Product Consistency Test (PCT: C 1285-02) and American nuclear society’s devised leaching test (ANS 16.1). It is noteworthy that the introduction of Cs and Cs/Mo in the GP matrix led to a reduction of more than 50% and 60% in compressive strength, respectively. This outcome may be attributed to the interference of Cs and Mo with the geopolymerization process, potentially causing the formation of new phases. However, it is crucial to emphasize that both developed matrices exhibited an acceptable normalized leaching rate of less than 10-5 g·m-2·d-1. This finding underscores the promising potential of the GP matrix for the immobilization of cationic and anionic radioactive species, paving the way for more sustainable nuclear waste management practices.

Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

The soils contaminated with radionuclides such as Cs-137 and Sr-90 should be solidified using a binder matrix, because radioactively contaminated soils pose environmental concerns and human health problems. Ordinary Portland cement has been widely used to solidify various radioactive wastes due to its low cost and simple process. In this study, simulant soil waste was solidified using cement waste form. The soils were collected around ‘Kori Nuclear Power Plant Unit 1’ and they were contaminated with the prepared simulant liquid waste containing Fe, Cr, Cs, Ni, Co, and Mn. The water-to-dry ingredients (W/D) ratio of cement waste form was 0.40. The cement paste was poured into a cubic mold (5×5×5 cm) and then cured for 28 days at room temperature. The 28-day compressive strength, water immersion, and EPA1311-toxicity characteristic leaching procedure (TCLP) tests were performed to evaluate the structural stability of cement waste form. The compressive strength was not proportional to soil waste loading, and the lowest compressive strength (4±0.1 MPa) was achieved in cement waste form containing 50wt% soil waste. After the water immersion test for 90 days, the compressive strength of cement waste form with 50wt% soil waste increased to 7.5±0.6 MPa, meeting the waste form acceptance criteria in the repository. It is believed that long-term water immersion test contributed to the additional curing and hydration reaction, resulting in the enhanced compressive strength. As a result of the TCLP test, the released amount of As, Ba, Cd, Cr, Pb, Se, Co, Cs, and Sr was less than the domestic and international standards. These results imply that cement waste form can be a promising candidate for the solidification of radioactive soil wastes.

A lot of solid wastes are generated when nuclear power plant is dismantled, and a lot of treatment costs and optimal waste treatment technologies are required to treat the generated solid wastes. Currently, there is no optimized reduction and solidification technology for each characteristics of radioactive dismantling waste, so the customized treatment technology for each waste is required to respond actively to this issue. This paper shows the evaluation results of molding and sintering characteristics using preliminary sample to derive operational characteristics and improvements for powder mixing device, molding device, and sintering device manufactured for solidification of dispersible radioactive waste. Zeolite was used as a preliminary sample for performing basic operation characteristics evaluation of each unit device. First of all, the basic operation characteristics of the powder mixing device was evaluated by analyzing the sample distribution, mixing degree, and tap density. It was confirmed that the preliminary sample was well mixed in all areas of the cylinder where the mixing was performed. In the tap density analysis, the increase effect of the volume reduction of the sample was confirmed according to the increase of the RPM speed (up to 2000 RPM). Since the particle size of zeolite sample is very small (nanometer size), the particular consistency of the change of average particle size with RPM speed couldn’t be confirmed, but the uniform of particle size distribution was confirmed with RPM speed size. The basic operation characteristics of the molding device was evaluated for each mold size (ID30, ID50, ID100) according to the moisture content (0-20%) and the molding pressure condition (25-200 MPa) for the preliminary sample. In the characteristics evaluation of the sintered body, the strength of the sintered body was much higher than that of the molded body. However, it was confirmed that as moisture evaporated during the sintering process according to the moisture content contained in the molded body, the swelling occurred in the sintered body due to vapor pressure, and this caused cracks in the longitudinal or transverse direction inside and outside the sintered body. Therefore, optimal moisture content conditions for sintering should be derived. In conclusion, if the operation characteristics and improvements of powder mixing, molding and sintering devices derived from this study are reflected and improved, it is judged that it is possible to derive the optimal process for solidification of dispersive radioactive wastes.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

Present study investigated the waste form integrity of melted products generated from PAM-MSO system, which is proposed and developed to compensate the drawbacks of each system. The disposal suitability of the melting solidification products generated from the plasma arc melting treatment of pulverized cement debris spiked by Pb, Cd and Cs, as indicators of typical hazardous metals and radionuclides existed in the low-level mixed waste in the KHNPPs. The final waste form obtained by the test was evaluated for suitability for disposal. The compressive strength was 261.10 MPa, showing much higher values when compared to other waste form products. The compressive strength of both the sample after irradiation with 107 Gy radiation and that after long-term submersion test (90 days) satisfied the disposal criteria. As a result of the leaching test conducted according to the ANS 16.1 test method, it was confirmed that the leaching index satisfies the disposal criteria.