In this experimental work, a p-type c-Si (100) substrate with 8 × 8 × 2 mm dimension was taken for TiCN thin-film coating deposition. The whole deposition process was carried out by chemical vapor deposition (CVD) process. The Si substrate was placed within the CVD chamber at base pressure and process pressure of 0.75 and 500 mTorr, respectively, in the presence of TiO2 (99.99% pure) and C (99.99% pure) powder mixture. Later on, quantity of C powder was varied for different set experiments. The deposition of TiCN coating was carried out in the presence of N2– H2–TiCl4–CH3CN gas mixture and 600 ℃ of fixed temperature. The time for deposition was fixed for 90 min with 10 and 5 ℃ min− 1 heating and cooling rate, respectively. Later on, heat treatment process was carried out over these deposited TiCN samples to investigate the changing characteristics. The heat treatment was carried out at 800 ℃ within the CVD chamber in the absence of any gas flow rate. The morphological properties of heat-treated samples have been improved significantly, evidence is observed from SEM and AFM analyses. The structural analysis by XRD has been suggested, upgradation in crystallinity of the heat-treated film as it possessed with sharp and higher intensity peaks. Evidence has been found that the electrochemical properties are enhanced for heat-treated sample. Raman spectroscopy shows that the intensity of acoustic phonon modes predominates the optic phonon modes for untreated samples, whereas for heat-treated samples, opposite trends have been observed. However, significant degradation in mechanical properties for heat-treated sample has been observed compared to untreated sample.

Large amounts of waste and wastewater from aquaculture have negatively impacted ecosystems. Among them, shrimp aquaculture wastewater contains large amounts of nitrogen contaminants derived from feed residues in an aerobic environment. This study isolated candidate strains from adult rockworms to treat shrimp aquaculture wastewater (SAW) in an aerobic environment. Among 87 strains isolated, 25 grew well at the same temperature as the shrimp aquaculture with excellent polymer degradation ability (>0.5 cm clear zone). Six isolates (strains AL1, AL4, AL5, AL6, LA10, and PR15) were finally selected after combining strains with excellent polymer degradation ability without antagonism. 16S rRNA sequencing analysis revealed that strains AL1, AL4, AL5, AL6, LA10, and PR15 were closely related to Bacillus paramycoides, Bacillus pumilus, Stenotrophomonas rhizophila, Bacillus paranthracis, Bacillus paranthracis, and Micrococcus luteus, respectively. When these six isolates were applied to SAW, they reached a maximum cell viability of 2.06×105 CFU mL-1. Their chemical oxygen demand (CODCr) and total nitrogen (TN) removal rates for 12 h were 51.0% and 44.6%, respectively, when the CODCr/TN ratio was approximately 10.0. Considering these removal rates achieved in this study under batch conditions, these six isolates could be used for aerobic denitrification. Consequently, these six isolates from rockworms are good candidates that can be applied to the field of aquaculture wastewater treatment.

A 11-year-old sexually intact female dog was presented because of a misshapen left elbow with a large mass. The masses on the elbow were diagnosed as high-grade canine mast cell tumor (cMCT) based on the cytological examination using fine-needle aspiration. Cytologically, the mass revealed moderate numbers of eosinophils, anisocytosis, anisokaryosis, and aberrant mitotic figures. The degree of cytologic dysplasia is diagnosed as moderate to severe. The blood analysis revealed low hematocrit and electrolytes. After examining the metastasis and surgical scope with radiographs and ultrasound, surgery was performed to removal the mass. Histopathologically, the mass revealed partially cytoplasmic granules, anisokaryosis of the nucleus, and mitotic counts. Scattered inflammatory cells and necrotic tissues were observed in the tumor. In this study, we found high-grade cMCT and treated the patient through a surgical approach.

To improve the quality of jujube (Zizyphus jujuba Miller var. hoonensis), which is a fruit of health functional, the effect of polyphenol preparation treatment on the fruit characteristics of two cultivars (cv. Bokjo and cv. Sangwang) of jujube was investigated. There was no difference in the height and breast diameter of jujubes tested between the polyphenol treatments and non treatment. Jujube trees treated with polyphenol preparation produced significantly more fruit than untreated in both cultivars. In cvultivar of Bokjo, the polyphenol preparation treatment increased the fruit's fresh weight and dry weight more than two times, respectively, compared to the untreated treatment. Polyphenol preparation tr eatments also changed the leaf characteristics of jujube trees. In the polyphenol-treated trees, leaf thickness tended to be thickest at the top and thinnest at the bottom. Polyphenol preparation treated jujube trees showed no difference in chlorophyll content. Moisture content was slightly higher in the untreatment than in the treatments. Visually, the polyphenol preparation treatment had a dark green color. Jujubes treated with polyphenol preparations showed differences in polyphenol content in fruits. The polyphenol content in both peel and flesh of the treatments were much higher than that of the untreatment. Reducing sugar was contained more in the peel than in the flesh and was higher in the untreatment than in the polyphenol preparation treatments. Treatment with polyphenol preparation showed differences in fruit appearance. As described above, it was found that the treatment of polyphenol preparation changed the leaves, fruit shapes and components of jujube trees. In particular, jujubes treated with polyphenol preparations are expected to contribute significantly to eco-friendly and highly functional jujube cultivation, as they appear to produce many fruits and increase the content of polyphenols and sugars.

Glutamine has been used to treat canine patients with parvoviral enteritis. However, little is known about the effect of L-alanyl-L-glutamine (Ala-Gln) supplementation in dogs with parvoviral enteritis. The objective of this study was to determine whether Ala-Gln supplementation can improve dog survival and ameliorate clinical signs without adverse effects. We conducted a randomized, double-blind, placebo-controlled clinical trial involving 39 client-owned dogs. The dogs were randomly assigned to two groups and administered either an Ala-Gln solution (Dipeptiven, 0.4 g/kg, n = 20) or an equivalent volume of placebo (n = 19) orally twice daily. Of the 39 dogs, 17 were vaccinated (n = 9 in the Ala-Gln-treated group and n – 9 in the placebo group). All dogs received standard treatment while hospitalized. The dogs were monitored according to a clinical scoring system and evaluated diagnostically daily for 11 days. Survival rates in both groups were quantified using Kaplan‒Meier survival curves and statistically compared using the log-rank test. The total score for clinical signs did not differ between the groups, except on day 2. The survival rates differed significantly (p=0.038). Three Ala-Gln-treated dogs (15.0%) died during the study, whereas eight dogs in the placebo group died (42.1%). No adverse effects were found to be associated with Ala-Gln treatment. Oral administration of Ala-Gln improves survival in dogs with parvoviral enteritis without causing adverse effects.

Titanium, which has excellent strength and toughness characteristics, is increasingly used in the aerospace field. Among the titanium alloys used for body parts, more than 80 % are Ti-6Al-4V alloys with a tensile strength of 931 MPa. The spark plasma sintering (SPS) method is used for solidification molding of powder manufactured by the mechanical milling (MM) method, by sintering at low temperature for a short time. This sintering method avoids coarsening of the fine crystal grains or dispersed particles of the MM powder. To improve the mechanical properties of pure titanium without adding alloying elements, stearic acid was added to pure titanium powder as a process control agent (PCA), and MM treatment was performed. The properties of the MM powder and SPS material produced by solidifying the powder were investigated by hardness measurement, X-ray diffraction, density measurement and structure observation. The processing deformation of the pure titanium powder depends on the amount of stearic acid added and the MM treatment time. TiN was also generated in powder treated by MM 8 h with 0.50 g of added stearic acid, and the hardness of the powder was higher than that of Ti-6Al-4V alloy when treated with MM for 8 h. When the MM-treated powder was solidified in the SPS equipment, TiC was formed by the solid phase reaction. The SPS material prepared as a powder treated with MM 8 h by adding 0.50 g of stearic acid also formed TiN and exhibited the highest hardness of Hv1253.

본 연구는 국내 산림생명자원 중 광나무의 대량생산을 위한 기초연구로, Priming 처리에 따른 발아 효율성 및 유묘 생육특성에 미치는 영향을 구명하고자 수행하였다. 종자 Priming처리는 대조구, GA3 (10, 100, 200 ㎎·L-1), Ca (NO3)2 (50, 100, 200 mM), KNO3 (50, 100, 200 mM)를 24시간 처리하여 발아특성을 분석하였고, 순화재배에 따른 생육특성 및 활착률을 조사하여 수확량을 비교·분석하였다. 발아특성은 25℃·KNO3 100 mM에서 발아율이 유의적으로(p<0.001) 높았으며, 평균발아일수는 10.3~18.1일로 15℃에서 대체적으로 빠른 발아일수를 나타냈다(p<0.001). 발아속도 및 발아균일지수 또한 25℃·KNO3 100 mM에서 유의적으로 높은 값을 보였다(p<0.001). 생육특성의 경우 2 5℃·KNO3 100 mM 처리에서 유근 길이(5.1±2.4 ㎝), 초장(5.7±0.7 ㎝), 근장(16.6±2.0 ㎝), 건중량(0.079 g)이 유의적으로 가장 높게 나타났다 (p<0.001). 유묘활력지수 또한 25℃·KNO3 100 mM 처리에서 1418.3으로 높았으며, 가장 높은 수확량(16.8 g/㎡)을 나타냈다. 결과적으로 25℃, KNO3 100 mM 처리 시 실내발아에서 발아효율성을 높일 수 있었으며, 순화재배 시 우수한 유묘 생육을 보여 수확량을 증진시킬 수 있었다.

Molten salt is one of the promising medium materials for molten salt reactors and energy storage systems. Molten salt is advantageous for better physical properties such as low melting point and high boiling point, high energy capacity, high thermal conductivity, and high thermal stability than other medium materials such as water or liquid metals. However, the corrosivity of the molten salt is one of the main factors that disturbs the various applications of the molten salt. On the other hand, metallic 3-D printing technologies have developed by leaps and bounds over the past 20 years and show potential for use in cutting-edge industries such as aerospace and military purposes. However, the biggest problem of 3-D printed products is that the mechanical and physical properties are very weak along the laminated plane that was generated during the manufacturing process. In particular, other research showed that corrosion is vulnerable through the laminated surface, and corrosion along the laminated plane is not completely mitigated through a general heat treatment process although the microstructure of the surface is evaluated to be partially mitigated by the heat treatment. In this study, molten salt corrosion behaviors of simple Ni-based alloy with a composition of 80Ni- 20Cr were analyzed. Ni-based alloys were fabricated by casting and 3-D printing, and some of the 3-D printed specimens were thermally treated at 1,273 K for 1 hour to examine the effects of heat treatment on corrosion behaviors. In molten eutectic NaCl-MgCl2 melts at 973 K, Ni-based alloys were corroded for 1, 3, 7, and 28 days and their microstructural changes were analyzed by SEM-EBSD-EDS and OM. The corrosion behaviors of the alloy were also evaluated by the salt composition measured with ICPOES. 3-D printed alloy with post-treatment showed more resistivity to the molten salt corrosion than as-fabricated 3-D printed alloy. However, the corrosion rate of the 3-D printed specimen after heat treatment was still higher than that made by casting.

Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

In this study, four technologies were selected to treat river water, lake water, and groundwater that may be contaminated by tritium contaminated water and tritium outflow from nuclear power plants, performance evaluation was performed with a lab-scale device, and then a pilot-scale hybrid removal facility was designed. In the case of hybrid removal facilities, it consists of a pretreatment unit, a main treatment unit, and a post-treatment unit. After removing some ionic, particulate pollutants and tritium from the pretreatment unit consisting of UF, RO, EDI, and CDI, pure water (2 μS/cm) tritium contaminated water is sent to the main treatment process. In this treatment process, which is operated by combining four single process technologies using an inorganic adsorbent, a zeolite membrane, an electrochemical module and aluminumsupported ion exchange resin, the concentration of tritium can be reduced. At this time, the tritium treatment efficiency of this treatment process can be increased by improving the operation order of four single processes and the performance of inorganic adsorbents, zeolite membrane, electrochemical modules, and aluminum- supported ion exchange resins used in a single process. Therefore, in this study, as part of a study to increase the processing efficiency of the main treatment facility, the tritium removal efficiency according to the type of inorganic adsorbent was compared, and considerations were considered when operating the complex process.

Radioactive carbon dioxide (14CO2) capture using innovative materials is desirable due to associated radiological hazards, and growing climate change. Mineral carbonation technology (MCT) is amenable to irreversibly capture CO2. Typically, MCT is attractive because capturing carbon through the chemical reaction between alkaline earth metal ions and CO2 forms insoluble and significantly stable carbonates. However, most applications of MCT have an intrinsic restriction regarding their operational conditions since no forward reaction occurs within realistic time scales. Thereby, the CO2 capture performance, such as CO2 capacity and carbonation reaction rate, of MCTs and their applications are severely restricted by the difficulty of operations under mild conditions. For example, natural minerals require aggressive carbonation reaction conditions e.g. high pressure (≥ 20 bar), high temperature (> 373 K), and pH-adjusted carrier solutions. To overcome such obstacles, the fabrication of alkaline earth oxides impregnated into an amorphous glass structure have been recently developed. They show enhanced rates of dissolution of alkaline earth metal ions and carbonation reaction due to the loosely packed glass structure and the generation of a surface coating silica gel, consequently facilitating CO2 capture under mild conditions. In this presentation, we report the synthesis and application of a crystallized glass tailored by controlled heat treatment for CO2 capture under mild conditions. The controlled heat treatment of an alkaline earth oxide-containing glass gives rise to a structural transformation from amorphous to crystalline. The structural characterizations and CO2 capture performance, including CO2 capacity, carbonation reaction rate, and the dissolution rate of alkaline earth metal ion, were analyzed to reveal the impact of controlled heat treatment and phase transformation.

The type of radioactive waste that may occur in the process of NPP dismantling can be classified into solid, liquid, gas, and mixed waste. Most of the radioactive waste generated during the dismantling of a NPP is metal solid waste, but liquid radioactive waste is also a very important factor in terms of radiation environmental impact assessment. In the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to design liquid radioactive waste processing system of Radioactive Waste Treatment Facility (RWTF). Depending on the amount of liquid radioactive waste generated, the type of liquid radioactive waste processing system included in the RWTF is different. In addition, in order to apply to the domestic RWTF, it is important to secure the site area occupied by the each system, the liquid radioactive waste treatment capacity of the system, and how to secure circulating water used for dilution and discharge of liquid radioactive waste. Therefore, this review aims to suggest an optimal method for the treatment system for liquid radioactive waste included in RWTF of Wolseong.

The decommissioning of Korea’s nuclear power facilities is expected to take place starting with the Kori Unit 1 followed by the Wolsong Unit 1. In Korea, since there is no experience of decommissioning, considerations of site selection for the waste treatment facilities and reasonable selection methods will be needed. Only when factors to be considered for construction are properly selected and their effects are properly analyzed, it will be possible to operate a treatment facility suitable for future decommissioning projects. Therefore, this study aims to derive factors to be considered for the site selection of treatment facilities and present a reasonable selection methodology through evaluation of these factors. In order to select a site for waste treatment facilities, three virtual locations were applied in this study: warehouse 1 to warehouse 3. Such a virtual warehouse could be regarded as a site for construction warehouses, material warehouses, annexed building sites, and parking lots in nuclear facilities. If the selection of preliminary sites was made in the draft, then it is necessary to select the influencing factors for these sites. The site of the treatment facility shall be suitable for the transfer of the waste from the place where the dismantling waste is generated to the treatment facility. In addition, in order for construction to take place, interference with existing facilities and safety should not be affected, and it should not be complicated or narrow during construction. Considering the foundation and accessibility, the construction of the facility should be economical, and the final dismantling of the facility should also be easy. In order to determine one final preferred plan with three hypothetical locations and five influencing factors, there will be complex aspects and it will be difficult to maintain consistency as the evaluation between each factor progresses. Therefore, we introduce the Analytic Hierarchical Process (AHP) methodology to perform pairwise comparison between factors to derive an optimal plan. One optimal plan was selected by evaluating the three virtual places and five factors of consideration presented in this study. Given the complexity and consistency of multiple influencing factors present and prioritizing them, AHP tools help users make decisions easier by providing simple and useful features. Above all, it will be most important to secure sufficient grounds for pairwise comparison between influencing factors and conduct an evaluation based on this.

The intermediate level spent resins waste generated from water purification for the the moderator and primary heat transport system during operaioin of heavy water reactor (HWR). Especially, moderator resins contain high level activity largely because of their C-14 content. So spent resins are considered as a problematirc solid waste and require special treatment to meet the waste acceptance criteria for a disposal site. Various methods have been studied for the treatment of spent resins which include thermal, destructive, and stripping methods. In the case of solidification methods, cement, bitument or organic polymers were suggested. In the 1990s, acid stripping using nitric acid and thermal treatment methods were actively investigated in Canada to remove C-14 nuclide from waste resin. In Japan, thermal distructive method was studied in the 1990s. Since 2005, KAERI developed acid stripping method using phosphate salt. However, acid stripping method are not suitable due to large amounts of 2nd waste containing acid solution with various nuclides. To solve this probelm, KAERI has been suggested the microwave treatment method for C-14 selective removal from waste resin in the 2010s. Pilot scale demonstration tests using radioactive waste resin generated from Wolsung unit 1 and unit 2 were successfully conducted and 95% of C-14 was selectively removed from the radioactive waste resin. In recent years, price of C-14 source is dramatically increased due to market growth of C-14 utilization and exclusive supply chain depending on China and Russia. High purity of C-14 were captured in HWR waste resin. Interest of C-14 recovery research from HWR waste resin is currently increased in Canada. In this study, microwave method is suggested to treat HWR waste resin with C-14 recovery process. Additionally, status of waste resin management and research trends of HWR waste resin treatment are introduced.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. In addition, according to the level of radioactivity, it can be divided into high level, intermediate level, low level, and clearance level waste. In the case of solid radioactive waste, it is necessary to secure disposal suitability in order to deliver it to a disposal facility, so safe and efficient treatment of a large amount of radioactive waste generated during decommissioning is one of the most important issues. For the treatment of radioactive waste generated during decommissioning, technologies in various fields such as cutting, decontamination, melting, measurement, and packaging are required. Therefore, this study intends to present and application plan for decommissioning domestic nuclear power plants through overseas case studies for the treatment of radioactive waste expected to occur during nuclear power plant decommissioning.

Various dry actives wastes (e.g., gloves, wipers, shoes, clothes) are generated during operation and maintenance of nuclear facilities. Among those, latex gloves gets interest because they contain both organic and inorganic compounds. CaCO3 is a common filler material for production of latex rubbers. Here, latex gloves were thermally treated in a closed vessel to separate the organic and inorganic compounds. Using the closed vessel is beneficial as it can prevent escape of any species, including radioactive nuclides in a real case, generated during the treatment. It was found that thermal decomposition of latex gloves occurred above 250°C. Latex gloves were decomposed to gas, liquid, and solid compounds. The gas product is thought to be volatile organic compounds (VOCs). The liquid product seems to be a mixture of oils and water. A CaCO3 phase was identified in the solid product, as expected. The VOCs can be easily separated at room temperature by purging in vacuum or inert atmosphere. The liquid-solid mixture can be separated by distillation. It is thought that gammaemitting nuclides, such as Cs-137, Sr-90, and Co-60, dominantly remain in the solid product. In the best situation, the solid product is the only subject to be transferred to final wasteform fabrication stream and thus volume of final waste can be reduced. Surrogates of contaminated latex gloves (containing Cs, Sr, and Co) were prepared and they were treated at 350°C in the closed vessel. How these contaminants behaves in this thermal process will be discussed in the presentation.

A large amount of small and medium-sized metal waste is generated during the decommissioning of nuclear power plants (NPPs). Metal waste is mostly contaminated with low-level radioactive, so it needs decontamination for self-disposal and recycling. A large amount of Organic Decontamination Liquid Waste during decontamination will be generated. The generated organic liquid waste is low in concentration, so the decomposition efficiency is low in the decomposition process. A conditioning process is necessary to concentrate at a high concentration. For effective treatment for Organic Decontamination Liquid Waste, the composition of organic liquid waste and conditioning process were analyzed. Organic acids, metal ions, radioactive nuclides, surfactants, etc. are present in the Organic Decontamination Liquid Waste, and suspended solids are sometimes generated by various reactions. According to previous studies, the concentration of organic acids including surfactants obtained results from several tens of ppm to a maximum of 1,000 ppm, so the maximum value of 1,000 ppm was assumed. For the composition and total amount of metal ions, the average value (52.7wt% Fe, 16.3wt% Ni, 15.1wt% Cr, 15.9wt% Mn) of the distribution of metal species removed by the actual decontamination process is applied, and the total amount is 1,000 ppm was assumed. As for the radionuclides, only 60Co and 137Cs, which are expected to be mainly present, were considered, and 60Co was assumed to be 2,000 Bq/g and 137Cs to be 360 Bq/g by referring to the literature. The amounts of suspended solids were assumed to be 500 ppm by referring to the characteristics of the liquid waste generated in the decontamination process of the NPPs. Based on the estimated value, a reaction formula was established and a simulated Organic Decontamination Liquid Waste was prepared. As a result of measurement using an analysis device, the composition of the estimated and simulated Organic Decontamination Liquid Waste had similar values. The conditioning and treatment process largely consists of pretreatment, conditioning, decomposition processes. Organic Decontamination Liquid Waste goes through a pretreatment process to remove impurities with large particles. In the conditioning process, treated water that has passed through the UF/RO membrane system is discharged into the environment. At this time, Concentrated water goes through a decomposition process for processing the Organic Decontamination Liquid Waste, and is discharged to the environment through a secondary RO membrane system. The conditioning process is the low-concentration Organic Decontamination Liquid Waste in the UF membrane system is forming a micelles in an RO membrane system, concentrating it to a high concentration and then go through a recirculation process in the UF membrane system. An experiment was conducted to confirm whether the concentration of surfactants occurred during the conditioning process. As a result of the experiment confirmed that the highly concentrated surfactant formed micelles and was filtered out in the UF membrane system.

Air conditioning facilities in nuclear power plants use pre-filters, HEPA filters, activated carbon filters, and bag filters to remove radionuclides and other harmful substances in the atmosphere. Spent filters generate more than 100 drums per year per a nuclear power plant and are stored in temporary radioactive waste storage. Plasma torch melting technology is a method that can dramatically reduce volume by burning and melting combustible, non-flammable, and mixed wastes using plasma jet heat sources of 1,600°C or higher and arc Joule heat using electric energy, which is clean energy. KHNP CRI & KPS are developing and improving waste treatment technology using MW-class plasma torch melting facilities to stably treat and reduce the volume of radioactive waste. This study aims to develop an operation process to reduce the volume of bag filter waste generated from the air conditioning system of nuclear power plants using plasma torch melting technology, and to stably treat and dispose of it. It is expected to secure stability and reduce treatment costs of regularly generated filter waste treatment, and contribute to the export of radioactive waste treatment technology by upgrading plasma torch melting technology in the future.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.