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.

To develop technology for extracting energy resources from seawater, we first investigated the research experiences of domestic experts. The survey items included the types of adsorbents that can adsorb dissolved resources in seawater, the subjects of experiments, and the scope of research. We divided the types of adsorbents into organic and inorganic categories and compared their adsorption performance. We also examined how adsorption experiments were conducted using simulated solutions and confirmed whether there were any experiences of conducting experiments in actual seawater. A total of 14 domestic research papers on extracting dissolved resources from seawater were reviewed, excluding topics such as removing dissolved resources from seawater and seawater desalination. This review provides an understanding of domestic research trends and will be helpful in setting directions for future research and development.

Seawater containing metals such as lithium and manganese is a “treasure trove” of infinite energy resources. Numerous domestic and foreign institutions are developing technologies to economically extract these resources from seawater. One method for extracting metal ions dissolved in seawater is the development of adsorbents with negative functional groups. Generally, adsorbents have adsorption performance that depends on factors such as seawater pH and temperature, but controlling the pH and temperature of seawater is practically impossible. On the other hand, thermal effluent discharged from power plants tends to be slightly higher in temperature than the surrounding environment. Therefore, this study investigates the potential for utilizing power plant effluent to extract dissolved resources in seawater. Results of investigations into several items related to the effluent from the Gori, Wolsong, Hanbit, and Hanul power plants are presented.

Aluminum’s exceptional properties, such as its high strength-to-weight ratio, excellent thermal conductivity, corrosion resistance, and low neutron absorption cross-section, make it an ideal material for diverse nuclear industry applications, including aluminum plating for the building envelope of nuclear power plants. However, plating aluminum presents challenges due to its high reactivity with oxygen and moisture, thus, complicating the process in the absence of controlled environments. Plating under an inert atmosphere is often used as an alternative. However, maintaining an inert atmosphere can be expensive and presents an economic challenge. To address these challenges, an innovative approach is introduced by using deep eutectic solvents (DES) as a substitute for traditional aqueous electrolytes due to the high solubility of metal salts, and wide electrochemical window. In addition, DESs offer the benefits of low toxicity, low flammability, and environmentally friendly, which makes DESs candidates for industrial-scale applications. In this study, we employed an AlCl3-Urea DES as the electrolyte and investigated its potential for producing aluminum coatings on copper substrates under controlled conditions, for example, current density, deposition duration, and temperature. A decane protective layer, non-polar molecular, has been used to shield the AlCl3-Urea electrolyte from the air during the electrodeposition process. The electrodeposition was successful after being left in the air for two weeks. This study presents a promising and innovative approach to optimizing aluminum electrodeposition using deep eutectic solvents, with potential applications in various areas of the nuclear industry, including fuel cladding, waste encapsulation, and radiation shielding.

In this study, we evaluated the performance of phosphate-functionalized silica in adsorbing uranium and provided insights into optimizing the initial conditions of the uranium solution (concentration and pH), which are often overlooked in uranium adsorption studies. While most studies take into account the effect of pH on both the surface charge of the adsorbents and the dissolved speciation of uranium in solution, they often overlook the formation of solid phases such as β-UO2(OH)2 (cr) and UO3· 2H2O(cr), leading to an overestimation of the adsorption capacity. To address this issue, we considered the speciation of U(VI) calculated using thermodynamic data. Our findings suggest that it is reasonable to evaluate the adsorption performance at pH 4 and concentration below 1.35 mM. The formation of β-UO2(OH)2 (cr) starts at 23 μM (pH 5) and 1 μM (pH 6) and increases sharply with increasing concentration. To avoid interference from the formation of solid phases, experiments should be conducted at lower concentrations, which in turn require very small msorbent/Vsolution ratios. However, controlling small amounts of sorbent can be challenging, and increasing the volume of the solution can generate significant amounts of radioactive waste. We also used UV-vis spectra analysis to investigate the formation of solid phases. We found that a 100 mg L-1 uranium solution resulted in the formation of colloidal particles in the solid phase after 2.5 hours at pH 6, while at pH 4, no significant changes in absorbance were observed over 120 hours, indicating a stable ion phase. Based on these conditions, we obtained an excellent adsorption capacity of 110 mg g-1.

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.

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.

Concrete decontamination tools capable of removing the nuclear contaminated surface are necessary to minimize the amount of concrete waste generated in the process of decontamination and dismantling of nuclear power plants. Laser scabbling is a decontamination technique that removes the contaminated surface layers concrete surface by inducing internal explosion. The application principle of laser scabbling technology uses the porous nature of concrete including moisture. When high thermal energy is applied to the concrete surface, an explosion at pores is induced along with an increase in water vapor pressure. High-powered laser beam can be an effective induction source of local explosive spalling on concrete surface. In this study, the scabbling test using a 5 kW highpowered fiber laser was conducted on the concrete blocks to establish the optimal conditions for surface decontamination. It was also measured the volume peeled off the concrete surface under the conditions of two different laser head speeds. Furthermore, we tested the removal efficiency of radioactive concrete particles generated during high-power fiber laser scabbling process. A 5 kW laser beam was applied to the concrete surface at two different laser head speeds - 120 mm/min and 600 mm/min. The laser beam repeatedly moved 200 mm horizontally and 40 mm vertically within the concrete block. The amount of surface concrete removed from concrete block was calculated from the measurement of the volume and mean depth using a 3D scanner device (laser-probed Global Advantage 9.12.8(HEXAGON)) for the two different the laser head speeds. By increasing the laser head speed, less explosive spalling occurred due to shorter contact time of the laser beam with the concrete. The laser head speed of 600 mm/min reduced about 89% of the waste generated by shallow depth of scabbling as compared to the waste generated at the laser head speed of 120 mm/min. The fiber laser scabbling system was developed for surface decontamination of radioactive concrete in nuclear power plants. Tests were performed to find the optimum parameters to reduce the generation of particulate waste from the contaminated concrete surface by controlling the laser head speeds. It was confirmed that the wastes from surface decontamination was reduced up to 89% by increasing laser head speed from 120 mm/min to 600 mm/min. It was also observed that the cylindrical tube effectively vacuumed the debris generated by the explosive spalling into the collector. Removal efficiencies of concrete particles were measured greater than 99.9% with ring blower power of 650 air watt of the filter system.

The removal of aqueous pollutants, including dye molecules from wastewater remains one of the pressing problems in the world. Because of chemical stability and conjugated structure, dye molecules cannot be easy decomposed by heat with oxidizing reagents such as H2O2 and light. The most common representative of widespread organic pollutant is methylene blue (MB) with molecular formula C16H18ClN3S, which is important colorant and used in various chemical and biological production industries and causes serious environment problems. Porous materials, including MOFs (metal-organic frameworks) have been applied for efficient MB photocatalytic degradation. However, one of the main barriers to using most MOFs to break down aromatic organics is wide band gap energy, which means that the catalyst can exhibit high photocatalytic performance only under UVlight irradiation. Moreover, most MOFs usually show the poor water stability of frameworks, which tend to dissolve in water with total destruction. In this work we report about two new copper based MOFs with high photocatalytic properties for efficient MB degradation from wastewater under UV-light and natural sunlight. Time, required for 100% MB degradation, equals 7 minutes under UV (source 4 W 254 nm VL-4.LC UV-lamp) and 60 minutes under natural sunlight irradiation in the presence of H2O2. Crystal structure information is provided using single crystal X-ray diffraction data. The composition and comparative characteristics of MOFs are given using powder X-ray diffraction, UV–visible diffuse reflectance spectroscopy, UVvisible spectroscopy and Fourier-transform infrared spectroscopy.

Laser scabbling has the potential to be a valuable technique capable of effectively decontaminating highly radioactive concrete surface at nuclear decommissioning sites. Laser scabbling tool using an optical fiber has a merits of remote operation at a long range, which provides further safety for workers at nuclear decommissioning sites. Furthermore, there is no reaction force and low secondary waste generation, which reduces waste disposal costs. In this study, an integrated decontamination system with laser scabbling tool was employed to test the removal performance of the concrete surface. The integrated decontamination system consisted of a fiber laser, remote controllable mobile cart, and a debris collector device. The mobile cart controlled the translation speed and position of the optical head coupled with 20 m long process fiber. A 5 kW high-powered laser beam emitted from the optical head impacted the concrete block with dimensions of 300 mm × 300 mm × 80 mm to induce explosive spalling on its surface. The concrete debris generated from the spalling process were collected along the flexible tube connected with collector device. We used a three-dimensional scanner device to measure the removed volume and depth profile.

KAERI has been developing a new decontamination process that does not contain any organic chemicals in the decontamination solution and minimizes the use of ion exchange resin in the solution as a purifying step. The process is hydrazine based reductive metal ion decontamination for decommissioning (HyBRID) and consists of N2H4, H2SO4 and Cu+ ions. The primary system of the LWR is composed of materials with high corrosion resistance, such as stainless steel and Inconel, but among the materials, the feeder and header of the primary system of the PHWR are composed of carbon steel (SA106Gr.B) with low corrosion resistance. Therefore, when decontamination of PHWRs, attention should be paid to corrosion of carbon steel. Since Fe3O4, a contaminating oxide film formed on the surface of carbon steel dissolves faster than ferrite or chromite formed on the surface of Inconel or stainless steel, the base material is exposed to the solution and is corroded during decontamination. When a large amount of iron ions is eluted into the decontamination agent due to corrosion of carbon steel, not only the soundness of the base metal is adversely affected, but also the amount of decontamination waste increases. The purpose of this study is to develop inhibitors that can minimize corrosion of carbon steel when decontamination of PHWRs using the HyBRID decontamination process. CG, 570S and PP3 were selected as corrosion inhibitors. In addition, corrosion tests of carbon steel were conducted in the HyBRID solution with corrosion inhibitors. The best corrosion inhibitors and optimal operating conditions were selected, and HyBRID decontamination agents with corrosion inhibitors were much better in corrosion resistance than existing commercial decontamination agents.

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.

Dry active wastes (DAWs) are a type of combustible radioactive solid waste, which includes decontamination paper, protective clothing, filters, plastic bags, etc. generated from operating nuclear facilities and decommissioning projects. The volume of DAWs could be increased over time, disadvantage to higher disposal costs and space utilitization of disposal site. Additionally, incineration methods cannot be applied to DAWs, unlike general environmental waste, due to concerns about air pollution and the release of harmful chemicals with radioactive nuclides into the atmosphere. Recently, KAERI developed an alternative thermochemical process for reducing the volume of DAW, which involves a step-wise approach, including carbonization, chlorination, and solidification. The purpose of this process is to selectively separate the radioactive nuclides from carbonized DAWs that are less than clearance criteria, which can be disposed of as non-radioactive waste. In this research, we investigated the thermal decomposition characteristics of DAWs using nonisothermal thermogravimetric analysis, which was performed with different categorized wastes and heating conditions. As a result, the cellulose DAWs such as decontamination paper and cotton were thermally decomposed in three or four-step depending on the heating conditions. On the other hand, the hydrocarbon and rubber DAWs such as plastic bags and latex were thermally decomposed in one or two-step. Therefore, it could be suggested the thermochemical treatment conditions that minimize the decomposition of DAWs by controlling the reaction steps, and we will try to apply these results for cellulose type DAWs such as decontamination paper and cotton, which is generated majorly from the nuclear facilities in the future.

Laser cutting has been recognized as one of key techniques in dismantling nuclear power plants as it has several advantages such as a remote operation and a reduced secondary waste. However, it generates a significant amount of aerosols that can pose a health risk to workers and further induce environmental pollution during the cutting operation. Thus, understanding the aerosol characteristics generated by the laser cutting is crucial for implementing an effective cutting operation and reducing the exposure to these hazardous particles. In this work, we established a methodology to collect the aerosols and investigate their properties in the laser cutting operation. We built an integrated laser cutting system for aerosol analyses, consisting of a high-power laser cutting module, a metal sample holder, an aerosol collector, and a closed chamber. We expect that this system will offer an opportunity for in-depth understanding of the aerosol properties, by connecting it with desired type of aerosol analysis platforms, and further safe dismantling operation of the nuclear power plants.

Laser cutting has many advantages, including high-speed cutting potential, no reaction forces, narrow kerf widths, ease of remote control, and more. This makes it the next generation cutting technology for nuclear decommissioning. For this reason, various groups in countries with nuclear power plants have been working on applying laser cutting to nuclear decommissioning. Our group has also been developing in-air and underwater laser cutting technologies. Previous research has focused on efficiently cutting thicker steels. To accomplish this, a cutting head with a long focusing element with a focal length of 600 mm was utilized. A long focusing head is advantageous for cutting thick objects at high speeds because it can maintain a high power density over a long distance. However, with such a long focused beam, the residual laser power that remains after passing through the target object can cut or damage other unwanted objects located behind the target. Utilizing a short focused element can solve this problem, but if the focal length is too short, the cutting capability will be reduced. In this work, we developed and applied a cutting head that utilizes a focused element with a short focal length of 300 mm. Cutting tests with this head allowed us to cut 10-60 mm thick stainless steel plates at a laser power of 6 kW. We also obtained the maximum cutting speed and kerf width for each thickness while increasing the laser power by 1 kW from 1 to 6 kW. The results obtained in this work are expected to be utilized for safe cutting in future nuclear decommissioning applications.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.

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.

Concrete radioactive waste is divided into surface-contaminated concrete and activated concrete, and although the generation rate varies depending on the operating conditions of the nuclear power plant, it is reported that the amount of surface-contaminated concrete generated is greater. It is reported in the ‘US-NRC Inventory Report’ that 99% of radionuclides in surface-contaminated concrete are distributed within 1 mm of the surface. Since concrete radioactive waste accounts for a large amount of generation after metal radioactive waste, it is necessary to reduce the amount of radioactive waste disposal by applying appropriate treatment techniques to surface-contaminated concrete. In this study, a similar contamination environment work space with the size of 5.4 (W) × 3.6 (L) × 2.5 (H) [m] in which concrete specimens can be fixed on the wall and floor was established. And an integrated decontamination equipment was verified the automation performance for ‘location accuracy’, ‘radioactive contamination level measurement’ and ‘concrete surface laser scabbling’. It was confirmed that the average was 8.3 [mm] in the evaluation of the ‘location accuracy’ for the remote control and movement of the integrated decontamination equipment. For performance verification of ‘radioactive contamination level measurement’ and ‘laser scabbling’, it were used that size of 30×30×8 [cm] ordinary concrete specimens and concrete radioactively contaminated with Co-60 below the regulatory exemption concentration. ‘Radioactive contamination level measurement’ is measured as much as the set range, divied and display the measured values in different colors on the map of the control program. Ordinary concrete specimens are 0.066~0.089 μ Sv/hr, and contaminated concrete specimens are 0.107~0.121 μ Sv/hr, and it was confirmed that they are expressed in different colors on the map. For ‘laser scabbling’, the performance according to the laser scabbling speed and reproducibility with ordinary concrete specimens was verified. As a result, a weight change of up to 1.48 kg was confirmed. Contaminated concrete specimens were subjected to a direct method using a surface contamination detector and an indirect method using a smear paper to measure surface contamination before and after scabbling, and the debris generated after scabbling was analyzed using HPGe.

The spent filters used to purify radioactive materials and remove impurities from primary systems at nuclear power plants (NPPs) have been stored for long periods in filter storage rooms at NPPs due to concerns about the unproven safety of the treatment method, absence of disposal facilities, and risk of high radiation exposure. In the storage room at Kori Unit 1, there are approximately 227 spent filters of 9 different types. The radiation dose rates of filters range from 0.01 to 500 mSv/hr. Recently, a comprehensive plan has been established for the treatment and disposal of radioactive waste that has not yet been treated to facilitate decommissioning of NPPs. As a follow-up measure, compression and packaging optimization processes are being developed to treat the spent filters. KHNP plans to dispose of the spent filters after compressing, packaging, and immobilizing them. However, the spent filters are currently stored without being sorted by type or radiation intensity. If the removal and packing of the filters are done randomly without a plan for the order of withdrawal and subsequent processes, issues may arise such as a decrease in drum loading efficiency and exceeding the dose limit of the package. In this study, the number of drums needed to pack the spent filters was calculated, considering the filter size, weight, quantity, dose rate, shielding thickness of drum, and loadable quantity in a shielding drum (SD). Then, the spent filters that can be loaded on each drum were classified into one group. In addition, the withdrawal order for each group was set so that the filter withdrawal, compression, and packaging processes could be performed efficiently. The spent filter groups are as follows: (1) compression/12 cm SD (17 groups), (2) compression/16 cm SD (6 groups), (3) non-compression/ intermediate storage container (17 groups, additional radiation attenuation required due to high dose rate), and (4) unclassified (5 groups, determined after measurement due to lack of filter information). The withdrawal order of the groups was determined based on several factors, including visual identification of the filter, ease of distribution after withdrawal, work convenience, and safety. Due to the decay of radioactivity over time, the current dose rate of the spent filters is expected to be much lower than at the time of waste generation. Therefore, in the future, sample filters will be taken from the storage room to measure their radioactivity and radiation dose rate. Based on these measurements, a database of radiological characteristics for the 227 filters will be created and used to revise the filter grouping.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.