This study demonstrates a different approach method to fabricate antimony selenide (Sb2Se3) thin-films for the solar cell applications. As-deposited Sb2Se3 thin-films are fabricated via electrodeposition route and, subsequently, annealed in the temperature range of 230 ~ 310oC. Cyclic voltammetry is performed to investigate the electrochemical behavior of the Sb and Se ions. The deposition potential of the Sb2Se3 thin films is determined to be -0.6 V vs. Ag/AgCl (in 1 M KCl), where the stoichiometric composition of Sb2Se3 appeared. It is found that the crystal orientations of Sb2Se3 thin-films are largely dependent on the annealing temperature. At an annealing temperature of 250 oC, the Sb2Se3 thin-film grew most along the c-axis [(211) and/or (221)] direction, which resulted in the smooth movement of carriers, thereby increasing the carrier collection probability. Therefore, the solar cell using Sb2Se3 thin-film annealed at 250 oC exhibited significant enhancement in JSC of 10.03 mA/cm2 and a highest conversion efficiency of 0.821 % because of the preferred orientation of the Sb2Se3 thin film.

Measurement of the physical properties of high-temperature molten salts is important for the efficient design and operation of molten salt reactors (MSR) in which the reactor coolant and nuclear fuel are in a homogeneous liquid state. Although some crucial physical properties such as viscosity, thermal conductivity, density, etc., have been drawing much attention, relative data, especially for molten chloride salts, are scarce. Thus, it is urgent to prepare the viscosity data as one of the key transport properties in thermal hydraulics analysis. However, it is not an easy task to measure the molten salt viscosity with high accuracy due to end effect, a small gap between the chamber and spindle, thermal expansion of the chamber and spindle at high temperatures in a rotational viscometer. Additionally, molten salt temperatures inside furnace are not uniform due to the large temperature gradient inside the chamber, and therefore the assumption of laminar condition can be violated. In this study, geometric factors, which can be a major interference in the torque measurement, were considered for the accurate determination of the viscosity. We established a high-temperature molten salt viscosity measurement system with Brookfield rotational viscometer. KNO3 molten salt was used as a model substance at a temperature range of 650–773 K. In-house designed spindles and chambers were made of corrosion-resistant alumina. Thermal expansion has a significant influence on the size and shape of the chamber and spindle. The effect of thermal expansion on the conventional correction method was examined with temperature variation and distribution. Gap size variation was also investigated in order to improve the accuracy.

This study presents a rapid and quantitative sequential separation method for H-3 and C-14 isotopes with distillation apparatus in environmental samples released from nuclear facilities. After adding 200 mg of granulated potassium permanganate and 500 mg of sodium hydroxide in 100 mL of sample solution, the sample solution was heated until approximately 10 mL of distillate, and the distillate fraction was removed. The sample solution was heated again until a minimum 10 mL of additional distillate was collected. 10 mL of distillate was transferred to the LSC vail and the measurement sample for H-3 was made by adding 10 mL of Ultima Gold LLT to the LSC vial. After adding 2.5 g of potassium persulfate, 2 mL of 1M silver nitrate and 15 mL of concentrated nitric acid to the remained sample solution, the sample solution was heated for 90 minutes and C-14 isotopes were adsorbed into 10 mL of Carbo-Sorb solution in glass vial. The measurement sample for C-14 was made by adding 10 mL of Permafluor to the C-14 fraction in glass vial. The purified H-3 and C-14 samples were measured by the liquid scintillation counter after quenching correction. The average recoveries of H-3 and C-14 with CRM were measured to be 96% and 85%, respectively. The sequential separation method for H-3 and C-14 investigated in this study was applied to activated charcoal filter produced from nuclear power plants after validating the reliability by result of proficiency test (KOLAS-KRISS, PT-2021-51).

According to the article 18 of NSSC notice “Regulations on the delivery of low-and intermediatelevel radioactive wastes”, the consignor shall establish and implement the quality assurance program about waste management to ensure conformity with the criteria set forth in the regulations and detailed criteria proposed by the disposal facility operator, including matters related to characterization of the waste concerned. To meet the above requirement, commercially available laboratory information management system, STARLIMS from Abbott Informatics was introduced in the late of 2019 and was customized to our standardized test method in 2020. In that time, Electronic Lab Notebook (ELN), which is an electronic system to create, store, retrieve, and share fully electronic records, was tailored to replace paper lab notebook. Scientific Data Management System (SDMS), which is computer system used to capture, centrally store, catalog, and manage data, was installed. Due to the parsing ability of SDMS, human error like mistake while data entry was reduced by extracting data from measurement sheet and exporting measurement data to designated area of ELN and this feature made work efficiency improved. Afterward, validation of STARLIMS was conducted following the procedure of user acceptance testing including Operational Qualification and Performance Qualification. As a result of these activities, STARLIMS has been officially operated and applied to means to manage test results since 2021. In 2021, for user-friendly environment, updating STARLIMS was also conducted by applying SDMS to import data from other radiometric measurements including gas proportional counter (GPC), liquid scintillation counter (LSC), and low-energy Ge detector (LEGe) besides HPGe detector for gamma measurement. From implementation to operation, it is confirmed that STARLIMS has been providing reliable and stable platforms to manage laboratory information regarding measurement records and playing a significant role in tool to meet the quality assurance required.

Density of chloride molten salts is an essential physical property in the reactor core design and thermal-hydraulic design simulation, especially in molten salt reactor (MSR) design currently under development in Korea. NaCl-MgCl2-UCl3 pseudo-ternary system is one of the various candidate chloride-based salt mixtures because it has relatively-low melting point, very low vapor pressure, high thermal conductivity, etc. However, to the best of our knowledge, the density data of NaCl-MgCl2- UCl3 have not yet been measured or published worldwide, and therefore the ballpark figures of the density should be given for the preliminary reactor design. In our present study, the density estimation of NaCl-MgCl2-UCl3 based on the pseudo-binary data, i.e., NaCl-MgCl2, MgCl2-UCl3, and NaCl- UCl3, reported in the literature previously were performed using the Redlich-Kister model. Binary interaction parameter for MgCl2-UCl3 was higher than that for NaCl-MgCl2 and lower than that for NaCl-UCl3. As an example, calculated density of 0.62 NaCl: 0.18 MgCl2: 0.20 UCl3 at 873 K was 2.578 g·cm−3. In our further study, the methodology using Redlich-Kister model will be applied to more complex multicomponent systems and to other physical properties such as viscosity, thermal conductivity, surface tension, etc.

The dose was evaluated for the workers transporting the spent resin drums from a spent resin mixture treatment facility. The treatment technology of spent resin mixture waste based on microwave was developed to compensate for the shortcoming of the existing one. The mechanism of the facility for the treatment is divided into separation, desorption, condensation and adsorption process. The treated spent resin that has passed through the microwave reactor flows into the spent resin storage tank. As the treatment time elapses, if spent resin accumulates in the spent resin storage tank, it is moved to the drum of the volume of 200 L. The drum must be moved by the worker, in which case radiation exposure to the drum transport worker occurs. It requires the dose evaluation for drum transport workers in terms of radiation safety. Dose evaluation was performed in consideration of the change in the composition ratio and weight of the spent resin mixture, where the working time for transportation was considered from 10 to 120 minutes in 10-minute increment. In the case of 100 kg of the spent resin mixture, the dose range was derived as 4.62×10−3 – 5.90×10−2 mSv for the 100 kg of spent resin, 4.72×10−3– 5.58×10−2 mSv for the 80 kg of spent resin and 20 kg of zeolite and activated carbon, and 5.38×10−3 – 6.32×10−2 mSv for the 60 kg of spent resin and 40 kg of zeolite and activated carbon. In the case of 150 kg of the spent resin mixture, the dose range was derived as 6.83×10−3 – 8.20×10−2 mSv for the 150 kg of spent resin, 7.13×10−3 – 8.22×10−2 mSv for the 120 kg of spent resin and 30 kg of zeolite and activated carbon, and 8.28×10−3 – 8.86×10−2 mSv for the 90 kg of spent resin and 60 kg of zeolite and activated carbon. The estimated maximum doses for each weight (100 kg and 150 kg of mixture) were confirmed to be 3.16×10−1% and 4.43×10−1% of the annual average dose limit of 20 mSv for radiation workers.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.

Detritiation of low-level tritiated water has become global issue after Fukushima accident. Several attempts have been made to reduce the radioactivity of Fukushima tritiated water below legal limit of nuclear plant effluents (~104 Bq·L−1). Various technologies such as water distillation, electrolysis, and catalytic exchange were tested to treat the tritiated water, however, those demand enormous expense to achieve the goal due to low process efficiency. It highlights that the performance enhancement of current technologies is necessary to improve economic feasibility. We have quantitatively evaluated the separation performance of various polymers toward low-level tritium (~105 Bq·L−1) through batch experiments. The polystyrene with grafted by 20 types of functional groups (Tris (2-aminoethyl) amine, dimethylaminomethyl, isocyanate, mercaptomethyl, aminomethyl, hydroxymethyl, triphenylphosphine, morpholine, 2-chlorotrityl amine, 4-(dimethylamino) pyridine, poly (vinyl chloride) carboxylated, poly (4-vinyl pyridine), p-toluenesulfonic acid, p-toluenesulfonyl hydrazide, piperidine, acetyl polystyrene, 2-chlorotrityl chloride, piperazine, diethylene triamine, poly (vinyl chloride)) were suspended in HTO solution (initial activity = 4.7 × 105 Bq·L−1). After the equilibration, the suspension was filtered using 3 kDa membrane filter and the activity in filtrates were quantified by LSC (HIDEX-300 SL). The results demonstrate the detritiation efficiency and separation factors of functional groups toward tritium. Carboxylic group (COOH) showed the most reactive performance as detritiation efficiency of ~4%. Compared to other functional groups, styrene sulfonyl groups including sulfonyl amide (SNH2) and sulfonyl hydrazide (SNHNH2) revealed promising performance for tritium separation as separation factor of 10.97 and 3.85, respectively. However, sulfonyl hydroxide (SOH) which is known as reactive functional group to tritium exchange showed the poor performance (detritiation efficiency: 0.68%; separation factor: 3.02). This study could suggest the promising functional groups for detritiation of low-level tritiated water which can be utilized to enhance the performance of current technologies. For example, reactive functional groups can be grafted on the surface of packing material within distillation tower resulting in the increasing detritiation efficiency.

This study is for evaluation and optimization of workers’ radiation exposure for dismantling Reactor Vessel (RV) at Kori unit 1 in connection with its decommissioning process for the purpose of establishing Radiation Safety Management Plan. This is because the safety of workers in a radiation environment is an important issue. The basis of radiological conditions of this evaluation is supposed to be those of 10 years after the permanent shutdown of Kori unit 1 when dismantling work of Reactor Vessel would suppose to be started. Dose rates of work areas were evaluated on the basis of spatial dose rate derived from activation level calculated by MCNP (Monte Carlo N-Particle Transport) and ORIGEN-S code. RV are radiated by neutrons during operation, creating an environment in which it is difficult for operators to access and work. Therefore, the RV must be dismantled remotely. However, due to work such as installing devices or dismantling surrounding structures, it is not possible to completely block the access of workers. Accordingly, the exposure of workers to the RV dismantling process should be assessed and safety management carried out. The dismantling process of Kori unit 1 RV was developed based on in-situ execution in atmospheric environment using the oxigen-propane cutting technology as the following steps of Preparation, Dismantling of Peripheral Structures, Dismantling of RV and Finishing Work. For evaluation of exposure of RV dismantling work, those processes of each steps are correlated with spatial dose rates of each work areas where the jobs being done. Results of the evaluation show that workers’ collective dose for RV dismantling work would be in the range of 536–778 man- mSv. The most critical process would be dismantling of upper connecting parts of RV with 170–256 manmSv while among the working groups, the expert group performing dismantling of ICI (In-core instrumentation) nozzles and handling & packaging of cut-off pieces is evaluated as the most significantly affected group with 37.5–39.4 man- mSv. Based on the evaluation, improvement plan for better working conditions of the most critical process and the most affected workers in terms of radiation safety were suggested.

The structural stability of the recycled concrete disposal container was evaluated and compared the applicability of the current design standards for recycled concrete of nuclear power plant. The structural stability requirement for concrete disposal containers is 37.7 MPa or more. As a result of the compressive strength test on recycled concrete, 50% of coarse aggregates of recycled concrete was 42.1 MPa. In addition, it was found that the bending strength and shear strength of recycled reinforced concrete beam exceeded the current design standard. Therefore, it is judged that recycled concrete containing coarse aggregates can be sufficiently utilized. It was possible to ensure the structural stability of the concrete container without changing the design specifications and reinforcing bars when recycled concrete is applied.

Currently, in domestic nuclear power plants (NPP), the spent filters (SFs) used for the purpose of reducing and purifying the radiation of the primary cooling water system are temporarily stored in an untreated state. In order to dispose of SFs, radioactive nuclide analysis (RNA) of SFs is required to be conducted. As segmented gamma scanner (SGS) is already being used in Kori NPP, utilizing SGS for RNA of SFs would be practical and economical. In this paper, factors required to be considered to improve accuracy of SGSs for RNA of SFs are studied. The analysis of the nuclide inventory of the packaging drum for radioactive waste should be performed by the indirect drum nuclide analysis method. The material of the SFs is iron (SS304) on the outside, and paper on the inside. In addition, to meet disposal acceptance criteria, radioactive waste drums are packaged in thick grouting or shielding drums. Therefore, it is necessary to derive an appropriate correction method for high inhomogeneity and thick media. Considering these factors, evaluating radionuclides inventory plans to measure gamma rays in SGS mode. Correct the gamma ray measurement by examining the medium attenuation factor and error factors. In this way, the inventory of gamma nuclides is calculated, and the specific radioactivity of beta ray and alpha particle emitting nuclides other than gamma rays is planned to be calculated by applying scaling factors.

Source localization technique using acoustic emission (AE) has been widely used to track the accurate location of the damaged structure. The principle of localization is based on signal velocity and the time difference of arrival (TDOF) obtained from different signals for the specific source. However, signal velocity changes depending on the frequency domain of signals. In addition, the TDOF is dependent on the signal threshold which affects the prediction accuracy. In this study, a convolutional neural network (CNN)-based approach is used to overcome the existing problem. The concrete block corresponding to 1.3×1.3×1.3 m size is prepared according to the mixing ratio of Wolseong low-to-intermediate level radioactive waste disposal concrete materials. The source is excited using an impact hammer, and signals were acquired through eight AE sensors attached to the concrete block and a multi-channel AE measurement system. The different signals for a specific source are time-synchronized to obtain TDOF information and are transformed into a time-frequency domain using continuous wavelet transform (CWT) for consideration of various frequencies. The developed CNN model is compared with the conventional TDOF-based method using the testing dataset. The result suggests that the CNN-based method can contribute to the improvement of localization performance.

The radwaste facility management team is preparing for clearance of 4 MCAs in The Radwaste Form Test Facility (RFTF). The targeted waste was used for clearance level radioactive waste sample analysis and has been used for this purpose since the early 2000s. Due to the characteristics of clearance level radioactive waste, the concentration of radioactivity is very low and MCA is used with Marinelli beakers the possibility of contamination is low. Moreover, the radiation detector should not be contaminated with radioactive materials, it should be less than the clearance level. These detectors were considered surface contamination materials. To detect the contaminated spot of each material, we scanned the whole surface of a material with a gamma survey meter. After that, we took a sample from 1×1 m2 and a total of 30 samples from each MCA. The wiped filter paper was analyzed with alpha, beta low background counting systems. The results of the analysis of the smear sample of total alpha and beta nuclide radioactivity were less than MDA (α: 2.88×10−5 Bq·cm−2, β: 3.07×10−5 Bq·cm−2). The major nuclide in this facility is Co-60 and Cs-137 therefore we analyzed gamma nuclide activity with HPGe. The maximum specific activity was Co-60: 2.31×10−5 Bq·cm−2, Cs-137: 1.96×10−6 Bq·cm−2. If it is satisfied with the clearance criteria, detectors will be reused at the Radioactive Waste Treatment Facility (RWTF) room # 7251 uncontrolled area.

Low and intermediate radioactive wastes in South Korea have been disposed in Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC), Gyeongju. This repository structure is planned to be operated few hundred years while toxicity of the waste is sufficiently decayed. The structural integrity of the repository is required to protect the waste in safe. The integrity of the structure is commonly estimated using acoustic emission (AE) method. The integrity of the structure using AE is obtained by following process: 1) Estimation of maximum acoustic crack energy of the structure, 2) Acoustic signal measurement and filtering, and 3) Measurement of simultaneous acoustic cracking energy. The damage of the structure can be obtained from cumulative cracking energy from the structure divided by the predicted maximum cracking energy of the structure. Estimation of maximum cracking energy is gained by the specimens whose components are identical to the repository structure. The cracking energy of the different specimens are obtained during uniaxial compressive test and volume of the specimen is calculated. Then, the fractal coefficient for the structure is obtained and the maximum crack energy of the target structure can be calculated. The specimens whose diameters vary from 50 mm to 150 mm and heights are twice of the diameter are made with same recipe of WLDC silo concrete. The uniaxial compression test is conducted with loading rate of 0.1 mm·min−1. The fractal coefficient is obtained by least square method from the volume-cumulative energy relationship.

Low to intermediate radioactive waste disposal concrete structures are subjected to coupled hydromechanical conditions and the identification of structural damage is crucial to ensure safe long-term disposal. Different damage models for concrete and the surrounding rock can affect the damage characteristics of radioactive waste disposal structures. In this study, the effects of different rock damage models are applied to the hydro-mechanical-damage coupled structural analysis of the Wolseong Low and Intermediate Level Radioactive Waste Disposal Center silo. A two-dimensional model of the disposal silo was modeled using the finite element analysis software COMSOL and the Mazars’ damage model was applied to the silo concrete. The Mazars’ model parameters were obtained from uniaxial compression and tensile tests on cylindrical concrete specimens after 28 days of water curing and further 32 days of wet curing at 75°C). The COMSOL embedded Richards equation module was used to simulate hydraulic analysis. Structural loading due to waste disposal was applied at the bottom of the silo structure and the damage evolution characteristics were investigated. The non-linear mechanical rock behavior obtained from laboratory tests (Hoek-Brown criterion, resonant column test, Mazar’s damage model) and field tests (Goodman Jack) were input to assess the effects of different rock damage models. The results highlight the importance of structural damage consideration when assessing the long-term stability and safety of underground radioactive waste disposal structures under coupled hydro-mechanical conditions.

Radioactive materials emitted from nuclear accident or decommissioning cause soil contamination over wide areas. In the event of such a wide area of contaminated soil, decontamination is inevitable for residents to reside and reuse as industrial land. There are many ways to decontaminate these contaminated soils, but in urgent situations, the soil washing, which has a short process period and relatively high decontamination efficiency, is considered the most suitable. However, the soil washing process of removing fine soil and cesium by using washing liquid as water and adding a flocculating agent (J-AF) generates slurry/sludge-type secondary waste (Cs-contaminated soil + flocculating agent). Since this form of sludge contaminants cannot be disposed, solidification is needed using an appropriate solidification agent to treat wastes for disposal. Therefore, this study devised a treatment method of contaminated fine soils occurring after the soil washing process. This investigation prepared the simulated wastes of contaminated fine soils generated after the soil washing, and pelletized the samples using a roll compactor under the optimum operating conditions. The optimum conditions of the device were determined in the pre-test. Roll speed, feeding rate, and hydraulic pressure were 1.5 rpm, 25 rpm, and 28.44 MPa, respectively. The waste forms were manufactured by incorporating created pellets (H 6.5 × W 9.4 mm) using polymers as solidification agents. Used polymers were main ingredient (YD-128), hardener (G-1034), and diluent (LGE). The optimum mixing ratio was YD-128 : G-1034 = 65 : 35 phr, and LGE was added in an amount of 10wt% of the total mixture. To confirm the disposal suitability of the manufactured waste forms, characterization evaluation was carried out (compressive strength, thermal cycling, immersion, and leaching test). Characterization evaluation revealed a minimum compressive strength of 23.1 MPa, far exceeding 3.44 MPa of the disposal facility waste acceptance criteria. Compressive strength increased to the highest value of 31.90 MPa after immersion test. To examine leaching characteristics, the pH, Electrical Conductivity (EC) and leachability index (􀜮􀯜) of leachates were identified. As results, pH and EC consistently increased or remained constant with leaching time. The average of Co, Cs and Sr nuclides was 17.76, 17.38 and 14.04, respectively, exceeding the value of 6 in the waste acceptance criteria. Effective waste treatment/ disposal can be achieved without increasing volumes of sludge/slurry by enhancing the technique of this research by performing additional studies in the future.

After the Fukushima accident in 2011, a huge amount of radioactively contaminated water is being generated by cooling the melted fuel of units 1, 2 and 3. Most of contaminated water is seawater and underwater containing not only salt elements but also nuclear fission products with radioactivity. To treat the contaminated water, Cs/Sr removal facilities such as KURION and SARRY are being operated by TEPCO. Additionally, three ALPS facilities are on operation to meet the regularity standards for discharge to the sea. However, massive secondary wastes such as Zeolite, sludge and adsorbent is being generated by these facilities for liquid water treatment. The secondary wastes containing various radionuclide with Cs and Sr is difficult to store due to highly radioactive concentration and corrosive properties. In Japan, a variety of technologies such as GeoMelt vitrification, In-Can vitrification and CCIM vitrification is considered as a promising solution. In this study, they were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for thermal treatment of sludge radwaste.

In KHNP CRI, the 100 kW PTM (plasma torch melting) system was designed for the treatment and disposal technology of various radioactive wastes including the metal, concrete, liquid waste and insulator. The facility consists of melting chamber, thermal decomposition chamber, waste feeding system and off-gas treatment system. In this study, to evaluate the applicability of the PTM system, demonstration test was conducted using the radiation hazmat suit as combustible waste. The plasma melting chamber is pre-heated by 2nd combustion device and plasma torch for 5 hours. The temperature inside the plasma melting chamber is approximately 1,600°C. The combustible waste was put into the melting chamber by the pusher feeding device with the throughput of maximum 50 kg/hour. During the test, the power of plasma torch is 60–96 kW on the transferred mod. It was evaluated in terms of long-term integrity of PTM system on operation according to the waste throughput ratio.

Plasma melting technology has been considered as promising technology for treatment of radioactive wastes. According to the IAEA TECDOC-1527 report (2006), the technology has an advantage that it can treat regardless of waste types which is both combustible and non-combustible wastes. In particular, it is expected that a large amount of concrete, a representative non-combustible wastes, will be generated during the operation and dismantling of nuclear power plants. In order to treat the concrete waste in plasma torch melting system, various factors could be considered like the slag of electric conductivity, viscosity and melting temperature. Above all, as a critical factor, the viscosity of the melt is very important to easily discharge the melt. The viscosity of slag (SiO2-CaO-Al2O3 system) can be lowered by adding a basic oxide such as CaO, Na2O, MgO and MnO. The basic oxides are donors of oxygen ions. These oxides are called notwork breakers, because they destroy the network of SiO2 by reacting with it. In this study, the slag composition of the concrete waste was developed to apply the plasma torch melting. Also, demonstration test was performed with the developed slag composition and 100 kW plasma torch melting system.

The feasibility study of synthesizing graphene quantum dots from spent resin, which is used in nuclear power plants to purify the liquid radioactive waste, was conducted. Owing to radiation safety and regulatory issues, an uncontaminated ion-exchange resin, IRN150 H/OH, prior to its use in a nuclear power plant, was used as the material of experiment on synthesis of graphene quantum dots. Since the major radionuclides in spent resin are treated by thermal decomposition, prior to conducting the experiment, carbonization of ion-exchange resin was performed. The experiment on synthesis of graphene quantum dots was conducted according to the general hydrothermal/solvothermal synthesis method as follows. The carbonized ion-exchange resin was added to a solution, which is a mixture of sulfuric acid and nitric acid in ratio of 3:1, and graphene quantum dots were synthesized at 115°C for 48 hours. After synthesizing, procedure, such as purifying, filtering, evaporating were conducted to remove residual acid from the graphene quantum dots. After freeze-drying which is the last procedure, the graphene quantum dots were obtained. The obtained graphene quantum dots were characterized using atomic force microscopy (AFM), Fourier-transform infrared (FT-IR) spectroscopy and Raman spectroscopy. The AFM image demonstrates the topographic morphology of obtained graphene quantum dots, the heights of which range from 0.4 to 3 nm, corresponding to 1–4 graphene layers, and the step height is approximately 2–2.5 nm. Using FT-IR, the functional groups in obtained graphene quantum dots were detected. The stretching vibrations of hydroxyl group at 3,420 cm−1, carboxylic acid (C=O) at 1,751 cm−1, C-OH at 1,445 cm−1, and C-O at 1,054 cm−1. The identified functional groups of obtained graphene quantum dots matched the functional groups which are present if it is a graphene quantum dot. In Raman spectrum, the D and G peaks, which are the characteristics of graphene quantum dots, were detected at wavenumbers of 1,380 cm−1 and 1,580 cm−1, respectively. Thus, it was verified that the graphene quantum dots could be successfully synthesized from the ionexchange resin.