Thermodynamic sorption modeling can enhance confidence in assessing and demonstrating the radionuclide sorption phenomena onto various mineral adsorbents. In this work, Ca-montmorillonite was successfully purified from Bentonil-WRK bentonite by performing the sequential physical and chemical treatments, and its geochemical properties were characterized using X-ray diffraction, Brunauer-Emmett-Teller analysis, cesium-saturation method, and controlled continuous acidbase titration. Further, batch experiments were conducted to evaluate the adsorption properties of Cs(I) and Sr(II) onto the homoionic Ca-montmorillonite under ambient conditions, and the diffuse double layer model-based inverse analysis of sorption data was performed to establish the relevant surface reaction models and obtain corresponding thermodynamic constants. Two types of surface reactions were identified as responsible for the sorption of Cs(I) and Sr(II) onto Ca-montmorillonite: cation exchange at interlayer site and complexation with edge silanol functionality. The thermodynamic sorption modeling provides acceptable representations of the experimental data, and the species distributions calculated using the resulting reaction constants accounts for the predominance of cation exchange mechanism of Cs(I) and Sr(II) under the ambient aqueous conditions. The surface complexation of cationic fission products with silanol group slightly facilitates their sorption at pH > 8.

During the initial cooling period of spent nuclear fuel, Cs-137 and Sr-90 constitute a large portion of the total decay heat. Therefore, separating cesium and strontium from spent nuclear fuel can significantly decrease decay heat and facilitate disposition. This study presents analytical technique based on the gas pressurized extraction chromatography (GPEC) system with cation exchange resin for the separation of Sr, Cs, and Ba. GPEC is a micro-scaled column chromatography system that allows for faster separation and reduction volume of elution solvent compared to conventional column chromatography by utilizing pressurized nitrogen gas. Here, we demonstrate the comparative study of the conventional column chromatography and the GPEC method. Cation exchange resin AG 50W-X12 (200~400 mesh size) was used. The sample was prepared at a 0.8 M hydrochloric acid solution and gradient elution was applied. In this case, we used the natural isotopes 88Sr, 133Cs, and 138Ba instead of radioactive isotopes for the preliminary test. Usually, cesium is difficult to measure with ICP-OES, because its wavelengths (455.531 nm and 459.320 nm) are less sensitive. So, we used ICP-MS to determine the identification and the recovery of eluate. In this study, optimized experimental conditions and analytical result including reproducibility of the recovery, total analysis time and volume of eluents will be discussed by comparing GPEC and conventional column chromatography.

Copper hexacyanoferrate (Cu-HCF), which is a type of Prussian Blue analogue (PBA), possesses a specific lattice structure that allows it to selectively and effectively adsorb cesium with a high capacity. However, its powdery form presents difficulties in terms of recovery when introduced into aqueous environments, and its dispersion in water has the potential to impede sunlight penetration, possibly affecting aquatic ecosystems. To address this, sponge-type aluminum oxide, referred to as alumina foam (AF), was employed as a supporting material. The synthesis was achieved through a dip-coating method, involving the coating of aluminum oxide foam with copper oxide, followed by a reaction with potassium hexacyanoferrate (KHCF), resulting in the in-situ formation of Cu-HCF. Notably, Copper oxide remained chemically stable, which led to the application of 1, 3, 5-benzenetricarboxylic acid (H3BTC) to facilitate its conversion into Cu-HCF. This was necessary to ensure the proper transformation of copper oxide into Cu-HCF on the AF in the presence of KHCF. The synthesis of Cu-HCF from copper oxide using H3BTC was verified through X-ray diffraction (XRD) analysis. The manufactured adsorbent material, referred to as AF@CuHCF, was characterized using Fourier-transform infrared spectroscopy (FTIR) and thermogravimetric analysis (TGA). These analyses revealed the presence of the characteristic C≡N bond at 2,100 cm-1, confirming the existence of Cu-HCF within the AF@CuHCF, accounting for approximately 3.24% of its composition. AF@CuHCF exhibited a maximum adsorption capacity of 34.74 mg/g and demonstrated selective cesium adsorption even in the presence of competing ions such as Na+, K+, Mg2+, and Ca2+. Consequently, AF@CuHCF effectively validated its capabilities to selectively and efficiently adsorb cesium from Cs-contaminating wastewater.

After the Fukushima accident in 2011, relevant concerns regarding the contamination of the natural environment rose abruptly. For example, water contaminated by radionuclides such as Cs and Sr may directly flow into the ocean and threaten the marine ecosystem. In this respect, costeffective and efficient decontamination techniques need to be developed and verified to remediate the contaminated water. Prussian blue (PB) is known as a representative material that can adsorb Cs by ion-trapping and is widely used for medical purposes. However, there is a limitation that PB itself is non-separable and highly mobile in aqueous system, so it needs a fixture, such as bentonite, to be collected after the adsorption. Furthermore, while the performance of PB toward Cs is relatively well known, its behavior toward Sr has rarely been reported. The object of this study is to investigate the sorption characteristics of Cs and Sr onto PB-functionalized bentonite at various conditions. The adsorbent employed in the present work was prepared by mixing bentonite, FeCl3, and K4[Fe(CN)6] at room temperature for 24 hours in the aqueous solution. The concentrations of FeCl3 and K4[Fe(CN)6] were set to a range of 5-200 % compared to the cation exchange capacity of bentonite. After that, the PB-functionalized bentonite was sieved with a mesh size of 63 μm and then reacted with the Cs and Sr solution at various liquid-to-solid (L/S) ratios of 2-10 g/L for up to 500 minutes. Moreover, synthetic seawater containing additional Cs and Sr was reacted with PBfunctionalized bentonite to characterize the ion selectivity of PB. After the completion of the adsorption experiment, a part of the adsorbent was separated and desorption of Cs and Sr with 2 M of nitric acid was performed. For the quantification of aqueous Cs and Sr concentrations, ICP-MS was employed after the filtration with a pore size of 0.45 μm. The result obtained in this study revealed a high sorption affinity of Cs and Sr onto PBfunctionalized bentonite. The analysis results also presented that the sorption reactions of Cs and Sr reached their steady state within 10 minutes of reaction time. Furthermore, the ion selectivity toward Cs and Sr was verified through sorption test with synthetic seawater. According to the high sorption affinity and selectivity, the PB-functionalized bentonite synthesized through this study is expected to be widely used for remediating the Cs- and Sr-contaminated groundwater and seawater, particularly in nuclear waste-relevant industries.

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

A low- and intermediate-level radioactive waste repository contains different types of radionuclides and organic complexing agents. Their chemical interaction in the repository can result in the formation of radionuclide-ligand complexes, leading to their high transport behaviors in the engineered and natural rock barriers. Furthermore, the release of radionuclides from the repository can pose a significant risk to both human health and the environment. This study explores the impact of different experimental conditions on the transport behaviors of 99Tc, 137Cs, and 238U through three types of barrier samples: concrete, sedimentary rock, and granite. To assess the transport behavior of the samples, the geochemical characteristics were determined using X-ray diffraction (XRD), X-ray fluorescence (XRF), Fouriertransform infrared spectroscopy (FTIR), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDS), and Brunauer-Emmett-Teller (BET) analysis. The adsorption distribution coefficient (Kd) was used as an indicator of transport behavior, and it was determined in batch systems under different conditions such as solution pH (ranging from 7 to 13), temperature (ranging from 10 to 40°C), and with the presence of organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA). A support vector machine (SVM) was used to develop a prediction model for the Kd values. It was found that, regardless of the experimental parameters, 99Tc may migrate easily due to its anionic property. Conversely, 137Cs showed low transport behaviors under all tested conditions. The transport behaviors of 238U were impacted by the order of EDTA > NTA> ISA, particularly with the concrete sample. The SVM models can also be used to predict the Kd values of the radionuclides in the event of an accidental release from the repository.

Korea Atomic Energy Research Institute (KAERI) is planning to disposal of the radioactive contaminated cement waste form to the final disposal facility. The final disposal facility require evaluation of immersion, compressive strength, and radionuclide inventory of radioactive wastes to meet the acceptance criteria for safe disposal. According to the LILW acceptance criteria of the Nuclear Safety and Security Commission ok Korea (NSSC), the disposal limit radioactivity of 129I (3.70×101 Bq/g) is lower than other radionuclides. 129I emits low energy as its disposal limit is low, so it is difficult to analyze in the presence of 137Cs and 60Co which emit high energy. Therefore, it is essential to an accurately separate and analyze iodine in radioactive waste. In this study, we focused on the determination of 129I in cement waste form containing 137Cs, 60Co. We added 1 g of 129I(11.084 Bg), 137Cs(1,300 Bq) and 60Co(402 Bq) to cement waste form, respectively. The separation of 129I in cement waste form was carried out using an acid leaching method. And, we confirmed the specific activity of 137Cs and 60Co at each separation step. It was observed that an acid leaching method showed the remove efficiency 137Cs(99.97%) and 60Co(99.94%), respectively. In addition, 129I was also analyzed at approximately 96.44% in simulated contaminated cement waste form. In conclusion, through this experiment, it was confirmed that 129I could be successfully separated and analyzed by using the acid leaching method in cement waste form containing excessive 137Cs and 60Co.

The removal of cesium (Cs) from contaminated clay minerals is still a challenge due to the limited efficiency of the process. Thus, this study aimed to enhance the removal for Cs+ ions during the conventional acid washing process by incorporating a bead-type adsorbent. Polyacrylonitrile-based nickel potassium hexacyanoferrate (NiFC-PAN) was utilized as the Cs adsorbent to selectively adsorb Cs+ ions in a strongly acidic solution that contained competing ions. To enable easy separation of clay particles and protect the adsorbent from harsh environmental conditions, PAN was deliberately constructed as large beads. The synthesized adsorbent (NiFC/PAN in a 2:1 ratio) displayed high selectivity for Cs+ ions and had a maximum capacity of 162.78 mg/g for Cs+ adsorption in 0.5 M HNO3 solution. Since NiFC-PAN exhibited greater Cs selectivity than the clay mineral (hydrobiotite, HBT), adding NiFC-PAN during the acid washing substantially increased Cs desorption (73.3%) by preventing the re-adsorption for Cs+ ions on the HBT. The acid treatment in the presence of NiFCPAN also significantly decreased the radioactivity of 137Cs-HBT from 209 to 27 Bq/g, resulting in a desorption efficiency of 87.1%. Therefore, these findings suggest that the proposed technique is a potentially useful and effective method for decontaminating radioactive clay.

Several tests should be performed to estimate the structural and chemical stability of the radioactive waste. Among the tests in Gyeongju LILW repository, the leaching test which follows ANS 16.1 standard test method should be conducted for Cs, Sr, and Co radionuclides and must satisfy leachability index larger than 6 which applies deionized water as a leachant. However, the expected leachant inside the silo is groundwater that contains various ions and a high pH condition is predicted due to the concrete structures inside the silo. According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. Cobalt precipitates when the leachant has high pH environment which is similar condition to the cement-saturated leachant. Unlike the cobalt, cesium is preferred to exist as ion in the high pH condition. Therefore, the significant effect of the chemical environment of the leachant on the leachability of the radionuclides should be considered for the waste acceptance criteria of the radioactive waste repository. From the ‘NRC, Technical position on the waste form, rev1’, the leaching test method should follow the ANS 16.1 methods by using deionized water as leachant, however, a new leachant showing more aggressive leachability can be applied instead of deionized water. In the other hand, ASTM C1308 leaching test method recommends applying actual groundwater of the repository as a leachant. FT-04-020, the leaching test method of France, suggests the ion composition of the groundwater including the pH value. Therefore, the adequacy of using deionized water as leachant for the leaching test method of Cs, Sr, and Co should be re-examined. In this study, the leaching behavior of Cs, Sr, and Co under the several leachant types is estimated. The cement solidified specimen containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water, simulated groundater, and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type was discussed. The result of this study is expected to be a background data that helps understanding the actual leaching behavior of the Cs, Sr, and Co in the Gyeongju LILW repository.

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

This study evaluated the synthesis of optimal materials for high efficiency adsorption and removal characteristics of Cs-137 for radioactive contaminated water, and considered thermal treatment methods to stabilize the spent adsorbent generated after treatment. We synthesized a composite adsorbent with a combination of impregnating metal ferrocyanide that improves the selectivity of Cs adsorption with zeolite capable of removing Cs as a support. The Cs removal efficiency of the composite adsorbent was evaluated, and the stability change of Cs according to the high-temperature sintering was evaluated as a stabilization method of the spent adsorbent. The metal ferrocyanide content of the adsorbent was in the range of 11.8~36.0%. The adsorption experiments were performed using a simulated liquid waste to have a total Cs concentration of 1 mg/L while containing a trace amount of Cs-137, and then gamma radioactivity was analyzed. In order to evaluate the stabilization of the spent adsorbent, heat treatment was performed in the range of 500~1,100°C, and the volatilization rate of Cs during heat treatment and the leaching rate of Cs after heat treatment were compared. In the adsorption experiment, the Cs removal efficiency was higher than 99%, regardless of the amount of metal ferrocyanide in the composite adsorbent. In the sintering experiment on the spent adsorbent, it was confirmed that there was no volatilization of Cs up to 850°C, and then the volatilization rate increased as the heating temperature increased. On the other hand, the leaching rate of Cs in the sintered adsorbent tends to significantly decrease as the heating temperature increases, so that Cs can be stabilized in the sintered body. In addition, as the content of metal ferrocyanide increases, the volatilization rate of Cs rapidly increases, indicating that the unstable metal ferrocyanide in the adsorbent may adversely affect the removal of Cs as well as the thermal treatment stability.

Korea faces decommissioning the nation’s first commercial nuclear power plant, the Kori-1 and Wolseong-1 reactors. In addition, other nuclear power plants that will continue to operate will also face decommissioning over time, so it is essential to develop independent nuclear facility decommissioning and site remediation technologies. Among these various technologies, soil decontamination is an essential not only in the site remediation after the decommissioning of the highly radioactive nuclear facility, but also in the case of site contamination caused by an accident during operation of the nuclear facility. But the soil, which is a porous material, is difficult to decontaminate because radionuclides are adsorbed into the pores. Therefore, with the current decontamination technology, it is difficult to achieve the two goals of high decontamination efficiency and secondary waste reduction at the same time. In this study, a soil decontamination process with supercritical carbon dioxide as the main solvent was presented, which has better permeability than other solvents and is easy to maintain critical conditions and change physical properties. Through prior research, a polar chelating ligand was introduced as an additive for smooth extraction reaction between radionuclides present as ions in soil and nonpolar supercritical carbon dioxide. In addition, for the purpose of continuity of the process, a candidate group of auxiliary solvents capable of liquefying the ligand was selected. In this research evaluated the decontamination efficiency by adding the selected auxiliary solvent candidates to the supercritical carbon dioxide decontamination process, and ethanol with the best characteristics was selected as the final auxiliary solvent. In addition, based on the decontamination effect under a single condition of the auxiliary solvent found in the Blank Test process, the possibility of a pre-treatment leaching process using alcohol was tested in addition to the decontamination process using supercritical carbon dioxide. Finally, in addition to the existing Cs and Sr, the possibility of decontamination process was tested by adding U nuclides as a source of contamination. As a result of this research, it is expected that by minimizing secondary waste after the process, waste treatment cost could be reduced and the environmental aspect could be contributed, and a virtuous cycle structure could be established through reuse of the separated carbon dioxide solvent. In addition, adding its own extraction capacity of ethanol used for liquefaction of solid-phase ligands is expected to maximize decontamination efficiency in the process of increasing the size of the process in the future.

Radioactive cesium is a heat generated and semi-volitile nuclide in spent nuclear fuel (SNF). It is released gasous phase by head-end treatment which is a pretreatment of pyroprocessing. One of the capturing methods of gasous radioactive cesium is using zeolite. After ion-exchanged zeolite, it is transformed to ceramic waste form which is durable ceramic structure by heat treatment. Various ceramic wasteforms for Cs immobilization have been researched such as cesium aluminosilicate (CsAlSi2O6), cesium zirconium phosphate (CsZr2(PO4)3), cesium titanate (CsxAlxTi8-xO16, Cs2TiNb6O18) and CsZr0.5W1.5O6. The cesium pollucite is composed to aluminosilicate framework and cesium ion incorporated in matrix materials lattices. Many researchers are reported that the pollucite have high chemical durability. In this study, the Cesium pollucite was fabricated using mixtures of aluminosilicate denoted Absorbent product (AP) and Cs2CO3 by calcination and pelletized by cold pressing. The characterization of fabricated pollucite powder and pellets was analyzed by XRD, TGA, SEM, SEMEDS and XRF. The chemical durability of pollucite powder was evaulated by PCT-A and ICP-MS and OES. Thus, the optimal pressure condition without breaking the pellets which is low Cs2O/AP ratio and pelletizing pressure was selected. The long-term leaching test was performed using MCC-1 method for 28 days with the fabricated pollucite pellets. The leachate of leaching test was allard groundwaster and Deionized water and replaced 5 contact periods which is 3 hours, 3 days, 7 days, 14 days and 28 days and analyzed by ICPMS. The leaching rate was shown two stages. The first stage was rapid and relatively large amount of nuclides were leached. The leaching rate was decreased in the second stage. The fractional release rate of this study was shown same trend. These results were similar to previous studies.

Various radionuclides are released and contaminate soils by the nuclear accidents, nuclear tests and disposal of radioactive waste. Among radionuclides, 137Cs is a harmful radioactive element that emits high-energy β particles and γ rays with a half-life of 30.2 years. 137Cs is difficult to extract because it is fixed to soil particles. For the volume reduction technology development of contaminated soil, this study tried to evaluate the irreversible Cs adsorption capacity of granite-originated soil. The soil sample used in the study was collected from C horizon of the soil developed in Mesozoic mica granite. The soil texture, mineralogy, organic content, pH, EC, cation exchange capacity (CEC), water-soluble cation and anion content of the soil samples were determined. A kinetic adsorption experiment and an isotherm adsorption experiment were performed to understand the overall Cs adsorption characteristics using 133Cs. The desorption of Cs by 0.1 mM KCl was also tested for the sample spiked with 133Cs and 137Cs. The soil sample showed a pH of 6.73, EC of 24.50 μS cm-1, and CEC of 1.34 cmolc kg-1, organic matter of 0.53% and sandy loam in texture. Quartz, feldspar and mica were identified as the major mineral components of bulk sample. The clay fraction consists of mica, hydroxyl-interlayer vermiculite (HIV), vermiculite and kaolinite. In the kinetic adsorption experiment, the Cs adsorption showed fast adsorption rates at the initial stage (6 hours) regardless of the 133Cs concentration, and the adsorption equilibrium state was reached after 48 hours. It was the most suitable for the pseudo second-order model. The 133Cs adsorption increased nonlinearly from low to high concentration, which was well match with the dual site Langmuir model. As a result of the desorption experiment, desorption was not performed up to 1.1 mg kg-1 in the presence of competitive ions K+, which is about 0.035% of CEC calculated by the isotherm model. The adsorption of Cs was controlled by frayed edge sites (FES) at a low concentrations and by basal sites or interlayer sites at a high concentration. Irreversible Cs fixation of by FES may be contributed by mainly weathered mica, and when these minerals are separated from the granite origin soil, the possibility of reducing the contamination concentration and volume of radioactive soil waste can be expected.

During electrorefining, fission products, such as Sr and Cs, accumulate in a eutectic LiCl-KCl molten salt and degrade the efficiency of the separation process by generating high heat and decreasing uranium capture. Thus, the removal of the fission products from the molten salt bath is essential for reusing the bath, thereby reducing the additional nuclear waste. While many studies focus on techniques for selective separation of fission products, there are few studies on processing monitoring of those techniques. In-situ monitoring can be used to evaluate separation techniques and determine the integrity of the bath. In this study, laser-induced breakdown spectroscopy (LIBS) was selected as the monitoring technique to measure concentrations of Sr and Cs in 550°C LiCl-KCl molten salt. A laser spectroscopic setup for analyzing high-temperature molten salts in an inert atmosphere was established by coupling an optical path with a glove box. An air blower was installed between the sample and lenses to avoid liquid splashes on surrounding optical products caused by laser-liquid interaction. Before LIBS measurements, experimental parameters such as laser pulse energy, delay time, and gate width were optimized for each element to get the highest signal-to-noise ratio of characteristic elemental peaks. LIBS spectra were recorded with the optimized conditions from LiCl-KCl samples, including individual elements in a wide concentration range. Then, the limit of detections (LODs) for Sr and Cs were calculated using calibration curves, which have high linearity with low errors. In addition to the univariate analysis, partial least-squares regression (PLSR) was employed on the data plots to obtain calibration models for better quantitative analysis. The developed models show high performances with the regression coefficient R2 close to one and root-mean-square error close to zero. After the individual element analysis, the same process was performed on samples where Sr and Cs were dissolved in molten salt simultaneously. The results also show low-ppm LODs and an excellent fitted regression model. This study illustrates the feasibility of applying LIBS to process monitoring in pyroprocessing to minimize nuclear waste. Furthermore, this high-sensitive spectroscopic system is expected to be used for coolant monitoring in advanced reactors such as molten salt reactors.

In this study, the positions of Cs-137 gamma ray source are estimated from the plastic scintillating fiber bundle sensor with length of 5 m, using machine learning data analysis. Seven strands of plastic scintillating fibers are bundled by black shrink tube and two photomultiplier tubes are used as a gamma ray sensing and light measuring devices, respectively. The dose rate of Cs-137 used in this study is 6 μSv·h−1. For the machine learning modeling, Keras framework in a Python environment is used. The algorithm chosen to construct machine learning model is regression with 15,000 number of nodes in each hidden layer. The pulse-shaped signals measured by photomultiplier tubes are saved as discrete digits and each pulse data consists of 1,024 number of them. Measurements are conducted separately to create machine learning data used in training and test processes. Measurement times were different for obtaining training and test data which were 1 minute and 5 seconds, respectively. It is because sufficient number of data are needed in case of training data, while the measurement time of test data implies the actual measuring time. The machine learning model is designated to estimate the source positions using the information about time difference of the pulses which are created simultaneously by the interaction of gamma ray and plastic scintillating fiber sensor. To evaluate whether the double-trained machine learning model shows enhancement in accuracy of source position estimation, the reference model is constructed using training data with one-time learning process. The double-trained machine learning model is designed to construct first model and create a second training data using the training error and predetermined coefficient. The second training data are used to construct a final model. Both reference model and double-trained models constructed with different coefficients are evaluated with test data. The evaluation result shows that the average values calculated for all measured position in each model are different from 7.21 to 1.44 cm. As a result, by constructing the double-trained machine learning model, the final accuracy shows 80% of improvement ratio. Further study will be conducted to evaluate whether the double-trained machine learning model is applicable to other data obtained from measurement of gamma ray sources with different energy and set a methodology to find optimal coefficient.

Concrete is one of the largest wastes, by volume, generated during the decommissioning of nuclear facilities, which significantly influences the projected costs for the disposal of decommissioning wastes. Concrete consists of aggregates and a cement binder. In radioactive concrete, the radioisotopes are mainly associated with the cement component. If the radioactive isotope can be separated from the concrete to below the clearance criteria, the volume of radioactive concrete waste could be reduced effectively. We were studied to separate the radioactive materials from the concrete by using the thermomechanical and chemical treatment processes, sequentially. From the study, separated aggregate could be treated to achieve the clearance level. However, these processes generate a large volume of secondary acidic radioactive wastewater, which might be a critical problem to reduce the volume of radioactive concrete waste. In this research, separating the 137Cs and 90Sr from dissolved concrete wastewater to below the discharge criteria by precipitation method, it would be released to the environment under industrial waste guidelines. The experiments were conducted to using a simulated radioactive wastewater, formed by the dissolution of concrete within HCl, which was spiking the 137Cs and 90Sr, respectively. In addition, we applied the chemical precipitation methods with wastewater, using ferrocyanide for 137Cs and BaSO4 coprecipitation for 90Sr. As a result, targeted radionuclides could be removed to the discharge level (137Cs: 0.05 Bq·ml−1, 90Sr: 0.02 Bq·ml−1) by precipitation method. Therefore, it could reduce the secondary wastewater effectively by precipitation method and enhance the additional volume reduction for radioactive concrete waste.

Radioactive Cesium is fission products of spent nuclear fuelwith high heat generating nuclide, having a 30 years half-life. Particularly, it is important to make stable waste form because Cs-137 have high solubility and mobility at ground water. The ceramic waste form has higher thermal and structural stability and lower solubility than glass and cement waste form. Various ceramic waste forms for Cs immobilization have been researched such as aluminosilicate (CsAlSi2O6), phosphate (CsZr2(PO4)3), titanate (CsxAlxTi8-XO16) and CsZr0.4W1.5O6. Cs pollucite is incorporated radio-Cesium to aluminosilicate framework by inorganic ion-exchange with zeolite. Therefore, it is an extremely stable structure. In previous study, we are prepared Cs pollucite pellet with various ratio of Cs precursor/matrix materials, and attempted to evaluate applicability as ceramic waste form. Cs pollucite is produced by mixing Mullite and SiO2 obtained by heat treatment Kaolinite with Cs2CO3 in ratios of 0.5, 0.6, 0.7, 0.8. Optimized ratio was 0.5 revealed single pollucite phase and the others exhibited CsAlSiO4 phase with pollucite. Cs pollucite of ratio 0.5 was pelletized under various conditions and evaluated performance as waste form. herein, the pellets were cracked on surface and edges broken. Therefore, Cs pollucite having high ratio of matrix materials contained Si and Al was prepared and pelletized, and then waste form was evaluated. The Cs pollucite powder is ratio of Cs precursor/matrix materials were 0.1, 0.2, 0.3, 0.4. Pollucite powder was mixed with 1.5, 2.0wt% Polyvinyl alcohol as binder, and dried at 70°C for overnight. Afterward, these powders obtained were pressed using punch-die apparatus at 50, 100 bar for 1 hour and the pellets with about dia. 25 mm and height 10 mm was acquired. These pellets were sintered at 1,400°C for 5 hours. Subsequently, the waste forms were evaluated physicochemical test such as compression strength, thermal conductivity, thermal expansion and leaching properties analysis.