Biomass-derived porous carbon is an excellent scientific and technologically interesting material for supercapacitor applications. In this study, we developed biomass-derived nitrogen-doped porous carbon nanosheets (BDPCNS) from cedar cone biomass using a simple KOH activation and pyrolysis method. The BDPCNS was effectively modified at different temperatures of 600 °C, 700 °C, and 800 ℃ under similar conditions. The as-prepared BDPCNS-700 electrode exhibited a high BET surface area of 2883 m2 g− 1 and a total pore volume of 1.26 cm3 g− 1. Additionally, BDPCNS-700 had the highest electrical conductivity (11.03 cm− 1) and highest N-doped content among the different electrode materials. The BDPCNS-700 electrode attained a specific capacitance of 290 F g− 1 at a current density of 1 A g− 1 in a 3 M KOH electrolyte and an excellent longterm electrochemical cycling stability of 93.4% over 1000 cycles. Moreover, the BDPCNS-700 electrode had an excellent energy density (40.27 Wh kg− 1) vs power density (208.19 W kg− 1). These findings indicate that BDPCNS with large surface areas are promising electrode materials for supercapacitors and energy storage systems.

This study was performed to evaluate the separation of Sr, Cs, Ba, La, Ce, and Nd using gas pressurized extraction chromatography (GPEC) with anion exchange resin for the quantitation of Neodymium. GPEC is a micro-scaled column chromatography system that provides a constant flow rate by utilizing nitrogen gas. It is overcome the disadvantages of conventional column chromatography by reducing the volume of elution solvent and shortening the analysis time. Here, we compared the conventional column chromatography and the GPEC method. The whole analysis time was decreased by nine times and radioactive wastes were reduced by five times using the GPEC system. Anion exchange resin 1-X4 (200~400 mesh size) was used. The sample was prepared at a 0.8 M nitric acid in methanol solution. The elution solvent was used at a 0.01 M nitric acid in methanol solution. Finally the eluate was analyzed by ICP-MS to determine the identification and recovery. In this case, we applied the natural isotopes of LREEs (139La, 140Ce, and 144Nd) and high activity nuclides (88Sr, 133Cs, and 138Ba) instead of radioactive isotopes for the preliminary test; as a result, unnecessary radioactive waste was not produced. The recoveries were 93.9%, 105.9%, 91.9%, 47.6%, 35.9%, and 79.9% of Sr, Cs, Ba, La, Ce, and Nd, respectively. The reproducibility of recoveries by GPEC were in the range 2.8%–10.9%.

Molten salts have gained significant attention as a potential medium for heat transfer or energy storage and as liquid nuclear fuel, owing to their superior thermal properties. Various fluoride- and chloride-based salts are being explored as potential liquid fuels for several types of molten salt reactors (MSRs). Among these, chloride-based salts have recently received attention in MSR development due to their high solubility in actinides, which has the potential to increase fuel burnup and reduce nuclear water production. Accurate knowledge of the thermal physical properties of molten salts, such as density, viscosity, thermal conductivity, and heat capacity, is critical for the design, licensing, and operation of MSRs. Various experimental techniques have been used to determine the thermal properties of molten salts, and more recently, computational methods such as molecular dynamics simulations have also been utilized to predict these properties. However, information on the thermal physical properties of salts containing actinides is still limited and unreliable. In this study, we analyzed the available thermal physical property database of chloride salts to develop accurate models and simulations that can predict the behavior of molten salts under various operating conditions. Furthermore, we conducted experiments to improve our understanding of the behavior of molten salts. The results of this study are expected to contribute to the development of safer and more efficient MSRs.

Since 2018, Central Research Institute of Korea Hydro & Nuclear Power (KHNP–CRI) has been operating an X-ray irradiation system with a maximum voltage of 160 kV and 320 kV X-ray tube to test personal dosimeters in accordance with ANSI N13.11-2009 “Personnel Dosimetry Performance- Criteria for Testing”. This standard requires that dosimeters for the photon category testing be irradiated with the X-ray beams appropriate to the ISO beam quality requirements. KHNP-CRI has implemented the fourteen X-ray reference radiation beams in compliance with ISO-4037-1, 2, and 3. When installing the X-ray irradiation system, KHNP-CRI evaluated the uncertainties of dose conversion coefficients for deep and shallow doses, based on “Catalogue of X-ray spectra and their characteristic data – ISO and DIN radiation qualities, therapy and diagnostic radiation qualities, unfiltered X-ray spectra” published by Physikalisch Technische Bundesanstalt (PTB). A CdTe detector (X-123, AMPTEK) with disk type collimators made of tungsten was used to acquire X-ray spectra. The detector was located at 1 m from the center of the target material in the Xray tubes. Six uncertainty factors for the dose conversion coefficients for the fourteen X-ray beams were chosen as follows; the minimum and maximum cut-off energies Emin and Emax, the air density (ρ), the accuracy of the high-voltage of the X-ray tube, statistics of the pulse height spectra and the unfolding method. For example, uncertainty of each quantity for a HK30 beam was calculated to be 0.3%, 2.32%, 0.19%, 1.25%, and 0.13%, and 0.18%, respectively. The combined standard uncertainty for the deep dose conversion coefficient of the HK30 beam was calculated to be 2.67%. The coverage factor corresponding to a 95 percent confidence interval was obtained as k = 1.8 using a Monte Carlo method, which is slightly lower the coverage factor of k = 1.95 for a Gaussian distribution. This seems to result from that two dominant uncertainties, the unfolding uncertainty and minimum cut-off energy uncertainty, follow a rectangular distribution.

As a result of various generation, transmutation, and decay schemes, a wide variety of radionuclides exist in the reactor prior to accident occurrence. Considering all of the radionuclides as the accident source term in an offsite consequence analysis will inevitably take up excessive computer resources and time. Calculation time can be reduced with minimal impact on the accuracy of the results by considering only the nuclides that have a significant effect on the calculation among the potential radioactive sources that may be released into the environment. In earlier studies related to offsite consequence analysis, it is shown that the principal criteria for the radionuclide screening applied are as follows; radionuclide inventory in the reactor, radioactive half-life, radionuclide release fraction to the environment, relative dose contribution of nuclides within a specific group, and radiobiological importance. As a result, it is confirmed that 54, 60, and 69 nuclides are applied to the risk assessment performed in WASH-1400, NUREG-1150, and SOARCA (State-of-the-Art Reactor Consequence Analyses) project in the United States, respectively. In addition, in this study, the technical consultations with domestic and foreign experts were carried out to confirm details on criteria and process for screening out radionuclides in offsite consequence analysis. In this paper, based on the literature survey and technical consulting, we derived the screening process of selecting a list of radionuclides to be considered in the offsite consequence analysis. The first step is to eliminate radionuclides with little core inventory (less than specific threshold) or very short half-lives. However, important decay products of radionuclides that have short half-lives should not be excluded by this process. The next step is to further eliminate radionuclides by considering contribution to offsite impact, which is defined as a product of radioactivity released to the environment (i.e. ‘inventory in the reactor’ times ‘release fraction to offsite’) and comprehensive dose (or risk) coefficient taking into account all exposure pathways to be included. The final step is to delete isotopes that contribute less than certain threshold to any important dose metric through additional computer runs for each important source term. Even though it is presumed that this process is applicable to existing light water reactors and the set of accidents that would be considered in PSA, some of the assumptions or specific recommendations may need to be reconsidered for other reactor types or set of accident categories.

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

During decommissioning and site remediation of nuclear power plant, large amount of wastes (including radioactive waste) with various type will be generated within very short time. Among those wastes, soil and concrete wastes is known to account for more than 70% of total waste generated. So, efficient management of these wastes is very essential for effective NPP decommissioning. Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes from the generation. The system is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. By adopting modular type, the system is good for dealing with variable situation where system capacity needs to be expanded or contracted depending on the decommissioning schedule, good for minimizing secondary waste generated during maintenance of failed part and also good for disassemble, transfer and assemble. The contamination evaluation module of the system has two sub module. One is for quick measurement with NaI(Tl) detector and the other is for accurate measurement with HPGe detector. For waste transfer, the system adopts LTS (Linear Transfer System) conveyor system showing low vibration and noise during operation. This will be helpful for minimizing scattering of dust from the waste container. And for real time positioning of waste container, wireless tag was adopted. The tag also used for information management of waste history from the generation. Once a container with about 100 kg of soil or concrete is loaded, it is moved to the weight measurement module and then it transfers to quick measurement module. When measured value for radioactivity concentration of Co- 60 and Cs-137 is more than 1.0 Bq/g, then the container is classified as waste for disposal and directly transferred to stacking and storage rack. Otherwise, the container is transferred to accurate measurement module. At the accurate module, the container is classified as waste for disposal or waste for regulatory clearance depending on the measurement result of 0.1 Bq/g. As the storage rack has a sections for disposal and regulatory clearance respectively, the classified containers will be positioned at one of the sections depending on the results from the contamination evaluation module. The system can control the movement of lots of container at the same time. So, the system will be helpful for the effective nuclear power plant decommissioning in view of time and budget.

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

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

During the operation of the nuclear power plant, various radioactive waste are generated. The spent resin, boron concentrates, and DAW are classified as a generic radioactive waste. They are treated and stored at radioactive waste building. In the reactor vessel, different types of radioactive waste are generated. Since the materials used in reactor core region exposed to high concentration of neutrons, they exhibit higher level of surface dose rate and specific activity. And they are usually stored in spent fuel pool with spent fuel. Various non-fuel radioactive wastes are stored in spent fuel pool, which are skeleton, control rod assembly, burnable neutron absorber, neutron source, in core detector, etc. The skeleton is composed of stainless 304 and Inconel-718. There are two types of control rod assembly, that are WH type and OPR type. The WH type control rod is composed of Ag-In-Cd composites. The OPR type control rod is composed of B4C and Inconel-625. In this paper, the characteristics and storage status of the non-fuel radioactive waste will be reported. Also, the management strategy for the various non-fuel radioactive waste will be discussed.

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.

Metakaolin-based geopolymers have shown promise as suitable candidates for 14C immobilization and final disposal. It has been shown that the physicochemical properties of metakaolin wasteforms meet, and often far exceeding, the strict compression strength and leaching acceptance criteria of the South Korea radioactive waste disposal site. However, it is not possible to analyze and characterize the internal structure of the geopolymer wasteform by conventional characterization techniques such as microscopy without destruction of the wasteform; an impractical solution for inspecting wasteforms destined for final disposal. Internal inspection is important for ensuring wastes are homogenously mixed throughout the wasteform and that the wasteform itself does not pose any significant defects that may have formed either during formulation and curing or as a result of testing prior to final disposal. X-ray Computed Tomography (XCT) enables Non-Destructive Evaluation (NDE) of objects, such as final wasteforms, allowing for both their internal and external, characterization without destruction. However, for accurate quantification of an objects dimensions the spatial resolution (length and volume measures) must be know to a high degree of precision and accuracy. This often requires extensive knowledge of the equipment being used, its precise set-up, maintenance and calibration, as well as expert operation to yield the best results. A spatial resolution target consists of manufactured defects of uniformed dimensions and geometries which can be measured to a high degree of accuracy. Implementing the use of a spatial resolution target, the dimensions of which are known and certified independently, would allow for rapid dimensional calibration of XCT systems for the purpose of object metrology. However, for a spatial resolution target to be practical it should be made of the same material as the intended specimen, or at least exhibit comparable X-ray attenuation. In this study, attempts have been made to manufacture spatial resolution targets using geopolymer, silica glass, and alumina rods, as well as 3D printed materials with varying degrees of success. The metakaolin was activated by an alkaline activator KOH to from a geopolymer paste that was moulded into a cylinder (Diameter approx. 25 mm). The solidified geopolymer cylinder as well as both the silica glass rod and alumina rod (Diameter approx. 25 mm) we cut to approximately 4 mm ± 0.5 mm height with additional end caps cut measuring 17.5 mm ± 2.5 mm height. All parts were then polished to a high finish and visually inspected for their suitability as spatial resolution targets.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

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.

It is important to make a strategy for clearance-level radioactive waste. Sampling and disposal plans should be drawn up with characteristics of target waste. In this paper, a target clearance-level radioactive waste is used in a laboratory for experiments with Cs-137 and Co-60, unsealed radioactive sources with gamma radiation isotopes. Therefore, it is enough to analyze with HPGe to check the contaminant level. The laboratory fume hood combined multiple materials, which means some are volume contamination and others are surface contamination. The wood, plastic, and drywall boards, which are absorbent volume contaminated parts and make up PVC pipes, base cabinet doors, backside baffles, etc., will be sampled with coring methods. The metals and glasses, which are unabsorbent, surface-contaminated parts, are sampled with smear methods. The work surface, baffles, exhaust plenum, and glass sash inside parts have a high possibility of being contaminated. The hood body, flame, base cabinet, PVC pipe (the rare end of the filter), and blower transition case have a low possibility of becoming contaminated. When we checked with HPGe, except for the work surface (which was below clearance level), other parts were less than MDA. The highest radionuclide concentration was in PVC pipe: Cs-137C 3.91E-02 (Bq/g), Co-60 4.54E- 03 (Bq/g). It is less than clearance level. Therefore, the waste was applied for the clearance level radioactive wastes and got permission from the regulatory body.

Level measurement of liquid radwaste is essential for inventory management of treatment system. Among various methods, level measurement based on differential pressure has many advantages. First, it is possible to measure the liquid level of the system regardless of liquid type. Second, as the instrument doesn’t need to be installed near the tank, there is no need to contact the tank when managing it. Therefore, workers’ radiation dose from the system can be decreased. Finally, although it depends on the accuracy, the price of the instrument is relatively low. With these advantages, in general, liquid radwaste level in a tank is measured using differential pressure in the treatment system. Not only the advantages described above, there are some disadvantages. As the liquid in the system is waste, it is not pure but has some suspended materials. These materials can be accumulated in tanks and pipes where the liquids move to come into direct contact with pneumatic pipes that are essential in differential pressure instruments. As a result, in case of a treatment using heat source, the accumulated materials may become sludge causing interference in pneumatic pipes. And this can change the pressure which also affects the level measured. In conclusion, in case of liquid storage tanks in which the situation cannot be checked, the proficiency of an operator becomes important.

Surface water temperature of a bay (from the south to the north) increases in spring and summer, but decreases in autumn and winter. Due to shallow water depth, freshwater outflow, and weak current, the water temperature in the central to northern part of the bay is greatly affected by the land coast and air temperature, with large fluctuations. Water temperature variations are large in the north-east coast of the bay, but small in the south-west coast. The difference between water temperature and air temperature is greater in winter and in the south-central part of the bay than that in the north to the eastern coast of the bay where sea dykes are located. As the bay goes from south to north, the range of water temperature fluctuation and the phase show increases. When fresh water is released from the sea dike, the surrounding water temperature decreases and then rises, or rises and then falls. The first mode of empirical orthogonal function (EOF) represents seasonal variation of water temperature. The second mode represents the variability of water temperature gradient in east-west and north-south directions of the bay. In the first mode, the maximum and the minimum are shown in autumn and summer, respectively, consistent with seasonal distribution of surface water temperature variance. In the second mode, phases of the coast of Seosan~Boryeong and the east coast of Anmyeon Island are opposite to each other, bordering the center of the deep bay. Periodic fluctuation of the first mode time coefficient dominates in the one-day and half-day cycle. Its daily fluctuation pattern is similar to air temperature variation. Sea conditions and topographical characteristics excluding air temperature are factors contributing to the variation of the second mode time coefficient.