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%.

The density of molten salts is the most important property in the development of molten salt reactor (MSR). The density value measured through the experiment is also very valuable as a gold standard for the validation of the prediction models based on molecular dynamics or other computational methods. To the best of our knowledge, the experimental density data of the ternary NaCl-MgCl2- UCl3 salt system as a MSR candidate fuel salt have never been reported previously. In this study, density measurement experiment of high-temperature molten salt of NaCl-MgCl2 and NaCl-MgCl2- UCl3 was conducted using a previously-developed density measurement system based on the maximum bubble pressure (MPB) method. As a result of the experiment, the density value of 62NaCl- 18MgCl2-20UCl3 molten salt at 873 K was 2.62 g/cm3. A density prediction value of 2.65 g/cm3 at 873 K was derived from the obtained results based on the rule of additivity of molar volume method. The predictred density of 62NaCl-18MgCl2-20UCl3 was consistent with the experimental value within 1%. The density measuring system used in this study is promising for the validation of other multicomponent molten salt systems.

Viscosity of molten salts is an essential property for the thermal hydraulic design and evaluation of molten salt reactor (MSR). Therefore, viscosity data is one of the fundamental physical property data required for safe process operation and countermeasures to severe accidents. In this study, based on our experience of developing a viscosity measurement system for high-temperature LiCl-KCl molten salt system, the viscosity of NaCl-MgCl2 and NaCl-MgCl2-UCl3 molten salts, which are considered promising salts in MSR, was measured. In order to investigate the physical properties of uranium in high-temperature NaCl-MgCl2 molten salt, a viscometer system for high-temperature viscosity measurement was specially designed. As a result of the measurement, the viscosity of the 58NaCl- 42MgCl2 molten salt was 2.73 cP at 838 K, 2.15 cP at 889 K, and 1.68 cP at 940 K. And the viscosity of 73NaCl-21MgCl2-6UCl3 molten salt was 3.79 cP at 877 K, 3.58 cP at 897 K, and 1.63 cP at 941 K. The repeatability of the measurement showed a precision of less than 3%. Although sufficientlyverified starting materials were not used, viscosity data were reported for the first time for NaCl- MgCl2-UCl3 molten salts.

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.

Tritium is a radioactive isotope of hydrogen with a half-life of about 12.3 years, and it is commonly found in the environment as a result of the production of Nuclear Power Plants. The World Health Organization (WHO) has established guidelines for the permissible levels of tritium in drinking water. The guideline value for tritium in drinking water is 10,000 Bq/L. It is important to note that the guideline value for tritium is not a legal limit, but rather a recommendation. National and local authorities may establish legal limits that are more restrictive than the WHO guideline value based on local conditions and risk assessments. The Australia and Finland have set a limit for tritium in drinking water at 76,103 Bq/L and 30,000 Bq/L respectively, which is more than three to seven times higher compare to guideline value of WHO. The United States Environmental Protection Agency (EPA) has set a maximum contaminant level (MCL) for tritium in drinking water at 20,000 picocuries per liter (pCi/L), which is equivalent to 740 Bq/L. The Health Canada has set a guideline value for tritium in drinking water at 7,000 Bq/L. Assuming drinking water corresponding to each tritium limit (or guideline value) for one year, the expected exposure dose is 0.01 mSv to 1 mSv. It means that the tritium in drinking water below the limits or guideline value does not pose a significant risk to human health.

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.

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.

Our research team has developed a gamma ray detector which can be distributed over large area through air transport. Multiple detectors (9 devices per 1 set) are distributed to measure environmental radiation, and information such as the activity and location of the radiation source can be inferred using the measured data. Generally, radiation is usually measured by pointing the detector towards the radioactive sources for efficient measurement. However, the detector developed in this study is placed on the ground by dropping from the drone. Thus, it does not always face toward the radiation source. Also, since it is a remote measurement system, the user cannot know the angle information between the source and detector. Without the angle information, it is impossible to correct the measured value. The most problematic feature is when the backside of the detector (opposite of the scintillator) faces the radiation source. It was confirmed that the measurement value decreased by approximately 50% when the backside of the detector was facing towards the radiation source. To calibrate the measured value, we need the information that can indicate which part of the detector (front, side, back) faces the source. Therefore, in this study, we installed a small gamma sensor on the backside of the detector to find the direction of the detector. Since this sensor has different measurement specifications from the main sensor in terms of the area, type, efficiency and measurement method, the measured values between the two sensors are different. Therefore, we only extract approximate direction using the variation in the measured value ratio of the two sensors. In this study, to verify the applicability of the detector structure and measurement method, the ratio of measured values that change according to the direction of the source was investigated through MCNP simulation. The radioactive source was Cs-137, and the simulation was performed while moving in a semicircular shape with 15 degree steps from 0 degree to 180 degrees at a distance of 20 cm from the center point of the main sensor. Since the MCNP result indicates the probability of generating a pulse for one photon, this value was calculated based on 88.6 μCi to obtain an actual count. Through the ratio of the count values of the two sensors, it was determined whether the radioactive source was located in the front, side, or back of the probe.

Gamma imaging devices that can accurately localize the radioactive contamination could be effectively used during nuclear decommissioning or radioactive waste management. While several hand-held devices have been proposed, their low efficiency due to small sensors have severely limited their application. To overcome this limitation, a high-speed gamma imaging system is under development which comprises two quad-type detectors and a tungsten coded aperture mask. Each quad-type detector consists of four rectangular NaI(Tl) crystals with dimensions of 146×146 mm2 and 72 square-type photomultiplier tubes (PMTs). The detectors are placed in front and back to serve as scatter and absorber, respectively, for Compton imaging. In addition, a coded aperture mask was fabricated in rank 19 modified uniformly redundant array pattern and placed in front of the scatter for coded aperture imaging. The system offers several advanced features including 1) high efficiency achieved by employing large-area NaI(Tl) crystals and 2) broad energy range of imaging by employing a hybrid imaging combining Compton and coded aperture imaging. The imaging performance of the system was evaluated through experiments in various conditions with different gamma energies and source positions. The imaging system provides clear images of the source locations for gamma energies ranging from as low as 59.5 keV (241Am) to as high as 1,330 keV (60Co). The imaging resolution was within the range of 7.5–9.4°, depending on gamma energies, when a hybrid maximum likelihood estimation maximization (MLEM) algorithm was used. The developed system showed high sensitivity, as the 137Cs source at distance, incurring dose rate lower than background level (0.03 μSv/h above background dose rate), could be imaged in approximately 2 seconds. Even under lower dose rate condition (i.e., 0.003 μSv/h above background dose rate), the system was able to image the source within 30 seconds. The system developed in the present study broadens the applicable conditions of the gamma ray imaging in terms of gamma ray energy, dose rate, and imaging speed. The performance demonstrated here suggests a new perspective on radiation imaging in the nuclear decontamination and radioactive waste management field.

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.

Disposal of radioactive waste requires radiological characterization. Carbon-14 (C-14) is a volatile radionuclide with a long half-life, and it is one of the important radionuclides in a radioactive waste management. For the accurate liquid scintillation counter (LSC) analysis of a pure beta-emitting C-14, it should be separated from other beta emitters after extracted from the radioactive wastes since the LSC spectrum signals from C-14 overlaps with those from other beta-emitting nuclides in the extracted solutions. There have been three representative separation methods for the analysis of volatile C-14 such as acid digestion, wet oxidation, and pyrolysis. Each method has its own pros and cons. For example, the acid digestion method is easily accessible, but it involves the use of strong acids and generates large amount of secondary wastes. Moreover, it requires additional time-consuming purification steps and the skillful operators. In this study, more efficient and environment-friendly C-14 analysis method was suggested by adopting the photochemical reactions via in-situ decomposition using UV light source. As an initial step for the demonstration of the feasibility of the proposed method, instead of using radioactive C-14 standards, non-radioactive inorganic and organic standards were investigated to evaluate the recovery of carbon as a preliminary study. These standards were oxidized with chemical oxidants such as H2O2 or K2S2O8 under UV irradiations, and the generated CO2 was collected in Carbo-Sorb E solution. Recovery yield of carbon was measured based on the gravimetric method. As an advanced oxidation process, our photocatalytic oxidation will be promising as a time-saving method with less secondary wastes for the quantitative C-14 analysis in low-level radioactive wastes.

When decommissioning a nuclear power plant, a large amount of radioactive waste is generated simultaneously. Therefore, efficient treatment of radioactive waste is crucial to the success of the decommissioning process. An utility or decommissioning contractor of NPP often build separate radioactive waste treatment facilities (RWTF) to handle this waste. In Korea, RWTFs are planned to be built for the decommissioning of the Kori Unit 1 and Wolsong Unit 1. In this study, we introduce an application case of using process simulation to derive the optimal layout design and investment plan for a radioactive waste treatment facility. In particular, the steam generator is the largest and most complex device processed in RWTF. Therefore, it is necessary to reflect the large equipment processing area that can treat steam generators in the design of RWTF. In this study, Siemens’ Plant Simulation® was used to derive an optimization plan for the dismantling area of large equipment in RWTF. First, a virtual facility was built by modeling based on the steam generator dismantling process and facilities developed by Doosan Enerbility. This was used to pre-validate the facility investment plan, discover wasteful factors in the logistics waste streams, and evaluate alternatives to derive, validate, and apply appropriate improvement alternatives. Through this, we designed a layout based on the optimal logistics waste streams, appropriate workstations, and the number of buffer places. In addition, we propose various optimization measures such as investment optimization based on optimal operation of facility resources such as facilities and manpower, and establishment of work standards.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.