Kale (Brassica oleracea var. acephala) is one of the most frequently consumed leafy vegetables globally, as it contains numerous nutrients; essential amino acids, phenolics, vitamins, and minerals, and is particularly rich in glucosinolates. However, the differences in the biosynthesis of glucosinolates and related gene expression among kale cultivars has been poorly reported. In this study, we investigated glucosinolates profile and content in three different kale cultivars, including green (‘Man-Choo’ and ‘Mat-Jjang’) and red kale (‘Red-Curled’) cultivars grown in a vertical farm, using transcriptomic and metabolomic analyses. The growth and development of the green kale cultivars were higher than those of the red kale cultivar at 6 weeks after cultivation. High-performance liquid chromatography (HPLC) analysis revealed five glucosinolates in the ‘Man-Choo’ cultivar, and four glucosinolates in the ‘Mat-Jjang’ and ‘Red-Curled’ cultivars. Glucobrassicin was the most predominant glucosinolate followed by gluconastrutiin in all the cultivars. In contrast, other glucosinolates were highly dependent to the genotypes. The highest total glucosinolates was found in the ‘Red-Curled’ cultivar, which followed by ‘Man-Choo’ and ‘Mat-Jjang’. Based on transcriptome analysis, eight genes were involved in glucosinolate biosynthesis. The overall results suggest that the glucosinolate content and accumulation patterns differ according to the kale cultivar and differential expression of glucosinolate biosynthetic genes.

Antioxidants are food additives that extend the shelf life of food products by preventing lipid rancidity caused by active oxygen. They can either be naturally-derived or manufactured synthetically via chemical synthesis. In this study, method validation of five synthetic antioxidants, namely butylated hydroxyanisole, butylated hydroxytoluene, tertiary butylhydroquinone, propyl gallate, and disodium ethylenediaminetetraacetic acid, was performed using a high performance liquid chromatography–ultraviolet visible detector, and the method applicability was evaluated by analyzing foods containing antioxidants. The coefficient of determination (R2) average was 0.9997, while the limit of detection and limit of quantification were 0.02–0.53 and 0.07–1.61 mg/kg, respectively. The intra and inter-day accuracies and precisions were 83.2±0.7%–98.7±2.1% and 0.1%–5.7% RSD, respectively. Inter-laboratory validation for accuracy and precision was conducted using the Food Analysis Performance Assessment Scheme quality control material. The results satisfied the guidelines presented by the AOAC International. In addition, the expanded uncertainty was less than 16%, as recommended by CODEX. Consequently, to enhance public health safety, the results of this study can be used as basis data for evaluating the intake of synthetic antioxidants and assessing their risks in Korea.

Viscosity is a fundamental physical property that is important in any system in which fluid movement occurs. In addition, most of the elements exist as ions in molten state in high-temperature molten salt, and electrical conductivity in such molten state is closely related to viscosity as a transport property. Molten salt reactor (MSR) and pyroprocess are representative processes dealing with high-temperature molten salts, actinide elements, and other radioactive materials. In MSR and pyroprocesses, the viscosity data must be provided as one of the fundamental physical property data required for safe process operations and countermeasures to severe accidents. In order to measure the viscosity of highly corrosive molten salt at high temperatures, we have built a in-house developed molten salt viscosity measurement system based on the Brookfield rotationary viscometer. We also developed a special correction technique to improve the accuracy of the viscosity measurement. In this study, the viscosity was measured at 500°C for NaCl-MgCl2 molten salt, which is selected as the base salt material of MSR system under development in Korea Atomic Energy Research Institute (KAERI), using our viscosity measurement system installed in a oxygen- and moisture-free Ar-atmosphere glovebox. Our viscosity measurement system was calibrated using a LiCl-KCl eutectic mixture with well-known viscosity value, and viscosity values obtained using our own correction methodology were compared with those of other conventional correction methods. In our further study, we plan to measure the NaCl-MgCl2-UCl3 system at various compositions and temperatures.

Nuclear spent fuel (SNF) disposal in deep geological repositories is considered as one of sound options for the long-term and safe sequestration of radiotoxic SNF and the sustainable use of nuclear energy. The chemical behaviors of various radionuclides originated from SNF should be well understood to evaluate the migrational behaviors of radionuclides and their reactions and interactions with various geochemical components. Formation of secondary minerals, colloids, other insoluble precipitates is of interest since the concentrations of radionuclides in groundwaters can be limited by the solubility of those solid phases. Particularly when evaluating their solubility, the use of well-defined solid materials in terms of chemical composition and molecular structure is crucial to obtain reliable measurement results. In this study, a synthetic calcium uranyl silicate (Ca-U(VI)-silicate, or uranophane) was prepared and characterized by using various analytical methods including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational (FTIR and Raman) spectroscopies. Uranyl silicate minerals are significant to the disposal of nuclear wastes. Our simulation demonstrates that uranophane (Ca[UO2SiO3OH]2·5H2O), one having a U:Si ratio of 1:1, can be a mineral species limiting U(VI) solubility under groundwater conditions in Korea. For the preparation of Ca-U(VI)-silicate, we applied a two-step hydrothermal synthetic procedure reported in literature with modification. Briefly, we conclude that the obtained mineral phase is the ‘α-uranophane’; our characterization results show that the structural and spectroscopic properties of the synthetic Ca-U(VI)-silicate agree well with those of α-uranophane. For instance, the pXRD patterns obtained from the solid show nearly identical diffraction peak positions with those from the reference XRD pattern. From IR and Raman spectroscopy it is noticed that the stretching modes of UO2 2+ and SiO4 4- ions result in strong absorption bands in a region of 700 ~ 1,100 cm-1. Elemental compositions of the synthetic solids were also estimated by using EDX analysis, which results in a Ca:U:Si ratio close to 1:2:2 on average. However, we found that it is difficult to obtain good crystallinity of uranophane, which can be observable by using SEM and its image analysis. We believe that this work serves as a model study to provide synthetic routes of radionuclide-related mineral phases and applicable solid phase characterization methods. In the presentation, the potential use of the U(VI)-silicate solid phase for the upcoming groundwater solubility measurements will be discussed. Keywords: Hexavalent Uranium, Silicate

This study presents a rapid and quantitative radiochemical separation method for Nb isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin after Fe coprecipitation. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the Nb isotopes were coprecipitated with Fe after filtering the leaching solution with 0.45 micron HA filter, while the Sr, Tc and Ni isotopes were in the solution. The Nb isotopes were separated in HCl medium with anion exchange resin. The purified Nb isotopes were measured using a low level liquid scintillation counter after installing quenching curve with standard Nb-94 isotopes. The separation method for Nb isotopes investigated in this study was applied to neutron dosimeter samples from the nuclear power plant after validating the Nb activity concentration with gamma spectrometry system.

The sorption/adsorption behavior of radionuclides, usually occurring at the solid-water interface, is considered to be one of the primary reactions that can hinder the migration of radiotoxic elements contained in the spent nuclear fuel. In general, various physicochemical properties such as surface area, cation exchange capacity, type of radionuclides, solid-to-liquid ratio, aqueous concentration, etc. are known to provide a significant influence on the sorption/adsorption characteristics of target radionuclides onto the mineral surfaces. Therefore, the distribution coefficient, Kd, inherently shows a conditiondependent behavior according to those highly complicated chemical reactions at the solid-water interfaces. Even though a comprehensive understanding of the sorption behavior of radionuclides is significantly required for reliable safety assessment modeling, the number of the chemical thermodynamic model that can precisely predict the sorption/adsorption behavior of radionuclides is very limited. The machine-learning based approaches such as random forest, artificial neural networks, etc. provide an alternative way to understand and estimate complicated chemical reactions under arbitrarily given conditions. In this respect, the objective of this study is to predict the sorption characteristics of various radionuclides onto major bentonite minerals, as backfill materials for the HLW repository, in terms of the distribution coefficient by using a machine-learning based computational approach. As a background dataset, the sorption database previously established by the JAEA was employed for random forest machine learning calculation. Moreover, the hyperparameters such as the number of decision trees, the number of variables to divide each node, and random seed numbers were controlled to assess the coefficient of determination, R2, and the final calculation result. The result obtained in this study indicates that the distribution coefficients of various radionuclides onto bentonite minerals can be reliably predicted by using the machine learning model and sorption database.

Currently, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated in Korea. For the disposal of the radioactive wastes, the transport demand is expected to increase. Prior to transportation, it is necessary to evaluate the radiation risk of transportation to confirm that is not high. In Korea, there is no transportation risk assessment code that reflects domestic characteristics. Therefore, foreign assessment codes are used. In this study, before developing the overland transportation risk assessment code that reflects domestic characteristics, we analyzed the radiation risk assessment methodology in transportation accident codes developed in other countries. RADTRAN and RISKIND codes were selected as representative overland transportation risk assessment codes. For the two codes we analyzed accident scenarios, exposure pathways, and atmospheric diffusion. In RADTRAN, the user classifies accident severity for possible accident scenarios, and the user inputs the probability for each accident severity. On the other hand, in the case of RISKIND, the accident scenarios are classified and the probabilities are determined according to the NRC modal study (LLNL, 1987) in consideration of the cask impact velocity, cask impact angle, and fire temperature. In the case of RISKIND, the accident scenarios are applied only to transportation of spent nuclear fuel, and cannot be defined for low and intermediate-level radioactive waste. However, in the case of RADTRAN, since the severity and probability of accidents are defined by user, it can be applied to low and intermediate-level radioactive wastes. As the exposure pathways considered in transportation accident, both RADTRAN and RISKIND consider external exposure (cloudshine and groundshine), and internal exposure (inhalation, resuspension inhalation and ingestion). In the case of RADTRAN, additionally, external exposure due to loss of shielding (LOS) is considered. Atmospheric diffusion calculation is essential to determine the extent to which radioactive materials are diffused. In both RADTRAN and RISKIND, atmospheric diffusion calculations are based on Gaussian diffusion model. Users must input Pasquill stability class, release height, heat release, wind speed, temperature and mixing height, etc. Additionally, RADTRAN can input weather information relatively simply by inputting only the Pasquill stability class fraction and selecting the US average weather option. This study results will be used as a basis for developing radioactive waste overland transportation risk assessment code that reflects domestic characteristics.

In accordance with the notification of the Nuclear Safety and Security Commission (NSSC), environmental impact assessments around nuclear power plants are conducted annually and the results are disclosed to the public. The effects of direct radiation exposure from nuclear power plants as well as liquid effluents and gaseous effluents are taken into consideration in the evaluation of dose calculation for residents. In the United States, regulatory guidelines on direct radiation exposure are described in Reg. Guide 4.1, and the effects of direct radiation are evaluated through regulatory guidelines in Korea. We are going to review optimal evaluation method by reviewing the direct radiation exposure evaluation method currently being conducted in domestic nuclear power plants and the direct radiation exposure evaluation method in overseas nuclear power plants such as in the United States.

The decommissioning of a nuclear power plant is a project that consists of several stages, and various technologies are applied when performing various tasks at each stage. And it is essential to secure safety and economic feasibility. As the paradigm has changed due to digital transformation in various industries, digitalization is applied to the life cycle of nuclear power plant from construction, operation and decommissioning project. Element technologies are being developed for decommissioning plan establishment, process design, econtamination method, decommissioning work process, waste management, environmental monitoring and radiation dose simulation. The utilization of digital twin in the decommissioning stage is classified into three categories. ① Process Monitoring (decommissioning work procedure, work progress (plan/actual), real-time work status and etc.) ② Facility Monitoring (real-time sensing and video data monitoring, decommissioning SSCs information, work alarm and etc.) ③ Safety Monitoring (work safety, radiation exposure, fire monitoring, work risk and etc.) A system suitable for the decommissioning stage and work should be developed in consideration of the target of use, development function, and when to create data according to the purpose of the system. Simulation module according to user purpose should be provided. In addition, data-base management should be performed according to the decommissioning characteristics in consideration of the data associated with the existing operating system. The system to be developed should support the project management to comply with the domestic standards and regulations to be determined in the future. This will improve the competitiveness of domestic and foreign markets.

It is likely to occur internal exposure for workers in Nuclear Power Plants (NPPs) due to the intake of radionuclide. To assess the internal exposure dose the measurement of activity for remain radionuclide is necessary. The Whole Body Counters (WBCs) are commonly used for measurement of remain radionuclide activity in human body. Korea Hydro & Nuclear Power Co., Ltd. (KHNP) conduct performance test of WBCs in all NPPs for every year to confirm the performance of equipment. The performance test is conducted using unknown sources and the participants of the comparison test submit the radionuclide and activity of the unknown sources measured by WBC as a result. The performance indicator and criteria for WBC recommended in the American National Standards Institute (ANSI) N13.30 report published in 2011 are applied. The performance indicator is Root Mean Squared Error (RMSE) and criteria is 0.25 or less. The results of performance test performed in 2022 for all WBC is meet the ANSI N13.30 criteria. And the RMSE values are confirmed from 0.01 to 0.23. This means that the residual radioactivity measurement results using WBC are reliable.

When the leakage of radioactive material or radiation to the environment or a concern, it is important to accurately understand the impact on the environment. Therefore, environmental effects evaluation using modeling based on meteorological data and source-term data is carried out, or environmental radiation monitoring which is an emergency response activity that directly measures dose is performed. As lessons learned from the Fukushima accident, environmental effects evaluation and modeling cannot utilize during the emergency and decision-making process for protective action for the public. Thus, rapid environmental radiation monitoring is required. In Korea, when an emergency is issued at a nuclear facility, urgent environmental radiation monitoring is conducted based on the national nuclear emergency preparedness and response plan, which can provide important information for decisionmaking on public protective actions. A review of strategies for urgent environmental radiation monitoring is important in performing efficient emergency responses. The main purpose of urgent environmental radiation monitoring is to gather data for decisionmaking on public protective actions to minimize the damage from the accident. For effective data collection and distribution, support from the national and local government and local public organizations and radiation expertise groups, and nuclear facility licensee are required. In addition, an emergency environmental radiation monitoring manual is required to immediately perform environmental monitoring in an emergency situation. The manual for emergency monitoring should include the activities to be conducted according to the phases of the emergency. The phases of the emergency are divided into pre-leakage, post-leakage, intermediate, and recovery. The reasons for establishing strategies are government and public information, the implementation of urgent population protection countermeasures, predicting and tracking plume trajectory, and detection of any release, the protection of emergency and recovery workers, the implementation of agricultural countermeasures and food restrictions, the implementation of intermediate- and recovery-phase countermeasures, contamination control. Besides meteorological data, ambient dose rate and dose, airborne radionuclide concentration, environmental deposition, food, water, and environmental contamination, individual dose, and object surface contamination data are also required for making information for the public.

Cement is widely used as representative industrial material. In Korea, about 50 million tons of cement are consumed every year. In the manufacture of cement, raw materials containing NORM such as fly ash and bauxite are used. Therefore, the workers can be subjected to radiation exposure. The major exposure pathway in NORM industries is internal exposure due to inhalation of aerosol. Internal radiation dose due to aerosol inhalation varies depending on physicochemical properties of the aerosol. Therefore, the objective of this study was to investigate aerosol properties influencing inhalation dose in cement industries. In this study, aerosol properties were measured for two cement manufacturers. A particulate size distribution and concentration at various processing areas in cement manufacturing industries in Korea were analyzed using a cascade impactor. The mass density of raw materials and byproducts were measured using pycnometer. Shape of particulates was analyzed using SEM. The radioactivity concentration of Ra-226, Ra-228 for U/Th decay series was measured using HPGe. Particulate concentration by size was distributed log-normally with maximum at particle size about 7.2 μm in manufacturer A and 5.2 μm in manufacturer B. The mass density of fly ash and cement were 2.3±0.06, 3.2±0.02 g/cm3 respectively in manufacturer A. In manufacturer B, the mass density of bauxite and cement were 3.4±0.02, 2.9±0.01 g/cm3 respectively. The shape of particulates appeared as spherical shape in manufacturer A and B regardless of sampling area. Thus, a shape factor of unity could be assumed. The radioactivity concentrations of Ra-226, Ra-228 were 82±9, 82±8 Bq/kg for fly ash, and 25±4, 23±3 Bq/kg for cement in manufacturer A. In manufacturer B, the radioactivity concentrations of Ra-226, Ra-228 were 344±34, 391±32 Bq/kg for bauxite, and 122±13, 145±12 Bq/kg for cement. The radioactivity concentrations of Ra-226, Ra-228 in cement were less than raw materials such as fly ash and bauxite. It is because the dilution of the radioactivity concentration occurred during mixing with other raw materials in cement production process. This study results will be used as database for accurate dose assessment due to airborne particulate inhalation by workers in cement industries.

In emergency situations such as nuclear accidents or terrorism, radioactive and nuclear materials can be released by some environmental reasons such as the atmosphere and underground water. To secure the safety of human beings and to respond appropriately emergency situation, it is required to designate high and low dose rate regions in the early stages by analyzing the location and radioactivity of sources through environmental radiation measurement. This research team has developed a small gamma probe which is featured by its geometrical accessibility and higher radiation sensitivity than other drone detectors. A plastic scintillator and Silicon Photomultiplier (SiPM) were applied to the probe to optimize the wireless measurement condition. SiPM has a higher gain (higher than 106) and lower operating voltage (less than 30 V) compared to a general photodiode. However, the electronic components in the SiPM are sensitively affected by temperature, which causes the performance degradation of the SiPM. As the SiPM temperature increases, the breakdown voltage (VBD) of the SiPM also increases, so the gain must be maintained by applying the appropriate VBD. Therefore, when the SiPM temperature increases while the VBD is fixed, the gain decreases. Thus, the signal does not exceed the threshold voltage (VTH) and the overall count is reduced. In general, the optimal gain is maintained by cooling the SiPM or through a temperature compensation circuit. However, in the developed system, the hardware correction method such as cooling or temperature compensation circuit cannot be applied. In this study, it was confirmed that the count decreased by up to 20% according to the increase in the temperature of the SiPM when the probe was operated at room temperature (26°C). We propose methods to calibrate the total count without cooling device or compensation circuit. After operating the probe at room temperature, the first measured count is set as the reference value, and the correction factor is derived using the tendency of the count to decrease as the temperature increases. In addition, since this probe is used for environmental radiation monitoring, periodic measurements are more suitable than continuous measurements. Therefore, the temperature of the probe can be maintained by adding a power saving interval to the operation sequence of the probe. These two methods use the operation sequence and measurement data, respectively. Thus, it is expected to be the most effective method for the current system where the temperature compensation through hardware is not possible.

When the nuclear accident like the Fukushima is occurred, it is required to immediately determine the location of radioactive materials and their activities. Various studies related the unmanned technique to detect and characterize the contaminated area have been conducted. The Korea Institute of Nuclear Nonproliferation and Control (KINAC) has developed a new gamma detection system which consists of nine probes using a silicon photomultiplier (SiPM) and plastic scintillator. The probe is the small gamma detector designed to be carried and dropped near the accident area by the unmanned aerial vehicle. In this paper, we developed the improved design related to the angular dependence of the radioactive contamination detection system with the purpose of increasing the detection efficiency. The detection efficiency, radiation shielding and back-scattering varies depending on the direction of incidence of radiation because the probe has vertical structure of consisting scintillator, photomultiplier, and electric circuits. That is, when the experimental conditions are same except the direction of gamma probe, the result of measurements is different. It causes errors in measuring the radioactivity and location of the radioactive source. Since the direction of the probe is arbitrarily determined during the deployment of the probe through the unmanned aerial vehicle, it is considered changing the design of the scintillator from a conventional 1.0" × 1.0" Φ cylindrical shape to a 1.0" Φ spherical shape. In case of using the spherical scintillator, it is confirmed that angular dependence was reduced through MCNP simulation. The difference in the measurement depending on the direction of the probe could be reduced through additional structure design. Finally, we hope that the developed detection system which has the probes with spherical shape of scintillator can measure the radioactivity and location of the radioactive source in a range of about 100 × 100 m2 by measuring for at least 5 minutes. The field test at Fukushima area will be carried out with JAEA members in order to prove the feasibility of the new system.

In liquid scintillation counting, sample radioactivity is analyzed by measuring photons emitted from counting vials. Quenching effect lowers photon intensity from samples, which leads to lower counting efficiency. So an appropriate quenching correction according to characteristics of samples is important. In this study, the quenching correction for H-3 analysis was conducted according to the characteristics of paper packaging material leached samples. The leached samples are made from H-3 leaching method which is in the process of development for H-3 contamination screening. There are several ways of quenching correction such as internal standard (IS) method, quench correction curve and triple-to-double coincidence ratio (TDCR) method, etc. For quench correction curve, quenched standard set, which has the same matrix as experimental samples, is needed to be prepared. Each leached sample, however, has different matrix and color depending on condition of leaching experiment, which means that it is not capable of preparing standard set having same matrix with the samples. In this study, the counting samples are used for plotting quench correction curve instead of quenched standard set. Spectral quench parameter of the external standard [SQP(E)] is used as quench indicating parameter (QIP). TDCR and counting efficiencies determined by IS method are used as counting efficiencies. The quench curve of TDCR versus SQP(E) has R2 = 0.55 and the curve of efficiency from IS method versus SQP(E) has R2 = 0.99. TDCR is known for approximate counting efficiency, however, TDCR as counting efficiency needs careful use for H-3 analysis of leached samples. The curve used efficiency from IS method is suitable for H-3 analysis of leached samples. In this study, the quench correction curve is prepared for H-3 analysis of leached samples of paper packaging material. SQP(E), TDCR and efficiency from IS method was used as parameters to plot the quench correction curve, and, the efficiency from IS method is suitable for H-3 analysis of the leached samples. The result of this study can be used for H-3 analysis of leached samples of paper packaging material.

In Korea, Kori Unit 1, a commercial pressurized water reactor (PWR), was permanently shut down in June 2017, and an immediate decommissioning strategy is underway. Therefore, it is essential to understand the characteristics of radioactive waste during the decommissioning process of nuclear power plants (NPP). Because radioactive waste must be handled with care, radioactive waste is treated in a hot cell facility. Hot cell facility handles radioactive waste, and worker safety is essential. In this study, it was dealt with whether or not the radiation safety regulations were satisfied when processing the core beltline metal of the dismantling waste treated at the post irradiation examination facility (PIEF) of the hot cell facility. Core beltline metal used for the pressure vessel in the reactor is carbon steel, and it is continuously irradiated by neutrons during the operation of the NPP. A radiological safety estimation of the behavior of radioactive aerosols during the cutting process within the PIEF was carried out to ensure the safety of the environment and workers. When processing the core beltline metal in PIEF, dominant six nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) of aerosol are generated. Accordingly each cutting device, amount of aerosol and value of dose is different. Using a 99.97% efficiency HEPA filter, the emission concentration of the dominant nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) in the air source term was satisfied with the emission control standard of Nuclear Safety Commission No. 2016-16. It was confirmed that the radioactivity concentration in the airborne source term inside the PIEF is in equilibrium state, when ventilation is considered. Also, the mass of aerosol and the concentration of airborne source term differed according to the thickness of the saw blade of the cutting tool, and the exposure dose of the worker was different through Monte Carlo N-Particle (MCNP). At that time, 60Co accounted for 95.4% of the exposure dose, showing that 60Co had the highest impact on workers, followed by 55Fe with 2.7%. The worker’s dose limit is satisfied in accordance with Article 2 of the Nuclear Safety Act and the dose limit of radiation-controlled area is found to be satisfied in accordance with Article 3 of the rules on technical standards for radiation safety management at this time.