Background: Sperm quality and the number of sperm introduced into the uterus during artificial insemination (AI) are pivotal factors influencing pregnancy outcomes. However, there have been no reports on the relationship between sperm concentration at AI and sperm quality in Hanwoo cattle. In this study, we examined sperm quality and pregnancy rates after AI using sperm inseminated at different concentrations. Methods: We evaluated the motility, viability, and acrosomal membrane integrity of sperm at different concentrations (10, 15, 18, and 20 million sperm/straw) in 0.5-mL straws. Subsequently, we compared the pregnancy rates after AI with different sperm concentrations. Results: After freeze-thawing, sperm at the assessed concentrations showed similar viability and acrosomal membrane integrity. After AI, cattle in the 10 million group had significantly lower pregnancy rates compared to those in the 18 and 20 million groups. Conversely, there were no statistically significant variances observed between cattle in the 10 and 15 million groups. Conclusions: Sperm at concentrations of 10, 15, 18 and 20 million per straw exhibited comparable motility, viability, and acrosomal membrane integrity. However, a concentration of at least 18 million sperm per straw is required to achieve a consistent rate of pregnancy rate in Hanwoo cattle after AI.

토복령은 우수한 항균, 항산화, 항염증 효능을 가진 소재로 알려져 있다. 이러한 토복령(Smilax china)의 추출물의 기능성을 화장품에 적용하기 위한 기초연구로써 토복령에서 발견되는 플라보노이드인 quercetin, catechin, naringenin의 농도별 경피 투과 특성을 조사할 필요성이 있다. Marzulli의 정의에 적 용한 결과 케르세틴의 Kp 값은 0.1 mg/mL에서 "빠름"으로 분류되었고, 0.2 및 0.4 mg/mL에서 "보통"으로 분류되었다. 특히, 농도가 증가함에 따라 투과 속도가 감소하는 경향이 있었다. 나린제닌의 경우 Flux 값은 각각 0.1, 0.2 및 0.4 mg/mL 농도에서 0.69, 1.07 및 1.42 μg/hr/cm²이었으며, 해당 Kp 값은 각각 6.95, 5.34 및 3.56이었다. 나린제닌의 Kp 값은 모든 농도에서 "보통" 범주에 속하며, 케르세틴과 관찰된 것과 같이 농도가 높아짐에 따라 투과 속도가 감소하였다. 카테킨의 경우 Flux 값은 각각 0.1, 0.2 및 0.4 mg/mL 농도에서 0.75, 1.09 및 1.66 μg/hr/cm²이었으며, 해당 Kp 값은 각각 7.55, 5.46 및 4.16이었다. 카테킨의 Kp 값은 모든 농도에서 일관되게 "보통"으로 분류되었다. 여드름 저해능 및 항염증 효능이 우수 한 토복령 추출물의 유효성분인 quercetin, catechin, naringenin의 경피 투과 특성이 보통 이상으로 나타나 기능성 화장품에 사용할 수 있는 우수한 천연물 소재인 것을 확인할 수 있었다.

The economy of Tanzania relies on agriculture as the main economic activity, because agriculture provides both food and income to the population of the country, especially the rural households. Improvement of market access to crops also increases the productivity. Therefore, in this study, descriptive statistics and statistical analysis of the probit model were used to analyze the factors that determine the market accessibility of small-scale farmers in Chemba, Tanzania. Cross-sectional data collected by the systematic sampling method for 200 corn peasants in Chemba, Tanzania were used. As a result of probit regression analysis, it was found that access to improved seeds and technology had a positive statistically significant effect on the accessibility to the market, while the age of the head of the household, production cost, distance to the market, and household size had a negative statistically significant effect. Therefore, in this study, policy establishment and implementation are recommended. A policy needs to be considered to reduce the transaction costs that eventually allows the farmers to increase the accessibility to markets, enables small-scale farmers to participate in cooperatives, lowers input costs and provides educational programs on quality products to increase their competitiveness in the marketplace.

Background: Porcine pluripotent stem cells (pPSCs) would provide enormous potential for agriculture and biomedicine. However, authentic pPSCs have not established yet because standards for pPSCs-specific markers and culture conditions are not clear. Therefore, the present study reports comparative pluripotency characteristics in porcine induced pluripotent stem cells (piPSCs) derived from different viral transduction and reprogramming factors [Lenti-iPSCs (OSKM), Lenti-iPSCs (OSKMNL) and Sev-iPSCs (OSKM)]. Methods: Porcine fibroblasts were induced into Lenti-iPSCs (OSKM) and Lenti-iPSCs (OSKMNL) by using Lentiviral vector and Sev-iPSCs (OSKM) by using Sendaiviral vector. Expressions of endogenous or exogenous pluripotency-associated genes, surface marker and in vitro differentiation in between Lenti-piPSCs (OSKM), Lenti-iPSCs (OSKMNL) and Sev-piPSCs (OSKM) were compared. Results: Colonial morphology of Lenti-iPSCs (OSKMNL) closely resembles the naïve mouse embryonic stem cells colony for culture, whereas Sev-iPSCs (OSKM) colony is similar to the primed hESCs. Also, the activity of AP shows a distinct different in piPSCs (AP-positive (+) Lenti-iPSCs (OSKMNL) and Sev-iPSCs (OSKM), but AP-negative (-) LentiiPSCs (OSKM)). mRNAs expression of several marker genes (OCT-3/4, NANOG and SOX2) for pluripotency was increased in Lenti-iPSCs (OSKMNL) and Sev-iPSCs (OSKM), but Sev-iPSCs (OSKM). Interestingly, SSEA-1 of surface markers was expressed only in Sev-iPSCs (OSKM), whereas SSEA-4, Tra-1-60 and Tra-1-81 were positively expressed in Lenti-iPSCs (OSKMNL). Exogenous reprogramming factors continuously expressed in Lenti-iPSCs (OSKMNL) for passage 20, whereas Sev-iPSCs (OSKM) did not express any exogenous transcription factors. Finally, only Lenti-iPSCs (OSKMNL) express the three germ layers and primordial germ cells markers in aggregated EBs. Conclusions: These results indicate that the viral transduction system of reprograming factors into porcine differentiated cells display different pluripotency characteristics in piPSCs.

This study was conducted to reset the maximum residue limit (MRL) for didecyldimethylammonium chloride (DDAC) in broiler chickens. The disinfectant containing DDAC (10%, w/w) was diluted 160 times and evenly sprayed on the bodies of twenty-four broiler chickens at a rate of 15 mL per day per bird for 7 days. After the disinfectant treatment, tissue samples were collected from six broiler chickens at 0.25, 1, 3 and 5 days, respectively. Residual DDAC concentrations in poultry tissues were determined using LC-MS/MS. Correlation coefficient (0.99 >), the limits quantification (LOQ, 2.0~10.0 μg/kg), recoveries (86.9~118.6%), and coefficient of variations (<19.98%) were satisfied the validation criteria of Korean Ministry of Food and Drug Safety. In all tissues except for liver, DDAC was detected more than LOQ at 5 days after the disinfectant treatment. In liver tissues, DDAC was detected more than LOQ at 3 days after treatment. According to the European Medicines Agency’s guideline on determination of withdrawal periods, withdrawal period of DDAC in poultry tissues was established to 26 days. In conclusion, the developed analytical method is sensitive and reliable for detecting DDAC in poultry tissues. When DDAC disinfectant is sprayed on a poultry house in the presence of broiler chickens, it is necessary to keep the disinfectant from contacting the body of the livestock.

This study evaluated the fungicidal efficacy of weakly acidic hypochlorous acid water (WAHW) against Microsporum canis (M. canis) and its therapeutic effect on M. canis-infected mouse skin. WAHW was produced by a WAHW generation module. A fungicidal efficacy test by the broth dilution method was used to determine the lowest effective concentration of the WAHW. The lowest effective concentration of WAHW was less than 10 ppm. For T-1, T-2, and T-3, 30 ppm of WAHW was applied to the infected skin once, twice, and three times a day, respectively, and for T-4, 50 ppm of WAHW was applied once a day. On the 3rd day after the initiation of treatment, skin scores in all of the WAHW-treated groups were significantly decreased compared to those in the positive control group (PC) (p<0.05), and there were no significant differences compared to the normal control (NC). The area of the infected skin in all of the WAHW-treated groups was significantly decreased compared to PC from the first day after the initiation of WAHW treatment (p<0.05). The results showed that WAHW had a fungicidal efficacy on M. canis at less than 10 ppm, and it was effective in improving skin symptoms when 30 ppm of WAHW was applied to the M. canis-infected area once a day for five days or 50 ppm of WAHW was applied once a day.

KORAD (Korea Radioactive Waste Agency, http://www.korad.or.kr) has stored slightly contaminated ascon (asphalt coated concrete mixture) that was introduced to Gyeongju repository about a decade ago waiting for a final disposal. It is believed to be mainly contaminated by radioisotope 137Cs due to impurities introduced from the outside during the ascon manufacturing process. We studied characteristics of the radioactive waste to see whether this material would be proper enough to be disposed in Gyeongju LILW repository or be other ways to reduce the disposal volume including self-disposal before its final disposal otherwise. KORAD looked into the properness of characteristics of ascon in terms of WAC (Waste Acceptance Criteria) documented by KORAD that includes general chemical and physical properties of asphalt, density, size of grains, content of organic material and possibility of existence of chelate materials that qualitatively limited to be disposed by the criteria. And other associated characteristics such as gas generation and bio degradation were also investigated. Based on the data obtained from the study, we proposed various plausible solutions in associated with operational and disposal safety and economic view points. This study will be used for KORAD’s decision on how to control and safely dispose the spent ascon within a reasonable time period. And also those experiences may be applied for other LILW issues that require treatment or conditioning of radioactive wastes in the future.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

The bentonite buffer material is a crucial component for disposing of high-level radioactive waste (HLW). Several additives have been proposed to enhance the performance of bentonite buffer materials. In this study, unconfined compression tests were conducted on bentonite mixtures as well as pure bentonite buffer material. Joomunjin and silica sands were added at a 30% ratio, and graphite was added at 3% along with bentonite. The unconfined compression strength (UCS) and elastic modulus of pure bentonite were found to be 20% to 50% higher than those of bentonite mixtures under similar dry density and water content conditions. This decrease in strength can be attributed to the reduced cross-sectional area available for bearing the applied load in the bentonitemixture. Furthermore, the 3% graphite-bentonite mixture exhibited a 10% to 30% higher UCS and elastic modulus compared to the 30% sand-bentonite mixtures. However, since the strength properties of additive-bentonite mixtures are lower than those of pure bentonite, it is essential to evaluate thermohydraulic-mechanical functional criteria when considering the use of bentonite mixtures as buffer materials.

In 2012, POSIVA selected a bentonite-based (montmorillonite) block/pellet as the backfilling solution for the deposition tunnel in the application for a construction license for the deep geological repository of high-level radioactive waste in Finland. However, in the license application (i.e. SC-OLA) for the operation submitted to the Finnish Government in 2021, the design for backfilling was changed to a granular mixture consisting of bentonite (smectite) pellets crushed to various sizes, based on NAGRA’s buffer solution. In this study, as part of the preliminary design of the deep geological repository system in Korea, we reviewed history and its rationale for the design change of Finland’s deposition tunnel backfilling solution. After the construction license was granted by the Finnish Government in 2015, POSIVA conducted various lab- and full-scale in-situ tests to evaluate the producibility and performance of two design alternatives (i.e. block/pellet type and granular type) for backfilling. Principal demonstration tests and their results are summarized as follows: (a) Manufacturing of blocks using three types of materials (Friedland, IBeco RWC, and MX-80): Cracking and jointing under higher pressing loads were found. Despite adjusting the pressing process, similar phenomena were observed. (b) 1:6 scale experiment: Confirmation of density difference inhomogeneity due to the swelling of block/pellet backfill and void filling due to swelling behavior into the mass loss area of block/pellet. (c) FISST (Full-Scale In situ system Test): Identification of technical unfeasibility due to the inefficient (too manual) installation process of blocks/pellets and development of an efficient granular in-situ backfilling solution to resolve the disadvantage. (d) LUCOEX-FE (Large Underground Concept Experiments – Full-scale Emplacement) experiment: Confirmation of dense/homogeneous constructability and performance of granular backfilling solution. In conclusion, the simplified granular backfill system is more feasible compared to the block/ pellet system from the perspective of handling, production, installation, performance, and quality control. It is presumed that various experimental and engineering researches should be preceded reflecting specific disposal conditions even though these results are expected to be applied as key data and/or insights for selecting the backfilling solution in the domestic deep geological repository.

The permanent disposal of discharged spent nuclear fuel (SNF) and contaminated radioactive waste generated from the subsequent chemical treatments of SNF has become a serious pending issue in many countries that operate the nuclear power plants. Among the diverse engineering solutions proposed for the disposal of high-level radioactive waste (HLW), deep geological disposal (DGD) has been considered as the most proven and safe option to prevent any significant release of radionuclides into the biosphere and to predictably ensure the long-term performance of disposal system. The DGD system consists of multiple structural components; the bentonite clay-based buffer and tunnel backfills are designed to perform the primary hydrogeochemical functions of 1) inhibiting the ingress of groundwater and reactive substances that could compromise the integrity of canister and 2) retarding the migration of released radionuclides into biosphere by providing the sufficient chemisorption sites. Montmorillonite, which is a 2:1 phyllosilicate mineral belonging to smectite group, constitutes the majority of bentonite, and it mainly predominate the swelling and chemisorption capacities of the clay material. Thus, it is essentially required to thoroughly understand the chemical interactions of major radionuclides and other important substances with montmorillonite in advance to accurately evaluate the long-term retention performance of engineered barriers and to reduce the uncertainties in the safety assessment of a deep geological repository (DGR) ultimately. Thus far, sorption of dissolved species onto mineral adsorbents has been generally described and quantified using the simple sorption-desorption distribution coefficient (Kd) concept; since any specific reaction mechanisms are not considered and reflected in the Kd concept, an empirical Kd value is intrinsically dependent on the aqueous conditions under which it was measured. In this framework, substantial scientific efforts have been made to develop a robust basis for geochemically parametrizing the sorption phenomena more reliably, and the application of thermodynamic sorption modeling (TSM), which is based on the chemical principle of mass action laws, has been studied with the aim of improving overall confidence in the description of radionuclide migration under a wide range of aquatic conditions. The disposal performance demonstration R&D division of KAERI introduced a new reference Ca-bentonite clay called Bentonil-WRK (Clariant Korea) for HLW disposal research in 2021 as the domestic Ca-bentonite sources have being depleted. We successfully separated and purified Ca-montmorillonite from the Bentonil-WRK clay, and its geochemical characteristics were meticulously studied by means of XRD, BET, CEC, FT-IR analyses and controlled acid-base titration. In this work, chemical sorption behaviors of aqueous iodide and benzoate, which are a major fission product in HLW and a model ligand of complex natural organic matters present in the deep geological environment, onto the purified Camontmorillonite were assessed under ambient conditions of S/L = 5 g/L, I = 0.01 M CaCl2, pH = 4- 9, pCO2 = 10-3.4 atm, and T = 25°C. Further, their unique adsorption envelopes and corresponding thermodynamic reaction constants refined from the diffuse double layer model (DDLM)-based inverse modeling of experimental sorption data were discussed.

The WRK (Waste Repository Korea bentonite) compacted bentonite medium has been considered as the appropriate buffer material in the Korean SNF (Spent nuclear fuel) repository site. In this study, hydraulic properties of the WRK compacted bentonite core (4.5 cm in diameter and 1.0 cm in length) as the buffer material were investigated in laboratory experiments. The porosity and the entry pressure of the water saturated core at different confining pressure conditions were measured. The average velocity of water flow in the WRK compacted bentonite core was calculated from results of the breakthrough curves of the CsI aqueous solution and the hydraulic conductivity of the core was also calculated from the continuous flow core experiments. Because various gases could be generated by continuous SNF fission, container corrosion and biochemical reactions in the repository site, the gas migration property in the WRK compacted bentonite core was also investigated in experiments. The gas permeability and the average of gas (H2) in the core at different water saturation conditions were measured. Laboratory experiments with the WRK Compacted bentonite core were performed under conditions simulating the DGR environment (confining pressure: 1.5- 20.0 MPa, injection pressure: 1.0-5.0 MPa, water saturation: 0-100%). The WRK Compacted bentonite core was saturated at various confining pressure conditions and the porosity ranged from 27.5% to 43.75% (average: 36.75%). The calculated hydraulic conductivity (K) of the core using experimental results was 8.69×10-11 cm/s. The gas permeability of the core when the water saturation 0~58 % was ranged of 19.81~3.43×10-16 m2, representing that the gas migration in the buffer depends directly on the water saturation degree of the buffer medium. The average gas velocity in the core at 58% of water saturation was 9.8×10-6 m/s, suggesting that the gas could migrate fast through the buffer medium in the SNF repository site. Identification of the hydraulic property for the buffer medium, acquired through these experimental measurements is very rare and is considered to have high academic values. Experimental results from this study were used as input parameter values for the numerical modeling to simulate the long-term gas migration in the buffer zone and to evaluate the feasibility of the buffer material, controlling the radionuclide-gas migration in the SNF repository site.

In general, systems are developed by repeatedly performing the processes of design, analysis, manufacturing, and performance testing. In particular, systems with temperature, pressure, and flow rate often utilize computational fluid dynamics tools at the design stage. In this paper, we aim to verify the reliability of the analysis results of Solidworks Flow Simulation, which is widely used in heat flow analysis at the design stage. A tube furnace was manufactured, various experiments were performed, and a study was conducted to compare the analysis results. The details of the experiment are as follows. First, an experiment was conducted in which the heater was heated to 900°C without insulating the exposed part of the tube. The detailed contents of the experiment are as follows; - Heating heater and measuring temperature without supplying flow inside the tube, - Tube flow supply (25°C, 15 lpm air) and heater heating/temperature measurement. Second, an experiment was performed in which the exposed part of the tube was insulated (thickness 50 mm) and the heater was heated to 900°C. The detailed contents of the experiment are as follows; - Insulate the outside of the tube except for the flanges at both ends of the tube, and heat the heater and measure the temperature without supplying flow inside the tube. - Insulate the outside of the tube except for the flanges at both ends of the tube, supply flow rate inside the tube (25°C, 15 lpm air) and measure heater heating/temperature. - Insulate the flange of the flow supply section, heat the heater and measure temperature without supplying flow inside the tube. - Insulate the flange of the flow supply section, heat the supply air (277°C, 15 lpm) and measure the temperature using a heating gun without heating the heater. - Insulate the flange of the flow supply section, supply heated air (277°C, 15 lpm) and measure heater heating/temperature. - Insulate the flange of the flow supply section and measure temperature according to heater heating (900°C) and supply temperature (25°C, 277°C 15 lpm). The following results were derived from the experimental and analysis results. - When the exposed part of the tube is insulated, the temperature inside the tube increases and the steady-state power decreases compared to non-insulated. - In areas with insulation, the temperature error between experiment and analysis results is not large. - When flow rate is supplied, there is a large temperature error in experiment and analysis results. - The temperature change after the center of the heater is not large for a temperature change of 15 lpm flow rate. From these results, it can be seen that Solidworks Flow Simulation has a significant difference from the experimental results when there is a flow rate in the tube. This was thought to be because the flow rate acts as a disturbance, and this cannot be sufficiently accounted for in the analysis. In the future, we plan to check whether there is a way to solve this problem.

After the decision of the Wolsong unit 1 permanent shutdown (2019), spent fuel stored in the spent fuel bay (hereafter, SFB) should be transported to a dry storage facility (MACSTOR or Canister) in order to decommission Wolsong unit 1. Accordingly, KHNP has established a shipment schedule for damaged fuel of Wolsong Unit 1 and is trying to complete the shipment according to the schedule. Wolsong is equipped with transportation casks and dry storage facilities, but baskets need to be manufactured separately. In addition, license approval is required for baskets, transport cask, and dry storage facilities for legal grounds to contain, transport, and store damaged fuels. In this paper, the initial model, upgrade model, and automation model of encapsulation equipment planned to be introduced in Canada to handle PHWR’s damaged fuel were compared, and the optimal model was selected in consideration of KHNP’s planned spent fuel shipment schedule. The PHWR’s damaged fuel encapsulation system is a system developed the PHWR’s damaged spent fuel to be handled in the same way as the existing PHWR when storing it in the dry storage facility and loading a basket for capsulation into transport cask. At the Gentilly-2 nuclear power plant in Canada, a manually operated encapsulation system was used due to the low quantity of damaged fuel, which can be encapsulated two bundles a day, and this model is an initial model. In the case of Wolsong Unit 1, it has about 300 damaged fuels, so it takes about nine months to work when using the initial model. The upgrade model developed to improve work efficiency and reliability has increased work efficiency through some automation, but it would take about eight months to process the damaged spent fuels of Wolsong Unit 1, and this model has not yet been manufactured and applied. Lastly, the automation model changed the work location outside the SFB and automated drainage/drying operations. It is easy to maintain and replace consumables because the work is carried out by lifting the damaged fuel to a shuttle outside the SFB surrounded by a shielding chimney. Considering the reduction of drainage/drying time, it is possible to save about four times as much time as the initial model. That is, if the automation model is used, it is judged that the supply of Wolsong Unit 1 can be processed in about two months. However, in terms of license, initial model and upgrade model are expected to be easier and the period is expected to be shortened. However, if licensing is carried out as soon as equipment design is completed, it is believed that the period can be shortened by parallel equipment manufacturing and licensing. It is judged that the best way to comply with the target schedule is to select an automation model with excellent work performance, develop equipment, and proceed with licensing at the same time. Accordingly, KHNP is in the process of designing equipment with the aim of using the automation model to take out damaged fuel for Wolsong Unit 1.

It has been investigated on the management of Strontium-90 in KAERI. It is needed to separate the solute from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt in our previous study. Strontium chloride was successfully carbonated by reactive distillation of SrCl2 – K2CO3 – LiCl – KCl system. In this study, it was tried to develop another route to recover strontium from the salt solution by a solid-solid reaction for avoiding the entrainment of product and the salt-K2CO3 reaction. Reactive distillation experiments were carried out for SrCl2 - K2CO3 – LiCl – KCl system. The carbonation temperature and pressure were 520°C and 0.8 bar. After the carbonation reaction, the temperature was elevated to 820°C to remove KCl from the reaction product. SrCO3 and KCl peaks were found in the XRD analysis of the residual product. It could be concluded that SrCl2 can be successfully carbonated after salt removal by the solid-solid reaction.

The Republic of Korea (ROK), as a member state of the IAEA, is operating the State’s System of Accounting for and Control (SSAC) and conducting independent national inspections. Furthermore, an evaluation methodology for the material unaccounted for (MUF) is being developed in ROK to enhance capabilities of national inspection. Generally, physical and chemical changes of nuclear material are unavoidable due to the operating system and structure of facilities, an accumulation of material unaccounted for (MUF) has been issued. IAEA developed statistical MUF evaluation method that can be applied to all facilities around the world and it mainly focuses on the diversion detection of nuclear materials in facilities. However, in terms of the national safeguard inspection, an evaluation of accountancy in facilities is additionally needed. Therefore, in this research, a new approach to MUF evaluation is suggested, based on the Guide to the Expression of Uncertainty in Measurement (GUM) that an evaluation of measurement uncertainty factors is straightforward. A hypothetical list of inventory items (LII) which has 6,118 items at the beginning and end of the material balance period, along with 360 inflow and outflow nuclear material items at a virtual fuel fabrication plant was employed for both the conventional IAEA MUF evaluation method and the proposed GUM-based method. To calculate the measurement uncertainty, it was assumed that an electronic balance, gravimetry, and a thermal ionization mass spectrometer were used for a measurement of the mass, concentration, and enrichment of 235U, respectively. Additionally, it was considered that independent and correlated uncertainty factors were defined as random factors and systematic factors for the ease of uncertainty propagation by the GUM. The total MUF uncertainties of IAEA (σMUF) and GUM (uMUF) method were 37.951 and 36.692 kg, respectively, under the aforementioned assumptions. The difference is low, it was demonstrated that the GUM method is applicable to the MUF evaluation. The IAEA method demonstrated its applicability to all nuclear facilities, but its calculated errors exhibited low traceability due to its simplification. In contrast, the calculated uncertainty based on the GUM method exhibited high reliability and traceability, as it allows for individual management of measurement uncertainty based on the facility’s accounting information. Consequently, the application of the GUM approach could offer more benefits than the conventional IAEA method in cases of national safeguard inspections where factor analysis is required for MUF assessment.

Characteristics and useful effects for human health of antioxidant carbon nanodots contained in microwave-assisted Opuntia humifusa extract (MA-OHE(CD)) were investigated in this work. MA-OHE(CD) was characterized using transmission electron microscopy, dynamic light scattering analysis, X-ray diffraction, UV–vis spectroscopy, X-ray photoelectron spectroscopy, and Fourier transform-infrared spectroscopy. Besides, total phenolic content and antioxidant activity of MA-OHE(CD) were measured. It was revealed that the MA-OHE(CD) increases aquaporin-3 expression in human epidermal keratinocytes cell with hydrophilic characteristics. Moreover, the bio-active compound extraction efficacy and antioxidant activity of microwave-assisted extraction were great when compared to maceration.